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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 EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT EMF-2087(P). REVISION 0 "SEM/PWR-98:ECCS EVALUATION MODEL FOR PWR LBLOCA APPLICATIONS" SIEMENS POWER CORPORATION (TAC NO. MA3457) 1 INTRODUCTION AND BACKGROUND Topical Report XN-NF-82-20(P), Revision 1, Supplement 5, "EXEM/PWR Large-Break LOCA .
ECCS TOODEE2 Updates," provided a description of and supporting technicalinformation for I all changes made to the TOODEE2 hot rod heat up computer code made since the code's approvalin 1986. Topical Report XN-NF-82-20(P), Revision 1, Supplement 6, documented the updates related to a change in the implementation of the fuel cooling test facility (FCTF) reflood ,
heat transfer correlation in the TOODEE2 hot rod heat up code and a correction to the Z- l Equivalent model. The staff subsequently approved the modifications to the EXEM/PWR Large-Break LOCA code (Reference 1).
Siemens Power Corporation (SPC) informed the staff of the potential for excessive variability in its EXEM/PWR LBLOCA evaluation model because of excessive variability in the RELAP4 code which forms a base in the EXEM/PWR code (Reference 2). Topical Report EMF-2087(P),
"SEM/PWR-98, Revision 0: ECCS Evaluation Model for PWR LBLOCA Applications," which presented modifications to the EXEM/PWR code to correct the excessive variability problems was submitted for staff review (Reference 3).
Review of the EMF-2087(P), Revision 0, topical report resulted in requests for additional information from the staff (Reference 4), and responses to those requests from SPC (Reference 5).
2 PROPOSED MODEL REVISION ,
While using the EXEM/PWR LBLOCA code, SPC found that some input changes that were expected to be inconsequential would result in large calculated changes in the peak cladding temperature (PCT). Revisions to the code to correct some of the PCT variability resulted in a reduction of more than 50* F for some plants. When PCT calculations result in a drop of 50' F or more, further changes are required to remove non-conservatism in the Dougall-Rohsenow correlation. Further, a 1997 inspection conducted at SPC by the NRC resulted in commitments for revisions of the emergency core cooling system (ECCS) models. The following modet revisions have been presented to correct the PCT variability problem, Dougall-Rohsenow correlation non-conservatism, and commitments made to the staff for ECCS model corNetion.
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.2 Numerical Scheme SPC found that there were several root causes to the " excess variability" problem and made several modifications to the numerical solution method and calculation methodology in order to !
fix the problem. RELAP4-EM originally used the Porsching method (Reference 6) to advance !
the solution in time. SPC modified the Porsching method to use more new time information in the numerical solution scheme effectively making the solution more implicit in time. '
Time step variatiori has been noted to cause calculational variability in large computer codes.
SPC modified the solution method to automatically control the timestep size based on convergence of the numerical scheme. Previously the user had to manually determine time step convergence.This procedure not only assures convergence, but also helps eliminate
" water packing" which makes the calculation oscillatory. The revised numerical scheme has i been used successfully in the boiling-water reactor (BWR) version of the code, giving more
. consistent, converged, and reliable results. !
- Volume Flow Definition -
- The code uses volume average flow in both the momentum equation and for heat transfer and critical heat flux The momentum equation calculation has been modified to use a simple l average of the inlet and outlet junction flows to avoid a discontinuity that occurs under certain conditions when the volume average flow exceeds the outlet flow but the inlet flow density is closest to the volume density. The volume averaged flow has been defined for use in the heat ,
transfer correlations as an integrated average of the absolute value of the flow through the ,
volume. This is always non-negative. Since the Richert-Franz correlation uses a drift flux !
model based on the direction of flow with respect to gravity, the volume averaged flow is ;
assigned a positive or negative value depending on whether it is upward or downward.
Combined System and Hot Channel Calculations !
' Historically, the computers available at the time of the development of the original version of the l SEM/PWR-98 code (the Exxon Water Reactor Evaluation Model) had insufficient memory and i storage to permit complete system blowdown and core hot channel calculations. Current
- generation computers now make it possible to perform the complete calculation set with a single input model. Thus, the previous procedure of performing separate calculations for system blowdown and core heat up have been combined through redefining the parameter ,
dimensions. This modification removes smallinconsistencies between the system calculation !
and the hot channel calculation that existed in the uncoupled calculation. It has not been necessary to modify the analytical methodology to effect this change. !
Elimination of Enthalov Transoort Model i As discussed in the performance of single combined system and hot channel calculations, advances in computer technology no longer necessitate use of coarse nodalization and
- enthalpy transport models to model a reactor coolant system. It is now possible to use denser nodalization schemes that avoid the discontinuous behavior of the enthalpy transport model. ;
- Now, the extent of core nodalization required to reach convergence is determined based on sensitivity studies. ,
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3 Consistency Between RODEX2. RELAP4-EM. and TOODEE2
- The original EXEM-PWR code package applied the RODEX2 fuel rod input only to the hot channel calculation.1This has now been expanded to permit the RODEX2 model to be applied to the average fuel rod, the average rod in the hot assembly, and the hot ' rod: A further modification was made to permit a gadolinia rod should a need arise for analysis of gadolinia bearing rods.
