ML19350B450
| ML19350B450 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 03/05/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19350B447 | List: |
| References | |
| TAC-07718, TAC-7718, NUDOCS 8103200561 | |
| Download: ML19350B450 (7) | |
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ENCLOSURE TROJAN REVISED ECCS ANALYSIS DOCKET NO. 50-344 LICENSE CHAliGE APPLICATION 60 DATED T4AY 7,1980 Sumary of Report To address the latest fuel rod model concerns and at the same time recover the margin that has been lost in connection with previous Emergency Core Cooling System (ECCS) Evaluation Model changes, and to accomnodate the possibility of plugging of as many as all Row I and Row 2 steam generator tubes, a new ECCS analysis has been performed forTrojan Iising the latest NRC-approved Westinghouse Evaluation flodel (February 1978 version).
This revised ECCS performance evaluation consists of two parts. The first provides a tabular summary of relevant input assumptions in the large LOCA analysis and calculated output parameters.
In addition, plots of key parameters during the course of the LOCA event &re provided. The second part is a quantitative discussion of the impact of incorporating i
NRC fuel rod models into the Westinghouse Evaluation ifodel. This dis-l cussion responds to the concern about fuel rod modeling and is consistent f
with current NRC policy regarding ECCS analyses pending final agreement l
on fuel rod modeling questions.
In summary, the revised ECCS evaluation is consistent with unrestricted l
operation of Trojan with up to 6 percent uniform steam generator tube plugging, with a maximum total peaking factor of 2.32.
Under these l
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l conditions, the calculatedpeak cladding temperature during the worst-case LOCA is 1970 degrees Fahrenheit. The impact of incorporating NRC fuel rod models is demonstrated to be more than compensated by inherent conservatisms in the Westinghouse Evaluation Model.
l 8108200 %
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. Steam generator tube plugging has a detrimental. affect on ECCS perf armance Plugging levels of up to several percent of the steam generator tubes are generally acceptable, however an analysis of ECCS performance must be made to demonstrate that the plugging does not in fact significantly degrade ECCS performance. Based on the Trojan steam generator inspections in late 1979, there is a possibility that additional Row I steam generator tubes may have to be plugged. Therefore, the ECCS evaluation reported in this LCA accounts for an appropriate level of plugged steam generator tubes'.
Background
Demonstration that the Plant's E'CCS meetscertain performance standards is a requirement of 10 CFR 50.46. This requirement has previously been met for Trojan by submitting ECCS performance analyses using NRC-approved Westinghouse ECCS Evaluation Models. Since the original licensing of Trojan, a number of shortcomings have been identified by Westinghouse and the NRC in the Westinghouse ECCS Evaluation Models; this has resulted in the need to reanalyze ECCS performance using improved models with the errors corrected. For example, a Trojan reanalysis was performed to accommodate the fact that the water temperature in the reactor vessel upper head region was initially thought tp be intermediate betwe?n the hot-and cold-leg temperature, but subseouent experiments shewed it to be closer to the hot-leg temperature. Also, a reanalysis was done to correct an inconsistency in the handling of the heat input from the zirconium-water reaction. The most recent ECCS Evaluation Model concern was identified in 1979 involving acceptability of bestinghouse fuel rod burst and strain nedeling in light of recent experimental results.
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The overall result of these ECCS reanalyses'has been to decrease the margin between the calculated Trojan ECCS performance and the limits set forth in 10 CFR 50.46.
The current situation for Trojan, even before considering the fuel rod modeling concerns, is that it is no longer 0
possible to meet the peak cladding temperature { PCT) limit of 2200 F with the original maximum peaking factor (FQ) of 2.32.
A Technical Specifi-cation change request is pending at the NRC (LCA 51) to reduce the 0
allowable maximum Fg to 2.25 in order to meet the 2200 F PCT limit, and Fq is presently administratively limited to 2.23. A further reduction in Fg could potentially impact on operational flexibility by prohibit-ing certain load-following maneuvers. Furthermore, the current Trojan ECCS performance evaluation is based on the October 1975 version of the Westinghouse Evaluation Model which has been superseded by the February 1978 version.
Staff Evalt'ation This revised ECCS performance evaluation has been made using an NRC-approved evaluation model which satisfies the requirements of 10 CFR 50.46 and Appendix K.
The currently applicable small-break LOCA analyses for Trojan are unaffected by the considerations that have led to having the design basis LOCA analyses performed. Therefore, it is appropriate that the current ECCS performance evaluation consists only of a spectrum of worst-case large-break LOCAs. The methods for accounting for steam generator tube plugging are an integral part of the Westinghouse evalu-ation model and have been previously approved by the NRC. The discussion of the impact of incorporating NRC~ fuel rod models is consistent with current NRC policy and adequately responds to the concern.
