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Category:TEXT-SAFETY REPORT
MONTHYEARML20217L8831999-10-21021 October 1999 Safety Evaluation Supporting Proposed Alternatives to Code Requirements Described in RR-V17 & RR-V18 ML20217L9371999-10-20020 October 1999 Safety Evaluation Supporting Licensee Proposed Alternative from Certain Requirements of ASME Code,Section XI for First 10-Yr Interval Request for Relief for Containment Inservice Insp Program ML20217K3301999-10-19019 October 1999 Safety Evaluation Supporting Amend 195 to License DPR-61 ML20217J0721999-10-18018 October 1999 Safety Evaluation of Topical Rept EMF-2158(P),Rev 0, Seimens Power Corp Methodology for Boiling Water Reactors, Evaluation & Validation of Casmo-4/Microburn-B2. Rept Acceptable for Licensing Evaluations of BWR Neutronics ML20217H8991999-10-18018 October 1999 SER Approving Licensee Requests for Relief NDE-R001 (Part a & B),NDE-R027,NDE-028,NDE-R029,NDE-R030,NDE-R032 & NDE-R035. Relief Request NDE-036,denied & Relief Request NDE-R-034, Deemed Unnecessary ML20217J4791999-10-18018 October 1999 SER Approving Exemption from Certain Requirements of 10CFR73 for Zion Nuclear Power Station,Units 1 & 2.NRC Concluded That Proposed Alternative Measures for Protection Against Radiological Sabotage Meets Requirements of 10CFR73.55 ML20217K9441999-10-15015 October 1999 SER Accepting Util Alternative Proposed Relief Request RR-ENG-2-4 for Second 10-year ISI Interval at Stp,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(i) ML20217K9151999-10-15015 October 1999 SER Authorizing Util Relief Request RR-ENG-2-3 for Second 10-year ISI Interval of Stp,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(i) ML20217G0931999-10-15015 October 1999 Safety Evaluation of Topical Rept BAW-10179P,Rev 3, Safety Criteria & Methodology for Acceptable Cycle Reload Analysis. Rev 3 Found Acceptable & Accurately Include Conditions & Limitations for Applicability of References ML20217K0651999-10-15015 October 1999 Safety Evaluation of Topical Rept BAW-10193P, RELAPS5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized-Water Reactors. Rept Acceptable for Referencing in Licensing Applications ML20217G0191999-10-15015 October 1999 Safety Evaluation Concluding That Licensee Followed Analytical Methods Provided in GL 90-05.Grants Relief Until Next Refueling Outage,Scheduled to Start on 991001.Temporary non-Code Repair Must Then Be Replaced with Code Repair ML20217G2161999-10-15015 October 1999 Errata Pages 2 & 3 for Safety Evaluation Supporting Amend 168 Issued to FOL DPR-63 Issued on 990921.New Pages Change Description of Flow Control Trip Ref Cards to Be Consistent with Application for Amend ML20217K9931999-10-14014 October 1999 Safety Evaluation Supporting Amend 234 to License DPR-56 ML20217H8501999-10-14014 October 1999 Safety Evaluation Supporting Amend 197 to License DPR-64 ML20217G9961999-10-14014 October 1999 SER Accepting First 10-year Interval Inservice Insp Requests for Relief for Plant,Units 1 & ML20217G2041999-10-13013 October 1999 Safety Evaluation Supporting Amend 179 to License DPR-28 ML20217J1101999-10-13013 October 1999 Safety Evaluation of Topical Rept TR-108727, BWRVIP Vessel & Internals Project,Bwr Lower Plenum Insp & Flaw Evaluation Guideline (BWRVIP-47). Rept Will Provide Acceptable Level of Quality for Exam of safety-related Components ML20217D3061999-10-13013 October 1999 SER Accepting Licensee Proposed Changes to Edwin I Hatch Nuclear Plant Emergency Classification Scheme to Add Emergency Action Levels Related to Operation of Independent Spent Fuel Storage Installation ML20217C9121999-10-12012 October 1999 SER Input Authorizing Licensee Proposed Request to Modify Definition of Core Alteration in Section 1.0 of TS & Update Sections 3/4.1,3.4.3 & 3/4.9 to Reflect Proposed Definition Change ML20217C1311999-10-0808 October 1999 Safety Evaluation Supporting Amend 153 to License DPR-3 