View source for RS-09-134, Supplemental Information Concerning Request for License Amendment to Revise TS 3.4.5, RCS Leakage Detection Instrumentation, TS 5.6.5, Core Operating Limits Report, & Renewed Facility Operating.

{{Adams
| number = ML092790207
| issue date = 10/05/2009
| title = Supplemental Information Concerning Request for License Amendment to Revise TS 3.4.5, RCS Leakage Detection Instrumentation, TS 5.6.5, Core Operating Limits Report, & Renewed Facility Operating.
| author name = Hansen J
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000237, 05000249, 05000254, 05000265
| license number = DPR-019, DPR-025, DPR-029, DPR-030
| contact person = 
| case reference number = RS-09-134
| document type = Letter type:RS
| page count = 21
| project = 
| stage = Supplement
}}

=Text=
{{#Wiki_filter:www.cxeloncorp.com Exelon Ceneration 4300 Winfield Road Nuclear Warrenville, IL 60555 10 CFR 50.90 October 5, 2009 RS-09-134 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

==Subject:==
Supplemental Information Concerning Request for License Amendment to Revise Technical Specification (TS) 3.4.5, "RCS Leakage Detection Instrumentation," TS 5.6.5, "Core Operating Limits Report (COLR)," and Renewed Facility Operating License

==References:==
: 1. Letter from J. L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Request for License Amendment to Technical Specification (TS) 3.4.5, "RCS Leakage Detection Instrumentation," TS 5.6.5, "Core Operating Limits Report (COLR)," and Renewed Facility Operating License," dated April 7, 2009
: 2. Letter from C. Gratton (U. S. NRG) to C. G. Pardee (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Request for Additional Information (TAC Nos. ME1053 thru ME1 056), II dated September 14, 2009 In Reference 1, Exelon Generation Company, LLC (EGG) submitted a request to amend Appendix A, "Technical Specifications," (TS) of Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively.

October 5,2009 U. S. Nuclear Regulatory Commission Page 2 The proposed amendment deletes a footnote from DNPS TS 3.4.5, "RCS Leakage Detection Instrumentation" that was incorporated as part of a limited duration emergency license amendment approved and implemented in August 2008, and is no longer applicable.
The proposed amendment corrects administrative errors in the titles of analytical methods (Le.,
NRC-approved topical report references) and deletes obsolete analytical methods in DNPS and QCNPS TS 5.6.5, "Core Operating Limits Report (COLR)," paragraph b.
Finally, the proposed amendment deletes a license condition from the DNPS Units 2 and 3 and QCNPS Units 1 and 2 renewed facility operating licenses that limited the maximum rod average burnup for each unit.
In Reference 2, the NRC provided a request for additional information (RAI) related to the proposed license amendment. In response to this request, EGC is providing the information in the attachments to this letter.
There are no regulatory commitments in this letter or attachments.
Should you have any questions or require additional information, please contact Mr. John L.
Schrage at (630) 657-2821.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 5th day of October 2009. :    Exelon Generation Company, LLC Response to Request for Additional Information :    Letter from C. P. Patel (U. S. NRC) to T. J. Kovach (Commonwealth Edison Company), "Commonwealth Edison Company Topical Report NFSR-0091,
                 'Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods' (TAC NO. M82731)," dated March 22,1993

ATTACHMENT 1 Exelon Generation Company, LLC Response to Request for Additional Information NRC RAI Question 1 "With respect to proposed change to Technical Specification (TS) 5.6 .5.b, "Core Operating Limits Report [COLR]," the list of approved methodologies consists of General Electric, Combustion Engineering, and Westinghouse topical reports for DNPS Units 2 and 3, and QCNPS Units 1 and 2.
Please provide the following information:
(1)     Clarify the need for different vendors' methodologies; (2)     Describe the current fuel loading pattern; (3)     Justify why other fuel vendor's methodologies are still listed in the COLR TS if there is no mixed core loading pattern; (4)     Identify which cycle-specific parameters listed in TS 5.6.5.a is [are] supported by methodologies listed in TS 5.6.5.b; (5)     Provide approved date and version for the methodologies listed in TS 5.6.5.b; (6)     Provide NRC approved letter or safety evaluation report for proposed TS 5.6.5.b.1 for DNPS and TS 6.6.5.b.2 [sic] for QCNPS, Commonwealth Edison Company Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods."
Exelon Generation Company, LLC lEGC) Response (1)     Clarify the need for different vendors' methodologies (3)     Justify why other fuel vendor's methodologies are still listed in the COLR TS if there is no mixed core loading pattern Dresden Nuclear Power Station (DNPS), Units 2 and 3 and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2 are currently operating with mixed cores that contain both Global Nuclear Fuel (GNF) GE14 and Westinghouse Optima2 fuel bundles.
Analytical methods from both fuel vendors are included in Technical Specification (TS) 5.6.5.b.1 to support the generation of the COLR for mixed cores.
(2)     Describe the current fuel loading pattern DNPS Units 2 and 3 and QCNPS Units 1 and 2 are currently operating with mixed cores of GNF GE14 fuel bundles and Westinghouse Optima2 fuel bundles. The oldest fuel in each core is GE14. Optima2 fuel has been loaded in recent cycles and will continue to be loaded in upcoming cycles, so that all four units will eventually discharge all GE14 fuel and the cores will consist entirely of Optima2 fuel. The historical and planned loading with mixed cores is described below for each unit.
Page 1 of 8

