RS-12-020, Supplemental Information Supporting Request for Technical Specification Change for Minimum Critical Power Ration Safety Limit
| ML12031A073 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/30/2012 |
| From: | Gullott D Exelon Nuclear, Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML120310248 | List: |
| References | |
| RS-12-020, TAC ME7331 | |
| Download: ML12031A073 (42) | |
Text
Exelon Generation Company,LtC wwwexe!oncorpcom 4300 Wmfjeld Road Nuclear Warrenvllle, Il 60555 Attachment 1 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2-390. When separated from Attachment 1, this document is decontrolled.
RS-12-020 10 CFR 50.90 January 30.2012 U* S. Nudear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit 1 Facility Operating License No* NPF-11 NRC Docket No. 50-373
Subject:
Supplementallnformaflln Supporting the Request for Technical Specification Change for Minimum Critica I Power Ratio Safety Limit (TAC No. ME7331)
References:
- 1. Letter from Mr. David M. Gullott (Exelon Generation Company. LLC (EGC>>
to U. S. NRC, "RequestforTechnical Specification Change for Minimum Critical PowerRatb Safety Limit." dated October 12,2011
- 2. Letter from U. S. NRC to Mr. Michael J. Pacilio (EGC). "LaSalle County Station, Unit 1 -Request for Addit'bnal Information Regarding Proposed Technical Specification Safety Limit Minimum Critical Power Ratio Changes (TAC No. ME7331)." dated January 19, 2012 In Reference 1, Exelon Generation Company. LLC (EGC) requested an amendment to Appendix A. Technical Specifications (TS). of Facility Operating License No. NPF-11 for LaSalle County Station (LSCS), Unit 1. The proposed change revises the value of the safety limit minimum critical power ratios (SLMCPRs) in TS Section 2.1.1, "Reactor Core SLs." These changes are needed to support the upcoming cyde of operation (i .e., Cycle 15) for LSCS, Unit 1. In Reference 2, the NRC requested that EGC provide additional information in support of their review of Reference 1. The EGC responses to the NRC requests are provided in the attachments to this letter. Specifically, the responses to NRC RAI-1 through 6 are provided in Attachment 1, the response to NRC RAI-2 is contained in Attachments 1 and 4, and the response to NRC RAI {J7 is provided in Attachment 4.
Attachment 1 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2-390. When separated from Attachment 1, this document is decontrolled.
January 30, 2012 U. S. Nuclear Regu story Commlssion Page 2 contains information that is proprietary to Global Nuclear Fuel (GNF) and should be withheld from public disclosure in accordance with paragraph (b}(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." An affidavit from GNF attesting to the proprietary nature of this information is provided in Attachment 2. Attachment 3 18 a non-proprietary version of Attachment 1.
The information provided In this letter does not affect the No Significant Hazards Consideration or the Environmental Consideration provided In Attachment 1 of the original license amendment request as described in the Reference 1 submittal.
In accordance with 10 CFR 50.91 (b), "State consultation," EGC is providing the State of Illinois with a copy of this letter and its attachment to the designated State Official.
This letter contains no new regulatory commitments. If you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30th day of January, 2012.
Respectfully, David M. Gullatt Manager - Licensing Exelon Generation Company, LLC Attachments:
- 1. Additional Information Supporting the Request for Technical SpeCification Change br Minimum Critical Power Ratio Safety Umit - Proprietary Version
- 2. Global Nuclear Fuel Affidavit Supporting Proprietary Nature of Information In Attachment 1
- 3. Additional Information Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Umit - Non-Proprietary Version
- 4. Exelon Generatlon Company, LLC Supplemental Response to NRC RAI-2 and Response to NRC RAI-7
- 5. LaSalle County Station, Unit 1, Cycle 15 Cycle Design Inputs and Requirements (CDIR)
Attachment 2 Global Nuclear Fuel Affidavit Supporting Proprietary Nature of Information in Attachment 1
Global Nuclear Fuel- Americas AFFIDAVIT I, Lukas Trosman, state as follows:
(1) I am Engineering Manager, Reload Design and Analysis, Global Nuclear Fuel -
Americas, LLC (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 Enclosures 1 of GNF's letter, CFL-EXN-HAI-12-008, C. Lamb (GNF-A) to J. Fisher (Exe1on), entitled "GNF Response to NRC RAIs for LaSalle Unit 1 Cycle 15 SLMCPR Submittal," dated January 27,2012. GNF-A proprietary information in Enclosure 1, which is entitled "Response to NRC RAIs for LaSalle Unit 1 Cycle 15 SLMCPR Submittal," is identified by a dotted underline inside double square brackets. ((r.hi~..~~nt~n£~..l§..~Jl
~~~mRh~c~~!J] A "((" marking at the beginning of a table, figure, or paragraph closed with a "))" marking at the end of the table, figure or paragraph is used to indicate that the entire content between the double brackets is proprietary. 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" (Exemption 4). The material for which exemption from disclosure is here sought 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, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (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 aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A; CFL-EXN-HAI-12-008 Affidavit Page 1 of 3
- d. Infonnation which discloses patentable subject matter for which it may be desirable to obtain patent protection.
The infonnation 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 10 CFR 2.390 (b) (4), the infonnation sought to be withheld is being submitted to NRC in confidence. The infonnation is of a sort customarily held in confidence by GNF-A, and is in fact so held. The infonnation 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 infonnation in confidence. Its initial designation as proprietary infonnation, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(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 infonnation in relation to industry knowledge, or subject to the tenns 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 detennination 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 infonnation, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The infonnation identified in paragraph (2) is classified as proprietary because it contains details of GNF-A's fuel design and licensing methodology. The development of this methodology, along with the testing, development and approval was achieved at a significant cost to GNF-A.
The development of the fuel design and licensing methodology along with the interpretation and application of the analytical results is derived from an extensive experience database that constitutes a major GNF-A asset.
(9) Public disclosure of the infonnation sought to be withheld is likely to cause substantial hann to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The infonnation 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 CFL-EXN-HA 1-12-008 Affidavit Page 2 of 3
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.
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 peIjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.
Executed on this 27th day of January 2012.
Lukas Trosman Engineering Manager, Reload Design and Analysis Global Nuclear Fuel- Americas, LLC CFL-EXN-HAI-12-008 Affidavit Page 3 of 3
Attachment 3 Additional Information Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit - Non-Proprietary Version
ENCLOSURE 2 CFL-EXN-HA 1-12-008 Response to NRC RAls for LaSalle Unit 1 Cycle 15 SLMCPR Submittal Non-Proprietary Information - Class I (Public)
INFORMATION NOTICE This is a non-proprietary version of CFL-EXN-HA1-12-008 Enclosure 1, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside an open and closed bracket as shown here (( )).
CFL-EXN-HA1-12-008 Non-Proprietary Information - Class I (Public) Page 1 of 10 REQUEST FOR ADDITIONAL INFORMATION LASALLE COUNTY STATION, UNIT 1 LICENSE AMENDMENT REQUEST REGARDING PROPOSED SAFETY LIMIT MINIMUM CRITICAL POWER RATION CHANGE DOCKET NO. 50-373 By letter dated October 12, 2011 (Agencywide Documents Access and Management System Accession (ADAMS) Accession No. ML112860067), Exelon Generation Company LLC (Exelon, the licensee) submitted a license amendment request (LAR) proposing to modify technical specification (TS) Section 2.1, "Safety Limits," for LaSalle County Station (LSCS), Unit 1. The requested change involved revised safety limit minimum critical power ratios (SLMCPRs) calculated as a result of the cycle-specific analysis performed by Global Nuclear Fuel (GNF) to support operation in the upcoming LSCS, Unit 1, Cycle 15. The U.S. Nuclear Regulatory Commission (NRC) staff has been reviewing the submittal and has determined that additional information is needed to complete its review.
CFL-EXN-HA 1-12-008 Non-Proprietary Information - Class I (Public) Page 2 of 10
RAI-01
In the LAR, Attachment 6, Tables RAI-06-1 and RAI-06-2, provide core map to show those bundles experienced 0.1 boiling transition criterion of limiting cases for single-loop operation (SLO) and two-loop operation (TLO). Please provide identification of bundle group and number of bundles in the Figure 1, Attachment 5, corresponding to their burnup status (once-burned, twice-burned, or fresh fuel) for Cycle 15.
