JAFP-18-0093, License Amendment Request - Reactivity Anomalies Surveillance

License Amendment Request - Reactivity Anomalies Surveillance

ML18275A060
Person / Time
Site:	FitzPatrick
Issue date:	10/02/2018
From:	David Gudger Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-18-0093
Download: ML18275A060 (15)

Text

200 Exelon Way Exelon Generation Kennett Square. PA 19348 www.exeloncorp com 10 CFR 50.90 JAFP-18-0093 October 2, 2018 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 James A. FitzPatrick Nuclear Power Plant, Unit 1 Renewed Facility Operating License No. DPR-59 NRG Docket No. 50-333

Subject:

License Amendment Request - Reactivity Anomalies Surveillance In accordance with 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) requests a proposed change to modify the Technical Specifications (TSs) concerning a change to the method of calculating core reactivity for the purpose of performing the reactivity anomaly surveillance at the James A. FitzPatrick Nuclear Power Plant (JAFNPP), Unit 1.

The proposed changes have been reviewed by the JAFNPP Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the Exelon Quality Assurance Program.

Exelon requests approval of the proposed amendment by October 2, 2019. Once approved, this amendment shall be implemented within 90 days of issuance. Additionally, there are no commitments contained within this letter.

Attachment 1 contains the evaluation of the proposed changes. Attachment 2 provides the marked-up TS and Bases pages. The Bases pages are being provided for information only.

In accordance with 10 CFR 50.91, Exelon is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Officials.

Should you have any questions concerning this letter, please contact Christian Williams at (610) 765-5729.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2nd of October 2018 ..

Respe tfully, J

~ *J f. :A--J~

Davi T. Gudger Manager, Licensing & Regulatory Affairs Exelon Generation Company, LLC

License Amendment Request Reactivity Anomalies Surveillance October 2, 2018 Page2 : Evaluation of Proposed Changes : Markup of Technical Specifications and Bases Pages cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, JAFNPP USNRC Project Manager, JAFNPP A. L. Peterson, NYSERDA

A TTACHMEMT 1 License Amendment Request James A. FitzPatrick Nuclear Power Plant, Unit 1 Docket No. 50-333 Evaluation of Proposed Changes

SUBJECT:

Reactivity Anomalies Surveillance 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

License Amendment Request Attachment 1 Reactivity Anomalies Surveillance Page 1 Docket No. 50-333 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Renewed Facility Operating Licenses DPR-59 for the James A. FitzPatrick Nuclear Power Plant (JAFNPP), Unit 1.

The proposed change would revise the Technical Specifications to allow performance of the surveillance on a comparison of predicted to actual (or 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 predicted reactivity behavior.

Currently, the JAFNPP Technical Specifications (TSs) require that the surveillance be done by comparing predicted control rod density (calculated prior to the start of operation for a particular cycle) to an actual control rod density. The comparison is done, as required by the surveillance requirements. The proposed revision will change the method by which the reactivity anomaly surveillance is performed and not the specified frequency for performing the surveillance.

The current TS requires that the reactivity equivalence of the difference between the actual rod density and the predicted rod density shall not exceed +/-1 % ~k/k.

The proposed TS and Bases would be revised to state that the reactivity difference between the actual keffective (keff) and the predicted keff shall not exceed +/-1% ~k/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 keff 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 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.

Marked up TS Bases pages are provided in Attachment 2 and are provided for information only.

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

License Amendment Request Attachment 1 Reactivity Anomalies Surveillance Page2 Docket No. 50-333 Evaluation of Proposed Changes TS surveillance actually begins with the predicted kett at rated conditions and the companion rod patterns derived using those predicted values of kett. A calculation is made of 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 kett. 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 kett with a "translation" of acceptance criteria to rod density.

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, it is not uncommon for reactivity anomaly concerns to arise during operations simply because of greater use of near-edge and shallow control rods than anticipated, when in fact no true anomaly exists. Use of actual to predicted kett instead of rod density eliminates the limitations described above, provides for a technically superior comparison, and is a very simple and straightforward approach.

