ML20138B323
| ML20138B323 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/21/1997 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20138B321 | List: |
| References | |
| NUDOCS 9704290123 | |
| Download: ML20138B323 (27) | |
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ATTACIIMENT A DESCRIPTION AND EVALUATION l OF PROPOSED CIIANGES !
ESK-97-089 I (Page 1 of 14) ;
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i Table of Contents
- 1. Background Information
- 2. Description of the Proposed Changes
- 3. Description of the Current Requirements
- 4. Bases for the Current Requirements
- 5. Need for the Revision of the Requirements and Basis for Exigency I
- 6. Description of the Revised Requirements
- 7. Basis for the Revised Requirements !
- 8. Schedule
- 9. References i
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9704290123 970421 PDR ADOCK 05000254 p PDR
ATTACIIMENT A DESCRIPTION AND EVALUATION OF PROPOSED CIIANGES ESK-97-089 (Page 2 of 14)
- 1. Background Information
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Change to Minimum Critical Power Ratio (MCPR) Safety Limit and MCPR Saf(tv Limit Bases i
Quad Cities Nuclear Power Station will be using Siemens Power Corporation Nuclear Division (SPC) ATRIUM-9B fuel assemblies for Unit 2 Cycle 15 operation. SPC has determined that the need exists to increase the size of the data base used to calculate the additive constant uncertainty for SPC's ATRIUM-9B 9x9 fuel design with an internal water channel. This uncertainty value is currently described in the NRC approved SPC ;
Critical Power Correlation, Reference 1. However, as part of their response to a NRC l inspection, SPC investigated the validity of that uncertainty. In Reference 5, SPC determined a new additive constant uncertainty for the ATRIUM-9B fuel by including additional experimental data from critical power tests of other fuel designs which share many of the same design features as the ATRIUM-9B fuel. The expanded datasets also include data that cover a wider range of pressures and flows and axial pawer shapes.
An interim approach, which provides additional conservatism, is being used to calculate the MCPR Safety Limit, during NRC review of Reference 5. The difference between the expanded data set's and the original data set's additive constant uncertainties is doubled for conservatism and added to the original additive constant uncertainty. The MCPR Safety Limit is then calculated using the resulting, additive constant uncertainty.
Reference 7 documents this conservative interim approach of doubling the difference in additive constant uncertainties. The new MCPR Safety Limit is 1.10 for Quad Cities Unit 2. This proposed amendment will change the MCPR Safety Limit for Unit 2 and the plant will operate with this conservatism until NRC approval of Reference 5. This proposed amendment also includes an insen to the MCPR Safety Limit Bases that states why the Unit 1 MCPR Safety Limit (1.07) and the Unit 2 MCPR Safety Limit (1.10) are different. ARer NRC approval of Reference 5, Comed intends to submit a Technical Specification change to lower the MCPR Safety Limit to a value supported by the new NRC approved uncertainty (based on the increased data base size) and add the Reference 5 topical to Section 6 of the Technical Specifications. l Addition of Siemens Power Corocration's (SPC) methodology for Application of the ANFB Critical Power Correlation to Coregident GE Fuel for Ouad Cities Unit 2 Cycle 15 to Section 6.9. A.6.b j In addition to the ATRIUM-9B fuel, Quad Cities Unit 2 Cycle 15 will operate with l l previously exposed GE fuel types (GE9 and GE10). Because the ANFB critical power '
correlation (Reference 1) will be used in establishing and monitoring MCPR limits for l
both the SPC fuel and the coresident GE fuel, ANFB additive constants and uncertainties were developed for GE9/10 fuel. These additive constants and uncertainties are 1
ATTACHMENT A DESCIUPTION AND EVALUATION OV PROPOSED CIIANGES ESK-97-089 (Page 3 of 14) discussed vad calculated in Reference 2. The additive constants and uncertainties in Reference 2 are only for the GE fuelin Quad Cities Unit 2 Cycle 15 and are therefore, not afkted by the concern that the data base for the ATRIUM-9B is insufficient to support One uncertainty values for the additive constants used for ATRIUM-98 Critical Power R atio (CPR) calculations.
This suSmittal proposes to add Reference 2, Siemens Power Corporation's (SPC) methorblogy for Application of the ANFB Critica! Power Correlation to Coresident GE Fuel fhr Quad Cities Unit 2 Cycle 15, to Section 6.9.A.6.b of the Technical -
Specitications. The SPC topical report and the Comed transmittalletter describe the meth odology used to determine the additive constants and the associated uncertainty for application of the ANFB Critical Power Correlation to GE fuel. The additive constant uncertainty for the GE fuel is also included in calculating the Quad Cities Unit 2 Cycle 15 MCPR Safety Limit. Since Section 6.9. A.6.b is to include the " analytical methods used to determine the operating limits" and because the Safety Limit is used as part of the operating limit determination, it is appropriate that Reference 2 be included.
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ATTACIIMENT A l DESCRIPTION AND EVALUATION l OF PROPOSED CIIANGES l ESK-97-089 (Page 4 of 14)
- 2. Description of the Proposed Changes l Change to Minimum Critical Power Ratio (MCPR) Safety Limit and MCPR Safety Limit Bases I This submittal proposes to change the Quad Cities Unit 2 MCPR Safety Limit from 1.07 to a greater limit of 1.10. This submittal also proposes to include an explanation why the Unit 1 MCPR Safety Limit (1.07) and the Unit 2 MCPR Safety Limit (1.10) are difTerent )
in the MCPR Safety Limit Bases. These changes willincorporate a new ATRIUM-9B ;
additive constant uncertainty calculated by SPC. j The new ATRIUM-9B additive constant uncertainty used to calculate the 1.10 MCPR !
