ML20057A614
| ML20057A614 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/03/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20057A602 | List: |
| References | |
| NUDOCS 9309150040 | |
| Download: ML20057A614 (7) | |
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UNITED STATES 5(
AE NUCLEAR REGULATORY COMMISSION
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W ASHINGTON, D.C. 2055AWM
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. im AND ini TO 3
FACILITY OPERATING LICENSE N05. DPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NOS. 1 AND 2 DOCKET N05. 50-282 AND 50-306
1.0 INTRODUCTION
By letter dated June 11, 1993, as revised June 30, 1993, the Northern States Power Company (NSP or the licensee) requested amendments to the Technical Specifications (TS) appended to Facility Operating License Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant, Unit Nos. I and 2.
Theproposedchangeswouldincreasethemaximumfuelenrichgentspecifiedin 2
2 TS 5.3. A.2 from 4.25 weight percent (w/o) U " to 5.0 w/o U Due to this proposed change, TS 5.6. A and 3.8.E regarding new and spent fuel storage have been modified to accommodate the higher enriched fuel. Technicai Specification table TS 4.1-2B and figures TS 3.8-1, TS 5.6-1 and TS 5.6-2 are being added or modified, as necessary, to update the surveillance and storage requirements. Additionally, proposed TS 3.3. A.I.a and TS 5.3. A.] would increase the minimum refueling water storage tank (RWST) boron concentration from 1950 ppm to 2500 ppm and allow the use of ZIRLO clad and natural uranium.
The staff's safety evaluation of the proposed changes follows.
Bases for proposed TS 3.8.E.1 and 3.8.E.2 are being incorporated into the fuel handling specification bases.
2.0 fvALUAT10N 2.1 Reactor Core Chanaes Prairie Island was previously licensed for the use of Zircaloy-4 clad material and slightly enriched uranium. TS 5.3.A.1 has been modified to allow the use of ZlRLO cladding material and natural uranium. The licensee states that the use of ZlRLO and natural uranium will be consistent with paragraph 4.2.1 of the Westinghouse Standard Technical Specifications, NUREG-1431.
Furthermore, the licensee states that the requested change allowing the use of natural uranium is merely a clarification because natural uranium has been previously used as an axial blanket and replacement fuel pins.
9309150040 930903 PDR ADOCK 05000282 P
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2 The ZIRLO clad material has evolved from Zircaloy-4 as a material that has better corrosion resistance and less radiation induced growth than its predecessor. ZIRLO has very similar neutronic properties to Zircaloy-4 and, therefore, should not significantly perturb tne spectrum from the previously analyzed state.
Furthermore, the licensee states that "...(the) use of ZlRLO clad fuel in the reactor cores will be evaluated with NRC approved methodologies prior to use." The staff considers the use of ZlRLO acceptable up to burnups of 60 GWD/MTV. This restriction will remain valid pending approval of extended burnup fuel cycles using Westinghouse VANTAGE + fuel (ref. 1).
The proposed use of natural uranium is acceptable because reload core designs are evaluated for each cycle. The results of these analyses will be compared to the limits in the safety analyses and the TS to ensure that operation in the proposed fashion is acceptable.
2.2 RWST Boron Concentration Increase The licensee has proposed, by modifying TS 3.3.A.I.a, increasing the minimum RWST boron concentration from 1950 ppm to 2500 ppm in anticipation that this change will be required due to an increase in fuel enrichment.
Increasing the minimum RWST minimum boron concentration to 2500 ppm will provide adequate negative rer.ctivity to ensure that the reactor will remain subcritical following a loss of cpglant accident (LOCA) for reload cores utilizing fuel enriched to 5.0 w/o V The licensee states that this will not cause any change to the current plant procedures because the RWST concentration is already administrative 1y controlled at concentrations higher than 2500 ppm.
The licensee states in its submittal that it will have to confirm that the 2500 ppm value is acceptable during the appropriate reload analysis.
If the reload analysis concludes that the results are not consistent with the increase to 2500 ppm, then the licensee will need to submit a separate license amendment request. The staff finds this change to be acceptable.
2.3 fuel Enrichment Increase ThelicenseehasmodifiedTSs5.6.Aand3.8.Ejpordertoallowforthe storage of fuel up to enrichments of 5.0 w/o U To accommodate the higher enriched fuel, two fuel storage configurations have been proposed; one for new fuel and one for spent fuel. Westinghouse analyzed the proposed fuel storage configurations using the AMPX (using ENDF/B-V) code package (ref. 2) for cross section development and the KENO-Va Monte Carlo Transport Theory code (ref. 2) for reactivity calculations. The NITAWL (pa-t of the AMPX package) program (ref. 2) began with 227 group cross sections and developed the geometrically specific energy self-shielded cross sections.
