ML061290590
| ML061290590 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/01/2006 |
| From: | Crouch W Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TVA-BFN-TS-455 | |
| Download: ML061290590 (27) | |
Text
Tennessee Valley Autiority, Post Office Box 2000, Decatur, Ajabama 35609-2000 May 1, 2006 TVA-BFN-TS-455 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:
In the Matter of
)
Docket No. 50-259 Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 1 -
TECHNICAL SPECIFICATIONS (TS)
CHANGE TS-455 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SLMCPR)
CYCLE 7 OPERATION Pursuant to 10 CFR 50.90, the Tennessee Valley Authority (TVA) is submitting a request for a TS change (TS-455) to license DPR-33 for BFN Unit 1. The proposed change revises the numeric values of SLMCPR in TS Section 2.1.1.2 for single and two reactor recirculation loop operation to incorporate the results of the Unit 1 Cycle 7 SLMCPR analysis.
The SLMCPR analysis report was prepared by Global Nuclear Fuel (GNF) for TVA in support of this proposed TS change.
On June 28, 2004, TVA submitted proposed TS change TS-431 to allow operation of Unit 1 at Extended Power Uprate (EPU) conditions.
Accordingly, the Unit 1 Cycle 7 SLMCPR analysis was performed based on EPU conditions.
A proprietary version of the GNF SLMCPR analysis report is provided in Enclosure 3. Some of the information in is considered proprietary and GNF requests that this proprietary information be withheld from public 3003C)
U.S. Nuclear Regulatory Commission Page 2 May 1, 2006 disclosure in accordance with 10 CFR 9.17(a)(4) and 10 CFR 2.390(a)(4). A GNF affidavit supporting this request is included in Enclosure 3. Enclosure 4 provides a non-proprietary version of the same report.
The SLMCPR analysis report is provided in a new template format, which was agreed to between GNF and NRC.
It is TVA's understanding that this template format is intended to provide NRC with sufficient information on GNF-based SLMCPR TS changes to minimize the need for supplemental submittals.
TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).
Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health.
Approval of TS-455 is needed for BFN Unit 1 Cycle 7 operation, which is scheduled to begin in Spring 2007.
Therefore, TVA is asking that this TS change be approved by February 1, 2007, and that the implementation of the revised TS be made within 60 days of NRC approval.
There are no regulatory commitments associated with this submittal.
If you have any questions about this submittal, please contact me at (256) 729-2636.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on May 2, 2006.
Sincerely,
'U4&.
&VJ William D. Crouch Manager of Licensing and Industry Affairs
Enclosures:
- 1. TVA Evaluation of the Proposed Change
- 2. Proposed Technical Specifications Changes (mark-up)
- 4. Non-Proprietary version of GNF Report cc: See page 3
U.S. Nuclear Regulatory Commission Page 3 May 1, 2006 Enclosures cc(Enclosures):
State Health Officer Alabama State Department of Public Health RSA Tower -
Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 Mr. Malcolm T. Widmann, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite Atlanta, Georgia 30303-8931 NRC Unit 1 Restart Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 Margaret Chernoff, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Browns Ferry Nuclear Plant (BFN) Unit 1 Technical Specifications (TS) Change TS-455 Safety Limit Minimum Critical Power Ratio (SLMCPR)
Cycle 7 Operation TVA Evaluation of the Proposed Change
1.0 DESCRIPTION
Pursuant to 10 CFR 50.90, the Tennessee Valley Authority (TVA) is submitting a request for a TS change (TS-455) to Operating License DPR-33 for BFN Unit 1. The proposed change revises the numeric values of SLMCPR in TS Section 2.1.1.2 for single and two reactor recirculation loop operation to incorporate the results of the Unit 1 Cycle 7 SLMCPR analysis.
Approval of TS-455 is needed for Cycle 7 operation, which is scheduled to begin in Spring 2007.
2.0 PROPOSED CHANGE
The proposed TS change revises the SLMCPR value in Unit 1 TS 2.1.1.2 to 1.11 for single recirculation loop operation and to 1.09 for two recirculation loop operation. A marked-up TS page is provided in Enclosure 2, which shows the TS revision.
3.0 BACKGROUND
Safety Limits (SLs) are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.
