ML19208C503

Safety Evaluation Supporting Amend 69 to License DPR-57

ML19208C503
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Site:	Hatch
Issue date:	08/06/1979
From:	Office of Nuclear Reactor Regulation
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ML19208C502	List: ML19208C501 ML19208C503 ML19208C505
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UNITED STATES rh NUCLEAR REGULATORY COMM!SSION

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 69 TO FACILITY OPERATING LICENSE NO. DPR-57 GEORGIA POWER COMPANY OGLETHORPE ELECTRIC MEMBERSHIP CORPORATION MUNICIPAL ELECTRIC ASSOCIATION GF GEORGIA CITY OF DALTON, GEORGIA EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 DOCKETNO.50-32(

1.0 Introduction Georgia Power Company (GPC) has proposed changes to the Techgal Specifications of the Edwin I. Hatch Nuclear Plant, Unit 1.

Tne proposed changes relate to the replacement of 164 fuel assemblies constituting refueling of the core for Cycle 4 operation at power le"els up to 2436 Mwt (100% power).

In support of the reload application, the proposed Technical Specification changes [{censee has provided nd the GE BWR supple-mental licensing submittal for Hatch-l. (3)

This reload involves 1cading of GE 8x8 retrofit (8x8R) fuel. The description of the nuclear and mechanical designs is contained in References 4 and 5.

Reference 4 also contains a complete set of references to topical reports which describe GE's analytical methods for nuclear, thermal-hydraulic, transient and accident calculations, and information regarding the applicability of these methods to cores containing a mixture of 8x8 and 8x8R fuel.

Values for plant-specific data such as steady state operating pressure, core flow, safety and safety / relief valve setpoints, rated thermal power, rated steam flow, and other design parameters are provided in Reference 4.

Additional plant and cycle dependent information is provided in the reload application (Reference 3) which closely follcws the outline of Appendix A of Reference 4.

Appendix C of Reference 4 includes a description of the staff's review, approval, and conditions of approval for the plant-specific data ad-dressed in Refere1ce 4.

The above-mentioned plant-specific data have been used in the transient and accident analysis provided with the reload application.

$70 7 900260 7

. Our safety evaluation (Reference 4) of the GE generic reload licensing topical report has also concluded that the nuclear and mechanical design of the 8x8R fuel, and GE's. analytical methods for nuclear and thermal-hydraulic calculations as applied to mixed cores containing 8x8 and 8x8R fuel, are acceptable. Approval of the application of the analytical methods did not include plants incorporating a prompt recirculation pump trip (RPT) or Thermal Power Monitor (TPM).

Because of our review of a large number of generic considerations related to use of 8x8R fuel in mixed loadings, and on the basis of the evaluations which have been presented in Reference 4, only a limited number of additional areas of review have been included in this safety evaluation report. For evaluations of areas not specifi-cally addressed in this safety evaluation report, the reader is re-ferred to Reference 4.

2.0 Evaluation 2.1 Nuclear Characteristics For Cycle 4 operation of Hatch-1,164 fresh gR fuel bundles of type 80RB265H will be loaded into the core.

The remainder of the 560 fuel bundles in the core will be 168 once-burned type 80RS265H bundles, 92 twice-burned type 8DB250 bundles, and 136 bundles from the initial core.

The fresh fuel will be loaded and the previously peripheral fuel will be shuffled inward to constitute an octant-symmetric core pattern.

Based on the data provided in Reference 3, both the control rod system and the standby liquid control system will have acceptable shutdown capability during Cycle 4.

2.2 Thermal Hydraulics 2.2.1 Fuel Cladding Integrity Safety Limit MCPR As stated in Reference 4, for BWR cores which reload with GE's retrofit 8x8R fuel, the safety limit minimum critical power ratio (SLMCPR) re-sulting from either core-wide or localized abnormal operational tran-sients is equal to 1.07.

When meeting this SLMCPR during a transient, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The 1.07 SLMCPR to be used for Cycle 4 is unchanged from the SLMCPR previously approved for Cycle 3.

The basis for this safety limit is addressed in Reference 4, while our generic approval of the limit is given in the staff evaluation included in Reference 4.