Response to RAI*01:
Table RAI-01-1 shown below contains the bundle group, number of bundles, bundle type, burnup status and fuel type (IAT) associated with the Cycle 15 core loading map presented in Figure 1 of Attachment 5 of the LAR. All of the data presented in Table RAI-01-1 is equivalent for both TLO and SLO.
Table RAI-01*1 Bundle Group, Number of Bundles, Bundle Type, Bumup Status and Fuel Type for Both TlO and SlO
((
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. J ;.. ~~*~f.y;;~j~**. ~';i~y~;~%~}:(;/y;>f~~~
- ~l"< ('CcY' ~,.;. *~.>.{~.~(.*~*i**r~~/*i<'0*
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CFL-EXN-HA 1-12-008 Non-Proprietary Information - Class I (Public) Page 3 of 10 RAI*02:
Core design is an iterative process designed to develop an optimal configuration that meets operational requirements. In the LAR, Attachment 7, for the slides titled "Pre-Estimation -
Linear Reactivity," please provide the most current updated parameters applicable to LSCS, Unit 1, Cycle 15. Also, provide the details of a plant-specific final core loading pattern as shown in Figure 1 including core design procedures, guidelines, criteria, and approved methodologies used for this Cycle 15 analysis with respect to a mixed core application.
Response to RAI*02:
Pre-estimating reload batch size and enrichment, either using the method on the slide entitled "Pre-Estimation - Linear Reactivity" or any other method, is not applicable to the final, analyzed core design.
Methods used to analyze the core-loading pattern, shown in Figure 1, are in accordance with GESTAR II. GESTAR II is the umbrella for all procedures, guidelines, criteria, and methodologies used for this analysis. There is no change in approved methodologies. This is a SLMCPR Technical Specifications change within approved methodologies. SLMCPR is not the primary driver in developing the fuel cycle core design. The energy plan, reactivity, and thermal margins are the primary drivers.
In the development of a mixed core, as in the development of a core containing all GNF fuel, the loading pattern is developed collaboratively by GNF and Exelon based on Exelon input. Among the inputs are:
- Cycle Energy Requirements - fuel bundle design (nuclear) and loading patterns
- Thermal Limit Margins
- Reactivity Margins - minimum shutdown margin, minimum and maximum hot excess reactivity
- Discharge Exposure Limitations and Other Limits as established by safety analysis
- Desired Control Rod Patterns - sequences and durations
- Channel Distortion Minimization
CFL -EXN-HA 1-12-008 Non-Proprietary Information - Class I (Public) Page 4 of 10 RAI*03:
GNF2 fuel deviates from traditional10x10 design through the introduction of a partial length rod configuration, the use of higher linear power, and the use of mixing vanes. The NRC staff considers this a new fuel design with regards to the four restrictions identified in the safety evaluation of General Electric (GE) Licensing Topical Reports NEDC-32601 P, NEDC-32694, and Amendment 25 to NEDE-24011-P-A. Given that LSCS, Unit 1, Cycle 15, uses a core loading pattern which includes GNF2 fuel, provide the following: (1) an evaluation of the four restrictions in NEDC-32601 P, NEDC-32694 and Amendment 25 to NEDE-24011-P-A and the applicability to mixed core with ATRIUM 10 fuel; (2) a description that explains under what conditions the methodologies listed in Section 1.0 of Attachment 5 are sufficient and applied to the LSCS, Unit 1, Cycle 15, application; and (3) a clarification for the statement "no new GNF2 fuel designs are being introduced in LSCS, Unit 1, Cycle 15," in Section 2.5 of Attachment 5.
Response to RAI*03*1:
The four restrictions for GNF2 were determined acceptable by the NRC review of the "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II)," NEDC-33270P, Revision 0, FLN-2007-011, March 14, 2007. Specifically, in the NRC audit report ML081630579 for the said document, Section 3.4.1 page 59 states:
"The NRC staff's SE of NEDC-32694P-A (Reference 69 of NEDE 33270P) provides four actions to follow whenever a new fuel design is introduced. These four conditions are listed in Section 3.0 of the SE. The analysis and evaluation of the GNF2 fuel design was evaluated in accordance with the limitations and conditions stated in the NRC staff's SE, and is acceptable."
Additionally, the NRC audit report, ML081630579, Section 3.4.2.2.1 page 59 states:
"The NRC staff finds that the calculational methods, evaluations and applicability of the OLMCPR and SLMCPR are in accordance with existing NRC-approved methods and thus valid for use with GNF2 fuel."
The four restrictions applied specifically to the mixed core were addressed during the transition from ATRIUM-10 to GNF2 fuel. These limitations were addressed for ATRIUM-10 as follows:
(1) The TGBLA fuel rod power calculational uncertainty for ATRIUM-10 was determined and verified.
(2) The rod power calculation uncertainties were used to reevaluate and confirm the R-factor uncertainty for ATRIUM-10.
(3) The applicability of the Minimum Critical Power Ratio (MCPR) Importance Parameter (MIP) criterion was previously reevaluated through the inclusion of plants containing ATRIUM-10 fuel in the data contained in Figure 5 of Attachment 5.
(4) The bundle power uncertainty associated with the core monitoring system was verified by Exelon as applied to ATRIUM-10.
CFL-EXN-HA 1-12-008 Non-Proprietary Information - Class I (Public) Page 5 of 10 Response to RAI-03-2:
There are three references listed in Section 1.0 of Attachment 5. The applicability of each of the three references is discussed. The three references are:
A. NEDC-32601 P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations,"
August 1999.
B. NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations,"
August 1999.
C. NEDC-32505P-A, "R-Factor Calculation Method for GE11, GE12 and GE13 Fuel,"
Revision 1, July 1999. Table 2 identifies the actual methodologies used for the Cycle 15 SLMCPR calculations. References A and B are directly applicable to the analysis. This process is fuel product independent as long as the R-Factors were appropriately generated.
Reference C is the generic R-Factor methodology report that describes the changed methodology that was adopted after part length rods were introduced. The NRC staffs Safety Evaluation (SE) for NEDC-32505P-A has a requirement that the applicability of the R-Factor methodology is confirmed when a new fuel type is introduced. The confirmation for GNF2 was determined to be acceptable by the NRC staff review of the "GEXL 17 Correlation for GNF2 Fuel," NEDC-33292P, Revision 0, FLN-2007-011, March 14, 2007 in the NRC audit report ML081630579, Section 3.5.5 page 62. The confirmation for the ATRIUM-10 GEXL97 correlation applicable to LSCS Unit 1 Cycle 15 was determined to be acceptable by the NRC staff review of the "GEXL97 Correlation for ATRIUM-10 Fuel," NEDC-33106P-A, Revision 2, June 2004.
Response to RAI-03-3:
GNF2 is an evolutionary fuel product based on GE14 that maintains the previously established 1Ox1 0 array and two water rod makeup.
CFL-EXN-HA1-12-008 Non-Proprietary Information - Class I (Public) Page 6 of 10 RAI*04:
The LSCS, Unit 1, Cycle 15, is a mixed core with once and twice burned ATRIUM 10 fuel.
Please provide: (1) a detailed description of the methodologies used and procedures applied to the LSCS, Unit 1, Cycle 15, calculation for the proposed SLMCPR values based on Figure 3, ; and (2) justification that the methodologies related to ATRIUM 10 fuel may not be needed in this application because none is listed in Section 1.0, Attachment 5.
Response to RAI*04*1:
While LSCS Unit 1 Cycle 15 contains ATRIUM-10 fuel that was not manufactured by GNF, the methodologies contained within NEDC-32601 P-A incorporate fuel-type dependency using fuel-specific inputs. A specific critical power correlation for ATRIUM-10 fuel, GEXL97, referenced in the response to RAI-03-2, was used in this calculation. Additionally, the following items in Figure 3, Attachment 5 were calculated specifically for the ATRIUM-1 0 fuel in the core:
- (J CPD (GEXL) - This information comes from the previously-approved GEXL-97 correlation referenced in the response to RAI-03-2.