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 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, JAFNPP utilizes the Global Nuclear Fuel (GNF) 3D MONICORE core monitoring software system . Reference 1 notes NRC acceptance of 3D MON ICORE core surveillance system power distribution uncertainties. The latest version of this product 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 2 notes NRC acceptance of PANAC11 . PANAC11 is the same 3D core simulator code used in core design and licensing activities. When a 3D MONICORE core monitoring case is run, the core kett (as computed by PANAC11) is also calculated and printed directly on each 3D MONICORE case output. This value can then be directly compared to the predicted value of kett as a measure of reactivity anomaly.

No plant hardware or operational changes are required with this proposed change.

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 JAFNPP, Unit 1 Technical Specifications serves to partly

License Amendment Request Attachment 1 Reactivity Anomalies Surveillance Page 3 Docket No. 50-333 Evaluation of Proposed Changes 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 Precedents 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.

• Duane Arnold - Amendment 304 (Reference 3)

• Peach Bottom - Amendments 284 and 287 (Reference 4)

• Limerick - Amendments 168 and 207 (Reference 5)

• Hatch - Amendments 207 and 263 (Reference 6)

In addition to the above amendments, the reactivity anomaly Limiting Condition for Operation (LCO) in the BWR/6 Standard Technical Specifications, NUREG-1434, Rev. 4.0, is written with the kett comparison, as opposed to the control rod density comparison.

Currently, 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 keff comparison .

4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) 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:

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

Response: No.

The proposed Technical Specification 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 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 Technical Specification requirements such that, rather than performing the surveillance by comparing predicted to actual control rod density, the surveillance is performed by a direct comparison of kett. Present day on-line core monitoring systems, such as the one in use at the James A. FitzPatrick Nuclear Power Plant (JAFNPP), Unit 1 are capable of performing the direct measurement of reactivity.

License Amendment Request Attachment 1 Reactivity Anomalies Surveillance Page4 Docket No. 50-333 Evaluation of Proposed Changes Therefore, since the reactivity anomaly surveillance will continue to be performed by a viable method, the proposed amendment does not involve a significant increase in the probability or consequence of a previously evaluated accident.

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

Response: No.

This Technical Specifications amendment request 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 anomaly Technical Specifications will only provide a new, more 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 and the possibility of a new or different kind of accident is not created.

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

Response: No.

This proposed Technical Specifications 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 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 Technical Specifications, only the method of performing the surveillance will be changed. Consequently, core reactivity assumptions made in safety analyses will continue to be adequately verified.

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

Based on the above, Exelon Generation Company, LLC, concludes that the proposed amendment does not involve a 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.

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

License Amendment Request Attachment 1 Reactivity Anomalies Surveillance Page 5 Docket No. 50-333 Evaluation of Proposed Changes 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 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 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 [provides NRC acceptance of 3D MONICORE core surveillance system power distribution uncertainties)

2) MFN-035-99, S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, "GESTAR 11"-lmplementing Improved GE Steady-State Methods (TAC No. MA6481)," November 10, 1999 [provides NRC acceptance of PANACEA Version 11)

3) Letter from M. L. Chawla (U.S. Nuclear Regulatory Commission) to Mr. Dean Curtland (NextEra) "Duane Arnold Energy Center - Issuance of License Amendment No. 304 RE: Revision to Technical Specification 3.1 .2, Reactivity Anomalies (EPID L-2017-LLA-0218)" dated March 9, 2018.

4) Letter from Richard B. Ennis (U.S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Nuclear) "Peach Bottom Atomic Power Station, Units 1 and 2 - Issuance of Amendments RE: Revision to Technical Specification 3.1.2, Reactivity Anomalies Surveillance (TAC Nos. ME6356 and ME6357" dated May 25, 2012.

5) Letter from Peter Bamford (U.S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Nuclear) "Limerick Generating Station, Units 1 and 2 - Issuance of License Amendments RE: Reactivity Anomalies Surveillance (TAC Nos. ME6348 and ME6349" dated March 14, 2012.