/

1012 238 2.2.2 Operating' Limit MCPR Various transient events can reduce the MCPR from its nornal operating level. To assure that the fuel cladding integrity SLMCPR will not be violated during any abnormal operational transient, the most limiting transients have ~been reanalyzed for this reload by the licensee., in order to determine which event results in the largest reduction in the minimum critical power ratio. These events have been analyzed for the exposed 717 and 8x8 fuel and the exposed and fre~sh 8x8R fuel. Addition of the largest reductions in critical power ratio to the SLMCPR estabi bhes the operating limits for each fuel type.

2.2.2.1 Transient Analysis Methods The generic methods used for these calculations, including cycle-independent initial conditions and transient input parameters, are described in Reference 4.

The staff evaluation, included as Appendix C of Reference 4, contains our acceptance of the cycle-independent v al ues. Additionally, Appendix C contains our evaluation of the transient analysis methods, together with a description and summary of the outstanding issues associated with these methods. Suppl e-mentary cycle-independent initial conditions and transient input paraneters used in the transient analyses appear in the tables in Sections 6 and 7 of Reference 3.

Our evaluation of the methods used to develop these supplementary input values is also included in Appendix C of Reference 4.

At the time we completed our evaluation of the generic methods, the acceptability of the GEXL critical power correlation for use in connection with the retrofit fuel design had not been adequately documented by GE. The staff found, however, that the then available 8x8R critical power test data was sufficient to support the accept-ability of GE's retrofit 8x8 fuel design fci SW] core reloads for Accordingly, we stated "

that future BWR one operating cycle.

core reload applications involving retr ofit 8x8 fuel for a second operating cycle would have to include additional information which adequately justified the correlation for application to 8x8R fuel operating beyond one cycle.

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1012 239

. GE has prepared a report on this subject (7'1 that provides the results of full scale critical power tests performed on 8x8R fuel bundles.

Thetestsincludedbothtransientandstgy-statesimu-lations and followed the same approved procedures used for the standard 8x8 and 7x7 fuel designs. The analysis of a total of 577 steady-state data points was performed using methods also previously approved by the staff. The data spanned a range of 1ocal power peaking and flow conditions. GE stated that the GEXL correlation was applicable to the retrofit fuel after adjustments were made to the additive constants used in the formulation of the rod-by-rod R-factors (Figure 3-1 of Reference 7).

Using the new additive constants, GE assessed the accuracy and pre-cision of the GEXL correlation. The results showed that the cor-relation provides for a mean predicted-to-measured critical power ratio of 0.9879 with a standard deviation of 0.0234.

The 8x8R GEXL correlation has a conservative bias when viewed over the entire range of its applicability (which is the same as the 8x8 correlation). Thus, the 8x8R GEXL correlation has better precision than the 7x7 and 8x8 GEXL correlations for predicting critical bundle powers when viewed over the entire range of applicability. Further-more, the 3.6% standard deviation and zero bias of the GEXL correla-tion bound the statistical characteristics of the 8x8R GEXL correla-tion used in the GETAB statistical analysis to derive the 1.07 safety 1imit MCPR.

The infomation furnished by GE (7) was intended to apply to all BWR cores that contain 8x8R fuel. This information is currently being re-viewed by the staff for generic application.

A' aough the evaluation is not yet complete, it was noted that the critical power test condi-tions specifically representative of normal conditions during second-cycle fuel operations exhibit a slightly non-conservative bias in predictions. This observation focuses in on a correlation behavioral concern not explicitly addressed in the overall GETAB methods approval (8) for the 7x7 and 8x8 fuel types. However, based on our review to date, we conclude that the 8x8R GEXL correlation has an acceptability and applicability equivalent to those of the 7x7 and 8x8 GEXL correlations previously approved by the staff.

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. Therefore, we conclude that there is sufficient conservatism implicit in the generic detertaination of the 1.07 safety limit MCPR to offset a possible nonconservatism associated with this copern for Cycle 4 of Hatcr. 1.

Additionally, the generic evaluation

/ considered an all 8x8R equilibrium core, whereas the Cycle 4 core involves 7x7, 8x8 and 8x8R fuel in a non-equilibrium condition.

In view of these con-servatisms (which are representative of a typical non-equilibrium 8x8R reload core) we believe that the overall thermal-hydraulic (GETAB) methods are adequate for establishing conservative MCPR operating limits for Cycle 4.

2.2.2.2 Transient Analysis Results The transients evaluated were the limiting pressure and power increase transient (turbine trip without bypass in this case), the limiting coolant temperature decrease transient (loss of a feedwater heater),

the feedwater controller failure transient, cnd the control rod with-drawal error transient.

