BSEP 02-0174, Request for License Amendment Regarding Technical Specification 2.1.1.2, Reactor Core Safety Limits Minimum Critical Power Ratio Safety Limit
| ML023290354 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/07/2002 |
| From: | Keenan J Carolina Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 02-0174, TSC-2002-10 | |
| Download: ML023290354 (20) | |
Text
CP&L AProgress Energy Company John S Keenan Vice President Brunswick Nuclear Plant NOV 0 7 2002 SERIAL: BSEP 02-0174 10 CFR 50.90 TSC-2002-10 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Brunswick Steam Electric Plant, Unit No. 2 Docket No. 50-324/License No. DPR-62 Request For License Amendment Regarding Technical Specification 2.1.1.2, Reactor Core Safety Limits Minimum Critical Power Ratio Safety Limit Ladies and Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power & Light (CP&L) Company is requesting a revision to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Unit No. 2. The proposed license amendment revises the Minimum Critical Power Ratio (MCPR) Safety Limit values contained in Technical Specification 2.1.1.2 from 1.09 to 1.11 for two recirculation loop operation and from 1.10 to 1.13 for single recirculation loop operation.
An evaluation of the proposed license amendment is provided in Enclosure 1, and is supported by a Global Nuclear Fuel - Americas, LLC (GNF-A) document in Enclosure 2 which provides a summary of analysis input parameters and results of a comparison of the revised Unit 2 Cycle 16 and previous Unit 2 Cycle 15 MCPR Safety Limit values. Some of the information contained in the document is considered proprietary by GNF-A and should be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4) and 10 CFR 2.790(a)(4). An affidavit attesting to this fact is provided in Enclosure 3. A non proprietary version of the GNF-A document is provided in Enclosure 4.
Refueling Outage 15 for Unit 2 (i.e., designated as B216R1) is scheduled to begin March 8, 2003. Unit 2 will be unable to resume power operation without receipt of approval for the revised MCPR Safety Limit values in Technical Specification 2.1.1.2.
Therefore, the NRC is requested to issue the requested license amendment no later than March 1, 2003. CP&L requests that the amendment, once approved, be issued effective immediately, to be implemented prior to resuming operation from Unit 2 Refueling Outage 15 for Cycle 16.
PO Box 10429 Southport, NC 28461 T> 9104572496 F> 9104572803
Document Control Desk BSEP 02-0174 / Page 2 CP&L has evaluated the proposed change in accordance with 10 CFR 50.91(a)(1), using the criteria in 10 CFR 50.92(c), and determined that this change involves no significant hazards considerations. CP&L has reviewed the proposed license amendment and determined the changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, in accordance with 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
In accordance with 10 CFR 50.91(b), CP&L is providing the State of North Carolina a copy of the proposed license amendment.
Please refer any questions regarding this submittal to Mr. Edward T. O'Neil, Manager - Regulatory Affairs, at (910) 457-3512.
Sincerely,
Document Control Desk BSEP 02-0174 / Page 3 WRM/wrm
Enclosures:
- 1. Evaluation of Proposed License Amendment Request
- 2. Global Nuclear Fuel - Americas, LLC Document Entitled "Additional Information Regarding the Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16" dated October 17, 2002 (Proprietary Information)
- 4. Global Nuclear Fuel - Americas, LLC Document Entitled "Additional Information Regarding the Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16" dated October 17, 2002 (Non-Proprietary Version)
- 5. Marked-up Technical Specification Page - Unit 2
- 6. Typed Technical Specification Page - Unit 2
- 7. List of Regulatory Commitments John S. Keenan, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Carolina Power &
Light Company.
