ML20126H877
| ML20126H877 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/22/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126H856 | List: |
| References | |
| NUDOCS 9301050347 | |
| Download: ML20126H877 (4) | |
Text
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'n UNITED STATES 3.
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1 NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATIOS BY THE Offl(LOf ILE. LEAR _ REACTOR RLQULATION MLATED 10_ AMENDMENT NO.176 TO FACILITY OPERAllflq_ LICENSE N0. DPR-53 AND AMENDMENT NO.15310 FACILITY OPERATING LICENSE NO. DPR-67 DALTIMORE GAS AND ELECTRL(lQliPANY CALVERT CllffS NUCLEAR EDWER PLANT. UNIT NOS. 1 AND 2 DD_CKET N05, 50-317 AND 50-318 1.0 11[1RQQEll0h By letter dated September 1, 1 ')2, as supplemented November 11, 1992, the Baltimore Gas and Electric Company (BC&E, the licensee) submitted a request for changes to the Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2, Technical Specifications (TS).
The requested changes would revise the Unit Nos. I and 2 spent fuel pool enrichment limit.
The spent fuel pool enrichment limit would be decreased from 5.0 weight percent (w/o) V-235 to a value of 4,52 w/t l' 235.
The change is being requested because of errors identified in the prer s calculations performed by Asea Brown Boveri-Combustion Enginer s (ABB-CE) to support the 5.0 w/o U-235 enrichment limit.
BG&E imposeo.uministrative limits on the maximum allowable enrichment, which were based on analysis performed when the errors were identified, until the requested decrease in the TS enrichment limit to 4.52 w/o U-235 is issued.
The November 11, 1992, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
The information provided supported the use of Pacific Northwest Laboratory (PNL) critical experiments which BG&E used to qualify the analytical methods used and BG&E's evaluation of the calculational uncertainty and bias.
2.0 EACKGR0VND The Unit I and Unit 2 spent fuel pool is common for both units.
The storage racks located in the common pool are identical except for the different poison (neutron absorber) material used in each.
The Unit I racks use a poison material made of a boron carbide composite which is not susceptible to shrinkage and associated gap formation.
The Unit 2 racks use Boraflex as the neutron poison material.
In the past, gaps have been observed in the Boraflex, in some cases, when the material is physically restricted and shrinks under irradiation. On September 8, 1987, the NRC issued Information Notice No. 87-43 alerting all operating licensees to this prcblem.
In early 1992, ABB/CE informed the NRC of errors in the spent fuel pool criticality calculations performed for several plants including Calvert 9301050347 921222 PDR ADOCK 05000317 P
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, Cliffs.
This prompted the issuance of NRC Information Notice No. 92-21 and Its Supplement.
BG&E had used these incorrect calculations to support a maximum allowed enrichment of 5.0 w/o U-235 for fuel in the Calvert Cliffs spent fuel pool.
3.0 [VALVATION l
Part of the discrepancy in the arevious ABB/CE spent fuel pool reactivity calculation was attributed to tie buckling term used in the CEPAK code spectral calculation to obtain four neutron energy group cross sections.
A geometric buckling term corresponding to a sparsely populated and unpoisoned array was used as an approximation of buckling in the poisoned configuration.
Although this approximation gave good agreement when applied to critical-experiments of both unpoisoned and lightly poisoned arrays, it is not appropriate for the specific configuration found in the Calvert Cliffs spent fuel racks where the assembly pitch is small and the fuel assembly is completely surrounded by a strong poison.
With this configuration, the buckling caused the fine group spectrum to be shifted to the thermal energies such that the effective broad group thermal removal cross section was overestimated, resulting in an underestimate of the effective multiplication CEPAK was,g). In the revised analysis, the geometric buckling supplied to factor (k derived from a transport theory solution for a fuel assembly in the storage rack environment.
The calculation was pet formed with the two-dimensional discrete ordinates transport code 00T-IV.
The staff concludes that the geometric buckling calculated in this manner is indicative of the neutronic environment of the fuel assembly in the spent fuel rack and is, therefore, acceptable.
The other discrepancy in the previous spent fuel 3001 calculations was attributed tr. the omission of self-shielding in t1e epithermal (group 3 of 4) poison absorption cross section.
CEPAK performs a one-dimensional thermal calculation (group 4) but a zero-dimensional fast and epithermal calculation i
(groups 1 through 3).
