ML17177A291
| ML17177A291 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities, LaSalle  |
| Issue date: | 02/27/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17177A290 | List: |
| References | |
| NUDOCS 9203040254 | |
| Download: ML17177A291 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL ON NEUTRONICS METHODS FOR BWR RELOAD DESIGN FOR CECO PLANTS COMMONWEALTH EDISON COMPANY DRESDEN UNITS 2 AND 3, QUAD CITIES UNITS 1 AND 2. LASALLE UNITS 1 AND 2 DOCKET NOS. 50-237. 50-249, 50-254, 50-265, 50-373. AND 50-374
1.0 BACKGROUND
By letter dated December 12, 1990, Commonwealth Edison Company (CECo) submitted the licensing topical titled, "Commonwealth Edison Company Topical Report - Benchmark of BWR Nuclear Design Methods," NFSR-0085.
Supplements 1 and 2 were submitted by letter dated May 8, 1991, and additional information was submitted on December 23, 1991.
The topical report summarizes the nuclear analysis methods to be employed by CECo.
These methods are based on the General Electric (GE) approved methodology.
The topical was submitted in support of reload nuclear design analyses for the Dresden, Quad Cities and LaSalle stations.
The supplements provide detailed comparisons to the Quad Cities Unit assembly and pin gamma scan results and they describe the means by which the neutronic methods will be used for the analysis of abnormal neutronic events.
The December 23, 1991, letter responds to NRC questions.
CECo has two fuel vendors for its three BWR stations; GE is the fuel supplier for Quad Cities and LaSalle stations and Advanced Nuclear Fuels (ANF) is the fuel supplier for Dresden station.
Due to the differences between the GE and ANF neutronic methodologies with respect to the Critical Power Correlation and the associated uncertainties, CECo does not intend to use the GE methods described in this report for neutronic licensing calculations for the Dresden Station. However, CECo plans to use the GE steady-state neutronic methods for Dresden fuel management and operational support analyses.
This report summarizes the nuclear analysis methods used by CECo in support of reload analysis for its BWR reactors and the benchmark data used to demonstrate CECo's capabili~y to independently perform the steady-state neutronic analysis needed for the steady-state licensing, operation, testing and surveillance of a BWR reload cycle. The methodology is based on GE proprietary codes TGBLA, for lattice physics calculations and PANACEA, for the three-dimensional (3D) core simulation. The methodology of this code package has previously been approved.
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The benchmarking analysis includes hot and cold critical eigenvalue data and a comparison of the predicted to measured Transverse lncore Probe (TIP}
readings. Supplement 1 presents CECo's gamma scan comparisons for both local fuel pins and assemblies. Supplement 2 presents comparisons to vendor results for the neutronic licensing activities, including calculation of the R value, SLCS shutdown margin, misoriented assembly, misloaded assembly, control rod drop accident, rod withdrawal error, and loss of feedwater heating transient.
2.0 EVALUATION The methodology which CECo proposes to use is a GE code package written for the VAX computer system.
It was installed, validated and verified on the CECo system.
CECo engineers underwent extensive training at the GE facilities.
This training included the performance of the full scope of neutronic calculations required for a reload design, and training in the acceptability and limitations of the computer programs for calculating neutronic parameters.
The methodology was used for the benchmark analysis, which was performed to demonstrate CECo's capability to independently perform the steady-state neutronic analysis portions of the reload design process.
I Since the methodology consisted of using computer codes which were previously approved, the review focused on the benchmarking data.
For the benchmarking of the critical eigenvalues and the TIP readings, results were from Quad Cities Units 1 and 2, Cycles 7 through 10, Dresden Unit 3, Cycles 8 through 11, and LaSalle Units 1 and 2, Cycles 1 through 3. This database includes fuel designs from 7X7 to 9X9 fuel pin arrays, various water rod configurations, axially dependent lattice designs throughout the enriched portion of the assembly and both GE and ANF fuel product lines.
Hot Critical Eigenvalues The hot critical eigenvalues as a function of cycle exposure were presented.
Hot critical eigenvalues should be consistent and predictable as a function of cycle exposure so that adequate projections of the critical eigenvalues can be made for the upcoming cycles.
The benchmark hot critical eigenvalues are a strong function of plant type and fuel product line but are consistent within these parameters.