ATTACHMENT 1 Exelon Generation Company, LLC Response to Request for Additional Information DNPS Unit2 In DNPS Unit 2, EGC first loaded Optima2 fuel in Fall 2007 for cycle 21, which is the current operating cycle. The Unit 2 core currently contains 480 GE14 and 244 Optima2 fuel bundles. For cycle 22, EGC will load 260 new Optima2 fuel bundles, for a total of 504 Optima2 and 220 GE14 fuel bundles. This cycle is expected to commence in late-November 2009. The DNPS Unit 2 core will consist entirely of Optima2 fuel beginning in Fall 2011 with cycle 23. EGC will also load Optima2 fuel for cycle 24.
DNPS Unit3 In DNPS Unit 3, EGC first loaded Optima2 fuel in Fall 2006 for cycle 20. The current operating cycle is cycle 21. The Unit 3 core currently contains 216 GE14 and 508 Optima2 fuel bundles, including 244 Optima2 bundles that were loaded for cycle 20 and 264 Optima2 bundles that were loaded for cycle 21. The DNPS Unit 3 core is expected to consist entirely of Optima2 fuel beginning in Fall 2010 with Cycle 22. EGC will also load Optima2 fuel for cycle 23.
aCNPS Unit 1 In QCNPS Unit 1, EGC first loaded Optima2 fuel in Spring 2007 for cycle 20. The current operating cycle is cycle 21. The Unit 1 core currently contains 196 GE14 and 528 Optima2 fuel bundles, including 260 Optima2 bundles that were loaded for cycle 20 and 268 Optima2 bundles that were loaded for cycle 21. The QCNPS Unit 1 core is expected to consist entirely of Optima2 fuel beginning in Spring 2011 with cycle 22.
EGC will also load Optima2 fuel for cycle 23.
aCNPS Unit 2 In QCNPS Unit 2, EGC first loaded Optima2 fuel in Spring 2006 for cycle 19. The current operating cycle is Cycle 20. The Unit 2 core currently contains 236 GE14 and 488 Optima2 fuel bundles, including 228 Optima2 bundles that were loaded in cycle 19 and 260 Optima2 bundles that were loaded in cycle 20. The QCNPS Unit 2 core is expected to consist entirely of Optima2 fuel beginning in Spring 2010 with Cycle 21.
EGC will also load Optima2 fuel for cycle 22.
Page 2 of 8

AITACHMENT1 Exelon Generation Company, LLC Response to Request for Additional Information (4)     Identify which cycle-specific parameters listed in TS 5.6.5.a is supported by methodologies listed in TS 5.6.5.b; Table 1 below provides a cross-reference of TS 5.6.5.a cycle-specific parameters to the TS 5.6.5.b analytical methodologies that are used to determine each of the cycle-specific parameters.
Table 1          Cross-Reference of Cycle-Specific Parameters and Analytical Methodologies TS 5.6.5.a                                              TS 5.6.5.b Applicable Analytical Cycle-Specific Parameter                                Methodology TS 5.6.5.a.1                                            NEDE-24011 TS 3.2.1                                                CENPD-300-P-A Average Planar Heat Generation Rate                     WCAP-15682-P-A (APLHGR)                                                WCAP-16078-P-A CENPD-390-P-A TS 5.6.5.a.2                                            NEDE-24011 TS 3.2.2                                                CENPD-300-P-A Minimum Critical Power Ration                           WCAP-16081-P-A (MCPR)                                                  CENPD-390-P-A TS 5.6.5.a.3                                            NEDE-24011 TS 3.2.3                                                CENPD-300-P-A Linear Heat Generation Rate                             WCAP-15836-P-A (LHGR)                                                  WCAP-15942-P-A CENPD-390-P-A TS 5.6.5.a.4                                            CENPD-300-P-A TS 3.3.2.1 Allowable Value Control Rod Block Instrumentation Setpoint for the Rod Block Monitor-Upscale Function TS 5.6.5.a.5                                            NEDO-32465-A SR 3.3.1.3.3 trip function setpoints Oscillating Power Range Monitor (OPRM) trip function setpoints Page 3 of 8