Energy and spatial weighing was performed using the XSDRNPM (part of the AMPX package) S n transport theory code (ref. 2). These codes are widely used for criticality analyses and have been bench-marked against the results from numerous critical experiments which closely simulate the Prairie Island spent fuel racks. All analyses were performed at the required 95/95 probability / confidence level.
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TSs 5.6.A and 3.8.E have been modified to update the fuel storage requirements
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to accommodate the 5.0 w/o fuel. The proposed changes will require the 14 central locations in the new fuel storage area to be blocked and mandate a two-region, checkerboard loading spent fuel storage configuration.
Spent fuel pool region I refers to the restricted, checkerboard loading area and spent fuel pool region 2 refers to the unrestricted area. Unrestricted when used in this context refers to fuel which can be placed in any configuration.
Regional requirements will be prescribed by a combination of original enrichment and burnup.
Figures TS 3.8-1 and TS 5.6-2 will be used to determine whether a fuel assembly will be considered restricted and, if so determined, whether it must be stored in a high burnup or low burnup location.
Figure TS 5.6-1 shows the checkerboard loading pattern proposed and used in the Westinghouse analyses. The rows between the restricted and unrestricted regions must either be filled with alternating unrestricted and restricted burned fuel or be left empty.
Analyses of the proposed new fuel storage rack were performed with a KENO-Va model using the following assumptions:
1.
Limiting assumptions about fuel type, density, and manufacturing were made.
For example, no credit is taken for natural or slightly enriched i
uranium axial blankets.
2.
No credit was taken for non-fuel neutron absorgi,ng matg'ials, i.e., rack structure, burnable and soluble poisons, and U and U or other fission i
products.
3.
Lateral (X and Y) dimensions were infinite. The Z dimension was assumed infinite for the full density moderation calculation and was explicitly modeled for the optimal density moderation case.
The results of the analyses demonstrate that the proposed new fuel storage configuration meets the acceptance criteria which require that k be less than 0.95 for full density conditions and less than 0.98 for optNal moderation conditions. The calculated k for normal conditions was 0.9312, g
and 0.9690 for optimal moderation conditions. The staff has concluded that the proposed changes are acceptable because the referenced analysis is very conservative, significantly bounding and it meets the acceptance criteria.
Analyses of the proposed spent fuel storage rack, region 1, were also performed using the KENO-Va code. The following assumptions were made:
1.
Limiting assumptions about fuel type (the analysis used 9FA assemblies),
density, and manufacturing were made. Once again, no credit was taken for axial blankets.
2.
Limiting assumptions about pool moderator and structural conpitions were made.
For example, the moderator was assumed to be 1.0 g/cm density w:ter at 68'F (20*C) and Boraflex loading was assumed at a minimum.
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Nocreditwp5stakenfornon-fuelneutronpoisonssuchasrackstructural material, U ' and U
, and soluble poisons.
4.
The rack was assumed infinite in the lateral (X and Y) dimensions and finite in the axial (Z) direction. A reflector of clean, full density water was assumed on the top and bottom of the stored fuel.
In order to account for burnup, a technique called reactivity equivalencing was used. This technique consists of plotting constant k,,, contours on a burnup vs. initial enrichment plot. This technique offers two advantages.
One, it implicitly considers the effect that the rack has on reactivity.
j Second, it facilitates simpler analysis because it allows one to analyze
" fresh" assemblies that have equivalent reactivity to higher enriched,
" burned" assemblies. This technique was used to calculate the " fresh" 2.5 w/o value used as " burned" fuel for the region I criticality calculation.
Reactivity equivalencing was done with the PHOENIX code (ref. 3) and has been accepted in many other fuel storage evaluations.
i for region 1, two cases were analyzed; one used nominal conditions and the other used worst case assumptions. The nominal calculations yielded a k,,, ions of i
0.9042 using a 95/95 uncertainty value of 10.0054. The worst case assumpt involved using the limiting values for manufacturing tolerances to limit Boraflex loading and optimize moderation.
Furthermore, fuel enrichments were increased by 0.05 w/o to account for enrichment variations. Using these worst case assumptions, k,, was calculated to 0.9448. K for both cases was calculated to be below the acceptance criteria requ,'i, ring a k, < 0.95.
Optimal moderation conditions are not required to be conside,r,ed for this case.
The staff finds the proposed changes acceptable because the referenced analysis is suitably conservative and it meets the acceptance criteria.
Regicn 2 analyses consisted of using reactivity equivalencing to extend the constant k,0 w/o Vcurvesto$.contoug5on the existing Prairie Island burnup vs. enrichme covereo the range from 3.87 to 4.27 w/o U{because the existing curves only This was necessar and this modified curve will be the basis for determining whether or not fuel can be stored in region 2 using the new storage technique. Because the proposed region 2 storage configuration will be bounded by the existing analyses (the extension of the existing burnup vs. enrichment curve ensures that the fuel storage configuration of region 2 will be no more react".ve than what is currently used in the spent fuel pool), no further analyses were necessary.