ML20217B5401999-10-0606 October 1999 Safety Evaluation Supporting Amend 193 to License DPR-40 ML20217B1641999-10-0505 October 1999 Safety Evaluation of Topical Rept BAW-10228P. Science. Rept Acceptable for Licensing Applications,Subject to Listed Conditions in Accordance with Fcf Agreement (Reference 4) ML20212M2141999-10-0505 October 1999 Safety Evaluation Concluding That Topical Rept EMF-2158(P), Rev 0,acceptable for Licensing Evaluations of BWR Neutronics Designs & Applications,As Per SPC Agreement (Ref 9) Subj to Stated Conditions ML20217B4331999-10-0505 October 1999 Safety Evaluation Supporting Amend 233 to License DPR-56 ML20212L0881999-10-0404 October 1999 SER Accepting Licensee Requests for Relief 98-012 to 98-018 Related to Implementation of Subsections IWE & Iwl of ASME Section XI for Containment Insp for Crystal River Unit 3 ML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20212J9251999-10-0101 October 1999 Safety Evaluation Accepting Licensee Relief Request IWE-3 for Second 10-year ISI for Plant ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c ML20212J8631999-10-0101 October 1999 Safety Evaluation Supporting Licensee Proposed Alternatives to Provide Reasonable Assurance of Structural Integrity of Subject Welds & Provide Acceptable Level of Quality & Safety.Relief Granted Per 10CFR50.55a(g)(6)(i) ML20212J2011999-09-30030 September 1999 Safety Evaluation Supporting Transfer of Dl Ownership Interest in Pnpp to Ceico ML20212K9781999-09-30030 September 1999 Safety Evaluation Accepting USI A-46 Implementation Program ML20212J1301999-09-30030 September 1999 Safety Evaluation Concluding That Topical Rept WCAP-12472-P-A,Addendum 1, Beacon-Core Monitoring & Operations Support System, Acceptable for Licensing Applications Subj to Pertinent Restrictions ML20212J9141999-09-29029 September 1999 Safety Evaluation of Topical Rept TR-108724, BWRVIP Vessel & Internals Project,Vessel Id Attachment Weld Insp & Flow Evaluation Guidelines (BWRVIP-48) ML20212J9661999-09-29029 September 1999 Safety Evaluation of Topical Rept TR-107285, BWRVIP Vessel & Internals Project,Bwr Top Guide Insp & Flaw Evaluation Guidelines (BWRVIP-26), Dated Dec 1996.Rept Acceptable ML20212F7671999-09-24024 September 1999 SER Granting Relief Request C-4 Pursuant to 10CFR50.55a(g)(6)(i) for Unit 2,during First 10-year ISI Interval & Relief Requests B-15,B-16 & B-17 Pursuant to 10CFR50.55a(g)(6)(i) ML20216H7091999-09-24024 September 1999 Safety Evaluation Supporting Amends 229 & 232 to Licenses DPR-44 & DPR-56,respectively ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20212F5641999-09-23023 September 1999 SER Concluding That All of ampacity-related Concerns Have Been Resolved & Licensee Provided Adequate Technical Basis to Assure That All of Thermo-Lag Fire Barrier Encl Cables Operating within Acceptable Ampacity Limits ML20212E6341999-09-23023 September 1999 Suppl to SE Resolving Error in Original 990802 Se,Clarifying Fact That Licensee Has Not Committed to Retain Those Specific Compensatory Measures That Were Applied to one-time Extension ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 ML20212H2381999-09-22022 September 1999 Safety Evaluation Supporting Amend 228 to License DPR-49 ML20212E6911999-09-21021 September 1999 Safety Evaluation Supporting Proposed EALs Changes for Plant Unit 3.Changes Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20212J0501999-09-21021 September 1999 Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1 ML20212D3831999-09-20020 September 1999 Safety Evaluation Supporting Proposed Rev to Withdrawal Schedule for First & Third Surveillance Capsules for BFN-3 RPV ML20212D1911999-09-20020 September 1999 SER Accepting Exemption from Certain Requirements of 10CFR50,App A,General Design Criterion 57 Closed System Isolation Valves for McGuire Nuclear Station,Units 1 & 2 ML20216F9831999-09-20020 September 1999 Safety Evaluation Supporting Amend 11 to License R-115 ML20216H9901999-09-20020 September 1999 Proposed Final Rept Impep Review of South Carolina Agree- Ment State Program 990712-16 ML20212D4471999-09-20020 September 1999 Safety Evaluation Supporting Amend 31 to License R-103 ML20212C2551999-09-17017 September 1999 Safety Evaluation Supporting Amend 175 to License DPR-28 1999-09-30
