ML20149H149
| ML20149H149 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 11/07/1994 |
| From: | Saccomando D COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20149H153 | List: |
| References | |
| NUDOCS 9411180090 | |
| Download: ML20149H149 (12) | |
Text
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'^ C mm:nwealth Edis:n c Braidwood Nuclear Power Station n
Z.c Route #1, Box 84 y
v Braceville, Illinois 60407 Telephone 815/458-2801 November 7,1994 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk
Subject:
Application for Amendment to Facility Operating Licenses - Design Features Byron Station Units 1 and 2 NPF-37/66: NRC Docket Nos. 50-454/455 Braidwood Station Units 1 and 2 NPF-72/77: NRC Docket Nos. 50-456/457
Reference:
J. Bauer letter to W. Russell transmitting Request to Amend Technical Specification to allow for a Positive Moderator Temperature Coefficient and Reduced Thermal Design Flow dated March 23,1994 Pursuant to 10CFR50.90, Commonwealth Edison Compsny IComEd) requests to amend Technical Specifications of Facility Operating License Numbers NPF-37, NPF-66, NPF-72, and NPF-77. The proposed amendment involves revision to Section 5.6, Design Features-Fuel Storage, which allows for the use of higher enrichment fuel and specifies the spent fuel storage requirements for Regions 1 and 2 of the spent j fuel pool. Please note that an amendment of similar scope has been previously I submitted and approved for V.C. Summer. The codes and methods used in the l Byron /Braidwood reanalysis are the same as those used for the other utilities; I specifically, reactivity equivalency is used in the analysis by crediting Integral Fuel Burnable Absorbers (IFBAs) and accumulated fuel assembly burnup.
l The amendment request is subdivided as follows:
Attachment A: Description and Safety Analysis of Proposed Changes Attachrtent B: Proposed Revision to the Technical Specifications Attach.nent C: Evaluation of Significant Hazards Consideration Attachment D: Environmental Assessment Attachment E: Change Summary k:/nla/treutu LSX23 0 y4111eooyc 941107 3, PDR ADOCK 05o00454 P PDR I l
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, . Document Control Desk November 7,1994 Attachment F: Byron and Braidwood Spent Fuel Rack Criticality Analysis Considering Boraflex Gaps and Shrinkage 1
1 The proposed changes have been reviewed and approved by each station's On-site Review Committee and Off-site Review Committee, in accordance with Comed procedures. Comed has reviewed this proposed amendment in accordance with 10 CFR 50.92(c) and has determined that no significant hazards consideration exists, This proposed license amendment request is considered a Cost Beneficial Licensing Action. This amendment in conjunction with the proposed amendment which allows for a positive moderator temperature coefficient (see the reference letter) would realize a saving of approximately $700,000 per unit per cycle. This savings will result from Comed having to purchase fewer fuel assemblies because using higher enriched fuel in the reactor core design reduces the number of fuel assemblies required per reload.
To take advantage of this cost savings as soon as possible, Comed request approval of this proposed Technical Specification amendment in advance of the next Byron refueling outage, B2R05, scheduled to begin in February 1995. Comed would like to receive the amendment before January 20,1995, so that there is sufficient time to move fuel from the new fuel storage vault into the fuel pool before the reactor is taken off line.
Comed is notifying the State of Illinois of our application for these amendments by transmitting a copy of this letter and the associated attachments to the designated State Official.
To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my personal knowledge, but on information furnished by other Comed employees, contractor employees, and/or consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.
