ML033650288
| ML033650288 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/19/2003 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC0918, TVA-BFN-TS-445 | |
| Download: ML033650288 (25) | |
Text
Proprietary Information Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 TVA-BFN-TS-445 December 19, 2003 10 CR 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.
50-296 Tennessee Valley Authority
)
BROWNS FERRY NUCLEAR PLANT (BFN) -
UNIT 3 -
TECHNICAL SPECIFICATIONS (TS)
CHANGE 445 -
SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SLMCPR)
CYCLE 12 OPERATION -
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
This letter is in response to the December 5, 2003, RAI regarding BFN TS change request 445.
The proposed amendment, which was submitted October 1, 2003, revises the numeric value of SLMCPR in TS 2.1.1.2 for one and two recirculation loop operation to incorporate the results of the Unit 3 Cycle 12 core reload analysis.
The SLMCPR values provided in TS-445 were determined by Framatome Advanced Nuclear Power (FANP) for TVA.
The NRC questions are repeated in the two enclosures along with the TVA responses for each question.
provides an RAI response, which includes information considered proprietary by FANP. Accordingly, FANP has requested that this proprietary response be withheld from public disclosure pursuant to 10 CFR 2.790.
In consideration, an affidavit, as required by 10 CFR 2.790(b)(1), is also included in Enclosure 1. provides a non-proprietary version of RAI response.
Proprietary Information Enclosure 1 Prted on cyd pa
U.S. Nuclear Regulatory Commission Page 2 December 19, 2003 TVA has determined this additional information response does not change the determination in the October 1, 2003, TS-445 submittal 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 the enclosures to the Alabama State Department of Public Health.
Approval of TS-445 is needed for BFN Unit 3 Cycle 12 operation, which begins in Spring 2004.
Therefore, TVA has previously requested that TS-445 be approved by February 1, 2004, and that the implementation of the revised TS be made within 60 days of NRC approval.
The final Unit 3 Cycle 12 core configuration is currently being modified based on the recent identification of a leaking bundle, which will result in a nominal increase in the batch size of fresh ATRIUM-10 fuel assemblies being used.
The modified core design will maintain the SLMCPR values being requested in TS-445.
There are no regulatory commitments associated with this submittal.
If you have any questions about this TS change, please contact me at (256)729-2636.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on this 19th day of December, 2003.
S' cerely, At-<A Manager of g
and I try Affairs Enclos es:
- 1. Affida oprietary Version of RAI Response
- 2. Non-proprietary Version of RAI Response Proprietary Information Enclosure 1
U.S. Nuclear Regulatory Commission Page 3 December 19, 2003 cc (Enclosures):
State Health Officer Alabama State Department of Public Health RSA Tower -
Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 Proprietary Information Enclosure 1 Technical Specifications (TS) Change 445 Safety Limit Minimum Critical Power Ratio (SLMCPR)
Unit 3 Cycle 12 Operation Affidavit and Proprietary Version of RAI Response
?
- r.
AFFIDAVIT STATE OF WASHINGTON
)
) ss.
COUNTY OF BENTON
)
- 1.
My name is Jerald S. Holm. I am Manager, Product Licensing, for Framatome ANP, Inc. ("FANP"), and as such I am authorized to execute this Affidavit.
- 2.
1 am familiar with the criteria applied by FANP to determine whether certain FANP information is proprietary. I am familiar with the policies established by FANP to ensure the proper application of these criteria.
- 3.
1 am familiar with the FANP information in the report denoted as Attachment A, Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Responses, dated December 15, 2003, and referred to herein as "Document." Information contained in this Document has been classified by FANP as proprietary in accordance with the policies established by FANP for the control and protection of proprietary and confidential information.
- 4.
This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by FANP and not made available to the public.
Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
- 5.
This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.
- 6.
The following criteria are customarily applied by FANP to determine whether information should be classified as proprietary:
(a)
The information reveals details of FANP's research and development plans and programs or their results.
