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| | issue date = 09/05/1980 | | | issue date = 09/05/1980 |
| | title = Forwards Supplemental ECCS Calculations in Response to NRC 800829 Ltr Re NUREG-0630 | | | title = Forwards Supplemental ECCS Calculations in Response to NRC 800829 Ltr Re NUREG-0630 |
| | author name = MITTL R L | | | author name = Mittl R |
| | author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY | | | author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY |
| | addressee name = MIRAGLIA F J | | | addressee name = Miraglia F |
| | addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) | | | addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) |
| | docket = 05000311 | | | docket = 05000311 |
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| =Text= | | =Text= |
| {{#Wiki_filter:I Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 September 1980 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: | | {{#Wiki_filter:I OPSE~ |
| Mr. Frank J. Miraglia, Chief Licensing Branch 3 Division of Licensing Gentlemen: | | Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 September 5~ 1980 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Frank J. Miraglia, Chief Licensing Branch 3 Division of Licensing Gentlemen: |
| CLADDING SWELLING AND RUPTURE CALCULATIONS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 PSE&G hereby submits, in the enclosure to this letter, the supplemental ECCS calculations related to "Cladding Swelling and Rupture Models for LOCA as requested in your letter of August 29, 1980. Should you have any questions in this regard, do not hesi-tate of contact us. CC: Mr. Leif Norrholm Salem Resident Inspector C02 The Energy People soo91ao 11c6 R. L. Mittl General Manager -Licensing and Environment Engineering and Construction 1// 95-0942 I " . :... * . .. e ATTACHMENT l A. Evaluation of the potential impact of using fuel rod models sented in draft NUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for Salem Units 1 and 2. This evaluation is based on the limiting break LOCA analysis fied as follows: BREAK TYPE -DOUBLE ENDED COLD LEG' GUILLOTINE BREAK DISCHARGE COEFFICIENT | | CLADDING SWELLING AND RUPTURE CALCULATIONS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 PSE&G hereby submits, in the enclosure to this letter, the supplemental ECCS calculations related to NUREG-0630~ |
| | "Cladding Swelling and Rupture Models for LOCA Analysis~" as requested in your letter of August 29, 1980. |
| | Should you have any questions in this regard, do not hesi-tate of contact us. |
| | R. L. Mittl General Manager - |
| | Licensing and Environment Engineering and Construction CC: Mr. Leif Norrholm Salem Resident Inspector C02 1// |
| | The Energy People soo91ao 11c6 95-0942 |
|
| |
|
| ===0.8 WESTINGHOUSE===
| | I " . :... |
| | | * ATTACHMENT l |
| ECCS EVALUATION MODEL VERSION FEBRUAR 1978 CORE PEAKING FACTOR 2.32 HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD -2130°F = PCT 8 ELEVATION | | .. e A. Evaluation of the potential impact of using fuel rod models pre-sented in draft NUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for Salem Units 1 and 2. |
| -6.0 Feet HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD - | | This evaluation is based on the limiting break LOCA analysis identi-fied as follows: |
| = PCTN -ELEVATION | | BREAK TYPE - DOUBLE ENDED COLD LEG' GUILLOTINE BREAK DISCHARGE COEFFICIENT 0.8 WESTINGHOUSE ECCS EVALUATION MODEL VERSION FEBRUAR 1978 CORE PEAKING FACTOR 2.32 HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD - 2130°F = PCT8 ELEVATION - 6.0 Feet HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD - 2003°~ = PCTN - |
| -7.5 Feet CLAD STRAIN DURING SLOWDOWN AT THIS ELEVATION | | ELEVATION - 7.5 Feet CLAD STRAIN DURING SLOWDOWN AT THIS ELEVATION 3.3 Percent MAXIMUM CLAD STRAIN AT THIS ELEVATION - 6.2 Percent Maximum temperatur~ for this node occurs when the core reflood rate is greater than 1.0 inch per second and reflood heat transfer is based on the flecht calculation. |
| | | AVERAGE HOT ASSEMBLY ROD BURST ELEVATION - 6.0 Feet HOT ASSEMBLY BLOCKAGE CALCULATED - 45 Percent |
| ===3.3 Percent===
| | : 1. BURST NODE The maximum potential impact on the ruptured clad node is expressed in letter NS-TMA-2174 in terms of the change in the peaking factor limit (FQ) required to maintain a peak clad temperature (PCT) of 2200°F and in terms of a change in PCT at a constant FQ. Since the clad-water reaction rate increases significantly at temperatures above 2200.°F, in-dividual effects (such as ~PCT due to changes in several fuel rod models) indicated here may not accurately apply over large ranges, but a simultaneous change in FQ which causes the PCT to remain in the neighborhood of 2200.°F justifies use cf this evaluation procedure. |
| MAXIMUM CLAD STRAIN AT THIS ELEVATION | |
| -6.2 Percent Maximum for this node occurs when the core reflood rate is greater than 1.0 inch per second and reflood heat transfer is based on the flecht calculation. | |
| AVERAGE HOT ASSEMBLY ROD BURST ELEVATION | |
| -6.0 Feet HOT ASSEMBLY BLOCKAGE CALCULATED | |