ATTACHMENT 1 Exelon Generation Company, LLC Response to Request for Additional Information (5)     Provide approved date and version for the methodologies listed in TS 5.6.5.b; The applicable version and NRC approval date for the analytical methodologies that are retained in TS 5.6.5.b are delineated in Table 2.
Table 2     Retained Analytical Methodologies Methodology            Applicable          NRC Approval Version                Date NFSR-0091              Revision 0           22 Mar 1993 NEDE-24011             Revision 15          17 Mar 2005 NEDO-32465-A           Revision 0           04 Mar 1996 CENPD-300-P-A          Revision 0           24 May 1996 WCAP-16081-P-A         Revision 0           23 Dec 2004 WCAP-15682-P-A         Revision 0           10 Mar 2003 WCAP-16078-P-A         Revision 0           15 Oct 2004 WCAP-15836-P-A         Revision 0           28 Sept 2005 WCAP-15942-P-A         Revision 0           06 Feb 2006 CENPD-390-P-A          Revision 0           24 Jul2000 (6)     Provide NRC approved letter or safety evaluation report for proposed TS 5.6.5.b.1 for DNPS and TS 6.6.5.b.2 [sic] for QCNPS, Commonwealth Edison Company Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods. provides the March 22, 1993 letter from C. P. Patel (U. S. NRC) to T. J. Kovach (Commonwealth Edison Company) transmitting the NRC's review and approval of Commonwealth Edison Company Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods."
Page 4 of 8

ATTACHMENT 1 Exelon Generation Company, LLC Response to Request for Additional Information NRC RAI Question 2 "With respect to the request of deletion of license condition from DNPS and QCNPS Facility Operating Licenses, please provide information as follows:
(1)      Describe the relationship between generic environmental assessments and an extended maximum rod average burnup limit of 62 gigawatt-days per metric ton of uranium; (2)      Clarify any relevant relationship and impact between the license condition and content given in the references (2) and (3) of References Section 7 .0 of the submittal; and (3)      Provide the proposed maximum burnup limit for the current reactor operation."
EGC Response (1)      Describe the relationship between generic environmental assessments and an extended maximum rod average burnup limit of 62 gigawatt-days per metric ton of uranium; (2)      Clarify any relevant relationship and impact between the license condition and content given in the references (2) and (3) of References Section 7 .0 of the submittal; The relationship between the existing DNPS and QCNPS fuel burnup license condition and 1) the generic environmental assessments; and 2) the content of the cited references is provided in the summary below. This summary describes the historical and regulatory basis for the addition of the license condition, the generic environmental assessments, and the applicability of the generic environmental assessments to DNPS and QCNPS, thus providing the regulatory basis for the deletion of the license condition.
Addition of License Condition In letters dated August 3,1999 and September 29,2000 (i.e., References (2) and (3) of the April 7, 2009 submittal), Commonwealth Edison (i.e., the predecessor to Exelon Generation Company, LLC) submitted license amendment requests (LARs) for DNPS and QCNPS to support a change in fuel vendor. In part, these LARs requested NRC approval to add new analytical methodologies to the COLR section of the DNPS and QCNPS TS.
The proposed COLR analytical methods included an NRC-approved licensing topical report (LTR) that permitted the use of extended burnup limits up to 62 gigawatt-days per metric ton of uranium (GWD/MTU) rod average (i.e., EMF-85-74(P), Revision 0, "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Supplements 1 and 2 (P)(A), Siemens Power Corporation, February 1998"). During the review of the LARs, the NRC indicated that a generic environmental assessment had not yet been completed to support the higher fuel burnup limit.
Page 5 of 8

