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| | issue date = 07/31/1989 | | | issue date = 07/31/1989 |
| | title = Forwards Addl Info Re Pressure Temp Limits,Per NRC 890526 Request & Discussion W/Tsao of NRC on 890724.Capsule Z Did Not Contain Any Weld Metal Specimens.Therefore,Discussion of Specimens Not Included in WCAP-11955 | | | title = Forwards Addl Info Re Pressure Temp Limits,Per NRC 890526 Request & Discussion W/Tsao of NRC on 890724.Capsule Z Did Not Contain Any Weld Metal Specimens.Therefore,Discussion of Specimens Not Included in WCAP-11955 |
| | author name = LABRUNA S | | | author name = Labruna S |
| | author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY | | | author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:Public Service Electric and Gas Company | | {{#Wiki_filter:Public Service Electric and Gas Company Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4800 Vice President - Nuclear Operations July 31, 1989 NLR-N89139 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen: |
| * Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4800 Vice President
| | REQUEST FOR ADDITIONAL INFORMATION PRESSURE TEMPERATURE LIMITS SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 The purpose of this letter is to provide PSE&G's responses to the questions forwarded in the request for additional information dated May 26, 1989. The attached responses were discussed with Mr. Tsao of the NRC staff on July 24, 1989. Figures 3.4-2 and 3.4-3 for Unit 1 were revised as a result of this conversation and are attached for incorporation into LCR 88-14. |
| -Nuclear Operations July 31, 1989 NLR-N89139 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen: | | Should you have any questions, please feel free to contact us. |
| REQUEST FOR ADDITIONAL INFORMATION PRESSURE TEMPERATURE LIMITS SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 The purpose of this letter is to provide PSE&G's responses to the questions forwarded in the request for additional information dated May 26, 1989. The attached responses were discussed with Mr. Tsao of the NRC staff on July 24, 1989. Figures 3.4-2 and 3.4-3 for Unit 1 were revised as a result of this conversation and are attached for incorporation into LCR 88-14. Should you have any questions, please feel free to contact us. Attachment 89-oaoso2:30--89o731 | | Sincerely, Attachment 89-oaoso2:30--89o731 --.----. |
| --.----. P PDR ADOCK 05000272 PNU Sincerely, I I | | PDR ADOCK 05000272 P PNU I I |
| * *----*-' Document Control Desk NLR-N89139 C Mr. J. c. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector 2 Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief
| | |
| * New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 7-31-89
| | Document Control Desk NLR-N89139 2 |
| *
| | * 7-31-89 C Mr. J. c. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 |
| * ATTACHMENT QUESTIONS | | * ATTACHMENT QUESTIONS & RESPONSES - PRESSURE/TEMPERATURE LIMITS (LCR 88-14) |
| & RESPONSES | | A. WCAP-11955 describes the test results of specimens in Capsule Z at Salem 1. The following questions relate to WCAP-11955 which provides the technical bases for the 12/28/88 submittal for Salem, Unit 1: |
| -PRESSURE/TEMPERATURE LIMITS (LCR 88-14) A. WCAP-11955 describes the test results of specimens in Capsule Z at Salem 1. The following questions relate to WCAP-11955 which provides the technical bases for the 12/28/88 submittal for Salem, Unit 1: 1. There was no discussion of weld metal in Section 4, Description of Program, in WCAP-11955. | | : 1. There was no discussion of weld metal in Section 4, Description of Program, in WCAP-11955. However, Table 4-1 discusses weld metal from Capsule Y. Table 4-2 and Table A-1 discusses a certain weld metal but we could not determine whether that weld metal was included in Capsule z. Was there a weld metal included in Capsule Z? |
| However, Table 4-1 discusses weld metal from Capsule Y. Table 4-2 and Table A-1 discusses a certain weld metal but we could not determine whether that weld metal was included in Capsule z. Was there a weld metal included in Capsule Z? Rl. Salem Unit 1 Capsule z did not contain any weld metal specimens. | | Rl. Salem Unit 1 Capsule z did not contain any weld metal specimens. Therefore, no discussion of weld metal results was included in WCAP-11955. |