One such SL included in BFN TS is the SLMCPR value in TS 2.1.1.2.
The SLMCPR limit is established such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling as a result of normal operation and abnormal operational transients, which in turn ensures fuel cladding damage does not occur.
A general discussion of the SLMCPR parameter is maintained in Section 3.7.7.1.1, Fuel Cladding Integrity Safety Limit, of the BFN Updated Final Safety Analysis Report.
E1-1
As noted above, the SLMCPR limit is established such that fuel design limits are not exceeded during steady state operation, normal operational transients, and abnormal operational transients.
As such, fuel damage is calculated not to occur if the limit is not violated.
However, because fuel damage is not directly observable, a stepback approach is used to establish corresponding MCPR Operating Limits.
In simple terms, the MCPR Operating Limits are established by summing the cycle-specific core reload transient analyses adders and the calculated SLMCPR values.
The MCPR Operating Limits are required to be established and documented in the Core Operating Limits Report (COLR) for each reload cycle by TS 5.6.5, COLR.
TS 3.2.2, MCPR, specifies the Limiting Conditions for Operation and Surveillance Requirements for monitoring MCPR against the MCPR Operating Limits documented in the COLR.
The absolute value of SLMCPR tends to vary cycle-to-cycle, typically due to the introduction of improved fuel bundle types, changes in fuel vendors, and changes in core loading pattern.
Following the determination of the cycle-specific SLMCPR values, the MCPR Operating Limits are derived.
The MCPR Operating Limits are maintained by the Licensee in the COLR in accordance with TS 5.6.5.a(3).
However, the cycle-specific SLMCPR numeric values are listed in TS 2.1.1.2 and must be revised using the License Amendment process.
The cycle-specific calculations for the Unit 1 Cycle 7 core design have been recently completed and a change to the TS 2.1.1.2 SLMCPR values for single and two recirculation loop operation is warranted.
Therefore, this proposed TS change is requesting that the SLMCPR numeric values in TS 2.1.1.2 be revised to reflect the results of the Cycle 7 analysis.
Approval of TS-455 is being requested for BFN Unit 1 Cycle 7 operation, which is scheduled to begin in Spring 2007.
Therefore, TVA is asking that this TS change be approved by February 1, 2007, and that the implementation of the revised TS be made within 60 days of NRC approval.
4.0 TECHNICAL ANALYSIS
The SLMCPR values have been determined by Global Nuclear Fuel (GNF) for TVA for Unit 1 Cycle 7 operation using plant-and cycle-specific fuel and core parameters. A proprietary version of the SLMCPR analysis letter report prepared by GNF in support of this proposed TS change is provided in Enclosure 3. Enclosure 4 provides a non-proprietary version of the same report.
The SLMCPR evaluation was based on the cycle-specific procedures and analytical methodologies described in "General Electric E1-2
Standard Application for Reactor Fuel," NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September 2005, which includes Amendment 25 (Reference 1).
Amendment 25 was approved by the NRC in a March 11, 1999, safety evaluation report.
Also used were Licensing Topical Reports (LTRs) NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations" (Reference 2), NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluation" (Reference 3), and NEDE-32505P-A, Revision 1, "R-Factor Calculation Method for GE-li, GE-12, and GE-13 Fuel" (Reference 4).
These LTR methodologies were previously used to justify SLMCPR TS changes for Unit 2 Cycle 13 (TS-420), which was approved by NRC in the Safety Evaluation Report (SER) dated February 28, 2003 (ADAMs Accession No. ML020940822) and for Unit 3 Cycle 11 (TS-416), which was approved by NRC in the SER dated March 29, 2002 (ML020940822).
Subsequent Units 2 and 3 cores used Framatome Advanced Nuclear Power fuel and SLMCPR analysis methods.
The Unit 1 Cycle 7 core is designed for Extended Power Uprate (EPU) operation and is primarily comprised of fresh fuel.
The core design uses 564 fresh GE14 and 108 fresh GE13 fuel assemblies, and 36 previously irradiated GE14 and 56 previously irradiated GE13 fuel assemblies, which were discharged from the BFN Unit 2 Cycle 13 core.
