ML20028D399
| ML20028D399 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/06/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20028D396 | List: |
| References | |
| NUDOCS 8301190070 | |
| Download: ML20028D399 (6) | |
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555 f.....o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 72 TO FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 1.0 Introduction In a letter dated February 20, 1981, the Power Authority of the State of New York (the licensee) requested an amendment to Appendix A of Facility Operating License No.
DPR-59 for the James A. FitzPatrick Nuclear Power Plant (Reference 1).
The proposed changes were supplemented by letters' of November 18, 1981 (Reference 4) and February 19,1982 (Reference 5).
The amendment would' allow use of redefined rated recirculation flow which corresponds to the core flow used in the analysis of safety limits.
Also the changes would permit the licensee to operate within an extended power /
flow operating envelol e bounded by a new operating line (References 2 and 3) i above the present Average Power Range Monitor (APRM) rod blockline.'
However, any rod movement would still be limited to the present APRM rod blockline.
This amendment would provide more operating flexibility during power ascension and reduction operations while maintaining the plant with-in design basis and previously established safety limits.
2.0 Discussion 2.1 Refinition of Recirculation Flow e
The present ' Technical Specifications (TS) define rated recirculation flow as 34.2 x 10 1,b/hr.
However, the safety analyses of transients and accidents employ rated recirculation flow defined as that value which pro-duces a core flow of 77.0 x 10' lb/hr.
Consequently, the present TS de-finition of 34.2 x 10' lb/hr rated recirculation flow is more conservative than that employed in the safety analysis, since the recirculation' flow rate, as currently defined, 'is greater than that necessary to produce a rated core flow of 77.0 x 10 lb/hr. Therefore, the 1.icensee has proposed to redefine recirculation flow as that recirculation flow which produces a rated core flow of 77.0 x 10' lb/hr.
The licensee would use the conser-vative recirculation flow value of 34.2 x 10'lb/hr at the beginning of the operating cycle, and then adjust the value based on actual measurement-as appropriate during the cycle.
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2.2 Extended Power / Flow Operating Line Operating flexibility during power ascension in proceeding from a low power /
low core-flow condition to a high power /high' core-flow condition is affected.
by many factors, including the power / flow ratios defined.by the power / flow curve.
The power / flow curve is the locus of power from a fixed rod pattern as a function of flow from which the occurrence of postulated transients will yield results within analyzed and acceptable limits.
Operation of the FitzPatrick Plant utilizing the power flow curves is described in Reference 6.
The restrictions imposed by the power / flow operator enve16pe assure acceptable pressure and thermal margins during postulated transients.
This requires an analysis of abnormal operating transients with degraded scram reactivity characteristics which are dependent on fuel exposure.
The licensee has proposed to extend the upper bound of the power / flow operating envelope; as follows:
the proposed power / flow line would follow the new APRM Rod Blockline (0.58W + 50%) up to an intercept point of 85%
power /61% flow, and then a linear path to 100% power /94% flow (100/94),
followed by constant 100% power to the 100/100 point.
As referenced to i.n the licensee's analysis, the ' hew APRM Rod Blockline" (0.58W + 50%) defines the new upper bound of the operating region and the rod block intercept p61nt (85/61).
However, the actual APRM Rod Blockline remains the same as in the present TS.
The new upper bound and present APRM rod blocks are defined as follows:
y, For the new upper bound of the power / flow operating region and rod block intercept point, S = 0.58W + 50%
For the APRM Rod Blockline,
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S e6 0.66W + 42%
- where, S = Setting in percent of rated power (2435'MWT) i W = Recirculation flow in percent of rated.
l l-The intercept point of 85% power /61% flow establishes the highest power level l
permitted when operating on the new upper bound.
This will be sufficiently high to provi'de the desired operational flexibility during power ascension, but low enough to ensure adequate safety margin from the analysis limit of the 104% power /100% flow point.
Abnormal operational transients have been analyzed for the planned operational conditions up to the thermal power conditions of 104% rated thermal power (2533 hWT).
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As an added conservatism, the actual APRM Rod Blockline is below the proposed new upper bound.
The licensee has performed an evaluation in, support of the proposed change
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which is summarized below:
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1.
The scram reactivit'y insertion characteristics were determined for ende of-cycle (E0C) conditions, at 104/100,, 85/61, and 100/94 power flow points (References 2 and 3).
These values were used in analyzing the most limiting abnornal operational transients: load rejection without bypass, turbine trip without bypass, feedwater controller failure, loss of feedwater heating, rod withdrawal error, and high pressure coolant injection.
Each transient analysis for Reload 3/ Cycle 4 (References 2 and 3) and Reload 4/ Cycle 5 (Reference 5) shows that the limiting transient was the load-rejection-without-bypass at the 104/100 point and that the existing operating limits for minimum critical power ratio (MCPR)'
were applicable to the new power / flow line and are therefore acceptable within the new operating envelope.
Thus, the licen' sing basis values still remain the most limiting values.
2.
Compliance with the ASME pressure vessel code was verified (Reference 2) for all main steamline isolation valve (MSIV) closures with flux scram.
Again, the limiti,ng condition occurs at the 104/100 point with the peak vessel bottom pressure of 1264 psig.