Please address any further comments or questions regarding this matter to this office.
incerely, Okr md>
Denise M. Va comando Nuclear Licensing Administrator Attachments cc: G. Dick, Byron Project Manager - NRR Signed and Sworn this 7th Day f N vember 1994 before me:
R. Assa, Braidwood Project Manager - NRR H. Peterson, Senior Resident Inspector - Byron S. Dupont, Senior Resident inspector - Braidwood
$Md Ndtary 'Publit' J. Martin, Regional Administrator - Region ill ,,,,,,,,,,,_ _
Office of Nuclear Facility Safety - IDNS " OFFICI AL SE AL "
HENRY L. BUCHHOLZ NOTARY PUBUC, STATE OF ILUNOIS MY COMMIS$10N EXPIRES 3/17/96
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ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES l
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i Description of the Proposed Chanaes Commonwealth Edison (Comed) proposes to revise Section 5, Design Features, of Technical Specifications for Byron and Braidwood stations. The proposed changes would allow an increase to the allowable nominal fuel enrichment from 4.2 to 5.0 weight percent uranium-235 (w/a U-235). Using higher enriched fuel in the design of reactor cores will result in fuel cost saving from the reduction in the number of new fuel assemblies required per reload and subsequent reduction in spent fuel storage l space. The proposed changes are supported by a criticality reanalysis of the spent l fuel pool. The criticality reanalysis considered the two storage regions and produced l separate criteria for each region. The changes includes: (1) increasing the allowable l storage enrichment in Region 1 and allowing the use of Integral Fuel Burnable Absorbers (IFBAs) for reactivity equivalencing, (2) revising the Region 2 discharge burnup curve to include nominal fuel enrichments up to 5.0 w/o U-235, and (3) making I editorial changes.
The marked up Technical Specification pages for each station indicating the proposed changes are provided in Attachment B. An Evaluation of No Significant Hazards Consideration is provided in Attachment C, and an Environmental Assessment is provided in Attachment D. A summary of the proposed changes is in Attachment E.
Attachment F is the report, " Byron and Braidwood Spent Fuel Rack Criticality Analysis ,
Considering Boraflex Gaps and Shrinkage," dated June 1994. A discussion of each l proposed change follows.
- 1. Proposed Change to the Allowable Storage Enrichment in Regicn 1 Descrintion and Bases of the Current Requirement The current Region 1 storage enrichment requirements are located in the
" Criticality Analysis of Byron and Braidwood Station Fuel Storage Racks." The original analysis was submitted to the NRC on August 15,1989, in accordance with Technical Specification 6.9.1.10. Region 1 is analyzed using an NRC accepted methodology to accommodate new fuel with a nominal enrichment of 4.2 w/o U-235 or spent fuel, regardless of the discharge fuel burnup. The purpose of the fuel storage requirements is to prevent inadvertent criticality in the fuel storage racks. The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95% probability at the 95%
confidence level that the K , of the fuel assembly array will not exceed 0.95 with full density moderation. The multiplication factor K ,is the ratio of the number of fissions in one generation divided by the number of fissions in the preceding generation.
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, . Dascription and Bases cf th7 Reauested Revision Comed proposes to redefine the maximum fuel storage enrichment limit for Region 1 as follows:
a) a maximum nominal initial U-235 enrichment of less than or equal to 4.2 weight percent, or b) a maximum nominal initial U-235 enrichment of 5.0 weight percent with sufficient Integral Fuel Burnable Absorbers present in each fuel assembly such that the maximum reference fuel assembly K-infinity (Km) is less than or equal to 1.470 at 68'F.
IFBAs consist of neutron absorbing material applied as a thin coating of ZrB2 on the outside of the fuel pellets, thereby rendering the material non-removable or an integral part of the assembly when manufactured. For IFBA credit, all 17X17 fuel assemblies placed in the Region 1 spent fuel pool racks must comply with the enrichment-IFBA requirements established in the revised criticality analysis or have a reference Km less than or equal to 1.470. The difference between Km and K,,is that Km does not assume any fast or thermal leakage. Therefore, in a finite geometry, neutrons may be lost and in an infinite geometry they will not be leaked out of the geometry.
