RS-11-073, License Amendment Request to Modify Technical Specifications Section 3.1.2, Reactivity Anomalies

License Amendment Request to Modify Technical Specifications Section 3.1.2, Reactivity Anomalies

ML111650145
Person / Time
Site:	Clinton
Issue date:	06/13/2011
From:	Hansen J Exelon Nuclear, Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-11-073
Download: ML111650145 (16)

Text

www.exelonc RS-1 1-073 June 13, 2011 Nuclear 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

License Amendment Request to Modify Technical Specifications Section 3.1.2, "Reactivity Anomalies" In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) of Facility Operating License No. NPF-62 for Clinton Power Station, Unit 1 (CPS).

The proposed amendment would revise the Limiting Condition for Operation (LCO) 3.1.2, "Reactivity Anomalies," through a revision to the method for calculating core reactivity for the purpose of performing an anomaly check. This request is subdivided as follows:

• provides an evaluation of the proposed change.

• provides the markup pages of existing TS to show the proposed changes.

• provides the markup pages of the existing TS Bases to show the proposed changes for information only and do not require NRC approval.

EGC requests approval of the proposed license amendment by June 13, 2012, with the amendment being implemented within 60 days.

U. S. Nuclear Regulatory Commission June 13, 2011 Page 2 In accordance with 10 CFR 50.91(a)(1), "Notice for Public Comment," the analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the NRC.

The proposed amendment has been reviewed by the CPS Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC is notifying the State of Illinois of this application for a change to the TS by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b).

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mitchel Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 13th day of June 2011.

Respectfully, J

y L-Hansen Manager - Licensing and Regulatory Affairs Attachments:

1.

Evaluation of Proposed Changes 2.

Markup pages of existing TS to show the proposed changes.

3.

Markup pages of existing TS Bases to show the proposed changes - for information only.

cc:

Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Changes

Subject:

License Amendment Request to Modify Technical Specifications Section 3.1.2, "Reactivity Anomalies" 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

6.0 REFERENCES

Page 1 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend the Technical Specifications (TS) of Operating License No. NPF-62 for Clinton Power Station, Unit 1 (CPS).

The proposed change would revise TS LCO 3.1.2, "Reactivity Anomalies," to allow performance of the surveillance on a comparison of predicted to monitored core reactivity. The reactivity anomaly verification is currently determined by a comparison of predicted vs. actual control rod density.

2.0 DETAILED DESCRIPTION The purpose of the reactivity anomaly surveillance is to compare the observed reactivity behavior of the core at hot operating conditions with the expected reactivity behavior calculated prior to the start of operation.

Currently, CPS TS require a comparison between predicted control rod density that is calculated prior to the start of operation for a particular cycle to an actual control rod density during the cycle. The comparison is done, as required by Surveillance Requirement (SR) 3.1.2.1, once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement, and each 1000 MWD/T thereafter during operations in Mode 1. This proposed TS change will not change the frequency of the SR, only the method by which the reactivity anomaly SR is performed.

The current LCO 3. 1.2 reads:

The reactivity difference between the monitored rod density and the predicted rod density shall be within +/- 1 % Ak/k.

The proposed LCO 3. 1.2 reads:

The reactivity difference between the monitored core keff and the predicted core keff shall be within +/- 1 % Ek/k.

The Condition statements will not be changed.

The SR will also be re-worded to replace rod density with keffective (keff) as appropriate.

The current method of performing the reactivity anomaly uses rod density for the comparison primarily because early core monitoring systems did not calculate core critical keff values for comparison to design values. Rod density was used instead as a convenient representation of core reactivity.

Allowing the use of a direct comparison of keff, as opposed to rod density, provides for a more direct measurement of core reactivity conditions and eliminates the limitations that exist for performing the core reactivity comparisons with rod density.

Page 2 of 6

ATTACHMENT 1 Evaluation of Proposed Changes

3.0 TECHNICAL EVALUATION

If a significant deviation between the reactivity observed during operation and the expected reactivity occurs, the reactivity anomaly surveillance alerts the reactor operating staff of a potentially anomalous situation, indicating that something in the core design process, the manufacturing of the fuel, or in the plant operation may be different than assumed. This situation would trigger an investigation and further actions as needed.