Safety Limit is conservatively determined as follows: The difference is calculated '
between the additive constant uncertainties prior to and afler the data set was expanded j from 125 points to 527 points. This difference is then conservatively doubled The i doubled difference is then added to the additive constant uncertainty based on the 125 point data set. Using this conservative interim methodology the new ATRIUM-9B l additive constant uncertainty is 0.029. This new conservative additive constant uncertainty was then utilized by SPC in conjunction with their NRC-approved Reference 6 methodology to determine the appropriate MCPR Safety Limit for Quad Cities Unit 2 j Cycle 15. The new MCPR Safety Limit has been determined to be 1.10. This proposed i l amendment will change the MCPR Safety Limit and the plant will operate with this j conservatism until the NRC approval of Reference 5. After NRC approval of Reference 5, Comed intends to submit a Technical Specification change to lower the MCPR Safety Limit to a value supported by the new NRC approved uncertainty (based on the increased data base size) and add the Reference 5 topical to Section 6 of the Technical Specifications.
Addition of Siemens Power Corporation's (SPC) methodology for Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Ouad Cities Unit 2 Cycle 15 to Section 6.9.A.6.b This submittal proposes to add Reference 2, Siemens Power Corporation's (SPC) l methodology for application of the ANFB Critical Power Correlation to coresident GE Fuel for Quad Cities Unit 2 Cycle 15, to Section 6.9.A.6.b of the Technical Specifications. The SPC topical report and the Comed transmittalletter describe the methodology used to determine the additive constant and uncertainties necessary to use the ANFB Critical Power Correlation for coresident GE9 and GE10 fuel. The additive constant uncertainty is used as input to CPR calculations. Since Section 6.9.A.6.b is to include the " analytical methods used to determine the operating limits", it is appropriate that Reference 2 be included.
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ATTACIIMENT A DESCRIPTION AND EVALUATION l
l OF PROPOSED CIIANGES ESK-97-089 l (Page 5 of 14) l l 3. Description of the Current Requirements Change to Minimum Critical Power Ratio (MCPR) Safety Limit and MCPR Safety Limit Bases The MCPR is the smallest CPR which exists in the core, where CPR is the ratio of that power in the assembly which is calculated by application of the NRC approved correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. The fuel cladding integrity Safety Limit is set such that boiling transition, which is a precursor to (mechanistic) fuel damage, is not calculated to occur in 99.9% of the fuel rods if the safety limit is met during postulated transients (anticipated operational occurrences). The current MCPR Safety Limit for Quad Cities Units 1 and 2 is 1.07.
Addition of Siemens Power Cornoration's (SPC) methodolouv for Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Ouad Cities Unit 2 Cycle 15 to Section 6.9. A.6.b The Administrative Control section of the Technical Specifications lists the NRC approved topical reports describing the analytical methods used to determine the operating limits (Specification 6.9.A.6.b). The methods listed include the General Electric Standard Application for Reactor Fuel (GESTAR), and Comed Reports l benchmarking BWR Nuclear Design methods and Neutronic Licensing Analyses.
Due to the change from General Electric (GE) fuel to Siemens Power Corporation (SPC) fuel, submittals have been made to propose adding a number of NRC approved SPC methodology topicals, including the SPC topical on their CPR correlation (ANFB),
to the Specification 6.9.A.6.b Reference list. The Comed Topical Report NFSR-0091,
" Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods" was also proposed to be added to the Reference list in Specification 6.9.A.6.b. References 3 and 4 transmitted a proposed license amendment to the NRC for their review and approval.
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ATTACIIMENT A DESCRIPTION AND EVALUATION OF PROPOSED CIIANGES FSK-97-089 (Page 6 of 14)
- 4. Bases for the Current Requirements Change to Minimum Critical Power Ratio (MCPR) Safety Limit and MCPR Safety Limit 11asn i
The current requirements are based on GE methods, including the use of the GE CPR correlation (GEXL) to calculate the CPR for the GE fuel bundles. Comed has submitted proposed changes to update the current MCPR Safety Limit requirements to include '
SPC methods used to ensure that the MCPR Safety Limit requirements are satisfied (References 3 and 4). When a unit loads SPC fuel, the GE critical power correlation (GEXL) will no longer be used; instead, the SPC critical power correlation (ANFB) will l be the correlation used to establish and monitor the CPR limits. The GE fuel will be monitored using the ANFB correlation with additive constants that are used to ensure the ANFB results are conservative. l 1
j Addition of Siemens Power Corooration's (SPC) methodolouv for Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Ouad Cities Unit 2 Cycle 15 to Section 6.9. A.6.b I l
The list of references in Section 6.9. A.6.b provides documentation in the Technical Specifications of the NRC approved methods used to determine the operating limits that are detailed in the Core Operating Limits Report (COLR). This Reference list was created in response to Generic Letter 88-16.
The reference list of Section 6.9.A.6.b has been updated by References 3 and 4 to contain the NRC approved SPC methods and the NRC approved Comed Topical Repod l
for CASMO/MICROBURN BWR Nuclear Design Methods. '
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- 5. Need for the Revision of the Requirements and Basis for Exigency Change to Minimum Critical Power Ratio (MCPR) Safety Limit and MCPR Safety Limit Bases The MCPR Safety Limit increase is due to a new ATRIUM-9B additive constant uncertainty calculated and conservatively applied by SPC to the ATRIUM-9B fuel. This l new ATRIUM-9B additive constant uncertainty is necessary because SPC has
! determined that the data base used to determine the ATRIUM-9B additive constant uncertainty should be expanded to include a wider range of test conditions. (The data base for ATRIUM-9B consisted of 125 data points at 1000 psi for mass fluxes from 0.5 2
to 1.5 Mlb/hr-fl .) In response to this need for additional data points., SPC compiled a larger set of data for the recalculation of the additive constant uncertainty. Data sets from fuel designs, which like ATRIUM-98, have full length rods,9x9 design, and inner water channels, were included in the statistical analysis to provide a data pool of 527 points at different pressures and flows. Additional nonuniform axial power shape data from the ATLAS test facility was also used for the evaluation of the axial power shape
- effect for ATRIUM-98. The revised methodology for treating the uncertainties is provided in the Reference 5 topical report.
