ML20078C315
| ML20078C315 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/20/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078C304 | List: |
| References | |
| NUDOCS 9501260259 | |
| Download: ML20078C315 (5) | |
Text
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NUCLEAR REGULATORY COMMISSION 2
WASHINGTON. D.C. *=A *1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 68 TO FACILITY OPERATING LICENSE NO. NPF-37,
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AMENDMENT NO. 68 TO FACILITY OPERATING LICENSE NO. NPF-66, AMENDMENT NO. 58 TO FACILITY OPERATING LICENSE NO. NPF-72, i
AND AMENDMENT NO. 58 TO FACILITY OPERATING LICENSE NO. NPF-77 COMMONWEALTH EDIS0N COMPANY BYRON STATION. UNIT NOS. 1 AND 2 BRAIDWOOD STATION. UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454. STN 50-455. STN 50-456 AND STN 50-457
1.0 INTRODUCTION
By letter dated November 7,1994, Commonwealth Edison Company (Comed, the licensee) requested changes to the technical specifications (TS) for Byron i
Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, to permit the use of higher enriched fuel and specifies the spent fuel storage recuirements for Regions 1 and 2 of the spent fuel pools. The staff requested acditional information on December 6,1994, which was provided by the licensee by letter dated December 16, 1994.
The December 16, 1994, submittal provided additional clarifying information that did not change the initial proposed no significant hazards consideration determination.
1 The proposed changes would allow for the storage of fuel with enrichment not to exceed a nominal 5.0 weight percent (w/o) U-235 in the spent fuel storage racks.
An enrichment manufacturing tolerance of-10.05 percent U-235 about the nominal-value was incorporated into the analysis.
The new fuel storage vaults for Byron and Braidwood Stations were previously approved for storage of fresh fuel _ enriched up to 5.0 percent and, consequently, were not addressed in this evaluation.
2.0 fyALUATION The licensee's analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed with the three-dimensional multi-group Monte Carlo computer code, KENO Va, using neutron cross sections generated by the AMPX code package from the 227 energy group ENDF/B-V data library.
Since the KEN 0 Va code package does not have depletion capability, burnup analyses were performed with the two-dimensional transport theory code, PHOENIX, using a 42 energy group nuclear data library.
PH0ENIX was also used to determine the reactivity effects of material and manufacturing tolerances.
These codes are P
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. widely used for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments. These experiments 1
simulate the Byron and Braidwood fuel storage racks with respect to parameters important to reactivity such as enrichment, assemb13 spacing, and absorber worth. The intercomparison between two independent cnethods of analysis (KENO Va and PHOENIX) also provides an acceptable techniqa for validating calculational. methods for nuclear criticality s afety. To minimize the statistical uncertainty of the KENO Va reactivity calculations, a minimum of j
60,000 neutron histories were typically accumulated in each calculation.
Experience has shown that this number of histories is quite sufficient to assure convergence of KENO Va reactivity calculations. The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the Byron and Braidwood spent fuel storage racks with a high degree of confidence.
The licensee reevaluated the spent fuel storage racks in Region 1 for 4.2 w/o U-235 enriched fuel moderated by pure water at 20 degrees Celsius with a density of 1.0 grams per cubic centimeter (gm/cc).
For the nominal. storage cell design in Region 1, uncertainties due to tolerances in fuel enrichment and density, fuel pellet dishing, storage cell I.D., cell lattice spacing, stainless steel thickness, Boraflex width and thickness, and boron-10 (B-10) loading were accounted for as well as eccentric fuel positioning. These uncertainties were appropriately determined at the 95/95 probability /confidencelevel.
In addition, calculational and methodology biases and uncertainties due to benchmarking, B-10 self shielding, and pool water temperature ranges-were included. The reactivity calculations also i
considered the effects of Boraflex shrinkage and gap formation. All Boraflex panels were modeled with four percent width shrinkage.
In addition, five axial shrinkage / gap scenarios were evaluated which covered the spectrum of shrinkage-to-gap ratios from 100 percent gaps and zero percent shrinkage through zero percent gaps and 100 percent shrinkage. Maximum length shrinkage was assumed to be four percent. Based on the results of blackness testing performed at other storage facilities,'and on upper bound values recommended by Electric Power Research Institute (EPRI), the staff concurs that these assumptions bound the current measured data and future development of additional shrinkage and gaps. Due to the presence of Boral inserts (described in Amendment No. 25 and Amendment No. 20 for Byron and Braidwood, respectively), and added to the flux trap gaps of the Region I racks prior to their initial installation, the worst case effects of Boraflex shrinkage and gap formation resulted in no increase in k,,, hen fully loaded with fuel for the Region 1 storage configuration. The final Region 1 design, w enriched to 4.2 w/o U-235, resulted in a k,,, of 0.9389 when combined with all known uncertainties. This meets the staff s criterion of.k no greater than 0.95 including all uncertainties at the 95/95 probability /c,o,n,fidence level and is, therefore, acceptable.
i To enable the storage of fuel assemblies with nominal enrichments greater than 4.2 w/o U-235, the concept of reactivity equivalencing was used.
In this technique, which has been previously approved by the NRC, credit is taken for the reactivity decrease due to the integral fuel burnable absorber (IFBA)
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> l material coated on the outside of the U0, pellet. Based on these calculations, the reactivity of the fuel rack array, when filled with fuel assemblies enriched to 5.0 w/o U-235 with each containing 64 IFBA rods, was found to be equivalent to the reactivity of the rack when filled with fuel assemblies enriched to 4.2 w/o and containing no IFBAs.