One can conceive of scenarios in which rack reactivity will increase, for example, a misloaded fuel assembly or an assembly being dropped into the pool.
The licensee has considered these events and because credit may be taken for the presence of a soluble poison (in this case, boron) for accident situations, the rack k,,, will remain well below the 0.95 acceptance criterion for these limiting cases.
In order to ensure the presence of boron, the licensee's proposed TS 3.8.E.2 requires that boron be present in the pool from the onset of fuel movement until a spent fuel pool verification is performed to ensure that no assemblies have been mis-positioned. The boron
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concentration during fuel movement is required to be 1800 ppm or greater.
Furthermore, surveillance requirements (in proposed TS table 4.1-2B) during fuel movement and storage have been updated to require weekly tests of the i
boron concentration when boron is required to be present in the pool. The staff has reviewed the proposed changes and finds them acceptable.
Based upon the above discussion, the staff concludes that the proposed TS changes relating the fuel storage, core design, and RWST boron concentration j
are acceptable. The staff's conclusion is based upon the following 1.
The referenced analysis used established techniques that were properlr l
verified.
l 2.
The analysis was suitably conservative and bounding.
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3.
The results met NRC acceptance criteria with respect to K,,,.
l 4.
The consequences of limiting accidents are acceptable.
i 5.
The proposed TS are consistent with the analyses provided.
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2.4 Desian Basis Accidents j
In its application, the licensee evaluated the possible consequences of postulated accidents, included means for their avoidance in the design and operation of the facility, and provided means for mitigation of their consequences should they occur. The licensee has evaluated the effect of the changes on the calculated consequences of a spectrum of postulated design
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basis accidents (DBA) and concluded that the effect of the proposed TS changes is small. The licensee concluded that the calculated consequences are within regulatory requirements and staff guideline dose values. Since the licensee proposes to utilize higher enrichment fuel, the staff reevaluated the fuel handling accident for Prairie Island to consider the effects of increased l
enrichment and burnups.
In its evaluation for Prairie Island, issued on September 28, 1972, the staff conservatively estimated offsite doses due to radionuclides released to the atmosphere from a fuel handling accident. The staff concluded that the plant miti,ative features would reduce the doses for this DBA to below the doses t
specified in NUREG-0800, "Stardard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants."
Since the applicant intends to utilize higher enrichment fuel, the staff reanalyzed the fuel handling DBA for this case. According to NUREG/CR-5009,
" Assessment of the Use of Extended Burnup Fuel in Lightjater Power Reactors,"
(February 1989) increasing fuel enrichment to 5.0 w/o U with a maximum l
burnup of 60,000 MWD /T increases the doses for a fuel handling accident by a factor of 1.2.
235 The licensee proposes to limit enrichment to 5.0 w/o U Therefore, the 1.2 factor increase in dose displayed in Table 1 below, bounds the dose consequences of the licensee's proposal.
In Table 1, the new and old DBA doses are presented and compared to the guidelines doses in NUREG-0800 (established based on 10 CFR Part 100).
6 TABLE 1 RADIOLC3ICAL CONSEQUENCES OF FUEL HANDLING DESIGN BASIS ACCIDENT (REM)
Exclusion Area low Population Zone Thyroid Thyroid Staff Evaluation 33 6
September, 1972 Bounding 39.6 7.2 Estimates for Higher Enrichment Fuel Burnup*
Regulatory 75 75 Requirement (NUREG-0800 Chapter 15.7.4)
Factor of 1.2 greater than original estimate for iodine.
The staff concludes that the only potential increased doses resulting from the fuel handling accidents with increased enrichment are the thyroid doses; these doses remain well within the dose limits set forth in NUREG-0800 and are, therefore, acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State Official was notified of the proposed issuance of the amendments. 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 impact was published in the Federal Reaister on August 30, 1993 (58 FR 45536).
Accordingly, based upon the environmental assessment, the Commission has determined that issuance of these amendments 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
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6.0 REFERENCES
1.
Letter from USNRC to S.R. Tritch (Westinghouse Electric Corporation),
accepting the " referencing of topical report WCAP-12610 VANTAGE + Fuel Assembly Reference Core Report," July 1, 1991.
2.
NUREG/CR-0200, " SCALE: A Modular Code System for Performing Standardized i
Computer Analysis for Licensing Evaluation, Volume 1," Oak Ridge Natior.al Laboratory, 1982.
3.
WCAP-11956, " Qualification of the PH0ENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," Westinghouse Electric Corporation, 1987 (Proprietary).
Principal Contributors:
T. Ulses J. Minns Date: Septirtxr 3,1993
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