ML17342B325

Safety Evaluation Supporting Licensee Use of Retran Computer Code for Sys Transient Analyses & Approving Licensee Topical Rept on PWR Lattice Physics Methodology

ML17342B325
Person / Time
Site:	Saint Lucie, Turkey Point, 05000000
Issue date:	04/19/1988
From:	Office of Nuclear Reactor Regulation
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ML17342B324	List: ML17342B323 ML17342B325
References
NUDOCS 8804290076
Download: ML17342B325 (7)

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UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D. C. 20555 ENCLOSURE 1

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE USE OF RETRAN COMPUTER CODE FOR SYSTEMS TRANSIENT ANALYSES FLORIDA POWER AND LIGHT COMPANY (FPSL)

TURKEY POINT Al'lD ST.

LUCIE PLANTS DOCKET NOS. 50-250/251 and 50-335/389

1.0 INTRODUCTION

The FP8L Topical Report NTH-G-6, "RETRAN Code, Transient Analysis Model Qualification," was submitted to demonstrate FPhL's technical competence to utilize the RETRAN computer code to perform systems transient analyses for their Turkey Point and St. Lucie plants.

This submittal was made in response to NRC generic Letter 83-11.

The subject topical report was submitted by FPSL in a letter dated January 7,

1986.

The licensee's original stated objectives of this report were:

(1) to present RETRAN base model verification results for each of the FPAL's plants, and (2) to demonstrate the proficiency of FP8L's personnel to perform systems transient analyses.

It was requested that these models be approved for non-LOCA transient analyses to support licensing actions.

During the staff review process, FPAL, in a letter dated March 2, 1987, revised their objectives of the topical report to:

(1) show that the model FPKL developed adequately simulates the behavior of the plants, and (2) show that the FPSL personnel have the capability to apply the models to various transients, so that in future licensing submittals, the licensee could reference the subject report to partially support its systems transient analyses using the RETRAN computer code.

In response to the staff requests for additional information, additional supporting mater'ials wer e submitted in FPSL letter dated April 10, 1987.

The staff has completed its review of the FPKL submittals with technical assistance from its consultants at the International Technical Services, Inc.

(ITS).

The staff evaluation is addressed below.

A technical evaluation report prepared by ITS is attached to this report.

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2. 0 EVALUATION FPSL has used a variety of RETRAN versions in the subject topical report..

RETRAN02/M0002 has been reviewed and approved by the staff for future use.

RETRAN02/HOD03 is currently under the staff review and has not yet been formally approved by the staff.

Nevertheless, for the purpose of demonstrating the licensee's technical competence for using the RETRAN code, it is acceptable to use any of the code versions.

Two-loop nodalizations were developed and used in the transient analyses for both the Turkey Point and St. Lucie plants.

Although both of these plant models are briefly described in the report and the nodalization diagrams are presented, the information provided in the topical report is not in sufficient detail to demonstrate that these RETRAN models are acceptable for appropriate system transient analyses for the Turkey Point and St. Lucie plants.

FP8L uses these'eveloped RETRAN models to perform system transient analyses and compare

+he results of these analyses to (1) the results submitted in the FSAR for the Turkey Point and St. Lucie plants, (2) Westinghouse generic study

results, and (3) some plant operating data.

From the information provided in the topical report and the licensee's response to the staff request for additional information, we conclude that the licensee has demonstrated the technical capability to utilize the RETRAN computer code to perform system transient calculations on their plants.

3.0 CONCLUSION

The staff concludes that the topical report does demonstrate that FPSL has the capability to use RETRAN computer code to perform systems transient calculations for the Turkey Point and St. Lucie plants and, therefore, fulfills the require-ments of Generic Letter 83-11.

The topical report is approved for future referencing to partially support the FPAL's licensing analyses.

However, additional comparisons between the RETRAN computed results and plant operating data together with appropriate nodalization, sensitivity studies and licensing assumptions will be necessary in future reports before these models are accepta-ble for licensing submittals.

Oated:

Apnl lK, 1988 Princi al Contributor:

C. Liang

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 ENCLOSURE 2

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT ON PWR LATTICE PHYSICS METHODOLOGY FLORIDA POWER AND LIGHT COMPANY DOCKET NOS.

50-250

-251 "335 AND -389

1. 0 INTRODUCTION By letter dated May 10, 1984, the Florida Power and Light Company, the licensee for the Turkey Point Units 3 and 4 and the St.

Lucie Units 1 and 2, submitted for review a topical report entitled, "PWR Lattice Physics Methods at Florida Power and Light Company" (Ref. 1).

Additional information was submitted on October 21, 1986, August 8, 1986, and June 26, 1987 (Refs. 2-4).

This report is the first of a three report series on the Florida Power and Light core physics methodology to support their reload methods but does not include the special assemblies for their flux reduction commitment for the pressurized thermal shock issue for the Turkey Point Units 3 and 4.

The Nuclear Performance Section of the SRXB reviewed the submitted information.

This review was accomplished by comparison of the results of a benchmarked standard problem with the results of the same problem derived with the proposed methodology (Ref. 2).

The standard problem was designed to test all of the essential computational capabilities of the methodology and was run by J.

Carew of BNL under the Technical Assistance FIN 8A-3845.

2. 0 EVALUATION The proposed methodology is based on the CHEETAH computer program (Refs.

5 8 6) to calculate core group cross-sections and neutron spectra.

This code is used for core fuel and weak absorber regions.

The heavy poison sections of the peripheral fuel assemblies designed to reduce the flux to the peripheral welds in Turkey Point 3 and 4 will not be treated with this program.

However, this

methodology contains a depletion model which is used to generate cross-sections for various stages of burnup.

