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{{#Wiki_filter:* Public Service .Electric and Gas Company P.O. Box 236 _ Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit MAR 2 o 1998 LR-N980129 U.
{{#Wiki_filter:I Public Service .Electric and Gas Company P.O. Box 236 _Hancocks Bridge, New Jersey 08038-0236 I
* S. Nuclear Regulatory Cmmnission Document Control Desk Washington, DC 20555 LER 272/98-006-00 SALEM GENERATING STATION -UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 5-0-272 Gentlemen:
Nuclear Business Unit MAR 2 o 1998 LR-N980129 U.
This Licensee Event Report entitled "ESF Actuation of *11 arid 12 Auxiliary Feedwater Pumps" is being submitted pursuant to the requirements of the Code of Federal Regulations  
* S. Nuclear Regulatory Cmmnission Document Control Desk Washington, DC 20555 LER 272/98-006-00 SALEM GENERATING STATION - UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 5-0-272 Gentlemen:
****10CFR50.73 (a_) (2) (iv)****.
This Licensee Event Report entitled "ESF Actuation of *11 arid 12 Auxiliary Feedwater Pumps" is being submitted pursuant to the requirements of the Code of Federal Regulations ****10CFR50.73 (a_) (2) (iv)****.
Attachment c Distribution LER File 3.7 Sincerely, A. c. Bakken III General Manager_ Salem Operations r-, -I .. dd I --9803310173 980320 PDR ADOCK 05000272 S PDR The power is in your hands. 95-2168 REV. 6/94 I I 
Sincerely, A. c. Bakken III General Manager_
. I * *,1. " . I NRCl"'ORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB*NO. 3150-0104 (4-95) EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. LICENSEE EVENT _REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.
Salem Operations Attachment
FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND . ' RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR (See reverse for required number of REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO digits/characters for each block) THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. I F1 .. ;1LITT NAllE (1) DOCKET NUllBER (2) PAGE (3) SALEMUNIT1 05000272 1 OF4 TITLE (4) ESF Actuation of 11 and 12 Auxiliary Feedwater Pumps EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8) YEAR' I SEQUENTIAL . I REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR MONTH DAY YEAR ' 2 21 98 98 006 00 3 20 98 FACILITY NAME DOCKET NUMBER ----OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11) . MODE (9) 4 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i)(B) 50.73(a)(2)(viii)
*~PJD/kjb c            Distribution LER File 3.7 r-,   -I
POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)
                                                                                            ~f_ . dd           I 9803310173 980320 PDR ADOCK 05000272 S                         PDR The power is in your hands.
LEVEL (10) 0 20.2203(a)(2)(i) 20.2203(a)(3)(ii)
95-2168 REV. 6/94
: 50. 73(a)(2)(iii) 73.71 -20.2203(a)(2)(ii) 20.2203(a)(4) x 50. 73(a)(2)(iv)
 