6) 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 Specification Limiting Conditions for Operation 3.1.2, "Reactivity Anomalies" (TAC Nos ME3006 and ME3007)"

dated November 4, 2010.

ATTACHMENT 2 James A. FitzPatrick Nuclear Power Plant (JAFNPP), Unit 1 Markup of Technical Specifications and Bases Pages Revised Pages (Unit 1) 3.1.2-1 3.1.2-2 B 3.1.2-1 B 3.1.2-2 B 3.1.2-3 B 3.1.2-4

Reactivity Anomalies 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Reactivity Anomalies LCO 3.1.2 The reactivity difference between the measured rod densitycore kett and the predicted FOd densitycore kett shall be within .+/- 1% t1k/k.

APPLICABILITY: MODES 1 and 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Core reactivity A.1 Restore core 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> difference not within reactivity difference limit. to within limit.

B. Required Action and B.1 Bein 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.

JAFNPP 3.1.2-1 Amendmentm

Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between Once within the measured rod densitycore ken and the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after predicted rod densitycore ketr is within :t 1% reaching

~k/k. equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 1000 MWD/T thereafter during operations in MODE 1 JAFNPP 3.1.2-2 Amendmentm

Reactivity Anomalies B3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Anomalies BASES BACKGROUND In accordance with the Updated Final Safety Evaluation Report (UFSAR) Section 16.6 (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 Anomalies are used as a measure of the predicted versus measured (i.e., monitored) core reactivity during power operation. The continual confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA) and transient safety analyses remain valid.

A large reactivity anomaly could be the result 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 SDM 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 SDM requirements (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") 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 (continued)

JAFNPP B 3.1.2-1 Revision Q

Reactivity Anomalies B 3.1.2 BASES BACKGROUND poisons (mainly xenon and samarium) are present in the fuel.

(continued) The predicted core reactivity, as represented by control rod Densitycore kP.1t, is calculated by the 3D Monicore System as a function of cycle exposure. This calculation is performed for projected operating states and conditions throughout the cycle. The core reaetivity is deterFAined froA'l control rod densities for act1:1al plant eonditions and is then eoFApared to the predicted value for the cycle exposure.The monitored katr is calculated by the core monitoring system at actual plant conditions and is compared to the predicted value at the same cycle exposure.

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

Measuring 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 predicted rod densitycore k,;tt for identical core conditions at BOC do not reasonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used to predict rod densitycore kett 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 rod densitycore keff from the predicted rod densitycore ketr that 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 10 CFR 50.36(c)(2)(ii) (Ref. 3).

LCO The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses. Large differences between measured and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the (continued)

JAFNPP B 3.1.2-2 Revision G

Reactivity Anomalies B 3.1.2 BASES LCO uncertainties in the "Nuclear Design Methodology" are larger (continued) than expected. A limit on the difference between the measured and the predicted rod deAsitycore k-;tt of+/- 1% ~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 measured 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 measuring 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, shuffling). The SDM test, required by LCO 3.1.1, provides a direct comparison of the predicted and measured core reactivity at cold conditions: therefore, the Reactivity Anomalies Specification 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 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 DBA 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.

(continued JAFNPP B 3.1.2-3 Revision G

Reactivity Anomalies B 3.1.2 BASES ACTIONS (continued)

If the core reactivity cannot be restored to within the 1% ~k/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 measured and predicted rod deAsitycore keff is within the limits of the LCO provides added assurance that plant operation is maintained within the assumptions of the DBA and transient analyses.

The 3D Monicore System calculates the rod deAsitycore k~tt for the reactor conditions obtained from plant instrumentation. A comparison of the measured rod deAsitycore kerf to the predicted roe deASitycore kett at the same cycle exposure is used to calculate 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 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 measured and predicted roe EieAsitycore kett 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 or core flow changes) at 75% RTP have been obtained. 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. The tests performed at this Frequency also use base data obtained during the first test of the specific cycle.

(continued)

JAFNPP B 3.1.2-4 Revision 0