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| | issue date = 07/14/1995 | | | issue date = 07/14/1995 |
| | title = LER 95-010-00:on 950615,RHR Pumps for long-term Flow Requirements for Both Units Declared Inoperable Due to RHR Flow Instrument Uncertainties.Evaluated EOPs.W/950714 Ltr | | | title = LER 95-010-00:on 950615,RHR Pumps for long-term Flow Requirements for Both Units Declared Inoperable Due to RHR Flow Instrument Uncertainties.Evaluated EOPs.W/950714 Ltr |
| | author name = PASTVA M J, SUMMERS J C | | | author name = Pastva M, Summers J |
| | 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:-. e Public_ Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit July 14, 1995 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT NO. 95-010-00 This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73(a) | | {{#Wiki_filter:e OPS~G Public_ Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit July 14, 1995 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT NO. 95-010-00 This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73(a) (2) (v) (A). |
| (2) (v) (A). SORC Mtg. 95-077 MJPJ:vs C Distribution LER File .'"' ,..... ::; .. '! ,.) 9507260220 950714 PDR ADOCK 05000272 S PDR Thl' is in \"t.!llr J. C. Summers General Manager -Salem Operations 95-2168 REV. 6194 NRG FORM 366 . U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5*92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS 1-ICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO (See reverse for required number of digits/characters for each block) THE PAPERWORK REDUCTION PROJECT (315().()104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3) Salem Generating Station Unit 1 05000 272 1 OFS TITLE 1 4 1 Inoperability. | | J. C. Summers General Manager - |
| Of Both Units' Residual Heat Removal--(RHR} P'umps For Long-Term* | | Salem Operations SORC Mtg. 95-077 MJPJ:vs C Distribution LER File |
| Flow -Reauirements Due To RHR Flow !ni::ti-11n anf-Tln,.1>i-t-'3inf--f | | ~ .'"' ,..... ::; ..'! ,.) |
| "" EVENT DATE (5) LER NUMBER (6 REPORT NUMBER (7} OTHER FACILITIES INVOLVED (8) SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR 05000 311 NUMBER NUMBER c::,,1 n.,; .. ? FACILITY NAME DOCKET NUMBER 06 15 95 95 --010. --00 07 14 95 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §; (Check one or more I (11) MODE (9) 5 20.402(b) 20.405(c) 50.73(a)(2) (Iv) 73.r,1 (b) I I 0 I 20.405(a)(1)(l) 50.36(c)(1) | | 9507260220 950714 PDR ADOCK 05000272 S PDR Thl' p~\\*cr is in \"t.!llr h.md~. |
| IX 50.73(a)(2)(v) 73.i1(c) 20.405(a)(1)(li) 50.36(c)(2) 50.73(a)(2)(vii)
| | 95-2168 REV. 6194 |
| OTHER -20.405(a)(1)(iii) 50.73(a) (2) (i) 50.73(a) (2) (viii) (A) (Specify in Abstracl 20.405(a)(1)(iv) 50.73(a) (2) (ii) 50.73(a)(2)(viii)(B) below and in Text, NRG Form 366A) 20.405 (a)(1 )(v) 50. 73 (a) (2) (iii) 50.73(a)(2)(x) | | |
| LICENSEE CONTACT FOR THIS LER 12) NAME TELEPHONE NUMBER (Include Area Code) M. J. Pastva Jr. LER Coordinator 609/339-5165 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORTC13l CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS TONPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR I YES X (If yes, complete EXPECTED SUBMISSION DATE) NO SUBMISSION DATE (15) 10 31 95 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) At 0745 hours on 6/15/95, the NRC was notified of a potential problem with the Residual Heat Removal (RHR) flow orifices installation and the potential for inaccurate operation of the RHR pump recirculation valves, which provide pump protection. | | NRG FORM 366 . U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5*92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD 1-ICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (315().()104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. |
| Testing on 6/30/95 satisfactorily demonstrated the operability and control function of the RHR flow orifices and the RHR pump recirculation valves. However, on 7/10/95 a potential design basis concern was identified that the 1000 gpm minimum continuous (long-term) flow requirements of the RHR pumps could not be assured. Subsequently, it was determined on 7/12/95 the concern with the RHR flow instrumentation included RHR operation per Emergency Operating Procedures (EOPs) to mitigate the effects of mainly a small break Loss of Coolant Accident scenario. | | FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3) |
| The cause of this occurrence is attributed to a design deficiency when the uncertainties for the RHR loop flow instruments were not accounted for. An investigation is in progress to determine | | Salem Generating Station Unit 1 05000 272 1 OFS 4 |
| *the cause(:;) | | TITLE 11 Inoperability. Of Both Units' Residual Heat Removal- -(RHR} P'umps For Long-Term* |
| for use of the incorrect instrument setpoints_ | | Flow - |
| Prior to subsequent entry of either Salem Unit into Mode 4, changes to plant operating procedures and EOPs will be evaluated, and appropriate changes will be implemented, as necessary. | | Reauirements Due To RHR Flow !ni::ti-11n anf- Tln,.1>i-t-'3inf--f "" |
| It is anticipated that, by 10/31/95 this report will be supplemented to further detail the root cause of this occurrence and any additional corrective actions. NRG FORM 366 (5-92) | | EVENT DATE (5) LER NUMBER (6 REPORT NUMBER (7} OTHER FACILITIES INVOLVED (8) |
| BLOCK NUMBER 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK NUMBER OF DIGITS/CHARACTERS TITLE UP TO 46 FACILITY NAME 8 TOTAL-DOC-KET NUMBER 3 IN ADDITION TO 05000 VARIES PAGE NUMBER UP TO 76 TITLE 6TOTAL 2 PER BLOCK EVENT DATE 7 TOTAL 2 FOR YEAR LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 2 PER BLOCK REPORT DATE UP TO 18 FACILITY NAME 8 TOTAL -DOCKET NUMBER OTHER FACILITIES INVOLVED
| | FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER c::,,1 ~~ n.,; .. ? 05000 311 FACILITY NAME DOCKET NUMBER 06 15 95 95 -- 010. -- 00 07 14 95 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §; (Check one or more I (11) |
| * 3 IN ADDITION TO 05000 1 OPERATING MODE . 3 POWER LEVEL 1 CHECK BOX THAT APPLIES REQUIREMENTS OF 10 CFR UP TO 50 FOR NAME 14 FOR TELEPHONE LICENSEE CONTACT CAUSE VARIES 2 FOR SYSTEM 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1 CHECK BOX THAT APPLIES SUPPL!=MENTAL REPORT EXPECTED 6 TOTAL 2 PER BLOCK EXPECTED SUBMISSION DATE - | | MODE (9) 5 20.402(b) 20.405(c) 50.73(a)(2) (Iv) 73.r,1 (b) |