Notary (Seal)
My commission expires: ti" 0-4 LA~j 4-.s."j'"r~ - -"
Document Control Desk BSEP 02-0174 / Page 4 cc (with enclosures):
U. S. Nuclear Regulatory Commission, Region H ATFN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Theodore A. Easlick, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Mr. Leonard N. Olshan (Mail Stop OWFN 8H12) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 cc (without Enclosure 2):
Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Ms. Beverly 0. Hall, Section Chief Radiation Protection Section, Division of Radiation Protection North Carolina Department of Environment and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221
BSEP 02-0174 Enclosure 1 Page 1 of 5 Evaluation of Proposed License Amendment Request
Subject:
Technical Specification 2.1.1.2, Reactor Core Safety Limits Minimum Critical Power Ratio Safety Limit 1.0 Description This letter is a request for approval of an amendment to Operating License DPR-62 for Carolina Power & Light (CP&L) Company's Brunswick Steam Electric Plant (BSEP), Unit No. 2. The proposed change revises the Minimum Critical Power Ratio (MCPR) Safety Limit values contained in Technical Specification 2.1.1.2 for Unit 2 due to the loading of additional GE14 fuel bundles. Specifically, the MCPR Safety Limit values contained in Technical Specification 2.1.1.2 must be revised from 1.09 to 1.11 for two recirculation loop operation and from 1.10 to 1.13 for single recirculation loop operation.
Refueling Outage 15 for Unit 2 is scheduled to begin March 8, 2003. Unit 2 will be unable to resume power operation without receipt of approval for the revised MCPR Safety Limit values in Technical Specification 2.1.1.2. Therefore, the NRC is requested to issue the requested license amendment no later than March 1, 2003.
2.0 Proposed Change The proposed amendment revises the MCPR Safety Limit values contained in Technical Specification 2.1.1.2 from 1.09 to 1.11 for two recirculation loop operation and from 1.10 to 1.13 for single recirculation loop operation. The MCPR Safety Limit values are being revised for Unit 2 based on the loading of new reload GE14 fuel bundles. The proposed Technical Specification states:
2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.2 With the reactor steam dome pressure Ž 785 psig and core flow > 10% rated core flow:
MCPR shall be > 1.11 for two recirculation loop operation or > 1.13 for single recirculation loop operation.
"BSEP02-0174 Enclosure 1 Page 2 of 5 3.0 Background Technical Specification 2.1.1.2 establishes MCPR Safety Limit values, which if met, ensure that no mechanistic fuel damage is calculated to occur. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling (i.e., transition boiling) have been used to designate the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to boiling water reactor fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. The MCPR Safety Limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9 percent of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.
The design process for each operating cycle core involves verification that appropriate safety limit values for the MCPR exist. Unit 2 Cycle 15 was the first operating cycle involving the loading of the GEl4 fuel. For Unit 2 Cycle 15, the core design did not require the MCPR Safety Limit values in Technical Specification 2.1.1.2 to be revised. For Cycle 16, additional GE14 fuel is being used in the core design as a replacement for some currently loaded GE13 fuel.
Evaluation of the Unit 2 Cycle 16 core design has determined that a revision of the MCPR Safety Limit values is necessary.
4.0 Technical Analysis The Global Nuclear Fuel - Americas, LLC (GNF-A) methodology for MCPR Safety Limit determination for each fuel design is contained in topical report NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-II)," Revision 14,and U.S.
Supplement, NEDE-24011-P-A-14-US, June 2000, which incorporates Amendment 25 (i.e.,
Reference 1). To address NRC concerns relating to the methodologies and procedures for determining cycle-specific MCPR Safety Limits, GNF-A (i.e., under the corporate name of General Electric) submitted several topical reports for NRC review and approval. These topical reports include: (1) a description of the procedures used to account for the reload-specific core design and operation in determining the cycle-specific MCPR Safety Limit in NEDC-32601P, "Methodology and Uncertainties for Safety Limit MCPR Evaluations;" (2) the power distribution uncertainty for the new General Electric 3D-MONICORE core surveillance system in NEDC-32694P, "Power Distribution Uncertainties for Safety Limit MCPR Evaluation;" and (3) the methodology and uncertainties required for the implementation of cycle-specific MCPR Safety Limits in Amendment 25 to NEDE-24011-P-A. By letter dated March 11, 1999, from Frank Akstulewicz, NRC, to Glen Watford, General Electric (i.e., Reference 2), the NRC approved the use of Amendment 25 to NEDE-24011-P-A.
