NG-17-0084, License Amendment Request (TSCR-164), Revision to Technical Specification 3.1.2, Reactivity Anomalies

License Amendment Request (TSCR-164), Revision to Technical Specification 3.1.2, Reactivity Anomalies

ML17111A631
Person / Time
Site:	Duane Arnold
Issue date:	04/20/2017
From:	Dean Curtland NextEra Energy Duane Arnold
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NG-17-0084
Download: ML17111A631 (18)

Text

NEXTera..,

EN E RGY~

DUANE ARNOLD April 20, 2017 NG-17-0084 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Duane Arnold Energy Center Docket No. 50-331 Renewed Facility Operating License No. DPR-49 License Amendment Request (TSCR-164), Revision to Technical Specification 3.1.2, Reactivity Anomalies In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), NextEra Energy Duane Arnold, LLC (NextEra) is submitting a request for an amendment to the Technical Specifications (TS) for the Duane Arnold

• Energy Center (DAEC) . The proposed amendment would revise TS 3.1.2, "Reactivity Anomalies," with a change to the method of calculating core reactivity for the purpose of performing the reactivity anomaly surveillance.

The Enclosure to this letter provides NextEra's evaluation of the proposed change.

Attachment 1 to the enclosure provides a markup of the TS showing the proposed changes, and Attachment 2 provides the proposed TS Bases changes. The changes to the TS Bases are provided for information only and will be incorporated in accordance with the TS Bases Control Program upon implementation of the approved amendment.

NextEra requests approval of the proposed license amendment by May 1, 2018, and implementation within 90 days.

In accordance with 10 CFR 50.91, a copy of this application , with enclosures ; is being provided to the designated State of Iowa official.

As discussed in the Enclosure , the proposed change does not involve a significant hazards consideration pursuant to 10 CFR 50.92, and there are no significant environmental impacts associated with the change. The Duane Arnold Energy Center Onsite Review Group has reviewed the proposed license amendment request.

This letter contains no new or revised regulatory commitments.

NextEra En ergy Duane Arn old , LLC, 32 77 DAEC Road, Palo , IA 52324

Document Control Desk NG-17-0084 Page 2 of 2 If you have any questions or require additional information, please contact Michael Davis, Licensing Manager, at 319-851-7032.

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

Executed on April 20, 2017.

)µa~

Dean Curtland p,,r Ot.,c c.,,u,...t Site Director, Duane Arnold Energy Center

• NextEra Energy Duane Arnold, LLC Enclosure cc: Regional Administrator, USNRC, Region Ill, Project Manager, USNRC, Duane Arnold Energy Center Resident Inspector, USNRC, Duane Arnold Energy Center A. Leek (State of Iowa)

ENCLOSURE 1 to NG-17-0084 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER License Amendment Request (TSCR-164), Revision to Technical Specification 3.1.2, Reactivity Anomalies EVALUATION OF PROPOSED CHANGE 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

4.4 CONCLUSION

S

5.0 ENVIRONMENTAL CONSIDERATION

S

6.0 REFERENCES

ATTACHMENT 1 - Markup of Technical Specifications ATTACHMENT 2 - Markup of Technical Specification Bases 6 pages follow

1.0

SUMMARY

DESCRIPTION NextEra Energy Duane Arnold, LLC (NextEra) hereby requests an amendment to the Duane Arnold Energy Center (DAEC) Technical Specifications (TS) to modify TS 3.1.2, "Reactivity Anomalies." The proposed change would allow performance of the reactivity anomaly surveillance on a comparison of monitored to predicted core reactivity. The reactivity anomaly verification is currently determined by a comparison of monitored vs. predicted control rod density.

2.0 DETAILED DESCRIPTION The proposed change revises TS 3.1.2, "Reactivity Anomalies", as shown below.

LCO 3.1.2 The reactivity difference between the monitored rod density core kett and the predicted rod density core kett shall be within +/- 1% .Llk/k.

SR3.1.2.1 Verify core reactivity difference between the monitored rod density core kett and the predicted rod density core kett is within +/- 1% .Llk/k.

The purpose of the reactivity anomaly surveillance is to compare the observed reactivity behavior of the core (at hot operating conditions) with the predicted reactivity behavior.