The IFBA credit requirements given in Table 7 of Attachment F were conservatively established to identify the minimum number of IFBA rods per assembly needed to allow fuel storage in the Region 1 spent fuel racks. These l requirements have several conservatisms built in that may not be applicable to l the final IFBA fuel assembly design. These conservatisms include allowances for minimum IFBA length, IFBA rod configuration, and IFBA rod repositioning effects that accompany increased IFBA boron-10 loadings. Furthermore, the requirements for the number of IFBA rods per assembly will be verified during the reload core design process. The Byron Unit 2 Cycle 6 design has been j verified to meet this requirement.
l The IFBA credit requirements given in Table 7 of Attachment F are also ,
conservative with respect to the Technical Specification requirement to maintain Km less than or equal to 1.470 in unborated water at 68'F. The infinite multiplication factor, Km, is used as a reference criticality reactivity point to determine acceptability of storage for the actual IFBA fuel assembly design.
This allows the as-designed IFBA loading, IFBA lengths, and IFBA rod positions to be taken into account.
By allowing a maximum fuel storage enrichment limit of 4.2 w/o U-235, or greater than 4.2 w/o U-235 with sufficient IFBAs present in each fuel assembly, the maximum spent fuel pool K., will then be less than 0.95. Storage of fuel assemblies with enrichments greater than 4.2 w/o U-235 in Region 1 is acceptable with reactivity equivalencing. Reactivity equivalencing occurs when the addition of IFBAs causes a reactivity decrease.
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The criticality analysis for Region 1 fuel storage has been verified to meet the design basis for preventing criticality outside the reactor. The design basis is I that there is a 95 percent probability at a 95 percent confidence level that the !
effective neutron multiplication factor, K,,,, of the fuel assembly array will be less than 0.95. The criterion includes uncertainties. This design basis is recommended by ANSI 57.2-1983 and the OT Position Paper for Review and i l
Acceptance of Spent Fuel Storage and Handling Applications, dated April 14, 1978. The proposed changes would maintain the design basis for preventing criticality outside the reactor core for fuel assemblies with enrichments up to 5.0 w/o U-235. The revised criticality analysis, dated June 1994 is included for reference in this amendment request as Attachment F. The revised criticality analysis will be sent, under separate cover, in accordance with Specification 6.9.1.10, upon approval of the proposed Technical Specification change. ,
i impact of the Proposed Chance This change allows storage of 17X17 fuel assemblies (Optimized Fuel Assemblies, VANTAGE 5, VANTAGE + and PERFORMANCE +) with a nominal enrichment of 4.2 w/o U-235 utilizing all available storage cells. Fresh and burned fuel assemblies with a nominal enrichment of up to 5.0 w/o can also be stored in these racks provided sufficient IFBAs are present within each fuel assembly. The criticality design basis remains applicable and continues to be met. (Note that the maximum fuel enrichment is the combination of the nominal fuel enrichment and the Department of Energy enrichment tolerance of 0.05 w/o.)
The Region 1 criticality analysis contained in the June 1994, report " Byron and Braidwood Spent Fuel Rack Criticality Analysis Considering Boraflex Gaps and Shrinkage," (Attachment F) uses a criticality analysis methodology that has been accepted previously by the NRC. The analysis and report were reviewed and approved by Comed and were verified to meet the criteria in the following standards: ANS-8.11, ANSI /ANS-8.1-1983, NRC OT Position Paper, Regulatory Guide 3.42, ANSI /ANS-57.2-1983, ANSI /ANS-57.3-1983 and Regulatory Guide 3.43.
The new fuel storage vault is designed to handle the increased enrichment.
The Byron and Braidwood new fuel vaults were previously analyzed using NRC accepted criticality analysis methodology in June 1989. This analysis was performed to increase the storage enrichment of the New Fuel Vault to 5.0 w/o U-235. The New Fuel Vault analysis was submitted to the NRC and is the current licensing basis.