| -45 Percent 1 . BURST NODE The maximum potential impact on the ruptured clad node is expressed in letter NS-TMA-2174 in terms of the change in the peaking factor limit (FQ) required to maintain a peak clad temperature (PCT) of 2200°F and in terms of a change in PCT at a constant FQ. Since the clad-water reaction rate increases significantly at temperatures above 2200.°F, dividual effects (such as due to changes in several fuel rod models) indicated here may not accurately apply over large ranges, but a simultaneous change in FQ which causes the PCT to remain in the neighborhood of 2200.°F justifies use cf this evaluation procedure. | |
| From NS-TMA-2174: | | From NS-TMA-2174: |
| For the Burst Node of the clad: 0.01 AFQ 150°F BURST NODE APCT Use of the NRC burst model could require an FQ reduction of 0.015 The maximum estimated impact of using the NRC strain model is a required FQ reduction of 0.03. Therefore, the maximum penalty for the Hot Rod burst node is: APCT 1 = (.015 + .03) (l50°F/.Ol) | | For the Burst Node of the clad: |
| = 675°F Margin to the 2200°F limit is: APCT 2 = 2200.°F -PCT 8 = 70°F The FQ reduction is required to maintain the 2200°F clad ture limit is. AFQ 8 = (APCT 1 -APCT ) (.Ol AFQ) 2 l 50°F = (675 -70) co1) 150 = .04 (but not less than zero). 2. NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the core mid-plane during the core reflood phase of the LOCA transient. | | 0.01 AFQ ~ ~ 150°F BURST NODE APCT Use of the NRC burst model could require an FQ reduction of 0.015 The maximum estimated impact of using the NRC strain model is a required FQ reduction of 0.03. |
| The tial impact on that maximum clad temperature of using the NRC fuel rod models can be estimated by examining two aspects of the analyses. | | Therefore, the maximum penalty for the Hot Rod burst node is: |
| The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation. | | APCT 1 = (.015 + .03) (l50°F/.Ol) = 675°F Margin to the 2200°F limit is: |
| Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can change the time at which burst is calculated. | | APCT 2 = 2200.°F - PCT 8 = 70°F The FQ reduction is required to maintain the 2200°F clad tempera-ture limit is. |
| Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. | | AFQ 8 |
| The possible PCT increase resulting from a change in strain (in the Hot Rod) is +20.°F per percent decrease in strain at the maximum clad temperature locations. | | = (APCT 1 - APCT ) (.Ol AFQ) 2 l 50°F |
| Since the clad strain calculated during the reactor coolant system blowdown phase of the accident is not changed by the use of NRC fuel rod models, the maximum decrease in clad strain that must be considered here is the difference bet\ .. !een the 11 maximum clad strain 11 and the 11 clad strain during blowdown 11 indicated above. | | = (675 - 70) co1) 150 |
| {-. '.......... Therefore: | | = .04 (but not less than zero). |
| t,PCT 3 20°F = C.Ol strain) (MAX STRAIN -SLOWDOWN STRAIN) = ( 20) (.062 -.033) . 01 = 58 The second aspect of the analysis that can increase PCT is the flow blockage calculated. | | : 2. NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the core mid-plane during the core reflood phase of the LOCA transient. The poten-tial impact on that maximum clad temperature of using the NRC fuel rod models can be estimated by examining two aspects of the analyses. The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation. Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can change the time at which burst is calculated. Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. The possible PCT increase resulting from a change in strain (in the Hot Rod) is +20.°F per percent decrease in strain at the maximum clad temperature locations. Since the clad strain calculated during the reactor coolant system blowdown phase of the accident is not changed by the use of NRC fuel rod models, the maximum decrease in clad strain that must be considered here is the difference bet\.!een the 11 maximum clad strain 11 and the 11 clad strain during blowdown 11 indicated above. |
| Since the greatest value of blockage indicated by the NRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivity formula shown in NS-TMA-2174. | | {-. '.......... ~* |
| Therefore, t,PCT 4 = l.25°F (50 -PERCENT CURRENT BLOCKAGE) | | Therefore: |
| + 2.36°F (75-50) = 1.25 (50 -45) + 2.36 (75-50) = 65°F If PCTN occurs when the core reflood rate is greater than 1.0 inch per second t,PCT 4 = 0. The total potential PCT increase for the non-burst node is then Margin to the 2200°F limit is = 2200°F -PCTN The FQ reduction required to maintain this 2200°F clad perature limit is (from NS-TMA-2174) ) t,FQN = -0. 14 but not less than zero. = 0 The peaking factor reduction required to maintain the 2200°F clad temperature limit is therefore the greater of t,FQ 8 and ACn ACn - | | t,PCT 3 20°F |
| VI' | | = C.Ol strain) (MAX STRAIN - SLOWDOWN STRAIN) 20 ) |
| -*-* | | = (. 01 (.062 - .033) |
| ,,.. -B. The effect on LOCA analysis results of using improved analytical and modeling techniques (which are currently approved for use in the Upper Head Injection plant LOCA analyses) in the reactor ant system blowdown calculation (SATAN computer code) has been via an analysis which has recently been submitted to the NRC for review. Recognizing that review of that analysis is not yet complete and that the benefits associated with those model improvements can change for other plant designs, the NRC has established a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod models. That credit for two, three and four loop plants is an increase in the LOCA peaking factor limit of 0. 12, 0. 15 and 0.20 respectively.
