ML20203D408
| ML20203D408 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 12/04/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203D363 | List: |
| References | |
| NUDOCS 9712160187 | |
| Download: ML20203D408 (10) | |
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t UNITED STATES j
j NUCLEAR REGULATORY COMMISSION W AsHtNGTON, D.C. 30kWMoo1 e
o 59.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.94 TO FACILITY OPERATING LICENSE NO. NPF-37.
AMENDMENT NO.94 TO FACILITY OPERATING LICENSE NO. NPF-86.
AMENDMENT NO.86 TO FACILITY OPERATING LICENSE NO. NPF-72.
AND AMENDMENT NO.86 TO FACILITY OPERATING LICENSE NO. NPF-77 COMMONWEALTH EDISON COMPANY BYRON STATION. UNIT NOS.1 AND 2 BRAIDWOOD STATION. UNIT NOS.1 AND 2 DOCKET NOS. STN 50-404. STN 50-455. STN 50-456 AND STN 50-457 1.0 WTRODUCTION On April 2,1997, the staff issued Ucense Amendment Nos. 86 for Byron Station, Units 1 and 2, and Ucense Amendment Nos. 78 for Braidwood Station, Units 1 and 2, granting partial credit for boron in the spent fuel pools to maintain the suberiticality. The amendments were granted because Commonwealth Edison Company (Comed) had noted degradation of the Boroflex panels in the spent fuel storage cells. However the amendments were to be in effect only until December 31,1997, while development of a long temt solution was underway, in a letter of June 30,1997 (Reference 1), supplemented by letter of September 23,1997 (Reference 2),
Comed requested changes to the Byron and Braidwood, Units 1 and 2, Technical Specifications (TS) to allow the use of credit fc' soluble boron in the spent fuol pool criticality analyses. These criticality analyses were performed using the methodology developed by the Westinghouse Owners Group (WOG) and described in WCAP-14416-NP-A, " Westinghouse Spent Fuel Rack Criticality Analysis Methodology"(Reference 3).
2.0 EVALUATION 2.1 Criticality Analysis The Byron and Braidwood spent fuel storage racks were analyzed using the Westinghouse methodology which has been reviewed and approved by the NRC (Reference 3). This methodology takes partial credit for soluble boron in the fuel storage pool criticality anaYses and requires conformance with the following NRC acceptance criteria for preventing criticality outside the reactor.
1) k, shall be less than 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties at a 95 percent probability,95 percent confidence (95/95) level as described in WCAP-14416-NP A; and hh p
P _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
2) k, shall be less than or equal to 0.95 if fully flooded with borated water, which includes an allowance for uncertainties at a 95/95 level as described ir WCAP-14416-NP-A.
The analysis of the reactivity effects of fuel storage in the Byron and Braidwood spent fuel racks was performed with the three-dimensional Monte Cario code, KENO-Va, with neutron cross sections generated with the NITAWL-il and XSDRNPM-S codes using the 227 group ENDF/B V cross-section library. Since the KENO-Va code package does not have bumup capability, depletion analyses and the determinatic.,a of small reactivity increments due to manufacturing tolerances were made with the two-dimensional transport theory code, PHOENIX-P, which uses a 42 energy group nuclear data library. The analytical methods and models used in the reactivity analysis have been benchmarked against experimental data for fuel assemblies similar to those for which the Byron and Braidwood racks are designed and have been found to adequately reproduce the critical values. This experimental data is sufficiently diverse to establish that the method bias and uncerteinty will apply to rack conditions which include close proximity storage and strong neutron absorbers. The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the Byron and Braidwood storage racks with a high degree of confidence.
The Byron and Braidwood spent fuel pools are composed of three different types of storage racks, designated as Region 1, Region 2, and Failed Assembly cells. Region 1 contains three distinct cell types denoted as Type 1, Type 2, and Type 3. All cells contain 7.75 inch wide Boraflex sheets of 0.075 inch thickness. The Type 1 interior cells also contain Boral sheets outside of the Boraflex on all four sides. The Type 2 cells are side peripheral cells with Boral sheets outside of the Boraflex only on the three interior sides. Type 3 cells are comer peripheral cells with Boral sheets outside of the Boraflex only on the two interior sides. The Region 2 racks contain Boraflex sheets on all four sides, but no Doral. The six Failed Assembly cells contain no neutron absorber but rely upon the 21 inch lattice spacing to meet criticality requirements.