The current method for the development of the reactivity anomaly curves used to perform the TS surveillance actually begins with the predicted critical keff at rated conditions and the companion rod patterns derived using those predicted values of keff. A calculation is made of the number of notches inserted in the rod patterns, and also the number of average notches required to make a change of +/- 1 % Ak/k around the predicted critical keff. The notches are converted to rod density and plotted with an upper and lower bound representing the +/- 1% Ak/k acceptance band as a function of cycle exposure. This curve is then used as the predicted rod density during the cycle. In effect, the comparison is still based on critical keff with a "translation" of acceptance criteria to rod density.

CPS utilizes the Global Nuclear Fuel (GNF) 3D MONICORE core monitoring software system.

NRC acceptance of the 3D MONICORE core surveillance system for power distribution uncertainties is documented in Reference 1. The latest version of this product incorporates the PANACEA Version 11 (i.e., PANAC1 1) core simulator code to calculate parameters such as core nodal powers, fuel thermal limits, etc., using actual, measured plant input data. NRC acceptance of PANACEA Version 11 is documented in Reference 2. PANAC1 1 is the same 3D core simulator code currently used in CPS core design and licensing activities.

When a 3D MONICORE core monitoring case is run, the core keff, as computed by PANAC1 1, is also calculated and printed directly on each 3D MONICORE case output. This value can then be directly compared to the predicted value of keff as a measure of reactivity anomaly.

The revised method for evaluating a potential reactivity anomaly compares the measured (i.e., monitored) core Keff to predicted core keff. Measured core keff is calculated by the 3D core simulator model in the plant's core monitoring system based on measured plant operating data.

The predicted core keff, as a function of cycle exposure, is developed prior to the start of each operating cycle and incorporates benchmarking of exposure-dependent 3D core simulator keff behavior in previous cycles and any fuel vendor recommended adjustments due to planned changes in fuel design, core design, or operating strategy for the upcoming cycle.

While being a convenient measurement of core reactivity, control rod density has its limitations, most obviously that all control rod insertion does not have the same impact on core reactivity.

For example edge rods and shallow rods have very little impact on reactivity while deeply inserted central control rods have a large effect. Thus, it is not uncommon for reactivity anomaly concerns to arise during operation simply because of greater use of near-edge or shallow control rods than anticipated, when in fact no true anomaly exists. Use of actual to predicted keff instead of rod density eliminates the limitations described above, provides for a technically superior comparison, and is a very simple and straightforward approach.

Page 3 of 6

ATTACHMENT I Evaluation of Proposed Changes These proposed changes will not affect transient and accident analyses because only the method of performing the reactivity anomaly surveillance is changing, and the proposed method will provide an adequate estimate as discussed above. Furthermore, the anomaly check will continue to be performed at the current required frequency.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria General Design Criteria 26, 28, and 29 require that reactivity be controllable such that subcriticality is maintained under cold conditions and specified applicable fuel design limits are not exceeded during normal operations and anticipated operational occurrences.

The reactivity anomaly check required by the CPS TS in LCO 3.1.2 serves to partly satisfy the above General Design Criteria by verifying that core reactivity remains within expected/predicted values.

Ensuring that no reactivity anomaly exists provides confidence of adequate shutdown margin as well as providing verification that the assumptions of safety analyses associated with core reactivity remain valid.

4.2 Precedent This Technical Specifications amendment was granted to the Edwin I. Hatch Nuclear Plant (TAC NOS. ME3006 AND ME3007) on November 4, 2010, via Amendment Nos. 263 and 207 for Units 1 and 2, respectively (i.e., Reference 3).

Additionally, the Reactivity Anomaly LCO in the BWR/6 Standard Technical Specifications, NUREG-1434, Revision 3.0 (i.e., Reference 4), is written with the keff comparison, as opposed to the control rod density comparison. The Bases changes are modeled after the applicable BWR/6 Standard Bases.

4.3 No Significant Hazards Consideration Determination The proposed changes revise the reactivity anomaly Technical Specifications (TS) limiting condition for operation (LCO) 3. 1.2, "Reactivity Anomalies," to allow a direct measurement of core reactivity by using the effective multiplication factor (i.e., keff) rather than the control rod density.

