ML20206C912
| ML20206C912 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/27/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20206C910 | List: |
| References | |
| NUDOCS 9905030247 | |
| Download: ML20206C912 (5) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR RE RELATED TO AMENDMENT NO.175 TO FACILITY OPERATING LICE FLORlDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302
1.0 INTRODUCTION
By letter dated October 30,1998, and supplemented by [[letter::3F0399-12, Forwards Response to NRC 990225 Telcon RAI Re Util LAR 245 Re Methodology Change for Boraflex Degradation.Ltr Establishes No New Regulatory Commitments|letter dated March 31,1999]], Florida Power Corporation (FPC) requested an amendment to the licensing basis for Crystal River Unit 3 (CR-3) to change the methodology for the Spent Fuel Pool B criticality analysis. Th proposed change is necessary due to Boraflex degradation in the Spent Fuel Pool B storag racks. The criticality effects of the proposed change to the licensing basis as well as the proposed changes to the Final Safety Analysis Report (FSAR) and the associated improved Technical Specification (ITS) Bases were included with the above submittal. The March 31 1999, supplement provided clarifying information and did not affect the original no significan hazards consideration determination.
2.0 EVALUATION CR 3 has two spent fuel pools designated as the "A" and "B" pools, which are physically jo together through a transfer canal. The A Spent Fuel Pool has high density storage rack modules which do not utilize Boraflex. The B Spent Fuel Pool has eight high density racks which are constructed with Boraflex. Fuel storage is divided into two Regions within the B pool. Region 1 was designed to accommodate new (fresh) fuel assemblies or fuel which has not experienced sufficient burnup to be stored in Region 2. Region 2 was designed to accommodate irradiated fuel, determined by bumup calculations. The Region 1 racks have a double layer of Boraflex panels within each cell with a one-inch water gap between each cell.
The Region 2 racks have only a single layer of Boraflex.
Boraflex is known to degrade under the influence of gamma radiation and chemical reaction with free radicals in the pool water. Over the first few years of use, the Boraflex will shrink, typically creating gaps distributed randomly in the axial direction. As the gamma dose increases, the Boraflex panels will slowly begin to deteriorate, losing the neutron absorbing component B.C. The proposed amendment request is intended to determine the potential effect of Boraflex degradation in Pool B on criticality safety. It is also intended to update the analyses, incorporating the more modem and improved methodologies that have become available in the last few years, and to confirm configurations for acceptable storage of fuel with enrichments up to 5.010.05 weight percent (w/o) U-235.
The analysis of the reactivity effects of fuel storage in the CR-3 spent fuel racks was performed primarily with the three-dimensional NITAWL-KENO 5a Monte Carlo code packag 9905030247 990427 Y
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NITAWL was used with the 238-group SCALE-4.3 cross section library and the Nordheim integral treatment for U 238 resonance shielding effects. Verification calculations were made with the MCNP4A Monte Carlo code. Since the KENO-Va code package does not have f
burnup capability, depletion analyses and the determination of small reactivity increments due J
to manufacturing tolerances were made with the two-dimensional transport theory code.
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CASMO4. The SCALE-4.3 system used in the reactivity analysis has been benchmarked agsinst experimental data for fuel assemblies similar to those for which the CR-3 racks are designed and has been found to adequately reproduce the critical values. This experimental data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include close proximity storage and strong neutron absorbers. The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the CR-3 storage racks with a high degree of confidence.
U.S. Nuclear Regulatory Commission (NRC) General Design Criterion (GDC) 62 of Appendix A to 10 CFR 50 requires the prevention of criticality in fuel storage and handling. The NRC acceptance criterion for preventing criticality in spent fuel storage areas is that, including uncertainties, there is a 95% probability at a 95% confidence level (95/95 probability / confidence) that the effective neutron multiplication factor (k,y) of the fuel assembly array will be no greater than 0.95.
For the nominal storage cell design, the racks were assumed to contain the most reactive fuel authorized to be stored without any control rods or bumable poison. These are the Babcock &
Wilcox 15x15 Mark B-10F and Mark B-10 fuel. The moderator was assumed to be pure water at a temperature within the design basis range corresponding to the highest reactivity. No credit was taken for radial neutron leakage or for neutron absorption in minor structural members. Uncertainties due to tolerances in U-235 enrichment and density, boron loading, Boraflex panel width, water gap (Region 1), cell box inner diameter or lattice pitch (Region 2),
and stainless steel thickness were accounted for as well as a method bias and uncertainty.
These uncertainties were appropriately determined at least at the 95/95 probability / confidence level. In addition, an allowance of 5% of the reactivity decrement from beginning of life to the burnup of interest was included for uncertainty in depletion calculations for those cases where burnup credit is used. These biases and uncertainties meet the previously stated NRC requirements and are, therefore, acceptable.
In the Pool B calculations, additional assumptions are made to consider the increase in reactivity due to Boraflex gapping. Although several design enhancements and measures were integrated into the CR-3 fuel racks to minimize Boraflex gap formation, the analysis assumes the presence of a random axial distribution of 4-inch gaps in all Boraflex panels. This is an acceptable conservative assumption based on existing industry-wide test results.