- (J RPEAK - This value was established in accordance with NEDC-32601 P-A and is based on the modeling uncertainties for ATRIUM-10 fuel in GNF methods established in Cycle 11, as well as current manufacturing and channel bow uncertainties relevant to this fuel.
Response to RAI*04*2:
The methodologies listed in Section 1.0, Attachment 5 are applicable to GNF2 and ATRIUM-10 designs, and are therefore applicable to LSCS Unit 1 Cycle 15.
CFL-EXN-HA1-12-008 Non-Proprietary Information - Class I (Public) Page 7 of 10 RAJ*05:
Please identify the breakdown of the 10x10 data shown in Attachment 5, Figure 5, by fuel type (Le., GE14, GNF2), because Figure 5 only shows combined data points for the two fuel types.
Also provide: (1) details of the application of Figure 5 data to a mixed core with ATRIUM 10 fuel; and (2) justification that the estimation formula for SLMCPR value is still valid for LSCS, Unit 1, Cycle 15, application.
Response to RAJ*05:
The 10x10 (GE14, GNF2) points shown in Figure 5 of Attachment 5 reflect transition cores with a mix of 1Ox1 0 fuel products. Thus, there are not specific GNF2 data points in Figure 5. The table shown below provides the GE14 and GNF2 batch sizes, and the corresponding
(( )) for the 10x10 (GE14, GNF2) points in the figure. Each row in the table below corresponds to one of the "10x10 (GE14, GNF2)" data points in Figure 5. The table is in ascending order of the abscissa of Figure 5 for ease of correlation to the figure. Sums of batch sizes and ((
)) may not add to 100% due to rounding and/or the presence of other fuel products in the core.
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k 31.0 38.5 28.6 71.4 64.4 35.6 31.0 38.5 31.0 38.5 64.4 35.6 28.6 71.4 28.6 71.4 67.4 32.6 64.4 35.6 67.4 32.6
)) 67.4 32.6 ))
CFL-EXN-HA1-12-008 Non-Proprietary Information - Class I (Public) Page 8 of 10 Response to RAI*05*1 :
Figure 5 is an updated version of Figure 111.5-2 from NEDC-32601 P-A (referenced in Section 1.0 of Attachment 5). Per the response to RAI 111.5 on NEDC-32601 P-A, 'The reduction in CPR margin ... required to place a nominal rod pattern nearer the operating limit is correlated to the natural logarithm of the ratio of the nominal MIP to the limiting MIP value. This correlation is shown in Figure 111.5-2 for all fuel types. The fact that all the data for different fuel types is interspersed about the same curve suggests that it is appropriate to establish a single threshold value for MIP that is independent of fuel type." Figure 5 was previously updated with points representing plants containing ATRIUM-10 fuel. The continued interspersion of the data about the correlation provides continued support for the conservatism of the current MIP criteria used in the SLMCPR process, and the independence of this criterion from fuel type.
Response to RAI*05*2:
While still used as a secondary reasonability check, the estimation formula is not part of the SLMCPR development process. It has no effect on the final SLMCPR.
CFL-EXN-HA 1-12-008 Non-Proprietary Information - Class I (Public) Page 9 of 10 RAI*06:
Please clarify that there is no effect of GNF2 bent spacer wing to LSCS, Unit 1, Cycle 15, operation. If there is an adverse impact, please provide an assessment of the impacts on operations and fuel thermal performance.
Response to RAI*06:
GNF2 bent spacer wing related Part 21 issues are not applicable to LSCS Unit 1 Cycle 15 because the GNF2 fuel in this cycle is not impacted by the Part 21 issue, as indicated in Section 2.13 of Attachment 5 of the LAR.
CFL-EXN-HA 1-12-008 Non-Proprietary Information - Class I (Public) Page 10 of 10 RAI*07:
Please provide an updated version of power/flow map for Cycle 15 operation including stability Option III features of scram region and controlled entry region for backup stability protection based on the Boiling-Water Reactor Owners Group position stated in NEDO-31960A for SLO and TLO.
Response to RAI*07:
Exelon will provide a response to this RAI.
Attachment 4 Exelon Generation Company, LLC Supplemental Response to NRC RAI-2 and Response to NRC RAI-7
Attachment 4 Page 1 of 2 NRC RAI-2. Core design is an iterative process designed to develop an optimal configuration that meets operational requirements. In the LAR, Attachment 7, for the slides titled "Pre-Estimation - Linear Reactivity,"
please provide the most current updated parameters applicable to LSCS, Unit 1, Cycle 15. Also, provide the details of a plant-specific final core loading pattern as shown in Figure 1 including core design procedures, guidelines, criteria, and approved methodologies used for this Cycle 15 analysis with respect to a mixed core application.
Exelon Generation Company, LLC (EGC) Supplemental Response to NRC RAI-2 The response provided by Global Nuclear Fuel (GNF) for RAI-2 in Attachment 1 describes the use of approved GNF methodologies and procedures in accordance with General Electric Standard Application for Reactor Fuel (GESTAR-II). Exelon procedures and guidelines were also used for LaSalle County Station (LSCS), Unit 1, Cycle 15 to direct the bundle design and core reload process. Exelon procedures NF-M-100, "Reload Control Procedure," and NF-AB-110, "Bundle and Core Design (BWR}," were previously provided to the NRC in the responses to an NRC request for additional information (RAI) related the Quad Cities Nuclear Power Station, Unit 1, Cycle 22 Minimum Critical Power Ratio Safety Limit (SLMCPR) license amendment request (Le., NRC Accession Number ML112650386), dated September 21, 2011, as Attachments 5 and 6, respectively.
Design criteria for LSCS, Unit 1 Cycle 15 are defined in the Cycle Design Inputs and Requirements (CDIR) document, which sets the cycle energy, thermal margins, and other design constraints. This document is included as Attachment 5.
NRC RAI-7. Please provide an updated version of power/flow map for Cycle 15 operation including stability Option III features of scram region and controlled entry region for backup stability protection based on the Boiling-Water Reactor Owners Group position stated in NEDO-31960A for SLO and TLO.
EGC Response to NRC RAI-7 The power/flow map for Cycle 15 will remain unchanged from LSCS, Unit 1, Cycle 14, which is the "LaSalle County Station, Unit 1, Cycle 14 and Expected Cycle 15, Power-to-Flow Map,"
provided in Attachment 10 of the October 12, 2011, license amendment request. The power/flow map for backup stability protection (BSP) at LSCS, Unit 1 conservatively treats the controlled entry region as an immediate exit region (Le., Region 2). Region 1 on the power/flow map is the scram region. The BSP region boundaries are calculated based on a specified core decay ratio per the approved stability methodology described in NEDE-24011-P-A, Revision 18, "General Electric Standard Application for Reactor Fuel (GESTAR II, U.S. Supplement),"
Section S.4.2.2, "Backup Stability Protection (BSP) for Option III," dated April 2011. The core decay ratio is a function of principal reactor core parameters (e.g., power and power distribution, flow, subcooling, fuel design, etc.). The core decay ratio is independent of the core flow mode (Le., the same for two loop operation (TLO) and single loop operation (SLO>>. Therefore, the calculated BSP regions are bounding and applicable for both TLO and SLO.
Attachment 4 Page 2 of 2 The power/flow map generally depicts a "natural circulation" flow line and a "maximum rod line."
The SSP region boundaries are calculated based on points on the natural circulation line and the maximum rod line. The SSP regions are depicted as areas between the maximum rod line, the natural circulation line, and the SSP region boundaries in the high-power, low-flow region of the map. However, the natural circulation line is approximate and the core flow measurement uncertainty is larger at low flow conditions. In the past, this has resulted in operating conditions in which the indicated powerlflow condition was below (Le., to the left of) the natural circulation line on the power/flow map. Also, industry operational experience has identified conditions in which operation above the maximum rod line has occurred. To address these situations, an operational decision was made to conservatively extend operating boundaries (e.g., rod lines, stability regions, etc.) back to "zero flow" and extend the SSP boundaries above the maximum rod line. These operational enhancements to the power/flow map have been made to provide additional guidance to address the unlikely, but possible circumstance of operating at those conditions.