The Unit 1 Cycle 7 core fuel assembly loading pattern is shown in Figure 1 of Enclosures 3 and 4.
Referring to Figure 1, the previously irradiated GE14 and GE13 fuel assemblies are used in the periphery of the core, and the fresh fuel assemblies occupy the core interior.
The previous Unit 1 core (Cycle 6) operated in the mid-1980s and was comprised of older GNF fuel assembly types.
In view of this, the customary comparison of the next (Cycle 7) core with the previous cycle's core is not meaningful and, therefore, is not included in SLMCPR analysis reports in Enclosures 3 and 4.
The Cycle 7 core is designed for a nominal 24-month EPU operating cycle.
In this regard, on June 28, 2004, TVA submitted proposed TS change TS-431 to permit operation of Unit 1 at EPU conditions (ML041840109).
Additionally, on March 7, 2006, TVA submitted a response (ML0607202480) to a Request for Additional Information (RAI) dated December 22, 2005 (ML053560120) on the TS-431 EPU application. Enclosure 1 of the March 7, 2006, RAI response provides additional details on the Unit 1 core design and SLMCPR analytic methodologies in the replies to items SRXB-A.9, SRXB-A.23, and SRXB-A.24. As discussed in the reply to RAI item SRXB-A.24, the Unit 1 Cycle 7 SLMCPR analysis includes a 0.02 CPR adder consistent with the approach recommended in NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains,"
February 2006.
This approach was found to be satisfactory for the Vermont Yankee EPU license amendment application, which was E1-3
approved by NRC on March 2, 2006 (ML0600500220).
A copy of NEDC-33173P is provided in Enclosure 9 of the March 7, 2006, RAI response.
5.0 REGULATORY SAFETY ANALYSIS The Tennessee Valley Authority (TVA) is submitting a Technical Specifications (TS) change request to Operating License DPR-33 for the Browns Ferry Nuclear Plant (BFN) Unit 1. The proposed change revises the Reactor Core Safety Limit Minimum Critical Power Ratio (SLMCPR) in TS Section 2.1.1.2 from 1.10 to 1.09 for two reactor recirculation loop operation and from 1.12 to 1.11 for single loop operation.
5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed TS change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment,", as discussed below:
- 1.
Does the proposed Technical Specification change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No The proposed amendment establishes a revised SLMCPR value for single and two recirculation loop operation.
The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident.
The proposed SLMCPR values preserve the existing margin to transition boiling and the probability of fuel damage is not increased.
Since the change does not require any physical plant modifications or physically affect any plant components, no individual precursors of an accident are affected and the probability of an evaluated accident is not increased by revising the SLMCPR values.
The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences.
The revised SLMCPR values have been determined using NRC-approved methods and procedures.
The basis of the MCPR Safety Limit is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated.
These calculations do not change the method of operating the plant and have no effect on the consequences of an evaluated accident. Therefore, the proposed TS change does not involve an increase in the probability or consequences of an accident previously evaluated.
E1-4
Does the proposed Technical Specification change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
No The proposed license amendment involves a revision of the SLMCPR value for single and two recirculation loop operation based on the results of an analysis of the Unit 1 Cycle 7 core.
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 the allowable methods of operating the facility.
This proposed license amendment does not involve any modifications of the plant configuration or changes in the allowable methods of operation.
Therefore, the proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3.
Does the proposed Technical Specification change involve a significant reduction in a margin of safety?
Response
No The margin of safety as defined in the TS bases will remain the same.
The new SLMCPR values were calculated using referenced fuel vendor methods and procedures, which are in accordance with the fuel design and licensing criteria.
The SLMCPR remains high enough to ensure that greater than 99.9 percent of all fuel rods in the core are expected to avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity.
Therefore, the proposed TS change does not involve a reduction in the margin of safety.
Based on the above, TVA concludes that the proposed TS change 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 The SLMCPR values included in this TS submittal have been determined in accordance with the referenced NRC-approved fuel vendor methodologies.
Accordingly, applicable regulatory requirements and criteria are met.
E1-5
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the TS changes will not be inimical to the common defense and security or the health and safety of the public.
6.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed TS changes would change 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 TS changes do 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 TS change 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 TS change.