[Table view]Some use of "" in your query was not closed by a matching "". Category:TOPICAL REPORT EVALUATION
MONTHYEARML20217J0721999-10-18018 October 1999 Safety Evaluation of Topical Rept EMF-2158(P),Rev 0, Seimens Power Corp Methodology for Boiling Water Reactors, Evaluation & Validation of Casmo-4/Microburn-B2. Rept Acceptable for Licensing Evaluations of BWR Neutronics ML20217K0651999-10-15015 October 1999 Safety Evaluation of Topical Rept BAW-10193P, RELAPS5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized-Water Reactors. Rept Acceptable for Referencing in Licensing Applications ML20217G0931999-10-15015 October 1999 Safety Evaluation of Topical Rept BAW-10179P,Rev 3, Safety Criteria & Methodology for Acceptable Cycle Reload Analysis. Rev 3 Found Acceptable & Accurately Include Conditions & Limitations for Applicability of References ML20217J1101999-10-13013 October 1999 Safety Evaluation of Topical Rept TR-108727, BWRVIP Vessel & Internals Project,Bwr Lower Plenum Insp & Flaw Evaluation Guideline (BWRVIP-47). Rept Will Provide Acceptable Level of Quality for Exam of safety-related Components ML20212M2141999-10-0505 October 1999 Safety Evaluation Concluding That Topical Rept EMF-2158(P), Rev 0,acceptable for Licensing Evaluations of BWR Neutronics Designs & Applications,As Per SPC Agreement (Ref 9) Subj to Stated Conditions ML20217B1641999-10-0505 October 1999 Safety Evaluation of Topical Rept BAW-10228P. Science. Rept Acceptable for Licensing Applications,Subject to Listed Conditions in Accordance with Fcf Agreement (Reference 4) ML20212J1301999-09-30030 September 1999 Safety Evaluation Concluding That Topical Rept WCAP-12472-P-A,Addendum 1, Beacon-Core Monitoring & Operations Support System, Acceptable for Licensing Applications Subj to Pertinent Restrictions ML20212J9661999-09-29029 September 1999 Safety Evaluation of Topical Rept TR-107285, BWRVIP Vessel & Internals Project,Bwr Top Guide Insp & Flaw Evaluation Guidelines (BWRVIP-26), Dated Dec 1996.Rept Acceptable ML20212J9141999-09-29029 September 1999 Safety Evaluation of Topical Rept TR-108724, BWRVIP Vessel & Internals Project,Vessel Id Attachment Weld Insp & Flow Evaluation Guidelines (BWRVIP-48) ML20216F4771999-09-16016 September 1999 Safety Evaluation of Topical Rept TR-108823, BWR Vessel & Internals Project,Bwr Shroud Support Insp & Flaw Evaluation Guidelines (BWRVIP-38).Requests That BWRVIP Be Reviewed & Resolve Issues & Incorporate Concerns in Revised BWRVIP-38 ML20211Q3171999-09-0909 September 1999 Safety Evaluation Accepting BWROG Rept, Prediction of Onset of Fission Gas Release from Fuel in Generic BWR, Dtd July 1996 ML20212B2501999-09-0202 September 1999 Safety Evaluation of TR WCAP-14696, WOG Core Damage Assessment Guidance, Rev 1.Rept Acceptable ML20211K5711999-09-0101 September 1999 FSER by NRR Re BWR Vessel & Internals Project,Instrument Penetration Insp & Flaw Evaluation Guidelines (BWRVIP-49), for Compliance with License Renewal Rule (10CFR54).TR Acceptable ML20209H9571999-07-15015 July 1999 Safety Evaluation Accepting EPRI Rept TR-105696-R1, BWR Vessel & Intervals Project:Reactor Pressure Vessel & Internals Examination Guidelines (BWRVIP-03) Rev 1, ML20209J1131999-07-15015 July 1999 Safety Evaluation of Topical Rept NSPNAD-8102,rev 7 Reload Safety Evaluation Methods for Application to PI Units. Rept Acceptable for Referencing in Prairie Island Licensing Actions ML20209F1571999-07-0808 July 1999 Safety Evaluation of Topical Rept TR-108695, BWR Vessel & Internals Project,Instrument Penetration Inspection & Flaw Evaluation Guidelines (BWRVIP-49). Rept