ML20011F146
| ML20011F146 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/23/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20011F144 | List: |
| References | |
| NUDOCS 9003010306 | |
| Download: ML20011F146 (7) | |
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NUCLEAR REGULATORY COMMISSION o
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, WASHINGTON, D. C. 20086
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' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR' REGULATION MLLATED TO AMLNUMtHI N05. 126 AND 130 TU FACIETTV OPERATING LICENSE N05. DPR-24 AND DPR-27 WISCONSIN ELECTRIC POWER COMPANY.
POINT BEACH NUCLEAR PLANT, UNIT Nos. I AND 2 D_0CKET N05. 50-266 AND 50-301 1.0 INTRO,DUCTIO!i By letter dated July 6,(1988 as supplemented November 1, 1989, Wisconsin l
Electric Power Company the licensee) applied for amendments to Facility l'
Operating Licenses DPR-24 and DPR-27 of the Point Beach Nuclear Plant to revise the maximum fuel enrichment limit specified in Technical Specification (TS)15.5.4.2. The application also modified the wording of the TS to
' permit the use of axial fuel-blankets, r
l Facility Operating Licenses DPR-24 and DPR-27 for the Point Beach Nuclear Plant were amended on October 5, 1984 (Amendments 86 and 90, respectively) to include the use of 0FA fuel with an allowable U-235 content limited to 39.4 grams per axial centimeter-(equivalent to 4.0 weight percent). The effect of the new 0FA-fuel on criticality, spent fuel cooling requirements, and' radiological consequences as well as the gama heating effects were evaluated and found to be acceptable. The facility operating licenses were amended again on April 14,'1989 (Amendments 117 and 120, respectively),for g
raising the allowable U-235 content per axial centimeter to 40.0 grams 0FA fuel.
g 2.0 EVALUATION L
The licensee requested that the U-235 loading limit'specified in TS 15.5.4.2 L
for fuel storage in the new fuel storage vault and the spent fuel storage pool be revised to 46.8 grams per axial centimeter for optimized fuel assemblies (OFA)andthattheuseofaxialfuelblanketsbepermitted. The U-235 loading level is not a parameter that is considered in accident analyses for operations of the Point Beach Nuclear Plant.
Rather, the safety margins are maintained by other limits such as power, power distribution, reactivity coefficients, burnup, etc., which are verified to be acceptable by cycle-specific reload analyses.
2.1 FUEL STORAGE
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The water in the spent fuel storage pool at the Point Beach Nuclear Plant contains at least 1800 parts per million of boron. The spent fuel storage racks consist of an array of individual square cross-section, rectangular cylinders with an outer dimension of approximately 10 inches and a length of about 14 feet. The racks are designed so that it is impossible to store fuel 9003o10306 900223 94
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assemblies within the racks in other than a valid storage position, thereby ensuringthenecessaryspacingbetweentheassemblies. The rectangular cylinders are edge-welded to 'orm a honeycomb structure. Two poison assenblies are. inserted into contiguous poison compartments formed inside each of the storage cylinders.
Each poison assembly supports two Boraflex sandwiches. Each sandwich consists of a F raflex sheet between two stainless steel plates. There are two Boraflex see
..ies between any two stored fuel assemblies in the spent fuel storage rat.
l Criticality analyses of _the Point Beach new and. spent fuel storage racks for storage of Westinghouse 0FA fuel were originally performed for the licensee by Pickard, Lowe and Garrick, Inc. (PLG), in support of their application to i
use OFA fuel at the Point Beach Nuclear Plant. The NRC issued amendments approving the use and storage of 0FA fuel with a U-235 content limited to 39.4 grams per axial centimeter as noted above.
The original PLG analyses utilized the LEOPARD and PDQ-7 computer programs.
LEOPARD and PDQ-7 calculational accuracies were verified by means of benchmark comparisons with critical assembly experiments, and conservative techniques were used for the determination of the infinite neutron multiplication factor, k-effective. The calculations were performed with the same methods that had been used and approved previously for the new and spent fuel racks at Point Beach.- The. licensee has used the same methodology to determine k-effective for higher enrichment fuel.
For 0FA fuel with a U-235 content of 4.75 weight percent, the maximum k-effective for the spent fuel racks was calculated to be 0.9406, including all biases and uncertainties. This maximum value of k-effective is acceptable because it is less than the 0.95 regulatory limit. Furthermore, it should be noted that all neutron multipli-cation factors were calculated assuming a-spent fuel pool filled with unborated water.