7.0 REFERENCES
- 1. General Electric Standard Application for Reactor Fuel (GESTAR-II), NEDE-24011-P-A-15, and the US Supplement, NEDE-24011-P-A-15-US, September 2005.
- 2. NEDC-32601P-A, Methodology and Uncertainties for Safety Limit MCPR Evaluations, August 1999.
- 3. NEDC-32694P-A, Power Distribution Uncertainties for Safety Limit MCPR Evaluation, August 1999.
- 4. NEDE-32505P-A, Revision 1, R-Factor Calculation Method for GE-11, GE-12, and GE-13 Fuel, July 1999.
E1-6 Browns Ferry Nuclear Plant (BFN) Unit 1 Technical Specifications (TS) Change TS-455 Safety Limit Minimum Critical Power Ratio (SLMCPR)
Cycle 7 Operation Proposed Technical Specification Changes (mark-up)
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
> 10% rated core flow:
}fC shall be >
for two recirculation loop operation or for single 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.
BFN-UNIT 1 2.0-1 Amendment No. 236 December 23, 1998 Browns Ferry Nuclear Plant (BFN) Unit 1 Technical Specifications (TS) Change TS-455 Safety Limit Minimum Critical Power Ratio (SLMCPR)
Cycle 7 Operation Affidavit and Proprietary Version of GNF Report
Affidavit Affidavit I, Jens G. M. Andersen, state as follows:
(1) I am Consulting Engineer, Thermal Hydraulic Methods, Global Nuclear Fuel -
Americas, L.L.C. ("GNF-A") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in the attachment, "Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry I Cycle 7" 29 March, 2006. GNF proprietary information is indicated by enclosing it in double brackets. In each case, the superscript notation
{3) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.
(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information," and some portions also qualify under the narrower definition of "trade secret," within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission. 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;
- b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of GNF-A, its customers, or its suppliers;
- d. Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, of potential commercial value to GNF-A;
- e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
Affidavit The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b., above.
(5) To address the 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited on a "need to know" basis.
(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF-A's fuel design and licensing methodology.
The development of the methods used in these analyses, along with the testing, development and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to GNF-A or its licensor.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The fuel design and licensing methodology is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.
The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A or its licensor.
Affidavit The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.
GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.
I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.
Executed at Wilmington, North Carolina, this 29th day of
- March, 2006.
Jens G. M. Andersen Global Nuclear Fuel - Americas, LLC Browns Ferry Nuclear Plant (BFN) Unit 1 Technical Specifications (TS) Change TS-455 Safety Limit Minimum Critical Power Ratio (SLMCPR)
Cycle 7 Operation Non-Proprietary Version of GNF Report
Attachment Mearch 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Chycle 7 Proprietary Information Notice This document is the GNF non-proprietary version of the GNF proprietary report. From the GNF proprietary version, the information denoted as GNF proprietary (enclosed in double brackets) was deleted to generate this version.
- 1.
Applicability This guidance is only applicable for plants fueled entirely with GE/GNF fuel bundle designs using NEDE-2401 1-P-A "General Electric Standard Application for Reactor Fuel" (Revision 15 or latest) in which the following methodologies are used to perform the Safety Limit Minimum Critical Power Ratio (SLMCPR) calculation:
- NEDC-32601P-A "Methodology and Uncertainties for Safety Limit MCPR Evaluations" (August 1999).
- NEDC-32694P-A "Power distribution Uncertainties for Safety Limit MCPR Evaluations" (August 1999).
NEDC-32505P-A "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel" (July 1999).
- NEDO-10958-A "General Electric BWR Thermal Analysis Basis (GETAB): Data Correlation and Design Application" (January 1977).
- 2.
Introduction This attachment documents the SLMCPR for Browns Ferry Unit 1 Cycle 7. Browns Ferry 1 Cycle 7 is the first cycle of operation after a long shut down period. The Browns Ferry 1 Cycle 7 core is very similar to an initial core. Only the core periphery contains irradiated bundles. The irradiated bundles are bundles from Browns Ferry 2 Cycle 13. All bundles in the interior part of the core are fresh bundles.