| | = 58 The second aspect of the analysis that can increase PCT is the flow blockage calculated. Since the greatest value of blockage indicated by the NRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivity formula shown in NS-TMA-2174. |
| C. The peaking factor limit adjustment required to justify plant operation for this interim period is determined as the appropriate credit identified in section (B) above, minus the calculated in section (A) above (but not greater than zero). FQ ADJUSTMENT= | | Therefore, t,PCT 4 = l.25°F (50 - PERCENT CURRENT BLOCKAGE) |
| .2 -.04 -7 0 }} | | + 2.36°F (75-50) |
| | = 1.25 (50 - 45) + 2.36 (75-50) |
| | = 65°F If PCTN occurs when the core reflood rate is greater than 1.0 inch per second t,PCT 4 = 0. The total potential PCT increase for the non-burst node is then Margin to the 2200°F limit is |
| | ~PCT 6 = 2200°F - PCTN The FQ reduction required to maintain this 2200°F clad tem-perature limit is (from NS-TMA-2174) |
| | ) |
| | t,FQN = -0. 14 but not less than zero. |
| | =0 The peaking factor reduction required to maintain the 2200°F clad temperature limit is therefore the greater of t,FQ 8 and ACn ~~* ACn - n~ |
| | Ul~N' VI' ~*~PENALTY - *-* |
| | B. The effect on LOCA analysis results of using improved analytical and modeling techniques (which are currently approved for use in the Upper Head Injection plant LOCA analyses) in the reactor cool-ant system blowdown calculation (SATAN computer code) has been qu~ntified via an analysis which has recently been submitted to the NRC for review. Recognizing that review of that analysis is not yet complete and that the benefits associated with those model improvements can change for other plant designs, the NRC has established a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod models. That credit for two, three and four loop plants is an increase in the LOCA peaking factor limit of |
| | : 0. 12, 0. 15 and 0.20 respectively. |
| | C. The peaking factor limit adjustment required to justify plant operation for this interim period is determined as the appropriate |
| | ~FQ credit identified in section (B) above, minus the ~FQPENALTY calculated in section (A) above (but not greater than zero). |
| | FQ ADJUSTMENT= .2 - .04 |
| | -7 0 |
| | }} |
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Rept Summarizes Liquid & Gaseous Releases & Solid Waste Shipments from Salem Generating Station for Period of Jan-June 1999 ML18107A5221999-09-0808 September 1999 Requests Approval to Use ASME Code Case N546,which Provides Alternative Qualification Requirements That Allow Personnel Most Familiar with Walkdown of Plant Sys,Like License Operators to Perform VT-2 Examinations ML18107A4981999-08-26026 August 1999 Forwards Response to NOV That Resulted from Predecisional Enforcement Conference Conducted on 990624.Corrective Actions:Communications to Supervisors Reinforced Employee Right & Duty to Raise Nuclear Safety Issues ML18107A5181999-08-26026 August 1999 Forwards Ninety Day Rept for ISI Activities Conducted at Sgs,Unit 2 During Ninth Extended Outage & Tenth Refueling Outage.List of Encl,Provided ML18107A5061999-08-26026 August 1999 Provides First Feedback from Observation of NRC Insp Under Pilot Nuclear Power Plant Insp Program.Attached Are Completed Insp Feedback Forms for Procedure 71111, Attachment 21 & Procedure 71151 ML18107A5051999-08-24024 August 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Request IAW Requirements of 10CFR50.91(b) & Has No Comments Re Amend to FOL Change Request S99-02 to Modify TS Re Penetration Valves ML18107A4921999-08-23023 August 1999 Provides Suppl Info Re 971024 Amend Request to Modify TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Util Withdraws from Request All Proposed Changes Associated with Filter Testing,Per Issuance of GL 99-02 ML18107A4911999-08-20020 August 1999 Forwards Revised Plant Attribute Libraries for Salem & Hope Creek Generating Stations,Iaw 10CFR50,App E,Vi, Emergency Response Data