The spent fuel storage rnks have previously been qualified for storage of various Westinghouse 17x17 fuel assembly types with maximum enrichments up to 5.0 weight percent (w/o) U-235 The maximum enrichment is based on a nominal value of 4.05 w/o U-235 plus a manufacturing tobrance of 0.05. The spent fuel rack Boraflex absorber panels were considered in this previcas analysis. Because of the Boraflex deterioration that has been observed in many spent fuel pools, the Byron and Braidwood spent fuel storage racks have been reanalyzed neglec'ing the presence of Boraflex to allow storage of all 17x17 fuel assemblies with nominal enrichments up to 5.0 w/o U-235 (enrichment tolerance of 10.05 w/o U-235) using credit for checkerboarding, bumup, bumable absorbers, and soluble boror,.
For the Region 1 and Region 2 storage racks, the moderator was assumed to be pure water at a temperature of 68 degrees Fahrenheit and a density of 1.0 gm/cc and the array was assumed to be infinite in lateral extent. Uncertainties due to tolerances in fuel enrichment and density, storage cell inner diameter, storage cell pitch, stainless, steel thickness, assembly position, calculational uncertainty, and methodology bias uncertainty were accounted for. For the Region 1 analysis, a 10.007 inch tolerance about the nominal Boral sheet thickness of 0.075 inches was also used. These uncertainties were appropriately determined at the 95/95
probability / confidence level. A methodology bias (determined from benchmark calculations) as well as a reactivity bias to account for the effect of the normal range of spent fue: pool water temperatures (50 degrees Fahrenheit to 160 degrees Fahrer,heit) were included. These biases and uncertainties meet the previously stated NRC requirements and are, therefore, acceptable.
For Region 1, the enrichment required to maintain h less than 1.0 with all cells filled with Westinghouse 17x17 OFA fuel assemblies (most reactive type) and no soluble boron in the pool waterwas found to be 4.70 w/o U-235. This resulted in a nominal 4 of 0.98264. The 95/95 h was then determined by adding the temperature and methodology biases and the statistical sum of independent tolerances and uncertainties te the nominal 4 values, as described in Reference 3. This resulted in a 95/95 4 of 0.99944. Since this value is less than 1.0 and was determined at a 95/95 probability /coafidence level, it meets the NRC criterion for precluding criticality with no credit for soluble boron and is acceptable.
Storage of fuel assemblies with nominal enrichments between 4.70 and 5.0 w/o U-235 is achievable by crediting the reactivity decrease associated with Integral Fuel Bumable Absorbers (IFBAs). IFBAs consist of a boron neutron absorbing material applied as a thin coating on the outside of the UO fuel pellet. Reactivity calculations were performed to 2
determine the number of IFBA rods needed for nominally enriched 5.0 w/o fuel which yield a y equivalent to or lower than that previously obtained for 4.70 w/o U-235 fuel. For this analysis, the fuel assembly is modeled at its most reactive point in cycle life. Each IFBA rod contains the minimum standard B-10 loading (1.50 mg/in) offered by Westinghouse for 17x17 OFA fuel.
Uncertainties associated with IFBA credit include a 5 percent manufacturing tolerance and a 10 percer.t calculational uncertainty on the B-10 loading. The staff finds these uncertainties adequately conservative and acceptable. The lowest IFBA pattem (16 rods) offered by Westinghouse was found to be Lcceptable, in order to assure that a 5 percent subcritical margin is maintained, credit for soluble boron was taken. The soluble boron credit calculations for Region 1 assumed the all cell storage configuration moderated by water borated to 400 ppm. As previously described, the individual tolerances and uncertainties and the temperature and methodology biases were added to the calculated nominal 4 to obtain a 95/95 value. The resulting 95/95 9 was 0.94569 for OFA fuel enriched to 4.70 w/o U-235. Since 6 is less than 0.95 with 400 ppm of boron and uncertainties at a 95/95 probability / confidence level, the NRC acceptance criterion for precluding criticality is satisfied. This is well below the minimum spent fuel pool boron concentration value of 2000 ppm required by TS 3.9.11 and is, therefore, acceptable, in order to determine the enrichment required to maintain y less than 1.0 in Region 2 with no credit for soluble boron or Doraflex, three configurations were analyzed. The first was all cell storage in which fuel assemblies with sufficiently low nominal enrichments were stored in every celi location. The second involved a 3-out of-4 checkerboarding pattern with one empty ce;! and fuel assemblies with nominal enrichments ep to 1.64 w/o U-235 in the other three cells. The third configuration used a 2-out-of-4 assemb!y checkerboard with two diagonally adjacent empty cells and two assemblies with nominal enrichments up to 4.10 w/o U-235 in the other diagonally adjacent cells. The criticality analyses considered all the fuel types currently stored in the spent fuel pool and in use at Byron and Braidwood. The reactivity of the Westinghouse OFA design
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l was found to bound other fuel types in use or stored at Byron and Braidwood (VANTAGE 5, VANTAGE +, and PERFORMANCE +).
