ML20072J756
| ML20072J756 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 08/23/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072J750 | List: |
| References | |
| NUDOCS 9408290161 | |
| Download: ML20072J756 (5) | |
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,%[v,f' W ASHINGTON, D.C. 2055 A001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 116TO FACILITY OPERATING LICENSE N0. NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY VIRGIL C. SUMMER NUCLEAR STATION. UNIT N0. 1 DOCKET N0. 50-395
1.0 INTRODUCTION
By letter dated December 13, 1993, as supplemented by letters dated February 2, 1994, and March 11, 1994, South Carolina Electric & Gas Company (SCE&G or the licensee) requested changes to the Virgil C. Summer Nuclear Station, Unit No. 1, (VCSNS or Summer) Technical Specifications (TS) to allow the use and subsequent storage of fuel initially enriched to 5 weight percent (w/o) Uranium 235 (U-235).
The March 11, 1994, letter provided clarifying information that did not change the initial determination of no significant hazards consideration as published in the FEDERAL REGISTER.
2.0 EVALUATION The 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, KEN 0 Va, using neutron cross sections generated by the AMPX code package from the 227 energy group ENOF/8-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, PH0ENIX, using a 25 energy group nuclear data library based on a modified version of the British WIMS cross section library.
These codes are widely used for the analysis of 4
fuel rack reactivity and have been benchmarked against results from numerous i
critical experiments.
These experiments simulate the VCSNS fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment, assembly spacing, and absorber thickness. The intercomparison between two independent methods of analysis (KENO-Sa and PH0ENIX) also provides an acceptable technique for validating calculational methods for nuclear criticality safety. To minimize the statistical uncertainty of the KENO-Sa reactivity calculations, a minimum of 60,000 neutron histories were accumulated in each calculation.
Experience has shown that this number of histories is quite sufficient to assure convergence of KEN 0 Va reactivity calculations.
The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the VCSNS storage racks with a high degree of confidence.
The spent fuel storage racks in Region I were reevaluated for 4.0 w/o U-235 enriched fuel based on the as-built boron-10 (B-10) loading in the Boraflex 9408290161 940823 PDR ADOCK 05000395 P
. panels suppl _ted by the Boraflex material vendor. The calculations were made for pure water moderator at 20' C with a density of 1.0 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, thickness, and length, and B-10 loading were accounted for as well as eccentric fuel positioning.
These uncertainties were appropriately determined at the 95/95 probability / confidence level.
In addition, calculational and methodology biases and uncertainties due to benchmarking, B-10 self shielding, and pool water temperature ranges were included as well as consideration of Boraflex gaps and shrinkage. The calculations assume that 75% of the Boraflex panels experience non-uniform shrinkage (random gaps) and the remaining 25% of the panels experience uniform shrinkage (pull-back) from the bottom.
Based on the results of recent blackness testing performed in the Region 1 racks, the staff concurs that these assumptions, in conjunction with the assumption of a 4%
width and length shrinkage, bound the current measured data and future development of additional shrinkage and gaps.
The final Region 1 design, when fully loaded with fuel enriched to 4.0 w/o U-235, resulted in a k of 0.9485 g
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 probbility/ confidence level and is, therefore, acceptable.
To enable the storage of fuel assemblies with nominal enrichments greater than 4.0 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) material coated on the outside of the U0a pellet.
The fuel assembly depletion calculations performed show that the maximum reactivity for rack geometry occurs at 0 burnup.
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 80 IFBA rods was found to be equivalent to the reactivity of the rack when filled with fuel assemblies enriched to 4.0 w/o and containing no IFBAs.
Since the worth of individual IFBA rods can change depending on position 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 10% on the total number of IFBA rods at the 5.0 w/o enrichment limit (i.e., about 8 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 reactivity point.
The PH0ENIX code was used for the fuel assembly k, calculations based on a unit assembly configuration in the VCSNS core geometry moderated by pure water at a temperature of 68 F with a density of 1.0 gm/cc.
A 1% reactivity bias was included to account for calculational uncertainties.
0
. Calculationsfor a fresh 4.0 w/o Westinghouse 17x17 0FA fuel assembly, which yields equivalent or bounding reactivity results relative to the other Westinghouse 17x17 fuel types, in VCSNS core geometry resulted in a reference k, of 1.460.
Since~~the fuel rack reactivity of a fresh 4.0 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 enrichment of 5.0 w/o U-235 with a maximum reference k, less than or equal to 1.460 at 68 F can be safely stored in the Region 1 racks.
The Region 2 spent fuel storage racks were reanalyzed for storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 2.5 w/o U-235. The same initial assumptions, biases, and uncertainties, as used for the Region 1 analyses, were included.
Since the blackness testing performed in the Region 2 racks did not indicate any gaps in any of the Boraflex panels inspected, no gaps were assumed in the analysis.
However, a 4% total width and length shrinkage was assumed in every Boraflex panel with the placement of the entire 4% of length shrinkage at the bottom of every panel.
