ML19351G322
| ML19351G322 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/10/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19351G319 | List: |
| References | |
| NUDOCS 8102230508 | |
| Download: ML19351G322 (6) | |
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l SAFETY EVALUATI0'l B,Y THE OFFICE OF NUCLEAR REACTOR REGULATI0f:
I SUPPORTIN'i AMENDMENT N0. 21 TO FACILITY OPERATING LICENS i
GEORGIA POWER COMPANY 0GLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALT0fi, GEORGIA EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO. 2 DOCKET N0. 50-366 1.
I_NTRODUCTION 17, 1980 Georgia Power Company (the licensee)
By letter dated October requested revisions to the Technical Specifications (TSs) appended tn t
Facility Operating License No. NPF-5 to complete the first refueling of the Edwin 1. Hatch Nuclear Plant, Unit No. 2 (Hatch 2), and begin Cycle 2 The original subnittal was revised on January 30, operation (Ref.1).
1981 (Ref. 2) to take advantaae of the fact that actual control blade insertion are typically faster than the scram tire scran times to 20' assumed in the licensing analyses, and the minimum critical power ratio (MCPR) operating limits calculated using actual scran time data are less i
limiting than those derived using the NRC staff's conservative fac tors.
(GE) plant-specific reload report (Ref. 3).
I In addition to the routine considerations in any reload application, the the NPF-5 license conditions:
2.C.(3)(a) Fuel submittal also addressed Abnormal Operational Transient Reanalysis, and Performance,2.C.(3)(c) 2.C.(3)(d) Boiling Transition Data, which are eligible for deletion.
The Hatch 2 Reload 1 involves loading 164 (P8x8R) fuel bundles P8DRB284LA.
used during Cycle 1.
l 2.
EVALUATION 2.1 TRANSIENTS To Various transient events will reduce the MCPR from its operating value.
assure that the fuel cladding integrity safety limit MCPR will not be violated withdrawal error, loss of feedwater heating and the pres l
This reanalysis calculates the reduction have been reanalyzed by the licensee.
b p
5] 0 2 2 3 09C$ib h
. in critical power ratio (CPR) for each of the most limiting transients and the Each of these events largest is used to establish the operating limit MCPR.
has been conservatively analyzed for each of the fuel types, i.e., 8x8R and The analysis P8x8R, and for the full range of exposure through the cycle.
shows that the most limiting transients for this cycle are the pressurization events, load rejection and feedwater controller failure.
For the analysis of the limiting pressurization events, the licensee used ODYN The licensee per the requirements of our acceptance of this cede (Ref. 4 and 5).
The change a:sures that the requirements has proposed a change to the MCPR TS.for credit of scram speed are satisfied over the range of cycle conditions and fuel types.
2.2 ACCIDENT ANALYSIS EMERGENCY CORE COOLING SYSTEM (ECCS) PERFORtiANCE I.NALYSIS 2.2.1 The licensee has reevaluated ECCS performance for the new reload fuel design by The results of methods that have been previously accepted by the NRC staff.
We have reviewed the plant-specific analysis are given in Section 14 of Ref. 3.
the infomation that has been submitted by the licensee and have concluded that all requirements of 10 CFR 50.46 and its Appendix K will be met when the reactor is operated in acco.Mance with the proposed maximum average planar linear heat These generation rate (MAPLHGR) limits versus average planar exposure values.
MAPLHGR limits vs. average planar exposure curves have been incorporated in the revised TSs.
2.2.2 CONTROL ROD DROP ACCIDENT Because the characteristic accident analysis input parameters for the worst case control rod drop accident were not bounded by all the assumptions of the bounding The analysis, the licensee reanalyzed this accident on a plant-specific basis.
results showed the peak fuel enthalphy to be less than the 280 calories per gram limit.
2.2.3 FUEL LOA 91NG ERROR t
The licensee has considered the effect of a possible fuel loading error on bundle An analysis of the most severe misoriented and mislocated fuel loading error per the accepted version of the generic reload topical (Ref. 6), shows CPR.
that the worst possible rotation or mislocation of a fuel bundle will not cause
(
l a violation of the safety limit MCPR.
l 2.2.4 OVERPRESSURIZATION ANALYSIS The overpressurization analysis of the main steamline isolation valve closure with high flux scram, which is the limiting overpressure event, has be with the 0DYN code.
tion limit that is adequate to account for the failure of one safety valve.