The revised MCPR Safety Limit analysis for BSEP, Unit 2 has been performed by GNF-A using the NRC-approved methods and procedures described in topical report NEDE-24011-P-A. Use of the NRC-approved methods ensures that the resulting MCPR Safety Limit values satisfy the
BSEP 02-0174 Enclosure 1 Page 3 of 5 fuel design safety criterion that more than 99.9 percent of the fuel rods in the core avoid boiling transition if the safety limit is not violated. As a result, the proposed MCPR Safety Limit value changes do not adversely impact any safety analysis assumptions or results. A summary of the relevant input parameters and results of a comparison of the revised Unit 2 Cycle 16 and previous Unit 2 Cycle 15 MCPR Safety Limit values is provided in Enclosure 2.
5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration CP&L has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The MCPR Safety Limit values are calculated to ensure that greater than 99.9 percent of the fuel rods in the core avoid transition boiling during any plant operation if the safety limit is not violated. The derivation of the MCPR Safety Limit values specified in the Technical Specifications, and their use to determine cycle-specific thermal limits, has been performed using the methodology discussed in "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-14 (i.e., GESTAR-il), and U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which incorporates Amendment 25. Amendment 25 was approved by the NRC in a March 11, 1999, safety evaluation report. Operational MCPR limits are applied that ensure the MCPR Safety Limit is not exceeded during all modes of operation and anticipated operational occurrences.
The revised MCPR Safety Limit values do not affect the operability of any plant systems nor do these revised values compromise any fuel performance limits; therefore, the probability of fuel damage will not be increased as a result of this change.
The MCPR Safety Limit values do not impact the source term or pathways assumed in accidents previously evaluated, and there are no adverse effects on the factors contributing to offsite or onsite radiological doses. In addition, the revised MCPR Safety Limit values do not affect the performance of any equipment used to mitigate the consequences of a previously evaluated accident and do not affect setpoints that initiate protective or mitigative actions.
Therefore, the proposed Technical Specification change does not involve a significant increase in the probability or consequences of a previously evaluated accident.
BSEP 02-0174 Enclosure 1 Page 4 of 5
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No Creation of the possibility of a new or different kind of accident would require 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. The proposed revision of the MCPR Safety Limit values does not involve any facility modifications, and plant equipment will not be operated in a different manner. No new initiating events or transients will result from the revised MCPR Safety Limit values.
As a result, no new failure modes are being introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No The margin of safety is established through the design of the plant structures, systems, and components; through the parameters within which the plant is operated; through the establishment of setpoints for actuation of equipment relied upon to respond to an event; and through margins contained within the safety analyses. The revised MCPR Safety Limit values will not adversely impact the performance of plant structures, systems, components, and setpoints relied upon to respond to mitigate an accident or transient. The MCPR Safety Limit values are calculated to ensure that greater than 99.9 percent of the fuel rods in the core avoid transition boiling during any plant operation if the safety limit is not violated, thereby ensuring that fuel cladding integrity is maintained. The revised MCPR Safety Limit values have been calculated using NRC approved methods and procedures and preserve the existing margin to transition boiling. Based on the assurance that the fuel design criteria are being met, the revised MCPR Safety Limit values do not involve a reduction in a margin of safety.
Based on the above, CP&L has concluded that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c)(1) requires that safety limits be included in the plant Technical Specifications.
Therefore, the MCPR Safety Limit is included in the BSEP, Unit 2 Technical Specifications. The MCPR Safety Limit values have been determined in accordance with NRC approved methodology described in "General Electric Standard Application for Reactor Fuel,"
BSEP 02-0174 Enclosure 1 Page 5 of 5 NEDE-240 11-P-A-14 (i.e., GESTAR-II), and U.S. Supplement, NEDE-240 11-P-A-14-US, June 2000.
6.0 Environmental Considerations A review has determined that the proposed license amendment changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
7.0 References
- 1. General Electric Licensing Topical Report NEDE 24011-P-A, "General Electric Standard Application for Reactor Fuel" (i.e., GESTAR II), Revision 14, and U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which incorporates Amendment 25.
- 2. Letter from Mr. Frank Akstulewicz, U. S. Nuclear Regulatory Commission, to Mr. Glen A. Watford, General Electric, "Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations;NEDC-32694P, Power DistributionUncertaintiesfor Safety Limit MCPR Evaluations;and Amendment 25 to NEDE-24011-P-A on Cycle Specific Safety Limit MCPR," TAC Nos. M97490, M99069, and M97491, March 11, 1999.