| LICENSEE EVENT REPORT {LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 2 of 5 Unit # 1 50-272 95-010-00 Plant and System Identification: | | IX I L:~~~~O) I 0 I 20.405(a)(1)(l) 20.405(a)(1)(li) 50.36(c)(1) 50.36(c)(2) 50.73(a)(2)(v) 50.73(a)(2)(vii) 50.73(a) (2) (viii) (A) 73.i1(c) |
| Westinghouse | | OTHER (Specify in Abstracl 20.405(a)(1)(iii) 50.73(a) (2) (i) below and in Text, NRG 20.405(a)(1)(iv) 50.73(a) (2) (ii) 50.73(a)(2)(viii)(B) Form 366A) 20.405 (a)(1 )(v) 50. 73 (a) (2) (iii) 50.73(a)(2)(x) |
| -Pressurized Water Reactor Energy Industry Identification System (EIIS) codes appear in the text as {xx} Identification of Occurrence: | | LICENSEE CONTACT FOR THIS LER 12) |
| Inoperability Of Both Units' Residual Heat Removal (RHR) Pumps For Long-Term Flow Requirements Due To RHR Flow Instrument Uncertainties Event Date: June 15, 1995 Report Date: July 14, 1995 This report was initiated by Incident Report No. 95-873 Conditions Prior to Occurrence: | | NAME TELEPHONE NUMBER (Include Area Code) |
| Both Units were in a self-imposed extended shutdown. , Mode 5 Reactor Power % Unit Load Mwe Description of Occurrence: | | M. J. Pastva Jr. LER Coordinator 609/339-5165 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORTC13l REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TONPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR X |
| | I YES (If yes, complete EXPECTED SUBMISSION DATE) |
| | NO SUBMISSION DATE (15) 10 31 95 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) |
| | At 0745 hours on 6/15/95, the NRC was notified of a potential problem with the Residual Heat Removal (RHR) flow orifices installation and the potential for inaccurate operation of the RHR pump recirculation valves, which provide pump protection. Testing on 6/30/95 satisfactorily demonstrated the operability and control function of the RHR flow orifices and the RHR pump recirculation valves. However, on 7/10/95 a potential design basis concern was identified that the 1000 gpm minimum continuous (long-term) flow requirements of the RHR pumps could not be assured. Subsequently, it was determined on 7/12/95 the concern with the RHR flow instrumentation included RHR operation per Emergency Operating Procedures (EOPs) to mitigate the effects of mainly a small break Loss of Coolant Accident scenario. The cause of this occurrence is attributed to a design deficiency when the uncertainties for the RHR loop flow instruments were not accounted for. An investigation is in progress to determine *the cause(:;) for use of the incorrect instrument setpoints_ Prior to subsequent entry of either Salem Unit into Mode 4, changes to plant operating procedures and EOPs will be evaluated, and appropriate changes will be implemented, as necessary. It is anticipated that, by 10/31/95 this report will be supplemented to further detail the root cause of this occurrence and any additional corrective actions. |
| | NRG FORM 366 (5-92) |
| | |
| | REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER OF TITLE NUMBER DIGITS/CHARACTERS 1 UP TO 46 FACILITY NAME 8 TOTAL- - |
| | 2 DOC-KET NUMBER 3 IN ADDITION TO 05000 3 VARIES PAGE NUMBER 4 UP TO 76 TITLE 6TOTAL 5 EVENT DATE 2 PER BLOCK 7 TOTAL 2 FOR YEAR 6 LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 7 REPORT DATE 2 PER BLOCK UP TO 18 FACILITY NAME 8 OTHER FACILITIES INVOLVED 8 TOTAL - DOCKET NUMBER |
| | * 3 IN ADDITION TO 05000 9 1 OPERATING MODE . |
| | 10 3 POWER LEVEL 1 |
| | 11 REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1 |
| | 14 SUPPL!=MENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15 EXPECTED SUBMISSION DATE 2 PER BLOCK |
| | |
| | LICENSEE EVENT REPORT {LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 2 of 5 Unit # 1 50-272 95-010-00 Plant and System Identification: |
| | Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes appear in the text as {xx} |
| | Identification of Occurrence: |
| | Inoperability Of Both Units' Residual Heat Removal (RHR) |
| | Pumps For Long-Term Flow Requirements Due To RHR Flow Instrument Uncertainties Event Date: June 15, 1995 Report Date: July 14, 1995 This report was initiated by Incident Report No. 95-873 |
| | *:t,:.: |
| | Conditions Prior to Occurrence: |
| | . |
| | * Ji, Both Units were in a self-imposed extended shutdown. |
| | Mode 5 Reactor Power % Unit Load Mwe Description of Occurrence: |
| At 0745 hours on June 15, 1995, the NRC was notified, in accordance with 10CFR50. 72 (b) (2) (i), of a potential problem with the RHR flow orifices 1(2)FE641A/B installation and potential for inaccurate operation of the RHR pump recirculation valves, which provide pump protection. | | At 0745 hours on June 15, 1995, the NRC was notified, in accordance with 10CFR50. 72 (b) (2) (i), of a potential problem with the RHR flow orifices 1(2)FE641A/B installation and potential for inaccurate operation of the RHR pump recirculation valves, which provide pump protection. |
| Testing on June 30, 1995 satisfactorily demonstrated the operability and control function of the RHR flow orifices and the RHR pump recirculation valves. However, on July 10, 1995, a potential design basis concern was identified that the 1000 gpm minimum continuous (long-term) flow requirements of the RHR pumps could not be assured. This determination followed recognition of that instrument loop uncertainties were not factored into the RHR flow indication setpoints. | | Testing on June 30, 1995 satisfactorily demonstrated the operability and control function of the RHR flow orifices and the RHR pump recirculation valves. However, on July 10, 1995, a potential design basis concern was identified that the 1000 gpm minimum continuous (long-term) flow requirements of the RHR pumps could not be assured. This determination followed recognition of that instrument loop uncertainties were not factored into the RHR flow indication setpoints. Subsequently, it was determined on July 12, 1995, the concern with the RHR flow instrumentation included RHR operation per Emergency Operating Procedures (EOPs) to |
| Subsequently, it was determined on July 12, 1995, the concern with the RHR flow instrumentation included RHR operation per Emergency Operating Procedures (EOPs) to *:t,:.: .* Ji, LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 3 of 5 Unit # 1 50-272 95-010-00 Description of Occurrence: (cont'd) mitigate the effects of a small break Loss of Coolant Accident (LOCA) scenario. | | |
