ML20196G687
| ML20196G687 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/23/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19318E104 | List: |
| References | |
| NUDOCS 8807060004 | |
| Download: ML20196G687 (10) | |
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\+..../ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO FUEL ENRICHMENT TECHNICAL SPECIFICATION CHANGE LICENSE NO. DPR-36 MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATOMIC POWER PLANT DOCKET NO. 50-309
1.0 INTRODUCTION
By letter of March 24, 1988 (Ref. 1) Maine Yankee Atomic Powee Company (MYC) requested a revision to the Technical Specifications (TS) for Maine Yankee (MY). The TS change (to Specification 1.3.A, "Reactor Core") relates to a proposed increase, to 3.7 percent, in the U-235 enrichment limit of fuel which may be placed in the MY reactor core. The priraary discussion in the submittal does not relate to characteristics of the core when using fuel with the higher enrichment, but rather to the criticality analysis of new and spent fuel storage areas containing fuel with the increased limit (or higher) enrichment. The request was accompanied by a topical report prepared by Yankee Atomic Electric Company (YAEC), YAEC-1637 (Ref. 2), describing calculations and providing results of the reactivity status of the MY storage areas with increased fuel enrichment levels. This submittal was preceded (February I?, 1988, Reference 3) by the submittal of another YAEC report.
YAEC-1622 (Ref. 4), describing the methodology used for the criticality analysis and providing its qualification and associated uncertainty factors.
This review is concerned primarily with these subjects of fuel storage criticality.
gRo7060004 880623 p ADOCK 05000309 PDR
2 A notice of consiceration of issuance for this amendment was published in the Federal Regir,er on April 29, 1988 (53 FR 15479). No request for hearing or petition fcr leave to ' intervene was filed.
The review of the request involves two areas: review of the safety of plant operation with the higher enrichment fuel and review of the safety of storage of the higher enrichment fuel.
The submittal and review are not directed at the effects of higher enrichment on operating core characteristics (e.g., peak power density) because such characteristics are not determined directly by peak enrichment (as long as they are in a reasonable range). They are rather determined by the interactions among a ,qumber of core component parameters and arrangements.
These include the specific patterns used for fuel loadings, burnable poisons, burnup and operational reactivity and power distribution control. There is no evident abstract limit on enrichment which can be directly related to core operating limits. Operating extremes of the fuel will depend on the many interacting core components and modes of operation. These are provided in the cycle reload design and analysis. The 3.7 percent enrichment limit requested by MYC is not unusually high, and is well within bounds (in some cases over 4.0 percent) which have been approved for other reactors. The licensee has indicated that it is expected that the Performance Analysis Report for the next MY cycle (11) design and subsequent cycle designs will show that the facility
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3 will meet required operating limits, even if.using 3.7 percent enriched new fuel assemblies. This will be confirmed or changes identified by the required reload analysis report.
On the above basis, it can be concluded that the increased fuel enrichment can be approved as far as core characteristics effects are concerned, provided all applicable acceptance criteria are met for Cycle 11 operation and that no other changes in any Technical Specifications are involved. If such changes are involved, the normal reload amendment request and NRC review process will assure that there will be adequate operational safety.
The remaining significant review area is the enrichment effect on meeting required criticality criteria in new and spent fuel storage areas.
The MY spent fuel storage area was reracked in 1985 and the criticality analysis supporting the licensing of these racks (Ref. 5) justified a maximum stored fuel enrichment of 3.5 percent. The new submittal presents a justification far in excess of 3.7 percent (up to 4.13 percent for some of the spent fuel storage area). There are no new racks involved in this increase, but some detail changes at the time of installation of the racks, and after the Reference 5 design analysis, has improved the reactivity margin to required limits compared to the previous analysis and has provided the incentive for the proposed increases in allowed enrichment. The racks are
4 arrays'~of unit cells with the fuel within steel boxes and outer steel "wrappers," with boral sheets between box and wrapper. There is a water gap between cells. The principal change has involved flooding the cell boral regions, providing greater flux trap neutron attenuation effects.