Acceptable.Rept Demonstrates That Aging Effects of Rv Components Adequate ML20209F1261999-07-0808 July 1999 Safety Evaluation of Topical Rept TR-108709, BWRVIP Vessel & Internals Project Low Alloy Steel Vessel Materials in BWR Environment (BWRVIP-60). Rept Acceptable for Assessment of SCC Growth in BWR Low Alloy Steel Pressure Vessels ML20209D9651999-07-0707 July 1999 Safety Evaluation of Topical Rept WCAP-14750, RCS Flow Verification Using Elbow Taps at Wesstinghouse 3-Loop Pressurized Water Reactors. Changes to TS Bases Acceptable ML20196J3791999-06-30030 June 1999 Safety Evaluation of TR WCAP-14750, RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop Pwrs. Rept Acceptable ML20196G6321999-06-15015 June 1999 Safety Evaluation of Topical Rept EMF-2087(P),Rev 0, SEM/PWR-98:ECCS Evaluation Model for PWR LBLOCA Application, Rept Acceptable ML20195J2681999-06-14014 June 1999 Safety Evaluation of Topical Rept TR-108726, BWR Vessel & Internals Project,Lpci Coupling Insp & Flaw Evaluation Guidelines (BWRVIP-42). Rept Acceptable for Insp of safety- Related LPCI Coupling Assemblies,Except Where Staff Differ ML20207H1521999-06-0909 June 1999 Safety Evaluation of Topical Rept TR-108708, BWRVIP Vessel & Internals Project,Underwater Weld Repair of Nickel Alloy Reactor Vessel Internals (BWRVIP-44), Sept,1997.Rept Acceptable ML20207G4971999-06-0808 June 1999 Safety Evaluation Re Mods to TR CENPD-266-P-A, Application of Dit Cross Section Library Based on ENDF/B-VI. Rept Acceptable ML20195D3061999-06-0202 June 1999 Safety Evaluation of TR SCE-9801-P, Reload Analysis Methodology for San Onofre Nuclear Generating Station,Units 2 & 3. Rept Acceptable ML20207C7321999-05-26026 May 1999 Safety Evaluation of Topical Rept BAW-2248, Demonstration of Mgt of Aging Effects for Reactor Vessel Internals. Rept Provides Individual B&W Nuclear Power Plant Utility Owner with Technical Details for for License Application Renewal ML20195J2271999-05-25025 May 1999 Safety Evaluation of CE Owner Group Topical Rept CE NPSD-951 Rev 1,justifying, Reactor Trip Circuit Breakers Surveillance Frequency Extension ML20207A6251999-05-21021 May 1999 Safety Evaluation of TR WCAP-14449(P), Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection. Rept Acceptable ML20207B0241999-05-18018 May 1999 Safety Evaluation of Topical Rept TR-107285, BWR Vessel & Intervals Project,Bwr Top Guide Insp & Flaw Evaluation Guidelines (BWRVIP-26), Dtd December 1996.Rept Acceptable ML20206K7691999-05-0808 May 1999 Topical Rept Evaluation of CENPD-389-P, 10x10 Svea Fuel Critical Power Experiments & CPR Correlations:SVEA-96+. Rept Acceptable ML20206D5441999-04-28028 April 1999 Safety Evaluation of Topical Rept TR-107284, BWRVIP Vessel & Internals Project,Bwr Core Plate Insp & Flaw Evaluation Guideline (BWRVIP-25). Rept Acceptable for Insp & Flaw Evaluation of Subject safety-related Core Interal ML20206D4951999-04-26026 April 1999 Safety Evaluation Supporting Topical Rept BAW-2251, Demonstration of Mgt of Aging Effects for Rv ML20205L9441999-04-0808 April 1999 Safety Evaluation of Topical Rept CENPD-289-P, Use of Inert Replacement Rods in Abb C-E Fuel Assemblies. Rept Acceptable ML20205L9671999-04-0707 April 1999 Safety Evaluation of Topical Rept TR-108727, BWRVIP Vessel & Internals Project,Bwr Lower Plenum Insp & Flaw Evaluation Guideline (BWRVIP-47). Rept Found Acceptable Except Where Staff Conclusions Differ from BWRVIP ML20205F0251999-03-21021 March 1999 Safety Evaluation of Topical Rept TR-108727, BWRVIP Vessel & Internals Project Vessel Id Attachmant Weld Insp & Flaw Evaluation Guidelines. Rept Acceptable ML20207E3821999-03-0202 March 1999 Topical Rept Evaluation of SL-5159(P), Methodology & Verification of Gapp Program for Analysis of Piping Systems with E-Bar Supports. Staff Finds Topical Rept Acceptable for Referencing in Licensing Applications ML20203H7381999-02-18018 February 1999 Safety Evaluation of Topical