| The EOPs for steam generator tube rupture secondary side breaks and inadvertent safety injection (SI) are also of concern. Analysis of Occurrence: | | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 3 of 5 Unit # 1 50-272 95-010-00 Description of Occurrence: (cont'd) mitigate the effects of a small break Loss of Coolant Accident (LOCA) scenario. The EOPs for steam generator tube rupture secondary side breaks and inadvertent safety injection (SI) are also of concern. |
| Review of plant drawings associated with RHR pump discharge piping indicated a potential operability/design problem involving the installation of the minimum recirculation control orifices on both Salem Units. Testing determined this issue was not an operability/design concern. However, this testing identified a concern with the accuracy of the RHR minimum recirculation valve control setpoints, which potentially affects the RHR pump continuous service operation. | | Analysis of Occurrence: |
| Subsequently, this concern was expanded to include securing the RHR pumps, per EOP guidance. | | Review of plant drawings associated with RHR pump discharge piping indicated a potential operability/design problem involving the installation of the minimum recirculation control orifices on both Salem Units. |
| | Testing determined this issue was not an operability/design concern. However, this testing identified a concern with the accuracy of the RHR minimum recirculation valve control setpoints, which potentially affects the RHR pump continuous service operation. Subsequently, this concern was expanded to include securing the RHR pumps, per EOP guidance. |
| Apparent Cause of Occurrence: | | Apparent Cause of Occurrence: |
| The cause of this occurrence is attributed to Manufacturing/Construction", as classified in NUREG-1022, Appendix B. This occurred when the instrument uncertainties for the RHR loop flow instruments were not accounted for in establishing the instruments' setpoints on the Salem Units. The results of an ongoing investigation to determine the cause(s) for use of the incorrect instrument setpoints, including contributing factors, as well as failed or deficient controls and barriers will be reflected in a supplement to this report. Prior Similar Occurrence: | | The cause of this occurrence is attributed to ~Design, Manufacturing/Construction", as classified in NUREG-1022, Appendix B. This occurred when the instrument uncertainties for the RHR loop flow instruments were not accounted for in establishing the instruments' setpoints on the Salem Units. |
| | The results of an ongoing investigation to determine the cause(s) for use of the incorrect instrument setpoints, including contributing factors, as well as failed or deficient controls and barriers will be reflected in a supplement to this report. |
| | Prior Similar Occurrence: |
| Review of documentation did not reveal a prior similar occurrence. | | Review of documentation did not reveal a prior similar occurrence. |
| Safety Significance: | | Safety Significance: |
| This occurrence is reportable pursuant to 10CFR50.73(a) | | This occurrence is reportable pursuant to 10CFR50.73(a) (2) (v) (A). |
| (2) (v) (A). | | |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 4 of 5 Unit # 1 50-272 95-010-00 Safety Significance: (cont'd) Potential RHR Operation At Flow Less Than 1000 gpm. Incorporating RHR flow instrumentation inaccuracies obtained through testing into in the setpoint calculation, the lowest expected minimum flow rate at closure of the recirculation valve for continuous flow service of the RHR pumps is estimated to be above 800 gpm. In addition, the RHR pump vendor has evaluated that the pumps are suitable to operate continuously at a flow rate of 800 gpm. As such, no safety concern exists with RHR operation at the current setpoint of 1000 gpm. Review of EOPs for accident conditions indicate the RHR pumps would be stopped in less than 45 minutes from initiation of an accident, if Reactor Coolant System pressure is above the pump shutoff head pressure and not injecting into the RCS. In accordance with the EOPs, an RCS pressure comparison check is made on cold leg injection flow indication. | | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 4 of 5 Unit # 1 50-272 95-010-00 Safety Significance: (cont'd) |
| If the flow indication is less than 200 gpm indicated, the operator is instructed to stop the involved RHR pump. Testing results support that the maximum combined flow through the RHR pump at that time would be 550 gpm recirculation flow through the pump minimum flow valves plus the 200 gpm flow, plus or minus process and loop uncertainties, for a total flow of approximately 750 gpm. RHR flow injection starts when RHR discharge pressure exceeds RCS pressure, at approximately 350 psig. For a small break LOCA that results in RHR flow to the RCS, the flow will increase in response to decreasing RCS pressure . . As such, RHR flow lower than 800 gpm is not expected to occur for an extended duration. | | Potential RHR Operation At Flow Less Than 1000 gpm. |
| Consequently, damage to the pump, from the effects of low flow for this short duration is expected to be minimal. The reduced RHR flow is not a potential concern for large break LOCAs since RCS pressure will depressurize rapidly. Corrective Action: Prior to subsequent entry of both Salem Units into Mode 4: Changes to plant operating procedures and EOPs will be evaluated, and if necessary, appropriate changes will be implemented on each Salem Unit. | | Incorporating RHR flow instrumentation inaccuracies obtained through testing into in the setpoint calculation, the lowest expected minimum flow rate at closure of the recirculation valve for continuous flow service of the RHR pumps is estimated to be above 800 gpm. In addition, the RHR pump vendor has evaluated that the pumps are suitable to operate continuously at a flow rate of 800 gpm. As such, no safety concern exists with RHR operation at the current setpoint of 1000 gpm. |
| LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 5 of 5 Unit # 1 50-272 95-010-00 Corrective Action: (cont'd) It is anticipated that by October 31, 1995, a supplement to this report will be submitted to further detail the root cause of this occurrence, as well as any additional corrective actions identified. | | Review of EOPs for accident conditions indicate the RHR pumps would be stopped in less than 45 minutes from initiation of an accident, if Reactor Coolant System pressure is above the pump shutoff head pressure and not injecting into the RCS. In accordance with the EOPs, an RCS pressure comparison check is made on cold leg injection flow indication. If the flow indication is less than 200 gpm indicated, the operator is instructed to stop the involved RHR pump. Testing results support that the maximum combined flow through the RHR pump at that time would be 550 gpm recirculation flow through the pump minimum flow valves plus the 200 gpm flow, plus or minus process and loop uncertainties, for a total flow of approximately 750 gpm. |
| MJPJ:vs REEF: SORC Mtg. 95-077 J. C. Summers General Manager -Salem Operations}} | | RHR flow injection starts when RHR discharge pressure exceeds RCS pressure, at approximately 350 psig. For a small break LOCA that results in RHR flow to the RCS, the flow will increase in response to decreasing RCS pressure . |
| | . As such, RHR flow lower than 800 gpm is not expected to occur for an extended duration. Consequently, damage to the pump, from the effects of low flow for this short duration is expected to be minimal. The reduced RHR flow is not a potential concern for large break LOCAs since RCS pressure will depressurize rapidly. |
| | Corrective Action: |
| | Prior to subsequent entry of both Salem Units into Mode 4: |
| | Changes to plant operating procedures and EOPs will be evaluated, and if necessary, appropriate changes will be implemented on each Salem Unit. |
| | |
| | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 5 of 5 Unit # 1 50-272 95-010-00 Corrective Action: (cont'd) |
| | It is anticipated that by October 31, 1995, a supplement to this report will be submitted to further detail the root cause of this occurrence, as well as any additional corrective actions identified. |
| | ~~ |
| | J. C. Summers General Manager - |
| | Salem Operations MJPJ:vs REEF: SORC Mtg. 95-077}} |
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Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0081998-12-24024 December 1998 LER 97-001-01:on 970215,failure to Perform TS Surveillance of Component Cooling Water Sys Check Valves Occurred.Caused by Inadequate Communication Between EOP Group & IST Reviewers.Procedure Revised.With 981224 Ltr ML18106B0021998-12-17017 December 1998 LER 98-015-01:on 980924,improper Installation of Test Equipment to RPS Occurred.Caused by Inadequate 10CFR50.59 Applicability Reviews During Past Revs.Revised Procedures. with 981217 Ltr ML18106A9551998-11-0303 November 1998 LER 96-013-01:on 960711,concluded That Current Gain & Bias Settings Had Rendered Overtemperature Delta Temp Protection Channels Inoperable.Caused by Scaling Error.Licensee Will Revise Scaling Calculations.With 981105 Ltr ML18106A9451998-10-30030 October 1998 LER 97-004-01:on 970408,failure to Comply with TS Action Statement,Dg Start & Inadequate Surveillance Testing,Was Noted.Caused by Inadequate Tracking of Inoperable Equipment. Discussed Event & Lessons Learned.With 981022 Ltr ML18106A9491998-10-22022 October 1998 LER 98-015-00:on 980924,identified Improper Installation of Test Equipment to Rps.Cause Indeterminate.Procedures for Installation of Test Equipment for Collection of State Point Data Were Placed on Administrative Hold.With 981022 Ltr ML18106A9301998-10-21021 October 1998 LER 98-014-00:on 980725,noted Improper Calibr of Liquid Radwaste Effluent Line Radiation Monitor.Caused by Inattention to Detail by Maint Personnel.Channel Calibr Was Successfully Performed on 1R18 on 980821.With 981019 Ltr ML18106A9071998-10-0101 October 1998 LER 98-014-00:on 980918,discovered That Fire Barrier Matl for HVAC Ducts Does Not Meet Required Level of Fire Resistance.Cause Indeterminate.Established Appropriate Compensatory Actions for Fire Barriers.With 981001 Ltr ML18106A8951998-09-28028 September 1998 LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr ML18106A8821998-09-21021 September 1998 LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr ML18106A8791998-09-16016 September 1998 LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr ML18106A8801998-09-0808 September 1998 LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr ML18106A8531998-08-27027 August 1998 LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr ML18106A8521998-08-27027 August 1998 LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr ML18106A8421998-08-24024 August 1998 LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr ML18106A8431998-08-24024 August 1998 LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr ML18106A8141998-08-13013 August 1998 LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr ML18106A8201998-08-13013 August 1998 LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr ML18106A6931998-06-29029 June 1998 LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged ML18106A6471998-06-0404 June 1998 LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr ML18106A6421998-06-0101 June 1998 LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr ML18106A6431998-05-29029 May 1998 LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr ML18106A6141998-05-18018 May 1998 LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr ML18106A5901998-05-0101 May 1998 LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr ML18106A5611998-04-20020 April 1998 LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected ML18106A6061998-04-0101 April 1998 Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences ML18106A4451998-04-0101 April 1998 LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr ML18106A4351998-03-30030 March 1998 LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr ML18106A3961998-03-20020 March 1998 LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr ML18106A5781998-03-20020 March 1998 Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs ML18106A4021998-03-20020 March 1998 LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr ML18106A4031998-03-20020 March 1998 LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr 1999-08-26
[Table view] Category:RO)