Sys, 3.b.Changes Are Identified by Rev Bars ML18107A4831999-08-18018 August 1999 Submits Licensee Comments on NRC 990730 Ltr Which Provided Notification of Close Out of TAC Numbers MA0567 & MA0568 Re GL 92-01,Rev 1,Suppl 1 ML18107A4801999-08-13013 August 1999 Requests That Pse&Gs Contact in NUREG-0383, Directory of Compliance for Radioactive Matl Packages, Be Changed ML18107A4751999-08-0505 August 1999 Forwards Fitness for Duty Performance Data Rept for Six Month Period Ending 990630 ML20210M7571999-08-0404 August 1999 Forwards Response to Requesting Addl Info Re Status of Decommissioning Funding for Lgs,Pbaps & Sngs. Attachment Provides Restatement of Questions Followed by Response ML18107A4431999-07-0606 July 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S98-17 Re Permissible Enrichment Values for New Fuel Storage & Has No Comments ML18107A4181999-06-30030 June 1999 Submits Response to NRC Request for Info Re Y2K Readiness at Npps,Per GL 98-01,suppl 1.Disclosure Encl ML18107A4131999-06-25025 June 1999 Provides Further Clarification of Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting Station SBO & Loca/Loop Loading Requirements,Per Util 990426 Ltr & Discussion with NRC ML18107A4061999-06-21021 June 1999 Provides Supplemental Info to Proposed License Amend, Modifying TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Info Re Acceptance Criterion Discussed ML18107A3691999-06-11011 June 1999 Forwards Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1.Original Submittal Contained Typo for year-to-date Value for Numbers of Hours Generator Was on Line (Service Hours) ML18107A3641999-06-0404 June 1999 Requests Enforcement Discretion for TS 3/4.6.2.3 Re Containment Cooling Sys for Salem Generating Station,Unit 1 ML18107A3561999-06-0303 June 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S99-05 & Has No Comments ML18107A3611999-05-27027 May 1999 Forwards Responses to NRC 990301 & 990323 RAIs for Salem & Hope Creek Generating Stations Relating to GL 96-05 ML18107A3301999-05-24024 May 1999 Forwards Suppl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety Related Power Operated Gate Valves. Encl Contains Methodology Used in Determination of Pressure Locking Susceptibility of PORVs Block Valves ML18107A3291999-05-20020 May 1999 Forwards Redacted Response to NRC 990322 RAI Re Notification of Licensed Operator That Tested Positive for Alcohol. Attachment 2 Withheld,Per 10CFR2.790(a)(6) ML18107A3031999-05-18018 May 1999 Provides Summary of Changes to NRC Commitments That Have Been Made But Not Reported by Other Means,Iaw with NEI Process for Managing NRC Commitments ML18107A2891999-05-13013 May 1999 Forwards Rev 36 to Pse&G Nuclear Business Unit Emergency Plan. Rev 36 Incorporates Changes to Section 1-3,6 & 7 & 9-17.Attached Copy Includes All Sections of EP for Completeness ML18107A2951999-05-12012 May 1999 Submits SG Tube Plugging Rept,Per Plant TS 4.4.6.5.a.Total of 47 Tubes Were Plugged During SG Tube Insps,Which Were Completed During Plant Tenth RFO ML18107A2861999-05-11011 May 1999 Forwards Rev 0 to NFS-0174, COLR for Salem Unit 2 Cycle 11. COLR Rept Was Received by Util as Part of Reload SE ML18107A2481999-04-29029 April 1999 Forwards 1998 Annual Radiological Environ Operating Rept for Salem & Hope Creek Generating Stations. Rept Summarizes Results of Radiological Environ Surveillance Program for 1998 ML18107A2511999-04-27027 April 1999 Submits 30-day Fuel Clad Temp Rept for Salem Generating Station,Units 1 & 2.