| Therefore, no discussion of weld metal results was included in WCAP-11955. | | : 2. Assuming there was no weld metal in Capsule z, then how was RTNDT of the weld metal calculated as shown on Pages A-7 and A-8. It appears that Westinghouse took the weld metal in Capsule Y (WCAP-11554) to be used in the P/T curves in WCAP-11955. Explain. |
| : 2. Assuming there was no weld metal in Capsule z, then how was RTNDT of the weld metal calculated as shown on Pages A-7 and A-8. It appears that Westinghouse took the weld metal in Capsule Y (WCAP-11554) to be used in the P/T curves in WCAP-11955. | | : 3. Table A-1 showed three weld metals. Which weld metal is the limiting one? That is, which weld metal was used in the calculations on Pages A-7 and A-8? |
| Explain. 3. Table A-1 showed three weld metals. Which weld metal is the limiting one? That is, which weld metal was used in the calculations on Pages A-7 and A-8? R2 & R3. The lower shell longitudinal weld seam 2-042C located on a 45° azimuth was considered the limiting weld material as indicated on Page A-7 of WCAP-11955. | | R2 & R3. The lower shell longitudinal weld seam 2-042C located on a 45° azimuth was considered the limiting weld material as indicated on Page A-7 of WCAP-11955. A chemical content of 0.35 cu and 1.00 Ni was used in developing t~g hea~up/cooldown (H/C) curves. A fluence of 7.68 X 10 n/cm at 15 EFPY for a weld located at a 45° azimuth from table 6-14 of WCAP 11955 was used for H/C curves. |
| A chemical content of 0.35 cu and 1.00 Ni was used in developing (H/C) curves. A fluence of 7.68 X 10 n/cm at 15 EFPY for a weld located at a 45° azimuth from table 6-14 of WCAP 11955 was used for H/C curves. 4. It appears that the chemistry factor of 212°F as shown on Page A-7 was obtained by using the copper and nickel contents of the weld metal. However, we could not determine how the chemistry factor of 158.9°F for the based plate was obtained. | | : 4. It appears that the chemistry factor of 212°F as shown on Page A-7 was obtained by using the copper and nickel contents of the weld metal. However, we could not determine how the chemistry factor of 158.9°F for the based plate was obtained. Regulatory Guide 1.99, Rev. |
| Regulatory Guide 1.99, Rev. 2, provides an alternative method (Section 2.1) to calculate the chemistry factor when two or more credible surveillance data sets become available. | | 2, provides an alternative method (Section 2.1) to calculate the chemistry factor when two or more credible surveillance data sets become available. |
| Explain how both chemistry factors (weld and base plate) were obtained. | | Explain how both chemistry factors (weld and base plate) were obtained. |
| *
| | |
| * R4. A chemistry factor of 272°F was obtained for weld metal from Table 1 of Regulatory Guide 1.99, Revision 2 for Cu 0.35 and 1.00 Ni. The base metal chemistry factor of 158.9°F was obtained from surveillance data for plate B2402-1 as follows: (OF) Capsule Flue2'ce n/cm RTNDT ff RTNDT x ff (OF) f f 2 T 0.24 x 1019 100 0.6138 61. 4 0.377 z 1. 33 x 1019 170 1.079 183.4 1.164 244.8 1.541 Chemistry factor = 244.8/1.541
| | R4. |
| = 158.9 ff -fluence factor 5. Figure A-2 in WCAP-11955 (corresponding to Figure 3.4-2 in the 12/28/88 submittal). | | A chemistry factor of 272°F was obtained for weld metal from Table 1 of Regulatory Guide 1.99, Revision 2 for Cu 0.35 and 1.00 Ni. |
| R5. Provide detailed/hand calculations to show that when RTNDT of 162°F is used in calculating P/T heatup curve it is more conservative than if RTNDT of 222.5°F is used. Following RTNDT values were calculated for 15EFPY (Page A-8 of WCAP-ll955) . Component EFPY Location ART Plate B2402-1 15 1/4T 204°F Plate B2402-l 15 3/4T 162°F Weld 15 1/4T 222.5°F Weld 15 3/4T 151. 5°F From the above table it is seen that for the 1/4T location, the limiting material is the weld and for the 3/4T location, the limiting material is the plate. Instead of generating two sets of heatup and cooldown curves, a conservative approach is taken by using 0 0 222.5 F RTNDT at the 1/4T and 162 F RTNDT at the 3/4T. This approacfi will envelop both the (B2402-1) and the weld. The Unit 1 Technical Specification heatup and cooldown curves submitted in LCR 88-14 have been revised to indicate both the 1/4T and 3/4T RTNDT values and are submitted herein. 6. Table 4-1 and Table A-1 show inconsistent copper and nickel contents for the base plate B2402-1 and weld metal. Explain. What are the exact percentages of copper and nickel contents used in the calculation of RTNDT of the B2402-1 plate and weld metal.