MONTHYEARML18107A5031999-08-26026 August 1999 LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr ML18107A4691999-07-28028 July 1999 LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr ML18107A4441999-07-0606 July 1999 LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr ML18107A4211999-07-0202 July 1999 LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr ML18107A4321999-07-0101 July 1999 LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr ML18107A4331999-07-0101 July 1999 LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr ML18107A3951999-06-17017 June 1999 LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0081998-12-24024 December 1998 LER 97-001-01:on 970215,failure to Perform TS Surveillance of Component Cooling Water Sys Check Valves Occurred.Caused by Inadequate Communication Between EOP Group & IST Reviewers.Procedure Revised.With 981224 Ltr ML18106B0021998-12-17017 December 1998 LER 98-015-01:on 980924,improper Installation of Test Equipment to RPS Occurred.Caused by Inadequate 10CFR50.59 Applicability Reviews During Past Revs.Revised Procedures. with 981217 Ltr ML18106A9551998-11-0303 November 1998 LER 96-013-01:on 960711,concluded That Current Gain & Bias Settings Had Rendered Overtemperature Delta Temp Protection Channels Inoperable.Caused by Scaling Error.Licensee Will Revise Scaling Calculations.With 981105 Ltr ML18106A9451998-10-30030 October 1998 LER 97-004-01:on 970408,failure to Comply with TS Action Statement,Dg Start & Inadequate Surveillance Testing,Was Noted.Caused by Inadequate Tracking of Inoperable Equipment. Discussed Event & Lessons Learned.With 981022 Ltr ML18106A9491998-10-22022 October 1998 LER 98-015-00:on 980924,identified Improper Installation of Test Equipment to Rps.Cause Indeterminate.Procedures for Installation of Test Equipment for Collection of State Point Data Were Placed on Administrative Hold.With 981022 Ltr ML18106A9301998-10-21021 October 1998 LER 98-014-00:on 980725,noted Improper Calibr of Liquid Radwaste Effluent Line Radiation Monitor.Caused by Inattention to Detail by Maint Personnel.Channel Calibr Was Successfully Performed on 1R18 on 980821.With 981019 Ltr ML18106A9071998-10-0101 October 1998 LER 98-014-00:on 980918,discovered That Fire Barrier Matl for HVAC Ducts Does Not Meet Required Level of Fire Resistance.Cause Indeterminate.Established Appropriate Compensatory Actions for Fire Barriers.With 981001 Ltr ML18106A8951998-09-28028 September 1998 LER 98-012-01:on 980725,noted That Afs Was Operated with Less than Required Number of Operable AFW Pumps.Caused by Improper Procedure Implementation.Runout Protection Pressure Device for 22 AFW Pumps Was Returned to Svc.With 980928 Ltr ML18106A8821998-09-21021 September 1998 LER 98-013-00:on 980820,noted Surveillance of Containment Penetration Overcurrent Protection Devices Missed.Caused by Human Error.Satisfactorily Tested Apprpriate Breakers & Disciplined Involved Personnel.With 980921 Ltr ML18106A8791998-09-16016 September 1998 LER 96-006-01:on 960717,determined That non-radioactive Liquid Basin Radwaste Monitor Inoperable During Low Head Conditions.Caused by Inadequate Design Change Package.Design Change 1EC3663-01 Has Been Installed.With 980916 Ltr ML18106A8801998-09-0808 September 1998 LER 98-013-00:on 980806,operation with TS Required Equipment OOS Was Noted.Caused by Human Error.Reviewed Processes & Practices Re Safety Sys Status Control,Procedure Rev & Extra Training.With 980908 Ltr ML18106A8531998-08-27027 August 1998 LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr ML18106A8521998-08-27027 August 1998 LER 98-011-00:on 980803,ESFA During a 4KV Automatic Transfer Test Was Noted.Caused by Premature Release of Control Console Pushbutton Due to Inadequate Procedural Step.Revised procedure.W/980827 Ltr ML18106A8421998-08-24024 August 1998 LER 98-012-00:on 980725,discovered That Plant Had Operated in Modes 1 & 2 w/twenty-two AFW Pumps Inoperable.Caused by Failure to Restore Pump Runout Protection Pressure Device to Svc.Returned Subject Device to svc.W/980824 Ltr ML18106A8431998-08-24024 August 1998 LER 98-009-00:on 980810,failure to Post Continuous Firewatch as Required by Fire Protection Plan Noted.Caused by Failure to Recognize Concurrent Conditions.Continuous Firewatch Was Posted Immediately & Repaired Smoke detectors.W/980824 Ltr ML18106A8141998-08-13013 August 1998 LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr ML18106A8201998-08-13013 August 1998 LER 98-012-00:on 980715,potential to Exceed Rating of Piping Due to Isolation of Overpressure Protection Line Was Noted. Caused by Inadequate Procedural Guidance.Appropriate Operations Dept Procedures Have Been revised.W/980813 Ltr ML18106A6931998-06-29029 June 1998 LER 98-003-00:on 980122,inappropriate Plugging of Tubes R9C60 & R10C60 in Salem Unit 2 Sg,Was Performed.Caused by Failure of Qualification,Verification & Validation Process. Tubes Reviewed to Verify No Others Inappropriately Plugged ML18106A6471998-06-0404 June 1998 LER 98-011-00:on 980505,improper Isolation of Single Cell Battery Charger from 125 Vdc Battery Was Noted.Caused by Inadequate 10CFR50.59 Applicability Review.Placed Procedure SC.MD-CM.ZZ-0024(Q) on Administrative hold.W/980604 Ltr ML18106A6421998-06-0101 June 1998 LER 98-010-00:on 931019,reactor Pressure Vessel Insp Plugs Were Out of Configuration,Was Noted.Caused by Personnel Error.Proper Configuration Was Restored Shortly After Discovery Prior to Entering Mode 2.W/980601 Ltr ML18106A6431998-05-29029 May 1998 LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr ML18106A6141998-05-18018 May 1998 LER 98-008-00:on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised procedure.W/980518 Ltr ML18106A5901998-05-0101 May 1998 LER 98-009-00:on 980405,epoxy Missing from Terminals of H Analyzer Was Noted.Caused by Inadequate Development of Procedure.H Analyzers Were Repaired & Review of Other Safety Related Equipment in Containment Was performed.W/980501 Ltr ML18106A5611998-04-20020 April 1998 LER 98-008-00:on 980323,inadequate Testing of Salem Unit 1 Containment Air Locks Resulted in Entering TS 3.0.3.Caused by less-than-adequate Work Practices During Replacement of Equalizing Valve.Salem Unit 2 Airlocks Were Inspected ML18106A6061998-04-0101 April 1998 Corrected LER 98-004-00:on 980302,failure to Comply W/Tss 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Have Been Taken to Correctly Document Safety Factors.Corrects Prior Similar Occurrences ML18106A4451998-04-0101 April 1998 LER 98-004-00:on 980302,failure to Perform TS 4.11.1.1.2 & 3.3.3.8 Was Noted.Caused by Organizational Deficiency.Steps Were Taken to Correctly Document Safety factors.W/980401 Ltr