ATTACHMENT 1 Exelon Generation Company, LLC Response to Request for Additional Information As stated in NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants," Volume 1, Addendum 1, the intent of this environmental assessment study was to analyze the cumulative impacts associated with transportation of spent nuclear fuel (SNF) to support the NRC's review of nuclear power plant license renewal requests.
In response to the NRC's concerns with the LARs, Commonwealth Edison proposed, by supplemental letter dated August 27,2001, the addition of a new license condition that limited the rod average burn up to 60 GWD/MTU, until an environmental assessment was completed that supported a higher burnup limit. The NRC found the proposed license condition to be acceptable. In safety evaluations (SEs) dated September 21,2000 and December 20,2001 (Le., References (4) and (5) of the April 7, 2009 submittal) the NRC approved the license amendment for DNPS and QCNPS. In those SEs, the NRC stated that the LTR supported increased burnup limits up to 62 GWD/MTU rod average.
Completion of Generic Environmental Assessment In the Federal Register, Volume 64, No. 171, dated September 3, 1999 (Le., 64 FR 48496), "Rules and Regulations, 10 CFR Part 51, "Changes to Requirements for Environmental Review for Renewal of Nuclear Power Plant Operating Licenses," the NRC identified changes to the requirements for the environmental review of requests for renewal of operating licenses. These changes included allowance for fuel burnup up to the 62 GWD/MTU limit, and was implemented with the revision of 10 CFR Part 51, Subpart A, Appendix B, "Environmental Effect of Renewing the Operating License of a Nuclear Power Plant," Table B-1, "Summary of Findings on NEPA Issues for License Renewal of Nuclear Power Plants."
Under the category of "Uranium Fuel Cycle and Waste Management, Transportation,"
Table B-1 stated that the impacts of transporting spent fuel enriched up to 5 percent and up to 62 GWO/MTU were consistent with the impact values in 10 CFR 51.52(c), Summary Table S-4, "Environmental Impact of Transportation of Spent Fuel and Waste to and from One Light-Water-Cooled Nuclear Power Reactor." In addition, the final rule changed the transportation of fuel and waste from Category 2 to Category 1, therefore not requiring site-specific analysis of in accordance with 51.53(c)(3)(i).
Applicability of Generic Environmental Assessment to DNPS and aCNPS This change to the regulations is further discussed in NUREG-1437, Volume 1, Addendum 1, Section 4, "Summary and Conclusions." NUREG-1437 contains all the environmental impact statements (EISs) for the plants that have been issued renewed facility operating licenses. The station-specific EISs for license renewal are provided as supplements to the NUREG. Supplement 16 is the QCNPS EIS and Supplement 17 is the DNPS EIS.
By letter dated October 28, 2004, the NRC issued Renewed Facility Operating Licenses (FOLs) DPR-19, DPR-25, DPR-29, and DPR-30 for DNPS Units 2 and 3 and QCNPS Units 1 and 2, respectively.
Page 6 of 8

ATTACHMENT 1 Exelon Generation Company, LLC Response to Request for Additional Information The October 28, 2004 NRC issuance letter stated that the technical basis for issuing the renewed licenses is described in NUREG-1796, "Safety Evaluation Report to the License Renewal of Dresden Nuclear Power Stations, Units 2 and 3, and Quad Cities Nuclear Power Stations, Units 1 and 2." The NRC issuance letter also stated that the results of the environmental reviews related to the issuance of the renewed licenses are contained in NUREG-1437, Supplement 16, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 16 Regarding Quad Cities Nuclear Power Station, Units 1 and 2," and Supplement 17, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 17 Regarding Dresden Nuclear Power Station, Units 2 and 3."
NUREG-1437, Supplements 16 and 17, Section 6.1, "The Uranium Fuel Cycle," provides an environmental assessment to support the 62 GWD/MTU limit for DNPS, Units 2 and 3, and QCNPS, Units 1 and 2. Specifically, in Section 6.1, the NRC confirmed that the impact values of transporting spent fuel from with an average rod burnup up to 62 GWD/MTU from DNPS and QCNPS are consistent with the impact values in 10 CFR 51.52(c), Summary Table S-4.* "Environmentallmpact of Transportation of Fuel and Waste to and from One Light-Water-Cooled Nuclear Power Reactor."
Based on the completion of the environmental assessment for DNPS and QCNPS (Le.,
NUREG-1437, Supplements 16 and 17), the license conditions limiting fuel burnup to 60 GWD/MTU rod average are no longer required. As such, these license conditions are obsolete, and deletion of the license conditions is an administrative change.
(3)     Provide the proposed maximum burnup limit for the current reactor operation.
Upon deletion of the license condition, EGC will administratively limit average rod burnup at DNPS and QCNPS to 62 GWD/MTU, in accordance with the applicable NRC-approved LTR for the resident fuel type.
Page 7 of 8