Attachment 5 LaSalle County Station, Unit 1, Cycle 15 Cycle Design Inputs and Requirements (CDIR)
NUCLEAR FUELS TRANSMITTAL OF DESIGN INFORMATION
~ SAFETY RELATED Originating Organization NF 10# NF1100177 o NON-5AFETY RELATED ~ NUClear Fuels Revlsion# 1 o REGULATORY RELATED o Other(specify) N/A SRRS # -3~A~.1~30~--
Page 1 of20 Station: LaSalle Unit: 1 Cycle: 15 Generic: ......;..;.;"-'------
NlA
Subject:
laSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements To: Ph" Hansett (laSalle - RX Eng. Manager) and eC/ECR#: 382897 Charlie Lamb (GNF - Customer Project Manager)
John L. McHale Date
'lJ13h/
r I Prepared by David A. Phegley 9!t:J~1/
Reviewed by Date John 1<. Wheeler Approved by qb+6.
Date Status of information: f2l Verified o Unverified o Engineering Judgment Action Tracking # for Method and SchedUle of Verification N/A for Unverified DESIGN INFORMATION:
Description of Rev. 1: This transmittal updates the CDIR to Revision 1 as indicated by Infonnation: revision bars.
Rev. 0: This information inCludes a listing of assumptions. design margin limitations. design methodology limitations, and expected station modifJCat1ons which wiD impact the L1C15 reload design.
Purpose of Infonnation: This data is to be used to create the L1C15 fuel and core design.
Source of Infonnation: References found within the document.
Supplemental Distribution:
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L1C15 CDIR Rev. 1 LaSalle Unit 1 Cycle 15 Cycle Design Inputs &
Requirements Revision 1 NF1100177 Rev 01 - Page 2 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L1C15 COIR Rev. 1 I COIR laSalle Unit 1 Cycle 1S t Revision 1 I John L. McHale Prepared by David A. Phegley Reviewed by Date John K Wheeler Approve by Signature Date NlA for Rev. 1 ROOT Approval Obtained Signature Date NJA for Rev. 1 RRB Approval Obtained Signature Date Section A - Core Desian Data Related Items Item I Criteria Value References/Comments ExeIon Nuclear Group's Planned OUtage End of Cycle 14 Shutdown Schedule, Revision 28 A1.a 2113112 Date L1C15 EUP
! Exelon Nuclear Group's Planned Outage A1.b Refueling Outage Duration 25 days Schedule. Revision 28
, L1C15 EUP ExeJon Nuclear Group's Planned Outage Beginning of Cycle 15 Schedule, Revision 28 A2 Startup Date 319112 L1C15 EUP Exelon Nuclear Group's Planned Outage End of Cycle 15 Shutdown 2110114 Schedule, Revision 28 A3.a Date L1C15EUP IL1C15 EUP A3.b Cycle 14 Thermal Power Level !
I 3546MWth ' This value is post-MUR. which occurred I I
ion 911712010. Prior to this date the I i Thermal Power Level was 3489 MVVth.
End of Cycfe 14 Nominal 116,300.8 MWdIST AA.a 11,968.6 MWdlMT Exposure IL1C15 EUP A4.b End of Cycle 14 Energy 2440869MWd L1C15EUP 15.900.8 MWdlST (EOC Cycle 14 Nominal Exposure-End of Cycle 14 Minimum A5 11.521.7MWdlMT 440.9 MWdIMT (400 MWdlST)
Exposure L1C15EUP A6.a I Cycle 15 EUP Required EOR Energy 2.332,840 MWd ,l1C15EUP I I NF1100117 Rev 01 - Page 3 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L 1C15 CDIR Rev. 1 Sedlon A - Core Deslan Data Related Items !
Item Criteria Value ReferencesJComments I!
EOR Calculaled k-eff within Cycle 15 EOR Exposure A.6.b following tolerance of target Exelon Design Requirement Acceptance Band
+0.0005 Ak I -0.0 Ak Cycle 15 Thermal Power 3546MWth A.6.c L1C15 EUP Leve!
Cycle 15 Operating A.6.d 98.375% L1C15 EUP (BOC to EOR)
Capacity Factor Cycle 15 EUP Total Cycle A.7 2,445,824 MWd L1C15EUP Energy No MFLPO points may exceed 0.885 in the design cycle.
A.8 Design MFLPO Unlit MFLPOsO.88 Reauired for both MB-2 and PANAC11.
No MFLCPR points may exceed 0.905 In the design cycle.
A.9 Design MFLCPR Umit MFLCPR s: 0.90 Required for both MB-2 and PANAC11.
No MAPRAT points may exceed 0.885 in A.10 Design MAPRAT Limit MAPRAT s 0.88 the design cycle. I Required for both MB-2 and PANAC11. I Based upon EOC 14 Minimum Cycle Exposure Exelon Methodology (MB2) Umits using GNF Methodology are
~ 1.O%AkIk based on using GRETA and a 0.003 Ak Design Cold Shutdown A.11 Margin (SOM) Umit local vs. distributed adjustment GNF Methodology:
~1.0%Ak SOM must be met at multiple temperatures between S8C1f' and 3200f based on short energy window for EOC14.
Exelon Methodoly (MB2):
No Adjustments or Penalties Required ExeJon Methodology (MB2):
GNF OBO 08-0004.01, Rev. 6 O.oo%AkIk Total SOM adjustments and A.12 penalties GRETA is used with GNF design GNF Methodology:
methodology only.
Local ve. distributed 0.003 Ak (within GRETA)
Local vs. distributed adjustment is needed for SOM and MSBWP calculations.
~ 1.0% Ak near BOC
~ 1.5% AI< at peak hot A.13 Hot excess reactivity limit NF-AB-110-2060. Revision 8 guidance.
excess
~ 0.4% Ak min to max NF1100177 Rev 01 - Page 4 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L 1C15 CDIR Rev. 1 Section A - Core Design Data Related Items Item Criteria Value References/Comments MB2:
CycteExp k-crificaJ H~WdIMTI (hot) 0.0 1.0060 5.0 1.0010 MB2:
10.0 0.9990 Developed based upon historical 13.0 0.9990 eigenvalue data from laSalle 1 and 17.0 1.0045 LaSalle 2 Cycles 11 through 14.
Design Cycle Hot Operating A.14 18.0 1.0045 K-effective Target PANAC11:
PANAC11: GNF document "CrHlcal Eigenvalue and Cycle Exp k-critical Thermal Margin Review. laSalle 1 Cycle (GW~T} (hot) 15." Approved 5117111.
0.0 1.0060 4.0 1.0060 14.0 1.0000 20.0 1.0000 MB2:
CycleExp k-criticaI
{gwdIMT} (cold) MB2:
0.0 0.9950 Developed based upon historical 4.0 0.9920 eigenvalue data from laSalle 1 and 20.0 0.9920 laSalle 2 Cycles 11 through 14.
Currant Cycle Cold Critical A.15 K-effective Target PANAC11: PANAC11:
Cycle Exp k-critical GNF document "Critical Eigenvalue and (GWQlMD !£Qkll Thermal Margin Review. laSalle 1 Cycle 0.0 1.0000 15." Approved 5117/11.
4.0 0.9960 20.0 0.9960 Cold Critical K-effectlve No additional adjustments other than A.16 None Adiustments those listed In A.11 and A.12 above.
Standby Uquld Control Assumes natural Boron enriched to 45 (SLC) System Shutdown A. 17 SLCS SOM > 1.0% weight % in the 8-10 Isotope (1571 ppm Capability Requirement -
(SDM) natural boron equivalent) at 68 OF.