1 Revision of Douaall-Rohsenow Correlation A requirement of 10 CFR 50, Appendix K, is that when a calculated PCT reduction of 50' F occurs for a plant, the known non-conservatism associated with use of the Dougall-Rohsenow l film boiling correlation must be removed. The Dougall-Rohsenow correlation was developed to
- l. be an extension of the Dittus-Bolter single-phase turbulent flow heat transfer correlation to two-phase flow. The problem with nonconservative heat transfer predictions by the Dougall-Rohsenow correlation has been shown to be caused by the equilibrium assumptions used in
. development of the correlation (Reference 7). SPC developed the Richert-Franz heat transfer correlation to correct the deficiencies of the Dougall-Rohsenow correlation in non-equilibrium
. conditions.. !
The Richert-Franz correlation is derived from the same basis as both the Dittus-Boeiter and Dougall-Rohsenow correlations (single phase liquid or steam, and film boiling). The Richert-Franz correlation is designed to take into account both fluids (steam and liquid) interacting at -
the wall. The two fluids are treated with their own properties. While the basic form of the correlation is the same as the two from which it was derived, the Richert-Franz correlation adds together two heat transfer coefficients. One coefficient represents superheated steam at the
- wall while the other represents saturated steam at the wall. The two fluids are based on their superficial velocities in the calculated two-phase flow. The superficial velocities are based on the Ohkawa-Lahey drift flux model used in the SPC BWR methodology (Reference 8)
. The Richert-Franz correlation has been shown to be conservative relative to the measured experimental data from tests performed by Edgerton, Germeshausen & Grier at the Idaho National Engineering Laboratory, tests performed by Combustion Engineering, and thermal hydraulic testing facility (THTF) experiments performed at Oak Ridge National Laboratory.
Revised End-of-Bvoass Model Appendix K to.10 CFR Part 50 provides a definition of end-of-bypass and means by which the end-of-bypass may be calculated. The SPC model has been changed so that instead of
- discarding both the ECCS and non-ECCS water at the end-of-bypass, just the ECCS water
. remaining lin the system is discarded. The primary system water that is calculated to remain in the system is retained. The model defines the end-of-bypass as the minimum of the time that.
sustained positive flow occurs from the upper to the lower downcomer volume, or the time that sustained positive flow occurs from the broken cold leg to the upper downcomer volume, less the time required to fill the portion of the cold leg from the ECCS injection point to the reactor vessel.
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. Revised Pumo Dearadation Model The original EXEM/PWR evaluation model used a two-phase pump degradation model based
- on.Semiscale pump data. Since that time, data more representative of large commercial .
' nuclear power plant pumps have become available. The two-phase pump degradation model ,
has been revised to use data obtained in the CE-EPRI Pump Two-Phase Performance Program j
'(Reference 9). The model is the same one developed for use in the SPC realistic loss-of-coolant accident (LOCA) model The new modelis pressure dependent. Two-phase pump degradation increases as pressure is reduced due to the increase in the ratio of vapor to liquid specific volume. Test cases using the two different pump degradation models showed little difference in computed PCT.
Inertial Flow Estimate for' Critical Flow Mgdgj The numerical method for flow predicti;n was modified to use an inertial flow model to determine if the flow reached the critical flow velocity before the end of the time step. If it does the flow is limited to the critical flow velocity. This modification prevents the over prediction of flow from the break and will only have a small effect concentrated at the beginning of the ;
calculation.-
]
kbble Mass Intearation Model The numerical solution method of the bubble mass integration model was improved. This change removes an approximation used in the previous integration model and will result in a more accurate solution of the equations.
Pumo Model Numerics The pump model numerical integration was modified to become implicit with respect to the pump flow. Since the pump head can depend strongly on the pump flow this improvement will improve the stability and accuracy of the calculation.
Claddina Creen Modelin RODEX2 SPC made several revisions to the cladding creep model in RODEX2. The revised model includes a term for thermal creep (MTYPE=4 option). The functional form of the model was changed to be of the same form as the approved RODEX3 model. The empirical coefficients were chosen to give more conservative results than the RODEX3 implementation which was designed for best estimate applications. The model was reviewed by Pacific Northwest National Laboratory and documented in a technical evaluation report (Reference 10). The staff finds the model acceptable for use in large-break LOCA analyses with rod-average bumups up to 62 GWd/MTU.
3 MODEL ASSESSMENT The assessment of the model changes can be broken down into sensitivity studies and benchmark cases. The benchmark cases are comprised of separate effects tests performed at I
l I
n .
L 5
l the thermal hydraulic testing facility (THTF) facility and integral system test L2-6 performed at the loss-of-fluid testing facility (LOFT).