(b)
Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c)
The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for FANP.
(d)
The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for FANP in product optimization or marketability.
(e)
The information is vital to a competitive advantage held by FANP, would be helpful to competitors to FANP, and would likely cause substantial harm to the competitive position of FANP.
- 7.
In accordance with FANP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside FANP only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8.
FANP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
i
- 9.
The foregoing statements are true and correct to the best of my knowledge, information, and belief.
/(Y~aJ I
if IC SUBSCRIBED before me this
) 6 4 day of..D.2 c' 2003.
1z
\\
Susan K. McCoyV\\
NOTARY PUBLIC, STATE OF WAS TON MY COMMISSION EXPIRES: 1/10/04 Technical Specifications (TS) Change 445 Safety Limit Minimum Critical Power Ratio (SLMCPR)
Unit 3 Cycle 12 Operation Non-proprietary Version of RAI Response
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Responses Attachment A Page A-i Attachment A Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Responses Nonproprietary Version December 15, 2003 RRS03022
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Attachment A Responses Page A-1 Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information I Responses NRC Question I Please provide a reference core loading pattern for cycle 12 operation and a detailed description of the final core and fuel design to achieve a 24-month fuel reload cycle at the licensed rated power of 3458 Mwt and operation to normal end of full power (EOFP); (cycle exposure of approximately 16,700 MwdIMTU). Please include post-EOFP and coastdown extensions.
The Browns Ferry Nuclear Plant (BFN) Unit 3 Cycle 12 reference core loading pattern used in the safety limit minimum critical power ratio (SLMCPR) analysis that the Technical Specifications (TS) - 445 submittal is based upon is shown in Figures A. I and A.2, and includes exposed GE1 3 and GE14 fuel assemblies, and a reload batch of new ATRIUM'"-1 0* fuel assemblies. Table A.1 provides an accounting of the fuel types for Cycle 12 operation and Figures A.3, A.4, and A.5 show details on the fresh Framatome ANP (FANP) ATRIUM-10 fuel assemblies. The core design accommodates the cycle length and energy needs and includes approximately 1,100 MWd/MTU post-EOFP and coastdown operations.
NRC Question 2 Provide a flow chart including input parameters to describe the analysis done for the safety limit minimum critical power ratio (SLMCPR) in the mixed core configuration. Also, identify the approved methodologies used in this analysis andjustify their applicability to the SLMCPR analysis for Cycle 12 operation. Please address the technical position and limitations, and conditions in the staff Safety Evaluation Report stated in ANF-524(P)(A) Revision 2 and ANF-524(P)(A) Supplements I and 2, EMF-2209(P)(A) Revision 1, and EMF-2158(P)(A), Revision 0.
An overall SLMCPR analysis flow chart is presented in Figure A.6. The statistical analysis portion of that figure is detailed in Figure A.7. The flow charts are consistent with the presentation and description in Reference A.1.
The NRC-approved topical report methodologies for the process are:
A.
EMF-2209(P)(A) Revision 1 (Reference A.2).
This report supports the applicability of the SPCB critical power correlation for ATRIUM-10 fuel. The NRC Safety Evaluation Report (SER) limitations, conditions, and restrictions applicable for the process are Items 1-3 in the Conclusions section of the SER, which deal with the range of applicability for the critical power correlation. Compliance with the range of applicability is handled within the SLMCPR code SAFLIM2. The remaining NRC SER condition (Item 4) is related to technology transfer to a utility. Since the application of the methodology was performed by FANP for TVA, Item 4 is not applicable.
ATRIUM is a trademark of Framatome ANP.
RR50302
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Attachment A Responses Page A-2 B.
EMF-2245(P)(A) Revision 0 (Reference A.3).
This report supports the application of an approved FANP critical power correlation for co-resident fuel. This methodology was used to characterize the GE13 and GE14 fuel types with the SPCB critical power correlation based on the indirect method as detailed in the topical report. The only NRC SER restriction in the topical is related to technology transfer to a utility. Since the application of the methodology was performed by FANP for TVA, this restriction is not pertinent.