Peak pellet exposure must remain within aD pellet exposure based limits provided ATRIUM-10(MWdlkgU): intheCOLR.
s 58.7 rod average ATRIUM-10 limits from L-G03372, exposure Revision 2.
s 54.0 bundle average A.18 Fuel Exposure Limits exposure GNF2:
NEDC-33270P Rev. 3 GNF2 Advantage GNF2: Generic Compliance With NEOE-24011-Peak PeHet- P-A (GESTAR II)
MWDIMT NEOE-24011-P-A-17-US. General Elec:trlc Standard Application for Reactor Fuel I(GESTAR II)
NF1100177 Rev 01 - Page 5 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L 1C15 CDIR Rev. 1 Section A - Core Design Data Related Items Item Criteria Value References/Comments End of Design Gycle Exelon DesIgn Target NEXRAT <0.97 A.19 Fuel Exposure Umits Multi-cycle analysis should include four Perform multk:ycle check of cycles past the design cycle.
NEXRAT None for ATRIUM*10 fuel A.20 Residence TIme limits GNF 08-0004.01 Rev. 6 laSalle currently does not have full core offload capability. Based upon the 2011 A,21 FuU Core Offload Capability None at this time schedule for Dey Cask Storage and rack constraints inserts, full core offload capability Will be
!regained by the end of 2011.
Attempt to maintain octant symmetry as much as Report non-symmetric pairs to reactor A,22.a possible.
Core Symmetry engineering prior to startup if It leads to non-symmetric thermaiUmits.
Maintain symmetry by bundle type and exposure.
Maintain bundles in same quadrant during shuffle as Quarter Core Shuffle practical.
A.22.b Exceptions shall be documented.
Restrictions No face-adjacent shuffles within a ceR.
Assumes SU date of 31912012. These sequence exchange dates take into Target Sequences A,23 See Attachment B account the calendar, Unit 2 exchanges.
Exchange Plans Unit 2 R14 Outage, and quarterly large load drops for surveillance tests.
Improved Low Leakage Core Design (ILLCO)
Implement ExeIon SCORE Principles as Possible A,23.b Core Loading Strategy Exelon Target Evaluate core loading plan changes using the GEH MCO software to achieve < 0.024% MCO Maintain coostant rod patterns between Sequence Exchanges Control Rod Pattem when possible. Avoid use of Group A24.a limitations 8 rods alOne in the core. Group 7A Exelon Target rods may only be placed at I POSitions ()().()8 or 48.
A,24.b Rod SeQuence A-2/ A-1 Maximum Continuous A,24.c "Clean" Rod Sequence Controlled Interval 3.1 = NF-AB440-1002, Rev. 2 Exchanges GWO/MT Core Average Exposure
! Prohibited Design Control A,24.d 02.04.42.44.46 Exelon Design Requirement Rod Positions NF1100177 Rev 01 - Page 6 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L 1C15 CDIR Rev. 1 Section A - Core Desian Data Related Items Item Criteria Value References/Comments Minimize Use of Shallow No shallow rods (> position A.24.e Exelon Design Requirement Rod Positions 24)
No rod pattern changes In summer 2012 A.24.f Minimize Rod Pattern Summer Is defined as 06I01NY -
Changes 08l311YY No rod pattern changes In summer 2013 A.24.g Maximum DevIation from +/-0.0005 ilk SOC - EOR IExelon Target Target E~value + 0.0005 L\k for EOR A.25.a 96.57-10525 MlbnJhr Flow is timlted by unit rod-line restrictions Core Flow Window 89%-97% of rated flow Ilsee A.25.bl.
=
Rodline [%CTP*1oo I (C , +
I
'Unit rOO-line restrictions (low (~*%FLOW) + (~*%FLO~>>]
High: 108%
A.25.b and high) for operation to be used In the reload design.
Low: 101% C,=22.191, ~.89714. Ca=-O.OO11905 I
Reference:
NF-LA-712-2501, Rev. 10 1.11DLO SLMCPR change due to GNF2 TBO.
A.26 Cycle 14 SLMCPR Perform preliminary check of SLMCPR 1.12 SLO Iprior to bundles being finalized.
L1C15 SLMCPR 1.11DLO Actual values are determined after A.27.a Reference Loading Pattern is finalized.
design assumption 1.12SLO Based upon stability analysis setting the Option B Umits:
A27.b Estimated OLMCPR limit for entire cycle.
1.55 SOC to EOC A 27 c Range of Acceptable OPRM OPRM setpoint~ 1.11
. . I Amplitude SetDolnt Values Values to be used for bundle and core ATRIUM-10: design work. If licensing activities '
Same as Cycle 14 change these values, then the updated APLHGR Umits values from the COLR shall be used for A28 GNF2: cycle operation.
Use GNF2-B36-P3 curves Exelon Target (PRIME03 is NRC Iaooroved for GNF2)
Values to be used for bundle and core design work. If licensing activities ATRIUM-10:
change these values. then the updated Same as Cycle 14 LHGR Steady State Umits values from the COLR shall be used for A.29 cycle operation.
GNF2:
Use GNF2-B36-P3 curves Exelon Target (PRIME03 is NRC aooroved for GNF2)
I Engineering Input - no reference needed.
Transient LHGR Umlts 1 MEOD Hmits are used and thus there are A.30 N/A no separate transient LHGR Hmits.
Not applicable to GNF Fuef NF1100177 Rev 01 - Page 7 of 20
laSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements LiCi5 CDIR Rev. 1 Section A - Core Design Data Related Items Item Criteria Value References/Comments Off-rated thermal limits are calculated Off-Rated thennallimits to during the Reload Licensing stage.
be used In the bundle and MCPR(P), MCPR(F). During the Bundle and Core Design A.31 core design phase prior to LHGRFAC(P). stage rated limits are used. At the power the completion of the cycle LHGRFAC(F) and flow conditions used during the core Reload Report design phase, off-rated limits would have little effect.
Fuel channe' distortion Exelon Channel Distortion NF-AB-105 (Rev. 13). Attachment 4, management limitations and Design I Monitoring Criteria Criterion 2 and 3 as well as the general A,32 mitigation strategy for considerations contained in the deskm See Attachment C attachment.
NF-AB-130-2620, Rev. 8 Individual notch worth I Minimum allowed calculated between 04-36 and total Ireactor period and worth between 36-48 must Notch worth must consider sequence A,33.a Imaximum alowed single steps beyond Groups 1-4 (Groups 7.8.9.
be less than the Ak that inotch worth for In-sequence would and/or 10) as necessary to satisfy result in a 50 second requirements up to +3.5% keff around Inotch worth determinations I period critical.
f IRange around the expected -1.2% to +3.5% keffaround NF-AB-130-2620.Rev.8 Icold critical eigenvalue the expected cold critical A.33.b Iwhen the notch worth and eigenvalue evaluated at 120 Bounding range based on uncertainties Istep worth limits are to be OF and 320 OF. for non-BOC critical enforced Step worth (sum of aft notch worth between movement limits) must be less than the minimum of:
0.OO5~
NF-AB-13Q.2620, Rev. 8 or Maximum rod step worth (7164) *' [ko - kARl - 0.0112)
Step worth must consider sequence A,33 allowed in the defined range (~) steps beyond Groups 1-4 (Groups 7.8,9,
.c ,around cold criticality for the where I startup sequence ko = the cold critical target keff at each exposure point and/or 10) as necessary to satisfy requirements up to +3.5% keff around i of interest critical.
I
=
kARl the cold all rods inserted keff at each I exposure point of interest Exelon:
CASM0-4IMICRO-B2 Code package(s) to be used MICROBURN-B2 A, 34 to determine the cold notch worths GNF:
TGBLAOS PANAC11 Core loading restrictions to Notch worths will be verified in the early Avoid triple loading fresh minimize risk of high notch stages of design to ensure a success A. 35 fuel toward the center worths during insequence path exists or the core design will be portion of the core criticals altered as needed.
NF1100177 Rev 01 - Page 8 of 20
LaSalle Unit 1 Cycle is Cycle Design Inputs & Requirements LiCiS CDIR Rev. 1 Section A - Core Design Data Related Items Item Criteria Value References/Comments IMust meet one of these Ithree (uncontrolled 40%
Ivoids)
- 1) LPF 51.4
- 2) Any combination of letter. GEH-HAOWX227-006, "lCS EOP/EPG Calculation LPF and APF such that lPF ,. APF so GNF2 Fuel Transition: F0906 Generic Assumptions and EPG Data," dated April 1. 2011.
