ML20044C379
| ML20044C379 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 03/17/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044C375 | List: |
| References | |
| NUDOCS 9303220387 | |
| Download: ML20044C379 (3) | |
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I'o UNITED STATES (g
NUCLEAR REGULATORY COMMISSION -
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WASHINGTON, D. C. 20555
'gQ. 5 gv.....f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT NFSR-0087-CECO PWR TRANSIENT ANALYSIS METHODOLOGY TOPICAL SUPPLEMENT COMMONWEALTH EDISON COMPANY ZION NUCLEAR POWER STATION. UNITS 1 AND 2
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DOCKET NOS. 50-295 AND 50-304
1.0 INTRODUCTION
By letter dated March 4,1991, Commonwealth Edison Company (CECO) submitted the supplement NFSR-0087, Revision 0, to Topical Report NFSR-0069, Revision-0.
NFSR-0069, entitled, " Transient Analysis Envelope for Zion Units 1 and 2,"
dated November 8,1989, reflects use of_ the Westinghouse Standard Thermal Design Procedure (STDP) for the analyses of departure:from nucleate boiling (DNB) limiting transients. The supplemental report presents sample calculations performed to demonstrate the applicability to' Zion of an upgraded methodology known as the Westinghouse Revised Thermal Design Procedure (RTDP).
This methodology will replace the STDP for use in;the analyses of DNB ' limiting.
transients and was approved by the staff in Topical Report.WCAP-11397-A dated-April 1989.
NFSR-0069 is currently under review by the staff and, until it is approved, supplement NFSR-0087 cannot be referenced for licensing analyses.
Commonwealth Edison Company plans to employ-'the upgraded methodology to increase calculated DNB margins for the first two reloads (in Cycle 13) of Westinghouse VANTAGE 5 fuel at Zion, Units 1_and 2.
Additionally, Ceco intends to implement the RTDP in future safety analyses for. all Ceco. PWRs using VANTAGE 5 and Westinghouse Optimized Fuel Assembly (OFA) fuels.
The RTDP statistically combines the uncertainties associated with plant operating parameters, nuclear and thermal parameters, fuel-fabrication-parameters, and DNB correlations to obtain an overall departureJ from nucleate boiling ratio.(DNBR) uncertainty factor. This factor;is then employed in computing:the Design Limit DNBR (DLDNBR).
In contrast, the STDP methodology statistically combin'es system parameter uncertainties separately from the DNB correl ation-uncertainty. The two are then combined directly, rather than statistically, resulting in an overconservative (i.e., higher).0LDNBR.
Because the DLDNBR already accounts. for:the above uncertainties, the transient:
analyses performed to determine minimum DNBR (and thus margin) for DNB limiting events can be performed using nominal or best estimate values of-system parameters. Application of the RTDP to Zion is consistent with-the Westinghouse application to a' representative plant as described in
' WCAP-11397-A. - However, the VIPRE and RETRAN thermal-hydraulic codes -have been employed in place of the Westinghouse THINC-IV and LOFTRAN codes-for DNBR computations.
9303220387 930317 yDR ADOCK 05000295 i
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L As a separate matter, NFSR-0087 additionally demonstrates applicability to-Zion of an NRC-approved Westinghouse methodology for calculating Reactor Protection System overtemperature delta-T (OT-delta T) and overpower delta-T (OP-delta T) safety analysis setpoints. This methodology is documented in WCAP 8745-P-A, " Design Bases for the Thermal Overpower Delta-T and Thermal Overtempcrature Delta-T Trip Functions," dated September 1986.
2.0 EVALUATION Computation of the DNBR uncertainty factor (and thus the DLDNBR) entails determination of the sensitivity of DNBR to key operating and fuel design parameters.
To determine these sensitivities, the licensee first defined various base cases to conservatively bound the range of possible operating conditions. The VIPRE code was then employed to perfonn the sensitivity and base case runs. The uncertainties assumed in the calculations are preliminary values for Zion, used only to demonstrate the computational methodology.
Finalized uncertainty values will be used for future li:.ensing analyses.
The computations resulted in DLDNBR values of 1.26 and 1.25 for typical cells and thimble cells, respectively. A 12 percent margin was then added to these-values to obtain the Safety Analysis Limit DNBR (SALDNBR) of 1.43 and 1.42, respectively. Although NFSR-0087 does not present comparative values of DLDNBR generated by the RTDP and STDP methodologies, WCAP-11397-A does make such a comparison for a representative plant.
In that case, values of 1.24 and 1.23 were found for typical and thimble cells, respectively, using RTDP; corresponding values of 1.35 and 1.34 were determined using STDP.
L For purposes of demonstrating applicability of the RTDP in DNB transient analyses, sample calculations were performed for the two most limiting DNB transients -- the Locked Rotor event and Loss of Reactor Coolant Flow event.
The minimum DNBR (MDNBR) computed for these cases was 1.73 and-1.84, respectively. Clearly, a margin to the SALDNBR is maintained.
To demonstrate applicability of the approved Westinghouse methodology for calculating OT-delta T and OP-delta T setpoints, sample calculations were i
performed to determine core thermal limit lines, the setpoint equation constants, and the f(delta I) reset function for the 0T-delta T trip. Core exit quality lines, DNB limit lines, and axial offset DNB envelopes were
.t computed using the VIPRE code. When the finalized setpoint equations. are j
generated for Zion and used in licensing analyses, the setpoint constants will need to be verified to demonstrate DNB margin through analysis of the RCCA Withdrawal at Power and the Loss of Load transients, i
It should be noted that the sample calculations presented in NFSR-0087_ do not include Intermediate Flow Mixing (IFM) grid inputs since the first two reloads of VANTAGE 5 fuel will not contain these grids.
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3.0 CONCLUSION
The staff has reviewed the Topical Report NFSR-0087, Revision 0, submitted by the licensee to demonstrate applicability of the Westinghouse methodologies discussed above to Zion, and has the following findings:
1.
The licensee has applied methodologies previously approved by the staff and in a manner consistent with the documented Westinghouse applications.
2.
The VIPRE thermal-hydraulic code employed in the sample calculations has been previously approved by the staff for all pressurized water reactor-(PWR) core thermal-hydraulic analyses except loss of coolant accidents (LOCA).
3.
The thermal-hydraulic code uncertainties assumed in the RTDP sample calculations are the same as those used in approved Westinghouse methodology.
However, since different codes were employed (VIPRE and RETRAN vs. THINC-IV and LOFTRAN), uncertainty values specific to these codes will need to be developed for use in licensing analyses.
4.
Nominal conditions used in the sample calculations bound all permitted ranges of plant operation.
5.
The DNB correlation statistics used in the sample calculations have been i
previously approved by the staff.
6.
Sensitivity factors have been calculated over a range of operating conditions that bound all possible steady-state and transient conditions.
7.
The preliminary values of uncertainties in operating and fuel design parameters assumed in the sample calculations appear realistic.
On the bases of the above, the staff finds that the licensee has acceptably demonstrated application of the Westinghouse RTDP methodology as well' as the methodology for cafculating OT-delta T and OP-delta T setpoints for use at Zion.
Topical Report NFSR-0087, Revision 0, may, therefore, be referenced in future licensing analyses, pending staff approval of NFSR-0069.
It is-understood, however, that since the uncertainties in the thermal-hydraulic codes and the operating and fuel design parameters assumed in the sample calculations are preliminary values, finalized values will need to be developed and appropriately incorporated. Additionally, IFM grid inputs will-need to be incorporated into analyses for future reloads'that contain these
. grids.
Principal Contributor:
H. Abelson l
Dated:
March 17,1993 1
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