Currently, DAEC TS 3.1.2 requires that the surveillance be done by comparing the monitored control rod density to the predicted control rod density, calculated prior to the start of operation for a particular cycle. The proposed amendment will change the method by which the reactivity anomaly surveillance is performed but not the specified frequency for performing the surveillance.

Current TS 3.1.2 requires that the reactivity equivalence of the difference between the monitored rod density and the predicted rod density shall not exceed +/-1 % Llk/k. The proposed amendment would revise TS 3.1.2 to state that the reactivity difference between the monitored core kettective (kett) and the predicted core kett shall not exceed +/-1 % Llk/k.

The current method of performing the reactivity anomaly surveillance uses rod density for the comparison primarily because early core monitoring systems did not calculate core critical kett values for comparison to design values. Instead, rod density was used as a convenient representation of core reactivity. Allowing the use of a direct comparison of core 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.

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 to 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 core kett at rated conditions and the companion rod patterns derived using those predicted values of kett* A calculation is made of Page 1 of 6

the number of notches inserted in the rod patterns, and also the number of equivalent notches required to make a change of +/-1 % reactivity around the predicted core keff* The rod density is converted to notches and plotted with an upper and lower bound representing the +/-1 % reactivity 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 indirect to critical core keff with a translation of acceptance criteria to rod density.

The revised method for evaluating a potential reactivity anomaly proposed in this amendment compares monitored core keff to predicted core keff* Monitored core keff is calculated by the 30 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 30 core simulator keff behavior in previous cycles and any 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 (inserted 1/3 of the way into the core or less) have very little impact on reactivity while deeply inserted central control rods have a larger effect.

Thus, a potential exists for reactivity anomaly concerns to arise during operations simply because of greater than anticipated use of near-edge and shallow control rods when in fact no true anomaly exists. Use of monitored to predicted core keff instead of rod density eliminates the limitations described above, provides for a technically superior comparison, and is a very simple and straightforward approach.

The proposed change 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 a technically superior comparison as discussed above. Furthermore, the reactivity anomaly surveillance will continue to be performed at the current required frequency.

Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified, and no margins of safety will be reduced.

Regarding the core monitoring system, OAEC recently transitioned from the 30 MONICORE to the ACUMEN system, both from Global Nuclear Fuel (GNF). Reference 1 notes NRC acceptance of 30 MONICORE core surveillance system power distribution uncertainties, whereas Reference 2 notes acceptance for transition to ACUMEN. The latest version of the 30 MONICORE and ACUMEN systems incorporates the PANACEA Version 11 (PANAC11) core simulator code to calculate parameters such as core nodal powers, fuel thermal limits, etc.,

using actual, measured plant input data. Reference 3 notes NRC acceptance of PANAC11.

PANAC11 is the same 30 core simulator code used in core design and licensing activities for OAEC. When a 30 MONICORE or ACUMEN core monitoring case is run, the core keff (as computed by PANAC11) is also calculated and printed directly on each 30 MONICORE or ACUMEN case output. This monitored value can then be directly compared to the predicted value of core keff as a measure of reactivity anomaly. No plant hardware or operational changes are required with this proposed change.

Page 2 of 6

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 surveillance required by the DAEC TS 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 The NRC has approved similar license amendments for the plants below that changed the method of performing the reactivity anomaly surveillance to use a comparison of monitored to predicted core kett*

• Peach Bottom - Amendments 284 and 287 [Reference 4]

• Hatch - Amendments 207 and 263 [Reference 5]

• Limerick - Amendments 168 and 207 [Reference 6]

• Brunswick - Amendments 187 and 218 [Reference 7]

In addition to the above amendments, reactivity anomaly surveillance requirement 3.1.2.1 in NUREG-1434, Standard Technical Specifications - General Electric BWR/6 Plants, Revision.

4.0, is written with the core kett comparison, as opposed to the control rod density comparison.

There are multiple BWRs that already use the core kett as the basis for the reactivity anomaly comparison. Reference 8 notes that the Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, and Quad Cities Nuclear Power Station, Units 1 and 2 TS use the core kett comparison.

4.3 No Significant Hazards Consideration The Duane Arnold Technical Specifications (TS) currently require performing the reactivity anomaly surveillance by comparing the monitored control rod density to the predicted control rod density calculated prior to the start of operation for a particular cycle. The proposed amendment will change the method by which the reactivity anomaly surveillance is performed by comparing monitored to predicted core reactivity.