Consequently, no spatial self-shielding of the fast and epithermal cross sections are performed explicitly by CEPAK but must be performed by ancillary codes and input to CEPAK, if needed.
Comparisons to explicit one-dimensional calculations for both thermal and fast neutron energies performed by the XSDRNPH code indicated that the group 3 poison cross section is significantly self-shielded and that the omission of self-shielding in the original poison cross sections resulted in an overestimate of poison worth (and underestimate of k,g) lysis were generated by a 123-group XSDRNPM by about 2%.
The group dependent poison cross sections in the revised ana calculation and collapsed to a broad four-group scheme.
The staff concludes that the resulting set of four-group poison cross sections properly account for epithermal self-shielding and are acceptable.
The manufacturer of the Calvert Cliffs storage racks-has indicated that there were no manufacturing directives which would have led to constraint of the Boraflex sheets during fabrication.
Therefore, the Boraflex sheets'would be less likely to form gaps upon shrinkage. However, a gap penalty has been applied to account for the possibility that they may exist.
Four-inch gap's-
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- t were assumed to exist in every sheet of Boraflex and the gaps in the four walls of any given rack cell were assumed to be axially aligned.
Based on blackness tests performed on Boraflex panels at other spent fuel storage pools, the assumption of a 4-inch gap size in the Calvert Cliffs analysis appears to be suitably conservative.
In the original Unit 2 analysis, it was very conservatively assumed that Boraflex gaas of 4-ihches were located ir, all box walls at the midplane of the fuel.
In t,1e revised analysis, the 4-inch gaps in adjacent rack cells were assumed to be staggered slightly, with a 2-inch vertical separation which leads to an alternative (checkerboard-type) pattern when looking at the entire pool. This is more conservative than assuming a random gap distribution which industry experience indicates is the more likely case.
The gaps were assumed to be distributed 3referentially around the axial centerline of the fuel assembly.
Since tie flux is highest at the fuel axial centerline, the worth of the gaps will be the greatest in this region.
Therefore, the calculation of the gap penalty at the central fuel region in the reanalysis is conservative and acceptable.
The two-dimensional 00T-IV trans) ort theory code with cross sections generated by CEPAK was used to determine tie rack k and the three-dimensional Monte Carlo code KENO IV with the AMPX system f,,r, cross section generation was used o
to determine the reactivity penalty associated with the assumed Boraflex gap distributions.
These codes are widely used in the nuclear industry and have been benchmarked against experimental data and have been found to adequately _
reproduce the critical values.
In addition, the CEPAK-DOT methodology has been shown to produce k values which are-in good agreement with the values produced by the AMPX-KEI/d methodology over a variation of boron poison loadings ranging from critical experiments to the actual Calvert Cliffs spent-fuel rack.
The intercomparison between different analytical methods is an acceptable technique for validating calculational methods for nuclear criticality safety.
The staff, therefore, finds the use of these codes acceptable.
The analysis for the Unit 2 racks assumed fuel assemblies with an enrichment of 4.30 w/o U-235 and produced a nominal k,,,ll thickness, and Boraflex gaps, of 0.92308. Uncertainties and penalties due to temperature, cell pitch, wa as well as a calculational uncertainty and methodology bias resulted in a final k,,the licensee has determined a derivative of enrichment with ak,,, o interesl,of 0.93494.
Over the relatively small enrichment range of -
0.1464 w/o enrichment per % Ak,,,.
Therefore, the maximum allowable enrichment which maintains k,,, no greater than 0.95 is:
4,30 + (0.95 - 0.93494)
- 100
- 0.1464 4.52 w/o.
Since the only calculational difference between the rack ' design for Unit 1 and Unit 2 is the penalty associated with'the Boraflex gapping for the Unit 2 racks, the use of the lower calculated enrichment limit for the Unit 2 racks for the entire pool is bounding and acceptable..Therefore, the staff has l
4-determined that the proposed enrichment limit of 4.52 w/o U-235 is acceptable for the Calvert Cliffs spent fuel pool.
4.0 STATE CONSULTATION
j In accordance with the iommission's regulations, the Maryland State official was notified of the proposed issuance of the amendments.
The State official had no comments.
5.0 ENVIRONM QTAL CONSIDERATION The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR
- Part 20.
The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is-no significant _ increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (57 FR 45075).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded,- based on the considerations discussed above, that:
(1)-there is reasonable assurance that the health and safety of_the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
L. Kopp Date: December 22,_1992 l
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