Initial conditions and transient input para-meters as specified in Sections 6 and 7 of Reference 3 were assumed.

The calculated system responses and aCPRs for the transients and conditions listed above have been analyzed by the licensee. Resul ts were as follows:

ACPR ACPR ACPR 7x7 8x8 8x8R Turbine Trip Without Bypass

.06

.10

.10 Loss of 100'F Feedwater Heater

.13

.14

.14 Feedwater Con-troller Failure

.06

.07

.07 Rod Withdrawal Error

.19

.17

.12 Fuel Loading Error, Rotated Sundle*

<.09

<.09

.09

• The.aislocated bundle error is considered separately in Section 2.3.3.

s' 1012 241 The above analyses include the effect of an End of Cycle recirculat:on pump trip (RPT) initiated by turbine stop valve closure or throttle valve fast closure. This RPT feature inserts negative reactivity into the reactor due to the rapid flow decrease and resultant increased voiding. Thus, the RPT helps shut down the reactor, effectively in-creasing the effectiveness of turbine-initiated scrams.

The tgsient analyses described abo [f67ere perforned with the REDY code.

A new improved code, ODYN, has been developed by GE.

The CDYN code, which uses a more physically correct model of the plant, generally predicts smaller ACPRs than the REDY code when the transient under study is fairly severe. However, as transient severity is less-ened, ODYN predicts a greater ACPR than REDY (Reference 10, p.1).

Both codes are run with conservative input values, but CDYN is a bgyr predictor of plant behavior once these input values are specified GE has stated (Reference 10) that REDY can still be used because the limiting transient has a ACPR sufficiently large to be above the region where REDY is non-conservative with respect to CDYN. We have proceeded on this basis in approving reloads thus far.

The add'. tion of the RPT feature has significantly reduced the aCPR associated with transients involving a turbine trip.

(Reductions as great as roughly a factor of 2 are presented on p.12 of Reference 10.)

This improvement has brough e reload 3 transient analysis into the region where GE's assertion is no longer valid for those tran-sients which involve a turbine trip. The limited data available to us clearly indicate that calculations which include axial effects and detailed steam line modeling predict more severe results than do point kinetics REDY calculations.

Therefore, unless more justification for the REDY-based calculations is forthcoming, th( transient analysis results must be conservatively adjusted to account for this effect. The analyses affected are the turbine trip without bypass transient and the feedwater controller failure transient. The loss of feedwater heater transient is much slower, and therefore should be well simulated by,1oint kinetics calculations. Moreover, although the loss of a feedwater heater leads to a reactor trip in this case, there is no ;urbine trip and thus no significant excitation of acoustic resonar.:es in the steam line. The remaining analyses (rod withdrawal errcr and rotated bundle) are not calculated with REDY and therefore are not affected.

s' 1012 242

. Thus, the turbine trip transient and the feedwater controller failure trarsient (which involves a turbine trip) must have their analyses adjusted to account for defects in the steam line and core axial response modeling. Comparisons of the REDY and CDYN calculations presented on p.12 of Reference 10 have enabled us to estimate a non-conservative trend for the "EDY-calculated aCPR values. Accordingly, we will require that the

.vR values used in the calculation of the operating limit MCPR be adjusted qwards for those transients involving a turbine trip. This results in toe following ACPRs:

aCPR aCPR aCFR 7x7 8x8 8x8R Turbine Trip Without Bypass

.11

.14

.14 Loss of 100 F Feedwater Heating

.13

.14

.14 Feedwater Con-troller Failure

.11

.12

.12 Rod Withdrawal Error

.19

.17

.12 Fuel Loading Error, Rotated Bundle

<.09

<.09

.09 Addition of the most severe aCPR to the safety limit (1.07) gives the appropriate operating limit MCPR for each fuel type. This will assure that the safety limit MCPR is not violated due to transients or fuel loading errors. Using the revised table of aCPRs, the operating limits become:

1.26 for 7x7 fuel (based on rod withdrawal error) 1.24 for 8x8 fuel (based on rod withdrawal error) 1.21 for Sx8R fuel (based on turb.ne trip without bypass nd loss of 100*F feedwater heating)

These values are numerically equal to those originally proposed by the licensee, since the adjustments discussed above did not make the affected analyses limiting.