C.
EMF-21 58(P)(A) Revision 0 (Reference A.4).
This report supports the application of the CASMO-4/MICROBURN-B2 methodology as a replacement for CASMO-3/MICROBURN-B. As stated in the NRC SER, the change does not require changes to safety analysis methodology. The treatment of the limitations, conditions, and restrictions in the NRC SER are provided below (paraphrased in some cases for brevity).
- 1.
CASMO-4/AICROBURN-B2 shall be applied within the range of validation criteria (Tables 2.1 and 2.2 of Reference A.4 and measurement uncertainties of Table 2.3 of Reference A.4.
FANP's application of this methodology meets these criteria.
- 2.
The CASMO-4/MICROBURN-B2 code system shall be validated for analyses of any new fuel design which departs from current orthogonal lattice designs and/or exceeds gadolinia and U-235 enrichments.
The fuel represented for BFN Unit 3 Cycle 12 operation does not violate this condition.
- 3.
The CASMO-4/MICROBURN-B2 code system shall only be used for BWR licensing analyses and BWR core monitoring applications.
BFN is a boiling water reactor, therefore, the SLMCPR analysis meets this condition.
- 4.
The review of the CASMO-4/MICROBURN-B2 code system does not imply a generic review of each code.
CASMO-4/MICROBURN-B2 is used as a code system for input into the SLMCPR analysis.
- 5.
ThL CASMO-4/MICROBURN-B2 code system is approved as a replacement for CASMO-3 /AICROBURN-B code system used in NRC-approved BWR methodology and core monitoring applications. Such replacements shall be evaluated to ensure affected methodology continues to comply with its SER restrictions and/or conditions.
Since the SLMCPR analysis consists of a core monitoring simulation, it is appropriate that the methods used for core monitoring are used in the SLMCPR analysis. Because CASMO-4/MICROBURN-B2 is used in the POWERPLEX)-111* core monitoring software POWERPLEX is a trademark of Framatome ANP registered in the United States and various other countries.
RRS032
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Attachment A Responses Page A-3 system (CMSS), the methods and corresponding uncertainties of CASMO-4
/MICROBURN-B2 are applicable for the SLMCPR analysis. This is supported by the response to NRC Question 11 in Reference A.1 Supplement 2 page 11, which states that the Reference A.1 methodology can be used with a core monitoring system if sufficient information is available about the uncertainties for the core monitoring code.
In this case, sufficient information is available for the uncertainties of POWERPLEX-I11 CMSS. As shown in Item D below, compliance with the NRC SER restrictions of Reference A. I is maintained.
- 6.
Any customer who proposes to use the CASMO-4IMJCROBURN-B2 code system independent of any fuel contract will be notified of conditions 1-4.
The BFN Unit 3 Cycle 12 analysis is not impacted by this condition since it is based on a fuel contract.
D.
ANF-524(P)(A) Revision 2 and Supplements 1 and 2 (Reference A.1)
This report supports the SLMCPR methodology. The NRC SER limitations, conditions, and restrictions applicable for the process are:
- 1.
The NRC-approved MICROBURN-B power distribution uncertainties should be used in the SLMCPR determination.
This NRC SER restriction was addressed in the letter from J.F. Mallay (FANP) to USNRC, EMF-2158(P) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2,"
NRC:99:050, December 20,1999. The NRC confirmed that this SER restriction does not apply to CASMO-4/MICROBURN-B2 in letter, Stuart A. Richards (USNRC) to J.F. Mallay (FANP),
Acceptance of Clarifications on Topical Report EMF-2158(P)
Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors:
Evaluation and Validation of CASMO-4IMICROBURN-B2 (TAC No. MA4592),"
February 29, 2000. As detailed in Item C, the approved uncertainties used are based on MICROBURN-B2.
- 2.