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2.2.5 THERMAL-HYDRAULIC STABILITY A themal-hydraulic stability analysis was performed with the methods l
l described in Ref. 6.
The results show that the channel hydrodynamic and reactor core decay ratios at the least stable operating state aretbelow the ultimate performance limit decay ratio of 1.0.
(The least stable operating state corresponds to the intersection of the natural circulation curve and the 105% rod line on the power-flow map.)
Generic concerns on operation at natural circulation conditions have been raised due to increasing decay ratios as equilibrium fuel cycles are approached and as reload fuel designs change.
These concerns relate to both the consequences of operation at decay ratios of 1.0 and the capability of the analytical methods to accurately predict decay ratios.
A requirement to preclude normal operation in the natural circulation mode has been instituted in the plant TSs. This restriction continues to provide a significant increase in reactor stability operating margins and is acceptable.
2.
2.6 CONCLUSION
Based on the foregoing, the proposed TSs and supporting analysis are acceptable.
2.3 LICENSE CONDITION 2.C.(3)(a), " FUEL PERFORMANCE" Condition 2.C.(3)(a) of the Hatch 2 license states:
" Georgia Power Company shall, prior to startup for that cycle of operation in which burnups greater than 20,000 megawatt days per ton of uranium are expected to be attained, provide for Commission review and obtain Commission approval of GEGAP-III calculations and other affected analyses utilizing fission gas release calculational methodology approved for burnups greater than 20,000 megawatt days per ton of uranium."
This license condition was imposed a's a result of our concern (Ref. 7) that fission gas release from the fuel may not be correctly calculated for burnups above 20,000 mwd /MTu.
The licensee has elected (Ref.1) not to provide revised calculations which account for enhanced fission gas release at high burnups for the proposed cycle of operation. As a basis for this decision, the licensee cited an NRC letter dated March 10,1980 (Ref. 8) on this subject.
The NRC letter, in turn, cit'es a GE letter (Ref. 9) that describes calculations to 33,000 IWd/MTu for a number of GE plant types and fuel designs. These calculations have been accepted (Ref. 8) on an interim basis for continued operation at other Boiling Water Reactor (BWR) facilities and would also apply to Hatch 2 for Cycle 2 operation.
The licensee has also noted (Ref.1) that a revised GE fuel performance model, GESTR, is currently under NRC staff review.
Until such time that the GESTR model is approved and incorporated into plant safety analyses,
. it is our intent to require revised calculations that account for burnup effects on fission gas release oMy when anticipated conditions approach or exceed 33,000 mwd /MTu, the highest value considered for loss of Coolant Accident (LOCA) an& lysis in Ref. 4.
Hatch 2 will not approach this value Since the present MAPLHGR tenninates at 30,000 mwd /MTu,(no in Cycle 2.
We therefore conclude that Condition 2.C.(3) a) other limit is needed.
may be deleted from the Hatch 2 operating license.
2.4 LICENSE CONDITION 2.C.(3)(c), " ABNORMAL OPERATIONAL TRANSIENT REANALYSIS" Facility Operating License No. NPF-5 Condition 2.C.(3)(c) requires a re-analysis of the limiting abnormal operational transient using the GE One Dimensional Core Transient Model 0DYN computer code prior to startup following the first refueling outage for Hatch 2.
We reviewed results of the abnormal transient reanalysis using the ODYN computer code (Ref. 3), which show that the operating MCPR limit of 1.24 for 8x8R/P8x8R fuel from beginning of cycle (BOC) to end of cycle (E0C) is acceptable (Ref.10). We note that the application of additional conservatism using adjustment factors based on measured scram time data resulted in a MCPR value for P8x8R fuel of 1.25.
This conservative result appears in TS Figure 3.2.3-2 and was caused by the statistical treatment of round-off error.
We find that the operating MCPR limits computed by the ODYN code are acceptable and have been applied conservatively.
We therefore conclude that Condition 2.C.(3)(c) may be deleted from the Hatch 2 operating license.
2.5 LICENSE CONDITION 2.C.(3)(d), "B0ILING TRANSITION DATA" Condition 2.C.(3)(d) of the Hatch 2 license states:
" Georgia Power Company shall, prior to startup following the first refueling outage, provide for Commission review and obtain Com-mission approval for the use of boiling transition data for 8x8 fuel bundles with two water rods in order to support the use of the i
l GEXL correlation for fuel bundle radial peaking patterns expected to be encountered during operation beyond the first cycle."