This increased data pool produced a new ATRIUM-9B additive constant uncertainty of 0.0195 as shown in Reference 5. In order to ensure conservative application of this uncertainty, the difference between the additive constant uncertainty for the data set of 527 points (0.0195) and the additive constant uncertainty for the data set of 125 points (0.01)is doubled and then added to the original additive constant uncertainty for the data set of 125 points. This new ATRIUM-9B additive constant uncertainty,0.029, is used to calculate the MCPR Safety Limit of 1.10.
The calculated MCPR Safety Limit is greater than the 1.07 MCPR Safety Limit currently in the Technical Specifications Section 2.1.B. Therefore, a revision is necessary to Section 2.1.B to incorporate the new Quad Cities Unit 2 MCPR Safety Limit. The conservative approach for calculating the MCPR Safety Limit will ensure fuel protection while Reference 5 is under NRC Staffreview l
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,-a e - p --1 4A ATTACIIMENT A DESCRIPTION AND EVALUATION OF PROPOSED CIIANGES ESK-97-089 (Page 8 of 14)
A_d. dition of Siemens Power Corporation's (SPC) methodoloav for Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Ouad Cities Unit 2 Cycle l 15 to Section 6.9. A.6.b l
Siemens Power Corporation's (SPC) methodology for application of the ANFB Critical i Power Correlation to coresident GE Fuel for Quad Cities Unit 2 Cycle 15, describes the l methodology used to determine the additive constants and the associated uncertainty for i application of the ANFB Critical Power Correlation to the coresident GE9 and GE10 fuel. The additive constant uncertainty is used as input to CPR calculations. Because the Safety Limit is used in calculating the operating limit, and because Generic Letter 88-i 16 indicated that Section 6.9. A.6.b is to include the " analytical methods used to determine the operating limits", Reference 2 is appropriate to be included in the Section 6 list of methods.
Basis for Exicency The revised MCPR Safety Limit and the addition of the SPC Topical Report for application of ANFB Critical Power Correlation to GE fuel to Section 6.9.A.6.b are necessary to support Quad Cities Unit 2 Cycle 15 startup. Therefore, this amendment is requested to be processed as an exigent Technical Specification amendment.
Approval of the proposed amendment on an exigent basis is warranted due to the short time frame between when SPC determined the need for a larger data base for determining the additive constant uncertainty (March 20,1997) and when Quad Cities Unit 2 Cycle 15 is scheduled to startup. This short time frame, combined with the time necessary to develop this request, will not allow the normal 30 day period for public comment. This condition was unavoidable and not created by the failure to make a timely application for a Technical Specification Amendment. This exigent amendment is needed to support Quad Cities Unit 2 Cycle 15 operation, which is currently scheduled to begin on May 19,1997. Without approval of this exigent amendment, Quad Cities Unit 2 will not be able to resume power operation.
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ATTACIIMENT A DESCRIPTION AND EVALUATION OF PROPOSED CllANGES '
ESK-97-089 (Page 9 of 14)
- 6. Description of the Revised Requirements Change to Minimum Critical Power Ratio (MCPR) Safejy Limit and MCPR Safety Limit Bases The new MCPR Safety Limit calculated for Quad Cities Unit 2 Cycle 15 is 1.10. This MCPR Safety Limit is based on SPC calculations using a conservative additive constant uncertainty for ATRIUM-9B fuel.
Addition of Siemens Power Corporation's (SPC) methodology for Application of the l ANFB Critical Power Correlation to Coresident GE Fuel for Ouad Cities Unit 2 Cycle l 15 to Section 6.9. A.6.b i
The following document is proposed for inclusion in Technical Specification Section l 6.9. A.6.b:
Comed letter," Comed Response to NRC StafrRequest for Additional Information (RAI) Regarding the Application of Siemens Power Corporation ANFB Critical Power Correlation to Coresident General Electric Fuel for LaSalle Unit 2 Cycle 8 and Quad
- Cities Unit 2 Cycle 15, NRC Docket No.'s 50-373/374 and 50-254/265", J.B. Hosmer to U.S. NRC, July 2,1996, transmitting the topical report, Application of the ANFB Critical Power Correlation to Coresident GE Fuelfor Quad Cities Unit 2 Cycle 15, EMF-96-051(P), Siemens Power Corporation - Nuclear Division, May 1996, and related information.
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ATTACIIMENT A l DESCRIPTION AND EVALUATION OF PROPOSED CIIANGES l ESK-97-089 (Page 10 of 14) ,
- 7. Basis for the Revised Requirements Change to Minimum Critical Power Ratio (MCPR) Safety Limit and MCPR Safety Limit Bases Quad Cities Nuclear Power Station will be using Siemens Power Corporation - Nuclear Division (SPC) ATRIUM-9B fuel assemblies for Unit 2 Cycle 15 operation. SPC has i determined that the need exists to increase the size of the data base for determining the additive constant uncertainty for SPC's ATRIUM-9B 9x9 fuel designs with an internal water channel. SPC has calculated a new additive constant uncertainty for the ATRIUM-9B fuel by including additional experimental data from critical power tests for other fuel designs which share many of the same design features as the ATRIUM-98 l
design. The information was selected to address the full operating range of the fuel.
