LIC-05-0127, Revised License Amendment Request to Support Use of Ms Fuel Cladding

Revised License Amendment Request to Support Use of Ms Fuel Cladding

ML053120421
Person / Time
Site:	Fort Calhoun
Issue date:	11/08/2005
From:	Ridenoure R Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-05-0127
Download: ML053120421 (22)

Text

Omaha Public power 0 Z t 444 South 16th Street Mall Omaha NE 68102-2247 November 8,2005 LIC-05-0127 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Reference:

1. Docket No. 50-285

2. Letter from Ross T. Ridenoure (OPPD) to Document Control Desk (NRC) dated August 11, 2005, Fort Calhoun Station Unit No. 1 - License Amendment Request to Support Use of M5 Fuel Cladding, and 10 CFR 50.46 and 10 CFR Appendix K Exemption Request (LIC-05-0089)

(ML052240083)

SUBJECT:

Fort Calhoun Station Unit No. 1 - Revised License Amendment Request to Support Use of M S

• Fuel

~ Cladding Pursuant to 10 CFR 50.90, Omaha Public Power District (OPPD) hereby requests the following amendment to the Fort Calhoun Station Unit 1 Technical Specifications (TS). The proposed amendment will modify TS 4.2.1, "Fuel Assemblies," to permit the use of AREVA (Framatome ANP) ~5~~ advanced alloy for fuel rod cladding and structural components such as guide tubes, intermediate spacer grids, end plugs and guide thimble tubes, beginning with Cycle 24. In addition, OPPD proposes to modify TS 5.9 to include the Framatome ANP Topical Report evaluating the impact of ~5~~ material properties on NRC approved methodology. ~5~~ is a proprietary, zirconium based alloy that is a variant of ZrlNb to replace zircaloy-4 in the construction of fuel assembly components. OPPD concludes that the proposed amendment presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c).

Reference 2 contained OPPD's original submittal with respect to use of ~5~~ clad fuel. This amendment request is being resubmitted to clarify the wording of Reference 2 with respect to referenced topical reports. This amendment request is also being resubmitted to provide the appropriate marked-up and clean typed Technical Specification pages.

Reference 2 also requests an exemption pursuant to 10 CFR 50.12 from 10 CFR 50.46, Acceptance Criteria for emergency core cooling systems for light-water nuclear ower reactors 4

and 10 CFR 50, Appendix K to Part 50 -- ECCS Evaluation Models. Since M5 cladding is a zirconium-based alloy that is chemically different than zircaloy or ZIRLO fuel cladding materials which are approved for use in these 10 CFR sections, a plant specific exemption from these regulations is required to support the use of ~5~~ cladding. Information supporting the exemption request is contained in Attachment 4 of Reference 2. OPPD has concluded that special circumstances defined by 10 CFR 50.12 exist to warrant the exemption and that granting the exemption request will not present undue risk to the public health and safety and is consistent with the common defense and security. No changes are proposed to the exemption request of Reference 2 in this submittal.

Employment with Equal Opportunity

U.S. Nuclear Regulatory Commission LIC-05-0127 Page 2 Attachment 1 provides the No Significant Hazards Evaluation and the technical bases for this requested change to the TS. Attachments 2 and 3 contain the revised marked-up (changes shown in italics) and clean-typed TS pages reflecting the requested TS changes.

The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(l) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in Attachment 1.

The NRC has approved similar TS changes for other plants. In particular, fuel with ~5~~

cladding is used at Oconee Units 1, 2, and 3, Three Mile Island Unit 1, Davis Besse, and Crystal River Unit 3, which are Babcock and Wilcox plants, and at North Anna Units 1 and 2 which are Westinghouse plants.

OPPD requests approval of the proposed amendment in this submittal and the exemption of Reference 2 by July 31, 2006 to support he1 procurement and core design for the Fall 2006 refueling outage. OPPD requests 120 days to implement this amendment. No commitments are made to the NRC in this letter.

I declare under penalty of perjury that the foregoing is true and correct. (Executed on November 8,2005)

If you have any questions or require additional information, please contact Mr. Thomas R. Byrne at (402) 533-7368.