These values are based on predicted End-of-Cycle (E0C) reactor kinetics with prompt Recirculation Pump Trip (RPT). We have not completed cur review of the hardware implementation of the prompt RPT feature for Hatch 1.

j Therefore, complete credit for the prompt RPT cannot be given.

1012 243

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The prompt RPT provides a reduction in recirculation flow which causes increased core void and thus introduces negative reactivity to mitigate transient power increases.

It is most effective in this objective at EOC, since cycle-dependent core characteristics result in large void reactivity feedback towards EOC.

With regard for the effect of the prompt RPT on transient consequences, we have reviewed previous, applicable transient analyses (Reference

18) whicli was approved by Amendment No. 52. We have concluded that core characteristics, e.g., void coefficient and scram worth, which are important inputs to the transients for which the prompt RPT has been designed, conservatively bound the present core character-istics. The results of these analyses shcw that the transient con-sequences would be bounded by the current ACPR results until about 1000 mwd /t before E0C. Therefore, the operating limit MCPR's will be conservative up to 2000 mwd /t before EOC. The Technical Spec-ifications have been amended to reflect this condition. MCPR limits for 2000 mwd /t-E004 to E0C 4 will be determined after completion of our review of the hardware implementation.

2.2.3.1 Thermal Power Monitor The Thermal Power Monitor (TPM), also called an APRM Simulated Thermal Power (STP) Trip in some documents, is a modification to the APRM trip system. The modified system generates two trips: a trip with a ficw biased setpoint and a second trip with a setpoint fixed at 120% power.

The flow biased setpoint is unchanged from that presently in the Tech-nical Specifications. However, the TPM conditions the APRM output to apply a time constant of about six seconds, which is less than but comparable to the fuel thermal time constants (seven to ten seconds).

Thus, the signal compared to the flow biased setpoint is a conserva-tive simulaton of fuel rod heat flux. This feature allows the plant operator to avoid spurious trips due to minor neutron flux overshoots when maneuvering the reactor.

The signal which is compared to the fixed 120% power setpoint is not modified. Thus, there is always a " fast scram" at 120% power in addition to the " heat flux" scram which may be below 120% power, depending on flow.

2.2.3.2 Effect of TPM on Safety Analyses Since all the transient analyses are done assuming full design flow, the TPM has no effect because the 120% " fast" trip is identical to the original system at full power. Any effect due to the TPM must be on analyses which are initiated from low flow conditions.

1012 244

. GE has addressed the analysis of the various transients initiated from low flow conditions on pp. 5-8 of Reference 4.

The generic analyses described there show that only the idle recirculation pump startup, recirculation flow controller failure (increasing),

feedwater controller failure (max demand), and rod withdrawal error can become more severe at low flow conditions. This is the basis for the flow-dependent multiplier (K ) in every GE plant's f

Technical Specifications.

The analyses supporting the K, factor did not take credit for the flow biasing, bN2jnstead conservatively assumed the trip to occur at 120% power.

Therefore, the analyses supporting the flow-dependent multiplier (K ) remain bounding.

f Similarly, the various accident analyces which involve a neutron flux induced trip (e.g., rod drop acc, dent) assume the trip to occur at 120". power regardless of initial power or flow conditions.

Therefore, the validity of _the accident analyses is also unaffected by the introduction of the TPM.

2.3 Accident Analyses 2.3.1 ECCS Apoendix K Analysis For the previous cycle, the licensee re-evaluated the adequacy of Hatch 1 ECCS pergance in connection with the retrofit 8x8 reload fuel design The methods used in this analysis were previously approved by the staff. For that cycle, we reviewed the ECCS analysis results submitted by the licensee and concluded that Hatch I would be in conformance with all the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 when operated in accordance with the MAPLHGR 8x8R versus Average Planar Exposure val'gwhich appeared in the proposed plant Technical Specifications.

Since the Cycle 4 reload fuel is of the same design as that loaded for Cycle 3, we find this same LOCA-ECCS safety analysis and related Technical Specifications to be equally acceptable for showing com-pliance with the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 40 for the current Cycle 4 reload fuel.

2.3.2 Control Rod Droo Accident For Hatch 1 Cycle 4, the accident reactivity shape function (cold) does not satisfy the requirements for the bounding analyses des-cribed in Reference 4.

Therefore, it was necessary for the licensee to perform a plant and cycle specific analysis for the control rod drop accident. The results of this aralysis indicated that the peak fuelythalpy for this event would be at most 197.35 calories per gram.