The ANFB additive constant uncertainty should be verified per fuel design.
The SPCB critical power correlation replaces the ANFB critical power correlation in the methodology. SPCB additive constant uncertainties were specifically determined for ATRIUM-10, GEI3, and GE14 fuel assemblies (References A.2 and A.3).
- 3.
The CPR channel bowing penalty for non-ANF fuel should be made using conservative estimates of the sensitivity of local powerpeaking to channel bow.
Channel bow specific for the GE13 and GE14 fuel types was explicitly used in the analyses, which results in conservative estimates of local power peaking to channel bow.
RARMSn
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Attachment A Responses Page A-4
- 4.
The methodology for evaluating the effect of fuel channel bowing is not applicable to reused second-lifetime fuel channels.
No reused second-lifetime fuel channels are being used for the BFN Unit 3 Cycle 12 core.
NRC Question 3 Please demonstrate, based on results of the SLMCPR analysis, that the margin for the proposed increase of the SLMCPR value is adequate. Also, identify the dominant fuel type with respect to cycle exposure during the Cycle 12 operation, and the main contributors for the increase of the proposed SLMCPR value.
Based on the BFN Unit 3 Cycle 12 reference core loading pattern used in the SLMCPR analysis, the 0.1% rods in boiling transition (BT) criterion for a two recirculation loop operation SLMCPR of 1.09 was met. Likewise, the criterion was met for an SLMCPR of 1.11 for single-loop operation. The margin was adequate.
Because of the inherent conservatism of the indirect method of Reference A.2 (the application of an approved CPR correlation to co-resident fuel) in determining additive constant uncertainty, a GE14 assembly will contribute more rods in BT than an ATRIUM-10 assembly if compared on an equal MCPR basis. At the beginning of cycle (BOC), the limiting GE14 and ATRIUM-10 assemblies have approximately the same MCPR margin. At later exposures, the limiting GE14 assembly falls off in power and increases in MCPR margin relative to the limiting ATRIUM-10 assembly. Therefore, the dominate MCPR fuel type with respect to cycle exposure is ATRIUM-10 for the cycle except at BOC.
Please refer to Table A.2 for additional detail on MCPR values versus exposure and the lowest MCPR fuel type.
Because TVA has switched from Global Nuclear Fuel methodology to FANP methodology for the Cycle 12 analysis, it is not informative to compare the previous cycle's results for explanations on the calculated 0.01 SLMCPR value increase. The change in methodology alone can cause a change in SLMCPR value.
NRC Question 4 Please describe the procedures used in the approved methodologies to generate fuel-and plant-related uncertainties listed in Table A. 1 of Attachment A of the October 1, 2003, submittal.
Please provide an equation, as a function of these uncertainties, that can be used as an input to the SLMCPR calculations.
The Fuel Uncertainties in Table A. 1 of Attachment A of the October 1, 2003 submittal for BFN3C1 2 are identified as [
] for a mixed core. The procedure used in the determination of the ATRIUM-10 additive constant uncertainty is described in Section 3.6.2 of Reference A.2. The procedure used in the determination of the GE13 and GE14 additive constant uncertainties applied the indirect method as described in Reference A.3. The procedure for determining the [
I is described in Section 9 of Reference A.4 (also refer to Answer 5). The [
, determined from RRS03022
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Attachment A Responses Page A-5
- l. The plant measurement uncertainties are not fuel design dependent; therefore, the same plant uncertainties used in the previous cycle SLMCPR analysis have been retained.
The SLMCPR is determined by a Monte Carlo analysis that convolves the uncertainties. Reference A.1 describes the procedure used for generation of the SLMCPR. Specifically, the SLMCPR is determined by a [
- l. Uncertainties are input as numerical values; therefore, an equation is not applicable.
NRC Question 5 In relation to the SLMCPR analysis, the submittal did not provide a reference for the conditions listed below. Please identify which approved methodology has the condition listed below.