ATTACHMENT 1 Exelon Generation Company, LLC Response to Request for Additional Information NRC RAI Question 3 "With respect to proposed deletion of footnote from DNPS TS 3.4.5, 'RCS Leakage Detection Instrumentation,' please describe work done by EGC to repair the failed DNPS Unit 3 component."
EGC Response The failed DNPS Unit 3 component that resulted in the addition of a TS 3.4.5 footnote in August 2008 was a drywell floor drain sump (DWFDS) Containment Isolation Valve (i.e, valve number 3-2001-105). The valve failed when the process diaphragm to valve stem connection separated, due in part to the orientation of the valve.
During the subsequent DNPS Unit 3 refueling outage (i.e., commencing in November 2008), EGC overhauled the DWFDS inboard and outboard containment isolation valves (Le., the 3-2001-105 and 3-2001-5 valves), as well as the drywell equipment drain sump (DWEDS) inboard and outboard containment isolation valves (Le., the 3-2001-106 and 3-2001-6 valves). This overhaul included replacement of the process diaphragms utilizing an enhanced maintenance procedure that addressed proper valve orientation.
Page 8 of 8

ATTACHMENT 2 Letter from C. P. Patel (U. S. NRC) to T. J. Kovach (Commonwealth Edison Company)
 "Commonwealth Edison Company Topical Report NFSR*0091, 'Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods' (TAC NO. M82731)," dated March 22, 1993

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2OfiI5.OOO1 March 22, 1993 Doc<et Nos. 50-237, 50-249 50-254, 50-265 and 50-373, 50-374 Hr. Thomas J. Kovach Nuclear licensing Manager Commonwealth Edison Company-Suite 300 OPUS West III 1400 OPUS Place Downers Grove, Illinois 60515

==Dear Mr. Kovach:==

==SUBJECT:==
COMMONWEALTH EDISON COMPANY TOPICAL REPORT NFSR-0091, "BENCHMARK OF CASMOjMICROBURN BWR NUCLEAR DESIGN METHODS" (TAC NO. Ma2731)
By letter dated December 31, 1991, Commonwealth Edison Company (CECo) submitted Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods." Supplements 1 and 2 were submitted by letters dated March 24, 1992, and May 22, 1992, respectively. Additional information was submitted by letter dated January 20, 1993.
We have reviewed Toeical Reeort NFSR-0091, SUrelements 1 and 2, and the additional informatlon submltted by CECo and lnd it acceetable for referencin~ for future reload c~c'es. Our findings are dlscussed in detail
     'in the enc osed evaluation. Thls completes our wOrk on TAC No. Ma2731.
Sincerely, Chandu P. Patel, Project Manager Project Directorate 111-2 Division of Reactor Projects - IIIjIVjV Office of Nuclear Reactor Regulation

==Enclosure:==

Evaluation cc w/enclosure:
See next page

Mr. Thomas J. Kovach                      Dresden Nuclear Power Station Commonwealth Edison Company               Unit Nos. 2 and 3 cc:
Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60690 Mr. C. Schroeder P1 ant Manager Dresden Nuclear Power Station 6500 North Dresden Road Morris, Illinois 60450-9765 U. S. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station 6500 North Dresden Road Morris, Illinois 60450-9766 Chairman Board of Supervisors of Grundy County Grundy County Courthouse Morris, Illinois 60450 Regional Administrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Ill1nois 60137 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Robert Neumann Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601

Hr. Thomas J. Kovach                  Quad Cities Nuclear Power Station Commonwealth Edison Company           Unit Nos. 1 and 2 cc:
Mr. Stephen E. Shelton Vice President Iowa-Illinois Gas and Electric Company P. O. Box 4350 Davenport, Iowa 52808 Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60690 Mr. Richard Bax Station Manager Quad Cities Nuclear Power Station 22710 206th Avenue North Cordova, 111 inoi s 61242 Resident Inspector U. S. Nuclear Regulatory Commission 22712 206th Avenue North Cordova, Illinois 61242 Chairman Rock Island County Board of Supervisors 1504 3rd Avenue Rock Island County Office Bldg.
Rock Island, Illinois 61201 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road, Bldg. '4 Glen Ellyn, Illinois 60137 Robert Neumann Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601