Rept BAW-2328, Blended U Lead Test Assembly Design Rept. Rept Acceptable Subj to Listed Conditions ML20203A2581999-02-0505 February 1999 Safety Evaluation of TR DPC-NE-3002-A,Rev 2, UFSAR Chapter 15 Sys Transient Analysis Methodology. Rept Acceptable. Staff Requests Duke Energy Corp to Publish Accepted Version of TR within 3 Months of Receipt of SE ML20203C1841999-02-0303 February 1999 Safety Evaluation of Topical Rept NEDC-32721P, Application Methodology for General Electric Stacked Disk ECCS Suction Strainer, Part 1.Concluded That Use of GE Hydraulics Design Method Acceptable for All Plants,With One Noted Exception ML20203A7461999-02-0202 February 1999 Safety Evaluation of Siemens Power Corp Topical Rept EMF-92-116(P), Generic Mechanical Design Criteria for PWR Fuel Design. Rept Acceptable ML20199L6651999-01-25025 January 1999 Topical Rept/Ser of BAW-10186P, Extended Burnup Evaluation. Rept Acceptable.Staff Finds That Improved Methodology Adequate & Acceptable for Fuel Reload Licensing Applications Subject to Listed Conditions ML20198G1851998-12-15015 December 1998 Safety Evaluation for Topical Rept WCAP-14572,rev 1, WOG Application of Risk-Informed Methods to Piping ISI Topical Rept ML20196A4191998-11-19019 November 1998 Safety Evaluation Accepting QA TR CE-1-A,Rev 66 Re Changes in Independent & Onsite Review Organization by Creating NSRB ML20195F7941998-11-17017 November 1998 Safety Evaluation of EPRI TR-106708 & TR-106893.Repts Found to Be Acceptable for Replacement &/Or Repair of BWRVIP Vessel & Internals Project,Internal Core Spray Components ML20195F7041998-11-17017 November 1998 Safety Evaluation Accepting Topical Rept NEDC-24154P, Supplement 1,for Referencing in Licensing Applications to Extent Specified & Under Limitations Delineated in Rept ML20195C6721998-11-10010 November 1998 Safety Evaluation of Topical Rept WCAP-15029, Westinghouse Methodology for Evaluating Acceptability of Baffle-Former- Bolting Distribution Under Faulted Load Conditions ML20155G3901998-11-0505 November 1998 Safety Evaluation of TR GENE-770-06-2, Addendum to Bases for Changes to Surveillance Test Intervals & Allowed Out-of- Svc Times for Selected Instrumentation Tss. Rept Acceptable ML20155G3031998-11-0505 November 1998 Safety Evaluation of TRs NEDC-30844, BWR Owners Group Response to NRC GL 83-28, & NEDC-30851P, TSs Improvement Analysis for BWR Rps. Rept Acceptable ML20155B6121998-10-28028 October 1998 Safety Evaluation of TR SNCH-9501, BWR Steady State & Transient Analysis Methods Benchmarking Topical Rept. Rept Acceptable ML20154F0711998-10-0606 October 1998 SE of TR WCAP-14036,Rev 1, Elimination of Periodic Protection Channel Response Time Tests. Rept Acceptable ML20155G2611998-10-0505 October 1998 Corrective Page 9 of Safety Evaluation of TR WCAP-14036,Rev 1, Elimination of Periodic Protection Channel Response Time Tests. Typos Made in Original Rept Re Components Covered by Solid State Protection Sys Were Corrected 1999-09-09
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Attachment SAFETY FVALUATION BY THE OFFICE _OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT CENPD-199-P REVISION 1-P-A. SUPPLEMENT 2-P "CE SETPOINT METHODOLOGY" ABB COMBUSTION ENGINEERING INCORPO%IED
- 1. INTRODUCTION In a letter of September 18,1997 (Ref.1). ABB Combustion Engineering.
Incorporated (ABB CE) submitted the topical report CENPD-199-P. Revision '
1-P-A, Supplement 2-P. "CE Setpoint Methodology" (Ref. 2) for U.S. Nuclear Regulatory Commission (NRC) review and approval. This supplement describes ,
several proposed modifications to the NRC-approved ABB CE setpoint metnodology described in CENPD-199-P. Revision 1-P-A (Ref. 3). These proposed modifications were also discussed on May 15. 1997, when representatives of the Florida Power and Light Company, licensee for the St. Lucie Nuclear Piant, and ABB-CE. met with members of the NRC staff at NRC Headquarters in Rockville.
Maryland.
