Text

Clarification of Proposed Change to Technical Specification 2.1.1.1 Reactor Core Safety Limits
	ML13347B236
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/11/2013
From:	Jeremy G. Browning; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	1CAN121301
	Download: ML13347B236 (8)
	v • d • e

s Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Jeremy G. Browning Site Vice President Arkansas Nuclear One 1CAN121301 December 11, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Clarification of Proposed Change to Technical Specification 2.1.1.1 Reactor Core Safety Limits Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

REFERENCE:

Entergy letter dated June 11, 2013, License Amendment Request - Revision to Technical Specification 2.1.1.1 Reactor Core Safety Limits" (1CAN061301)

(ML13162A736)

Dear Sir or Madam:

Via the reference, Entergy Operations, Inc. (Entergy) requested an amendment to the Arkansas Nuclear One, Unit 1 Technical Specification 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.

During the review of the reference submittal, the NRC determined that the burnup dependence of the proposed limit was unclear. It could be interpreted as a step decrease or a linear function that decreases with burnup. Subsequently the NRC requested the burnup dependence of the proposed fuel pin centerline temperature limit be clarified.

During a November 6, 2013, conference call with the NRC, Entergy clarified that the burnup dependence is a linear function. Entergy stated that the reference request would be revised to include this clarification. Attached are the revised pages.

With respect to the referenced Entergy request, changes proposed in this letter have been evaluated and Entergy has determined that the changes do not invalidate the assessment of the no significant hazards consideration included in the reference letter.

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

No new regulatory commitments have been identified in this letter.

1CAN121301 Page 2 of 2 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 November 11, 2013.

Sincerely, Original signed by Jeremy G. Browning JGB/rwc Attachments:

1. Replacement Markups of Technical Specification and Technical Specification Bases Pages

2. Replacement Clean (Revised) Technical Specification Pages cc: Mr. Marc L. Dapas 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

Attachment 1 to 1CAN121301 Replacement Markups of Technical Specification and Technical Specification Bases Pages

SLs 2.0 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, and £ 4901 °F, decreasing linearly 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.

ANO-1 2.0-1 Amendment No. 215,226,

Reactor Core SLs B 2.1.1 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.

ANO-1 B 2.1.1-1 Amendment No. 215,218 Rev. 9,

Reactor Core SLs B 2.1.1 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.

ANO-1 B 2.1.1-4 Amendment No. 215 Rev. 9,

Attachment 2 to 1CAN121301 Replacement Clean (Revised) Technical Specification Pages

SLs 2.0 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, and £ 4901 °F, decreasing linearly by 13.7 °F per 10,000 MWD/MTU of burnup for COPERNIC applications.