Figure 1 shows the Browns Ferry 1 Cycle 7 core and core inventory. Figure 2 shows the Power/Flow map for Browns Ferry 1 Cycle 7 (assuming NRC approval of Extended Power Uprate (EPU)).
Page 1 of 11 0000-0043-8309
Attachment March 29,2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7 OD 16 El QWWEBET M 4 El UE)1 E-E MME E-YD0MMTEa 1 MI3 ffD 1E-33 15D 12 W
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E3 E 30 B153 E1JE UE!I1D ffE -3 ff M E-E] ff 1a E~ll ffM-E1 E-0 f E~l E-3 E-E3 OMM 34~~~~~~~~~
E3 BEaa3S irZtE t ffl tE1 OD 22 M E3 900- fM ETF3 Ex-g M1!
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a 2
1 3 015 I1 1517 1921225 7 3133 35 73941 43 45 474 41838 789 Figure 1-Reference Loading Pattern and Fuel Assemblies - Browns Ferry 1 Cycle 7 Number Cycle Code Bundle Name Loaded Loaded A
GE13-P9DTB391 -13GZ-1OOT-146-T6-3964 56 BF213 B
GE14-PiODNAB200-3GZ-1OOT-150-T6-2609 36 BF213 C
GE 3-P9DTB1 56-NOG-1 OOT-1 46-T6-2887 56 7
D GE14-PIODNABi57-NOG-1OOT-150-T6-2889 224 7
E GE14-P1ODNAB377-16GZ-1OOT-150-T6-2890 96 7
F GE14-PIODNAB402-16GZ-1OOT-150-T6-2891 32 7
G GE14-P1 ODNAB350-16GZ-1OOT-150-T6-2892 32 7
H GE14-P1 ODNABi47-NOG-1OOT-150-T6-2893 20 7
I GE14-PIODNAB419-16GZ-1OOT-150-T6-2894 32 7
J GE14-PIlODNAB368-15GZ-lOOT-150-T6-2895 72 7
K GE14-PIODNAB402-19GZ-1OOT-150-T6-2896 24 7
L GE13-P9DTB163-NOG-1OOT-146-T6-2888 52 7
M GE14-PI ODNAB377-17GZ-1OOT-150-T6-2897 32 7
Page 2 of 11 0000-0043-8309
Attachment March 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7 TVAP~wTrw/FkMapMUWMEA)
CnrH16iir) so a
U I0 so 1W itO Ui l
3 2
4a FI I.
Cor Flaw %
Figure 2 - Browns Ferry I Cycle 7 Power/Flow Operating Map Page 3 of 11 0000-0043-8309
Attachment March 29,2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7
- 3.
Discussion Browns Ferry 1 Cycle 7 is the first cycle after a long shut down period and, therefore, no comparisons with previous cycle results are relevant and no previous cycle data is provided in this document.
However, the Browns Ferry Cycle 7 SLMCPR results are well in line with other similar reactor/cycles results. Table 1 provides the Browns Ferry 1 Cycle 7 core loading description. Table 2 lists the SLMCPR calculations applied methodologies.
The SLMCPR evaluations for Browns Ferry 1 Cycle 7 were performed using NRC approved methodology and uncertainties. In general, the calculated safety limit is dominated by two key parameters: (1) flatness of the core bundle-by-bundle MCPR 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. These parameters for Browns Ferry 1 Cycle 7 are summarized in Table 3.
The core loading pattern for Browns Ferry 1 Cycle 7 is provided in Figure 1. The impact of the fuel loading pattern on the calculated SLMCPR is correlated to the values of MCPR importance Parameter (MIP) and R-Factor Importance Parameter (RIP). The MIP value for Browns Ferry 1 Cycle 7 limiting case analyzed at End of Cycle (EOC) is provided in Table 3.
The pin-by-pin power distributions are characterized in terms of R-Factors using the NRC approved methodology. The Browns Ferry 1 Cycle 7 limiting case is at EOC, and the RIP value, considering the participation of the contributing bundles, is provided in Table 3.
These calculations use the GEXL14 correlation for GE14 fuel and GEXL09 for the GE13 fuel. ((
11.