The nominal 4 for the all cell storage configurstion was determined to be 0.96885 for Westinghouse OFA fuel enriched to 1.14 w/o U-235 with no credit for soluble boron or Boraflex.
The 95/95 L wec then determined by adding the temperature and methodology biases and the statistical sum of independent tolerances and uncertainties to the nominal 4 values, as described in Reference 3. This resulted in a 95/95 h of 0.99631. For the 3-out-of-4 checkerboard configuration with one empty cell and the remaining cells containing assemblies having a nominal enrichment no greater than 1.64 w/o U-235 and no credit for soluble bomn or Boraflex, the 95/95 4 was 0.99662. The 2-out-of-4 checkerboard configuration with two diagonally adjacent cells containing assemblies having a nominal enrichment no greater than 4.10 w/o U-235 and the remaining cells empty and no credit for soluble boron or Boraflex, the 95/95 h was 0.99664. Since these values are less than 1.0 and were determined at a 95/95 probability / confidence level, they moot the NRC criterion for precluding criticality and are acceptable.
The soluble boron credit calculations assumed 1.14 w/o U-235 fuelin the all cell storage configuration moderated by water borated to 150 ppm. This is well below the minimum spent fuel pool boron concentration value of 2000 ppm required by TS 3.9.11 and is, ther6 Tore, acceptable. As previously described, the individual tolerances and uncertainties and the temperature and methodology biases were added to the calculated nominal 4 to obtain a 95/95 value. The resulting 95/95 y was 0.94743. Since h is less than 0.95 with 150 ppm of boron and uncertainties at a 95/95 probability / confidence level, the NRC acceptance criterion for precluding criticality is satisfied.
For the checkerboard configurations described above, a soluble boron concentration of 200 ppm was found to result in a 95/95 4 of 0.93999 and 0.94663, for the 3-out-of-4 and the 2-out-of 4 checkerboard configurations, respectively. Therefore, these also meet the NRC acceptance criterion for precluding criticality and are acceptable.
The concept of reactivity equivalencing due to fuel bumup was used to achieve the storage of fuel assemblies with enrienments higher than 1.14 w/o U 235 for the all cell stcrage configuration and fuel assemblies with enrichments higher than 1.64 and 4.10 w/o U-235 for the 3-out-of 4 and 2-out-of-4 checkerboard configurations, respectively. The NRC has previously accepted the use of reactivity equivalencing predicated upon the reactivity decrease associated with fuel depletion. This analysis also includes spent fuel decay time credit, which results from the radioactive decey of isotopes in the spent fuel to daughter isotopes. The loss in reactivity due to the radioactive decay of the spent fuel results in reducing the minimum bumup needed to meet the reactivity requirements, in the decay time methodology, the fission product isotopes are frozen at the concentrations existing at the time of discharge from the core, except for Xe-135, which is removed. These calculations are performed at different discharge bumups. The fuelis depleted using a high soluble boror,i%down curve to enhance the buildup of plutonium making the fuel more reactive in the spent fuel storage racks. Credit is taken only for the decay of actinides, one of the major contributors being the decay of Pu-241 to Am-241. Calcu!ations by Westinghouse from 100
hours after shutdown (at which time the major fission product Xe-135 has essentially decayed away) to 30 years following shutdown have shown that decay of the fission products has the effect of continuously reducing the reactivity of the spent fuel. However, no cred t for fission product decay is used in the decay time credit. Based on there conservative assumptions, the staff concludes that the proposed use of decay time credit is acceptable.