| | The base metal chemistry factor of 158.9°F was obtained from surveillance data for plate B2402-1 as follows: |
| *
| | Flue2'ce RTNDT (OF) 2 Capsule n/cm ff RTNDT x ff (OF) ff T 0.24 x 1019 100 0.6138 61. 4 0.377 z 1. 33 x 1019 170 1.079 183.4 1.164 244.8 1.541 Chemistry factor = 244.8/1.541 = 158.9 ff - fluence factor |
| * R6. Table 4-1 reports the copper and nickel values obtained for the Salem surveillance materials.
| | : 5. Figure A-2 in WCAP-11955 (corresponding to Figure 3.4-2 in the 12/28/88 submittal). |
| Table A-1 identifies the original chemistry reported for the plates and forgings used in the vessel. In this weld metal a chemical composition of 0.35 Cu and 1.00 Ni was used. In the base metal see the response identified in A-4. B. Other Questions | | Provide detailed/hand calculations to show that when RTNDT of 162°F is used in calculating P/T heatup curve |
| | ~ha~ it is more conservative than if RTNDT of 222.5°F is used. |
| | R5. Following RTNDT values were calculated for 15EFPY (Page A-8 of WCAP-ll955) . |
| | Component EFPY Location ART Plate B2402-1 15 1/4T 204°F Plate B2402-l 15 3/4T 162°F Weld 15 1/4T 222.5°F Weld 15 3/4T 151. 5°F From the above table it is seen that for the 1/4T location, the limiting material is the weld and for the 3/4T location, the limiting material is the plate. |
| | Instead of generating two sets of heatup and cooldown curves, a conservative approach is taken by using 0 0 222.5 F RTNDT at the 1/4T and 162 F RTNDT at the 3/4T. |
| | This approacfi will envelop both the pla~e (B2402-1) and the weld. The Unit 1 Technical Specification heatup and cooldown curves submitted in LCR 88-14 have been revised to indicate both the 1/4T and 3/4T RTNDT values and are submitted herein. |
| | : 6. Table 4-1 and Table A-1 show inconsistent copper and nickel contents for the base plate B2402-1 and weld metal. Explain. What are the exact percentages of copper and nickel contents used in the calculation of RTNDT of the B2402-1 plate and weld metal. |
| | |
| | R6. |
| | Table 4-1 reports the copper and nickel values obtained for the Salem surveillance materials. Table A-1 identifies the original chemistry reported for the plates and forgings used in the vessel. In this weld metal a chemical composition of 0.35 Cu and 1.00 Ni was used. In the base metal see the response identified in A-4. |
| | B. Other Questions |
| : 1. 10 CFR 50, Appendix G, Section IV.2 specifies pressure and temperature requirements for the vessel closure flange region. Discuss whether Salem 1 and 2 P/T limits satisfy the requirements. | | : 1. 10 CFR 50, Appendix G, Section IV.2 specifies pressure and temperature requirements for the vessel closure flange region. Discuss whether Salem 1 and 2 P/T limits satisfy the requirements. |
| R1. Salem Unit 1 and 2 P/T limits satisfy the requirements of 10CFR50 Appendix G, Section IV.2 as noted on page A-5 of WCAP-11955 for Unit 1 and page A-6 of WCAP-11554 for Unit 2. 2. The the and was for Salem Unit 2 P/T curves (Figure 3.4-2 and 3.4-3) in 12/28/88 submittal show two RTNDT values at 1/4T 3/4T locations. | | R1. Salem Unit 1 and 2 P/T limits satisfy the requirements of 10CFR50 Appendix G, Section IV.2 as noted on page A-5 of WCAP-11955 for Unit 1 and page A-6 of WCAP-11554 for Unit 2. |
