Text

License Amendment Request, Revision to Technical Specification 2.1.1.1 Reactor Core Safety Limits
	ML13162A736
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/11/2013
From:	Jeremy G. Browning; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	1CAN061301
	Download: ML13162A736 (14)
	v • d • e

1CAN061301 June 11, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request Revision to Technical Specification 2.1.1.1 Reactor Core Safety Limits Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

REFERENCES:

1. BAW-10231(P)(A), Revision 1,COPERNIC Fuel Rod Design Computer Code, January 2004 (ML040150701)

2. NRC Letter to Mr. Joseph W. Shea, Manager, Corporate Nuclear Licensing, Tennessee Valley Authority, dated September 26, 2012, Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Revise the Technical Specification to Allow Use of AREVA Advanced W17 High Thermal Performance Fuel (TS-SQN-2011-07)

(TAC Nos. ME6538 and ME6539) (ML12249A394)

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests an amendment to the Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS) 2.1.1.1. The requested amendment is to add the determination of the maximum local fuel pin centerline temperature using the NRC reviewed and approved COPERNIC fuel performance computer code (Reference 1). The ANO-1 TSs currently provide similar information for other fuel performance computer codes.

provides a description and assessment of the proposed changes including the technical analyses; regulatory analyses; and environmental considerations. Attachment 2 provides markup pages of existing TSs and TS Bases to show the proposed change. provides revised (clean) TS pages.

Reference 2 provides the NRC approval of a similar request.

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Jeremy G. Browning Vice President - Operations Arkansas Nuclear One

1CAN061301 Page 2 of 3 Entergy requests approval of the proposed license amendment by July 1, 2014, with the amendment being implemented within 90 days of approval.

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

In accordance with 10 CFR 50.91(b)(1), a copy of this application is being provided to the designated Arkansas state official.

If you have any questions or require additional information, please contact Stephenie Pyle at 479-858-4704.

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

Executed on June 11, 2013.

Sincerely, Original signed by Mike Chisum for Jeremy Browning JGB/rwc Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification and Bases Changes (mark-up)

3. Proposed Technical Specification Changes (clean)

1CAN061301 Page 3 of 3 cc:

Mr. Arthur T. Howell Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam MS O-8B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

to 1CAN061301 Analysis of Proposed Technical Specification Change to 1CAN061301 Page 1 of 4 DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE

1.0 DESCRIPTION

This letter is a request to amend Operating License DPR-51 for Arkansas Nuclear One, Unit 1 (ANO-1).

The maximum fuel centerline melt temperature is given by the relationships defined by Technical Specification (TS) 2.1.1.1 for the respective fuel designs and is dependent upon which computer code is used in the analysis. The ANO-1 TS requirements are revised to include another fuel performance computer code used to determine the maximum fuel centerline temperature. This code, COPERNIC, has previously been reviewed and approved by the NRC (Reference 1).

2.0 BACKGROUND

Fuel centerline melting occurs when the local linear heat rate, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of radioactivity to the reactor coolant.

The COPERNIC code is an improved fuel performance code for fuel rod design and analysis of natural, slightly enriched (up to 5 percent) uranium dioxide fuels and urania-gadolinia fuels with advanced cladding material, M5. The code includes fuel thermal conductivity degradation (TCD) with burnup. TCD is a physical phenomenon caused by irradiation damage and the progressive buildup of fission products in fuel pellets resulting in reduced thermal conductivity of the pellets. The COPERNIC code was approved to a peak rod average burnup of 62 gigawatt-days per metric ton of uranium (GWD/MTU) (Reference 1).

The ANO-1 Cycle 24 was completed on March 24, 2013. During the design of the reload core for Cycle 24, the COPERNIC code was used to determine the fuel centerline temperature, transient cladding strain, and internal pin pressure. ANO-1 TS 5.6.5 requires that the analytical methods utilized to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in Reference 2. The Cycle 24 Core Operating Limits Report (COLR) lists the latest revision of Reference 2 is Revision 8. Revision 8 includes Reference 1. The appropriate limits listed in the COLR were developed using the COPERNIC computer code.