To determine the amount of soluble boron required to maintain k,,50.95 for storage of fuel assemblies in Region 2 with enrichments up to 5.0 w/o U-235, a series of reactivity calculations was performed to generate a set of enrichment versus fuel assembly ciischarge bumup ordered pairs which all yield an equivalent k,,when stored in the Byron and Braidwood spent fuel storage racks. These are shown in TS Figures 5.6-1 through 5.6-3 and represent combinations of fuel enrichment and discharge bumup which yield the same rack k,, as the rack loaded with 1.14,1.*4, and 4.10 w/o fuel (at zero bumup) for all cell, 3-out-of-4, and 2-out-of-4 storage configus tions, respectively. Decay time credit is included for the sti cell and 3-out-of-4 storage
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confirJrations. Uncertainties associated with bumup credit includs a reactivity uncertainty of 0.01 Ak at 30,000 MWD /MTU applied linearly to the bumup credit requ"rr.ent to account for calculational and depletion uncertainties and 5 percent on the cM~'
' umup to account for bumpp measurement uncertainty. The NRC staff conclude o
- : certainties are acceptable. The amount of additional scluble boron, ab:
tr.
- > quired above, that is needed to acceant for these uncertainties is 400 ppm fc -
wafiguration,350 ppm for e.
the 3-out of-4 checkerboard configuration, and 50 ppm i
.-out-of-4 checkerMard configuration. This results in a total soluble boron credit of 550 ppm for the all cells and 3-out-of-4 configurations and 250 ppm for the 2-out-of-4 configuration. These are well below the mirdnum spent :uel pool boron concentration value of 2000 ppm required by TS 3.9.11 and are, therefore, acceptable.
l Although most accidents will not result in a reactivity increase, two ace; dents can be postulated for each storage configuration which would increase reactivity beyond the analyzed conditions.
The first would be a loi.r. of fuel pool coc, ling system and a rise in pool water temperature frorn 160 degrees Fahrenheit to 240 degrees Fahrenheit. The second would be a mistoad of an assembly into a cell for which the restrictions on location, enrichment, or bumup are not satisfied. Calculations have shown that the mistoad assembly accident for a 2-out-of-4 checkerboard configuration results in the highest reactivity increase. The reactivity increase requires an additional 1400 ppm of soluble boron. However, for such events, the double contingency principle can be applied. This states that the assumption of two unlikely, independent, cor. current events is not mquired to ensure protection against a criucality accident.
Therefore, the minimum amoJnt of boron req' aired by TS 3.9.11 (2000 ppm)is more than sufficient to cover any accident and the presence of the additional boron above the a
concentration required for normal conditions and reactivity equivalencing (550 ppm maximum) can be assumed as a realistic initial condition since not assuming its presence would be a second uc.t kely event. Similar calculations for the six Failed Assembly spent fuel racks when all cc% comam. fuel enriched to 5.0 w/o U-235 resulted in a 95/95 k,, of 0.96330 with no credit taken for soluble boron and 0.93543 when 100 ppm of soluble boron is included. An additiunal 100 ppm of soluble boron is required for accidents, resulting in a total boron credit of 200 ppm, well below the minimum spent fuel pool boron concentration value of 2000 ppm require d by TS 3.9.11 and, therefore, acceptable.
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In ordar to prevent an undesirable increase in reactivity, the boundaries between the different storage configurations were analyzed. For Region 1, the boundary between the rack modules must use a 3-out of-4 checkerboard arrangement since the peripheral cell exterior sides do not contain Boral sheets. Severalinterface requirements apply within Region 2. The bouridary between all cell storage and 2-out-of-4 or 3+ut-vf-4 checkerboard storage must either be separated by a vacant row of *, ells or configured such that the first row of carryover la the checkerboard storage zone uses 1.64 w/o fuel assemblies attemating with empty cells. The boundary between 2-out-of-4 and 3-out-of-4 checkerboard storage must either be separated by a vn. cant row of cells or the interface must be configured such that the f%t row of carryover in the 2-out-of-4 storage zone uses 4.10 w/o assemblies attemating with empty cells. Since the criticality analysis does not consider the Boraflex neutron absorber panels, the boundary between Region 1 and Region 2 must be configured such that one row of vacarit cells is maintained between the regions.
The TS changes proposed as a result of the revised criticality analysis are consistent with the changes stated in the NRC Safety Evaluation (SE) for WCAP-1441C-P (Reference 4).