Sincerely, -

,)/ice ~ r e s f k n t

~ttachmdnts:

1. OPPD's Evaluation of the proposed change@)

2. Markup of Technical Specification Pages

3. Clean Typed Technical Specification Pages c: Division Administrator - Public Health Assurance, State of Nebraska

LIC-05-0127 Page 1 ATTACHMENT 1 Fort Calhoun Station Unit No. 1 - Revised License Amendment Request to Support Use of ~5~~ Fuel Cladding , and 10 CFR 50.46 and 10 CFR Appendix K Exemption Request

LIC-05-0127 Page 2 Attachment 1 Fort Calhoun Station Unit No. 1 - Revised License Amendment Request to Support Use of ~5~~ Fuel Cladding, and 10 CFR 50.46 and 10 CFR Appendix K Exemption Request DESCRIPTION PROPOSED CHANGE BACKGROUND TECHNICAL ANALYSIS REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria ENVIRONMENTAL CONSIDERATION PRECEDENCE REFERENCES

LIC-05-0127 Page 3 Fort Calhoun Station Unit No. 1 - Revised License Amendment Request to Support Use of ~5~~ Fuel Cladding, and 10 CFR 50.46 and 10 CFR Appendix K Exemption Request

1.0 DESCRIPTION

This letter is a request to amend Operating License DPR-40 for Fort Calhoun Station Unit No. 1 (FCS). The proposed changes to Technical Specifications (TS) Design Features, TS 4.2.1, "Fuel Assemblies," and to TS 5.9 "Reporting Requirements" would permit the use of the ~5~~ advanced alloy. FCS is planning to use an enhanced AREVA fuel design, which uses ~5~~ material for fuel cladding and other assembly structural components, for replacement fuel assemblies in future core reload designs starting with Cycle 24.

2.0 PROPOSED CHANGE

The proposed change will add the allowance to use the ~5~~ advanced alloy fuel to FCS TS Section 4 Design Features, Section 4.2.1, thereby permitting the use of ~5~ cladding for replacement fuel assemblies in future core reloads. The ~5~~ fuel cladding is chemically different than zircaloy, which is currently specified in TS 4.2.1. A modification of TS 5.9 "Reporting Requirements" to include the Framatome Topical report BAW-10240(P)(A), Revision 0, "Incorporation of ~5~~ Properties in Framatome ANP Approved Methods," that evaluate .the ~5~~ cladding and structural components is also proposed. The approved version of this topical report will be specified in the FCS Core Operating Limits Report (COLR) per the allowance of TSTF-363. TSTF-363 allows licensees to use current topical reports to support limits in the COLR without having to submit an amendment request every time the topical report is revised.

An exemption to 10 CFR 50.46 and 10 CFR Appendix K has also been proposed in accordance with 10 CFR 50.12 to support the use of ~5~~ cladding. This was included as Attachment 4 to Reference 8.10.

In summary, OPPD proposes to amend the FCS TS to permit the use of the ~5~

advanced alloy as fuel rod cladding and fuel assembly structural components.

3.0 BACKGROUND

Currently FCS fuel cladding is zircaloy-4, which is allowed by TS 4.2.1. The fuel rod cladding is designed to maintain its integrity for the anticipated operating transients throughout core life. The effects of gas release, fuel dimensional changes, and corrosion-induced or irradiation-induced changes in the mechanical properties of cladding are

LIC-05-0127 Page 4 considered in the design of fuel assemblies. The zircaloy-4 cladding is designed to withstand strain resulting from combined effects of reactor pressure, fission gas pressure, fuel expansion, and thermal and irradiation growth. Materials testing and actual reactor in-service operation with zircaloy cladding have demonstrated that zircaloy-4 material has sufficient corrosion resistance and mechanical properties to maintain the integrity and serviceability required for the design burnup.

In order to provide an improvement in performance and improved margins during normal operation, AREVA has developed the ~5~~ advanced fuel rod cladding and fuel assembly structural material. ~5~~ is an alloy comprised primarily of zirconium (98.9%), niobium (1%) and oxygen (0.1%). The absence of tin in ~5~~ has resulted in superior corrosion resistance and reduced irradiation-induced growth relative to standard zircaloy (1.7% tin) and low tin zircaloy (1.2% tin). The addition of niobium increases ductility, which is desirable to avoid brittle failures.