ML18106A4351998-03-30030 March 1998 LER 98-006-00:on 980227,incorrect Scaling of First Stage Turbine Impulse Pressure Transmitters Noted.Cause Indeterminate.Implemented Procedure Changes & re-scaled Affected Turbine Impulse Pressure transmitters.W/980330 Ltr ML18106A3961998-03-20020 March 1998 LER 98-005-00:on 980219,inoperability of Twelve EDG Fuel Oil Transfer Pump (FOTP) Noted.Caused by Installation of Incorrect Control Switch.Installed Correct off-auto-manual Switch & Verified Operability of Twelve FOTP.W/980320 Ltr ML18106A5781998-03-20020 March 1998 Corrected LER 98-005-00:on 980219,inoperability of 12 Fuel Oil Transfer Pump (Fotp),Noted.Caused by Installation of Incorrect Control Switch.Field Insp Performed to Verify Configuration of Switches for 11,21 & 22 FOTPs ML18106A4021998-03-20020 March 1998 LER 98-007-00:on 980218,failure to Establish Containment Integrity (Closure) Prior to Fuel Movement Was Noted.Caused by Failure to Identify & Include Condensate Pot Vent in Appropriate Valve Lineup.Valves identified.W/980320 Ltr ML18106A4031998-03-20020 March 1998 LER 98-006-00:on 980221,ESF Actuation of 11 & 12 Auxiliary Feedwater Pumps Occurred.Caused by Human Error.Operators Promptly Established Feedwater to All SG & Restored Proper Water levels.W/980320 Ltr 1999-08-26
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML18107A5581999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Salem,Unit 2.With 991014 Ltr ML18107A5571999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Salem,Unit 1.With 991014 Ltr ML18107A5301999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Salem,Unit 2.With 990913 Ltr ML18107A5311999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Salem,Unit 1.With 990913 ML18107A5031999-08-26026 August 1999 LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr ML18107A5201999-08-12012 August 1999 Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#9) Second Interval,Second Period, First Outage (96RF). ML18107A4811999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Salem,Unit 1.With 990813 Ltr ML18107A4821999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Salem,Unit 2.With 990813 Ltr ML18107A4691999-07-28028 July 1999 LER 99-008-00:on 990714,determined That Limit Switch Cables Were Subject to Multiple Hot Shorts in Same Fire Area.Caused by Inadequate Original Post Fire Safe Shutdown Analysis.All Limit Switch Cables for MOVs Were Reviewed.With 990728 Ltr ML18107A4441999-07-0606 July 1999 LER 99-007-00:on 990605,surveillance for Quadrant Power Tilt Ratio (QPTR) Was Missed.Caused by Human Error.Qptr Calculation Was Performed & Personnel Involved Have Been Held Accountable IAW Pse&G Policies.With 990706 Ltr ML18107A4211999-07-0202 July 1999 LER 99-005-00:on 990605,11 Containment Declared Inoperable. Caused by Valves 11SW72 & 11SW223 Both Leaking.Procedure S1.OP-ST.SW-0010(Q) Was Enhanced to Provide Specific Instructions to Ensure Proper Sequencing.With 990702 Ltr ML18107A4331999-07-0101 July 1999 LER 99-002-01:on 990405,determined That 2SA118 Failed as Found Leakrate Test.Caused by Foreign Matl Found in 2SA118 valve.2SA118 Valve Was Cycled Several Times & Seat Area Was Air Blown in Order to Displace Foreign Matl.With 990701 Ltr ML18107A4321999-07-0101 July 1999 LER 99-006-01:on 990501,determined That There Was No Flow in One of Four Injection Legs.Caused by Sticking of Valve in Safety Injection Discharge Line to 21 Cold Leg.Valve Was Cut Out of Sys & Replaced.With 990701 Ltr ML18107A5211999-07-0101 July 1999 Rev 0 to Sgs Unit 2 ISI RFO Exam Results (S2RFO#10) Second Interval,Second Period,Second Outage (99RF). ML18107A4351999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Salem,Unit 1.With 990713 Ltr ML18107A4341999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Salem,Unit 2.With 990713 Ltr ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML18107A3951999-06-17017 June 1999 LER 99-004-00:on 990520,reactor Tripped from 100% Power,Due to Negative Flux Trip Signal from Nuclear Instrumentation. Cause Has Not Been Determined.Discoloration Was Identified on One of Penetrations.With 990617 Ltr ML18107A3661999-06-0909 June 1999 LER 99-003-00:on 990513,unplanned Entry Into TS 3.0.3 Was Made.Caused by Human error.Re-positioned Creacs Supply Fan Selector Switches & Revised Procedures S1 & S2.OP-ST.SSP-0001(Q).With 990609 Ltr ML18107A3551999-06-0202 June 1999 LER 99-005-00:on 990504,failure to Meet TS Action Statement Requirements for High Oxygen Concentration in Waste Gas Holdup Sys Occurred.Caused by Inability of Operators. Existing Procedures Will Be Evaluated.With 990602 Ltr ML18107A3441999-06-0101 June 1999 Interim Part 21 Rept Re Premature Over Voltage Protection Actuation in Circuit Specific Application in Dc Power Supply.Testing & Evaluation Activities Will Be Completed on 990716 ML18107A3541999-06-0101 June 1999 LER 99-006-00:on 990501,HHSI Flow Balance Discrepancy Was Noted During Surveillance.Caused by Sticking of Check Valve in SI Discharge Line to 21 Cold Leg.Valve 21SJ17,was Cut Out of Sys & Replaced.With 990601 Ltr ML18107A3681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 1.With 990611 Ltr ML18107A3721999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Salem Generating Station,Unit 2.With 990611 Ltr ML18107A2931999-05-12012 May 1999 LER 99-002-00:on 990413,determined That Number 12 Auxiliary Bldg Exhaust Fan Was Rotating Backwards.Caused by mis-wiring of Motor Due to Human Error by Maint technician.Mis-wiring Was Corrected & Fan Was Returned to Svc.With 990512 Ltr ML18107A2781999-05-10010 May 1999 LER 99-004-00:on 990411,automatic Actuation of ESF Occurred During Reactor Vessel Head Removal in Support of Refueling Operations.Caused by High Radiation Condition.Containment Atmosphere Was Monitored.With 990505 Ltr ML18107A2791999-05-0404 May 1999 LER 99-003-00:on 990406,all Salem Unit 2 Chillers Rendered Inoperable.Caused by Human Error.Lessons Learned from Event Were Communicated to All Operators by Including Them in Night Orders.With 990504 Ltr ML18107A2741999-05-0303 May 1999 LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr ML18107A3711999-04-30030 April 1999 Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1 ML18107A3151999-04-30030 April 1999 Submittal-Only Screening Review of Salem Generating Station Individual Plant Exam for External Events (Seismic Portion), Rev 1 ML18107A2991999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Salem