Precedents
- 3. Letter from the U. S. Nuclear Regulatory Commission to Mr. J. S. Keenan, "Brunswick Steam Electric Plant, Unit 1 - Issuance of Amendment Regarding Revision of Safety Limit Minimum Critical Power Ratio (TAC No. MB2952)," dated March 22, 2002.
BSEP 02-0174 Enclosure 4 Global Nuclear Fuel - Americas, LLC Document Entitled "Additional Information Regarding the Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16" dated October 17, 2002 (Non-Proprietary Version)
Attachment Additional Information Regarding the 17 October 2002 Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16 References
[1] Letter, Frank Akstulewicz (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations;NEDC-32694P, PowerDistribution Uncertaintiesfor Safety Limit MCPR Evaluation;and Amendment 25 to NEDE-2401 1-P-A on Cycle Specific Safety Limit MCPR,"
(TAC Nos. M97490, M99069 and M97491), March 11, 1999.
[2] Letter, Thomas H. Essig (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Report NEDC-32505P, Revision 1, R-Factor CalculationMethodfor GEl1, GEl2 and GEl3 Fuel," (TAC Nos. M99070 and M95081), January 11, 1999.
[3] General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977.
[4] Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention tc R. Pulsifer (NRC), "Confirmation of 10x10 Fuel Design Applicability to Improved SLMCPR, Power Distribution and R-Factor Methodologies", FLN-2001-016, September 24, 2001.
[5] Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue (NRC), "Confirmation of the Applicability of the GEXL14 Correlation and Associated R-Factor Methodology for Calculating SLMCPR Values in Cores Containing GE14 Fuel", FLN-2001-017, October 1, 2001.
[6] Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue (NRC), "Final Presentation Material for GEXL Presentation- February 11, 2002", FLN-2002-004, February 12, 2002.
Comparison of Brunswick Unit 2 SLMCPR Values for Cycles 16 and 15 Table 1 summarizes the relevant input parameters and results of the safety limit MCPR (SLMCPR) determination for the Brunswick Unit 2 Cycle 16 and Cycle 15 cores. The SLMCPR evaluations were performed using NRC approved methods and uncertainties"]. These evaluations yield different calculated SLMCPR values because different inputs were used. The quantities that have been shown to have some impact on the determination of the SLMCPR are provided.
In comparing the Brunswick Unit 2 Cycle 16 and Cycle 15 SLMCPR values it is important to note the impact of the differences in the core and bundle designs. These differences are summarized in Table 1.
In general, the calculated safety limit is dominated by two key parameters: (1) flatness of the core bundle-by-bundle MCFR distributions and (2) flatness of the bundle pin-by-pin power/R-factor distributions. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR.
((I))
The uncontrolled bundle pin-by-pin power distributions were compared between the Brunswick Unit 2 Cycle 16 bundles and the Cycle 15 bundles. Pin-by-pin power distributions are characterized in terms of R-factors using the NRC approved methodology12]. For the Brunswick Unit 2 Cycle 16 limiting case analyzed at EOR-1K, (( )) the Brunswick Unit 2 Cycle 16 bundles are slightly flatter than the bundles used for the Cycle 15 SLMCPR analysis.
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Attachment Additional Information Regarding the 17 October 2002 Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16 The revised power distribution model has been justified, reviewed and approved by the NRC (reference NEDC-32601P-A). The conservatism that remains even when applying the revised model to calculate a lower SLMCPR was documented as part of the NRC review and approval. It was noted on page A-24 of NEDC-32601P-A (( ))
Summary
(( )) have been used to compare quantities that impact the calculated SLMCPR value. Based on these comparisons, the conclusion is reached that the Brunswick Unit 2 Cycle 16 core/cycle has a significantly flatter core MCPR distribution (( )) and slightly flatter pin-by-pin bundle power distributions (( )) than what was used to perform the Cycle 15 SLMCPR evaluation.
The calculated 1.11 Monte Carlo SLMCPR for Brunswick Unit 2 Cycle 16 is consistent with what one would expect (( )) the 1.11 SLMCPR value is appropriate when the approved methodology given in NEDC-32601P-A is used.