~5~~ has completed several cycles of irradiation in US and European reactors. Results from the irradiation of the ~5~~ fuel rod cladding has demonstrated that the maximum fuel rod corrosion rate is 40 to 50% that of low-tin zircaloy-4. In addition, the hydrogen pickup is a quarter of that experienced with zircaloy-4. Similar improvements have been shown for the fuel assembly structural components, such as guide tubes and spacer grids.

The fuel rod growth measurements have shown a reduced irradiation-induced growth of approximately 80% relative to standard zircaloy-4. The ~5~~ cladding will provide additional margin to the fuel assembly and fuel rod growth limits for fuel assemblies with high burnups. Since fuel rod bow is driven by the irradiation growth of the fuel rods, the reduced fuel assembly growth will help reduce irradiation-induced fuel rod bow and distortion, which can be detrimental to fuel handling activities. Since the creep rate of

~5~~ is considerably slower than that of standard zircaloy-4 cladding, the creep collapse life of ~5~~ fuel rods is much greater than the standard rods and is not limiting at burnups up to 62 GWDMTU. This decrease in creep collapse rate can benefit the fuel rod internal pressure performance.

4.0 TECHNICAL ANALYSIS

Topical Report BAW-10227P-A, Revision 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel (Reference 8.1) approved by the NRC on June 18, 2003 provides the technical licensing basis for the use of ~5~~ fuel cladding material and structural material. The ~5~~ cladding is an AREVA proprietary material com rised of approximately 99% zirconium, and 1% niobium. As mentioned in Section 3, M5T E cladding provides improved performance in fuel cladding corrosion and hydrogen pickup, fuel assembly and fuel rod growth, fuel rod bowing, and fuel rod cladding creep over standard zircaloy-4 cladding. The ~5~~ fuel cladding alloy has been tested in both

LIC-05-0127 Page 5 reactor and non-reactor environments to establish its superior mechanical and structural properties.

AREVA has evaluated the properties of ~5~~ and determined that the use of ~5~~ as cladding and structural material would have either no significant impact or would produce an improvement in performance and increased margins for the following parameters and analyses:

Fuel assembly and rod growth Fuel assembly handling and shipping loads Fuel rod internal pressure Fuel rod cladding transient strain Fuel centerline melting temperature Fuel rod cladding fatigue Fuel rod cladding creep collapse Fuel rod bow High temperature swelling and rupture High temperature oxidation AREVA has determined that the ~5~~ advanced alloy will perform acceptably at all normal operating conditions.

AREVA has performed an evaluation of the LOCA and non-LOCA performance of the

~5~~ cladding alloy for the generic accident scenarios described in Reference 8.1. The LOCA evaluation is performed with a set of analyses to show compliance with 10 CFR 50.46. A comparison of results obtained using the base evaluation model methods with zircaloy-4 cladding and the results obtained for an identical case using

~5~~ swelling and rupture model shows that the ~5~~ cladding performance should not adversely affect core operation or operating limits.

AREVA has performed a plant-specific realistic Large Break LOCA (RLBLOCA) for FCS using approved RLBLOCA methodology (Reference 8.3). OPPD has submitted a separate, but related license amendment request based on the AREVA RLBLOCA analysis as Reference 8.15.

AREVA will perform an assessment of the impact of the ~5~~ alloy on the safety performance of nuclear fuel. The results of these calculations are not expected to differ substantially from zircaloy-4 based calculations and no limiting criteria are expected to be challenged.

LIC-05-0127 Page 6 AREVA has determined that .the use of the ~5~~ alloy will have no significant adverse impact on radiological doses, which may result from any accident involving the radionuclides in the gap or fuel pellet.

An overview of IWC approved OPPD methodology for FCS reload core analysis is included in OPPD-NA-8301, Revision 8, Omaha Public Power District Reload Core Analysis Methodology Overview (Reference 8.1 1). Neutronics design methods implemented for FCS core reload analysis are described in the NRC approved document, OPPD-NA-8302, Revision 6, Omaha Public Power District Reload Core Analysis Methodology, Neutronics Design Methods and Verification (Reference 8.12). The use of

~5~~ cladding or structural material does not affect the neutronics characteristics of the fuel material. The basic thermo-physical properties and neutronics characteristics of

~5~~ advanced alloy are quite similar to those of zircaloy-4. This ensures that the NRC approved CASMO-4lSIMULATE-3 methodology (Reference 8.14) that is described in OPPD-NA-8302 is appropriate for the FCS core reload process such as cross section generation, core simulation and depletion analysis, and neutronics parameters generation for use in transient and accident analyses.