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Plant Large & Small LOCA & Small Break LOCA Analyses ML18107A2371999-04-26026 April 1999 Forwards Corrected Response to NRC RAI Re Licensee Request for Change to TS Permissible Enrichment Values for New Fuel Storage.Incorrect Attachment Was Provided with Util 990412 Ltr to Nrc.Encl Supersedes 990412 Submittal ML18107A2631999-04-26026 April 1999 Provides Clarification on Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting SBO & Loca/ LOOP Loading Requirements ML18107A2411999-04-22022 April 1999 Forwards Draft Revised Pages 4.1 & 4.2 of Nuclear Business Unit Emergency Plan for Hope Creek & Salem Generating Stations.Changes Are Noted in Italics ML18107A1841999-04-14014 April 1999 Forwards PSEG Annual Rept for 1998, & PECO Annual Rept for 1998. Stockholders Annual Rept of Each Owner & Cash Flow Statements Showing 1998 Actual & 1999 Projected Cash Flow with Explanation Encl ML18107A1981999-04-12012 April 1999 Responds to 990312 RAI Re Request for Change to TSs Permissible Enrichment Values for New Fuel Storage,Which Was Submitted on 990202 ML18107A1691999-04-12012 April 1999 Forwards Proprietary & non-proprietary Epips,Including Rev 17 to EPIP 807,rev 1 to NC.EP-EP.ZZ-0801(Q) & Rev 2 to NC.EP-EP.ZZ-0806(Q) & Revised EPIPs Table of Contents. Proprietary Info Withheld ML20205K4541999-04-0808 April 1999 Forwards Revised Info Re 990330 NRC Nuclear Power Reactor Licensee Financial Qualifications & Decommissioning Funding Assurance Status Rept ML18106B1491999-04-0505 April 1999 Forwards Drafts of Proposed Changes to Pages 4.1 & 4.2 of Emergency Plan,Which Are Contained on Page 4.2 & Noted in Italics & Underlined ML20205F8981999-03-31031 March 1999 Provides Info Re Status of Decommissioning Funding for LGS, Units 1 & 2,PBAPS,Units 1,2 & 3 & Sgs,Units 1 & 2,per Requirements of 10CFR50.75(f)(1) ML18106B1431999-03-31031 March 1999 Forwards Pse&G Rept on Financial Min Assurance for Period Ending 981231 for Hope Creek,Salem,Units 1 & 2 & Pbaps,Units 2 & 3,IAW 10CFR50.75 ML18107A2201999-03-30030 March 1999 Forwards Final Exercise Rept for 980303,full-participation Plume Exposure Pathway Exercise & 980505-07, full-participation Ingestion Pathway Exercise of Offsite Radiological Emergency Response for Salem & Hope Creek 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARML18095A4881990-09-17017 September 1990 Requests Regional Waiver of Compliance from Tech Spec 3.6.2.3, Containment Cooling Sys. Waiver Requested in Order to Allow Replacement of Containment Fan Cooler Unit Motor #22 W/O Requiring Plant Shutdown ML18095A4901990-09-13013 September 1990 Provides Supplemental Info Applicable to Clarification of 10CFR50,App R Exemption Request Re Fire Suppression Sys for Panel 335,per NRC Request ML20059E6821990-09-0404 September 1990 Forwards Info Re Temporary Mod to Security Plan Concerning Protected Area.Info Withheld ML18095A4641990-08-31031 August 1990 Forwards Revised Response to NRC Bulletin 88-004 Re Potential pump-to-pump Interaction.Util Pursuing Permanent Solution to Issue & Will Implement Appropriate Permanent Field Change by End of Unit 1 10th Refueling Outage ML18095A4621990-08-31031 August 1990 Provides Revised Response to Generic Ltr 89-13, Svc Water Problems Affecting Safety-Related Equipment. Only HXs Exhibiting Unsatisfactory Test Results Will Be Inspected & Possibly Cleaned ML18095A4431990-08-30030 August 1990 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept,Jan-June 1990 & Rev 6 to Odcm. ML18095A4531990-08-30030 August 1990 Forwards RERR-28, Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1990 & Revised Odcm.W/O Revised ODCM ML18095A4391990-08-29029 August 1990 Forwards Semiannual Rept Re fitness-for-duty Performance Data for 6-month Period Ending 900630,per 10CFR26.71(d).Rept Includes Testing Results,Random Testing Program Results & Confirmed Positive Tests for Specific Substances ML18095A4421990-08-28028 August 1990 Clarifies 900710 Request for Amends to Licenses DPR-70 & DPR-75,changing Sections I & M.Under Proposed Change,Section I Should Be Changed to Read Section 2.J for License DPR-70 & Section M Changed to Read Section 2.N for License DPR-75 ML20059B6611990-08-22022 August 1990 Confirms That 10 Anchor/Darling Model S350W Swing Check Valves Installed at Plant,Per NRC Bulletin 89-002.All 18 Valves Inspected & Retaining Block Studs Replaced W/Upgraded Matl.No Crack Noted on Any Studs Which Were Replaced ML20059C2861990-08-21021 August 1990 Provides