The 1985 spent fuel pool reracking involved two phases and three types of storage cell details, differentiated by the boral material used and thus by the boron effective thickness av.d water gap thickness. The Phase I racks have all type 1 cells and the Phase II racks (adjacent to the Phase I racks) have some of each type. The most limiting (highest k-effective) cell type is type 3 and the analysis presented for Phase II racks assumes all type 3 cells. The cell type difference leads to different proposed limits of 4.13 and 3.72 percent enrichment for the Phase I and Phase II spent fuel racks respectively.
The new fuel racks have also been analyzed for increased enrichment and the submittal indicates that fuel with enrichment up to 5.5 percent can be accommodated within the required reactivity criteria.
2.0 Methodolooy Oualification Evaluation The methodology used for the analysis of the MY new and spent fuel racks is briefly described in Reference 2 and more generally in Reference 4 which also provides the methodology qualification and the associated bias and uncertainty analyses and resulting values to be used in the design calculations.
5 Twc types of methods are used. Monte Carlo calculations are done with the NITAWL-S/ KENO-Va cross section and neutronics code package, and the transport theory or transport-diffusion theory calculation is done with the CASM0-3 or CASM0/ CHART /PDQ code packages. Both of these types of methods as well as the specific individual code components, including both the cross sections and the analysis methods, are industry standards, commonly used in storage criticality analyses. A wide background exists for their use, and staff approval has been given for numerous storage analyses using the same or similar methodologies.
Thus the selection of methods for the various MY calculations is appropriate.
Several cross section sets are available for these methodologies. YAEC has chosen to provide in YAEC-1622, qualification calculations for both a 123 group and a (more recently derived) 27 group set of cross sections for the Monte Carlo calculations. Both sets are widely used for such analyses. Only the 123 group set was used for the MY rack calculations, however. For the transport-diffusion qualifications both a 70 and a 40 group set were examined, but only the 40 group set was used for MY. These sets, and the selection for MY are reasonable and acceptable.
The qualification process has involved calculating a series of Babcock & Wilcox critical experiments designed to be relevant to high density storage configurations. These experiments are commonly used for methodolooy cualification and provide an acceptable basis for method uncertainty quantification. The review has indicatem that the experiments were
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6 appropriately modeled and the statistical analyses of the results followed suitable, standard procedures to provide calculational biases and uncertainties. Results are within expected ranges. It is concluded that acceptable qualification and associated uncertainty values have been provided for the YAEC use of NITAWL-S/ KENO-Va (with 123 or 27 groups) and CASM0/ CHART /PD0 (with 70 or 40 groups) for calculating the criticality of fuel storage systems.
3.0 STORAGE ANALYSIS EVALUATION The NRC storage criterion (see NUREG-0800, NRC Standard Review Plan, Sections 9.1.1 and 9.1.2) for spent fuel pools (in ficoded conditions) is 0.95, and for new fuel racks (i.e., dry storage) is 0.95 flooded and 0.98 with optimum low density moderation (e.g., from fire fighting sprays). These limits are interpreted as including all relevant uncertainties at a 95/95 (probability /
confidence) level, No credit is given for boron in the pool in meeting these limits. Credit may be taken for boron in the pool for analyses of accident conditions (e.g., dropped fuel assembly), however, and such accident events are not considered in combination with optimum moderator conditions in new fuel (dry) storage. The YAEC analyses for the proposed enrichment increase for MY follow these NRC criteria and positions and are acceptable.
For the analyses of the spent fuel racks the YAEC calculations used KEN 0 (123)
7 as the reference methodology. The details of the rack unit cell with a fresh (representative MY design) fuel assembly are modeled in quarter cell geometry with reflecting boundary conditions. CASMO(40) calculations,doneasan additional consistency check, were modeled as a (diagonal) half cell with reflective boundaries. CASM0/PDQ calculations were not needed for the MY analyses. Calculations were done for Type 1 and Type 3 cells, representing, as previously explained, the Phase I and II racks, respectively. A series of calculations were done (with both methods) as a function of fuel enrichment.
A comparison of the KENO and CASM0 results indicate that they are in agreement within the 1-sigma level of the Monte Carlo statistical uncertainty. This provides assurance of the correctness of the reference KEN 0 calculations.
CASM0 was used for mechanical and composition variation sensitivity calculations to provide uncertainty values to be combined with the methodology uncertainties. The analysis of this physical uncertaint, gave a typical 95/95 value of about 0.0063 delta-k.