| We could not aeeerrnine which RTNDT used for the heatup curve and which RTNDT was usea the cooldown curve. Discuss. R2. The following RTNDT values were used to develop the heatup and cooldown curves for Unit 2. Component Long. Weld Long. Weld 10 10 Location 1/4T 3/4T ART 178.6°F 116.1°F RTN of 178.6°F at 1/4T location is used to develop coo£down curves and RTNDT of 178.6°F (at 1/4T), 116.1°F (at 3/4T) are used to aevelop heatup curves. 3. Table B 3/4.4-1 of the Salem Unit 2 Technical Specification revision in the 12/28/88 submittal: | | : 2. The Salem Unit 2 P/T curves (Figure 3.4-2 and 3.4-3) in the 12/28/88 submittal show two RTNDT values at 1/4T and 3/4T locations. We could not aeeerrnine which RTNDT was used for the heatup curve and which RTNDT was usea for the cooldown curve. Discuss. |
| The limiting longitudinal weld in Figure 3.4-2 and 3.4-3 of the Salem Unit 2 Technical Specification revision show the copper content of 0.35% and nickel content 1.00%. This chemical composition is inconsistent with the weld metals in table B 3/4.4-1. | | R2. The following RTNDT values were used to develop the heatup and cooldown curves for Unit 2. |
| * R3. Although the chemical composition for Salem Unit 2 weld metal identified in Table B 3/4.4-1 of the Technical Specification is inconsistent with Figures 3.4-2 and 3.4-3, a more conservative chemical content (0.35% Cu and 1.00% Ni) was used for the H/C curves. | | Component Location ART Long. Weld 10 1/4T 178.6°F Long. Weld 10 3/4T 116.1°F RTN of 178.6°F at 1/4T location is used to develop coo£down curves and RTNDT of 178.6°F (at 1/4T), 116.1°F (at 3/4T) are used to aevelop heatup curves. |
| ..... <.> -Cll Q. -w a: Cll Cll w a: Q. Q w -u -Q z -Material Pr.rty Initial RTNDT' -56 F | | : 3. Table B 3/4.4-1 of the Salem Unit 2 Technical Specification revision in the 12/28/88 submittal: The limiting longitudinal weld in Figure 3.4-2 and 3.4-3 of the Salem Unit 2 Technical Specification revision show the copper content of 0.35% and nickel content 1.00%. |
| * RTNDT After 15 EFPY: 2500 2250 2000 1750 1500 1250 1000 i-. CooldCMI 750
| | This chemical composition is inconsistent with the weld metals in table B 3/4.4-1. |
| * Rates .. *F/Hr .. .. o-.. 500 250 .. .. .. 6 j-"""" .. 50 100 l/4T = 222.5°F 3/4T = 162°F URiCClt)tible Operation
| | * R3. Although the chemical composition for Salem Unit 2 weld metal identified in Table B 3/4.4-1 of the Technical Specification is inconsistent with Figures 3.4-2 and 3.4-3, a more conservative chemical content (0.35% Cu and 1.00% Ni) was used for the H/C curves. |
| ,,,, . .... .... , _.__ -,., ,,,. '-"'" '-... ... _..__ ,,. ..,.. i--150 200 I I . ' , .. ... '/' V.1 'fT I"/'/, "'/OIJ' ,,,-/ AccepUble Operation 250 300 I I j I I 350 INDICATED TEMPERATURE (DEG.F) CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS I 7 I I Figure 3.4-3 Salem Unit 1 Reactor Coolant System Cooldown Limitations Applicable for Cooldown Rates up to 100°F/HR for the Service Period up to 15 EFPY SALEM -UNIT 1 3/4 4-27 500 _________________
| | |
| _ | | Material Pr.rty B:sis*-- |
| ' . | | Initial RTNDT' RTNDT After 15 EFPY: |
| * Initial RT NOT: 45 ° F RT NOT After lSEFPY: 2500 2250 2000 1750 3/4T, 162°F l/4T, 222.5°F r r I I I I I ! I I f I I I r I I r ! Leak Test I Limit I I Heetup
| | -56 F l/4T 3/4T |
| -I 1500 111 a. .._ 1250 111 111 lo.I f 1000 Q lo.I ... c u 750 Q z 500 250 00 50 100 Up To I 61 l *F /Hr" I I ) I Unacceptable | | = |
| / Operation f/ f_; D I/ -.... Acceptible Operation 150 200 250 300 I I I I I I I I I 7 I I I 7 I . 1orft1calit)
| | = |