There is no adverse impact on the ability of the Spent Fuel Pool cooling system to maintain the bulk pool temperature within limits. The UFSAR analysis performed to calculate the maximum fuel cladding temperature and spent fuel pool cooling include assumptions which bound the use of more highl; anriched fuel assemblies. Although fuel enrichment is not a specific assumption in any k: / nl a /brdwd/ f ue12.wp f / 5
of these analysis, the heat load of a typical core offload may change with higher enrichments. The average burnup of the offload will be increased since fewer assemblies will be used per cycle; however, the new heat load will continue to be bounded by the UFSAR analysis because the spent fuel pool racks have been analyzed for a total core offload with all fuel assemblies having 4.5 years of operating time.
The radiological consequences analysis continues to bound the licensed fuel burnup and enrichment at Byron and Braidwood stations. The radic.hgical consequences analysis results are a function of the core inventory of radioactive isotopes. Since the maximum fuel burnup limits and fuel peaking factors will not be exceeded, the assumed fission product inventory will remain valid. Therefore, the limits of 10 CFR 100 continue to be met. Additionally, Byron and Braidwood addressed the issue of the impact on the radiation levels at the pool surface to the worker during non-accident conditions. These conditions are not changed as the result of this submittal, because the average fuel assembly burnup limit (isotopic inventory) and maximum power produced in each fuel assembly will not be changed by the increased fuel enrichment.
Using fuel with a higher enrichment in reactor core design will result in fuel cost savings from the reduction in the number of new fuel assemblies required per reload. Additionally, there is a savings in fuel storage space within the spent fuel pool.
- 2. Proposed Change to the Allowable Storage Enrichment in Region 2 Description and Bases of the Cunent Requirement The current Region 2 storage enrichment requirements are located in the
" Criticality Analysis of Byron and Braidwood Station Fuel Storage Racks." The original analysis was submitted to the NRC on August 15,1989, in accordance with Technical Specification 6.9.1.10. Region 2 is analyzed using NRC accepted methodology to accommodate fuel storage with nominal initial enrichments of up to 4.2 w/o U-235 that have accumulated minimum burnups within an acceptable bound as defined in Technical Specification Figure 5.6-1.
The design criteria are also met when a checkerboard loading pattem is used, regardless of assembly burnup, as stated it. Specification 5.6.1.1.a. The purpose of the fuel storage requiremen's is to prevent inadvertent criticality in the fuel storage racks. The design basis for preventing criticality outside the reactor is that, including uncertainties, there h a 95% probability at the 95%
confidence level that the K, of the fuel assembly array will not exceed 0.95 with full density moderation.
Description and Bases of the RequeMy Revision Comed proposes to redefine the nominal initial fuel storage enrichment limit in Region 2 as follows:
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a) Fuel assemblies may be stored in this region with a maximum nominal initial U-235 enrichment of 1.6 weight percent with no burnup and up to 5.0 weight percent U-235 with a minimum discharge burnup as specified in Figure 5.6-1, or b) Fuel assemblies with a maximum nominal initial U-235 enrichment of greater than 1.6 and less than or equal to 4.2 weight percent, assemblies that do not meet the minimum burnup specified in Figure 5.6-1, shall be loaded in a checkerboard pattern for storage in this region.
Comed also proposes to revise Figure 5.6-1 to reflect the increased enrichment liinit. The proposed changes would maintain the design basis for preventing criticality outside the reactor core for fuel assemblies with enrichments up to 5.0 w/o U-235. The changes are consistent with the June 1994 Westinghouse report, " Byron and Braidwood Spent Fuel Rack Criticality Analysis Considering Boraflex Gaps and Shrinkage," which uses an NRC accepted criticality analysis methodology.
The effect of top and bottom shrinkage has a negligible effect on the burnup credit in the Region 2 spent fuel pool racks. Any increase in flux at the ends of the fuel assembly increases the axial leakage experienced by the fuel assembly. Therefore, slight increases in the flux at the ends of the fuel assembly are offset by the increase in axial leakage. Other conservatisms in the burnup credit curve include a conservative estimate that 50% of the boraflex panels experience nonuniform shrinkage (random gaps) and the remaining boraflex panels experience uniform shrinkage (pullback) from the bottom end.