A Part 21 communicated by MFN 05-058 and MFN 05-095 (ADAMS Accession Nos. ML051790237 and ML052690084, respectively) identified a potential problem with the GEXL14 correlation due to GE14 Zircaloy spacer spring deformation during critical power testing at the ATLAS facility. This issue is directly addressed in the Unit 1 Cycle 7 SLMCPR analysis by applying modified R-Factors for all of the GE-14 bundles in the core. These modified R-Factors were generated by applying conservative additive constants to the susceptible rod locations. A change in the GEXL14 analytical solution was also applied to assure convergence for the resulting higher bundle R-Factors consistent with the approach discussed in MFN 05-095. No additional actions are required to address this Part 21 issue for Unit 1.
- 4.
Deviations in Uncertainties The SLMCPR evaluation procedure is shown in Figure 4.1 of NEDC-32601P-A and reproduced in Figure 3 of this report.
The SLMCPR evaluations for Browns Ferry 1 Cycle 7 were performed using NRC-approved methodology and uncertainties. Tables 4 and 5 provide a list of uncertainties that deviate from the NRC approved values. Note only the GEXL R-Factor uncertainty is different than the approved value. ((
]1. This Page 4 of 11 0000-0043-8309
Attachment March 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7 change affects only the a RPEAK as shown in Figure 4.1 of NEDC-32601P-A and Figure 3 of this document.
Browns Ferry 1 is a D lattice plant and, therefore, is not susceptible to control blade shadow corrosion induced channel bow. ((
11 Browns Ferry 1 Cycle 7 SLMCPR is analyzed only at rated flow (100%) because the licensed flow window is very narrow: 99°0 to 105% flow.
4.1 Departure from Approved Methodologies No departures from the approved methodologies were used in the Browns Ferry 1 Cycle 7 SLMCPR calculations.
4.2 GE14 Fuel Axial Power Shape Penalty
))
4.3 Methodology Restrictions The four restrictions identified on Page 3 of NRC's Safety Evaluation relating to the General Electric Licensing Topical Reports NEDC-32601P, NEDC-32694P, and Amendment 25 to NEDE-2401 1-P-A (March 11, 1999) are addressed in References 1, 2, and 3.
4A Core Monitoring System The Browns Ferry 1 Cycle 7 core will be monitored with the GE 3D MONICORE core monitoring system.
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Attachment March 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7
))
Figure 3 - Reference Figure Page 6 of 11 0000-0043-8309
Attachment March 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7
- 5.
Summary The calculated 1.07 Dual Loop Operation (DLO) SLMCPR and 1.09 Single Loop Operation (SLO)
SLMCPR for Browns Ferry 1 Cycle 7 are consistent with expectations given the ratios for MIP and RIP that have been calculated, the axial power shapes in the core, and the methodology and uncertainties applied. Correlations of MIP and RIP directly to the calculated SLMCPR have been performed for Browns Ferry 1 Cycle 7, which show that these values are appropriate when the approved methodology and the reduced uncertainties given in NEDC-32601P-A and NEDC-32694P-A are used.
The final DLO SLMCPR of 1.09 and final SLO SLMCPR of 1.11 were determined by adding a 0.02 ACPR to the corresponding calculated values. This approach is consistent with that used in the Vermont Yankee EPU license application, which was reviewed and subsequently approved by NRC on March 2, 2006 (Ref: Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: Extended Power Uprate (TAC No. MC0761), and is in accordance with Licensing Topical report NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains", February 2006, which was submitted by GE to NRC on February 10, 2006.
- 6.
References
- 1. Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to R Pulsifer (NRC), "Confirmation of IOx1O Fuel Design Applicability to Improved SLMCPR, Power Distribution and R-Factor Methodologies", FLN-2001-016, September 24, 2001.
- 2. 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.
- 3. 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.
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Attachment March 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7 Table 1 - Description of Core Previous Previous Curet Curren DCycle Cycle Cycle Min t Core D crponMini Core Rated Core Core Flow RtdCr Flow Flow Flow Number of Bundles in the core N/A N/A N/A 764 Limiting Cycle Exposure Point N/A N/A N/A EOC (BOC/MOC/EOC)
Cycle Exposure at Limiting Point N/A N/A N/A 13000 (MWd/STU)
% Rated Flow N/A N/A N/A 100 Reload Fuel Type N/A N/A N/A GE14/GE13 Latest Reload Batch Fraction, %
N/A N/A N/A 88 Latest Reload Average Batch Weight %
N/A N/A N/A 2.63 Enrichment Core Fuel Fraction GE14 (%).