A36.a Limitations: 1.4" 2.0 Core and Bundle Peaking 3) lPF so 1.59 for all Factors bundle axial power
=
lPF local Peaking Factor shapes peaked at
=
APF Axial Peaking Factor or beiow the midplane for a I maximum APF of 2.0 EOP/EPG Calculation Assumptions and Letter, GEH-HAOWX227-006. "lCS A36.b Limitations: s38GWDIST GNF2 Fuel Transition: F0906 Generic EOC Core Average EPG Data." dated April 1. 2011.
Exposure EOP/EPG Calculation Assumptions and Letter, GEH-HAOWX227-006. "lCS A36.c limitations: so 5 years GNF2 Fuel Transition: F0906 Generic EOC Core Average EPG Data." dated April 1. 2011.
Effective Fun Power PerIod EOP/EPG Calculation Letter. GEH-HAOWX227-006, "lCS Assumptions and Cl: 3.779 w/0 U235 A36.d GNF2 Fuel Transition: F0906 Generic limitations:
EPG Data." dated April 1, 2011.
Reload Average Enrichment Engineering Input - no reference needed.
L1C14 MSBWP was 02. If the 02 position EOPIEPG Calculation All rods at position 02 or Is not met. must contact station Assumptions and deeper and the strongest EPGlSAG coordinator.
A36.e rod out yields limitations:
MSBWP ~ 1.oo%~SoM $OM must be met at multiple temperatures between 68°F and 32QOF based on short energy window for EOC14.
The 45% enriched B-10 shall be factored into these calculations (see item A 17).
EOPIEPG Calculation Assumptions and Cold: SoM ~ 1.00 %AkIk A36J SOM must be met at multiple Limitations: Hot: SOM ~ 0.00 %~ temperatures between 680F and 3200F SLC System Requirements based on short anergy window for I EOC14.
Evaluate core loading plan Engineering input - no reference needed.
A.37 Moisture Carryover using GEH MCO software to this value is an administrative limit used achieve 5 0.024% MeO.
at LaSalle.
NF1100177 Rev 01 - Page 9 of 20
laSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L 1C15 CDIR Rev. 1 Section A - Core Design Data Related Items Item Criteria Value ReferenceS/Comments TOOl NFM9900058 Rev. O.
EOC13 CAVE}( < 36.000 AREVA Letter CMP:05:057. "Transmittal MWdlMTU& of FANP Letter Report on Disposition of Initial core average Several Safety Scope of Work Items".
A38 Decay Heat load enrichment for new fuel > C.M. Powers to R.J. Chin, dated 3.5 w/o (treat all fuel as 1011212005.
fresh)
The values listed may be updated as part of NFl to GNF2.
Fluence analysis owned by station Fluence Analysis See DA L-002869, current A39 design engineering if further Information Assumptions revision needed.
GNF2 - GEXl17 0.94 < R-Factor < 1.30 GNF 08-0003.26 R8 (ill aI bows, controlled! GNF AG-0024.01 R3 CPR correlation type(s) and uncontrolled. a/l exposures) GEXl17 for GNF2 A40 ATRIUM-10 GEXL97(03) limitations NEDC-33292 R3. June 2009 1.02 < R-Factor< 1.20 GEXl97(03) for A10 (ill aI bows, controlledl NEDC-33106 R1. June 2003 uncontrolled aI
- 1. Bundle enrichment range from 3.38 to 4.50 w/o U235.
- 2. Bundle uranium loading AmerGen Calculation No. C-1101-202-(nominal) < 197.0 kgU. E520-443. Revision 1. "PWR & BWR
- 3. Rod exposure < 58.700 Isotopic Inventories for Spent Fuel Pool Spent fuel pool gamma MWDIMTU. Gamma Heating Study."
A.41 heating constraints on 4. Average bundle power bundle and core design (reactor rated thermal These criteria are bounding values and power I # bundles) < are the basis for spent fuel pool heating 5.566MWt calculations as shown in the above
- 5. Radial Peaking Factor < document 2.00.
- 5. Axial Peaking Factor <
1.SO.
No fuel on the outer two Fuel promotion from exterior rows In l1C14 may be A42 location to interior location NF*AB-110-2210, Rev. 13 shuffled inward toward the limitations for design center of the core.
These values are representative values as they change constantly as the steam A.43 Reactor dome pressure and 1020psia flow. etc. change in the plant The reactor feedwater temperature 428.5deg. F heat balance in the 3d simulator codes is used to calculate these values as needed if they are not directly input One bundle will be considered for A.44 I Reinsert fuel bundles for 0-1 reinsertion Into l1C15. This bundle (43C135) has one cycle of operation this cycle's design from l2C13 and was identified as a faHed assembly during l2R13.
NF1100111 Rev 01 - Page 10 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L1C15 CDIR Rev. 1 I Section B - Fuel Bundle Desian Data Related Items Item Criteria Value References/Comments "LaSalle Unit 1 and Unit 2 Nuclear Power Unit 1 Spent fuel pool Station Spent Fuel Pool Criticality Safety B.1.a criticality limits for See Attachment D Analysis with GNF2 Fuel," AREVAANp*
GNF2fuei 3003(P). Revlslon 1. June 2011 Unit 2 Spent fuel pool "LaSalle Unit 1 and Unit 2 Nuclear Power criticality for Station Spent Fuel Pool Criticality Safety B.1.b GNF2fuei See Attachment D Analysis with GNF2 Fuel," AREVA ANP-
"Include Boraflex 3OO3(P), Revision 1, June 2011 Assumotlon "LaSalle Unit 1 and Unit 2 Nuclear Power Unit 2 Spent fuel pool StatIon Spent Fuel Pool Criticality Safety B.i.e criticality compliance with See Attachment D AnalysiswithGNF2 Fuel," AREVAANP-NETCO tnsens 3003(P), Revision 1. June 2011 For GNF2 Fuel at all axial levies:
- 1) Fuel may be stored with or without channels "laSalle Unit 1 and Unit 2 Nuclear Power Station New Fuel Storage Vault Criticality
- 2) Maximum Lattice Safety Analysis for GNF2 Fuel," AREVA New fuel vault criticality Average EnrIchment. wt% ANP-3008(P), Revision 0, June 2011 B.2 criteria for reload fuel U-23S:4.70
.. Face adjacent gadollnla rods are
- 3) Minimum Number of treated as a single gadolinia rod.
Rods containing Gd:P3: 6*
- 4) Minimum wt% Gd:P3 in these Gd rod: 3.0 RAJ-II shipping cask will be Criticality criteria for fuel used. Criticality criteria are B.3 GNF DB-0003.26 Rev. 8 shipping containers provided in the GNF2 1 Desfoo Basts Document LS Exempt from LS must meet 10 CFR 5O.68(b) in lieu of 10 CFR 70.24 criticality B.4 10CFR70.24 10 CFR 70.24 as is documented in monitoring exemptions for criticality monitorlnQ UFSAR Section 9.1 Fuel type to be New fuel type Is being introduced for B.S.a GNF2 manufactured (fresh fuel) laSalle 1.
Cycle 16: GNF2 Extended Power Uprate (EPU)
Fuel Type for Multl-cycle Cycle 17: GNF2 (EPU)
B.S.b information for Cycles 17.18, and19 is Analysis Cycle 18: GNF2 (EPU) documented In Attachment A.
CVcle 19: GNF2 jEPU)
II B.5.c Fuel Manufacturing Constraints I
GNF Fuel Contract GNF Design BasIs Document GNF DB-0003.26 Rev. 8 Source for fuel product line Nuclear Design Bases -
B.6 dimensions Fuel Bundles (GNF2)
GNF OB-0003.26 Rev. 8 Assumptions made for this B.7 cvcIe's channel bow 35 mils Value determined by GNF I NF1100177 Rev 01 - Page 11 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L 1C15 CDIR Rev. 1 Section B - Fuel Bundle Design Data Related Items Item Criteria Value References/Comments GNF2 - GEXl17 0.94 < R-Factor < 1.30 GNF 08-0003.26 RS
(@ all bows, controlled{ GNF AG-0024.01 R3 CPR correlation limitations uncolltrolled. all expclSUfEIS) GEXL17 for GNF2 B.S (R-Factor) ATRIUM-10 GEXL97(03) NEDC-33292 R3 t June 2009 1.02 < R-Factor < 1.20 GEXL97(03) for A10
(@ all bows. controlled{ NEDC-33106 R1. June 2003 uncontrolled all
<1.4 for 40 VH at 0.0 GWDIST (HOTUNC) Exelon targets. May be exceeded with
<1.2 for 40VH at 20 Exelon approval.