The FCS core thermal hydraulics, transient and accident analysis methods and computer codes for core reload analysis are described in the NRC approved document, OPPD-NA-8303, Revision 6, Omaha Public Power District Reload Core Analysis Methodology Transient and Accident Methods and Verification (Reference 8.13). Use of these documents was approved by the NRC in Reference 8.14. The non-DNBR limiting events such as loss of load and feedwater malfunctions were analyzed using the OPPD-NA-8303 listed Combustion Engineering (CE) methodology. The acceptance criteria for these events are that the reactor coolant and main steam pressures should be less than 110% of design values (i.e., challenges to the pressure boundary) and that the Departure from Nucleate Boiling (DNB) and Linear Heat Generation Rate (LHGR) Specified Acceptable Fuel Safety Limit (SAFDL) be met. The DNB and LHGR SAFDL criteria which affect fuel cladding integrity are not of a major concern since DNB Ratio increases during these events and the Peak LHGR margin required is much less limiting than for other Abnormal Operational Occurrences (AOOs). Thus, since the acceptance criteria for these two events are associated with the integrity of the RCS pressure boundary and secondary system pressure boundary, the use of ~5~~ cladding will have no impact on the two analyses of record (AOR) using CE methodology.

The non-DNBR limiting events, which were analyzed using the OPPD-NA-8303 documented CE methodology will be evaluated during the reload process to determine whether they can either be dispositioned or a reanalysis is necessary. If it is determined that a reanalysis is required, those transients and accidents shall be reanalyzed using the EMF-23 1O(P)(A) methodology.

The AREVA topical report, BAW-10240(P)(A) (Reference 8.7) justifies the use of ~5~~

with all of the Framatome ANP topical reports referenced in section 5.9 of the FCS

LIC-05-0 127 Page 7 Technical Specifications and section 2.0 of the COLR except for the topical report ANF-89-15 1(P)(A). The topical report ANF 151. (P)(A) contains a non-LOCA transient analysis methodology similar to EMF-2310(P)(A). The change in cladding material from zircaloy-4 to ~5~~ potentially affects the fuel rod modeling in ANF-89-15 1(P)(A).

ANF-89-151(P)(A) addresses the fuel rod modeling that is applied in the ANF-RELAP code; specifically, confirmatory RODEX2 calculations are performed for a given fuel design to verify that the pellet-to-clad gap conductance is within the range of the sensitivity study performed in support of the ANF-89-15 1(P)(A) methodology.

RODEX2 calculations for a fuel design similar to that for FCS have verified that gap conductance for fuel rods with ~5~ cladding is within the range supported by the fuel rod sensitivity study performed for the ANF 151(P)(A) methodology. Further, the gap conductance used in the ANF-RELAP analyses is within the range supported by the fuel rod sensitivity study. Thus the ANF 151(P)(A) approved methodology does not require revision to support ~5~ cladding.

Those transient and accident analyses performed using ANF 151(P)(A) methodology will be examined during every cycle reload process to determine whether those analyses can either be dispositioned or a reanalysis is necessary. If it is determined that a reanalysis is required, those transients and accidents shall be reanalyzed using the EMF-23 1O(P)(A) methodology.

The report BAW-10240(P)(A) demonstrates that neutronic, non-LOCA and DNE3 related topical reports do not require revision to address the use of ~5~~ cladding. There is a negligible impact of the ~5~~ cladding on non-LOCA transients and no impact on the DNE3 correlations or neutronic methods. It is thus concluded that it is acceptable to reference these topical reports in the FCS TS without modification for ~5~~ cladding.

This proposed amendment does not involve application or use of risk-informed decisions.

5.0 REGULATORY SAFETY ANALYSIS The technical analysis performed to justify the use of the fuel assemblies containing

~5~~ material will be performed with methods contained in NRC approved topical reports. Since the ~5~~ material has either a small or beneficial impact of the safety analyses it is expected that no significant impact on the safety analyses will be observed.