I Sensitivity Studies -
In keeping with the requirements of 10 CFR 50.46, and 10 CFR 50, Appendix K, that sensitivity j calculations be performed for LOCA codes, SPC has determined appropriate nodalization to l
- represent the reactor core and steam generators through sensitivity studies. In addition, i sensitivity studies have been performed on time step convergence, axial nodalization, radial fuel rod noding, break spectrum, worst single failure, and pump degradation model. Since the Dougall-Rohsenow correlation has been replaced with the Richert-Franz correlation, validation against three sets of film boiling heat transfer data was also performed.
A full break spectrum analysis was performed on a three loop Westinghouse plant using four break sizes (0.4,0.6,0.8 and 1.0) for both the double-ended cold leg guillotine (DECLG) and cold-leg split (CLS), and three axial power shapes, bottom of core (BOC), middle of core (MOC), and top of core (TOC). The limiting break was found to shift from the 0.8 DECLG MOC case to the 1.0 DECLG MOC. The PCT decreased from 1988F to 1925 F.
Assessment of the RODEX2 (MTYPE=4) model changes is described in Reference 10.
Benchmark Cases 1
Large system analysis computer codes are not only to be assessed against separate effects !
data, but also by comparison with integral systems test data so that the overall performance of l the code and the interaction of the various models can be determined. SPC has assessed the l SEM/PWR-98 code against a series of THTF tests as well as against the LOFT L2-6 test. The l THTF test assessments demonstrate the conservative nature of the code in predicting both fluid conditions and heat transfer in film boiling heat transfer; The LOFT test was performed in a i scaled, integral nuclear reactor and demonstrates the ability of the code to predict the system i blowdown and recovery phenomena. The LOFT calculations were performed both with the i Appendix K features tumed off and with them tumed on. I i
Comparisons between SEM/PWR-98 and THTF tests indicate the THTF data are virtually all bounded by the SEM/PWR-98 predictions. The few data points that are not bounded exceed ;
' the predicted values by a small amount, and are not limiting.
The SEM/PWR-98 predictions of the LOFT L2-6 test are reasonably close to the data for ;
system primary pressure, break flow, cold leg flow, accumulator level and pressurizer level. In ,
both the EM and non-EM mode the SEM/PWR-98 prediction does not pick up the LOFT fuel l rod quench that occurs following blowdown because of the CHF locking required in Appendix K
~ Models. Both cases predict a higher blowdown peak cladding temperature, and a much higher PCT at the start of reflood. Thus, the code is conservative relative to the LOFT test PCT, 4 CONCLUSION L The staff has determined that the modifications made by SPC adequately resolve the previous ,
code calculation problems. SPC has performed adequate testing and validation of the new
F . ,
s' l-6 models to show that they conservatively predict the peak cladding temperature for a LBLOCA.
The revised model complies with the required model features of 10 CFR 50, Appendix K.
5' REFERENCES
- 1. Letter from T. H. Essig, NRC, to J. S. Holm, SPC, " Acceptance for Referencing of the Topical Report XN-NF-82-20(P), Revision 1, Supplement 6, 'EXEM/PWR Large Break -
LOCA ECCS TOODEE2 Updates," dated June 5,1998.
I
- 2. Letter from J. F. Mallay, SPC, to NRC, "RELAP4 Excessive Variability," dated March 17,
~1998.
' 3. Le' tter from J. F. Mallay, SPC , to NRC, Request for' Review of EMF-2087(P), Revision 0,
. SEM/PWR-98:ECCS Evaluation Model for PWR LBLOCA Applications," dated
'4. August 31,1998.
- 5. Letter from E. Y. Wang, NRC, to J. Mallay, SPC, Request for Additional information to the Topical Report EMF-2087(P), Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications" TAC NO. MA3457, December 4,1998.
1
- 6. Letter from J. Mallay, SPC to NRC, Request for Additional Information to the Topical j Report EMF-2087(P), Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR i
' LBLOCA Applications" TAC NO. MA3457, December 18,1998.
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i
- 7. Porsching, T. A., Murphy, J. H., Redfield, J. A., " Stable Numerical Integration of Conservation Equations for Hydraulic Networks," Nuclear Science and Technology, Vol. 43 (February 1971), pp. 218-225.
= 8. Morris, D. G., Mullins, C. B., Yoder Jr. G. L., "An Expsrimental Study of Rod Bundle Dispersed-Flow Film Boiling with High-Pressure Water," Nuclear Technology Vol. 69, {
. pp. 82-93.,1985 '
- 9. Ohkawa, K., Lahey Jr., R. T., "The Analysis of Proposed BWR Inlet Flow Blockage Experiments," Department of Nuclear Engineering, Rensselair Polytehnic Institute, Troy, 1 New York, 12181,~ 1978.
- 10. " Pump Two-Phase Performance Program," EPRI NP-15556, Volumes 1-8, September 1980. -;
- 11. Beyer, C. E., " Technical Eva!uation Report of the Topical Report EMF-2087, Entitiled )
_'SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications'," Contract DE- !
AC06-76RLO'1830, May 1999 i 12.' Letter from J. Mallay, SPC to NRC, " Response to Request for Additional Information l
t regarding EMF-2087(P)," Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications" TAC NO. MA3457, NRC:99:010, Dated April 20,1999.
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