| L 1111ft Bu--* Jn lnservfcu Hydrostitfc Test Tenp. (JJ;*F) fo;-the Servic. Period Up
| | --~-~--' |
| * 15 EFPY I I I I I I I I 350 400 INOICATED TEMPERATURE (DEG.r) CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS 450 Fiqure 3.4-2 | | 222.5°F 162°F 2500 I |
| * Salem Unit 1 Reactor Coolant System Heatup Limitations Applicable for Heatup Rates up to 60°F/HR for the Service Period up to 15 EFPY SALEM -UNIT l 3/4 4-26 I I 500}} | | 7 2250 I |
| | I 2000 I |
| | I 1750 j I |
| | ..... 1500 I Cll Q. I I |
| | w a: |
| | ~ |
| | Cll Cll 1250 w |
| | a: ,' |
| | 1000 .. |
| | Q. |
| | URiCClt)tible u |
| | Q w |
| | ~ |
| | 750 .* |
| | i-Operation I"/'/, |
| | '/' |
| | V.1 'fT Q |
| | z CooldCMI Rates |
| | *F/Hr _.__ |
| | ~ . |
| | ~ |
| | ,,,, "'/OIJ' |
| | ,,,- / |
| | 500 .. - ,., |
| | .... o- '- ... ... |
| | '-"'" ~ ~ |
| | . ~ |
| | ..,.. AccepUble 250 . |
| | 6 j-"""" i-- |
| | Operation |
| | ~ |
| | 50 100 150 200 250 300 350 500 INDICATED TEMPERATURE (DEG.F) |
| | CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS Figure 3.4-3 Salem Unit 1 Reactor Coolant System Cooldown Limitations Applicable for Cooldown Rates up to 100°F/HR for the Service Period up to 15 EFPY SALEM - UNIT 1 3/4 4-27 |
| | ~: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ |
| | |
| | '. ~ |
| | Initial RT NOT: 45 °F -~----------==------=-=-=--=-=--~- |
| | RT NOT After lSEFPY: 3/4T, 162°F l/4T, 222.5°F 2500 r r I I I I I ! I I I I f I I I r I I r ! I I I |
| | Leak Test I I Limit I I 2250 I I I I 7 I |
| | 2000 I |
| | I 1750 I 7 |
| | - I I Heetup ~ates |
| | ~ 1500 Up To 61 l *F/Hr" I . |
| | 111 a. |
| | .._ I |
| | ~ 1250 I |
| | ~ |
| | 111 ) |
| | 111 lo.I f 1000 I Q |
| | Unacceptable / |
| | ...c lo.I Operation f/ |
| | u 750 f_; |
| | Q D z I/ |
| | 500 |
| | .... - 1orft1calit) |
| | L1111ft Bu--* |
| | ~ |
| | Jn lnservfcu Acceptible Hydrostitfc Operation Test Tenp. |
| | 250 (JJ;*F) fo;- |
| | the Servic. |
| | Period Up |
| | * 15 EFPY I I I I I I I I 00 50 100 150 200 250 300 350 400 450 500 INOICATED TEMPERATURE (DEG.r) |
| | CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS Fiqure 3.4-2 |
| | * Salem Unit 1 Reactor Coolant System Heatup Limitations Applicable for Heatup Rates up to 60°F/HR for the Service Period up to 15 EFPY SALEM - UNIT l 3/4 4-26}} |
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Forwards Discharge Monitoring Rept for May 1999,containing Info as Required by Permit NJ0005622.Rept Prepared Specifically for EPA & Nj Dept of Environ Protection
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
ML18107A3371999-05-21021 May 1999
Forwards NPDES Discharge Monitoring for Salem Generating Station for Apr 1999, Containing Info as Required by Permit NJ0005622
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
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
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
1999-09-08
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML18107A5601999-10-18018 October 1999
Submits 30-day Fuel Clad Temp Rept,Iaw 10CFR50.46.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Salem Generating Station Large & Small Break LOCA Analyses
ML18107A5531999-10-0808 October 1999
Forwards Summary Rept of Plant Startup & Power Ascension Testing for Sgs,Unit 2 Cycle 11,per Requirements of TS 6.9.1.1
ML18107A5561999-10-0707 October 1999
Requests Relief Associated with Containment Examinations at Hope Creek & Salem Generating Stations.Attachment 1 Includes Proposed Alternatives & Supporting Justification for Relief Requests