Mr. Thomas J. Kovach                  LaSalle County Station Commonwealth Edison Company           Unit Nos. 1 and 2 cc:
Phillip P. Steptoe, Esquire           Robert Cushing Sidley and Austin                     Chief, Public Utilities Division One First National Plaza              Illinois Attorney General's Office Chicago, I11in01s 60603               100 West Randolph Street Chicago, Illinois 60601 Assistant Attorney General 100 West Randolph Street              Michael I. Miller, Esquire Suite 12                              Sidley and Austin Chicago, Illinois 60601               One First National Plaza Chicago, Illin01s 60690 Resident Inspector/LaSalle, NPS U. S. Nuclear Regulatory Commission   Mr. G. Diederich Rural Route No. 1                     LaSalle Station Manager P. O. Box 224                         LaSalle County Station Marseilles, Illinois 61341            Rural Route 1 P. O. Box 220 Chairman                              Marseilles, Illinois 61341 LaSalle County Board of Supervisors LaSalle County Courthouse Ottawa, Illinois 61350 Attorney General 500 South 2nd Street Springfield, Illinois 62701 Chairman Illinois Commerce Commission Leland Building 527 East Capitol Avenue Springfield, Illinois 62706 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, I111nois 62704 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Illinois 60137 Robert Neuman Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601

UNITED STATES UCLEAR REGULATORY CO                   18810 WASHINGTON, O.C. 2llIiIiIHlOO1 EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT NFSR-0091 BENCHMARK OF CASMO/MICROBURN BWR NUCLEAR DESIGN METHODS COMMONWEALTH EDISON COMPANY DRESpEN. UNITS 2 ANP 3. QUAD CITIES. UNITS 1 AND 2. LASALLE. UNITS 1 AND 2 DOCKET NOS, 50-237. 50-249. 50-254. 50-265. 50-373. AND 50-374 1,0 BACKGROUND By letter dated December 31, 1991, Commonwealth Edison Company (CECo) submitted the licensing Topical Report NFSR-0091, "Commonwealth Edison Company Topical - Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods,"
(Reference 1), Supplements 1 and 2 were submitted by letters dated March 24, 1992 (Reference 2) and Hay 22, 1992 (Reference 3), Additional information was submitted by letter dated January 20, 1993 (Reference 4). The topical summarizes nuclear analysis methods to be employed and benchmark results obtained by CECo in support of performing reload design activities for its Boiling Water Reactors (BWRs), the Dresden, Quad Cities, and LaSalle Stations.
These nuclear analysis methods are based upon the CASMO-3G and MICROBURN-B neutronic design computer codes utilized by Siemens Nuclear Power Corporation (SNP).
The March 24, 1992, letter submitted Supplement 1 and attachments which provide detailed comparisons to vendor results for cold and hot eigenvalues, fuel pin and assembly gamma scan data, Traversing Incore Probe data and cold rod worth data for Dresden, Units 2 and 3, and Quad Cities, Unit 1.
Supplement 1 summarizes the application procedures to be used by CECo in performing the neutronic licensing calculations. It also includes comparisons between CECo and SNP calculations for the two CECo BWRs at the Dresden Station.
The May 22, 1992, submittal included Supplement 2 as well as the assessment of a minor error in the MICROBURN computer code. Supplement 2 summarizes the LaSalle, Unit 2, benchmark results obtained by CECo in support of performing reload design activities for BWRs.
The January 20, 1993, submittal provided additional comparisons of CECo versus vendor neutronics licensing analyses results based upon calculations performed for Dresden, Unit 2, Cycle 14.
CECo has two fuel vendors for its three BWR stations: GE is the fuel supplier for the Quad Cities, and LaSalle Stations, and SNP is the fuel supplier for the Dresden Station. Due to the differences between the GE and SNP neutronic