As required by 10 CFR 50.91 (a), NextEra has evaluated the proposed change to the Duane Arnold TS using the criteria in 10 CFR 50.92 and determined that the proposed change does 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 3of6

The proposed change does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. The proposed change would only modify how the reactivity anomaly surveillance 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 TS requirements such that, rather than performing the surveillance by comparing monitored to predicted control rod density, the surveillance is performed by a direct comparison of core kett* Present day on-line core monitoring systems, such as 3D MONICORE and ACUMEN, are capable of performing the direct measurement of reactivity.

Therefore, since the reactivity anomaly surveillance will continue to be performed by a viable method, the proposed change does not involve a significant increase in the probability or consequence of a 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 The proposed change does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems. All systems important to safety will continue to be operated and maintained within their design bases. The proposed changes to the Reactivity Anomalies TS will only provide a new, more efficient method of detecting an unexpected change in core reactivity.

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

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

Response: No The proposed change is to modify the method for performing the reactivity anomaly surveillance from a comparison of monitored to predicted control rod density to a comparison of monitored to predicted core kett* The direct comparison of kett provides a technically superior method of calculating any differences in the expected core reactivity.

The reactivity anomaly surveillance will continue to be performed at the same frequency as is currently required by the TS, only the method of performing the surveillance will be changed. Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified. The proposed change has no impact to the margin of safety.

Based on the above, NextEra concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.

Page 4 of 6

4.4 Conclusions 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

S 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 types or a significant increase in the amounts of any effluents 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 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. Letter from Frank Akstulewicz (NRC) to G. A. Watford (General Electric Company),

"Acceptance for Referencing of Licensing Topical Reports NEDC-32601 P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-24011-P-A on Cycle-Specific Safety Limit MCPR (TAC Nos. M97490, M99069, and M97491,"

March 11, 1999.

2. Letter from Kevin Hsueh (NRC) to Jerald G. Head (GE-Hitachi Nuclear Energy Americas), "Final Safety Evaluation for Amendment 42 to Global Nuclear Fuel -

Americas Topical Report NEDE-24011-P-A-US General Electric Standard Application for Reactor Fuel (GESTAR II) Supporting the Transition from the 30-MONICORE Core Monitoring System to ACUMEN (CAC No. MF7438)," August 31, 2016.

3. Letter from Stuart A. Richards (NRC) to G. A. Watford (GE Nuclear Energy),

"Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, "GESTAR II" -

Implementing Improved GE Steady-State Methods (TAC No. MA6481)," November 10, 1999.

4. Letter from Richard B. Ennis (U.S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Nuclear), "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments RE: Reactivity Anomalies Surveillance (TAC Nos. ME6356 and ME6357),"

dated May 25, 2012.

5. Letter from Robert E. Martin (U.S. Nuclear Regulatory Commission) to M. J. Ajluni (Southern Nuclear Operating Company), "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.

Page 5 of 6

6. Letter from Peter Bamford (U.S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Nuclear), "Limerick Generating Station, Units 1 and 2 - Issuance of Amendments RE: Reactivity Anomalies Surveillance (TAC Nos. ME6348 and ME6349),"

dated March 14, 2012.

7. Letter from David C. Trimble (U.S. Nuclear Regulatory Commission) to C. S. Hinnant (Carolina Power & Light Company), "Issuance of Amendment No. 187 to Facility Operating License No. DPR-71 and Amendment No. 218 to Facility Operating License No. DPR-62 Regarding a Change in the Methodology for Detecting a Reactivity Anomaly

- Brunswick Steam Electric Plant, Units 1 and 2 (TAC Nos. M97688 and M97689),"

dated September 5, 1997.

8. Letter from M. D. Jesse (Exelon Generation Company) to U.S. Nuclear Regulatory Commission, "License Amendment Request - Reactivity Anomalies Surveillance," dated June 2, 2011.

Page 6 of 6

ATTACHMENT 1 MARKUP OF TECHNICAL SPECIFICATIONS 2 pages follow

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

~core keff *

• APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Core reactivity A.1 Restore core reactivity 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> difference not within difference to within limit.

limit.