ML18107A5501999-10-0505 October 1999
Provides Current Status of Pse&G Actions Re GL 98-01, Y2K Readiness of Computer Sys at Npps, for Salem Nuclear Generating Station,Units 1 & 2 & Hope Creek Nuclear Generating Station
ML18107A5521999-10-0505 October 1999
Encourages NRC to Support Abb Combustion Engineering Nuclear Power Request for Priority Review of Generic TR Re Crossflow Ultrasonic Flow Measurement Sys
ML18107A5591999-10-0505 October 1999
Informs That Nj Dept of Environ Protection Has No Comments on License Change Request S99-07 for Sgs,Units 1 & 2
ML18107A5341999-09-22022 September 1999
Provides Data Re Operator Licensing Exam for Salem & Hope Creek Station,In Response to NRC Form 536 (7-1999)
ML20212B3631999-09-14014 September 1999
Forwards Rev 13 to Salem - Hope Creek Security Plan,Iaw 10CFR50.54(p).Summary of Proposed Changes to Plan,Encl. Encl Withheld
ML18107A5321999-09-13013 September 1999
Forwards Revised 10CFR50.92 Evaluation to Clarify Util Response to Question Number 1 Re Amend to Modify TS 3/4 8.1, AC Power Sources.
ML18107A5331999-09-13013 September 1999
Provides Notification That PSEG Intends to Utilize ASME Code Case N-481 During Second ISI Interval at Sgs Units 1 & 2
ML18107A5351999-09-13013 September 1999
Informs That NRC Has Reviewed Pse&G Request Proposing to Modify TS Which Allow EDG to Be Operated for 24 Surveillance Test During Any Mode,Iaw 10CFR50.91(b) & Has No Comments
ML18107A5231999-09-0808 September 1999
Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1999. 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
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
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
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
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
[Table view] |
Text
Public Service Electric and Gas Company Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4800 Vice President - Nuclear Operations July 31, 1989 NLR-N89139 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR ADDITIONAL INFORMATION PRESSURE TEMPERATURE LIMITS SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 The purpose of this letter is to provide PSE&G's responses to the questions forwarded in the request for additional information dated May 26, 1989. The attached responses were discussed with Mr. Tsao of the NRC staff on July 24, 1989. Figures 3.4-2 and 3.4-3 for Unit 1 were revised as a result of this conversation and are attached for incorporation into LCR 88-14.
Should you have any questions, please feel free to contact us.
Sincerely, Attachment 89-oaoso2:30--89o731 --.----.
PDR ADOCK 05000272 P PNU I I
Document Control Desk NLR-N89139 2
- 7-31-89 C Mr. J. c. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625
- ATTACHMENT QUESTIONS & RESPONSES - PRESSURE/TEMPERATURE LIMITS (LCR 88-14)
A. WCAP-11955 describes the test results of specimens in Capsule Z at Salem 1. The following questions relate to WCAP-11955 which provides the technical bases for the 12/28/88 submittal for Salem, Unit 1:
- 1. There was no discussion of weld metal in Section 4, Description of Program, in WCAP-11955. However, Table 4-1 discusses weld metal from Capsule Y. Table 4-2 and Table A-1 discusses a certain weld metal but we could not determine whether that weld metal was included in Capsule z. Was there a weld metal included in Capsule Z?