This also assumes the boraflex panel starts at 6.0625 inches from the bottom of the rack. An additional 1% AK penalty at 30,000 MWD /MTU is applied linearly to the burnup credit curve. This uncertainty is applied to the PHOENIX calculational results, which start at zero for zero burnup and increase linearly with burnup. This bias is considered to be very conservative based on good agreement between PHOENIX predictions and measurements, and on conservative estimates of fuel assembly reactivity variances with depletion history.
Normal blackness testing of the Byron and Braidwood spent fuel racks provides the data to support the boraflex gap and shrinkage assumptions made in the Byron and Braidwood criticality analysis. Boraflex blackness testing results to date are bounded by the assumptions made in the criticality analysis.
Additional testing, beyond current procedure, is not deemed necessary since the criticality analysis assumptions bound the EPRI database on boraflex.
Comed is retaining the provision to use checkerboard loading in Region 2 for fuel assemblies having a nominal initial enrichment of up to 4.2 w/o U-235. The basis for the current requirement, described in the June 1989 report " Criticality K t /fil a / brdwd/ f uel 2.wpf / 7
Analysis of Byron /Braidwood Station Fuel Storage Racks,"is not affected by the !
revised criticality analysis.
This amendment does not address checker boarding of fuel assemblies with enrichments greater than 4.2% in Region 2 that do not meet the minimum required burnup. Analysis to support checkerboarding will be obtained and approved prior to placing these fuel assemblies into Region 2. This updated analysis will be required to be approved prior to discharging >4.2% fuel with 2 cycles of burnup.
Impact of the Proposed Chance This change allows the storage of 17X17 fuel assemblies (Optimized Fuel Assembly, VANTAGE 5, VANTAGE + and PERFORMANCE +) in Region 2 with nominal enrichments up to 5.0 w/o U-235 utilizing all available storage cells providing that they meet the appropriate burnup for that enrichment as indicated in the revised Technical Specification Figure 5.6-1. This curve includes a 3% >
uncertainty factor to account for burnup calculation uncertainty. The analysis was performed using a criticality analysis methodology that was previously accepted by the NRC. The proposed change to increase the allowable storage enrichment continues to meet the design basis for storage of spent fuel in Region 2.
The analysis that developed this updated curve is contained in the Westinghouse report " Byron and Braidwood Spent Fuel Rack Criticality Analysis Considering Boraflex Gaps and Shrinkage." The methodology used in the analysis has been accepted by the NRC. The report is part of the Criticality Analysis of Byron and Braidwood Station Fuel Storage Racks defined in Technical Specifications. The revised analysis and report were reviewed and approved by Comed. The criticality analysis is based on assumptions that are conservative and bounding with respect to upper bound values for shrinkage and gaps recommended by EPRI. The original criticality analysis did not address boraflex gaps and shrinkage.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective neutron multiplication factor, K,,, of the fuel assembly array ,
will be less than 0.95 as recommended by ANSI 57.2-1983 and OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, dated April 14,1978. The analysis for Region 2 fuel storage was verified to meet the above design basis. The revised criticality analysis that considers boraflex gaps and shrinkage is included for information in this amendment request as Attachment F.
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Checkerboard loading of fuel assemblies was not analyzed in the revised l criticality analysis. Checkerboard fuel loading is still acceptable in Region 2 for !
fuel assemblies with an initial enrichment of up to 4.2 w/o U-235. The original !
criticality analysis that was submitted to the NRC shows that these fuel I assembibs can be safety accommodated without exceeding the criticality analysis acceptance criteria. The checkerboard pattern criticality analysis is only applicable for Region 2 and for fuel assembly enrichments of less than or equal to 4.2 w/o U-235.