Correction to 900810 Response to Request for Addl Info Re Util Request for Restatement of OL Expiration Dates ML18095A4151990-08-10010 August 1990 Forwards Response to Request for Addl Info Re Reinstatement of OL Expiration Dates Based on Original Issuance of Ols. Advises That Correct Expiration Date for OL Proposed to Be 200418 ML18095A4091990-08-0909 August 1990 Forwards Responses to NRC Comments Re Plant Simulator Certification for 10CFR55.45(b)(2),per 891228 Ltr ML18095A4061990-08-0808 August 1990 Forwards Corrected marked-up Pages for Tech Spec Table 3.3-11 Re Subcooling Margin Monitor & Reactor Vessel Level Instrumentation Sys,Per 900223 Ltr.Administrative Changes Made as Indicated ML18095A3861990-07-30030 July 1990 Forwards Listing of Station Blackout Major Audit Items Resolution Scope,Per Station Blackout Schedule Commitment ML18095A3661990-07-26026 July 1990 Forwards Decommissioning Repts for Hope Creek,Peach Bottom & Salem Nuclear Generating Stations ML18095A3761990-07-26026 July 1990 Forwards Decommissioning Repts & Certification of Financial Assurance for Plants ML18095A3721990-07-24024 July 1990 Forwards Rept & Certification of Financial Assurance for Decommissioning for Plants,Per 10CFR50.75 ML18095A3751990-07-18018 July 1990 Provides Status of Commitments Made to NRC by Util in 900109 Ltr Re NUREG-0737,Item II.D.1,per 900628 Telcon ML18095A3741990-07-18018 July 1990 Provides Supplemental Info Re Facility sub-cooling Margin Monitor ML18095A3611990-07-18018 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. ML18095A3621990-07-18018 July 1990 Forwards Corrected Tech Spec Page 3/4 3-5 for License Change Request 89-12 Submitted on 891227 & 900521 ML18095A3591990-07-13013 July 1990 Corrects Typo in 900702 Response to Generic Ltr 90-04 Re Schedule for Completion of Remaining Open Items ML18095A3471990-07-11011 July 1990 Responds to NRC 900611 Ltr Re Violations Noted in Insp Repts 50-272/90-14 & 50-311/90-14.Corrective Actions:Directive from Radiation Protection Mgt to All Radiation Protection Personnel Issued Re Control of Compliance Agreement Sheets ML18095A3451990-07-10010 July 1990 Forwards Addl Info Re License Change Request 89-03 Re Reactor Trip Sys Instrumentation ML18095A3461990-07-10010 July 1990 Responds to NRC 900608 Ltr Re Violations Noted in Insp Repts 50-272/90-12 & 50-311/90-12.Corrective Actions:Assessment of ECCS & Component Performance Undertaken & ECCS Flow Testing Procedure Upgraded to Address Human Factors ML18095A3491990-07-10010 July 1990 Forwards Jn Steinmetz of Westinghouse 900614 Ltr Re Reassessment of Util Response to Bulletin 88-002 ML18095A3481990-07-10010 July 1990 Submits Supplemental Rept Identifying Root Cause of Missed Commitment & Corrective Actions to Assure Future Compliance Re Implementation of Mods to Facility PASS ML18095A3441990-07-0909 July 1990 Provides Written Notification Re Change in Calculated Peak Clad Temp,Per 900606 Verbal Notification ML18095A3281990-07-0202 July 1990 Responds to NRC 900530 Ltr Re Violations Noted in Insp Repts 50-272/90-09 & 50-311/90-09.Corrective Actions:Util Intends to Use Nuclear Shift Supervisor as Procedure Reader & EOP, Rev 2 Under Development ML18095A3301990-07-0202 July 1990 Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues.Table Describing Status of Generic Safety Issue Implementation Encl ML18095A3391990-06-29029 June 1990 Forwards Correction to 890913 License Change Request 88-09, Consisting of Tech Spec Page 3/4 4-13 ML18095A3221990-06-28028 June 1990 Provides Supplemental Info Re 900223 Proposed Revs to Tech Specs for Reactor Vessel Level Instrumentation Sys.Tables 3.3-11a & 3.3-11b Should Be Combined Into Single Table ML18095A3231990-06-28028 June 1990 Responds to NRC 900518 Ltr Re Violations Noted in Insp Repts 50-272/90-10,50-311/90-10 & 50-354/90-07.Two Noncited Violations Disputed.Util fitness-for-duty Program Exceeds Part 26 Requirements for Positive Blood Alcohol Limits ML18095A3241990-06-28028 June 1990 Forwards Retyped