OTHER 20.2203(a)(2)(iii) 50.36(c)(1)
--~_,
: 50. 73(a)(2)(v)
      . I NRCl"'ORM366 (4-95)
Abstract below or in C Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT _REPORT (LER)
LICENSEE CONTACT FOR THIS LER (12) NAME TELEPHONE NUMBER (Include Area Code) Philip J. Duca, Salem Licensing Engineer (609) 339-2381 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE I CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR IYES xi NO SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). DATE (15) xx xx xx ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) .. At 1332 on February 21, 1998 with Salem* Unit 1 operating in Mode 4 (Hot Shutdown), 11 and 12 Auxiliary Feedwater Pumps automatically started on Lo-Lo Steam Generator level in 14 Steam Generator.
* APPROVED BY OMB*NO. 3150-0104 EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.
The cause of this event was human error. The Main Steam lines were being warmed through the Main Steam Stop Bypass Valves. These valves had been opened wider during the night shift increasing the steaming rate of the steam generator$.
REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.
The operators did not adequately monitor steam generator water levels nor did they anticipate the increased steaming rate. Contributing causes were ineffective shift turnover and poor team communication.
COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND FORWARD
The operators promptly established feedwater to all steam generators and restored water levels. Other corrective actions induded discussion of the event and lessons learned by the personnel involved with all operating crews, and a reemphasis of responsibilities and the importance of safe operations by the shift superintendents.
                                                          .'                                       RECORDS MANAGEMENT BRANCH (T-6 F33), U.S.                 NUCLEAR (See reverse for required number of                               REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF digits/characters for each block)                                MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
Additionally, personnel involved have been held accountable in accordance with PSE&G procedures and policies. A 4 hour telephone notification was made to the NRC at 1500 on February 21, 1998 in accordance with 10CFR50.72(b)(2)ii; this report is being submitted in accordance with 10CFR50.73(a)(2)(iv);
F1..;1LITT NAllE (1)                                                                       DOCKET NUllBER (2)                                 PAGE (3)
both required for "any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature (ESF) ...... ". NRC FORM 366 (4-95)
SALEMUNIT1                                                       05000272                               1 OF4 TITLE (4)
* NRC FORM 366A (4-95) U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) LICENSEE EVENT REPOIJT (LER) TEXT CONTINUATION DOCKET NUMBER (2) LER NUMBER (6) YEAR I SEQUENTIAL I REVISION NUMBER NUMBER PAGE (3) SALEM UNIT1 05000272 -006 --00 2 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) PLANT AND SYSTEM IDENTIFICATION Westinghouse  
ESF Actuation of 11 and 12 Auxiliary Feedwater Pumps EVENT DATE (5)                   LER NUMBER (6)                 REPORT DATE (7)                     OTHER FACILITIES INVOLVED (8)
-Pressurized Water Reactor Auxiliary/Emergency Feedwater System {BA/-}*
FACILITY NAME                            DOCKET NUMBER MONTH      DAY    YEAR' YEAR I   SEQUENTIAL . I REVISION   MONTH   DAY     YEAR 2       21     98       98     --  006     --    00         3     20     98 FACILITY NAME                             DOCKET NUMBER OPERATING                 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)
            . MODE (9)         4         20.2201(b)                         20.2203(a)(2)(v)                 50.73(a)(2)(i)(B)                   50.73(a)(2)(viii)
POWER                     20.2203(a)(1)                     20.2203(a)(3)(i)                 50.73(a)(2)(ii)                     50.73(a)(2)(x)
LEVEL (10)         0         20.2203(a)(2)(i)                   20.2203(a)(3)(ii)                 50. 73(a)(2)(iii)                   73.71 x
20.2203(a)(2)(ii)                 20.2203(a)(4)                     50. 73(a)(2)(iv)                     OTHER 20.2203(a)(2)(iii)                 50.36(c)(1)                       50. 73(a)(2)(v)                 Spec~in    Abstract below or in   C Form 366A 20.2203(a)(2)(iv)                 50.36(c)(2)                       50.73(a)(2)(vii)
LICENSEE CONTACT FOR THIS LER (12)
NAME                                                                                             TELEPHONE NUMBER (Include Area Code)
Philip J. Duca, Salem Licensing Engineer                                                                             (609) 339-2381 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE       SYSTEM     COMPONENT     MANUFACTURER       REPORTABLE             CAUSE       SYSTEM     COMPONENT     MANUFACTURER       REPORTABLE TONPRDS                                                                           TONPRDS IYES SUPPLEMENTAL REPORT EXPECTED (14)
(If yes, complete EXPECTED SUBMISSION DATE).
Ixi  NO EXPECTED SUBMISSION DATE (15)
MONTH xx DAY xx YEAR xx ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
At 1332 on February 21, 1998 with Salem* Unit 1 operating in Mode 4 (Hot Shutdown), 11 and 12 Auxiliary Feedwater Pumps automatically started on Lo-Lo Steam Generator level in 14 Steam Generator.
The cause of this event was human error. The Main Steam lines were being warmed through the Main Steam Stop Bypass Valves. These valves had been opened wider during the night shift increasing the steaming rate of the steam generator$. The operators did not adequately monitor steam generator water levels nor did they anticipate the increased steaming rate. Contributing causes were ineffective shift turnover and poor team communication.
The operators promptly established feedwater to all steam generators and restored water levels. Other corrective actions induded discussion of the event and lessons learned by the personnel involved with all operating crews, and a reemphasis of responsibilities and the importance of safe operations by the shift superintendents. Additionally, personnel involved have been held accountable in accordance with PSE&G procedures and policies.
A 4 hour telephone notification was made to the NRC at 1500 on February 21, 1998 in accordance with 10CFR50.72(b)(2)ii; this report is being submitted in accordance with 10CFR50.73(a)(2)(iv); both required for "any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature (ESF) ...... ".
NRC FORM 366 (4-95)
 