CENPD-199-P Revision 1-P-A. describes the methodology used by ABB-CE to calculate lirait'ng safety system settings (LSSS) for the local power density and thermal margin trip systems and limiting conditions for operation (LCO) to assure that the specified acceptable fuel design limits are not exceeded during the design basis anticipated operational occurrences (A00s). The Combustion Engineering nuclear steam supply systems (NSSS) for which the methodology is applicable are those incorporating the analog reactor protection system and licensed after 1971.
- 2.
SUMMARY
OF TOPICAL REPORT Supplement 2-P to topical report CENPD-199-P, Revision 1-P-A. describes the following modifications and extensions to the ABB-CE setpoint methodology previously approved by the NRC.
- 1) an alternate method (Xenon Swing) for determining axial xenon concentrations used in the axial shape analysis 9803130100 980305 PDR ADOCKOS2Og3
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- 2) elimination of total planar radial peaking factor. F,,, monitoring for ABB-CE plarits having analog protection systems
- 3) use of previously approved three-dimensional (3-D) neutronics and the, mal-1,ydraulics codes for use in setpoint analyses as an alternative to one-dimensional /two-dimensional (1-D/2-D) synthesis methods
- 4) application of NRC approved ABB-CE thermal-hydraulic departure from nucleate boiling (DNB) andysis methodology to mixed cores (i.e .
cores coritaining fuel provided by two different vendors) containing similar fuel having non-mixing grid designs Appendix A to the Supplement describes the application of ABB-CE thermal-hydraulic DNB analysis methods to transition cores containing a mix of Siemens and A03-CE fuel designs. As an example, the methodology is applied to fuel
. designs for St. Lucie Unit 1.
- 3. TECHNICAL EVALUATION OF REPORT Axial Power Shane Generatio_n The NRC-approved methodology for generating axial power shapes for setpoint detendnations is described in CENPD-199-P. Revision 1-P-A. and is known as the Free Xenon Oscillation methodology. The methodology relies on creating axial power distributions driven by xenon oscillations. An equilibrium axial power distribution. at full power with all control reds withdrawn, is perturbed so that a divergent axial power oscillation is initiated. The oscillation is allowed to diverge without any control action for an appropriate period of time. Although this methodology produces axial shapes which span a broad axial shape index (ASI) range, the Doppler feedback has to be artificially reduced so that it is essentially eliminated from calculations near beginning of cycle (BOC) and up to middle of cycle (MOC) in order to generate axial shapes over a wide range of ASIS. Some of the shapes are produced by conditions which are significantly outside the limiting condition for operation (LCO) or trip boundaries on ASI, and which cannot actually be present at these times in cycle.
For these reasons. ABB-CE proposes to modify the current Free Xenon Oscillation axial shape methodology by an alternate approach termed he Xenon Swing methodology which is described in Supplement 2-P. As an alten stive to the artificial reduction in Doppler feedback and arbitrary perturbation used in the Free Xenon Oscillation methodology, the Xenon Swing methodology uses a
3 .
maneuver to initiate the xenon transient. Thus, the generation of axial power shapes using the proposed Xenon Swing methodology is more realistic than the Free Xenon Oscillation methodology since no artificial adjustments are made to the Doppler feedback during the transient. The resulting axial power shapes are more severe than those actually expected in operation since the power / control rod maneuver that initiates the Xenon Swing is selected to maximize the perturbation in the xenon distribution. Near BOC. when the core is axially stable. the Xenon Swing methodology covers the entire range of ASIS which can realistically be obtained by a severe xenon transient. At high power levels near end of cycle (EOC). ASIS outside the trip limits can be generated.
Comparisons of flyspeck distributions obtained from using the Xenon Swing to those obtained from the Free Xenon Oscillation methodology were reviewed by the staff. Based on these comparisons, and on the more realistic axial shapes generated, we conclude that the Xenon Swing methodology is an appropriately conservative, alternate method of generating axial shapes for the setpoint analysis.
Elimination of F,, Monitorina ABB-CE plants with analog protection systems using the current setpoint methodology (Ref. 3) traditionally have had technical specifications (TS) which monitor the total planar radial peaking factor. F,,. as well as the total integrated radial peaking factor. F,. ABB-CE has proposed to eliminate monitoring of F,, when the 1-D/2-D synthesis methodology is used. This would be done by calculating the upper bound of the ratio F,,/F, using an NRC-approved 3-D neutronics code such as ROCS (Ref. 4). F,,in the linear heat rate and DNBR setpoint analyses would then be obtained from the product of F, and the upper bound value of the ratio F,,/Fr . In this proposed procedure. F, would continue to be monitored by TS as is currently done. The elimination of F,, monitoring simplifies the effort to confirm compliance to the TS during power distribution surveillance.