The revised ECCS analyses were made with an evaluation mdel in whicn certain calcalational errors and inconsistencies have been corrected.
fio change in actual plant operating conditions is intended; it is only intended that the analysis of the ECCS during plant operations be correct and appropriately conservative.
The staff has reviewed the calculation of the containment backpressure in the revised ECCS analysis and determined that the analysis was perfor ed using a previously approved model and containrent parameters.
The fiRC staff has been generically evaluating three r.aterials models that are used in ECCS evaluation models. Those models are cladding rupture temperature, cladding burst strain, and fuel assecly flow blockage. We have (a) met and discussed our review with Hestinghouse an'd other industry representatives (1), (b) published !!UREG-0530,
" Cladding Swelling and Rupture Models for LOCA Analysis" (2), and (c) required fuel vendors and LWR licensees using Zircaloy cladding to confirm that their plants would continue to be in confor=ance with the
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ECCS criteria of 10 CFR 50.46 if the materials models of tiUREG-0630 were substituted for those models of their ECCS evaluation rodels (3, 4).
Until we complete our generic review and inplement new acceptance criteria for cladding models, we are requiring tnat plant ECCS reanalyses be accompanied by supplemental calculations performed with the materials models of NUREG-0630. For these supplemental calculations, we are accepting other compensatory nodel changes that may not yet be approved by the NRC, but are consistent with the changes allcwed for the confir-matory operating reactor calculations mentioned above.
5-By letters dated Janu'ary 9 and May 7, 1980 (5,5), PGEco provided supplemental *ECCS calculations. The calculation in the May sucmittal updated the January analysis by (a) increasing the water tergerature in the reactor vessel upper head region (b) correcting the inconsistency in the heat input from the zirconium - water reaction, (c) accounting for steam generator tube plugging, (d) using the February 1972 ECCS evaluation model, which superseded the October 1975 version, (e) using a higher total peaking factor, and (f) accounting for a non-conservatism recently identified by Westinghouse in their February 1978 ECCS evaluation model. As described in their letter dated ibyerrber 16, 1979 (7),
Westinghouse had discovered that LOCA analyses of actual plant heatup rates were at relatively slew temperature-ramp rates; whereas, the 1978 ECCS evaluation model was, in part, based on cladding burst tests that we.e conducted at relatively fast temperature-ramp rates.
The Trojan submittal assessed the combined impact of this calculational 0
error and the tRJREG-0630 models to be worth 855 F peak cladding temoerature above that previously calculated. Westinghouse has shown that a reduction 0
in total peaking factor, Fg, of 0.042 would offset the 855 F increase in peak cladding temperature. Westinghouse also identified a margin in Fg available through the use of a new thermohydraulic model (ucper head injection ttchnology) already approved for some applications. This margin was worth 0.20 in Fg. Thus no Fq reduction is required for Trojan.
Staff Findings The staff has reviewed the calculation of the containment backpressure in the revised ECCS analysis and deter =ined that the analysis was per-formed using a previously approved model and containment parameters. He find the revised analysis acceptable.
6 The staff has reviewed tne LCA response to cur latest fuel and eccel cor.cerns and we conclude that FGECo has satisfied our concerns related to tne swelling and rupture issue for plant coeration at a total peaking factor of 2.32.
This revised ECCS performance evaluation has been made using an NRC approved evaluation mocel which satisfies the requirements of 10 CFR 50.46 and 10 CFR 50 Appendix K. A 6 percent uniform steam generator tuce plugging is incorpor-ated into the model. A maximum total peaking factor of 2.32 was used, resulting 0
in a peak clad temperature of 1970 F for the worst-case Le.;A.
This LCA supersedes LCA 51, and justifies the withdrawal of LCA 51 since the revised analysis demonstrates that a reduction in Utes total peaking factor is not required.
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References l
1.
Memorandum from R. P. Cenise US?iRC, to R. J. Mattson, "Sux.ary Minutes of Meeting on Cladding Rupture Temperature, Cladding Strain, and Assesly Flcw Blockage," Novecer 20. 1979.
2.
D. A. Powers and R. O. " eyer, " Cladding Swelling and Rupture Models for LOCA Analysis," UStiRC report NUREG-0630, April 1980.
3.
Letter from D. G. Eisenhut, USNRC, to all Operating Light Water Reactors, dated Noveder 9,1979.
4 Meer)randum from H. R. Denton, USNRC, to Comissioners, " Potential Deficiencies in ECCS Evaluation Models," t;ovember 26, 1979.
5.
Letter from C. Goodwin, Jr., PGECo, to D. G. Eisenhut, USf1RC, dated January 9,1980.
6.
Letter from C. Goodwin, Jr., PGECo, to R. A. Clark, USNRC, dated :4ay 7,1980.
7.
Letter from T.M. Anderson, W, to D. G. Eisenhut, USNRC, tis-TMA-2163, dated November 16, 1979.
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