Unit 1.With 990514 Ltr ML18107A2971999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Salem Unit 2.With 990514 Ltr ML18107A2351999-04-23023 April 1999 LER 99-001-00:on 990330,MSSV Failed Lift Set Test.Caused by Setpoint Variance Which Is Result of Aging.Valves Were Adjusted & Retested to Ensure TS Tolerance.With 990423 Ltr ML18107A2881999-04-0707 April 1999 Rev 0 to NFS-0174, COLR for Salem Unit 2 Cycle 11. ML18107A1821999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Salem,Unit 1.With 990414 Ltr ML18107A1831999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Salem,Unit 2.With 990414 Ltr ML18106B1471999-03-29029 March 1999 LER 99-001-00:on 990228,reactor Scram Was Noted as Result of Turbine Trip.Caused by Operator Error.Lesson Plans Revised to Explicitly Demonstrate Manner in Which Valve Functions. with 990329 Ltr ML18106B1021999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Salem Unit 2.With 990315 Ltr ML18106B1011999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Salem Unit 1.With 990315 Ltr ML18106B0931999-02-25025 February 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Caused by Crack Due to Improper Location of Heated Bar.Only One Part Out of 7396 Pieces in Forging Lot Was Found to Be Cracked.Affected Util,Notified ML18106B0701999-02-16016 February 1999 LER 98-015-00:on 981208,inadvertent Discharge Through RHR Relief Valve During Startup Was Noted.Caused by Operator Performing Too Many Tasks Simultaneously.Appropriate Actions Have Been Taken IAW Policies & Procedures.With 990216 Ltr ML18106B0551999-02-0101 February 1999 Part 21 Rept Re Possible Matl Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Defect Is Crack in Center of Forging.Analysis of Part Is Continuing & Further Details Will Be Provided IAW Ncr Timetables.Drawing of Part,Encl ML18106B0561999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 2.With 990212 Ltr ML18106B0571999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Salem Generating Station,Unit 1.With 990212 Ltr ML20205P1671999-01-31031 January 1999 a POST-PLUME Phase, Federal Participation Exercise ML18106B0441999-01-29029 January 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee Part Number SS-6-T.Caused by Crack in Center of Forging. Continuing Analysis of Part & Will Provide Details in Acoordance with NRC Timetables ML18106B0491999-01-28028 January 1999 LER 98-007-01:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Caused by Transgranular Stress Corrosion Cracking.Replaced Affected Tubing.With 990128 Ltr ML18106B0401999-01-18018 January 1999 LER 98-016-00:on 981219,ECCS Leakage Was Outside of Design Value.Caused by Leakage Past Seat of 21RH34 Manual Drain. Valve 21RH34 Was Reseated.With 990118 Ltr ML18106B0251998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Salem Unit 2.With 990115 Ltr 1999-09-30
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e OPS~G Public_ Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit July 14, 1995 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT NO. 95-010-00 This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73(a) (2) (v) (A).
J. C. Summers General Manager -
Salem Operations SORC Mtg.95-077 MJPJ:vs C Distribution LER File
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9507260220 950714 PDR ADOCK 05000272 S PDR Thl' p~\\*cr is in \"t.!llr h.md~.
95-2168 REV. 6194
NRG FORM 366 . U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5*92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD 1-ICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (315().()104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)
Salem Generating Station Unit 1 05000 272 1 OFS 4
TITLE 11 Inoperability. Of Both Units' Residual Heat Removal- -(RHR} P'umps For Long-Term*
Flow -
Reauirements Due To RHR Flow !ni::ti-11n anf- Tln,.1>i-t-'3inf--f ""
EVENT DATE (5) LER NUMBER (6 REPORT NUMBER (7} OTHER FACILITIES INVOLVED (8)
FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER c::,,1 ~~ n.,; .. ? 05000 311 FACILITY NAME DOCKET NUMBER 06 15 95 95 -- 010. -- 00 07 14 95 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §; (Check one or more I (11)
MODE (9) 5 20.402(b) 20.405(c) 50.73(a)(2) (Iv) 73.r,1 (b)
IX I L:~~~~O) I 0 I 20.405(a)(1)(l) 20.405(a)(1)(li) 50.36(c)(1) 50.36(c)(2) 50.73(a)(2)(v) 50.73(a)(2)(vii) 50.73(a) (2) (viii) (A) 73.i1(c)
OTHER (Specify in Abstracl 20.405(a)(1)(iii) 50.73(a) (2) (i) below and in Text, NRG 20.405(a)(1)(iv) 50.73(a) (2) (ii) 50.73(a)(2)(viii)(B) Form 366A) 20.405 (a)(1 )(v) 50. 73 (a) (2) (iii) 50.73(a)(2)(x)
LICENSEE CONTACT FOR THIS LER 12)
NAME TELEPHONE NUMBER (Include Area Code)
M. J. Pastva Jr. LER Coordinator 609/339-5165 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORTC13l REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TONPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR X
I YES (If yes, complete EXPECTED SUBMISSION DATE)
NO SUBMISSION DATE (15) 10 31 95 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
At 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br /> on 6/15/95, the NRC was notified of a potential problem with the Residual Heat Removal (RHR) flow orifices installation and the potential for inaccurate operation of the RHR pump recirculation valves, which provide pump protection. Testing on 6/30/95 satisfactorily demonstrated the operability and control function of the RHR flow orifices and the RHR pump recirculation valves. However, on 7/10/95 a potential design basis concern was identified that the 1000 gpm minimum continuous (long-term) flow requirements of the RHR pumps could not be assured. Subsequently, it was determined on 7/12/95 the concern with the RHR flow instrumentation included RHR operation per Emergency Operating Procedures (EOPs) to mitigate the effects of mainly a small break Loss of Coolant Accident scenario. The cause of this occurrence is attributed to a design deficiency when the uncertainties for the RHR loop flow instruments were not accounted for. An investigation is in progress to determine *the cause(:;) for use of the incorrect instrument setpoints_ Prior to subsequent entry of either Salem Unit into Mode 4, changes to plant operating procedures and EOPs will be evaluated, and appropriate changes will be implemented, as necessary. It is anticipated that, by 10/31/95 this report will be supplemented to further detail the root cause of this occurrence and any additional corrective actions.