N/A N/A N/A 78.5 Core Fuel Fraction GE13 (%),
N/A N/A N/A 21.5 Core Average Weight % Enrichment N/A N/A N/A 2.70 Table 2 - SLMCPR Calculation Methodologies Previous Previous Current DecipinCycle Cycle CurreMnt Cycle Desciption n Core Rated Core Cycre no Rated Core now Flow Co Fow ow Non-power Distribution Uncertainty NNA NIA NIA 3EDC-N/A NA N/A 32601-P-A Power Distribution Methodology N-A NIA NIA 3E64-N/A NA N/A 32694-P-A Power Distribution Uncertainty N/A N/A NIA NEDC-yN/A N/A N/A 32694-P-A Core Monitoring System N/A I
N/A I
N/A 13D M O N IC O R E Page 8 of I 1 0000-0043-8309
. j Attachment March 29,2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7 Table 3 - Monte Carlo Calculated SLMCPR vs. Estimate Previous Previous Current Current Quanity DecripionCycle Cycle CylMi Cye Mi Description Min Core Rated Core Cycle Mmn Rated Core Flow Flow Core Flow Flow MCPR Importance Parameter (MIP)
JL R-Factor Importance Parameter (RIP)
.f Base SLMCPR Estimate - Using MIP*RIP
[
Correlation List Separately the SLMCPR change due to
[
any Methodology deviations, Penalties or Uncertainty Deviations from approved values (e.g. GEXL R-Factor Uncertainty Increase, Double Hump Axial Power Shape Penalty, etc.)
Total Estimated SLMCPR A__
Calculated SLMCPR (DLO)
N/A N/A N/A 1.07 Calculated SLMCPR (SLO)
N/A N/A N/A 1.09 Target OLMCPR N/A N/A N/A 1.430 Core MCPR (for limiting rod pattern)
N/A N/A N/A 1.414 Final SLMCPR (DLO)
(See SFPmary)
(See Summary)
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Attachment March 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7 Table 4 - Non-Power Distribution Uncertainties Nominal PrvosCurrent (NRC Previous Previous Cu nt Cu ept Quantity Approved)
Cycle M RCaytced Cycle Mn Cyle Impact on Value Car Flow Core Flow Core Flow Core Flow NEDC 32601-P-A Feedwater II Flow Feedwater Temperature Measurement Reactor
[]
Pressure Measurement X
Core Inlet 0.2 N/A N/A N/A 0.2 N/A Temperature Measurement Total Core 6.0 SLO /
N/A N/A N/A 6.0 SLO /
N/A Flow 2.5 DLO 2.5 DLO Measurement Channel flow
]
area variation Channel 5.0 N/A N/A N/A 5.0 N/A friction factor multiplier Channel to
((
]
channel non-uniformity friction factor multiplier Page 10 of 11 0000-0043-8309
6 Attachment March 29, 2006 Additional Information Regarding the Cycle Specific SLMCPR for Browns Ferry 1 Cycle 7 Table 5 - Power Distribution Uncertainties Nominal Previous l Previous Current I
(NRC Cycle Cycle Cycle Cylent pcto Quantity Approved)
Min Rated Min Rated SLMCPR Value Core Core Core Core SMow I__
Flow Flow Flow CoeFw NEDC 32694-P-A, 3D-MONICORE GEXL R-Factor
_11_1 Random 2.85 SLO /1.2 N/A N/A N/A 2.85 SLO N/A effective TIP DLO 1.2 DLO reading TIP integral JL1L Four bundle
((
))
power distribution surrounding TIP location Contribution to
[
bundle power uncertainty due to LPRM update Contribution to
]
bundle power due to failed TIP Contribution to
[
bundle power due to failed LPRM Total uncertainty
((
in calculated bundle power Uncertainty of
[]
TIP signal nodal uncertainty TIP random nodal error Page 11 of 11 0000-0043-8309