Since this is well belcw the criterion of 280 calories per gram, we find the results of this analysis to be acceptable.

1012 245

. 2.3.3 Fuel Loading Error Potential fuel loading errors involv'ing miscriented bundles have been explicitly included in the calculation of the operating limit MCPR.

Potential errors involving bundles loaded into incorrect positions have also been analyzed by a method which considers the initial MCPR of each bundle in the core, and the resultant MCPR was shown to be greater than 1.07.

This GE method for analysis of mis-orienth5jnd misloaded bundles has been reviewed and approved by the staff.

The analyses which have been performed for potential fuel loading errors for Hatch 1 Cycle 4 are acceptable for assuring that CPRs will not be below the safety limit MCPR of 1.07.

2.3.4 Overoressure Analysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, has been perfomed in accordance with the requirements of Reference 4.

As specified in the staff evaluation included in Reference 4, the sensitivity of peak vessel pressure to failure of one safety valve has also been eval uated. We agree that there is sufficient margin between the peak calculated vessel pressure and the design limit pressure to allow for the failure of at least one valve. Therefore, the limiting over-pressure event as analyzed by the licensee is considered acceptable.

2.4 Thermal Hydraulic Stability The r; ilts of the thermal hydraulic stability analysis (3) show that the channel hydrodynamic and reactor core decay ratios at the natural circulation - 105% rod line intersection (which is the least stable physically attainable point of operation) are belcw the sta-bility limit.

Because operation in the natural circulation mcde at greater than 1% rated thermal power will be prohibited by Technical Specifica-tions 3.6.J.1, there will be added margin to the stability limit and this is acceptable.

2.5 Startup Test Program The licensee has not changed his startup test program from that approved for the previous cycle. This program therefore remains acceptable.

1012 246 g

. 3,0 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 standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), 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.

4.0 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 accidents previously considered and does not involve a significant decrease in a safety margin, the change does not involve a significant hazards considera-tion, (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:

August 6,1979 1012 247

~,

, References 1.

Letter, Chas. F. Whitmer (GPC) to Director of Nuclear Reactor Regulation (NRC), dated March 22, 1979.

2.

" Proposed Changes to Technical Specifications," Attachments 1 and 2 'f Reference 1.

3.

" Supplemental Reload Licensing Submittal for Edwin I. Hatch Nuclear Plant Unit 1 Reload 3," NED0-24175, January 1979.

4.

" General Electric Boiling Water Reactor Generic Reload Application,"

NEDE-24011-P-A, May 1977.

5.

Letter, R. E. Engel (GE) to U. S. Nuclear Regulatory Commission, dated January 30, 1979.

6.

" General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NED0-10958, November 1973.

7.

" Basis for 8x8 Retrofit Fuel Thermal Analysis Application," NEDE-24131, enclosed in letter, R. Gridley (GE) to D. Eisenhut and D. Ross (NRC),

dated October 5, 1978.

8.

Letter, W. Butler (NRC) to I. Stuart (GE), dated October 2,1974.

9.

" Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NE00-10802, February 1973.

10.

" Impact of One-Dimensional Transient Model on Plant Operations Limits,"

enclosure of letter, E. D. Fuller (GE) to U. S. Nuclear Regulatory Com-mission, dated June 26, 1978.

11.

" Basis for Installation of Recirculation Pump Trip System," NED0-24119, April 1978.

12.

" Technical Specifications and Bases for Edwin I. Hatch Nuclear Plant Unit 1," Appendix A to Operating License DPR-47, Specification 2.1. A.1.C.

13.

" Loss of Coolant Accident Report for Edwin I. Hatch Nuclear Plant Unit 1," NED0-24086, December 1977, enclosure 2 of letter, C. Whitmer (GPC) to V. Stello (NRC), datcd January 5,1978.

1012 248

. 14.

"Edwin I. Hatch Nuclear Plant Unit No.1 Reload 2 Safety Evaluation Report," April 11, 1978.

15. Letter, D. G. Eisenhut (NRC) to R. Engel (GE), dated May 8,1978.

16.

" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," NEDO-24154, October 1978.

17.

Letter, R. J. Mattson (NRC) to General Electric Company (Attn. G. G. Sherwood), dated March 20, 1979.

18.

Charnly, J., " Supplemental Reload Licensing Submittal for Edwin I.

Hatch Nuclear Plant Unit 1 Reload 2," NED0-24078, Class 1, November 1977.

1012 249

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