When F,, is modeled implicitly as a multiple of F,, there is no need for the TS tradeoff curve to determine the power reduction required if the measured F,,
exceeds the analysis value. The setpoint analysis will determine allowed power versus F,, for the linear heat rate LSSS and LCO analyses in the same way that previous F,, tradeoff curve confirmations were done.
The rationale for eliminating F,, monitoring rests on the fact that F,, and F, are not independent. since F, is the integral of F,, (weighted by the core average axial power distribution) over the height of the core. Indeed. ABB-CE l
4 a
has presented data showing that F,, and F, are correlated. Therefore, we i conclude that the proposed modification to the setpoint methodology is an I acceptable alternative to monitoring of F,,, and the elimination of F,,
monitoring and the F,, tradeoff curve is acceptable. This procedure would be used when the 1-D/2-D synthesis methodology is used. The elimination of F,,
monitoring as a result of the use of 3-D calculations as an alternative to the 1-D/2-D synthesis method is discussed below.
Use of 3-D Physics and Detailed Thermal-Hydraulics The current sctpoint analysis process for ABB-CE NSSS plants with analog ;
protection systems involves a synthesis of axial power distributions generated I by HERMITE (Ref. 5) and radial peaking factors and pin-by-pin power distributicns based on ROCS (Ref. 4) calculations. Thermal-hydraulics calculations are performed using the CETOP-D code (Ref. 6) which is benchmarked to the detailed TORC code (Ref. 7).
I The use of a 3-D calculation not only models core characteristics more realistically it also allows elimination of certain calculations which are performed in the 1-D/2-9 synthesis. The calculation of F,,/F, becomes unnecessary because the 3-D code calculates the 3-D peaking factor. Fo.
directly. Therefore, the need for F,, monitoring is eliminated since limits can be imposed directly on Fa and since the relationship between F, and F,, is implicit in the 3-D calculation. Likewise, the relationship between core average axial shape index and peripheral axial shape index discussed in Ref. 3 would be calculated directly by the 3-D analysis rather than off-line calculations, thereby eliminating the need for adjunct 3-D calculations or bounding assumptions.
The influence of control rod bank position on radial peaking factor is also calculated directly, eliminating the need for adjunct 3-D calculations and bounding bank distortion factors. The 3-D code calculates the hot channel axial power distribution directly, eliminating the need for the pseudo-hot channel.
The hot and average channel axial power distributions from a 3-D neutronics code such as ROCS can be processed for use in the calculation of DNB by the simplified CETOP-D code or more detailed 3-D power distributions can be processed by the detailed TORC thermal-hydraulic code.
Therefore, since a 3-D neutronics code provides a more accurate methodology for obtaining physics data for setpoint and thermal-hydraulics analyses, the staff concludes that it is suitable to use NRC-approved 3-D neutronics codes
~
1 5
such as ROCS and detailed thermal-hydraulics codes such as TORC as an alternative to the synthesis of axial and radial power distribution calculations.
APB-CE DNB Analysis Methodoloav for Mixed Core Acolications ABB-CE uses the CETOP-D and TORC thermal-hydraulics codes with the NRC-approved CE-1 critical heat flux (CHF) correlation (Ref. 8 and 9) for DNB analyses of cores with ABB-CE 14x14 and 16x16 fuel. The CE-1 correlation is imbedded in both codes. In support of the modifications to the setpoint methodology, which relies on DNB calculations using an NRC-approved CHF correlation. ABB-CE has presented data to justify applicability of the CE-1 CHF correlation to mixed cores containing similar, non-mixing grid assembly designs.
The data base for the CE-1 CHF correlation was obtained from a series of CHF tests performed at the Columbia University Heat Transfer Test Facility. The test models were 5x5 array bundles (with and without guide tubes) modeling typical ABB-CE 14x14 and 16x16 fuel assembly geometries. The four 14x14 tests were conducted on test sections containing either 25 heated rods or 21 heated rods and an unheated guide tube type rod. Standard non-mixing vane grids were used in two of the test series and reinforced non-mixing vane grids were used in the other two. The two different grid loss coefficients used in each test series were used in the test data reduction process to determine local coolant conditions needed as input to development of the CE-1 CHF correlation.
A typical St. Lucie Unit 1 mixed core containing Siemens fuel residing with ABB-CE Guardian Grid fuel was used to evaluate the acceptability of the application of the.ABB-CE thermal-hydraulic DNB analysis methods, including the CE-1 CHF correlation, to DNB analyses of transition cores containing similar, non-mixing grid fuel designs. The basic fuel assembly design and relevant geometry characteristics of the two fuel types shown in Table A.1 (Ref. 2) are quite similar. However, grid hydraulic resistance differences, although within the range of the CHF database, would be expected to induce crossflow and non-uniform axial flows over the entire bundle length which must be appropriately treated.