The relevant method bias and uncertainty values are given in Reference 4 for KEN 0(123). The delta-k values are 0.00395 and (at a 95/95 level) 0.01245 respectively. For the number of gencrations used in the Monte Carlo calculation the statistical uncertainty is about 0.0041. These all combine for a typical total uncertainty of about 0.019 delta-k. This value (as well as the components) is reasonable and falls within expected ranges and is acceptable.
8 Combining the uncertainty-bias results with the KEN 0 calculated k for type 1 and 3 cells as a function of enriunment of the stored assemblies leads to a value for an enrichnent limit for the storage of current typical MY type fuel assemblies in the Phase I and,II rucks. These are 4.13 and 3.72 percent, respectively.
Aanormal or accident conditions, e.g., fuel assemblies outside but adjacent to a rack, were analyzed in the original MY rerack submittal (Ref. 5). The NRC gives credit for the boron in the pool water for these types of events. This boron provides a very large subcritical margin, which is not significantly affected by the changes in enrichment involved in these present analyses.
There is, therefore, no effect on the above limits from abnormal conditions.
It is thus concluded that the 4.13 and 3.72 percent enrichment values for the Phase I and II racks are acceptable limits for MY fuel assembly storage and that therefore, acceptable spent fuel storage is cvailable for the 3.7 percent enriched fuel of the proposed "Reactor Core" TS change. It is noted that, with the TS limit at 3.7 percent and presumably therefore, no greater than 3.7 percent fuel available, either the Phase I or II rackr may be used for fuel storage. However, if in the future greater than 3.72 percent fuel is to be stored, information on the procedurcs used to assure that the higher enrichment is stored only in Phase I racks should be presented for staff review.
The new fuel racks analyses were also done using KEN 0 (123) for the reference
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analysis. Because of the large shutdown margin available for these racks CASMO check calculations were unnecessary. Since the new fuel racks are not compact arrays, modeling involved a four cell mockup-incorporating the cell and subunit separation spaces involved in the rack geometry. Since for this dry storage arrangement the NRC requires analysis as a function of a full range of assumed water density in and between cells to find an "optimum" reactivity, the calculations covered this range as well as a range of enrichment values. These calculations determined an optimum density of about 5 percent, and a maximum enrichment of over 5.5 percent to stay under the NRC requireo low density optimum k-effective limit of 0.98. The density study also indicated that at full density (optimum high derisity) the k-effective rould be less than 0.95. These values included the relevant 95/95 uncertainty and bias values as previously discussed for the spent fuel pool. It is therefore acceptable to store typical current MY type fuel assemblies with enrichments up to 5.5 percent. This is, of course, compatible with the storage of 3.7 percent fuel corresponding to the proposed TS change.
4.0 CONCLUSION
S MYC has proposed a chan;2 for MY TS 1.3.A providing for an increase to 3.7 percent in the fuel assembly U-235 enrichment permitted to be used in the reactor core. The concerns related to the use of this fuel in the core are minimal and are addressed as a normal part of the analyses done for the next and future cycle reloads. The primary review area here is capability of the fuel storage areas to meet NRC reactivity criteria when containing fuel of this enrichment, or higher. The review has concluded that the new fuel racks can contain up to 5.5 percent enriched fuel of current MY design and the spent fuel racks 4.13 and 3.7 percent for Phase I and 11 respectively, within the required reactivity limits. The proposed TS change is in accordance with staff positions and previous TS approvals. Our review has concluded that appropriate material has been submitted and the TS change is acceptable.
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-The staff 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, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 REFERENCES
- 1. Letter from J. Randazza, MYC, to Director NRR, dated March 24, 1988, "Technical Specification Proposed Change 138: Fuel Enrichment Limit."
- 2. _YAEC-1637, "Criticality Analysis of Maine Yankee's Spent Fuel Pool and New Fuel Value," February 1988.
- 3. Letter from G. Whittier, MYC, to NRC, dated February 12, 1988, "Fuel Storage Criticality Analysis Methodology."
- 4. YAEC-1622, "Validation of the YAEC Criticality Safety Methodology,"
January 1988.
- 5. YAEC-1285, "Maine Yankee Spent Fuel Rack Modification - Criticality Analysis," March 1983.
Principal Contributor: H. Ritchings Date: June 23, 1988