methodologies with respect to the Critical Power Correlation and the associated uncertaintles t Co does not intend to use the Siemens methods described in this report for analysis w lC etermine the crlfica power ratio impact for ynits not fyeled by SHP (cyrrentlv the Quad Cities, and LaSalle Stations). A similar topical using the GE methodology was approved in February 1992.
NFSR-0091 summarizes the nuclear analysis methods used by CECo in support of reload analysis for its BWR reactors and the benchmark data used to demonstrate CECots capability to independently perform the steady state neutronic analysis needed for the steady-state licensing, operation t testing and surveillance of a BWR reload cycle. The CASMO-3G/HICROBURN methodology which CECo uses has previously been approved.
2.0 EVALUATION The nuclear analysis methods used by CECo are based on the MICBURN-3/CASMQ-3G/MICROBURN-B neutronic computer codes used by SNP. These codes were obtained by CECo from SHP and installed without modification on a PRIME computer system equivalent to the SNP PRIME system. The CECo engineers underwent extensive training at the Siemens facilities. This training included the performance of the full scope of neutronic calculations required for a reload design, and training in the acceptability and limitations of the computer programs for calculating neutronic parameters. The methodology was used for the benchmark analysis t which was performed to demonstrate CECo's capability to independently perform the steady-state neutronic analysis portions of the reload design process.
Since the methodology consisted of using computer codes which were previously approved, the review focused on the benchmarking data. For the benchmarking of the critical eigenvalues, data was used from Dresden t Unit 2, Cycles 6 through 12, Dresden, Unit 3, Cycles 7 through 12, Quads Cities t Unit I, Cycles 1 through 5 and LaSalle, Unit 2, Cycles 1 through 4. This database included fuel designs from current and past fuel product lines from both Siemens and General Electric. This includes 7X7, axa, and 9X9 fuel pin arrays, various water rod configurations and axially dependent gadolinia distributions.
Hot Critical Eigenvalues Figures showing hot critical eigenvalues as a function of cycle exposure were presented. Also given were the mean and standard deviation of the hot critical eigenvalues for each cycle, as well as the overall mean and standard deviation for each reactor and the overall statistics of the entire database.
The overall mean and standard deviation are 1.0056 and 0.0022 ~K for the Dresden, Units 2 and 3 t and the Quads Cities, Unit I, data. The LaSalle, Unit 2, data had a mean K-effective of 1.0017 and a standard deviation of 0.0013~K. Hot critical eigenvalues should be consistent and predictable as a function of cycle exposure so that adequate projections of the critical eigenvalues can be made for the upcoming cycles. The unit to unit standard deviations are very comparable and are consistent with the overall and cycle

specific standard deviations. Comparisons of the Siemens and CECo results for the hot critical eigenvalues for Dresden, Unit 2, Cycles 10 and 11 and for Dresden, Unit 3, Cycles 10 and 11 showed good agreement. The mean and standard deviation for these cycles are as follows:
CECo Data                    Siemens Data Unit/Cycle        Mean          SD              Mean         SO 02 C10             1.0053       0.0008          1.0039       0.0005 02 Cll            1.0046        0.0014          1.0037       0.0010 03 CI0            1.0049        0.0016          1.0045       0.0014 03 C11             1.0070       0.0015          1.0065       0.0014 Cold Critical Eigenvalues The cold eigenvalue predictions are presented in tabular form. As with the hot critical eigenvalues, the cold critical eigenvalues must also be consistent and predictable as a function of exposure because they are required to calculate core subcriticality. The values are consistent and predictable as a function of the plant. The overall mean and standard deviation for the entire database are 1.0055 and 0.0031 ~K. For the four cycles in which there are predictions of cold values by both CECo and Siemens the average eigenvalues are 1.0078 and 1.0076 for CECo and Seimens.
TIP Results Measured and calculated Traversing Incore Probe (TIP) data have been compared and are summarized. The TIP standard deviations were calculated over the entire axial length of the core, which is 24 nodes. TIP standard deviations are within the current design criteria of 10 percent nodally and 6 percent radially for nearly all the data. For the Dresden, Unit 2, Cycle 10 through 12 and Dresden, Unit 3, Cycles 9 through 11 data, the CECo average of the standard deviations was 5.5% and 7.48% for radial and nodal, respectively.
The SNP values were 5.61% and 7.48% for radial and nodal, respectively. The LaSalle, Unit 2, predicted versus measured TIP data had an overall standard deviation of 3.59% and 6.83% for radial and nodal, respectively. If, during operation. the standard deviations approach the 10% or 6% values, they are beyond what has been historically seen and will be investigated.
Assembly Gamma Scans Predictions of measured gamma scan data demonstrate the ability to predict power distribution in a BWR core. Assembly gamma scan measurements were performed at the end of Cycles 2, 4, and 5 of Quad Cities, Unit 1. The results of these assembly gamma scans show the nodal standard deviation varying from 3.69% to 5.3% and the radial standard deviation varying from 1.75% to 3.29%. These values are slightly better than deviations reported for a similar database. Assembly integrated power comparisons for Cycles 2 and 4 were presented. The SNP assembly integrated standard deviations for Cycles 2 and 4 are 1.98% and 2%, respectively, while the CECo values are 1.43% and