NRC FORM 366A (4-95)
FACILITY NAME (1)
LICENSEE EVENT REPOIJT (LER)
TEXT CONTINUATION DOCKET NUMBER (2)
U.S. NUCLEAR REGULATORY COMMISSION LER NUMBER (6)           PAGE (3)
YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SALEM UNIT1                                     05000272 -      98 -- 006 --         00     2   OF     4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Auxiliary/Emergency Feedwater System {BA/-}*
* Energy Industry Identification System (EllS) codes and component function identifier codes appear as {SS/CCC}.
* Energy Industry Identification System (EllS) codes and component function identifier codes appear as {SS/CCC}.
CONDITIONS PRIOR TO OCCURRENCE At the time of the occurrence Salem Unit *1 was in Mode 4 at 0 % Power with the Residual Heat Removal System in service providing core cooling. DESCRIPTION OF OCCURRENCE On February 19, the night shift (1900-0700) commenced warming up the Main Steam lines in accordance with procedures being utilized to take the plant from Cold Shutdown to Hot Standby. Warming up the Main Steam lines requires opening of the Main Steam Stop Bypass Valves (11-14MS18 valves). In Mode 4 feedwater is manually supplied to the steam generators on a periodic basis. This was being accomplished by starting and stopping of the Auxiliary Feedwater pumps. Establishing flow through the MS18 valves increased the steaming rate of the steam generators, thereby_increasing the required frequency of providing feedwater to the steam generators to maintain
CONDITIONS PRIOR TO OCCURRENCE At the time of the occurrence Salem Unit *1 was in Mode 4 at 0 % Power with the Residual Heat Removal System in service providing core cooling.
* water levels. The MS18 valves were opened approximately 1 to 2% of valve open position.
DESCRIPTION OF OCCURRENCE On February 19, the night shift (1900-0700) commenced warming up the Main Steam lines in accordance with procedures being utilized to take the plant from Cold Shutdown to Hot Standby.
The increased steaming rate required the starting of Auxiliary Feedwater pumps approximately once per 12 hour shift in order to maintain steam generator levels. On February 20, the night shift further opened all MS18 valves to approximately 4% valve open position.
Warming up the Main Steam lines requires opening of the Main Steam Stop Bypass Valves (11-14MS18 valves). In Mode 4 feedwater is manually supplied to the steam generators on a periodic basis. This was being accomplished by starting and stopping of the Auxiliary Feedwater pumps.
The steam generators were filled to greater than 33% narrow range level at 0451 on the morning of February 21. The additional steam demand, which caused water levels to fall at an increased rate, was not anticipated by the coming day shift (0700-1900).
Establishing flow through the MS18 valves increased the steaming rate of the steam generators, thereby_increasing the required frequency of providing feedwater to the steam generators to maintain
14 Steam Generator Narrow Range Level dropped from 36% to 9% in approximately 8.5 hours. Prior to the event steam generator water level narrow range chart recorders and steam generator water level program deviation console alarms on Control Console 2 were inoperable due to Advanced Digital Feedwater Control System testing which was in progress.
* water levels. The MS18 valves were opened approximately 1 to 2% of valve open position. The increased steaming rate required the starting of Auxiliary Feedwater pumps approximately once per 12 hour shift in order to maintain steam generator levels. On February 20, the night shift further opened all MS18 valves to approximately 4% valve open position. The steam generators were filled to greater than 33% narrow range level at 0451 on the morning of February 21. The additional steam demand, which caused water levels to fall at an increased rate, was not anticipated by the on-coming day shift (0700-1900). 14 Steam Generator Narrow Range Level dropped from 36% to 9% in approximately 8.5 hours.
NRC FORM 366A (4*95)
Prior to the event steam generator water level narrow range chart recorders and steam generator water level program deviation console alarms on Control Console 2 were inoperable due to Advanced Digital Feedwater Control System testing which was in progress.
!, I I ,. ** )
NRC FORM 366A (4*95)
* NRC FORM 366A (4-95) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) YEAR I SEQUENTIAL I REVISION NUMBER NUMBER SALEM UNIT 1 05000272 98 --006 --00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) DESCRIPTION OF OCCURRENCE (continued)
 