1 The licensee addressed the effects of higher enrichment fuel and increased burnup on spent fuel pool cooling and ganna heating. The original 0FA analysis provided by the licensee in support of license amendments 86 and 90 to Facility Operating Licenses DPR-24 and DPR-27, respectively, quantified those effects for spent fuel with an initial enrichment of up-to 4.0 weight percent U-235. The licensee has completed further analysis which demonstrates conservative changes in the those parameters of interest for the more highly enriched fuel at increased burnup. As a result, the licensee concluded that the original analysis completed for the OFA fuel bounds the expected effects resulting from the proposed increase in allowable U-235 content for OFA fuel.
t The new fuel storage vault is designed to hold new fuel assemblies in specially constructed dry storage racks. The center-to-center spacing of the assemblies in the new fuel storage vault is 20 inches.
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Similar analyses were performed for the new fuel ~ storage vault-in the 1,
flooded state. For the new fuel storage vault, all calculations were performed using new Westinghouse OFA fuel with an enrichment of 5.5 weight percent. The use of 5.5 weight percent fuel is conservative.
In addition to the baseline case for a flooded cavity, calculations were performed for elevated temperatures and for mist conditions with water densities ranging-from 3 to 80 percent of maximum water density. The LEOPARD /PDQ model bias for these cases is 0.0071 delta-k and the calculational uncertainty is 0.0029 delta-k.
In addition, the increase in k-effective due to fuel position uncertainties and due to the maximum fuel pellet' density were taken into account. The combination of these biases and uncertainties yielded a maximum expected value of k-effective for the new fuel storage racks of 0.9221 for the fully flooded case. This value of k-effective is acceptable because it is less than the 0.95 regulatory limit.
The licensee )resently has no plans to utilize axial fuel blankets at Point Beach but wisies to have tne operational flexibility to do so. The licensee-has confirmed the staff's. understanding that its use of an optional axial-zoned core-loading scheme refers to the use of a single enrichment throughout the active portion of the fuel assembly with axial blankets of natural (non-enriched) fuel above and below the active portion of the fuel assembly. The storage of such fuel in either the new fuel vault or the spent fuel storage pool is not a concern since the' replacement of the more highly enriched fuel with non-enriched fuel at the top and bottom will further reduce the neutron multiplication factor, k-effective. As a result, the use of 0FA fuel with axial fuel blankets is acceptable.
2.2 DESIGN BASIS FUEL HANDLING ACCIDENT i
L The licensee has requested authorization to increase fuel enrichment to 4.75 weight percent of U-235 with an expected average fuel burnup of 45,000 t
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MWD /MT. Both the staff and the licensee have evaluated the potential impact l
of this change on the radiological assessment of design basis accidents l
(DBAs) which were previously analyzed in the licensing of the Point Beach Units 1 and 2 nuclear power plants.
The staff reviewed the licensee's submittals and also reviewed a publication which was prepared for the NRC entitled, " Assessment of the Use of Extended Burnup Fuel in Light Water Reactors," NUREG/CR 5009, February 1988. The NRC contractor, the Pacific Northwest Laboratory (PNL) of Battelle Memorial Institute, examined the changes that could result in the NRC design basis l
accident (DBA) assumptions, as described in appropriate sections of the L
staff's Standard Review Plan and/or Regulatory Guides, that could result from the storage and handling of extended burnup fuel, which would require higher initial enrichment. The staff agrees with PNL that the only DBA that could be affected by the use of extended burnup fuel, in even a minor way, l
would be the potential thyroid doses that could result from a postulated design basis fuel handling accident.
PNL estimates that I-131 fuel gap activity in the peak fuel rod with 60,000 MWD /MT burnup (5.292 initial weightpercentU-235)couldbeashighas12%. This value is approximately
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,.4 20% higher than the value normally used by the staff in evaluating fuel handling accidents (Regulatory Guide 1.25 " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the
! Fuel Handling and Storage Facilities for Boiling and Pressurized Water Reactors").
'The staff, therefore, reevaluated the fuel handling accident for Point Beach Units-1 and 2 with an increase in iodine gap activity in the fuel damaged in a postulated fuel handling accident. Table 1 presents the conservatively analyzed fuel handling accident doses presented in the d
staff's Safety Evaluation dated April 4, 1tr79 related to Amendments 35 and 41.to-license Nos. DPR-24 and DPR-27, respectively. These estimates were based on a presumed fuel peaking. factor of 1.65.
The licensee's Final Safety Analysis Report (FSAR), updated through October 1,1988, assumes a peaking factor of 1.8 (p.14.2.1-11). The current staff dose estimates result from multiplying its previous estimates by 1.2 x 1.8/1.65 = 1.31, to account for these two factors. The results are also shown in Table 1.