(a) 50 percent of the local power range monitors (LPRMs) out of service (LPRM bypass model on or off);
(b) up to two traversing incore probe (TIP) machines out of service, or the equivalent number of TIP channels; (c) 2500 effective full power hour (EFPH) LPRM calibration interval; and (d) no reused channels.
Also, please describe the rationale for each of the above conditions.
Items (a)-(c) are core monitoring equipment out-of-service conditions. These equipment out-of-service (OOS) conditions are considered typical for operating domestic BWR plants using the POWERPLEX-I1l CMSS and are conditions that determine the [
] used in the SLMCPR analysis. With an increase in the number of LPRMs OOS, TIPs OOS, and the extended LPRM calibration interval, the [
I increases. Reference A.4 Section 9 discusses the [
l The method for calculating the increased [
I is presented in the May 6, 1996 letter, H.D. Curet (SPC) to H.J. Richings (NRC),
uPOWERPLEXS Core Monitoring: Failed or Bypassed Instrumentation and Extended Calibration."
Item (d) deals with channel bow used in the SLMCPR analysis. The channel bow methodology presented in Reference A.1 does not allow for second-lifetime channels. No reused second-lifetime fuel channels are being used for the BFN Unit 3 Cycle 12 core.
MRRS
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Attachment A Responses Page A-6 References A.
ANF-524(P)(A) Revision 2 and Supplements 1 and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuel Corporation, November 1990.
A.2 EMF-2209(P)(A) Revision 1, SPCB Critical Power Correlation, Siemens Power Corporation, July 2000.
A.3 EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
A.4 EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4 and MICROBURN-B2, Siemens Power Corporation, October 1999.
RRS0322
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Responses Attachment A Page A-7 Table A.1 BFN Unit 3 Cycle 12 Core Composition Fuel Description GEl3-P9DTB400-13GZ-1 00T-l46-T GEI3-P9DTB414-15GZ-1 OOT-146-T GE14-PlODNAB402-15GZ-10OT-150-T GE14-Pl ODNAB401 -14GZ-1 OOT-1 50-T ATRIUM-10 AlO-3812B-13GV80 ATRIUM-1 0 Al 0-4075B-1 5GV8O ATRIUM-1 0 Al 0-4087B-1 3GV80 Cycle Loaded 10 10 11 11 12 12 12 Number of Assemblies 124 72 220 64 64 152 68 Serial Number Prefix YJV YJV JLB JLB FCA FCA FCA RRS03022
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Responses Attachment A Page A-8 Table A.2 BFN Unit 3 Cycle 12 MCPR Versus Exposure
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JLB409 4
YJV270 2