Calculations performed for storage racks similar to VCSNS have indicated that positioning all of the Boraflex shrinkage at the bottom results in the most conservative k,.
The maximum k crTterion of 0.95.,,, for Region 2 is 0.9442, within the NRC acceptance To enable the storage of fuel assemblies initially enriched to greater than 2.5 w/o U-235, 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 are 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 3.9-1 in which a fresh 2.5 w/o enriched fuel assembly yields the same rack reactivity as an initially enriched 5.0 w/o assembly depleted to 21,600 MWD /MTV. This curve includes a reactivity uncertainty of 0.0072 due to depletion calculations.
Region 3 has been analyzed for the storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 1.4 w/o U-235. The same initial assumptions, biases and uncertainties used in the Region 2 analyses were also used for Region 3 except for the Boraflex related uncertainties.
The Region 3 racks do not contain Boraflex. The maximum k within the NRC acceptance criterion of 0.95.,,, for Region 3 is 0.9441, As for Region 2, burnup credit reactivity equivalencing was used to allow storage of fuel assemblies with initial enrichments greater than 1.4 w/o U-235.
Figure 3.9-2 shows that fresh 1.4 w/o enriched fuel is equivalent to initially enriched 5.0 w/o fuel which has achieved a burnup of 48,000 MWD /MTV.
The curve includes a reactivity uncertainty of 0.0160 due to depletion calculations.
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 heatup or cooldown events or the misloading of an assembly with a burnup and enrichment combination outside of the acceptable area in Figure 3.9-1 or 3.9-2, which l
J
, could lead to-an increase in reactivity.
For such events, credit may be taken for the presence of approximately 2000 ppm of boron in the pool water required during fuel handling operations since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (D tble Contingency Principle). The reduction in k,by credible accidents.
additioncauseY,causedbytheboron,morethanoffsetsthereactivity In fact, the licensee has determincd that only 400 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.
The new (fresh) fuel racks have been previously analyzed for storage of Westinghouse 17x17 fuel assemblies with enrichments up to 5.0 w/o U-235 and the criticality w 11ysis was included with this amendment request.
For the fully flooded condition, k, did not exceed 0.95, including appropriate allowances for biases and,un, certainties.
For the low density optimum moderation condition, k did not exceed 0.98.
Therefore, the criticality analyses of the fresh fYe#1 racks meet the applicable NRC criteria and are acceptable. No credit for IFBAs was included in these analyses. However, due to the restrictions required on spent fuel storage, the proposed TS changes require fuel assemblies with enrichments above 4.0 w/o U-235 to contain IFBAs such that the maximum reference fuel k, is no greater than 1.460 in unborated water at 68 F.
The following Technical Specification changes have been proposed as a result of the requested enrichment increase.
The staff finds that these changes are consistent with the above evaluation and, therefore, are acceptable.
(1)
Figure 3.9-1 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 gaps in Region 2 of the spent fuel pool.
(2)
Figure 3.9-2 has been revised to place restrictions on fuel burnup as a function of initial enrichment up to 5.0 w/o U-235 in Region 3 of the spent fuel pool.
Region 3 does not contain Boraflex.
(3)
TS 5.3.1 has been revised to permit reload fuel with a maximum enrichment of 5.0 w/o U-235 and to incorporate the recommendations described in NRC Generic Letter GL 90-02, Supplement 1.
It also requires fuel to contain sufficir u iFBAs in order to comply with the requirements of TS 5.6.1.1.a.
(4)
TS 5.6.1.1 has been revised to de seate the m tirements 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.
(5)
TS 5.6.1.2 has been revised to permit the storage of 5.0 w/o U-235 fuel in the new fuel storage racks, remove the reference to Section 4.3 of the FSAR, and to add the requirements for k.
o
. 1 Based on theyeview described above, the staff finds the criticality aspects of the proposed enrichment increase to the VCSNS new and spent fuel pool storage racks are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.
The staff concludes that Westinghouse 17x17 fuel from VCSNS may be safely stored in Region 1 of the spent fuel pool provided that the U-235 enrichment does not exceed 5.0 w/o and there are sufficient IFBAs such that the maximum reference fuel assembly k, does not exceed 1.460 at 68" F.
Any of these fuel assemblies may also be stored in Region 2 or 3 of the spent fuel pool provided it meets the burnup and enrichment limits specified in TS Figure 3.9-1 or 3.9-2, respectively.
Although the VCSNS TS have been modified to specify the above-mentioned fuel as acceptable for storage in the fresh or spent fuel racks, evaluations of reload core designs (using any enrichment) will be performed on 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.
Based on the foregoing evaluation, the staff finds the proposed changes acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of South Carolina official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an Environmental Assessment and Finding of No Significant Impo;t has been prepared and published in the Federal Reaister on August 15, 1994, (59 FR 41799). 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.
5.0 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 amendment vill not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
L. Kopp Date: August 23, 1994 l