ML20128N581
| ML20128N581 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/17/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20128N577 | List: |
| References | |
| NUDOCS 9302230303 | |
| Download: ML20128N581 (5) | |
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UNITE D STATES 0"
NUCLEAR REGULATORY COMMISSION j
j WASHINGTON, D. C. 20555
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SAFETY EVALVATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.170 TO FACILITY OPERATING LICENSE NO. OPR-50 METROPOLITAN EDISON. COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION IBREE MILE ISLAND NVCLEAR STATION. UNIT N0._1 DOCKET N0.: 50-289 1.0 JBTRODUCTION By letter dated August 25, 1992, GPU Nuclear Corporation (GPUN, the licensee) requested changes to the Three Mile Island, Unit No.1 (TMI-1) Technical Specifications (TS) to allow an increased limit for fuel enrichment. The proposed changes would allow for the storage of new fuel with an enrichment not to exceed a nominal 5.0 weight percent (w/o) U-235 in the TMI-1 new fuel storage vault and in Region I of spent fuel pool A.
The changes would also allow a slight increase in allowable maximum nominal enrichment of fuel to be stored in spent fuel pool B to 4.37 w/o V-235, 2.0 EVALVATION The analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed with the two-dimensional multi-group transport theory computer code, CASM0-3.
Independent verification calculations were also made with a Monte Carlo technique using the KENO-Sa computer package with the 27-group SCALE cross section library. KENO-Sa was also used for the new fuel storage vault analysis. These codes are widely used for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments. These experiments simulate the TMI-1 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-5a and CASMO-3) is 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 250,000 neutron histories in 500 generations of 500 neutrons each were accumulated in each calculation.
Experience has shown that this number of histories is quite sufficient to assure convergence of KENO-Sa reactivity calculations. The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the TMI-l storage racks with a high degree of confidence.
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For the new (unirradiated) fuel storage vault, the fuel assemblies'are stored in racks in parallel rows with a nominal center-_to-center _ distance of 21-1/8 inches in both directions..The new fuel storage vault.is' intended-for the-receipt and storage of fresh fuel under normally dry conditions where the-9 reactivity is very low due to the absence of neutron moderation. To assure criticality safety under accident conditions and to conform to the 1
requirements of General Design Criterion (GDC)'62, the NRC has two separate criteria for new fuel storage racks which must be satisfied.
- If fully loaded with fuel of the. highest anticipatedLreactivity and flooded with clean unborated water, the maximum reactivity, including uncertainties, shall not exceed -a k,,, of 0.95.
- If fully loaded with fuel of the highest anticipated reactivity and moderated by the optimum low density moderator ~(1.e., fog or foam), the maximum reactivity, including uncertainties, shall not exceed a k,,, of 0.98.
In order to meet these two criteria for fuel-enriched to a maximum'o'f 5.0 w/o U-235,. it is necessary to block-off and maintain two empty rows of storage locations. These twelve total-storage locations (aligned in.two rows. of six -
locations each) transverse rows number four'and eight..
The results of the reactivity calculations for the new fuel vault show that the limiting reactivity condition for low-density. optimum moderation occurs at-a hypothetical moderator density of 9%.. This resulted in a k of 0.9534 which included 'all appropriate uncertainties at the 95% probabTlity/95%
confidence' level (95/95 probability / confidence), thus' meeting -the NRC criterion of 0.98.
For the fully _ flooded. accident condition, the maximum k,, including 95/95 uncertainties, was 0.9487 which is within.the NRC-crYterionof_0.95and,therefore, acceptable.
The spent fuel storage racks in Pool: A were reevaluated for 5.0 w/o U-235:
enriched fuel based on the as-built boron-10' (B-10) ~1oading in the Boral-..
- panels (rather than the original; design loading)..The calculationsLwer made-at 20 C which. corresponds:to the highest-possible reactivity. -For the nominal storage: cell design in Region:I, uncertainties due to boroniloading tolerances, boral width tolerances, tolerances inicel1J1attice spacing,.
stainless steel thickness tolerances, and fuel enrichment and density' tolerances were accounted for.as well _ as -eccentric-- fuel. positioning.- These-uncertainties were appropriately determined at the 95/95 probability /-
- confidence level. _ In addition, a calculational _ bias and uncertainty were determined-from benchmark calculations. The final ' Region I: design,- when fully:
loaded with' fuel enriched to 5.0 w/o:U-235, resulted in a-_k of 0.9470 when ~
combined with' all known uncertainties. This meets the staffs criterion-of k,,, no greater,than 0.95 including all uncertainties'at the 95/95-probability / confidence-level and is, therefore, acceptable.