Exelon Generation Company, LLC (EGC) has evaluated the proposed changes to the TS using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. An 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 Page 4 of 6

ATTACHMENT 1 Evaluation of Proposed Changes This proposed TS change does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The amendment would only change how the reactivity anomaly check is performed. Verifying that the core reactivity is consistent with predicted values ensures that accident and transient safety analyses remain valid. This amendment changes the LCO 3.1.2 and Surveillance Requirement (SR) 3.1.2.1 requirements such that the check is performed by a direct comparison of keff rather than by comparing predicted to actual control rod density.

On-line core monitoring systems, such as the one currently in use at Clinton Power Station, Unit 1 (CPS), are capable of performing the direct measurement of reactivity.

Therefore, since the reactivity anomaly check will continue to be performed by a viable method, the proposed amendment does not involve a significant increase in the probability or consequence of any previously evaluated accident.

2.

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

Response: No This TS amendment request does not involve any changes to the operation, testing, or maintenance of any safety -related, or otherwise important to safety, system. All systems that are important to safety will continue to be operated and maintained within their design bases. The proposed changes to LCO 3.1.2 and SR 3.1.2. 1 will only provide a new, efficient method of detecting an unexpected change in core reactivity.

Since all systems continue to be operated within their design bases, no new failure modes are introduced, nor is the possibility of a new or different kind of accident created.

3.

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

Response: No This proposed TS amendment proposes to change the method for performing the reactivity anomaly surveillance from a comparison of predicted to actual control rod density to a comparison of predicted to actual keff. The direct comparison of keff provides a more direct method of calculating any differences in the expected core reactivity. The reactivity anomaly check will continue to be performed at the same frequency as is currently required by the TS, only the method of performing the check will be changed.

Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified.

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

Based upon the above analysis, EGC concludes that the requested change does not involve a significant hazards consideration, as set forth in 10 CFR 50.92(c), "Issuance of Amendment," and, accordingly, a finding of no significant hazards consideration is justified.

Page 5 of 6

ATTACHMENT I Evaluation of Proposed Changes 4.4 Conclusions 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 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.

5.0 ENVIRONMENTAL 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, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the amounts of any effluents that may be released off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria 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 amendment.

6.0 REFERENCES

1. MFN-003-99, F. Akstulewicz (NRC) to G. Watford (GE), Safety Evaluation Report for GE Licensing Topical Report NEDC-32694P, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations" (TAC No. M99069), March 11, 1999 2.

MFN-035-99, S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-2401 1 -P-A, "GESTAR II" -Implementing Improved GE Steady-State Methods (TAC No. MA6481)," November 10, 1999 3.

Letter from R. E. Martin (NRC) to M. J. Ajluni (Southern Nuclear Operating Company, Inc),

"Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Issuance of Amendments Regarding Revision to Technical Specifications Limiting Condition for Operation 3.1.2, "Reactivity Anomalies" (TAC NOS. ME3006 AND ME3007)," dated November 4, 2010.

4.

NUREG-1434, Volume 1, Revision 3.0, "Standard Technical Specifications General Electric Plants, BWR/6 Specifications," dated June 2004 Page 6of6

ATTACHMENT 2 CLINTON POWER STATION, UNIT 1 Docket No. 50-461 License No. NPF-62 Markup Pages of Existing Technical Specifications to Show the Proposed Changes MARKUP OF EXISTING REVISED TS PAGES 3.1-5 3.1-6

Reactivity Anomalies 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Reactivity Anomalies LCO 3.1.2 The reactivity difference between the monitored rod density and the predicted rod de city shall be within +/- 1% Ak/k.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> A.

Core reactivity difference not within limit.

A.1 Restore core reactivity difference to within limit.

B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> CLINTON 3.1-5 Amendment No. 95

Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1.2.1 Verify core reactivity difference between the monitore and the predicted core keff

^,a a;ii ; E.. is within +/- 1°s Ak/k.

FREQUENCY Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 1000 MWD/T thereafter during operation in MODE 1 CLINTON 3.1-6 Amendment No.