Reference 5 details the statistical analysis performed on the data. l SPC increased its ATRIUM-9B critical power test data base from 125 data points at 2
1000 psi with mass fluxes ranging from 0.5 to 1.5 Mlb/hr-ft , to 527 data points that cover a much wider range of pressures, mass fluxes, and axial power shapes.
The Experimental Critical Power Ratio (ECPR) and the standard deviation of the ECPR for each of the 527 data points were statistically examined by an Analysis of Variance. l The results of the Analysis of Variance for each of the pressures are a mean ECPR, a standard deviation of ECPR degrees of freedom, and equivalent sample size.
The overall uncertainty for CPR is statistically calculated using the standard deviation of the pooled data and the variance between the means associated with the axial power shapes. An upper 95% confidence limit standard deviation is calculated based on Chi-Square for the calculated degrees of freedom The overall standard deviation in ECPR is converted to an additive constant uncertainty. The conversion is derived from the ratios l of the ANFB correlation standard deviation to the additive constant standard deviation !
for the ATRIUM-9B data. This calculation is explained in detail in Reference 5 and summarized below:
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l Overall Standard Deviation = 0.0388 (Reference 5)
CPR standard deviation = 1.99 (Reference 5) 0.0388
0.0195
1.99 Additive Constant Uncertainty for ATRIUM -9B (data set of 527 points) l
l ATTACIIMENT A l DESCRIPTION AND EVALUATION OF PROPOSED CIIANGES ESK-97-089 (Page 11 of 14)
For additional conservatism, this calculated additive constant uncertainty is not directly applied to the MCPR Safety Limit calculation. Rather, a conservative additive constant uncertainty is used to calculate a new MCPR Safety Limit for Quad Cities Unit 2 Cycle
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The difference is calculated betwr;en the additive constant uncertainties after and prior to i the data set being expanded to include 527 points. This difference is then conservatively !
doubled. The doubled difference in the additive constant uncertainties is then added to j the additive constant uncertainty prior to the expansion of the data set (based on 125 l data points). This calculation is summarized below:
l Additive Constant Uncenainty for ATRIUM - 9B (data set of 125 points) = 0.010 Additive Constant Uncenainty for ATRIUM - 9B (data set of 527 points) = 0.0195 Additive Constant Uncenainty used to calculate more conservative Q2C15 MCPR Safety Limit =
(0.010 + 2(0.0195 - 0.010)) = 0.029 The resulting conservative additive constant uncertainty of 0.029 for ATRIUM-9B fuel, is used to calculate the new MCPR Safety Limit of 1.10 for Quad Cities Unit 2 Cycle 15.
This proposed amendment increases the MCPR Safety Limit for Quad Cities Unit 2 from 1.07 to 1.10. This submittal also proposes to include an explanation of why the Unit 1 MCPR Safety Limit (1.07) and the Unit 2 MCPR Safety Limit (1.10) are different in the MCPR Safety Limit Bases. The plant will operate with this conservatism until the NRC approval of Reference 5. After NRC approval of Reference 5, Comed intends to submit a Technical Specification change to lower the MCPR Safety Limit (based only on the additive constant uncertainty for ATRIUM-9B fuel data set of 527 points) and add the Reference 5 topical to Section 6 of the Technical Specifications.
Addition of Siemens Power Corooration's (SPC) methodolouv for Apolication of the ANFB Critical Power Correlation to Coresident GE Fuel for Ouad Cities Unit 2 Cycle 15 to Section 6.9. A.6.b l In addition to the ATRIUM-9B fuel, Quad Cities Unit 2 Cycle 15 will operate with l previously exposed GE fuel types (GE9 and GE10). Because the ANFB critical power
! correlation (Reference 1) will be used in establishing and monitoring Minimum Critical l Power Ratio (MCPR) limits for both the SPC fuel and the coresident GE fuel, ANFB additive constants and uncertainties were developed for GE9/10 fuel. These additive
, constants and uncertainties are calculated and discussed in Reference 2.
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ATTACIIMENT A DESCRIPTION AND EVALUATION OF PROPOSED CIIANGES ESK-97-089
- (Page 12 of 14)
Therefore, this submittal proposes to add Reference 2, Siemens Power Corporation's l (SPC) methodology for application of the ANFB Critical Power Correlation to l coresident GE Fuel for Quad Cities Unit 2 Cycle 15, to Section 6.9. A.6.b of the i Technical Specifications. This SPC topical report and Comed transmittal letter describe the methodology used to determine the additive constants and the associated uncertainty for application of the ANFB Critical Power Correlation to GE9 and GE10 fuel. The additive constant uncertainty for the GE fuel is included in calculating the Quad Cities ,
t Unit 2 Cycle 15 MCPR Safety Limit. Because the Safety Limit is used to calculate the operating limit and because Generic Letter 88-16 indicated that Section 6.9.A.6.b is to include the " analytical methods used to determine the operating limits", it is appropriate that Reference 2 be included in the list of references in Section 6.
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ATTACIIMENT A DESCRIPTION AND EVALUATION OF PROPOSED CHANGES ESK-97-089 (Page 13 of 14) l 8. Schedule Approval of the proposed amendment on an exigent basis is warranted due to the short time frame between when SPC determined the need for a larger data base for determining the additive constant uncertainty (March 20,1997) and when Quad Cities Unit 2 Cycle 15 is scheduled to startup (May 19,1997). This short time frame, combined with the time necessary to develop this request, will not allow the normal 30 day period for public comment. This cond; tion was unavoidable and not created by the i failure to make a timely application for a Technical Specification Amendment. Without l approval of this exigent amendment, Quad Cities Unit 2 will not be able to resume power operation.