This program encompasses the full range of anticipated temperatures from cold to full power, enrichments from

l. 5 to 4. 5 w/o U-235 and burnups from '0 to 65,000 WD/MTU.

2. 1 CHEETAH In ut/Out ut The code requires a description of the unit fuel cell, the geometry, and the materials.

In addition, the needed burnup steps and a fine group cross section library are required.

The fine group cross-sections are described in 295 thermal and 54 fast groups from the ENOF/ 8-I and -II files.

The output includes input verification, neutron spectra, number densities, few group microscopic cross sections, macroscopic parameters and depletion data.

The micro" and macro-scopic cross-sections for 2 and 4 groups are appropriate for PDg-7 input.

Fuel number densities are given for each burnup step.

2.2 The CHEETAH Model CHEETAH assumes an infinite array of unit cells arranged in a square or hexagonal lattice.

Materials constituting lattice irregularities are accounted for by postulating an extra region.

Hence, CHEETAH is a zero dimension, multigroup point depletion code.

The fast neutron spectra are calculated using the B-1 approximation.

The thermal neutron spectra are calculated using a modified Amouyal"Benoist method (Refs.

7, 8,

& 9).

The Doppler resonance broadening for U-238 and the 1.056 eV P -240 resonance are u

accounted for.

The Oancoff correction for the lattice shadowing effects is applied.

Fission product cross-sections are generated from a polynomial fit.

A depletion calculation is performed over specified burnups and assembly temperatures.

The Xenon and Samarium buildup are accounted for using the method of the CANDLE code (Ref. 10).

2.3 Verification of the Results The main result of a CHEETAH calculation is a set of group cross-sections for certain geometrical and material configuration.

The cross-sections can be used as input to a diffusion code to yield integral parameters which can be measured.

Diffusion codes usually include biases, approximations and other compensation of the computed value which are combined with the effect of the cross-sections.

In this manner a set of cross-sections by themselves cannot be reviewed for acceptability because there are no integral directly measurable quantities which correspond to them.

Therefore, here the review is limited to the methodology of the cross-section derivation rather than the acceptability of the cross-sections themselves.

The methodology as described in the submittal is similar to the accepted standards of similar methods.

It accounts for all of the phenomena which have an effect on the cross-section values.

The derived cross sections were used with diffusion codes for the estimation of critical experiments (Ref. ll, 12, 8 13).

The estimated values of k ff or critical bucklings were very close to eff those measured in these criticals.

The cross-sections were also used to derive isotopic distributions in depleted assemblies and compared them to the corresponding measured values.

In all of the comparisons the fissionable discharge was very closely predicted.

Also good agreement was obtained in cross comparisons with other methodologies.

Finally, a standard problem was run at BNL by staff consultants.

The results were acceptable, provided that this methodology will not be used for the calculation of temperatur e dependent parameters.

3. 0 CONCLUSIONS We have reviewed the information submitted by Florida Power and Light Company regarding their nuclear cross-section methodology for future licensing applications for the Turkey Point and St.

Lucie reactors.

The review was

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h based on (a) the description of the methodology and comparison to experiment and similar methodologies,,and (b) the results of a standard problem run by staff consultants.

The methodology is similar to other methodologies used for cross-section generation..

The results of calculations of integral parameters based on cross-sections derived from this methodology compared well with measured data as well as'omputations with other methodologies.

We find the proposed methodology acceptable.

However, the methodology must be used in conjunction with a diffusion or transport code, but no such code has been submitted for review.

4.0 REFERENCES

1.

Letter from J.

W. Williams Jr., Florida Power and Light Company, to S.

A. Varga, NRR, dated May 10, 1984 (L-86-125).

2.

Letter from C.'0.

Wood, Florida Power and Light Company, to J.

E.

Lyons, NRR, dated October 21, 1986 (L-86-432).

3.

Letter from J.

F.

Carew, BNL, to E. Knuckles, Florida Power and Light,.

dated August 8, 1986.

4.

Letter from E. Knuckles, Florida Power and Light Company, to L. Lois,

NRR, dated June 26, 1987.

5.

"CHEETAH-P" report of the nuclear fuel management and analysis

package, publication No. 84004100, Nuclear Associates I'nternational Corporation, July 1974.

6.

NAI-71-13, "NAI Modified LEOPARD," Nuclear Associates International Corporation, dated December 10, 1973.

7.

Bahl, H., et al.,

"MUFT-4, Fast Neutron Spectrum Code for the IBM-704,"

WAPD-TM-72, dated July 1957.

5 8.

Amster, H., et al.,

"The Calculation of Thermal Constants Averaged Over a Wigner-Wilkins Flux Spectrum:

Description of the SOFOCATE Code,."

WAPD-TM-39, dated January 1957.

9.

Amouyal, A., et al.,

"New Method of Determining the Thermal Utilization Factor in a Unit Cell," Journal of Nuclear Energy, Vol. 6, pp 79-98, 1957.

10.

Marlowe, 0. T., et al.,

"CANDLE-4 One-Dimensional Few-Group Depletion Code for the IBM-704," WAPD-TM-53, dated May 1957.

ll.

Strawbridge, L. E., et al., "Criticality Calculations in Uniform Water-Moderated Lattices," Nuclear Science and Engineering, Vol. 23, pp 58-73, September 1965.

12.

Callaway-l, Licensing Application for Expanded Spent Fuel Storage

Rack, Docket 50-483, (Table 9. 1A-3).

13.

Babcock and Wilcox, Standard Nuclear Steam

System, B-SAR-205, Vol. 2, pg 4. 3 (Table 4. 3-15).

Dated:

April 19, 1988 Princi al Contributor:

L. Lois