There is no impact on the ability of the Spent Fuel Pool cooling system to maintain the bulk pool temperature within limits. The UFSAR analysis performed to calculate the maximum fuel cladding temperature and spent fuel pool cooling include assumptions which bound the use of more highly enriched fuel assemblies. Although fuel enrichment is not a specific assumption in any of these analysis, the heat load of a typical core offload may change with higher enrichments. The average burnup of the offload will be increased since fewer assemblies will be used per cycle; however, the new heat load will continue to be bounded by the UFSAR analysis because the spent fuel pool racks have been analyzed for a total core offload with all fuel assemblies having 4.5 years of operating time.
The new fuel storage vault is designed to handle the increased enrichment.
The Byron and Braidwo.- iew fuel vaults were previously analyzed using NRC accepted criticality anal) . methodology in June 1989. This analysis was performed to increase the storage enrichment of the New Fuel Vault to 5.0 w/o U-235. The New Fuel Vault analysis was submitted to the NRC and is the current licensing basis.
The radiological analyses are not affected by the proposed increase in fuel enrichment. As with Region 1, the radiological analyses are not a function of fuel enrichment but are a function of fuel assembly burnup (isotopic inventory) and power level. Since the presently used limits of lead rod average burnup (60,0000 MWD /MTU as provided in WCAP 10444-P-A) and peaking factors (provided in the Byron and Braidwood Technical Specifications) will not be exceeded, the conclusion presenedt in the UFSAR will be bounding. Therefore, there will be no adverse impact on the radiation levels at the pool surface to the worker during non-accident conditions, and the limits of 10 CFR 100 continue to be met.
Using higher enriched fuel in the reactor core design reduces the number of fuel assemblies required p- reload. Comed will save money by requiring fewer new fuel assemblies and, therefore, using less spent fuel storage space.
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- 3. Proposed Editodal Changes Description and Bases of the Requested Revisions Specification 5.6.1.1 is rearranged and reformatted so that the Region 1 requirements are provided in Specification 5.6.1.1.a. and Region 2 requirements are provided in Specification 5.6.1.1.b. The reference to the FSAR is changed to reflect the Updated Final Safety Analysis Report (UFSAR).
Impact of the Proposed Chance The proposed changes are administrative in nature, and do not reduce the requirements of any Technical Specification. The new' format more clearly shows the requirements for each region.
Discussion of Criticality Analysis The proposed Technical Specification changes are based on a revised criticality analysis, which is provided in Attachment F. Key informatian is provided below.
Technical Specification 1.9.a is a definition of Criticality Analysis of Byron and Braidwood Station Fuel Storage Rack. On August 15,1989, Comed submitted two reports to the NRC, in accordance with Specification 6.9.10, that make up the defined analysis. The two parts were the June 1989, " Criticality Analysis of the Byron /Braidwood Fresh Fuel Racks" and the August 1989 " Criticality Analysis of Byron /Braidwood Station High Density Fuel Racks." The August 1989 report is superseded by the June 1994 Westinghouse report " Byron and Braidwood Spent Fuel Rack Criticality Analysis Considering Boraflex Gaps and Shrinkage," except for the provision for checkerboard loading in Region 2. The revised report is sent under separate cover to the NRC, in accordance with Specification 6.9.10. Therefore, the current criticality analysis, as defined by Technical Specifications, would include the June 1994 report considering boraflex gaps and shrinkage, the original (June 1989) analysis of the fresh fuel racks, and the portion of the August 1989 report that discusses checkerboard loading in Region 2.
i The criticality reanalysis is based on maintaining K,,less than or equal to 0.95. For each spent fuel pool region, the most reactive or limiting fuel assembly type is ,
analyzed to establish the reference K., and confirm that the 0.95 limit is not exceeded. To provide for future fuel management flexibility, storage limits were developed for enrichments up to and including 5.0 w/o U-235 by taking credit for integral fuel burnable absorbers and accumulated fuel assembly burnup.