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ATTACIIMENT A DESCRIPTION AND EVALUATION OF PROPOSED CIIANGES ESK-97-089 i (Page 14 of 14) l
- 9. References
- 1) ANFB CriticalPower Correkstion, ANF-1125(P)(A) with Supplements I and 2, Advanced Nuclear Fuels Corporation, April 1990. i l 2) Comed letter," Comed Response to NRC Staff Request for Additional !
Information (RAI) Regarding the Application of Siemens Power Corporation ANFB Critical Power Correlation to Coresident General Electric Fuel for LaSalle Unit 2 Cycle 8 and Quad Cities Unit 2 Cycle 15, NRC Docket No.'s 50-373/374 and 50-254/265", J.B. Hosmer to U.S. NRC, July 2,1996, transmitting the ;
topical report, Application of the ANFB CriticalPower Correlation to l Coresident GE Fuelfor Quad Cities Unit 2 Cycle 15 EMF-96-051(P), Siemens j Power Corporation - Nuclear Division, May 1996, and related information.
l 3) E.S. Kraft, Jr. to USNRC letter dated June 10,1996, Quad Cities Nuclear Power l Stations Units 1 and 2, Application for Amendment Request to Facility Operating !
Licenses DPR-29 and DPR-30 Technical Specification Changes for Siemens j l Power Corporation (SPC) Fuel Transition, NRC Docket Nos. 50-254 and 50- ;
265. I
- 4) E.S. Kraft, Jr. to USNRC letter dated February 17,1997, Quad Cities Nuclear i Power Stations Units 1 and 2 Supplement to Application for Amendment of l Facility Operating License DPR-29 and DPR-30 Technical Specifications NRC Docket Nos. 50-254 and 50-265. l l 5) ANFB Critical Power Correlation Uncertainty For Limited Data Sets, ANF-l 1125(P), Supplement 1, Appendix D, Siemens Power Corporation - Nuclear l Division, Submitted on April 18,1997.
l l 6) Advanced Nuclear Fuels Corporation Critical Power Methodologyfor Boiling l Water Reactors'AdvancedNuclear Fuels Corporation Critical Power Methodologyfor Boiling Water Reactors: Methodologyfor Analysis of l Assembly Channel Bowing Effects'NRC Correspondence, ANF-524(P)(A);
l Revision 2, Supplement 1 Revision 2, Supplernent 2, Advance Nuclear Fuels '
Corporation, November 1990.
- 7) Siemens Power Corporation letter," Interim Use ofIncreased ANFB Additive Constant Uncertainty", HDC:97:033, H.D. Curet to Document Control Desk, April 18,1997.
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ATTACIIMENT B l
SUMMARY
OF PROPOSED CIIANGES ESK-97-089 (Page 1 of 6) l Topic Affected Description of Change Pages l
l MCPR Safety 2-1 Increase the Unit 2 MCPR Safety Limit from 1.07 to 1.10 Limit l
MCPR Safety B 2-3 Insert an explanation why the Unit 1 MCPR Safety Limit l Limit Bases (1.07) is different from the Unit 2 MCPR Safety Limit I (1.10).
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1 l Section 6-16 Add Comed letter," Comed Response to NRC Staff :
6.9. A.6.b Request for Additional Information (RAI) Regarding the !
Methodology Application of Siemens Power Corporation ANFB Critical
( References Power Correlation to Coresident General Electric Fuel for l LaSalle Unit 2 Cyclc 8 and Quad Cities Unit 2 Cycle 15,
! NRC Docket No.'s 50-373/374 and 50-254/265", J.B.
Hosmer to U.S. NRC, July 2,1996, transmitting the topical report, Application of the ANFB Critical Power Correlation to Coresident GE Fuelfor Quad Cities Unit 2 Cycle 15, EMF-96-051(P), Siemens Power Corporation -
Nuclear Division, May 1996, and related information, to the list of references.*
- Note that revisions to Section 6.9. A.6.b are pending as described in the following previous transmittals relating to SPC fuel. The revisions made in this submittal take into account the changes proposed in previous transmittals. These transmittals are listed ,
below:
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- 1. E.S. Kraft, Jr. to USNRC letter dated June 10,1996, Quad Cities Nuclear Power Stations Units 1 and 2 '
Application for Amendment Request to Facility Operating I Licenses DPR-29 and DPR-30 Technical Specification Changes for l Siemens Power Corporation (SPC) Fuel Transition, l NRC Docket Nos. 50-254 and 50-265 l
- 2. E.S. Kraft, Jr. to USNRC letter dated Febniary 17,1997, l Quad Cities Nuclear Power Stations Units 1 and 2 Supplement to Application for Amendment of Facility Operating License DPR-29 and DPR-30 Technical Specifications NRC Docket Nos. 50-254 and 50-265
ATTACIIMENT B
SUMMARY
OF PROPOSED CIIANGES ESK-97-089 (Page 2 of 6)
Inserts to Quad Cities Technical Specifications Insert #1 Page B2-3, Section 2.1.B The Unit 1 MCPR Safety Limit is 1.07, based on General Electric methods for ;
calculating the MCPR Safety Limit. The Unit 2 MCPR Safety Limit is 1.10, based on i l Siemens Power Corporation (SPC) methods for calculating the MCPR Safety Limit. I t
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Page 6-16, Section 6.9.A.6.b (19) Comed letter," Comed Response to NRC Staff Request for Additional Information (RAI) Regarding the Application of Siemens Power Corporation ANFB Critical Power Correlation to Coresident General Electric Fuel for LaSalle l Unit 2 Cycle 8 and Quad Cities Unit 2 Cycle 15, NRC Docket No.'s 50-373/374 '
and 50-254/265", J.B. Hosmer to U.S. NRC, July 2,1996, transmitting the topical report, Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15, EMF-96-051(P), Siemens Power Corporation - Nuclear Division, May 1996, and related information.