5.1 NO SIGNIFICANT HAZARDS CONSIDERATION OPPD has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

LIC-05-0127 Page 8

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

Response: No.

The NRC approved topical report BAW-1027P-A (Reference 8.1) that provides the licensing basis for ~5~~ cladding and structural material, has shown that the

~5~~ alloy exhibits superior properties to the currently used zircaloy-4 material.

The cladding by itself does not initiate an accident and therefore does not affect accident probability. It has been determined that ~5~~ cladding will not significantly affect the consequences of an accident.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously analyzed.

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

Response: No.

The proposed change does not result in changes in the operation or overall configuration of the facility. Topical report BAW-10227P-A (Reference 8.1) demonstrated that the ~5~~ alloy will perform similar to or better than zircaloy-4, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident.

Since the material properties of ~5~~ alloy are similar to or better than zircaloy-4, there will not be any significant change in the types of effluents that may be released off-site. There will not be any significant increase in occupational or public radiation exposure.

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

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

Response: No.

AREVA has performed generic LOCA and non-LOCA evaluations and demonstrated the use of the ~5~~ material will have only a small, or beneficial, impact on the event consequences.

LIC-05-0127 Page 9 Plant-specific analyses using NRC approved methodology for the mixed core will demonstrate that the reactor core safety limits will continue to be met.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, OPPD concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 APPLICABLE REGULATORY REQUIREMENTSICRITERIA The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

5.2.1 Regulations The proposed amendment to allow the use of M5 fuel rod cladding must comply with Criterion 10 of 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants. OPPD has determined that the proposed change that allows the use of ~5~~ fuel rod cladding material requires exemptions from 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors and 10 CFR 50, Appendix K, ECCS Evaluation Models.

Attachment 4 provides the basis and justification for exemption from these regulations.

5.2.2 Design Basis The proposed change to use the ~5~~ he1 rod cladding will not affect the design bases of the plant and is therefore acceptable. The AOOs and postulated accidents listed in Chapter 14 of the FCS USAR are either analyzed or dispositioned for each cycle of operation. All incidents listed in Chapter 14 of the USAR are analyzed using NRC approved methodologies to show that no SAFDLs are exceeded. To assure that adequate protection is provided for the public, conservative assumptions are incorporated into the analyses.

5.2.3 Approved Methodologies An overview of NRC approved OPPD methodology for FCS reload core analysis is included in OPPD-NA-8301, Revision 8, Omaha Public Power District Reload Core Analysis Methodology Overview (Reference 8.1 1). Neutronics design methods implemented for FCS core reload analysis are described in the NRC approved document, OPPD-NA-8302, Revision 6, Omaha Public Power District Reload Core Analysis Methodology, Neutronics Design Methods and Verification

LIC-05-0127 Page 10 (Reference 8.12). The use of ~5~~ cladding or structural material does not affect the neutronics characteristics of the fuel material. The basic thermo-physical properties and neutronics characteristics of ~5~~ advanced alloy are quite similar to those of zircaloy-4. This ensures that the NRC approved CASMO-4lSIMLTLATE-3 methodology (Reference 8.14) that is described in OPPD-IVA-8302 is appropriate for the FCS core reload process such as cross section generation, core simulation and depletion analysis, and neutronics parameters generation for use in transient and accident analyses.

The FCS core thermal hydraulics, transient and accident analysis methods and computer codes for core reload analysis are described in the NRC approved document, OPPD-NA-8303, Revision 6 , Omaha Public Power District Reload Core Analysis Methodology Transient and Accident Methods and Verification (Reference 8.13). Use of these documents was approved by the NRC in Reference 8.14. The non-DNBR limiting events such as loss of load and feedwater malfunctions were analyzed using the OPPD-NA-8303 listed CE methodology. The acceptance criteria for these events are that the reactor coolant and main steam pressures should be less than 110% of design values (i.e.,

challenges to the pressure boundary) and that the DNB and LHGR SAFDL be met. The DNB and LHGR SAFDL criteria which affect fuel cladding integrity are not of a major concern since DNB Ratio increases during these events and the Peak LHGR margin required is much less limiting than for other AOOs. Thus, since the acceptance criteria for these two events are associated with the integrity of the RCS pressure boundary and secondary system pressure boundary, the use of ~5~~ cladding will have no impact on the two AORs using CE methodology.