YJV372 YJ3VSS YJV345 FCA274 JL8496 FCA206 JLB390 FCA166 JLB367 FCA058 JLB475 1CA042 JLB609 FCA022 YJV164 JLB564 JLB574 YJV221 FCA023 JLB611 ECA043 JLB464 FCA059 JjB541 FCA167 JLB421 FCA207 JLB548 FCA275 YJV380 YJV209 YJV336 YJV308 JLB4 65 FCA262 JLB565 FCA190 JLB387 FCA062 JLB395 FCA118 JLB519 FCA038 JLB391 FCAO1O YJV251 YJV253 FCA011 JLB412 FCA039 JLB536 FCA119 JLB418 FCA063 JLB411 FCA191 JLB575 FCA263 JLB535 YJV239 YJV366 YJV391 YJV170 FCA278 JLB392 FCA210 JL8372 FCA170 JLB380 FCA134 JLB604 FCA106 JLB376 FCA026 JLB606 FCQ002 FCAO03 JLB580 FCA027 JLB400 FCA107 JLB576 FCA135 JLB397 FCA171 JLB423 FCA211 JLB425 FCA279 YJV224 YJV385 YJV344 YJV335 JLB378 EA266 JLB571 FCA194 JL9373 FCA154 JLB493 FCA054 JLB510 FCA090 JLB504 FCA014 JLB602 JLB612 FCAO15 JLB401 FCA091 JLB531 FCA055 JLB527 FCA15S JLB415 FCA195 JLB584 FCA267 JLB403 YJV368 YJV379 YJV252 YJV257 FCA282 JLB570 FCA214 JLB506 FCA174 JLB374 ECA138 JLB503 FC046 JLB569 FCA082 JLB377 FCA006 FCA007 JLB532 FCA083 JLB573 FCA047 JLB546 FCA139 JL8526 FCA175 JLB544 FCA215 JLB581 FA283 YJV261 YJV233 YJV156 YJV137 JLB463 FCA270 ns.368 FCA198 31B370 FCA158 JLB393 FCA122 nsB388 FCA094 JLB497 FA018 3B499 JLB422 FCA019 JLB420 FCQ095 JLB424 FCA123 nsB427 FCA159 JLB402 FCAl 99 JLB398 FA271 JLB533 YJV193 YJV211 YJV346 YJV279 JLB385 FA254 JLB394 F1A78 JLB514 FCA142 nsB382 FAlIO 31.371 FCA030 J1B384 FCA066 FCA067 3ns426 FCAO31 31B409 ErAlI1 JLB405 FCA143 31B543 FCA17 9 JLB399 FCA255 JLB406 YJV312 YJV411 YJV259 YJV274 YJV340 FCA202 JLB567 FCA162 3ns396 FC2 6 JLB366 FC098 JLB383 FCA074 JLB476 JLB545 FCA075 JLB407 FCA099 JLB539 FCA127 JLB419 FCA163 JLB579 FCA203 YJV365 YJV246 YJV231 YJV157 YJV399 YJV337 FCA250 FA182 J1B363 FCA146 31B507 FCA1 14 JLB562 FCA086 JLB365 FAO70 FCA071 JLB417 FCA087 31B583 FCA115 JLB413 FCA14 7 JLB530 FCA183 FCA251 YJV367 YJV406 YJV213 YJV386 YJV161 YJV149 FCA24 6 JLB369 FQ130 JL8561 FCA102 nss568 FCA078 JLB389 JLB528 FCA079 JLB578 FCA103 JLB572 FCA131 JLB404 FCA247 YJV206 YJV215 YJV374 YJV400 YV339 JLB491 FCA242 JLB563 FCA234 JLB379 FCA226 ns.386 FCA218 FCA219 JLB416 FCA227 JL8414 FCA235 ns.610 FCA243 JLB538 YJV375 YJV403 YJV348 YJV305 YJV169 JL8466 FCA238 31B375 FCA230 31B362 FCA222 31B437 31B547 FCA223 JLB529 FCA231 JLB410 FCA239 JLB534 YJV217 YJV229 YJV412 YJV160 YJV203 YJV310 YJV260 YJV163 YJV172 YJV277 YJV347 YJV388 YJV266 YJV223 YJV216 YJV269 YJV218 YJV1 48 YJV214 YJV152 YJV276 YJV301 YJV341 YJV349 YJV298 YJV334 YJV370 YJV273 YJV376 YJV377 YJV265 YJV248 YJV230 Figure A.1 BFN Unit 3 Cycle 12 Reference Loading Pattern (Continued)
( 3 to =