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ATTACHMENT B
SUMMARY
OF PROPOSED CIIANGES ESK-97-089 (Page 3 of 6)
B. Summary of Proposed Changes Attached are the marked-up Technical Specification pages for Quad Cities Nuclear Power Station, as they are proposed in this submittal.
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, GAFETY LIMITS 2.1 l
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 7
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M SAFETY LIMITS -
THERMAL POWER, low Pressure or Low Flow l
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! 2.1. A THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, i
APPLICABILITY: OPERATIONAL MODE (s) 1 and 2. )
1 ACTION:
l With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specificaticn 6.7.
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- 1. J o Ar Us + 1 THERMAL POWER, Hioh Pressure and Hioh Flow w l
l i l 2.1.B The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.0*i' ith the i l reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow. During single recirculation loop operation, this MCPR limit shall be increased by 0.01.
APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.
l ACTION:
i With MCPR less than the above applicable limit and the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.
Amendment Nos. *8*
QUAD CITIES - UNITS 1 & 2 21 l .
j SAFETY LIMITS B 2.1 BASES i
approach. Much of the data indicates that BWR fuel can survive for an extended period in an environment of transition boiling.
L se r- + #/ j 2.1.C Reactor Coolant System Pressure The Safety Limit for the reactor coolant system pressure has been selected such that it is at a I pressure below which it can be shown that the integrity of the system is not endangered. The I reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of !
fission products. It is essential that the integrity of this system be protected by establishing a i pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.
The reactor coolant systern pressure Safety Limit of 1345 psig, as measured by the vessel steam space pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor vessel.
The 1375 psig value is derived from the design pressures of the reactor pressure vessel and coolant system piping. The rerpective design pressures are 1250 psig at 575*F and 1175 psig at 1 560*F. The pressure Safety Lirait was chosen as the lower of the pressure transients permitted !
by the applicable design codes, ASME Boiler and Pressure Vessel Code Section ill for the pressure vessel, and USASI B31.1 Code for the reactor coolant system piping. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 !
= 1375 psig), and the USASI Code permits pressure transients up to 20% cver design pressure (120% x 1175 = 1410 psig). The Safety Limit pressure of 1375 psig is referenced to the lowest elevation of the reactor vessel. The design pressure for the recirculation suction line piping (1175 psig) was chosen relative to the reactor vessel design pressure. Demonstrating compliance of peak vessel pressure with the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping with the USASI limit (1410 psig). Evaluation methodology to assure that this Safety Limit pressure is not exceeded for any reload is documented by the specific fuel vendor.
The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is a factor of 1.5 below the yield strength of 40,100 psi at 575*F. At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yield strength.
The relationships of stress levels to yield strength are comparable for the primary system piping and provides similar margin of protection at the established pressure Safety Limit.
The normal operating pressure of the reactor coolant system is nominally 1000 psig. Both pressure relief and safety relief valves have been installed to keep the reactor vessel peak pressure below 1375 psig. However no credit is taken for relief valves during the postulated full closure of all MSIVs without a direct (valve position switch) scram. Credit, however, is taken for the neutron flux scram. The indirect flux scram and safety valve actuation provide adequate margin below the allowable peak vessel pressure of 1375 psig.
Amendment Nos. 171 & W QUAD CITIES - UNITS 1 & 2 B23
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R: porting R2quirem nts 6.9 ADMINISTRATIVE CONTROLS (3) Commonwealth Edison Topical Report NFSR-0085, Supplement 1,
" Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons," (latest approved revision).
gh (4) Commonwealth Edison Topical Report NFSR-0085, Supplement 2,
" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing ;
Analyses," (latest approved revision).
- c. The core operating limits shall be determined so that all applicable limits (e.g., fuel ;
thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits I such as shutdown margin, and transient and accident analysis limits) of the safety l analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
6.9.B Special Reports Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report. '
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l QUAD CITIES - UNITS 1 & 2 6-16 Amendment Nos. 171 a u7 l
ATTAClIMENT C l EVAlf"ATION OF SIGNIFICANT IIA. RDS CONSIDERATIONS i ESK-97-089 (Page 1 of 6)
References for the Evaluation of Significant Hazards Considerations:
l 1) Comed letter," Comed Response to NRC Staff Request for Additional Information (RAI) Regarding the Application of Siemens Power Corporation ;
ANFB Critical Power Correlation to Coresident General Electric Fuel for '
l LaSalle Unit 2 Cycle 8 and Quad Cities Unit 2 efcle 15, NRC Docket No.'s l 50-373/374 and 50-254/265", J.B. Hosmer to U.S. NRC, July 2,1996, transmitting the topical report, Application of the ANFB CriticalPower Correlation to Coresident GE Fuelfor Quad Cities Unit 2 Cycle 15, EMF-96-051(P), Siemens Power Corporation - Nuclear Division, May 1996, and related information.
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- 2) SPC document, ANFB CriticalPower Correla' ion Uncertainty For Limited Data Sets, ANF-1125(P), Supplement 1, Appe, dix D, Siemens Power I Corporation - Nuclear Division, Submitted on April 18,1997.
- 3) Comed letter," Application of Siemen's Power Corporation ANFB Critical l Power Correlation to Coresident General Electric Fuel for LaSalle Unit 2 l Cycle 8", G.G. Benes to U.S. Nuclear Regulatory Commission, dated March 1 l 8,1996.
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- 4) NRC SER letter, " Safety Evaluation for Topical Report EMF-96-021(P), I Revision 1, ' Application of the ANFB Critical Power Correlation to Coresident GE Fuel for LaSalle Unit 2 Cycle 8' (TAC No. M94964)", D.M.
i Skay to I. Johnson, dated September 26,1996. l l
- 5) Siemens Power Corporation letter," Interim Use ofIncreased ANFB Additive i Constant Uncertainty", HDC:97:033, H.D. Curet to Document Control !