Rl. Salem Unit 1 Capsule z did not contain any weld metal specimens. Therefore, no discussion of weld metal results was included in WCAP-11955.
- 2. Assuming there was no weld metal in Capsule z, then how was RTNDT of the weld metal calculated as shown on Pages A-7 and A-8. It appears that Westinghouse took the weld metal in Capsule Y (WCAP-11554) to be used in the P/T curves in WCAP-11955. Explain.
- 3. Table A-1 showed three weld metals. Which weld metal is the limiting one? That is, which weld metal was used in the calculations on Pages A-7 and A-8?
R2 & R3. The lower shell longitudinal weld seam 2-042C located on a 45° azimuth was considered the limiting weld material as indicated on Page A-7 of WCAP-11955. A chemical content of 0.35 cu and 1.00 Ni was used in developing t~g hea~up/cooldown (H/C) curves. A fluence of 7.68 X 10 n/cm at 15 EFPY for a weld located at a 45° azimuth from table 6-14 of WCAP 11955 was used for H/C curves.
- 4. It appears that the chemistry factor of 212°F as shown on Page A-7 was obtained by using the copper and nickel contents of the weld metal. However, we could not determine how the chemistry factor of 158.9°F for the based plate was obtained. Regulatory Guide 1.99, Rev.
2, provides an alternative method (Section 2.1) to calculate the chemistry factor when two or more credible surveillance data sets become available.
Explain how both chemistry factors (weld and base plate) were obtained.
R4.
A chemistry factor of 272°F was obtained for weld metal from Table 1 of Regulatory Guide 1.99, Revision 2 for Cu 0.35 and 1.00 Ni.
The base metal chemistry factor of 158.9°F was obtained from surveillance data for plate B2402-1 as follows:
Flue2'ce RTNDT (OF) 2 Capsule n/cm ff RTNDT x ff (OF) ff T 0.24 x 1019 100 0.6138 61. 4 0.377 z 1. 33 x 1019 170 1.079 183.4 1.164 244.8 1.541 Chemistry factor = 244.8/1.541 = 158.9 ff - fluence factor
- 5. Figure A-2 in WCAP-11955 (corresponding to Figure 3.4-2 in the 12/28/88 submittal).
Provide detailed/hand calculations to show that when RTNDT of 162°F is used in calculating P/T heatup curve
~ha~ it is more conservative than if RTNDT of 222.5°F is used.
R5. Following RTNDT values were calculated for 15EFPY (Page A-8 of WCAP-ll955) .
Component EFPY Location ART Plate B2402-1 15 1/4T 204°F Plate B2402-l 15 3/4T 162°F Weld 15 1/4T 222.5°F Weld 15 3/4T 151. 5°F From the above table it is seen that for the 1/4T location, the limiting material is the weld and for the 3/4T location, the limiting material is the plate.
Instead of generating two sets of heatup and cooldown curves, a conservative approach is taken by using 0 0 222.5 F RTNDT at the 1/4T and 162 F RTNDT at the 3/4T.
This approacfi will envelop both the pla~e (B2402-1) and the weld. The Unit 1 Technical Specification heatup and cooldown curves submitted in LCR 88-14 have been revised to indicate both the 1/4T and 3/4T RTNDT values and are submitted herein.
- 6. Table 4-1 and Table A-1 show inconsistent copper and nickel contents for the base plate B2402-1 and weld metal. Explain. What are the exact percentages of copper and nickel contents used in the calculation of RTNDT of the B2402-1 plate and weld metal.
R6.
Table 4-1 reports the copper and nickel values obtained for the Salem surveillance materials. Table A-1 identifies the original chemistry reported for the plates and forgings used in the vessel. In this weld metal a chemical composition of 0.35 Cu and 1.00 Ni was used. In the base metal see the response identified in A-4.