Based on all of the facts, observations and arguments presented above, it is concluded that the calculated SLMCPR value of 1.11 for the Brunswick Unit 2 Cycle 16 core is appropriate. It is reasonable that this value is much larger than the 1.08 value calculated for the previous cycle.
For single loop operations (SLO) the calculated safety limit MCPR for the limiting case is 1.13 as determined by specific calculations for Brunswick Unit 2 Cycle 16.
(( 1]
Supporting Information The following information is provided in response to NRC questions on similar submittals regarding changes in Technical Specification values of SLMCPR. NRC questions pertaining to how GE14 applications satisfy the conditions of the NRC SER[1] have been addressed in Reference [4]. Other generically applicable questions related to application of the GEXL14 correlation and the applicable range for the R-factor methodology are addressed in Reference [5]. Only those items that require a plant/cycle specific response are presented below since all the others are contained in the references that have already been provided to the NRC.
The core loading information for Brunswick Unit 2 Cycles 15 and 16 is provided in Figures 1 and 2, respectively. The impact of the fuel loading pattern differences on the calculated SLMCPR is correlated to the values of (( ))
The power and non-power distribution uncertainties that are used in the analyses are indicated in Table 1. The referenced document numbers have previously been reviewed and approved by the NRC.
Prepared by: Verified by:
G.M. Baka R.H. Szilard Technical Program Manager Technical Program Manager Brunswick Unit 2 Project Global Nuclear Fuel - Americas
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Attachment Additional Information Regarding the 17 October 2002 Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16 Table 1 Comparison of the Brunswick Unit 2 Cycle 16 and Cycle 15 SLMCPR QUANTITY, DESCRIPTION Brunswick Unit 2 Brunswick Unit 2 Cycle 15 Cycle 16 Number of Bundles in Core 560 560 Limiting Cycle Exposure Point BOC EOR-1K Cycle Exposure at Limiting Point [MWd/STU] 181 15525 Reload Fuel Type GE14 GE14 Latest Reload Batch Fraction [%] 39.3% 42.5%
Latest Reload Average Batch Weight % 3.99% 4.24%
Enrichment Batch Fraction for GEl4 39.3% 81.8%
Batch Fraction for GE13 60.7% 18.2%
Core Average Weight % Enrichment 3.99% 4.10%
Core MCPR (for limiting rod pattern) 1.A205 1.4804
[r(( III 11 Power distribution methodology GETAB Revised NEDO-32601P-A NEDO-32601P-A Power distribution uncertainty GETAB GETAB NEDO-10958-A NEDO-10958-A Non-power distribution uncertainty Revised Revised NEDC-32694P-A NEDC-32694P-A Calculated Safety Limit MCPR 1.08 1.11
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Attachment Additional Information Regarding the 17 October 2002 Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16 Figure I Reference Core Loading Pattern - Cycle 15 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 1 2 3 4 5 7 8 8 8 8 8 8 8 8 7 52 1
50 2 Bunde IAT 7 9 10 13 10 13 13 10 13 10 9 7