Westinghouse submitted a revised topical report, WCAP-14416-NP-A, Revision 1, which incorporated the changes stated in the NRC SE. Also, since tha staff disagreed with the proprietary finding of the original WCAP-14416-P, Westinghouse's revised topical report was submitted as a nonproprietary version. Based on this cone' g with the approved methodology and on the above evaluation, the staff finds these TS changes acceptable. The proposed associated Bases changes adequately descrite these TS changes and are also acceptable.
2.2 Boron Dilution of the Spent Fuel Pool in accordance with the NRC Safety Evaluation of the Westinghouse methodology described in WCAP-14416-NP A (Reference 4), the licensee Wformed a boron dilution analysia. to ensure that sufficient time is available to detect and mitigate the dilution prior to exceeding the 0.95 k,,
design basis. Potential events were quantified to show that sufficient time wi:1 te available to enable adequate detection and suppression of any dilution event.
Deterministic dilution event calculations were performed for Byron and Braidwood to define the dilution times and volumes necessary to dilute the spent (Jet pool from the minimum TS boron concer.tration of 2000 ppm to a soluble boron concentration where k,, of 0.95 would be approached (550 ppm). The volume required to dilute the spent fuel pool from the TS l'mit to 550 ppm 12 612,000 gallons. The various events that were considered included dilution from the boron recovery system recycle holdup tanks, primary water system, chemical volume and control system blender, demineralized water system, fire protection system, reverse osmosis system, spent fuel pool demineralizers, station heating system. and component cooling veter system, and other events that may affect the boron concentration of the pool, such as seismic ovents, pipe break, and loss of offsite power.
The licensee's evaluation concluded that the most limiting event was determined to be using the p,, nary water connection to the spent fuel pool cleanup subsystem for either makeup or for trant 'emng spent resin. It is the only unborated water authorized for use under ncrmal plant conditions. A dilution event would occur if the process isolation valve was inadvertently left l
l i
l open. It would take approximately 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> at a ate of 220 gpm to dilute the spent fael pool to 550 ppm. It is reasonable to expect that alarms oi,,, ' tor rounds (every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) would detect the increase in spent fuel pool sevel prior to reaching 5:x,,spm and the event would be terminated by plant personnel. No single tank source is large enough at Byron to dilute the spent fuel pool below 550 ppm. At Braidwood, the only source large enough to dilute the spent fuel pool below 550 ppm (508 ppm) is the condensate storage tank (CST) which has a capacity of 650,000 gallons. The water from the CSTs also allow the highest dilution rate of 420 gpm. At this dilttion rate, it would take 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to dilute it e spent fuel pool to 550 ppm. Again, it is reasonable to expect that this increase in the spent fuel pool level would be detected by alarms or by operator rounds and termina'.ed by plant personnel. The evaluation of the dilution sources determined that two scenarios do not change the level of the spent fuel pool during dilution.
Ti.ese two scenarios involve the transfer of water from the reverse osmosis system and the spent fuel pool demineralizers. The limiting dilution would require 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br />. The TS surveillance requirement of sampling every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> would detect the dilution event in sufficient time. The staff finds that a boron dilution event to 550 ppm for the Byron and Braidwood plants is unlikely and the licensee's analysis is acceptable.
The licensee's evaluation concluded that an unplanned or inadvertent event which would result in the dilution of the spent fuel pool boron concentration from 2000 ppm to 550 ppm is not a credible event. The staff fir.ds that the combination of the TS-controlled spent fuel pool concentration and 48-hour sampling requirement, the alarms, operator rounds, and other administrative controls should adequately detect a d,!ution event prior to 19 reaching 0.95 (550 ppm) and, therefore, the analysis and proposed TS controls are acceptable for the boron dilution aspects of the request.
Additionally, the enticality analysis for the spent fuel storage pool show that IQ remains less than 1.0 at a 95/95 probability / confidence level even if the pool were completely filled with unborated water. Therefore, even if the spent fuel storage pool were diluted to zero ppm, the fuelis expected to remain suberitical.
3.0
SUMMARY
Based on the review described above, the staff finds the criticality aspects of the propt, sed Byron and draidwood license amendment request are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling. The analysis assumed credit for solubli bo:en, as ailowed by WCAP-14416-NP-A, but no credit for the Goraflex neutron absorber panels. The amount of soluble boron required for each storaga configuration analyzed is shown in attached Table 1.