GWD/ST (HOTUNC)
Local Peaking Factor <1.S5 for 40VH at 0.0 Based upon experience with TGBLA06, B.9 Constraints GWD/ST (HOTCON) this is applied to TGBLA06 and compared to CASM0-4.
<1.090 for 70 VH at 20 GWD/ST, VAN zone for fuel NF-AB-110-2000. Rev. 4.
on first row (HOTUNC)
Minimize movement of fuel This takes up an inordinate amount of bundles to a face adjacent time on the refuel floor and in most cases 8.10.a Core loadng restrictions position in the same fuel Is not necessary In achieving a good cell. core design.
Number of cycles on same 8.10.b Core loading restrictions GE SIL 320. NF-AB-110-2210 (Rev. 13) water face < 2 No gad rods face adjacent to water rods and minimize Limitations on placement of gad rods in Exelon Requirement B.11 placement/enrichment of location D-4 and G-7 (face GNF 08-0003.26 Rev. 8 gadoJinia rods adjacent to 2 short PLRs)
No gad rods face adjacent to one another Limitations on the amount Per GNF2 design basis B.12 of gadollnia allowed In fuel GNF 08-0003.26 Rev. 8 document.
pellets Target bundle average Past GNF2 Bundle DesIgn used by B.13.a 3.8-4.2%
Enrichment Exeion.
Per GNF2 DesIgn Basis j I
B.13.b Allowed Enrichments GNF 08-0003.26 Rev. 8 Document 1 Max. Number of Fuel Rod Per GNF2 Design Basis B.13.c GNF 08-0003.26 Rev. 8 Enrichments Document Fuel Rod Enrichment Axial Per GNF2 Design Basis B.13.d GNF 0B-0003.26 Rev. 8 Variation Document Target minimum length for 8.14 6 Inches GNF 0B-0003.26 Rev. 8 bottom fuel zone Natural Uranium Top and GNF 08-000326 Rev. S, top zone was B.15 6 Inches i Bottom 11 inches in previous cyde.
NF1100177 Rev 01 - Page 12 of 20
LaSalle Unit 1 Cycle is Cycle Design Inputs & Requirements l1CiS CDIR Rev. 1 Section B - Fuel Bundle Deskin Data Related Items Item Criteria Value References/Comments MICROBURN-B2:
Derived from PANAC11 MICROBURN-B2 methodology will utilize and thus the limit below R-Factor penalties as provided by GNF Maximum allowable control applies. to account for CBH CPR Impacts In the B.16 blade history delta PANACEA: fresh fuel using the GEXL11 CPR MFLCPR penalty IRFGEN=2 correlation.
Maximum delta MFlCPR <
0.005 between IRFGEN =1 NF-AB-110-2010. Rev. 9 and IRFGEN =2 Rotated Bundle shaH not set OLMCPR.
B.17 Rotated bundle delta CPR <0.31 GNF DB-0003.26 Rev. 8 Exefon Target Fuel manufacturing constraints which may be B.18 None challenged In this reload design Integrated Exposure Peak Exelon target Exposure peaking limits on (REX) < 1.15 at 45 B.19 Confirm acceptable NEXRAT in multi-the lattice designs GWO/ST, 40% VHf BASE cycle.
lattice Maximum number of unique B.20 rod types 23 GNF 08-0015 Rev. 0 Exelon target value. This Is an engineering value being set by the COIR.
Goal is to minimize Gd suppression Gad suppression> -1.5%
Target gadolinium penalty on LHGR. Desire is to meet B.21 Maximum gadolina suppression penalty
=
suppression -5.0%
target for all lattices at all exposures. For limited cases, it may exceed target but must stay below maximum urness approved by the BWR Design Manager.
< 1.08 @ 0 GWO/ST (hot, Exelon Target. This Is an engineering B.22 Target R-Factors uncontrolled) for fresh fuel value being set by the COIR.
in core interior I B.23 Target local peaking factors SeeB.9 i I B.24 Gad Rod Concentrations Per GNF2 Design Basis Document GNF 08-0003.26 Rev. 8 I Maximum Gad B.25 8% GNF OB-0003.26 Rev. 8 Concentration Gad Rod concentration Per GNF2 Design Basis I B.26 I Axial Variation Document GNF 08-0003.26 Rev. 8 I Gad Rod I Tie Rod Per GNF2 Design Basis B.27 Locations GNF OB-0003.26 Rev. 8 Document Max. Number of Fresh B.28 Bundle Types 8 GNF OB-0015 Rev. 0 Minimum number of B.29 16 GNF 08-0015 Rev. 0 bundles per batch I
Maximum number of unique I B.30 Gad Rod types 9 GNF 08-0015 Rev. 0 Maximum number of unique B.31 7 GNF OB-0015 Rev. 0 Gad Pellet types NF1100177 Rev 01 - Page 13 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L1C15 CDIR Rev. 1 Section C - Miscellaneous Design Data Related Items Item Criteria Value References/Comments Will only be needed at POWERPLEX-III stage but the general approach is used I Fuel conditioning guldeUnes NF-AB-440 and C.1 in design stage to set limits for fuel to be used in this design NF-AB-44Q-1002 shuffles (power changes) and rod pattern creation (fuel condltioninQ "friendly").
Exelon:
CASM0-4 (2.05.14)
Computer code{s) to be MICRO-B2 (UMAR04R1) used for lattice and core MICROBURN-B2 C.2 design along with any (UAPR05R1) special version requirements GNF:
TGBLA06IPANAC11 Power-flow map - may be referenced if In procedure LaSale Station Procedure or other controlled C.3 LOA-RR-101 document which is made Current Revision available to the reload vendor listing of required Technical SpecifICatIon changes for the reload (I.e. SLMCPR (TBD) due to the It has not been determined if a SLMCPR CA changed SLMCPR. introduction of GNF2 change will be required.
methodology. computer code, etc.)
High Duty Blades:
WestinghoUse CR99 listing of core component
- -7 changes planned (control NA300 LPRM instruments are currently C.5 blade types. TIPILPRM Low Duty Blades used for replacements In LS units.
GE Marathon Ultra MD types, ete.)
- -59 LPRM 1_'
GNF2NFI ATWS Analysis for GNF2 Any studies or projects Source Term Analysis with If ATWS analysis does not pass with which are in progress which GNF2, including any follow-C.G could impact the reload on design basis accident currently SLC capability. may need to use GE14 bundle design.
design or reload licensing recalculations I
Fluence effects due to reduced size of top natural lattice In GNF2 NF1100177 Rev 01 - Page 14 of 20
laSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L1C15 CDIR Rev. 1 Section C - Miscellaneous Design Data Related Items Item Criteria Value ReferencesIComments Peak rod average exposure
< 54 GWD/MT or, Reg. Guide 1.183 "Alternative Limits related to the 1! peak rod average Radiological Source Terms for exposure ~ 54 GWDIMT, radiological alternate Evaluating Design Basis Accidents at then peak rod average C.7 source term analysis (rod Nuclear Power Reactors" LHGR must not exceed bumups. kWlft nmHations.
6.3 kW/ft peaking factor limits, etc.) l-003067, Rev. 2 - Fuel Handling Accident Radial Peaking Factor maintained :'S 1.7.
Items which are determined C.S None specified at this time to be Critical to Quality.