* PAGE (3) 3 OF 4 The Unit 1 Reactor Operator (Nuclear Control Operator) logged 14 Steam Generator Narrow Range Level at 32% at 0730. At 1330, while performing shiftly logs, the Reactor Operator noticed 14 Steam Generator Level to be 12%. The Reactor Operator was about to start the Auxiliary Feedwater pumps and refill the steam generators just as the automatic action occurred.
)
At 1332, 11 and 12 Auxiliary Feedwater pumps automatically started on Lo-Lo steam generator level when 14 Steam Generator  
NRC FORM 366A (4-95)
*water level reached 9% narrow range level. Operators promptly established feedwater to all steam generators to restore water levels. At the time of the event, 11 Steam Generator water level was 21 %, 12 Steam Generator level was 31 %, and 13 Steam Generator level was 32%. Following the event, all steam generator water level narrow range chart recorders and program water level deviation alarms were returned to service. All steam generator levels were restored to greater than 33% as indicated on the narrow range level instruments.
LICENSEE EVENT REPORT (LER)
CAUSE OF OCCURRENCE The cause of this event was human error. The *unit 1 Reactor Operator, Plant Operator (the other contra.I room Nuclear Control Operator), and Control Room Supervisor did not adequately monitor steam generator water levels nor did they anticipate the increased feedwater requirements.
TEXT CONTINUATION DOCKET NUMBER (2)
The Reactor Operator did not determine the rate at which steam generator water levels were decreasing.
U.S. NUCLEAR REGULATORY COMMISSION LER NUMBER (6)               PAGE (3)
Steam generator narrow range water levels are logged every 6 hours. The Unit 1 Reactor Operator logged 14 Steam Generator Narrow Range Level at 32% at 0730. At 1330, while performing his logs, he noted 14 level to be 12% based on one of three narrow range level instruments.  
FACILITY NAME (1)
*The other two narrow range level instruments were at 9%. When level reached 9 %, the autostart of the auxiliary feedwater pumps occurred as designed.
YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SALEM UNIT 1                                   05000272       98 -- 006       --     00     3  OF    4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
DESCRIPTION OF OCCURRENCE (continued)
* The Unit 1 Reactor Operator (Nuclear Control Operator) logged 14 Steam Generator Narrow Range Level at 32% at 0730. At 1330, while performing shiftly logs, the Reactor Operator noticed 14 Steam Generator Level to be 12%. The Reactor Operator was about to start the Auxiliary Feedwater pumps and refill the steam generators just as the automatic action occurred. At 1332, 11 and 12 Auxiliary Feedwater pumps automatically started on Lo-Lo steam generator level when 14 Steam Generator
  *water level reached 9% narrow range level. Operators promptly established feedwater to all steam generators to restore water levels. At the time of the event, 11 Steam Generator water level was 21 %, 12 Steam Generator level was 31 %, and 13 Steam Generator level was 32%.
Following the event, all steam generator water level narrow range chart recorders and program water level deviation alarms were returned to service. All steam generator levels were restored to greater than 33% as indicated on the narrow range level instruments.
CAUSE OF OCCURRENCE The cause of this event was human error. The *unit 1 Reactor Operator, Plant Operator (the other contra.I room Nuclear Control Operator), and Control Room Supervisor did not adequately monitor steam generator water levels nor did they anticipate the increased feedwater requirements. The Reactor Operator did not determine the rate at which steam generator water levels were decreasing.
Steam generator narrow range water levels are logged every 6 hours. The Unit 1 Reactor Operator logged 14 Steam Generator Narrow Range Level at 32% at 0730. At 1330, while performing his logs, he noted 14 level to be 12% based on one of three narrow range level instruments. *The other two narrow range level instruments were at 9%. When level reached 9 %, the autostart of the auxiliary feedwater pumps occurred as designed.
A contributing cause* of this event was ineffective shift turnover and poor team communication.
A contributing cause* of this event was ineffective shift turnover and poor team communication.
Although the additional steam demand on the steam generators was discussed during the individual watch turnovers, the subject was never discussed at the Pre-watch Shift Brief nor did the Unit 1 Control Room Crew discuss any increased monitoring of steam generator water levels. The operating crew also did not discuss what steam generator water level should be maintained.
Although the additional steam demand on the steam generators was discussed during the individual watch turnovers, the subject was never discussed at the Pre-watch Shift Brief nor did the Unit 1 Control Room Crew discuss any increased monitoring of steam generator water levels. The operating crew also did not discuss what steam generator water level should be maintained. The Operations Superintendent was not informed that the MS18 valves were opened an additional 2%.
The Operations Superintendent was not informed that the MS18 valves were opened an additional 2%. NRC FORM 366A (4*95) 1 NRC r-"ORM 366A (4-95)
NRC FORM 366A (4*95)
* U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) YEAR I SEQUENTIAL I REVISION NUMBER NUMBER SALEM UNIT1 05000272 98 --006 --00 4 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) PRIOR SIMILAR OCCURRENCES A review of LERs issued in the past two years did not identify any similar occurrences regarding the autostart of the auxiliary feedwater pumps SAFETY CONSEQUENCES AND IMPLICATIONS The Auxiliary Feedwater pump automatic start on Lo-Lo steam generator level functioned as designed.
 