Pages to 871224 License Change Request 87-15 & Modified,Per 900620 Ltr ML18095A3211990-06-26026 June 1990 Requests 30-day Extension Until 900730 to Provide Completion Schedule to Resolve Audit Findings Re Station Blackout ML18095A3161990-06-25025 June 1990 Forwards Supplemental Info Re Response to Generic Ltr 88-14. All Committed Actions Complete as of 900613 ML18095A3141990-06-25025 June 1990 Provides Schedule Change for Implementation of Control Room Mods.Schedule Modified to Address Overhead Annunciator Human Engineering Discrepancies During Phase III ML18095A3201990-06-25025 June 1990 Responds to NRC 900524 Ltr Re Violations Noted in Insp Repts 50-272/90-11 & 50-311/90-11.Corrective Actions:All Overdue Operations & Maint Procedure Files Reviewed for Outstanding Rev Requests & Procedure Upgrade Program Initiated ML18095A3001990-06-20020 June 1990 Provides Addl Info Re Application for Amend to Licenses DPR-70 & DPR-75 Concerning Turbine Valve Surveillance Interval,Per 900320 Request.Util Adding Direction to Personnel If Unnacceptable Flaws Found ML20043H6221990-06-20020 June 1990 Provides Supplemental Info Re NRC Bulletin 88-008 for Fifth Refueling Outage.Detailed Test Rept Being Prepared to Document Results of Each Individual Insp Re Insulation, Hangers & High Energy Break Areas ML18095A2991990-06-20020 June 1990 Forwards Westinghouse Affidavit Supporting 900412 Request for Withholding Proprietary Info from Public Disclosure Per 10CFR2.790 ML18095A2721990-06-0808 June 1990 Responds to NRC 900329 Ltr Re Weaknesses Noted in Insp Repts 50-272/90-80 & 50-311/90-80.Corrective Actions:Fire Doors Placed on Blanket Preventive Maint Work Order & Damaged Fire Doors Will Be Repaired Immediately ML18095A2711990-06-0606 June 1990 Submits Info in Support of 900522 Verbal Request for Relief from Requirements of ASME Section XI ML18095A2611990-06-0101 June 1990 Forwards Corrected Operating Data Rept, Page for Apr 1990 Monthly Operating Rept ML18095A2521990-06-0101 June 1990 Forwards Application in Support of Request for Renewal of NJPDES Permit NJ0005622,per Requirements of Subsection 3.2 of Plant Environ Protection Plan,Nonradiological ML18095A2591990-06-0101 June 1990 Forwards Corrected Unit Shutdown & Power Reductions, Page for Apr 1990 Monthly Operating Rept ML18095A2411990-05-30030 May 1990 Submits Special Rept 90-4 Addressing Steam Generator Tube Plugged During Fifth Refueling Outage.Plugging Completed on 900516.Cause of Tube Degradation Attributed to Normal Wear Due to Erosion/Corrosion Factors ML18095A2431990-05-30030 May 1990 Informs of Util Plans Re Facility Cycle 6 Reload Core, Expected to Achieve Burnup of 16600 Mwd/Mtu.All Postulated Events within Allowable Limits Based on Review of Basis of Cycle 6 Reload Analysis & Westinghouse SER ML18095A2531990-05-29029 May 1990 Provides Addl Info Re End of Life Moderator Temp Coefficient.Feedback Used in Steam Line Break Has No Relationship to Full Power Moderator Density Coefficient 1990-09-04
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I OPSE~
Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 September 5~ 1980 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Frank J. Miraglia, Chief Licensing Branch 3 Division of Licensing Gentlemen:
CLADDING SWELLING AND RUPTURE CALCULATIONS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 PSE&G hereby submits, in the enclosure to this letter, the supplemental ECCS calculations related to NUREG-0630~
"Cladding Swelling and Rupture Models for LOCA Analysis~" as requested in your letter of August 29, 1980.
Should you have any questions in this regard, do not hesi-tate of contact us.
R. L. Mittl General Manager -
Licensing and Environment Engineering and Construction CC: Mr. Leif Norrholm Salem Resident Inspector C02 1//
The Energy People soo91ao 11c6 95-0942
I " . :...
.. e A. Evaluation of the potential impact of using fuel rod models pre-sented in draft NUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for Salem Units 1 and 2.