NRG FORM 366 (5-92)
REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER OF TITLE NUMBER DIGITS/CHARACTERS 1 UP TO 46 FACILITY NAME 8 TOTAL- -
2 DOC-KET NUMBER 3 IN ADDITION TO 05000 3 VARIES PAGE NUMBER 4 UP TO 76 TITLE 6TOTAL 5 EVENT DATE 2 PER BLOCK 7 TOTAL 2 FOR YEAR 6 LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 7 REPORT DATE 2 PER BLOCK UP TO 18 FACILITY NAME 8 OTHER FACILITIES INVOLVED 8 TOTAL - DOCKET NUMBER
- 3 IN ADDITION TO 05000 9 1 OPERATING MODE .
10 3 POWER LEVEL 1
11 REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1
14 SUPPL!=MENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15 EXPECTED SUBMISSION DATE 2 PER BLOCK
LICENSEE EVENT REPORT {LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 2 of 5 Unit # 1 50-272 95-010-00 Plant and System Identification:
Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes appear in the text as {xx}
Identification of Occurrence:
Inoperability Of Both Units' Residual Heat Removal (RHR)
Pumps For Long-Term Flow Requirements Due To RHR Flow Instrument Uncertainties Event Date: June 15, 1995 Report Date: July 14, 1995 This report was initiated by Incident Report No.95-873
Conditions Prior to Occurrence:
.
- Ji, Both Units were in a self-imposed extended shutdown.
Mode 5 Reactor Power % Unit Load Mwe Description of Occurrence:
At 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br /> on June 15, 1995, the NRC was notified, in accordance with 10CFR50. 72 (b) (2) (i), of a potential problem with the RHR flow orifices 1(2)FE641A/B installation and potential for inaccurate operation of the RHR pump recirculation valves, which provide pump protection.
Testing on June 30, 1995 satisfactorily demonstrated the operability and control function of the RHR flow orifices and the RHR pump recirculation valves. However, on July 10, 1995, a potential design basis concern was identified that the 1000 gpm minimum continuous (long-term) flow requirements of the RHR pumps could not be assured. This determination followed recognition of that instrument loop uncertainties were not factored into the RHR flow indication setpoints. Subsequently, it was determined on July 12, 1995, the concern with the RHR flow instrumentation included RHR operation per Emergency Operating Procedures (EOPs) to
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 3 of 5 Unit # 1 50-272 95-010-00 Description of Occurrence: (cont'd) mitigate the effects of a small break Loss of Coolant Accident (LOCA) scenario. The EOPs for steam generator tube rupture secondary side breaks and inadvertent safety injection (SI) are also of concern.
Analysis of Occurrence:
Review of plant drawings associated with RHR pump discharge piping indicated a potential operability/design problem involving the installation of the minimum recirculation control orifices on both Salem Units.
Testing determined this issue was not an operability/design concern. However, this testing identified a concern with the accuracy of the RHR minimum recirculation valve control setpoints, which potentially affects the RHR pump continuous service operation. Subsequently, this concern was expanded to include securing the RHR pumps, per EOP guidance.
Apparent Cause of Occurrence:
The cause of this occurrence is attributed to ~Design, Manufacturing/Construction", as classified in NUREG-1022, Appendix B. This occurred when the instrument uncertainties for the RHR loop flow instruments were not accounted for in establishing the instruments' setpoints on the Salem Units.
The results of an ongoing investigation to determine the cause(s) for use of the incorrect instrument setpoints, including contributing factors, as well as failed or deficient controls and barriers will be reflected in a supplement to this report.
Prior Similar Occurrence:
Review of documentation did not reveal a prior similar occurrence.
Safety Significance:
This occurrence is reportable pursuant to 10CFR50.73(a) (2) (v) (A).
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 4 of 5 Unit # 1 50-272 95-010-00 Safety Significance: (cont'd)
Potential RHR Operation At Flow Less Than 1000 gpm.
Incorporating RHR flow instrumentation inaccuracies obtained through testing into in the setpoint calculation, the lowest expected minimum flow rate at closure of the recirculation valve for continuous flow service of the RHR pumps is estimated to be above 800 gpm. In addition, the RHR pump vendor has evaluated that the pumps are suitable to operate continuously at a flow rate of 800 gpm. As such, no safety concern exists with RHR operation at the current setpoint of 1000 gpm.
Review of EOPs for accident conditions indicate the RHR pumps would be stopped in less than 45 minutes from initiation of an accident, if Reactor Coolant System pressure is above the pump shutoff head pressure and not injecting into the RCS. In accordance with the EOPs, an RCS pressure comparison check is made on cold leg injection flow indication. If the flow indication is less than 200 gpm indicated, the operator is instructed to stop the involved RHR pump. Testing results support that the maximum combined flow through the RHR pump at that time would be 550 gpm recirculation flow through the pump minimum flow valves plus the 200 gpm flow, plus or minus process and loop uncertainties, for a total flow of approximately 750 gpm.
RHR flow injection starts when RHR discharge pressure exceeds RCS pressure, at approximately 350 psig. For a small break LOCA that results in RHR flow to the RCS, the flow will increase in response to decreasing RCS pressure .
. As such, RHR flow lower than 800 gpm is not expected to occur for an extended duration. Consequently, damage to the pump, from the effects of low flow for this short duration is expected to be minimal. The reduced RHR flow is not a potential concern for large break LOCAs since RCS pressure will depressurize rapidly.
Corrective Action:
Prior to subsequent entry of both Salem Units into Mode 4:
Changes to plant operating procedures and EOPs will be evaluated, and if necessary, appropriate changes will be implemented on each Salem Unit.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page 5 of 5 Unit # 1 50-272 95-010-00 Corrective Action: (cont'd)
It is anticipated that by October 31, 1995, a supplement to this report will be submitted to further detail the root cause of this occurrence, as well as any additional corrective actions identified.
~~
J. C. Summers General Manager -
Salem Operations MJPJ:vs REEF: SORC Mtg.95-077