Desk, April 18,1997. !
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l The MCPR Safety Limit for Quad Cities Unit 2 Cycle 15 is proposed to be increased from 1.07 to 1.10. This increase is a result of applying a new ATRIUM-9B additive ,
constant uncertainty in the analyses for Quad Cities Unit 2 Cycle 15. This new additive I constant uncertainty was calculated based on an expanded pool of critical power test data for the ATRIUM-9B fuel design and is conservatively applied in the determination of the MCPR Safety Limit. The resulting MCPR Safety Limit,1.10, will ensure conservative operation of Quad Cities Unit 2 Cycle 15. I 1
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ATTACIIMENT C EVALUATION OF SIGNIFICANT IIA 7mARDS CONSIDERATIONS ESK-97-089 l (Page 2 of 6) l
- b. Addition of Siemens Power Corporation's (SPC) methodology for application of the ANFB Critical Power Correlation to coresident GE Fuel for Quad Cities Unit 2 Cycle 15 to Section 6.9. A.6.b This submittal proposes to add Reference 1, Siemens Power Corporation's (SPC) methodology for application of the ANFB Critical Power Correlation to coresident GE Fuel for Quad Cities Unit 2 Cycle 15, to Section 6.9.A.6.b of the Technical Specifications. This SPC topical report and Comed transmittalletter describe the methodology used to determine the additive constants and the associated uncertainty for application of the ANFB Critical Power Correlation to the Quad Cities Unit 2 Cycle 15 GE9 and GE10 fuel. The additive constant uncertainty for the GE fuel is also included in calculating the Quad Cities Unit 2 Cycle 15 MCPR Safety Limit. Because the Safety Limit is used to calculate the operating limit and because Generic Letter 88-16 indicated that Section 6.9. A.6.b is to include the " analytical methods used to determine the
- operating limits", it is appropriate that Reference 1 be included in the list of methods in Sectnn 6.
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l Comed has evaluated the proposed Technical Specification amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazard consideration established in 10CFR50.92(c), operation of i Quad Cities Unit I and 2 in accordance with the proposed amendments will not represent a significant hazards consideration for the following reasons:
l These changes do not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
The probability of an evaluated accident is derived from the probabilities of the j individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established consistent with NRC approved l methods to ensure that fuel performance during normal, transient, and accident l conditions is acceptable. The proposed Technical Specifications amendment conservatively establishes the MCPR Safety Limit for Quad Cities Unit 2, such
- that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences. Additionally, methodologies are being added to the Section 6.9. A.6.b list of methodologies utilized in determining core operating limits.
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ATTACllMENT C j EVALUATION OF SIGNIFICANT IIAZARDS CONSIDERATIONS ESK-97-089 (Page 3 of 6) l
- a. MCPR Safety Limit and MCPR Safety Limit Bases Change The probability of an evaluated accident is not increased by increasing the MCPR Safety Limit to 1.10 and changing the MCPR Safety Limit Bases. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected. ;
l l This Technical Specification amendment proposes to change the MCPR Safety Limit to protect the fuel during normal operation as well as during any transients
, or anticipated operational occurrences. The method that is used to determine the ATRIUM-9B additive constant uncertainty is conservative, such that, the resulting MCPR Safety Limit is high enough to ensure that less than 0.1% of the fuel rods are expected to experience boiling transition if the limit is not violated.
Operational limits will be established based on the proposed MCPR Safety Limit to ensure that the MCPR Safety Limit is not violated during all modes of
- operation. This will ensure that the fuel design safety criteria, more than 99.9%
l of the fuel rods avoiding transition boiling during normal operation as well as anticipated operational occurrences, is met. The method for calculating an
- ATRIUM-9B additive constant uncertainty, is described in Reference 2 and is
! based on an expanded pool of data for the ATRIUM-9B fuel design (527 data points). The additive constant uncenainty from Reference 2 is then used to determine the change from the additive constant uncertainty using the original pool of data (125 data points). This difference is conservatively doubled and added to the additive constant uncertainty using the original pool of data (125 data points). Reference 5 documents the conservative interim approach of doubling the difference in additive constant uncertainties. The resulting additive constant uncertainty is used to determine the Quad Cities Unit 2 Cycle 15 MCPR !
Safety Limit. Since the new MCPR Safety Limit was determined using a l conservative ATRIUM-9B additive constant uncertainty; and the operability of plant systems designed to mitigate any consequences of accidents have not changed, the consequences of an accident previously evaluated are not expected to increase.
- b. Addition of Siemens Power Corporation's (SPC) methodology for j Application of the ANFB Critical Power Correlation to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15 to Section 6.9. A.6.b l
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ATTACHMENT C i
' ' EVALUATION OF SIGNIFICANT
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IIAZARDS CONSIDERATIONS ESK-97-089 l l (Page 4 of 6) ;
l I The probability of an evaluated accident is not increased by adding Reference 1, to Section 6.9.A.6.b. Reference 1 describes the methodology used to determine i the additive constants and the associated uncertainty of the Quad Cities Unit 2 Cycle 15 GE9 and GE10 fuel for the ANFB critical power correlation. The ,
additive constant and the associated uncenainties for the GE9 and GE10 fuel are !
used to calculate the MCPR Safety Limit, which in turn is used to establish the ;
MCPR operating limit for Quad Cities Unit 2 Cycle 15 operation. Therefore, l
l adding Reference 1 to Section 6.9.A.6.b of the Technical Specifications updates the Reference list to include a methodology used for determining Quad Cities Unit 2 Cycle 15 operationallimits.
l Adding Reference 1 to the Reference list in Section 6.9.A.6.b also will not increase the consequences of an accident previously evaluated. Reference 1 l determines the additive constants and the associated uncertainty for the GE fuel in Quad Cities Unit 2 Cycle 15. It also provides input for determining the MCPR Safety Limit. Because Reference I contains conservative methods and l
calculations and because the operability of plant systems designed to mitigate any l consequences of accidents have not changed, the consequences of an accident previously evaluated will not increase.