It was not noted at the time of the reload core design efforts that the TS required revision.

The COLR limits and the TS limits protect the safety limits, including the fuel centerline melt temperature. In accordance with NRC Administrative Letter 98-10, Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety, administrative controls to ensure continued compliance have previously been established and will remain in place until this request is approved.

to 1CAN061301 Page 2 of 4

3.0 ASSESSMENT

The intent of the subject safety limit is to prevent fuel centerline temperature from reaching the melting point, which conservatively assures that there will be no breach in cladding integrity. The COPERNIC code was approved for use with M5 cladding material. M5 is currently in use in the ANO-1 fuel design.

In the safety evaluation for Reference 1, the NRC listed one condition. That condition is:

Licensees that reference this topical report still need to meet 10 CFR 51.52, Environmental effects of transportation of fuel and waste - Table S-4.

Upon further review of Reference 1 and the safety evaluation, it was determined that the environmental effects analyses of transportation in accordance with 10 CFR 51.52, is performed independent of COPERNIC analyses.

In accordance with 10 CFR 50, Appendix A, General Design Criteria (GDC) 10, Reactor Design, the acceptance criteria for normal operation and Anticipated Operating Occurrences (AOOs) is that the Specified Acceptable Fuel Design Limits (SAFDLs) not be exceeded. The SAFDL of interest is the Peak Fuel Centerline Temperature limit. This SAFDL is discussed in Section II(B)iv of Reference 3, which states in part:

Overheating of Fuel Pellets: It has also been traditional practice to assume that failure will occur if centerline melting takes place. For normal operation and anticipated operation occurrences, centerline melting is not permitted. The centerline melting criterion was established to assure that axial or radial relocation of molten fuel would neither allow molten fuel to come into contact with the cladding nor produce local hot spots. The assumption that centerline melting results in fuel failure is conservative.

The melting point of the fuel is dependent on fuel burnup. The best-estimate melting point of new fuel is 4901 °F based on the COPERNIC code. The melting temperature is adjusted downward from this temperature depending on the amount of burnup. The form and adjustment (decreasing by 13.7 °F per 10,000 MWD/MTU of burnup) for the COPERNIC code is consistent with the current determination using the TACO2 or TACO3 code.

4.0 REGULATORY ANALYSIS

4.1 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Entergy Operations, Inc. (Entergy) has evaluated the proposed changes to the TS using the criteria in Section 50.92 to Title 10 of the Code of Federal Regulations (10 CFR) and has determined that the proposed changes do not involve a significant hazards consideration.

The proposed amendment would revise TS requirements related to the fuel centerline melt temperature safety limit (TS 2.1.1.1). A new temperature versus burnup relationship based on the COPERNIC fuel performance computer code will be added to the current listing of the relationships using the TACO2 and TACO3 computer codes.

to 1CAN061301 Page 3 of 4 As required by 10 CFR 50.91(a), the Entergy analysis of the issue of no significant hazards consideration is presented below:

1.

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

Response: No The proposed change does not require any physical change to any plant systems, structures, or components, nor does it require any change in systems or plant operations. The proposed change does not require any change in safety analysis methods or results. Operations and analysis will continue to be in accordance with the ANO-1 licensing basis. The peak fuel centerline temperature is the basis for protecting the fuel and is consistent with safety analysis.

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

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 adds a new fuel centerline melt temperature versus burnup relationship based on an NRC reviewed and approved fuel performance computer code. The accident analyses presented in the ANO-1 Safety Analysis Report indicate that the fuel centerline temperature is not approached or exceeded for any of the events or Anticipated Operational Occurrences. The existing analyses, which are unchanged, do not affect any accident initiators that would create a new accident.