The non-DNBR limiting events, which were analyzed using the OPPD-NA-8303 documented CE methodology will be evaluated during the reload process to determine whether they can either be dispositioned or a reanalysis is necessary. If it is determined that a reanalysis is required, those transients and accidents shall be reanalyzed using the EMF-23 1O(P)(A) methodology.

The AREVA topical report, BAW-10240(P)(A) (Reference 8.7) justifies the use of ~5~~ with all of the Frarnatome ANP topical reports referenced in section 5.9 of the Ft. Calhoun Technical Specifications and section 2.0 of the COLR except for the topical report ANF-89-15 l(P)(A). The topical report ANF-89-15 1(P)(A) contains a non-LOCA transient analysis methodology similar to EMF-2310(P)(A). The change in cladding material from zircaloy-4 to ~5~~

potentially affects the fuel rod modeling in ANF-89-15 l(P)(A), ANF 151(P)(A) addresses the fuel rod modeling that is applied in the ANF-RELAP code; specifically, confirmatory RODEX2 calculations are performed for a given fuel design to verify that the pellet-to-clad gap conductance is within the range of the sensitivity study performed in support of the ANF 15l(P)(A)

LIC-05-0127 Page 11 methodology. RODEX2 calculations for a fuel design similar to that for FCS have verified that gap conductance for fuel rods with ~5~~ cladding is within the range supported by the fuel rod sensitivity study performed for the ANF 15l(P)(A) methodology. Further, the gap conductance used in the ANF-RELAP analyses is within the range supported by the fuel rod sensitivity study. Thus the ANF 151(P)(A) approved methodology does not require revision to support

~5~~ cladding.

Those transient and accident analyses performed using ANF-89-15 1(P)(A) methodology will be examined during every cycle reload process to determine whether those analyses can either be dispositioned or a reanalysis is necessary. If it is determined that a reanalysis is required, those transients and accidents shall be reanalyzed using the EMF-231O(P)(A) methodology.

The report BAW-10240(P)(A) demonstrates that neutronic, non-LOCA and DNB related topical reports do not require revision to address the use of ~5~~

cladding. There is a negligible impact of the ~5~~ cladding on non-LOCA transients and no impact on the DNB correlations or neutronic methods. It is thus concluded that it is acceptable to reference these topical reports in the FCS TS without modification for ~5~~ cladding.

OPPD has submitted a separate, but related license amendment request based on the AREVA RLBLOCA analysis as Reference 8.15.

5.2.4 Analysis AREVA has incorporated NRC approved ~5~~ material properties (Reference 8.1) into a set of approved AREVA methodologies for fuel mechanical analysis, realistic large break LOCA analysis, small break LOCA analysis and non-LOCA analysis (Reference 8.7). AREVA has performed an evaluation of the LOCA and non-LOCA performance of the ~5~~ cladding alloy for the generic LOCA and non-LOCA accident scenarios described in Reference 8.1. A comparison of results obtained using the base evaluation model methods with zircaloy-4 cladding and the results obtained for an identical case using the ~5~~ swelling and rupture model shows that the ~5~~ cladding performance should not adversely affect core operation or operating limits.

5.2.5 Conclusion 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

LIC-05-0127 Page 12 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.

6.0 ENVIRONMEIVTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 5 1.22(~)(9).Therefore, pursuant to 10 CFR 5 1.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 PRECEDENCE 7.1 Letter from David E. LaBarge QWC) to William R. McCollum (Duke Energy) dated June 21,2000, Oconee Nuclear Station Units l , 2 , and 3, Re: Issuance of Amendments (TAC Nos. MA8674, MA8675, and MA8676) (ML003726452) 7.2 Letter from Douglas V. Pickett (NRC) to Guy G. Campbell (FirstEnergy) dated March 15,2000, Issuance of Amendment - Davis Besse Station (TAC No.