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. Since the original design calculations for Region II in Pool A had already extended the analysis to include 5.0-w/o U-235 enriched fuel, little change resulted from the revised calculations at-the lower-(more reactive) water temperature and the use of the as-built B-10 loading in the Boral poison panels. Therefore, the combination of initial enrichment and cumulative burnup for spent fuel storage in Region II, as given-in TS Figure 5-4, remain essentially the same.
The storage racks in Pool B are unpoisoned and use a stainless steel and water flux trap between cells as a means of controlling reactivity.
These racks are-designed to accommodate fuel of various initial enrichments which have accumulated sufficient minimum burnups. The reactivity analysis for these racks show that the staff's 0.95 k limit is met for fresh (unirradiated)-
fuel of 4.37 w/o enrichment. CalcuYationstodeterminetherequireddischarge fuel burnup of 5.0 w/o initially enriched fuel showed that a burnup of only 2.48 MWD /KgU was adequate for unrestricted storage in the Pool B racks. These minimum burnup requirements are given in TS Figure 5-5.
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.the misloading of an assembly with a burnup and enrichment combination outside of
-the acceptable area in Figure 5-4 or 5-5 or dropping an assembly between the pool wall and the fuel _ racks, which could lead to an increase in reactivity.
However, for such events credit'may be taken for the presence of approximately 600 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 (Double Contingency Principle).
The reduction in k caused by the boron more than offsets the reactivity addition caused by c,r,e,dible accidents.
The following TS changes have been proposed as a result of the requested:
enrichment increase.
(1) TS 5.3.1,6 indicates the increased limit for fuel enrichment to' a nominal 5.0 w/o U-235 for new reload fuel assemblies and rods.
(2) TS 5.4.1.a indicates that nominal 5.0 w/o U-235 fuel can be stored in the New Fuel Storage Vault or Spent Fuel Pool _A, Region I storage locations without exceeding a k,,, of 0.95. -It-also indicates that Region 7 of Spent Fuel Pool A can accept fresh fuel with low enrichment and Spent Fuel Pool B can accept fresh fuel up to 4.37 w/o U-235 enrichment withoutl exceeding a k,,, of 0.95.
(3) TS 5.4.2 identifies the acceptability.of storing high enrichment burned fuel in Spent Fuel Pool A, Region II and Spent Fuel Pool B locations'as restricted by Figures 5-4 and 5-5, respectively. The number of storage locations in the Dry New Fuel Storage Area is decreased by 12 to; reflect the number of locations required to be vacant of fissile.or. moderating material. Section f-specifies a new value for the maximum allowable
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grams of U-235 per axial centimeter of fuel. assembly to support the increase to 5.0 w/o U-235 new fuel.
Figure 5-4 reflects changes resulting from a reduction in the design temperature to 4* C and the use of as-built B-10 loading in the Boral panels.
Figure 5-5 depicts the acceptable burnup domain for determining the acceptability for storage of high enriched spent fuel in Pool B based on the revised calculations described previously.
The staff finds these changes acceptable.
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SUMMARY
Based on the review described above, the staff finds the criticality. aspects of the proposed enrichment increase to the THI-l new and spent fuel pool storage racks are acceptable and meet the requirements of GDC 62 for the prevention of criticality in fuel storage and handling.
The staff concludes that fuel from THI-l may be safely stored in Region I of Pool A prefided that the U-235 enrichment does not exceed 5.0 w/o. Any of these fuel assemblies may also be stored in Region II of Pool A or in Pool B provided it meets the burnup and enrichment limits specified in TS Figure 5-4 or 5-5, respectively.
Although the THI-l 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, of course, 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.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment. 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 l
finding of no significant impact have been prepared-and published in the Federal Reaister on December 2, 1992 (57 FR 57079). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.
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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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
L. Kopp s
Date:
February 17, 1993 m
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