Comparisons of the GE and CECo results for the hot critical eigenvalues for Quad Cities 1, Cycle 10, and LaSalle 2, Cycle 3, showed good agreement.
The mean and standard deviation of the two cycles are as follows:
Unit/Cycle L2C3 QlClO CECo Data Mean SD 0.9984 1.0017 0.0015 0.0010 GE Data Mean SD 0.9985 1.0023 0.0016 0.0011 Cold Critical Eigenvalues The cold critical eigenvalue predictions are presented in tabular form.
As with the hot critical eigenvalues, the cold critical eigenvalues must also be consistent and predictable as a function of exposure because they are required to calculate core subcriticality. The values presented are consistent and predictable as a function of plant type.
CECo and GE calculated cold critical eigenvalues at the beginning of cycle are as follows:
Unit/Cycle LIC5 QlClO CECo Cold Critical Eigenvalue 1.0026 1.0053 GE Cold Critical Eigenvalue.
1.0027 1.0045 These are typical values.
The maximum deviation was 0.0018 and the minimum deviation was 0.0001, with the absolute average of 0.00066.
TIP Results Measured and calculated Traversing Incore Probe {TIP) data have been compared and are summarized.
The TIP standard deviations were calculated over the entire axial length of the core, which is 24 nodes.
TIP standard deviations for the last two cycles of each plant are within the current design criteria of 10 percent nodally and 6 percent radially. If the standard deviations approach these values, they are beyond what has been historically seen and will be investigated.
Assembly Gamma Scans Assembly gamma scans measurements were taken at the Quad Cities Nuclear Power Station, Unit 1, at the end of Cycles 2, 4, and 5.
The power distributions from these measurements are compared to the PANACEA calculated power distributions to demonstrate that the bundle-by-bundle power predictions are correct. The results of these assembly gamma scans show the nodal standard deviation varying from 4.71 percent to 5.29 percent and the radial standard deviation varying from 2.42 percent to 4.24 percent. These values are comparable to those reported by GE for a similar database. Assembly integrated power comparisons for each of the three cycles were presented. These results showed better agreement between measurement and prediction than comparable results
. calculated by GE.
Pin Gamma Scans Pin gamma scan measurements were performed at Quad Cities Unit 1 at the end of Cycles 2, 3, and 4.
Seven assemblies were evaluated in this comparison.
These assemblies have an average enrichment of between 2.12 and 2.62 w/o U235, include 7X7 and BXS lattice arrays, and exposures between 6.1 and 15.8 GWd/STU.
Results of the pin gamma scan comparisons showed a standard 0
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e e deviation of 3.12 percent for all rods. However, the peak power rods had a standard deviation of only 2.6 percent. These results demonstrate excellent agreement of measured and predicted data.
Neutronic Licensing Calculations The application procedures which CECo uses in the evaluation of cycle-specific neutronic licensing events is the same as those described in the General Electric Document GESTAR II, which has previously been reviewed and approved by the NRC.
Comparisons of CECo's results to GE's results was made for various abnormal neutronic licensing events, including the calculation of the critical power ratio for these events.
The following cycle-specific neutronic licensing analyses were compared.
(1)
(2)
(3)
(4)
(5)
(6)
(7)
Shutdown Margin Standby Liquid Control System Fuel Loading Error - Misoriented Assembly Fuel Loading Error - Misloaded Assembly Control Rod Drop Accident Control Rod Withdrawal Error Loss of Feedwater Heating In all cases the difference between the CECo and GE results was minimal.
3.0 CONCLUSION
S CECo has demonstrated through the extensive benchmarking provided in Topical Report NFSR-0085~ "Benchmark of BWR Nuclear Design Methods" and the two supplements, capability to perform the neutronic analyses required for the steady-state licensing, operation, testing, and surveillance of a BWR reload cycle.
The overall neutronic design process used by CECo is the same as that described in approved GE documents.
Since the CECo benchmarking results are comparable to or better than the results obtained by GE, it is acceptable for CECo to use calculational uncertainties identical to those used by GE~ Thus, we find Topical Report NFSR-0085 acceptable for referencing for future reload cycles.
Principal Contributor:
M. Chatterton Dated:
February 27, 1992
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