MA3552) (ML003696350) 7.3 Letter from Timothy G. Colbum (NRC) to Mark E. Warner (Amergen Energy Company) dated May 10,2001, TMI-1 Amendment Re: Expanded Use of M5 Cladding Alloy (TAC No. MB0788) (ML011300351) 7.4 Letter from Brenda Mozafari (NRC) to Dale E. Young (Crystal River Plant) dated October 1,2003, Crystal River Unit 3 - Issuance of Amendment Regarding Technical Specification Change Request For the Use of M5 Advanced Alloy Fuel Cladding (TAC No. MB6590) (ML032760276) 7.5 Letter from Stephen Monarque (NRC) to David A, Christian (Virginia Electric and Power Company) dated April 1,2004, North Anna Power Station, Unit 2 -

Issuance of Amendment Re: Use of Framatome ANP Advanced Mark-BW Fuel (TAC NO. MB47 15) (ML040960040)

LIC-05-0127 Page 13

8.0 REFERENCES

8.1 BAW- 10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," Framatome Cogema Fuels, June 2003.

EMF 153(P)(A), EMF 153(P)(A) Supplement 1, "HTP: Departure From Nucleate Boiling Correlation For High Thermal Performance Fuel," Siemens Power Corporation, March 1994.

EMF-2 103(P)(A), Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," Framatome ANP Richland, Inc., April 2003.

EMF-2328(P)(A), Revision 0, "PWR Small Break LOCA Model, S-RELAP5 Based," Framatome ANP, Inc., March 2001.

XN-75-21(P)(A), Revision 2, "XCOBRA-IIIC, A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation,"

Exxon Nuclear Company, January 1986.

EMF-23 1O(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP, Inc., May 2004.

BAW-10240(P)(A), Revision 0, "Incorporation of ~5~~ Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., May 2004.

EMF-1961(P)(A), Revision 0, "Statistical SetpointITransient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation, July 2000.

ANF 151(P)(A), Revision 0, "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, May 1992.

Letter from Ross T. Ridenoure (OPPD) to Document Control Desk (NRC) dated August 11,2005, Fort Calhoun Station Unit No. 1 - License Amendment Request to Support Use of M5 Fuel Cladding, and 10 CFR 50.46 and 10 CFR Appendix K Exemption Request (LIC-05-0089) (ML052240083)

OPPD-NA-8301, Revision 8, "Omaha Public Power District Reload Core Analysis Methodology Overview."

8.12 OPPD-NA-8302, Revision 6, "Omaha Public Power District Reload Core Analysis Methodology, Neutronics Design Methods and Verification."

8.13 OPPD-NA-8303, Revision 6, "Omaha Public Power District Reload Core Analysis Methodology, Transient and Accident Methods and Verification."

LIC-05-0 127 Page 14 8.14 Letter from Alan B. Wang (NRC) to R. T. Ridenoure (OPPD) dated March 11, 2005, Fort Calhoun Station , Unit No. 1 - Issuance of Amendment 233, (NRC 003 1) (ML050750534) 8.15 Letter from R. T. Ridenoure (OPPD) to Document Control Desk (NRC) dated September 30, 2005, Fort Calhoun Station Unit No. 1 - License Amendment Request to Support Use of AREVA Realistic Large Break Loss of Coolant Accident Methodology (LIC-05-0106) (ML052770174)

LIC-05-0 127 Page 1 ATTACHMENT 2 Markup of Technical Specification Pages

TECHNICAL SPECIFICATIONS 4.0 DESIGN FEATURES 4.1 Site The site for Fort Calhoun Station Unit No. 1 is in Washington County, Nebraska, on the west bank of the Missouri River and approximately nineteen miles north, northwest of the city of Omaha, Nebraska. The exclusion area, as defined in 10 CFR Part 100, Section 100.3(a), consists of approximately 1242 acres. The exclusion area boundary extent includes approximately 660 acres in Washington County, Nebraska, owned by the Omaha Public Power District (OPPD), and 582 acres in Harrison County, Iowa, on the east bank of the river directly opposite the facility, on which the District retains perpetual easement rights. The minimum exclusion area boundary point is located approximately at the 187.0 degree radial from the outer wall of the containment building and at a distance of 910 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 133 fuel assemblies. Each assembly shall consist of a matrix of zircaloy, e~ZIRLOB, or M5 clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 49 control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, or hafnium metal as approved by the NRC.

4.3 Fuel Storaqe Criticalitv 4.3.1 .I The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.5 weight percent,

b. keff 10.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.5 of the USAR, 4.0 - Page 1 Amendment Nos. 28,36,?C-)9,

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued) 1 OPPD-NA-8301, "Reload Core Analysis Methodology Overview" approved version as specified in the COLR.