4-48 3 8 8 8 9 13 13 9 13 9 9 13 9 13 13 9 8 81 8 8 12 13 9 10 10 7 46 4 7 10 10 9 13 12 8 13 7 12 12 7 13 44 5 8 10 8 9 12 12 9 9 7 12 10 10 12 7 9 9 12 12 91811018 42 8 11 91 13 12 7 9 9 12 12 7 7 1212191917 12 131911018 40 7 8 9 112 I12 7 12 12 12 9 12 10 10 12 9 12 12 12 7 12 12 9 8 9 12 7 38 12 9 99 7 12 13 9 7 8 7 9 13 12 7 12 9 12 9 12 12 9 12 12 9 12 38 9 12 9 9 12 12 7 9 12 12 9 9 12 12 9 7 12 12 9 9 12 13 9 7 7 9 13 34 10 9 12 9 9 9 7 12 9 9 12 7 9 9 9 12 9 9 8 13 10 8 8 10 13 8 9 8 13 13 7 12 9 12 12 7 12- 10 12 12 10 12 7 12 12 9 12 7 13 9 13 8 32 11 8 10 13 7 12 12 V 9 12 12 10 12 10 10 12 10 12 12 9 12 12 12 7 13 11 8 30 12 8 13 9 12 10 7 10 12 9 9 12 10 7 7 10 12 9 9 12 10 7 10 12 9 13 8 28 13 8 13 9 12 10 7 10 12 9 9 12 10 7 7 10 12 9 9 12 1--0 7 10 12 9 13 8 26 14 8 10 13 7 12 12 12 9 12 12 10 12 10 10 12 10 12 12 9 12 12 12 7 13 10 8 24 15 8 13 9 13 7 12 9 12 12 7 12 10 12 12 10 12 7 12 12 9 12 7 13 9 13 8 22 16 8 10 13 8 9 9 12 9 9 91 7 12 9 9 12 7 9 9 1 2 9 9 8 13 10 8 20 17 7 9 13 7 12 9 12 12 7 9 12 12 9 9 12 1 12 9 7 1 12 ]12 7 1 92 13 18 18 16 17 9 12 9 12 9 12 7 12 13 9 7 19 7
1 9 13 12 7 12 9
_1 12 1
9 1
12 9
4 12
~ 12 ~ - 4 - - + + + 4 4 --
14 12 12 7 12 12 12 9 12 10 10 12 9 12 12 121 7 12 12 9 8 20 8 9 12 21 13 12 7 9 9 12 12 7 7 112 12 9 917112 13 9 10 8 8 11 9 8 10 8 9 12 12 9 9 7 121010127 9+/- 9 10 22 13 7 12 12 7 13 8 12 13 10 7 08 23 7 10 10 9 13 12 8 9 11 06 24 8a8 8 9 13 9 13 13 9 8 8 8 9 13 13 9 131 9 04 25 7 9 10 13 10 13 13 10 113 10 ! 7 02 26 7 8 8 8 8 8 8 8 18 7 719 21 23 25 27 29 31 33 3537 39 41 43 45 47 49 51 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 AT#in # Cyce Loaded Bundle Name Core Fresh GE13-P'90TB395-12G5 0-100T-146-T 7 72 0 13 8 68 0 13 GE13-P90TB393-4G6 0M9G5 0-100T-146-T 9 136 0 14 GE13-Pg0T8403-7G6 0/7G5 0-100T-146-T 14 GE13-P9gTB403-5GM 0/7G5 0-10OT-146-T 10 60 0 11 4 0 14 GE13-P9DTB403-5G6 0(7G5 0-10OT-146-T 12 160 160 15 GE14-P1ODB1399-I6GZ-100T-150-T-2418 15 GE14-P1ONAB398-13GZ-100T-150-T-2417 13 60 60 Tota 560 220
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Attachment Additional Information Regarding the 17 October 2002 Cycle Specific SLMCPR for Brunswick Unit 2 Cycle 16 Figure 2 Reference Core Loading Pattern - Cycle 16 13 14 15 16 17 18 19 20 21 22 23 24 25 25 1 2 3 4 6 6 7 8 9 10 11 12 9 9 9 9 9 9 9 9 9 9 52 1
50 2 Bnde lAT 9 12 13 17 17 17 17 17 17 13 12 9 16 13 15 12 12 15 13 16 12 12 9 10 9 48 9 10 9 12 12 46 9110t131171 17115113 116112 16 16 12 16 13 15 17 17 13 10 9 9 10121413161 13 15 12 15 12 12 15 12 15 13 16 13 14 12 11 9 44 15 12 14 13 10 42 101314121512 16 12 14 12 14 14 12 14 12 16 12 40 17 13 15 12 14 12 14 12 14 12 12 14 12 14 12 14 12 15 12 17 9
-j g 117 1 12114112 14 12 14 12 14 14 12 14 12 14 12 14 12 16 17 12 9I 9 38 15 12 12 9 38 9 12 12 15 13 16112 14 12 14 13 14 10 10 14 13 14 12 14 12 16 13 13 16 13 9 34 91316131512 141214 10 14 12 14 14 12 14 10 14 12 14 12 15 9 17 13 16 12 14 12 14 13 14 13 14 12 12 14 13 14 13 14 12 14 12 16 13 17 9 32 15 12 15 17 10 30 17 15 12 15 12 14 12 14 12 14 12 14 14 12 14 12 14 12 14 12 10 10 14 12 14 10 10 14 12 14 10 14 12 14 12 16 12 13 10 28 12 14 10 14 12 14 12 16 12 17 10 12 14 12 14 10 14 12 14 10 10 14 11 12 12 16 26 14 12 14 12 14 12 15 12 15 17 9 1 2 1 12 14 12 14 12 14 14 12 24 11 17 12 12 14 12 16 13 17 9 13 14 12 12 14 13 14 13 14 17 12 16 12 14 12 14 13 14 9