The ASME Boiler and Pressure Vessel code ~1imit is 1375 psig.
3.
A reanalysis of the rod withdrawal accident based on 'the proposed new upper bound was conducted with acceptable results.
The actual rod block during operations will occur with less rod withdrawal (i.e., at 0.66 W +
42%), which is conservative.
4.
ECCS analysis verified the applicability of the extended region for Cycles 3, 4, and 5 (References 2, 3, and 5).
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5.
The Minimum Critical Power Ratio (MCPR) requirements along the proposed new upper. bound (Reference 2) show small increases in the MCPR requirements at lower load conditions during abnormal transients and are therefore accepta bl e.
This trend is consistently demonstrated and the increments are approximately same for 7 x 7, 8 x 8, 8x 8R fuels.
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Thermal hydraulic analysis was performed for the proposed new limiting power / flow line (References 2, 3).
The decay ratios determined from the limiting reactor core stability conditions show that, at the inter-section of natural circulation and extrapolated rod blocklines, the ratios were 0.76, 0.85, and 0.87 for Reload 2/ Cycle 3, Reload 3/ Cycle 4, and Reload 4/ Cycle 5 respectively, well within the bound of the ultimate performance criteria of 1.0.
At the most responsive intersection of natural circulation and extrapolated rod blocklines, the channel performance calculations for Cycle 3 and Cycle 4
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yielded decay ratios of 0.23 and 0.22 respectively for 7 x 7 fuel channels, 0.39 and 0.38 respectively for 8 x 8 fuel channels, and 0.34 and 0.30 re-spectively for 8 x 8R fuel channels.
Reload Core 4 is composed of only 8 x 8, 8 x 8R and P8 x 8R fuel channels, and their respective decay ratios are 0.37, 0.30, and 0.30. Again, they are well below the ultimate per-formance criteria of 1.0.
In addition, operation in the natural circulation l
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mode above 60% power level (as the most responsive conditions) is not a normal mode of operation.
Furthermore, the intersection of extrapolated rod blocklines and the natural circulation line is outside of the operating bound.
The reactor core stability conditions are therefore acceptable.
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3.0 Evaluation i
3.1 Redefinition of Recirculation Flow We have reviewed the information provided by the licensee in support of the proposed modification. We find that the proposed modification is consistent with the c.nalysis e.mployed in the JAFNPP Final Safety Analysis Report.
We also find that the proposed change has been evaluated for its impact on the flow-biased APRM rod block and trip setpoints.
For a nominal rated re-circulation value of 33 x 10' lb/hr which produces' 77 x 10' lb/hr, APRM Flux Scram and RcJ Block Trip Settings are approximately 3% lower than by using the present TS rated flow of 34.2 x 10' lb/hr.
Therefore, the proposed change would be within the bound of the analysis.
In order to maintain an equal margin of safety as given in the previous analysis, the rated recirculation flow would be calibrated conservatively against the rated core flow of 77 x 10' 1b/hr.
In addition, conservatism was incorporated in the transient analysis in estimating the controlling factors such that the void coefficient used was about 25% greater than the nominal maximum value expected during the core lifetime, and analyzed active core flow was 88% of total core flow.
On the basis of the foregoing considerations, the change proposed by the licensee is acceptablem.
3.2 Extended Power / Flow Operating Line We have reviewed the information provided by the licensee in support of this proposed modification.
We find that the licensee has conducted appropriate load-line limitj analyses to verify:
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Reactor Core Stability at the highest power / lowest flow point:
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That the highest power / lowest flow point is a more limiting condition l
than any other condition within the expanded operating envelope; 3.
The impact on Emergency Core Cooling System (ECCS) performance; 4.
Abnormal, operational transients for points along the proposed power /
flow curve; 5.
Scram Reactivity Insertion capabilities.
The licensee has also provided a safety analysis demonstrating that transients and accidents more severe than those analyzed at the 104/100 point will.not occur during operation within or along the proposed power / flow line.
Based on our review, we find that the change proposed by the licensee will allow reactor power ascension to proceed safely alog the modified power / flow line, and is therefore, acceptable.
5 4.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in.any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the stand environmental impact and, purswant to 10 CFR 551.5(d)(4) point of
, that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
...__ J 5.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significent hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
January 6,1982 Principal Contributors:
J. Chung; J. Hegner i
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6 6.0 References 1.
PASNY letter to USNRC dated February 20, 1981, proposed changes to Technical Specifications.
2.
NEDO-24243."_ General Ele.ctric Boiling Wa_t.er Reactor Load Lin.e Limit.
Analysis for James A. FitzPatrick Nuclear Power Plant, Februafy 1980.
3.
NED0-24243, Supplement 1, " General Electric Boiling Water. Reactor Load Line., Limit Analysis for James A. Fitzpatrick Nuclear Power Plant Cycle 4", July 1980.
4.
PASNY letter to USHRC dated November 18', 1981, Changes to the Proposed' Technical Specifications for Reload 4/ Cycle 5.
5.-
PASNY letter to USNRC dated February 19, 1982, Proposed Changes to the Technical Specifications for Reload 4/ Cycle 5 - Revision 1.
6.
Final Safety Analysis Report for the James A. FitzPatrick Nuclear Power Plant; Section 3.7; Docket No. 50-333.
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