The foiiowing storage configurations and U-235 enrichment limits for Westir.ghouse 17x17 OFA fuel acsemblies were detarmined to be acceptable.
For Region 1:
Assemblies with initial nominal enrichments no greater than 4.70 w/o U-235 can be stored in any cel: location. Fuel assemblies with initial nominal enrichments greater than this and up to 5.0 w/o U-235 must contain at least 16 IFBA rods.
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For Region 2:
Assemblies with initial nominal enrichments no greater than 1,14 w/o U-235 can be stored in any cell location. Fuel assemblies with initial nominal snrichments greater than this and up to 5.0 w/o U-235 must satisfy a minimum bumup requirement as shown in TS Figure 5.61.
Assemblies with initial nominal enrichments no greater than 1.64 w/o U-235 can be stored in a 3-out-of-4 checkerboard arrangement. Fuel assemblies with initial nominal ennchments greater than this and up to 5.0 w/o U-235 must satisfy a minimum bumup requirement as shown in TS Figure 5.6-2.
Assemblies with initial nominal enrienments no grotter than 4.10 w/o U-235 can be E
stored in a 2-out-of-4 checkerboard arrangement. Fuel assemblies with initial nominal enrichmen'.s groater than this and up to 5.0 w/o U-235 must satisfy a minimum bumus
- requirement as shown in TS Figure 5.6-3.
l For Failed Assembly Cells:
4 -
Assemblies with initial nominal enrichments no greater than 5.0 w/o U-235 can be stored j
in any celllocation.
l With respect to possible inadvertent dilution of boron from the spent fuel pool, the staff finds that the combination of the TS controlled spent fuel pool concentration and 48-hour sampling
+
requirement, the alarms, operator rounds, and other administrativa controls should adequately detect a dilution event prior to ly reaching 0.93 (550 ppm) and, therefore, the analysis and proposed TS controls are acceptable for the borcn dilution aspects of the request, i
4.0 STATE CONSULTATION
- In accordance with the Commission's regulations, the Illinois State official was notified of the 4
proposed issuance of the amendments. The State official had no comments; f
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change i
survaillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that Inay be released offsite, and that there is no significant increase in individual or cumulative l'
occupational radiation expo lsure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 54868). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),
1 no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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E.0 QONCt.USION The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable ast.urance that the health and safety of the public will not be endangered by operation in the pmposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the isseance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Attachment:
Summary of Soluable Boron CF-At Requirements for Byron /Braidwood, Units 1 and 2 Principal Contributors: L.Kopp D. Jackson Da%: December 4, 1997 i
, I.
7.0 REFERENCES
1.
Letter, J. B. Hosmer, Comed, to U. S. NRC,
Subject:
" Application for Amendment to Appendix A, Technical Specifications, for Facility Operating Licenses: Byron and Braidwood Nuclear Power Stations, Units 1 and 2, Boron Credit in the Spent Fuel Storage Peol," dated June 30,1997.
2.
Letter, J. B. Hosmer, Comed, to U. S. NRC,
Subject:
"Supplemer,t to the Application for Amendrnent Request for Appendix A, Technical Specifications, to Facility Operating Licenses: NPF-37, NPF-66, NPF-72, and NPF-77, Boron Credit in the Spent Fuel Pool,"
dated September 25,1997.
3.
Westinghouse Electric Corporation, WCAP-14416-NP-A, " Westinghouse Spent Fuel Rack Criticality Analysis Methodology," Newrnyer, W. D., dated November 1996.
4.
Letter, T. E. Collins, NRC, to T. Greene, Westinghouse Owne,s' Group,
Subject:
" Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Critica!ity Analysis Methodology," (TAC No. M93254), dated October 25, 1996.
I
a TABLE 1 Summary of Soluble Boron Credit Requirements for Byron /Braidwood Units 1 and 2 Total Soluble Soluble Boron Boron Credit Soluble Boron Required for Required Storage Required for Reactivity Without Region Configuration 6 s0.95 Euivalencing Accidents (ppm)
(ppm)
(ppm) 1 All Cell Storage 400 0
400 2
All Cell Storage ISO 400 550 2
3-out-of-4 Checkerboard 200 350 550 2
2-out-of-4 Checkerboard 200 50 250 Failed
/1 Cell Assembly Storage 100 n/a 100 e