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l 2. Create the possibility of a new or different kind of accident from any l l accident previously evaluated: ;
Creation of the possibility of a new or different kind of accident would require i the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation. This Technical Specification submittal does not involve any modifications of the plant configuration or allowable modes .
of operation. This Technical Specification submittal involves a) an added !
conservatism in the Quad Cities Unit 2 MCPR Safety Limit due to analytical l changes and use of an expanded database, and b) an additional reference incorporated in Section 6.9.A.6.b describing the methodology used to determine the additive constants and additive constant uncertainty for GE9 and GE10 fuel for Quad Cities Unit 2 Cycle 15. Therefore, no new precursors of an accident are created and no new or different kinds of accidents are created.
ATTACIIMENT C EVALUATION OF SIGNIFICANT i
IIAZARDS CONSIDERATIONS l ESK-97-089 l (Page 5 of 6) 1
- 3. Involve a significant reduction in the margin of safety for the following reasons:
The MCPR Safety Limit provides a margin of safety by ensuring that less than O.1% of the rods are expected to be in boiling transition if the MCPR limit is not l violated. The proposed Technical Specification amendment reflects MCPR Safety Limit results from conservative calculations by SPC using the new .
ATRIUM-9B additive constant uncertainty. These new ATRIUM-9B additive constant uncertainty calculations are based on a larger pool of data than previous
- calculations (527 data points versus 125 data points). Additionally, the additive ;
constant uncertainty resulting from the statistical analyses of the larger pool of data is conservatively applied to calculate a new MCPR Safety Limit of 1.10, i which is more restrictive than the current MCPR Safety Limit of 1.07.
l SPC has increased its ATRIUM-9B critical power test data base from 125 data points at 1000 psi with mass fluxes ranging from 0.5 to 1.5 Mlb/hr-ff, to 527 data points that cover a wider range of operating pressures, flows, and axial l power shapes.
The Experimental Critical Power Ratio (ECPR) and the standard deviation of the l
, ECPR for each of the 527 data points are statistically examined by an Analysis of I l Variance. The results of the Analysis of Variance of the Pressure Groups are a i mean ECPR, a standard deviation of ECPR, degrees of freedom, and equivalent sample size.
The overall uncertainty for CPR is statistically calculated using the standard deviation of the pooled data and the variance between the means associated with the axial power shapes. An upper 95% confidence limit standard deviation is calculated based on Chi-Square for the calculated degrees of freedom. This ]
overall standard deviation in ECPR is converted to an additive constant l uncertainty. This conversion is derived from the ratios of the ANFB correlation standard deviation to the additive constant standard deviation for the ATRIUM-9B data.
This calculated additive constant uncertainty is not directly applied to the MCPR Safety Limit calculation. A conservative ATRIUM-98 additive constant l uncertainty is used to calculate a new MCPR Safety Limit for Quad Cities Unit 2 l Cycle 15.
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- The difference is calculated between the additive constant uncertainties after and prior to the data set being expanded to include 527 points. This difference is then conservatively doubled and added to the additive constant uncertainty prior to the expansion of the data set (based on 125 data points).
ATTACIIMENT C EVALUATION OF SIGNIFICANT IIAZARDS CONSIDERATIONS ESK-97-089 (Page 6 of 6)
The resulting additive constant uncertainty,0.029, is used to calculate a new MCPR Safety Limit value of 1.10 for Quad Cities Unit 2 Cycle 15.
Because a conservative method is used to apply the ATRIUM-9B additive constant uncertainty to the MCPR Safety Limit calculation, a decrease in the margin to safety will not occur due to changing the MCPR Safety Limit. The revised Safety Limit will ensure the appropriate level of fuel protection.
Additionally, operational limits will be established based on the proposed MCPR Safety Limit to ensure that the MCPR Safety Limit is not violated during all modes of operation. This will ensure that the fuel design safety criteria, more than 99.9% of the fuel rods avoiding transition boiling during normal operation l as well as anticipated operational occurrences, is met.
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The man;in of safety is not decreased by adding the Reference to Section 6.9. A.6.b of Siemens Power Corporation's (SPC) methodology for application ef l the ANFB Critical Power Correlation to coresident GE Fuel for Quad Cities Unit i 2 Cycle 15. While this methodology is in review by the NRC, and pending approval for application to Quad Cities Unit 2 Cycle 15, it is the same methodology previously reviewed and approved for use at LaSalle Unit 2 (References 3 and 4).
This proposed amendment does not involve a significant relaxation of the criteria used to -.
establish the safety limits, a significant relaxation of the bases for the limiting safety system settings, or a significant relaxation of the bases for the limiting conditions for operations. Therefore, based on the guidance provided in 10CFR50.92(c), the proposed change does not constitute a significant hazards consideration.
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ATTACIIMENT D l
' i ENVIRONMENTAL ASSESSMENT APPLICABILITY REVIEW l ESK-97-089 i (Page1of1) i Comed has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10CFR51.21. It has been determined that the proposed changes meet the criteria for !
categorical exclusion as provided for under 10CFR51.22(c)(9). This conclusion has i been determined because the changes requested do not pose significant hazards considerations or do not involve a significant increase in the amounts, and no significant l changes in the types of any efIluents that may be released off-site. Additionally, this request does not involve a significant increase in individual or cumulative occupational i radiation exposure. ;
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