Since the worth of individual IFBA rods can change depending on position 4
within the assemblies due to local variations in thermal neutron flux, the licensee has included a conservative reactivity margin to assure that the IFBA requirement remains valid at intermediate enrichments where standard IFBA patterns may not be available.
In addition, to account for calculational uncertainties, the IFBA requirements also include a conservatism of approximately ten percent on the total number of IFBA rods at the 5.0 w/o enrichment limit (i.e., about 6 extra IFBA rods for a 5.0 w/o fuel assembly).
The staff concludes that sufficient conservatism has been incorporated to bound the calculational assumption that the IFBA requirements were based on the standard IFBA patterns used by Westinghouse.
As an alternative method for determining the acceptability of fuel storage in Region 1, the infinite multiplication factor, k, is used as a reference i
reactivity point. The PHOENIX code was used for the fuel assembly k, l
calculations based on a unit assembly configuration in the Byron and Braidwood core geometry moderated by pure water at a temperature of 20 degrees Celsius with a density of 1.0 gm/cc. A one percent reactivity bias was included to account for calculational uncertainties. Calculations for a fresh 4.2 w/o Westinghouse 17x17 0FA fuel assembly, which yields equivalent or bounding reactivity results relative to the other Westinghouse 17x17 fuel types, in the core geometry resulted in a reference k, of 1.470. Since the fuel rack j
reactivity of a fresh 4.2 w/o assembly is less than 0.95 and has been shown to be equivalent to a 5.0 w/o assembly with the standard number of IFBA rods, an assembly of maximum nominal enrichmer,t of 5.0 w/o U-235 with a maximum reference k, less than or equal to 1.470 at 20 degrees Celsius can be safely i
stored in the Region I racks.
The Region 2 spent fuel storage racks were reanalyzed for storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 1.6 w/o U-235. The same initial assumptions, biases and uncertainties as used for the Region 1 analyses were included.
In addition, the effects of Boraflex l
shrinkage and gaps were accounted for by assuming that 50 percent of the panels experience non-uniform shrinkage (random gaps) and the remaining 50 l
percent of the panels experience uniform shrinkage (pullback) from the bottom end. Although calculations have shown that positioning all of the Boraflex shrinkage at the bottom results in the most conservative k ', the staff considers this an overly conservative estimate and concurs,'that the assumption used is more realistic and still conservative since, based on Blackness test results, not all panels would be expected to undergo a four percent total shrinkage at one end. The maximum k,f, for Region 2 is 0.9449; within the NRC acceptance criterion of 0.95.
o-o To enable the storage of fuel assemblies initially enriched to greater than 1.6 w/o U-235 in Region 2, the concept of burnup credit reactivity equivalencing was used. This is predicated upon the reactivity decrease associated with fuel depletion and has been previously accepted by the staff for spent fuel storage analysis. For burnup credit, a series of reactivity calculations is performed to generate a set of initial enrichment-fuel assembly discharge burnup ordered pairs which all yield an equivalent k,,,
less than 0.95 when stored in the spent fuel storage racks. This is shown in Figure 5.6-1 of the TSs in which a fresh 1.6 w/o enriched fuel assembly yields the same rack reactivity as an initially enriched 5.0 w/o assembly depleted to 46,442 megawatt days per metric ton of uranium (MWD /MTU). This curve includes a three percent penalty factor to account for the uncertainty in measured burnup for each individual assembly.
Most abnormal storage conditions will not result in an increase in the k,,, of the racks. However, it is possible to postulate events, such as cooldown events or the misloading of an assembly with a burnup and enrichment combination outside of the acceptable area in Figure 5.6-1, which could lead to an increase in reactivity. However, for such events credit may be taken for the presence of approximately 2000 parts per million (ppm) of boron in the pool water (refueling canal) as required by TS 3.9.1 during fuel handling operations and by plant procedures during all other times 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 caused by the boron more than offsets the reactivity addition caused by c,r,e,dible accidents.
In fact, the licensee has determined that only 300 ppm of boron is necessary to mitigate the worst postulated accident in any pool region. Therefore, the staff criterion of k,,, no greater than 0.95 for any postulated accident is met.
3.0 TECHNICAL SPECIFICATION REVISIONS The following Technical Specification changes have been proposed as a result of the requested enrichment increase. The staff finds these changes acceptable.
(1) Figure 5.6-1 has been revised to place restrictions on fuel burnup as a function of initial enrichment up to 5.0 w/o U-235 and to account for the effects of Boraflex panel shrinkage and oaps in Region 2 of the spent fuel pool.
(2) TS 5.6.1.1 has been revised to clearly delineate the requirements for each region of the spent fuel pool, to add the new minimum burnups as a function of initial enrichment, to permit the storage of 5.0 w/o U-235 fuel, and to add the requirements for k,,.
3.1
SUMMARY
Based on the review previously described, the staff finds the criticality aspects of the proposed enrichment increase to the Byron and Braidwood spent
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a cycle-by-cycle basis as part of the reload safety evaluation process.
Each reload design is evaluated to confirm that the cycle core design adheres to the limits that exist in the accident analyses and TS to ensure that reactor operation is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact has been prepared and published in the Federal Reaister on January 20, 1995 (60 FR 4200).
Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant-effect on the quality of the human environment.
6.0 CONCLVSlDH 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: January 20, 1995 h
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