While steam generator level was dropping, the rate of decrease was relatively slow. The operators had ample time to react to the event prior to level in the steam generator reaching the extent (reaching hot and dry conditions) where damage could occur. Level in the other three steam generator levels remained above the autostart initiation level. At the time of the event core cooling was being provided by 11 Residual Heat Removal Loop. The steam generators were not being relied on for decay heat removal. Therefore, the health and safety of the public were not affected.
1 NRC r-"ORM 366A (4-95)
CORRECTIVE ACTIONS 1. The operators promptly established feedwater to all steam generators and restored proper water levels. 2. All steam generator water level narrow range chart recorders and program water level deviation console alarms were returned to operable status on February 21 , 1998. 3. The Reactor Operator and Plant Operator at the time of the event have discussed this event and lessons learned with all operating crews. 4. All on shift Operations Superintendents have discussed this event and have led a discussion with their respective crews regarding their responsibilities and the importance of safe operations.
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION DOCKET NUMBER (2)
U.S. NUCLEAR REGULATORY COMMISSION LER NUMBER (6)           PAGE (3)
FACILITY NAME (1)
YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SALEM UNIT1                                     05000272       98 -- 006 --         00     4   OF     4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
PRIOR SIMILAR OCCURRENCES A review of LERs issued in the past two years did not identify any similar occurrences regarding the autostart of the auxiliary feedwater pumps SAFETY CONSEQUENCES AND IMPLICATIONS The Auxiliary Feedwater pump automatic start on Lo-Lo steam generator level functioned as designed. While steam generator level was dropping, the rate of decrease was relatively slow. The operators had ample time to react to the event prior to level in the steam generator reaching the extent (reaching hot and dry conditions) where damage could occur. Level in the other three steam generator levels remained above the autostart initiation level. At the time of the event core cooling was being provided by 11 Residual Heat Removal Loop. The steam generators were not being relied on for decay heat removal. Therefore, the health and safety of the public were not affected.
CORRECTIVE ACTIONS
: 1. The operators promptly established feedwater to all steam generators and restored proper water levels.
: 2. All steam generator water level narrow range chart recorders and program water level deviation console alarms were returned to operable status on February 21 , 1998.
: 3. The Reactor Operator and Plant Operator at the time of the event have discussed this event and lessons learned with all operating crews.
: 4. All on shift Operations Superintendents have discussed this event and have led a discussion with their respective crews regarding their responsibilities and the importance of safe operations.
: 5. All personnel involved have been held accountable in accordance with PSE&G's procedures and policies.
: 5. All personnel involved have been held accountable in accordance with PSE&G's procedures and policies.
NRC FORM 366A (4-95)}}
NRC FORM 366A (4-95)}}