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 The proposed change does not require any change in safety analysis methods or results. Therefore, by adding the fuel centerline temperature and burnup relationship as defined by the COPERNIC code to the TS, the margin as established with the ANO-1 TS and SAR are unchanged.

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

to 1CAN061301 Page 4 of 4 4.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA ANO-1 was not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC), but is designed to be consistent with the GDCs applicable to this application as described above.

4.3 PRECEDENT Sequoyah Nuclear Plant, Units 1 and 2 filed an application for a Technical Specification change to allow the use of AREVAs Advanced W17 High Thermal Performance Fuel. As part of that application was a change to the Reactor Core Safety Limit associated with the local fuel pin centerline temperature. The change was based on the use of the COPERNIC computer code. The change proposed in the Sequoyah Safety Limit is identical to the change proposed in this request.

Reference 4 is the NRCs Safety Evaluation approving the Sequoyah request, including the change to the Safety Limit.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed TS change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change 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 TS change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed TS change.

6.0 REFERENCES

1.

BAW-10231(P)(A), Revision 1,COPERNIC Fuel Rod Design Computer Code, January 2004 (ML040150701)

2.

BAW-10179P-A, Safety Criteria and Methodology for Acceptable Cycle Reload Analyses, Rev. 8, Framatome ANP, Inc., Lynchburg, VA, May 2010.

3.

NUREG-0800, Standard Review Plan, Section 4.2, Fuel System Design, Revision 2, July 1981

4.

NRC Letter to Mr. Joseph W. Shea, Manager, Corporate Nuclear Licensing, Tennessee Valley Authority, dated September 26, 2012, Sequoyah Nuclear Plant, Units 1 and 2 -

Issuance of Amendments to Revise the Technical Specification to Allow Use of AREVA Advanced W17 High Thermal Performance Fuel (TS-SQN-2011-07) (TAC Nos. ME6538 and ME6539) (ML12249A394) to 1CAN061301 Proposed Technical Specification and Bases Changes (mark-up)

SLs 2.0 ANO-1 2.0-1 Amendment No. 215,226, 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, the maximum local fuel pin centerline temperature shall be 5080 - (6.5 x 10-3 x (Burnup, MWD/MTU)°F) for TACO2 applications, and 4642 - (5.8 x 10-3 x (Burnup, MWD/MTU)°F) for TACO 3 applications, or 4901 °F, decreasing by 13.7 °F per 10,000 MWD/MTU of burnup for COPERNIC applications.

2.1.1.2 In MODES 1 and 2, the departure from nucleate boiling ratio shall be maintained greater than the limits of 1.3 for the BAW-2 correlation, 1.18 for the BWC correlation, and 1.132 for the BHTP correlation.

2.1.1.3 In MODES 1 and 2, Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the Variable Low RCS Pressure-Temperature Protective Limits as specified in the Core Operating Limits Report, so that the safety limits are met.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2750 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed:

2.2.1 In MODE 1 or 2, if SL 2.1.1.1 or SL 2.1.1.2 is violated, be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 In MODE 1 or 2, if SL 2.1.1.3 is violated, restore RCS pressure and temperature within limits AND be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.3 In MODE 1 or 2, if SL 2.1.2 is violated, restore compliance within limits AND be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.4 In MODES 3, 4, and 5, if SL 2.1.2 is violated, restore RCS pressure to 2750 psig within 5 minutes.

2.2.5 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , notify the NRC Operations Center, in accordance with 10 CFR 50.72.

Reactor Core SLs B 2.1.1 ANO-1 B 2.1.1-1 Amendment No. 215,218 Rev. 9, B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires that reactor core SLs ensure specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and abnormalities.

This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (95/95 DNB criterion) that DNB will not occur and by requiring that the fuel centerline temperature stays below the melting temperature.

Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature and pressure can be related to DNB through the use of a critical heat flux (CHF) correlation. The BAW-2 (Ref. 2), BWC (Ref. 3),

and BHTP (Ref. 9) correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The BAW-2 correlation applies to Mark-B fuel, the BWC correlation applies to Mark-BZ fuel, and the BHTP correlation applies to Mark-BHTP fuel. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady state operation, normal operational transients and anticipated transients is limited to 1.30 (BAW-2), 1.18 (BWC), and 1.132 (BHTP).

The 95 percent confidence level that DNB will not occur is preserved by ensuring that the DNBR remains greater than the DNBR design limit based on the applicable CHF correlation for the core design. In the development of the applicable DNBR design limit, uncertainties in the core state variables, power peaking factors, manufacturing-related parameters, and the CHF correlation may be statistically combined to determine a statistical DNBR design limit. This statistical design limit protects the respective CHF design limit. Additional retained thermal margin may also be applied to the statistical DNBR design limit to yield a higher thermal design limit for use in establishing DNB-based core safety and operating limits. In all cases, application of statistical DNB design methods preserves a 95 percent probability at a 95 percent confidence level that DNB will not occur (Ref. 4).

The restrictions of this SL prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor coolant.

Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. The maximum fuel centerline temperatures are given by the relationships defined in SL 2.1.1.1 for the respective fuel designs and are dependent on whether the TACO2 (Ref. 5), or TACO3 (Ref. 6), or COPERNIC (Ref. 10) analysis was utilized. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Reactor Core SLs B 2.1.1 ANO-1 B 2.1.1-4 Amendment No. 215 Rev. 9, SAFETY LIMIT VIOLATIONS The following SL violation responses are applicable to the reactor core SLs.

2.2.1 AND 2.2.2 If SL 2.1.1.1, SL 2.1.1.2, or SL 2.1.1.3 is violated, the requirement to go to MODE 3 places the plant in a MODE in which these SLs are not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the plant to a MODE of operation where these SLs are not applicable and reduces the probability of fuel damage.

2.2.5 If SL 2.1.1.1, SL 2.1.1.2, or SL 2.1.1.3 is violated, the NRC Operations Center must be notified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in accordance with 10 CFR 50.72 (Ref. 8).

REFERENCES

1.

SAR, Section 1.4, GDC 10.

2.

BAW-10000A, Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, Babcock & Wilcox, Lynchburg, VA, May 1976.

3.

BAW-10143P-A, BWC Correlation of Critical Heat Flux, Babcock & Wilcox, Lynchburg, VA, April 1985.

4.

BAW-10179P-A, Safety Criteria and Methodology for Acceptable Cycle Reload Analyses, Rev. 68, Framatome ANP, Inc.Babcock & Wilcox, Lynchburg, VA, May 2010August 2005.

5.

BAW-10141P-A, Rev. 1, TACO2 Fuel Pin Performance Analysis, Babcock & Wilcox, Lynchburg, VA, June 1983.

56.

BAW-10162P-A, TACO3 Fuel Pin Thermal Analysis Code, Babcock & Wilcox, Lynchburg, VA, October 1989.

7.

SAR, Chapters 3 & 14.

8.

10 CFR 50.72.

9.

BAW-10241P-A, Rev. 1, BHTP DNB Correlation Applied with LYNXT, Framatome ANP, Lynchburg, VA, July 2005.

10.

BAW-10231(P)(A), Rev. 1, COPERNIC Fuel Rod Design Computer Code, January 2004.

to 1CAN061301 Proposed Technical Specification Changes (clean)

SLs 2.0 ANO-1 2.0-1 Amendment No. 215,226, 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, the maximum local fuel pin centerline temperature shall be 5080 - (6.5 x 10-3 x (Burnup, MWD/MTU)°F) for TACO2 applications, 4642 - (5.8 x 10-3 x (Burnup, MWD/MTU)°F) for TACO 3 applications, or 4901 °F, decreasing by 13.7 °F per 10,000 MWD/MTU of burnup for COPERNIC applications.