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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
(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.
-.
Of Both Units' Residual Heat Removal--(RHR} P'umps For Long-Term*
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).
Flow -Reauirements Due To RHR Flow !ni::ti-11n anf-Tln,.1>i-t-'3inf--f  
J. C. Summers General Manager -
"" 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)
Salem Operations SORC Mtg. 95-077 MJPJ:vs C         Distribution LER File
IX 50.73(a)(2)(v) 73.i1(c) 20.405(a)(1)(li) 50.36(c)(2) 50.73(a)(2)(vii)
                        ~  .'"' ,..... ::; ..'! ,.)
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)
9507260220 950714 PDR ADOCK 05000272 S                                   PDR Thl' p~\\*cr is in \"t.!llr h.md~.
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.
95-2168 REV. 6194
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  
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.
*the cause(:;)
FACILITY NAME (1)                                                                                           DOCKET NUMBER (2)                                     PAGE (3)
for use of the incorrect instrument setpoints_
Salem Generating Station Unit 1                                                                                 05000 272                               1 OFS 4
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.
TITLE 11 Inoperability. Of Both Units' Residual Heat Removal- -(RHR} P'umps For Long-Term*
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)
Flow -
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
Reauirements Due To RHR Flow !ni::ti-11n anf- Tln,.1>i-t-'3inf--f ""
* 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 -
EVENT DATE (5)                           LER NUMBER (6                   REPORT NUMBER (7}                         OTHER FACILITIES INVOLVED (8)
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:
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)
Westinghouse  
MODE (9)               5         20.402(b)                             20.405(c)                                 50.73(a)(2) (Iv)                   73.r,1 (b)
-Pressurized Water Reactor Energy Industry Identification System (EIIS) codes appear in the text as {xx} Identification of Occurrence:
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)
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:
OTHER (Specify in Abstracl
Both Units were in a self-imposed extended shutdown. , Mode 5 Reactor Power % Unit Load Mwe Description of Occurrence:
-
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 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}}

Revision as of 10:11, 21 October 2019

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
ML18101A845
Person / Time
Site: Salem PSEG icon.png
Issue date: 07/14/1995
From: Pastva M, Summers J
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-95-010-01, LER-95-10-1, NUDOCS 9507260220
Download: ML18101A845 (7)
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Text

e OPS~G

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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

~ .'"' ,..... ::; ..'! ,.)

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

  • 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 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

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