The resulting doses are small fractions of the applicable regulatory requirements of 10 CFR Part 100.
The staff concludes that the bounding doses potentially increased are the thyroid doses at the Exclusion-Area and Low Population Zone boundaries' resulting from postulated fuel handling accidents, that-these doses remain well within the 300 rem thyroid exposure guideline values set forth in 10 CFR Part 100, that the small calculated increase is not significant and that the Technical Specification change requested by the licensee is acceptable.
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-2.3 OTHER CONSIDERATIONS On April 4, 1979, the NRC staff issued a safety evaluation report approving the surveillance program for Boraflex to be used as a neutron attenuation l
material'in the spent fuel storage pools at the Point Beach Nuclear Plant.
The surveillance program would verify the continued integrity of the Boraflex i
material in the spent fuel storage racks. On February 11, 1987, the licensee l-reported to the NRC the results of the Boraflex examinations performed in 1985 and 1986. Ten Boraflex coupons were found to have significant decreases in sample-thickness, width, and weight, and were fragile. Nevertheless, the neutron attenuation capability of these coupons was not significantly reduced.
Further, the full length Boraflex inserts examined showed no degradation other than some-discoloration along the edges of the irradiated insert.
As a result of the observed degradations in the Boraflex coupons, the licensee proposed to replace the existing Boraflex surveillance procedure with a new surveillance program. On April 13, 1989, the licensee submitted for NRC approval i
- a. revised proposal to modify the Boraflex surveillance program. Further, by letter dated November 1, 1989, the licensee described corrective actions to be i
taken upon finding degraded Boraflex to ensure subcriticality of'the stored fuel assemblies. On February 21, 1990, the NRC staff issued a safety evaluation i
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5-i approving the revised Boraflex surveillance program. The safety evaluation i
concluded that the proposed surveillance program was sufficient to detect deteriorationofthephysicalintegrityoftheBoraflexthatmig'eofthe ht lead to the loss of neutron attenuation capability during the design li l
spent fuel storage racks. Further, the staff concluded that should significant I
loss of Boraflex neutron attenuation be found, the licensee can take corrective actions to ensure suberiticality of the stored fuel assemblies.
On November 21,1989, the NRC issued Bulletin 89-03, " Potential Loss e' Required Shutdown Margin During Refueling Operations." The bulletin noted o
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that as a result of the longer fuel operating cycles, utilities have been increasing the enrichment od reload fuel. Some of these fresh reload fuel assemblies may be highly reactive under certain refueling conditions. Although L
analyses are performed for PWRs to confirm that the refueling boron concentra-tion is sufficient to maintain the required shutdown margin for the final core configuration, the staff was concerned that these analyses may not be s.
sufficient to assure that the shutdown margin will be maintained for all intermediate fuel assembly positions.
The licensee responded to Bulletin 89-03 with a letter dated January 11, 1990.
In their response, the licensee indicated that actions, procedures, and training necessary to comply with the actions requested by Bulletin l
89-03 would be completed prior to shutdown for the Point Beach Unit 1 l
refueling outage. This outage is currently scheduled to begin on March 30, l
1990.
By letter dated February 15, 1990, the NRC staff notified the licensee that its response was satisfactory.
3.0 FINDINGS The NRC staff has reviewed the request by the Wisconsin Electric Power Company j
toincreasetheallowableU235contentforoptimizedfuelassemblies(OFA) to 46.8 trams por axial centimeter and to permit the use of axial fuel blankcts. Based on this review, the staff has concluded that the storage and use of such fuel at the Point Beach Nuclear Plant is acceptable and that the Technical Specifications submitted for this license amendment accurately reflect the modifications necessary to accommodate future fuel reloads.
Therefore, the proposed changes to TS 15.5.4.2 are acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35< an environmental assessment and finding of no significant impact has been w epared and published in the Federal Register on February 23,1990(55ER6564). Accordingly, based upon the environmental assessment, the Comission has determined that the issuance of these amendments will not have a significant effect on the quality of the human environment.
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f 5.0 CONCLUS10N The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safet not be endangered by operation in the proposed menner, y of the public willand(2)su will be conducted in compliance with the Commission's regulations, and 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 Contributors:
J. Martin W. Swenson Dated: February 23, 1990
Attachment:
Table 1 d
Table 1 i
Estinated Doses as a Consequence of DBA fuel Handling Accident Exclusion Area Boundary
- Thyroid Whole Body April 4, 1979 NRC Staff Estimate 36 rem 0.13 rem Current NRC Staff Estimate 47 rem 0.17 rem 10 CFR Part 100 300 rem 25 rem
- Low Population Zone doses are less than Exclusion Area Boundary doses l-i l
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