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Informaton Attachment A Responses Page A-I1 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 30 21 26 21 31 26 31 25 32 25 32 25 33 25 22 22 34.0 18.1 34.1 0.0 20.7 0.0 22.1 0.0 21.8 0.0 21.4 0.0 22.0 35.3 39.1 28 26 21 31 26 31 25 31 25 32 25 32 25 33 21 21 18.1 33.7 0.0 20.2 0.0 21.9 0.0 21.5 0.0 21.4 0.0 21.6 0.0 38.8 39.6 26 21 31 25 31 25 32 25 31 25 32 26 33 25 21 22 33.9 0.0 20.1 0.0 21.2 0.0 21.6 0.0 21.6 0.0 20.2 0.0 20.2 38.0 41.3 24 31 26 31 25 32 26 32 25 32 26 32 25 33 21 22 0.0 20.8 0.0 21.4 0.0 20.7 0.0 21.4 0.0 20.5 0.0 22.1 0.0 38.6 42.1 22 26 31 25 32 25 31 25 32 25 32 26 33 25 21 21 20.5 0.0 22.1 0.0 22.5 0.0 21.8 0.0 21.5 0.0 20.3 0.0 21.1 35.6 36.9 20 31 25 32 26 31 25 32 25 32 25 32 26 33 21 21 0.0 21.9 0.0 20.7 0.0 22.4 0.0 18.6 0.0 21.9 0.0 18.6 0.0 38.3 41.1 18 25 31 25 32 25 32 25 32 25 32 25 33 25 21 21 22.3 0.0 21.6 0.0 21.6 0.0 21.2 0.0 21.5 0.0 20.6 0.0 20.0 32.6 41.7 16 32 25 31 25 32 25 32 25 32 25 33 25 21 21 0.0 21.6 0.0 21.4 0.0 18.6 0.0 22.1 0.0 20.6 0.0 20.9 37.3 42.0 14 25 32 25 32 25 32 25 32 26 32 21 22 21 22.4 0.0 21.6 0.0 21.6 0.0 21.5 0.0 20.6 0.0 34.9 36.1 41.3 12 32 25 32 25 32 25 32 25 32 33 21 22 22 0.0 20.7 0.0 19.5 0.0 22.2 0.0 19.4 0.0 0.0 35.4 39.8 40.5 10 25 32 26 32 26 32 25 33 22 22 22 21.3 0.0 20.6 0.0 20.3 0.0 20.6 0.0 35.0 34.6 41.0 8
33 25 33 25 33 26 33 25 21 22 0.0 21.8 0.0 22.0 0.0 18.6 0.0 21.0 35.4 39.7 6
25 33 25 33 25 33 25 21 21 21 21.9 0.0 20.5 0.0 21.8 0.0 20.0 37.7 41.7 41.7 4
21 22 21 21 22 21 21 22 Nuclear Fuel Type 36.7 35.6 37.2 38.2 36.2 38.4 32.8 40.4 BOC Exposure (GWd/MTU) 2 22 21 22 22 22 21 21 39.5 39.3 41.5 42.0 40.9 42.1 41.9 Fuel Type Description Cycle Loaded 21 GE13-P9DTB400-13GZI-IOOT-146-T 10 22 GE13-P9DTB414-15GZ-10OT-146-T 10 25 GE14-PlODAB402-15GZ-10OT-150-T 11 26 GE14-PIODNAB401-14GZ-10OT-150-T 11 31 ATRIUM-10 A1O-3812B-13GV80 12 32 ATRIUM-10 A10-4075B-15GV80 12 33 ATRIUM-10 A10-4087B-13GV80 12 Figure A.2 BFN Unit 3 Cycle 12 Lower-Right Quarter Core Layout by Fuel Type RRS=3022
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Al OP-0720L-OGO Figure A.3 Elevation View for the BFN Unit 3 Cycle 12 ATRIUM-10 Al0-3812B-13GV80 Fuel Assembly Design RRSO0022
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I Figure A.4 Elevation View for the BFN Unit 3 Cycle 12 ATRIUM-10 AIO-4075B-15GV80 Fuel Assembly Design RRS03022
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A OT-4295L-1 OG50 Figure A.5 Elevation View for the BFN Unit 3 Cycle 12 ATRIUM-10 Al 04087-13GV80 Fuel Assembly Design RRS03022
Browns Ferry Nuclear Plant Unit 3 TS-445 Request for Additional Information Responses Attachment A Page A-1 5 Figure A.6 Overall SLMCPR Analysis Flow Chart "S03022
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Figure A.7 Statistical Analysis of SLMCPR Flow Chart RRSOa22