Latest revision as of 04:57, 3 February 2020

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
ML18106A403
Person / Time
Site: Salem PSEG icon.png
Issue date: 03/20/1998
From: Bakken A, Duca P
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-98-006, LER-98-6, LR-N980129, NUDOCS 9803310173
Download: ML18106A403 (5)
v • d • e


Text

I Public Service .Electric and Gas Company P.O. Box 236 _Hancocks Bridge, New Jersey 08038-0236 I

Nuclear Business Unit MAR 2 o 1998 LR-N980129 U.

  • S. Nuclear Regulatory Cmmnission Document Control Desk Washington, DC 20555 LER 272/98-006-00 SALEM GENERATING STATION - UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 5-0-272 Gentlemen:

This Licensee Event Report entitled "ESF Actuation of *11 arid 12 Auxiliary Feedwater Pumps" is being submitted pursuant to the requirements of the Code of Federal Regulations ****10CFR50.73 (a_) (2) (iv)****.

Sincerely, A. c. Bakken III General Manager_

Salem Operations Attachment

  • ~PJD/kjb c Distribution LER File 3.7 r-, -I

~f_ . dd I 9803310173 980320 PDR ADOCK 05000272 S PDR The power is in your hands.

95-2168 REV. 6/94

--~_,

. I NRCl"'ORM366 (4-95)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT _REPORT (LER)

  • APPROVED BY OMB*NO. 3150-0104 EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.

COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND FORWARD

.' RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR (See reverse for required number of REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

F1..;1LITT NAllE (1) DOCKET NUllBER (2) PAGE (3)

SALEMUNIT1 05000272 1 OF4 TITLE (4)

ESF Actuation of 11 and 12 Auxiliary Feedwater Pumps EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

FACILITY NAME DOCKET NUMBER MONTH DAY YEAR' YEAR I SEQUENTIAL . I REVISION MONTH DAY YEAR 2 21 98 98 -- 006 -- 00 3 20 98 FACILITY NAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)

. MODE (9) 4 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i)(B) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 0 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50. 73(a)(2)(iii) 73.71 x

20.2203(a)(2)(ii) 20.2203(a)(4) 50. 73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50. 73(a)(2)(v) Spec~in Abstract below or in C Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Code)

Philip J. Duca, Salem Licensing Engineer (609) 339-2381 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS IYES SUPPLEMENTAL REPORT EXPECTED (14)

(If yes, complete EXPECTED SUBMISSION DATE).

Ixi NO EXPECTED SUBMISSION DATE (15)

MONTH xx DAY xx YEAR xx ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

At 1332 on February 21, 1998 with Salem* Unit 1 operating in Mode 4 (Hot Shutdown), 11 and 12 Auxiliary Feedwater Pumps automatically started on Lo-Lo Steam Generator level in 14 Steam Generator.