ATTACHMENT 3 CLINTON POWER STATION UNIT 1 Docket No. 50-461 License No. NPF-62 Markup Pages of Existing Technical Specifications Bases to Show the Proposed Change MARKUP OF EXISTING TS BASES PAGES B 3.1-9 B 3.1-10 B 3.1-11 B 3.1-12

Reactivity Anomalies B 3.1.2 BASES BACKGROUND is critical at RTP, the excess positive reactivity is (continued) compensated by burnable absorbers (if any), control rods, and whatever neutron poisons (mainly xenon and samarium) are present in the fuel.

core keffective (keff)

The measured or The predicted core reactivity, as represented by eentro1 red monitored core keff dente, is calculated by a 3D core simulator code as a is calculated by function of cycle exposure.

This calculation is performed for projected operating states and conditions throughout the the core Ncycle.

se monitoring system for actual plant conditions and is then compared to the predicted value for the cycle exposure.

APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluation (Ref. 2).

In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core reactivity.

These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks.

Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity.

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and core keff Predicte for identical core conditions at BOC do not re sonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used to predicred Gle sit), may not be accurate. If reasonable con agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured value.

Thereafter, any significant deviations in the measurecjr^a dens ty from the predicted rod de ity hat eff develop during fuel depletion may be an indication that 'the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred.

Reactivity anomalies satisfy Criterion 2 of the NRC Policy Statement.

(continued) core keff CLINTON B 3.1-9 Revision No. G

Reactivity Anomalies B 3.1.2 BASES (continued)

The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses.

Large differences between monitored and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected.

A limit on the difference between the monitore re density and the predicted remit-f 1% Ok/k has been established based on engineering judgment.l A > 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.

APPLICABILITY In MODE 1, most of the control rods are withdrawn and steady state operation is typically achieved.

Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly.

In MODE 2, control rods are typically being withdrawn during a startup. In MODES 3 and 4, all control rods are fully inserted, and, therefore, the reactor is in the least reactive state, where monitoring core reactivity is not necessary. In MODE 5, fuel loading results in a continually changing core reactivity. SDM requirements (LCO 3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and an SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling).

The SDM test, required by LCO 3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions; therefore, reactivity anomaly is not required during these conditions.

ACTIONS A_1 Should an anomaly develop between measured and predicted core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions. Restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly.

This evaluation normally reviews the core (continued)

LCO core keff core keff CLINTON B 3.1-10 Revision No. 4

Reactivity Anomalies B 3.1.2 BASES ACTIONS A.1 (continued) conditions to determine their consistency with input to design calculations.

Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions.

The required Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is based on the low probability of a DBA during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.

B.1 If the core reactivity cannot be restored to within the 1% Ak/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Verifying the reactivity difference between the monitored core keff and predicte re-'

-1enS41-ZX is within the limits of the LCO provides furt er assurance that plant operation is maintained within the assumptions of the DBA and transient analyses.

The Core Monitoring System calculates the r-ad deny for the reactor conditions obtained from plant instrumentation.

A comparison of the monitored rod density to the predicteda rod density at the same cycle exposure is core keff used to calculatre the reactivity difference.

The comparison is required when the core reactivity has potentially changed by a significant amount. This may occur following a refueling in which new fuel assemblies are loaded, fuel assemblies are shuffled within the core, or control rods are replaced or shuffled.

Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod (continued)

CLINTON B 3.1-11 Revision No. 0 core keff core keff

Reactivity Anomalies B 3.1.2 BASES SURVEILLANCE SR 3.1.2.1 (continued)

REQUIREMENTS or a control rod from another core location. Also, core reactivity changes during the cycle.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the monitored and predicte5^-ed density values can be made. For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement) at >_ 80% RTP have been obtained for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 1000 MWD/T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity.

This comparison requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results.

Therefore, the comparison is only done when in MODE 1.

With regard to core reactivity differences values obtained pursuant to this SR, as determined from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 3).

REFERENCES 1.

10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.

2.

USAR, Chapter 15.

3.

Calculation IP-0-0002.

core keff CLINTON B 3.1-12 Revision No. 4--6