This evaluation is based on the limiting break LOCA analysis identi-fied as follows:
BREAK TYPE - DOUBLE ENDED COLD LEG' GUILLOTINE BREAK DISCHARGE COEFFICIENT 0.8 WESTINGHOUSE ECCS EVALUATION MODEL VERSION FEBRUAR 1978 CORE PEAKING FACTOR 2.32 HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD - 2130°F = PCT8 ELEVATION - 6.0 Feet HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD - 2003°~ = PCTN -
ELEVATION - 7.5 Feet CLAD STRAIN DURING SLOWDOWN AT THIS ELEVATION 3.3 Percent MAXIMUM CLAD STRAIN AT THIS ELEVATION - 6.2 Percent Maximum temperatur~ for this node occurs when the core reflood rate is greater than 1.0 inch per second and reflood heat transfer is based on the flecht calculation.
AVERAGE HOT ASSEMBLY ROD BURST ELEVATION - 6.0 Feet HOT ASSEMBLY BLOCKAGE CALCULATED - 45 Percent
- 1. BURST NODE The maximum potential impact on the ruptured clad node is expressed in letter NS-TMA-2174 in terms of the change in the peaking factor limit (FQ) required to maintain a peak clad temperature (PCT) of 2200°F and in terms of a change in PCT at a constant FQ. Since the clad-water reaction rate increases significantly at temperatures above 2200.°F, in-dividual effects (such as ~PCT due to changes in several fuel rod models) indicated here may not accurately apply over large ranges, but a simultaneous change in FQ which causes the PCT to remain in the neighborhood of 2200.°F justifies use cf this evaluation procedure.
From NS-TMA-2174:
For the Burst Node of the clad:
0.01 AFQ ~ ~ 150°F BURST NODE APCT Use of the NRC burst model could require an FQ reduction of 0.015 The maximum estimated impact of using the NRC strain model is a required FQ reduction of 0.03.
Therefore, the maximum penalty for the Hot Rod burst node is:
APCT 1 = (.015 + .03) (l50°F/.Ol) = 675°F Margin to the 2200°F limit is:
APCT 2 = 2200.°F - PCT 8 = 70°F The FQ reduction is required to maintain the 2200°F clad tempera-ture limit is.
AFQ 8
= (APCT 1 - APCT ) (.Ol AFQ) 2 l 50°F
= (675 - 70) co1) 150
= .04 (but not less than zero).
- 2. NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the core mid-plane during the core reflood phase of the LOCA transient. The poten-tial impact on that maximum clad temperature of using the NRC fuel rod models can be estimated by examining two aspects of the analyses. The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation. Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can change the time at which burst is calculated. Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. The possible PCT increase resulting from a change in strain (in the Hot Rod) is +20.°F per percent decrease in strain at the maximum clad temperature locations. Since the clad strain calculated during the reactor coolant system blowdown phase of the accident is not changed by the use of NRC fuel rod models, the maximum decrease in clad strain that must be considered here is the difference bet\.!een the 11 maximum clad strain 11 and the 11 clad strain during blowdown 11 indicated above.
{-. '.......... ~*
Therefore:
t,PCT 3 20°F
= C.Ol strain) (MAX STRAIN - SLOWDOWN STRAIN) 20 )
= (. 01 (.062 - .033)
= 58 The second aspect of the analysis that can increase PCT is the flow blockage calculated. Since the greatest value of blockage indicated by the NRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivity formula shown in NS-TMA-2174.
Therefore, t,PCT 4 = l.25°F (50 - PERCENT CURRENT BLOCKAGE)
+ 2.36°F (75-50)
= 1.25 (50 - 45) + 2.36 (75-50)
= 65°F If PCTN occurs when the core reflood rate is greater than 1.0 inch per second t,PCT 4 = 0. The total potential PCT increase for the non-burst node is then Margin to the 2200°F limit is
~PCT 6 = 2200°F - PCTN The FQ reduction required to maintain this 2200°F clad tem-perature limit is (from NS-TMA-2174)
)
t,FQN = -0. 14 but not less than zero.
=0 The peaking factor reduction required to maintain the 2200°F clad temperature limit is therefore the greater of t,FQ 8 and ACn ~~* ACn - n~
Ul~N' VI' ~*~PENALTY - *-*
B. The effect on LOCA analysis results of using improved analytical and modeling techniques (which are currently approved for use in the Upper Head Injection plant LOCA analyses) in the reactor cool-ant system blowdown calculation (SATAN computer code) has been qu~ntified via an analysis which has recently been submitted to the NRC for review. Recognizing that review of that analysis is not yet complete and that the benefits associated with those model improvements can change for other plant designs, the NRC has established a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod models. That credit for two, three and four loop plants is an increase in the LOCA peaking factor limit of
- 0. 12, 0. 15 and 0.20 respectively.
C. The peaking factor limit adjustment required to justify plant operation for this interim period is determined as the appropriate
~FQ credit identified in section (B) above, minus the ~FQPENALTY calculated in section (A) above (but not greater than zero).
FQ ADJUSTMENT= .2 - .04
-7 0