Two full scale 14x14 test bundles were used to demonstrate the capability of l the TORC code to predict axial flow redistribution. The fuel assemblies had I the same basic geometry but contained either standard grids or advanced spacer grids with different hydraulic characteristics located at the same elevation in the upper portion of the assemblies. A comparison of the flow split between assemblies is presented in Figure A.1 (Ref. 2). The results indicate
6 good agreement between TORC predictions and measurements. Therefore, we conclude that TORC accurately predicts the flow conditions in adjacent fuel bundles even when significant differences in grid loss coefficients exist.
Since the hydraulic -esistance mismatch between the Siemens and ABB-CE spacer grids is bounded by ti.at for the grids used in the dual bundle test, the crossflow and resultant axial flow split between the two fuel types in a mixed core of Siemens and ABB-CE 14x14 fuel will be bounded by those in the dual bundle test. Therefore, the ABB-CE thermal-hydraulic design methodology, which is based on the TORC code with the CE-1 CHF correlation, can be used for mixed cores containing coresident Siemens and ABB-CE 14x14 fuel assemblies within the range of coolant conditions spanned by the CE-1 CHF correlation.
- 4.
SUMMARY
AND CONCLUSIONS The staff finds the application of CENPD-199-P. Revision 1-P-A Supplement 2-P. acceptable for referencing in license applications for ABB-CE plants with analog protection systems. Specifica _ w : conclude that:
a) The Xenon Swing methodolev e s.eptable method of generating axial shapes for the s. -
n ysis.
b) The elimination of the TS on F,,is acceptable.
c) The use of NRC-approved 3-D neutronics codes such as ROCS and detailed thermal-hydraulics codes such as TORC are acceptable alternatives to the 1-D/2-D synthesis of axial and radial power distribution calculations for setpoint analysis, d) The use of the TORC code and CE-1 CHF correlation *is acceptable for DNB analysis of mixed cores containing coresident Siemens and ABB-CE 14x14 fuel assemblies with similar, non-mixing grid design characteristics which fall within the range of the CE-1 CHF correlation data base.
1
- 5. REFERENCES
- 1. Letter from I. C. Rickard (ABB-CE) to Document Control Desk (NRC). .
" Topical Report CENPD-199-P. Revision 1-P-A. Supplement 2-P. 'CE Setpoint Methodology. ' September,1997." LD-97-026. September 18.
199/.
7
- 2. "CE Setpoint Methodology." CENPD-199-P. Rev.1-P-A. Supplement 2-P.
. ABB Combustion Engineering Nuclear Operations. September 1997.
- 3. "CE Setpoint Methodology." CENPD-199-P. Rev.1-P-A. Combustion Engineering. Inc., January 1986.
- 4. "The ROCS and DIT Computer Codes for Neclear Design." CENPD-266-P-A.
Combustion Engineering. Inc., April 1983.
- 5. "HERMITE. A Multi-Dimensional Space-Time Kinetics Code for PWR Transients." CENPD-188-A. Combustion Engineering. Inc.. March 1976.
- 6. "CETOP-D Code Structure and Modeling Methods for Arkansas Nuclear One Unit 2." CEN-214(A)-P. Combustion Engineering. Inc., July 1982.
- 7. " TORC Code. Verification and Simplified Modeling Methods."
CENPD-206-P-A Combustion Engineering. Inc., June 1981.
- 8. "C-E Critical Heat Flux. Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids. Part 1 Uniform Axial Power Distributions." CENPD-162-P-A Combustion Engineering. Inc. ,
September 1976.
- 9. "C-E Critical Heat Flux. Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids. Part 2 Nonuniform Axial Power Distributions." CENPD-207-P. Combustion Engineering. Inc., June 1976.
r L .
C ABB-Combustion Engineering, Inc.
cc: Mr. Charles B. Brinkman, Director l Nuclear Systems Licensing i
ABB-Combustion Engineering, Inc. l Post Omco Box 500 l' 2000 Day Hill Road Windsor, Connecticut 06095-0500 Mr. lan C. Rickard, Director-Operations Licensing ABB-Combustion Engineering Nuclear Operations Post Office Box 500 2000 Day Hill Road Windsor, Connecticut 06095-0500 Mr. Charles B. Brinkman, Manager Washington Nuclear Operations ABB-Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 1
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