8. Required Action and 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 /> associated Completion Time not met.

DAEC 3.1-5 Amendment ~

Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between the Once within monitored rod density and the predicted f0G 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after density is within +/- ~k. reaching

~

• core keff equilibrium

~core keff conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 1000 MWD/T thereafter during operations in MODE1 DAEC 3.1-6 Amendment ~

ATTACHMENT 2 MARKUP OF TECHNICAL SPECIFICATIONS BASES 5 pages follow

Reactivity Anomalies B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Anomalies BASES BACKGROUND In accordance with the UFSAR (Ref. 1), reactivity shall be controllable such that subcriticality is maintained under cold conditions and acceptable fuel design limits are not exceeded during normal operation and abnormal operational transients.

Therefore, reactivity anomaly is used as a measure of the predicted versus measur ore reactivity during power operation.

The continual con

• tion of core reactivity is necessary to ensure tha esign Basis Accident (OBA) and transient safety ana s remain valid. A large reactivity anomaly could be the (i.e., monitored) esult of unanticipated changes in fuel reactivity or control rod worth or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SOM or violation of acceptable fuel design limits.

Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SOM demonstrations (LCO 3.1 .1, "SHUTDOWN MARGIN (SOM)") in assuring the reactor can be brought safely to cold, subcritical conditions.

When the reactor core is critical or in normal power operation , a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers, producing zero net reactivity.

In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive reactivity beyond that required to sustain steady state operation at the beginning of cycle (BOC) . When the reactor is critical at RTP and operating moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, and whatever neutron poisons (mainly xenon and samarium) that are (continued)

DAEC B 3.1-9 Amendment ~

Reactivity Anomalies B 3.1.2 core keff BASES BACKGROUND (continued)

APPLICABLE ccurate prediction of core reactivity is either an explicit or implicit SAFETY assumption in the accident analysis evaluations (Ref. 2). In ANALYSES particular, SOM 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 The monitored core keff is evaluations rely on computer codes that have been qualified calculated by the core against available test data, operating plant data, and analytical monitoring system at benchmarks. Monitoring reactivity anomaly provides additional actual plant conditions assurance that the nuclear methods provide an accurate and is compared to the representation of the core reactivity.

predicted value at the same cycle exposure. 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 predicted f0G core keff ~ density for identical core conditions at BOC do not reasonably agree, then the assumptions used in the reload cycle de~core keff analysis or the calculation models used to predict rod de

• may not be accurate. If reasonable 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 measured

• from the predicted f0G core keff ~ density that develop during fuel pletion may be an indication that the assumptions of the DB and transient analyses are no longer valid, or that an unexpe ted change in core conditions has occurred.

core keff Reactivity anomalies satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

(continued)

DAEC B3.1-10 Amendment 223

Reactivity Anomalies B 3.1.2 BASES (continued)

LCO 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 OBA 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 monitored and the predicted core keff ---7 rod density of+/- 1% b.k/k has been established based on engineering judgment. 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. SOM requirements (LCO 3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and an SOM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, shuffling). The SOM 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 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 (continued)

DAEC B3.1-11 Amendment 223

Reactivity Anomalies B 3.1.2 BASES ACTIONS A.1 (continued) reviews the core 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 OBA occurring 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.

If the core reactivity cannot be restored to within the 1% ilk/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 and predicted

• is within the limits of the LCO provides core keff - -------.:*........ assurance that plant operation is maintained within the assumptions of the OBA and transient analyses. The Plant Process Computer calculates the

• for the reactor conditions obt

• ant instrumentation. A comparison of core keff - - -trn:r-rrrorn ored

• to the predicted

• at the same cycle e e *s used to calculate the reactivity di e . The parison is required when the core reactivity has potential y core keff core keff 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 (continued)

DAEC B3.1-12 Amendment 223

Reactivity Anomalies B 3.1.2 BASES SURVEILLANCE SR 3.1.2.1 (continued)

REQUIREMENTS subsequent replacement with a new control rod 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 predicted can be made. For the purposes of this SR, the reactor is ass ed to be at equilibrium conditions when steady state operations (no core keff control rod movement or core flow changes) at z 75% RTP have been obtained . Additionally, the Reactor Engineer or individual fulfilling this role will normally be involved with determining equilibrium xenon conditions. 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.

REFERENCES 1. UFSAR, Sections 3.1 .2.3.7, 3.1 .2.3.9, and 3.1.2.3.10.

2. UFSAR, Chapter 15.

DAEC B 3.1-13 TSCR-444A