16 13 9 22 14 10 14 12 14 12 15 13 14 12 14 10 14 12 14 14 12 1 1 15 12 15 12 -
- 20 15 12 12 9 18 15 13 16 12 14 12 14 13 14- 10 - 10 -
14 12 14 12 14 12 16 13 12+/- 12 9
16 9 10 17 16 12 14 12 14 12 14 12 14 14 12 14 12 14 12 14 12 16 17 9 14 9 17 13 15 12 14 12 14 12 14 12 12 14 12 14 12 14 12 15 13 17 9 14 13 10 12 12 16 12 14 12 14 14 12 14 12 16 12 15 13 10 16 13 15 12 15 12 12 15 12 15 13 16 13 14 12 10 9 08 I I I I I 12 I 16I 13I 15 9101317171513161216161216131517171311119
___ ___III 16 12 1.! J 12 9 10 9 06 91 1019 12 15 12 12 15 12 12 15 13 12 16 13 17 17 1"7 17 17 13 12 9 04 12 13 17 02 9 9 9 9 9 9 9 9 9 9 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 01 03 05 BLtde Name IAT IT # In Core #
Fresh CyeLae Qyde Loaded GE13-,P9TB403-7G6 0t7G5 0-100T-146-T 9 62 0 14 GE13WTB40-5,G6 0Q7G5 0-100T-146-T 10 36 0 14 GE13-P9DTB4O3-5G6 0/7G5 0-100T-146-T 11 4 0 14 GE14-P10ONAB399-16GZ-100T-150-T-2418 12 160 0 15 GE14-P1ONAB398-13GZ-100T-150-T-2417 13 60 0 15 14 120 120 16 GE14-P100NAB420-18GZ-100T-150-T-2572 GE14-P100NAB419-G7f.07G6 0/3G2.0-100T-150-T-2573 15 40 40 16 GE14-P100NAB425-3G7.0'14G6 0_1G20-100T-150-T-2574 16 40 40 16 17 38 38 16 GE14-P10ONAB439-12G6 0-100T-150-T-2575 Total 560 238
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((enclosed by double brackets )) DRF # 0000-0005-1288
BSEP 02-0174 Enclosure 5 Marked-up Technical Specification Page - Unit 2
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs or core 2.1.1.1 With the reactor steam dome pressure < 785 psig flow < 10% rated core flow:
THERMAL POWER shall be
- 25% RTP.
psig and core 2.1.1.2 With the reactor steam dome pressure z 785 flow Ž,10% rated core flow:
MCPR sh 11 be >. for two recirculation loop operation M3 or Ž . for singe recirculation loop operation.
the top 2.1.1.3 Reactor vessel water level shall be greater than of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be : 1325 psig.
2.2 SL Violations shall be completed within With any SL violation, the following actions 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
2.0-1 Amendment No.(N Brunswick Unit 2
BSEP 02-0174 Enclosure 6 Typed Technical Specification Page - Unit 2
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:
THERMAL POWER shall be
- 25% RTP.
2.1.1.2 With the'reactor steam dome pressure Ž 785 psig and core flow *x 10% rated core flow:
MCPR shall be 1.11 for two recirculation loop operation or z 1.13 for single recirculation loop operation.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be* 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
Brunswick Unit 2 2.0-1 Amendment No. I
BSEP 02-0174 Enclosure 7 Page 1 of 1 List Of Regulatory Commitments The following table identifies those actions committed to by Carolina Power & Light (CP&L)
Company in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to the Manager - Regulatory Affairs at the Brunswick Steam Electric Plant.
Commitment Schedule
- 1. No commitments were made in this request. N/A