1.75%. Thus showing that CECo does at least as good a job of predicting power distributions as SNP.
Pin Gamma Scans The ability to predict pin-by-pin fuel assembly lattice power distributions, i.e., local peaking factors, is an indicator of the adequacy of the lattice physics methodology. This ability is assessed through comparison of predicted versus measured fuel pin gamma scan data. Pin gamma scan measurements were performed at Quad Cities, Unit I, at the end of Cycles 2, 3, and 4. Seven assemblies were used. They have an average enrichment of between 2.12 and 2.62 w/o U-235, include 7X7 and aX8 lattice arrays, GD~03 loadings from 1.5 to 3.0 w/o per pin, and two to four GD2~ pins per assemblY. The statistical results obtained from the differences in the predicted and measured data are presented. Supplement 1 presented the comparison of CECa and SNP data. The CECo overall standard deviation was 2.65% and the SNP overall standard deviation was 2.71%. The peak power rods had an even lower standard deviation of 1.59%. This demonstrates the ability of CECa to predict local peaking factors.
Neutronic Licensing Calcy1ations The application procedures which CECo will use in the evaluation of cycle-specific neutronic licensing events are the same as those used by Siemens Nuclear Power Corporation, which were previously reviewed and approved by the NRC. Comparisons of CECa's results to SNP's results were made for various abnormal neutronic licensing events, including the calculation of the critical power ratio for these events. These comparisons were made for Dresden, Unit 3, Cycle 13 and Dresden, Unit 2, Cycle 14. The following cycle-specific neutronic licensing analyses were compared.
: 1. Shutdown Margin
: 2. Standby liquid Control System
: 3. Fuel Loading Error - Misoriented Assembly
: 4. Fuel Loading Error - Misloaded Assembly
: 5. Control Rod Drop Accident
: 6. Control Rod Withdrawal Error
: 7. Loss of Feedwater Heating In all cases the difference between the CECa and SNP results was very small.
In almost all cases the CECa results were more conservative. Because the same neutronic methods are used together with application procedures which ensure compliance with the licensing bases and use of these procedures is benchmarked against the SNP results, the same uncertainties will be applied.
MICROBURN Error Assessment After completion of the Topical Report (Reference 1), CECa discovered a minor error in the MICROBURN computer code obtained from SNP. The impact of this error, which involved the manipulation of the samarium number density array to

reflect fuel shuffling, occurs near the beginning of cycle (BOC) only. This error was assessed by SNP and was shown to affect hot and cold eigenvalues by approximately 0.0005 ~ or less. Because the impact of comparisons to plant data is very slight and well within the uncertainty of the calculations for hot and cold eigenvalues, and there is no impact on CECo comparisons to SNP results (since both CECo and SNP results are equally affected), CECo did not repeat the entire benchmark. CECo did perform sample cases to demonstrate the small impact. Comparisons of hot critical eigenvalues results, cold eigenvalue results, and Traversing Incore Probe (TIP) calculated versus measured results for two cycles of Dresden, Unit 2, were provided. The comparisons provided the results that incorporate correction of the error and results that contain the error. The differences are negligible.

==3.0 CONCLUSION==

has demonstrated through the extensive Principal Contributor:  M. Chatterton Dated: March 22, 1993

==4.0 REFERENCES==
: 1. P.L. Piet (CECo) letter to T.E. Murley (NRC), submitting CECo Topical Report NFSR-0091, dated December 31, 1991.
: 2. P.L. Piet (CECo) letter to T.E. Murley (NRC), submitting CECo Topical Report NFSR-0091 Supplement I, dated March 24, 1992.
: 3. P.L. Plet (CECo) letter to T.E. Murley (NRC), submitting CECo Topical Report NFSR-0091 Supplement 2, dated May 22, 1992.
: 4. P.L. Piet (CECo) letter to T.E. Murley (NRC), submitting additional information on CECo Topical Report NFSR-0091, dated January 20, 1993.}}