The concern with boiling transition data for 8x8R fuel and the GEXL cor-relation was documented during the original Hatch 2 licensing process (Ref.
11 and 12).
We have reviewed the GE submittals on this subject (Ref.13,14 and 15) and found that the GEXL correlation for fuel bundle radial peaking factors is acceptable for Sx8R fuel reload application (Ref.16).
l As stated in Ref.11 for BWR cores which reload with GE's retrofit P'
- fuel, I
the allowable MCPR, resulting from either core-wide or localized abno operational transients, is equal to 1.07.
With this MCPR safety limit, at least 99.9% of the fuel rods in the core are expected to a^ void boiling transi-l tion.
. The 1.07 safety limit minimum critical power ratio (SLMCPR) proposed by the licensee for Cycle 2 represents a.01 increase from the 1.06 SLMCPR applicable during Cycle 1.
T$e basis for the revised safety limit is addressed in Ref.17, while cur generic approval of the new limit is given in Ref.11. This change is consistent with the criteria of Standard Review Plan 4.4 and on that basis has been found acceptable in Ref.11.
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We therefore conclude that Condition 2.C.(3)(d) may be deleted from the Hatch 2 operating license.
ENVIRONMENTAL CONSIDERATIONE We have deterTnined that this 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 this amendment involves an action which is insigni-ficant from the standpoint of environmental impact, and pursuant to 10 CFR Section Sl.5(d)(4) that an environmental impact statement, or negative declara-tion and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
CONCLUSIONS We have concluded, based on the considerations discussec above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant 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 Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.
REFERENCES 1.
W. A. Widner (GPC) letter to the U. S. N. R. C. dated October 17, 1980.
2.
W. A. Widner (GPC) letter to the U. S. N. R. C dated January 30, 1981.
3.
" Supplemental Reload Licensing Submittal for Hatch Nuclear Power Station Unit 2 Reload-1," Y1003J01A!0, July 1980.
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4.
Memorandum from P. S. Check to T. M. Novak and R. L. Tedesco, " Safety Evaluation for ' Qualification of the One-Dimensional Core Transient Model For Boil.ing Water Reactors,' NE00-24154 and NEDE-24154P, Volume I, l
II and III," October 22, 1980, 5.
Memorandum from P. S. Check to T. M. Novak and R. L. Tedesco, " Supplemental Safety Evaluation for ODYN Code," November 20, 1980.
6.
General Electric Company Topical Report NEDE-240ll-P-A, " Generic Reload l
Fuel Application."
REFERENCES (Cont ',d._)
7.
D. F. Rcss, Jr. (NRC) memorandum to D. B. Vassallo (NRC) on "SER Input for Hatch, Unit 2" dated April 25, 1977.
8.
T. A. Ippolito (NRC) letter to C. F. Whitner (GPC) dated March 10, 1980.
9.
G. G. Sherwood (GE) letter 4o D.
F,. Ross, Jr. (NRC) dated December 22, 1976.
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- 10. Memorandum for T. Nohak froin P. S$ Check, SSER for ODYN Code, November 20, 1980.
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Letter, D. Eiseehut '(NRC) 'to R. Gridley (GE). " Safety Evaluation for the 11.
General Electric Topical Report, Generic Reload Fuel Application NEDE-240ll-P "
May 12, 1978.
12.
Safety Evaluation Report Related to Operation of Edwin I. Hatch Nuclear Plant Unit No. 2, NUREG-0411, June 1978.
- 13. Letter, R. E. Engel (GE) to L. S. Rubenstein (NRC), Response to NRC Concerns on the 8x8R GEXL Correlation, August 26, 1980.
Letter, J. F. Quirk (GE) to 0. D. Parr (NRC), " General Electric Licensing 14.
Topical Report, NEDE-240ll-P-A, Generic Reload Fuel Application, Appendix D,"
l February 28, 1979.
Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " General Electric 15.
Licensing Topical Report NEDE-240ll-P-A, Generic Reload Fuel Application Appendix D Submittal." December 14, 1979.
Menorandum for T. Novak from L. Rubenstein, SER for the GEXL Correlation 16.
for 8x8R Fuel Reload Application per the Appendix D submittals of Generic l.
Reload Fuel Application NEDE-240ll-P dated February 28, 1979 and December 14, 1979, February 1981.
17.
" Generic Reload Fuel Application," General Electric Report, NEDE-240ll-P-3, March 1978.
l Dated: February 10, 1981 i
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