B. Other Questions
- 1. 10 CFR 50, Appendix G, Section IV.2 specifies pressure and temperature requirements for the vessel closure flange region. Discuss whether Salem 1 and 2 P/T limits satisfy the requirements.
R1. Salem Unit 1 and 2 P/T limits satisfy the requirements of 10CFR50 Appendix G, Section IV.2 as noted on page A-5 of WCAP-11955 for Unit 1 and page A-6 of WCAP-11554 for Unit 2.
- 2. The Salem Unit 2 P/T curves (Figure 3.4-2 and 3.4-3) in the 12/28/88 submittal show two RTNDT values at 1/4T and 3/4T locations. We could not aeeerrnine which RTNDT was used for the heatup curve and which RTNDT was usea for the cooldown curve. Discuss.
R2. The following RTNDT values were used to develop the heatup and cooldown curves for Unit 2.
Component Location ART Long. Weld 10 1/4T 178.6°F Long. Weld 10 3/4T 116.1°F RTN of 178.6°F at 1/4T location is used to develop coo£down curves and RTNDT of 178.6°F (at 1/4T), 116.1°F (at 3/4T) are used to aevelop heatup curves.
- 3. Table B 3/4.4-1 of the Salem Unit 2 Technical Specification revision in the 12/28/88 submittal: The limiting longitudinal weld in Figure 3.4-2 and 3.4-3 of the Salem Unit 2 Technical Specification revision show the copper content of 0.35% and nickel content 1.00%.
This chemical composition is inconsistent with the weld metals in table B 3/4.4-1.
- R3. Although the chemical composition for Salem Unit 2 weld metal identified in Table B 3/4.4-1 of the Technical Specification is inconsistent with Figures 3.4-2 and 3.4-3, a more conservative chemical content (0.35% Cu and 1.00% Ni) was used for the H/C curves.
Material Pr.rty B:sis*--
Initial RTNDT' RTNDT After 15 EFPY:
-56 F l/4T 3/4T
=
=
--~-~--'
222.5°F 162°F 2500 I
7 2250 I
I 2000 I
I 1750 j I
..... 1500 I Cll Q. I I
w a:
~
Cll Cll 1250 w
a: ,'
1000 ..
Q.
URiCClt)tible u
Q w
~
750 .*
i-Operation I"/'/,
'/'
V.1 'fT Q
z CooldCMI Rates
~ .
~
,,,, "'/OIJ'
,,,- /
500 .. - ,.,
.... o- '- ... ...
'-"'" ~ ~
. ~
..,.. AccepUble 250 .
6 j-"""" i--
Operation
~
50 100 150 200 250 300 350 500 INDICATED TEMPERATURE (DEG.F)
CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS Figure 3.4-3 Salem Unit 1 Reactor Coolant System Cooldown Limitations Applicable for Cooldown Rates up to 100°F/HR for the Service Period up to 15 EFPY SALEM - UNIT 1 3/4 4-27
~: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
'. ~
Initial RT NOT: 45 °F -~----------==------=-=-=--=-=--~-
RT NOT After lSEFPY: 3/4T, 162°F l/4T, 222.5°F 2500 r r I I I I I ! I I I I f I I I r I I r ! I I I
Leak Test I I Limit I I 2250 I I I I 7 I
2000 I
I 1750 I 7
- I I Heetup ~ates
~ 1500 Up To 61 l *F/Hr" I .
111 a.
.._ I
~ 1250 I
~
111 )
111 lo.I f 1000 I Q
Unacceptable /
...c lo.I Operation f/
u 750 f_;
Q D z I/
500
.... - 1orft1calit)
L1111ft Bu--*
~
Jn lnservfcu Acceptible Hydrostitfc Operation Test Tenp.
250 (JJ;*F) fo;-
the Servic.
Period Up
- 15 EFPY I I I I I I I I 00 50 100 150 200 250 300 350 400 450 500 INOICATED TEMPERATURE (DEG.r)
CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS Fiqure 3.4-2
- Salem Unit 1 Reactor Coolant System Heatup Limitations Applicable for Heatup Rates up to 60°F/HR for the Service Period up to 15 EFPY SALEM - UNIT l 3/4 4-26