I Fuel type to be used in C.9 downstream multicycle GNF2 I analyses NF1100177 Rev 01 - Page 15 of 20
laSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L1C15 CDIR Rev. 1 I Section 0 Item Criteria Value References/Comments Planned Station 0.1 GNF2 Introduction Modifications of Impact Planned Setpoint Changes II 0.2 of Impact None RDOTIRRBlSMDI Planned Station Minor 0.3 None RDOTIRRB/SMOI Modifications of Impact Planned Component 0.4 Chanaes of Impact See items C.5 and C.G ROOT/RRB/SMDI NF1100177 Rev 01 - Page 16 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements L 1C15 CDIR Rev. 1 Attachment A LaSalle 1 Cycle 15 EUP I Rated LaSa'le1 ReactorTMmtaI Cvcle 14 3.54&
Cvcle 15 CYCle 16 CYCle 17 Cyele18 Cycle 19 Power,MWth (Post-MUR) 3.54& 3,54& 3.988
- 3,988 Beglnnlng-of.CycIe (SOC)
Date 03105110 03109#12 03101/14 03111116 03112118 04101120 End-of.Cvcle (Eoo) Dat8 02113112 02110114 02122116 03104118 03114120 03115122 Post-outage Length. Dave: CYCle 25 19 18 18 18 18 Mid-Cycle 0 0 0 0 0 0
~tlng Cycle Length.
Dave 110 103 123 123 723 723 (BOCtoEOC}
Total Cycle Length. Days (SAJtoSlu) 735 722 141 741 741 141 (Includes Planned OUtages)
Full Power CC1pabIIity (End ofRatecn: Dave 110.0 66B.7 884.4 682.8 682.4 680.9 MWd 2A40,869 2,332,840 2,383173 2,614166 2673101 2,667,295
- 17. 18,940.2 11,012.4 18924.3 18912.6 18,871.5 EFPO 688.3 657.9 612.2 67G.7 670.3 688.8 Full Power Cycle Extension C : Dave 0.0 0.0 0.0 0.0 0.0 0.0 BWRs: ICF/FFWTR MWd 0 0 0 0 0 0
- PWRs: ACTRlAPSR Pull MWdIMT 0.0 i 0.0 0.0 0.0 0.0 0.0
~
EFPD 0.0 0.0 0.0 0.0 0.0 Total Full Power CaDabiIltv: Dave 710.0 666.7 884.4 682.8 882.4 MWd 2.440.669 2,332,840 2.383173 2.674,766 2673,101 2,667,295 MWdIMT 1196B.6 16940.2 11012.4 18924.3 18,912.6 18871.5 EFPO 688.3 657.9 672.2 610.7 67G.3 688.8 Coaetdown Cycle ExIsnsion: Days 0.0 34.3 38.8 40.2 40.8 42.1 MWd 0 112.983 126.227 147.351 148.808 153874 MWdIMT 0.0 820." 900.8 1042.5 1052.8 1.088.7 EFPO 0.0 31.9 35.6 36.9 37.3 38.6 Cycle Operating Capacity Factor. % 98.95 98.11 97.90 97.88 91.87 91.14 (Excludes refueling oulagtt)
Cvcle capacity Factor. % 93.65 .52 95.50 95.49 95.47 I (IncludeS refueling outage &
i MOO)
I Total CYCle Enenw: MWd MWdIMT 2,440669 11968.6 2445824 17,160.7 2,510000 17,913.2 2~ 2821910 19,965.4 2,821169 19960.2 EFPD $88.3 689.1 101.8 701.1 707.6 707.4 NF1100177 Rev 01 - Page 17 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements l1C15 CDIR Rev. 1 Attachment B Target Sequence Exchange Plans Sequence Sequence Sequence Cumulative Rod Start End Interval Exposure Group Date Date MWd/ST MWd/ST I
A2 3/9/12 5/20/12 1678 1678 A1 5/21/12 9/2/12 2413 4090 A2 9/3/12 12/16/12 2413 6503 A1 12/17/12 3/24/13 2252 8755 A2 3/25/13 5/26/13 1448 10203 Al 5/27/13 9/8/13 2413 12616 A2 9/9/13 12/15/13 2252 14868 ARO 12116/13 1/5/14 483 15351 Coast 1/6/14 2/10/14 804 16155 NF1100177 Rev 01 - Page 18 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements l1C15 COIR Rev. 1 Attachment C Channel Distortion Design I Monitoring Criteria Shadow..cooosion Based Threshold
- 1. No cen may contain one or more channels with >= 64 EFIO for Ziro-2 channels or >= 120 EFIO for Zirc-4 channels ~ >=36 GWdIMT channel exposure - all fuel violating this criteria must be rechanneled. discharged, or moved within the loading.
- 2. Cells that will contain one or more channels with >= 64 EFIO (Zlrc-2 or Zirc-4 channels) at end-of-cycle must be monitored.
- a. Monitoring to begin prior to any channel In the cen exceeding 33 GWdlMT channel exposure.
- b. Rechannel at site discretfon to minimize/eliminate required monitoring.1 Fast Fluence Gradient Based Threshold
- 1. ATRIUM*10 Fuel Only - No cell configuration may violate the Control Rod Friction Surveillance Recommendations (Interim Supplemental Surveillance for AREVA Fuel Channels in Core Periphery Locatlons) - rechannel. discharge. or move fuel within the loading to ensure no cells violate this criteria.
- 2. No cell configuration may violate the GE Nuclear Energy SIL No. 320, Supplements 1, 2. and 3 (Mitigation of the Effects of Peripheral Core Location on Fuel Channel Bowing C-Lattice Criteria)
- rechannel, discharge. or move fuel Within the loading to ensure no cells violate this criteria.
- 3. No peripheral cell may contain two or more channels with >= 43 GWdIMT channel exposure at end-of..cycle - rechannel! discharge, or move fuel within the loading to ensure no cells violate this criteria.
- 4. Peripheral cells that will contain one channel with >= 43 GWdIMT channel exposure at end-of-cycle must be monitored.
- a. Monitoring to begin prior to the leading channel exceeding 37 GWdlMT channel exposure.
- b. Rechannel at site discretion to minimize I eliminate required monitoring. 1 1 Note that ATRIUM-1 0 fuel bundles may only be re-channeled at the end of their first operating cycle and not later.
NF1100177 Rev 01 - Page 19 of 20
LaSalle Unit 1 Cycle 15 Cycle Design Inputs & Requirements LiCiS CDJR Rev. 1 Attachment D Spent Fuel Pool Criticality limits The fuel may be stored In the Spent Fuel Pool provided the lattice average enrichment Is less than 5.0 wt% U-235. and the k... of each enriched lattice does not exceed the following in-rack k... values at any point during its lifetime.
Zone Distance from BAF Max. in-rack It..
TS3 > 126"toTAF 0.9185 TS2 > 96" to 126" 0.8869 TS1 O"to96" 0.8843 Fuel lattices that meet the lJ,.235 enrichment and gadollnia requirements described below have been shown to meet these requirements.
Distance from BAF t!: 10 Gad Rods I lattice ~ 13 Gad Rods I Lattice t!: 10 Gad Rods I lattice
> 126"toTAF s 4.84 wt% U-235 s 4.95 wt% U-235 s 4.95 wt% U-235 2: 4.00 wt% Gd203 2: 4.00 wt<'AI Gd203 2: 5.00 wt% Gd203
> Y'* to 126" s 4.68 wt% lJ,.235 S 4.85 wt% U-235 S 4.95 wt% U-235 2: 6.00 wt% Gd203 2: 6.00 wt% Gd203 2: 7.00 wtOJo Gd203
>X"toY" s 4.13 wt% U-235 s 4.40 wt% lJ,.235 s 4.40 wt% U-235 2: 6.00 wtOJo Gd203 2: 6.00 wt% Gd203 2: 7.00 wtOJo Gd203 O"toX" s 4.27 wt% lJ,.235 S 4.48 wt<'k U-235 S 4.63 wt<'/o U-235 2: 6.00 wtOJo Gd203 2: 6.00 wt<'k Gd203 2: 7.00 wt<'AI Gd203 Note: Elevations X" and Y'* are proprietary to GNF. GEH, or GE. If needed, Nuciear Fuels can be contacted to discuss these values.
Note: Enriched lattices within each bundle must meet one of the three possibilities for the respective "Distance from BAP range. Not allattices within the bundle have to meet the requirements from the same column.
NF1100177 Rev 01 - Page 20 of 20