The criticality analysis performed for both storage regions produced separate criteria defining the storage limits applicable to each region as follows-l A : / nl a /t,r dad / f ue 12.wpf /10
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- 1. New and freshly discharged fuel assemblies with a maximum nominal enrichment of 4.2 w/o U-235 may be stored in Region 1. Fuel assemblies with a maximum nominal enrichment of 5.0 w/o U-235 may be stored in Region 1 when there are sufficient integral fuel burnable absorbers such that the maximum reference fuel Km is less than or equal to 1.470 at 68'F. ;
- 2. Fuel assemblies with a maximum nominal enrichment of 1.6 w/o U-235 with no burnup and up to 5.0 w/o U-235 with a minimum burnup as specified in proposed Technical Specification Figure 5.6-1 may be stored in Region 2. Fuel assemblies with a maximum nominal enrichment of greater than 1.6 w/o U-235 and less than or equal to 4.2 w/o U-235 that do not meet the minimum burnup specified in Figure 5.6-1 shall be loaded in a checkerboard pattern in Region 2.
Most accident conditions will not result in an increase in K.,. However, as discussed in Attachment F, accidents can be postulated that could cause reactivity to increase.
For these accident conditions, the double contingency principle of ANSI /ANS 8.1-1983 can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against criticality accidents.
Thus, for these conditions, the presence of soluble boron in the spent fuel pool water can be assumed as a realistic initial condition, since not assuming its presence would be a second unlikely event.
The most severe accident scenario is misloading a fresh fuel assembly at 4.2 w/o U-235 in the middle of a 5x5 array of Region 2 spent fuel rack cells with fresh assemblies at 1.6 w/o U-235. (With 5.0 w/o U-235 fuel, the above scenario is still the most severe accident scenario due to the fact that fuel that is 4.2 to 5.0 w/o U-235 has IFBA rods.) Calculations indicate this event could increase reactivity by as much !
as 0,0438 AK. To bound this increase, it is conservatively estimated that 300 ppm of soluble boron is required. Technical Specification surveillance 4.9.1.2 verifies that refueling canal boron concentration is greater than or equal to 2000 ppm at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in Mode 6. Administrative procedures require determinations of spent i fuel pool boron concentration at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during refueling operations l and weekly at all other times. Normally, a minimum concentration of 2000 ppm is j maintained, however, it may be allowed to decrease under controlled conditions, such ,
as during clean up via a reverse osmosis unit. Sampling frequency is increased when i boron concentration drops below 2000 ppm. In all cases, the boron concentration remains well above the 300 ppm boron required for the most severe accident scenario, described above. Should a postulated accident occur that causes reactivity to increase, K , will be maintained less than or equal to 0.95 due to the negative reactivity esset of tiie dissolved boron.
The Westinghouse criticality methodology used in the Byron /Braidwood analysis used the KENOVa Monte Carlo code. The typical Westinghouse KENOVa Monte Carlo calculation involves more than 60,000 neutron histories. The KENOVa default value is 30,900. The KENOVa edits or output, which show the average K,, by generation skipped, are examined to assure adequate convergence. These edits provide a visual inspection on the overall convergence of the KENOVa Monte Carlo results.
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Schedule Requirements The Technical Specification change is required to support the fuel load for Byron Unit 2 Cycle 6. The original criticality analysis allows storage of fuel with a nominal enrichment of up to 4.2 w/o U-235. Comed is planning to load fuel having 4.4 w/o U-235 during the refueling outage that is scheduled to begin February 10,1995.
Comed would like to receive the amendment before January 20,1995, so that there is sufficient time to move fuel from the new fuel storage vault into the fuel pool before the reactor is taken off line.
Identification and discussion of any irreversible consequences l
There were no irreversible consequences identified.
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