The analysis also assumed a concurrent loss of up to 20% of the Boraflex (8,C). In response to NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks," the licensee has stated that pool silica levels indicate some Boraflex degradation due to water ingress may be occurring. Boraflex degradation in the Pool B racks was projected using a i
calculated degradation rate based on the worst case weight loss of measured Boraflex l
samples. The current worst case calculations project that the Boraflex in the Pool B racks will have degraded to the point of 20% loss of neutron absorption in the year 2019. The estimated current (March 1999) weight loss using the same degradation rate is 5.3%. Although the i
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, Boraflex weight loss consists of several constituents in addition to boron, the analysis conservatively assumes that the entire weight loss is attributable to boron. Therefore, the 20%
loss assumed in the current analysis is acceptable.
Region 1 of Pool B is designed to accommodate a checkerboard pattern of fresh 5.0 w/o U-235 fuelintermixed with fuel of various initial enrichment vs. bumup combinations as specified in CR-3 Technical Specification (TS) Figure 3.7.15-2. Region 2 of Fool B is designed for fuel of various initial enrichment vs. bumup combinations as shown in TS Figure 3.7.15-3.
The licensee's analysis using the acceptable methods discussed above has shown that the bumup/ enrichment curves in the CR-3 TSs have sufficient margin to accommodate up to a 20% loss in Boraflex concurrent with a random distribution of 4-inch gaps and still maintain Pool B at less than or equal to 0.95k,when fully loaded and flooded with unborated water. In addition, Region 2 was evaluated with a 3-out-of 4 loading pattem. The results indicate a significantly greater reactivity margin available for this configuration to accommodate more reactive fuel (lower burnup) or greater Boraflex degradation than currently assumed.
Most abnormal storage conditions will not result in an increase in the k, of the spent fuel storage racks. However, it is possible to postulate events, such as the inadvertent misloading of an assembly in the spent fuel storage racks with a bumup and enrichment combination outside of the acceptable areas in Figures 3.7.15-2 or 3.7.15-3, which could lead to an increase in reactivity. The largest reactivity increase was caused by the inadvertent loading of a fresh Mark B-10F assembly enriched to 5.0 w/o U-235 into a fully loaded rack. For this condition, credit may be taken for the presence of 1925 ppm of soluble boron in the pool water, which is assured by TS 3.7.14, since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (Double Contingency Principle) The reduction in k,, due to only 350 ppm of boron offsets the reactivity addition caused by any credible accident.
The staff has reviewed changes to the following portions of the FSAR and the TS Bases.
Based on the above evaluation, the staff finds these changes acceptable.
- 1) FSAR Section 9.3.2,6.1, " Spent Fuel Pools Supplemental Cooling"
- 2) FSAR Section 9.6.1.2.2, " Spent Fuel Storage"
- 3) FSAR Section 9.6.2.4, " Safety Provisions"
- 4) FSAR Table 9-14," Fuel Storage Racks Suberiticality Margin-5.0% Enrichment"
- 6) ITS Bases B 3.7.15," Spent Fuel Assembly Storage"
3.0 STATE CONSULTATION
Based upon a letter dated March 8,1991, from Mary E. Clark of the State of Florida, Department of Health and Rehabilitative Services, to Deborah A. Miller, Licensing Assistant, U.Si NRC, the State of Florida does not desire notification of issuance of license amendments.
4.0 ENVIRONMENTAL CONSIDERATION
S The amendment changes requirements 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
F 4-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 this amendment involves no r,ignificant hazards consideration and there has been no public comment on such finding i
j (63 FR 71966). Accordingly, the amendment meets the eligibility criteria for categorical i
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 amendment.
5.0 CONCLUSION
Based on the review described above, the staff finds the criticality aspects of the proposed i
license amendment for CR-3 are acceptable and meet the requirements of General Design l
Criterion 62 for the prevention of criticality in fuel storage and handling. The revised FSAR Sections and ITS Bases changes correctly reflect the results of the new criticality analysis and are acceptable. The staff concludes 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 j
activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public, t
l Principal Contributor: Larry Kopp l
Date: April 27,1999 l
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3 l-Mr. John Paul Cowan CRYSTAL RIVER UNIT NO. 3 i
Florida Power Corporation cc:
Mr. R. Alexander Glenn Ms. Sherry L. Bernhoft, Director Corporate Counsel Nuclear Regulatory Affairs (SA2A)
Florida Power Corporation Florida Power Corporation MAC-A5A Crystal River Energy Complex P.O. Box 14042 15760 W. Power Line Street St. Petersburg, Florida 337.33-4042 Crystal River, Fiorida 34428-6708 Mr. Charles G. Pardee, Director Senior Resident inspector Nuclear Plant Operations (NA2C)
Crystal River Unit 3 i
. Florida Power Corporation U.S. Nuclear Regulatory Commission Crystal River Energy Complex 6745 N. Tallahassee Road 15760 W. Power Line Street Crystal River, Florida 34428 Crystal River, Florida 34428-6708 Mr. Gregory H. Halnon Mr. Michael A. Schoppman Director, Quality Programs (SA2C)
Framatome Technologies Inc.
Florida Power Corporation 1700 Rockville Pike, Suite 525 Crystal River Energy Complex Rockville, Maryland 20852 15760 W. Power Line Street Crystal River, Florida 34428-6708 Mr. William A. Passetti, Chief Department of Health Bureau of Radiation Control 2020 Capital Circlel, SE, Bin #C21 Tallahassee, Florida 32399-1741 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 Mr. Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 Chairman Board of County Commissioners Citrus County 110 North Apopka Avenue inverness, Florida 34450-4245 i
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