2. OPPD-NA-8302, "Neutronics Design Methods and Verification",

approved version as specified in the COLR.

3. OPPD-NA-8303, "Transient and Accident Methods and Verification",

approved version as specified in the COLR.

4. WCAP-12610-P-A, "VANTAGE + Fuel Assembly Report," April 1995 (Westinghouse Proprietary) as approved in the Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 178 to Facility Operating License No. DPR-40, Omaha Public Power District, Fort Calhoun Station Unit No. 1, Docket No. 50-285, dated October 25, 1996.

5. XN-75-32(P)(A) Supplements 1, 2, 3, & 4, "Computational Procedure for Evaluating Fuel Rod Bowing," approved version as specified in the COLR.

6. XN-NF-82-06(P)(A) and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

7. XN-NF-85-92(P)(A), "Exxon Nuclear Uraniurr~DioxideIGadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

8. ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWdIMTU," approved version as specified in the COLR.

9. EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.

10. BA W-10240(P)(A), "Incorporation of ~5~~ Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., approved version as specified in the COLR.

c. The core operating limits shall be determined so that all applicable limits (e.g.,

f i ~ ethermal l mechanical limits, core thermal hydraulics limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as shutdown margin (SDM),

transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.0 - Page 9 Amendment No.

1 4 1 l A 4 1 C 7 -7 1 1 1 ' , I ' J 1 , I 1 , 1 9

LIC-05-0127 Page 1 ATTACHMENT 3 Clean Typed Technical Specification Pages

TECHNICAL SPECIFICATIONS 4.0 DESIGN FEATURES The site for Fort Calhou~iStation Unit No. 1 is in Washington County, Nebraska, on the west bank of the Missouri River and approximately nineteen miles north, northwest of the city of Omaha, Nebraska. The exclusion area, as defined in 10 CFR Part 100, Section 100.3(a), consists of approximately 1242 acres. The exclusion area boundary extent includes approximately 660 acres in Washington County, Nebraska, owned by the Omaha Public Power District (OPPD), and 582 acres in Harrison County, Iowa, on the east bank of the river directly opposite the facility, on which the District retains perpetual easement rights. The minimum exclusion area boundary point is located approximately at the 187.0 degree radial from the outer wall of the containment building and at a distance of 910 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 133 fuel assemblies. Each assembly shall consist of a matrix of zircaloy, ZIRLOB, or M5 clad fuel rods with an initial I composition of natural or slightly enriched uranium dioxide (UOp) as fuel material. Lirr~itedsubstitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 49 control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, or hafnium metal as approved by the NRC.

4.3 Fuel Storaqe 4.3.1 Criticalitv 4.3.1 .I The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.5 weight percent,

b. k , ~50.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.5 of the USAR, 4.0 - Page 1 Amendment No. 20,3&4@&

47Q I , V, 7'25 LV

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reportinq Requirements (Continued)

1. OPPD-NA-8301, "Reload Core Analysis Methodology Overview" approved version as specified in the COLR.

2. OPPD-NA-8302, "Neutronics Design Methods and Verification",

approved version as specified in the COLR.

3. OPPD-NA-8303, "Transient and Accident Methods and Verification",

approved version as specified in the COLR.

4. WCAP-12610-P-A, "VANTAGE + Fuel Assembly Report," April 1995 (Westinghouse Proprietary) as approved in the Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.

178 to Facility Operating License No. DPR-40, Omaha Public Power District, Fort Calhoun Station Unit No. I,Docket No. 50-285, dated October 25, 1996.

5. XN-75-32(P)(A) Supplements I, 2, 3, & 4, "Computational Procedure for Evaluating Fuel Rod Bowing," approved version as specified in the COLR.

6. XN-NF-82-06(P)(A) and Supplements 2,4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.

7. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium DioxideIGadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

8. ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWdIMTU," approved version as specified in the COLR.

9. EMF-92-1 16(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.

10. BAW-10240(P)(A), "Incorporation of ~5~~ Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., approved version as specified in the COLR..

c. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulics limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as shutdown margin (SDM), transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.0 - Page 9 Amendment No.