The cause of this event was human error. The Main Steam lines were being warmed through the Main Steam Stop Bypass Valves. These valves had been opened wider during the night shift increasing the steaming rate of the steam generator$. The operators did not adequately monitor steam generator water levels nor did they anticipate the increased steaming rate. Contributing causes were ineffective shift turnover and poor team communication.

The operators promptly established feedwater to all steam generators and restored water levels. Other corrective actions induded discussion of the event and lessons learned by the personnel involved with all operating crews, and a reemphasis of responsibilities and the importance of safe operations by the shift superintendents. Additionally, personnel involved have been held accountable in accordance with PSE&G procedures and policies.

A 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> telephone notification was made to the NRC at 1500 on February 21, 1998 in accordance with 10CFR50.72(b)(2)ii; this report is being submitted in accordance with 10CFR50.73(a)(2)(iv); both required for "any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature (ESF) ...... ".

NRC FORM 366 (4-95)

NRC FORM 366A (4-95)

FACILITY NAME (1)

LICENSEE EVENT REPOIJT (LER)

TEXT CONTINUATION DOCKET NUMBER (2)

U.S. NUCLEAR REGULATORY COMMISSION LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SALEM UNIT1 05000272 - 98 -- 006 -- 00 2 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Auxiliary/Emergency Feedwater System {BA/-}*

  • Energy Industry Identification System (EllS) codes and component function identifier codes appear as {SS/CCC}.

CONDITIONS PRIOR TO OCCURRENCE At the time of the occurrence Salem Unit *1 was in Mode 4 at 0 % Power with the Residual Heat Removal System in service providing core cooling.

DESCRIPTION OF OCCURRENCE On February 19, the night shift (1900-0700) commenced warming up the Main Steam lines in accordance with procedures being utilized to take the plant from Cold Shutdown to Hot Standby.

Warming up the Main Steam lines requires opening of the Main Steam Stop Bypass Valves (11-14MS18 valves). In Mode 4 feedwater is manually supplied to the steam generators on a periodic basis. This was being accomplished by starting and stopping of the Auxiliary Feedwater pumps.

Establishing flow through the MS18 valves increased the steaming rate of the steam generators, thereby_increasing the required frequency of providing feedwater to the steam generators to maintain

  • water levels. The MS18 valves were opened approximately 1 to 2% of valve open position. The increased steaming rate required the starting of Auxiliary Feedwater pumps approximately once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift in order to maintain steam generator levels. On February 20, the night shift further opened all MS18 valves to approximately 4% valve open position. The steam generators were filled to greater than 33% narrow range level at 0451 on the morning of February 21. The additional steam demand, which caused water levels to fall at an increased rate, was not anticipated by the on-coming day shift (0700-1900). 14 Steam Generator Narrow Range Level dropped from 36% to 9% in approximately 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Prior to the event steam generator water level narrow range chart recorders and steam generator water level program deviation console alarms on Control Console 2 were inoperable due to Advanced Digital Feedwater Control System testing which was in progress.

NRC FORM 366A (4*95)

)

NRC FORM 366A (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET NUMBER (2)

U.S. NUCLEAR REGULATORY COMMISSION LER NUMBER (6) PAGE (3)

FACILITY NAME (1)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SALEM UNIT 1 05000272 98 -- 006 -- 00 3 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DESCRIPTION OF OCCURRENCE (continued)

Following the event, all steam generator water level narrow range chart recorders and program water level deviation alarms were returned to service. All steam generator levels were restored to greater than 33% as indicated on the narrow range level instruments.

CAUSE OF OCCURRENCE The cause of this event was human error. The *unit 1 Reactor Operator, Plant Operator (the other contra.I room Nuclear Control Operator), and Control Room Supervisor did not adequately monitor steam generator water levels nor did they anticipate the increased feedwater requirements. The Reactor Operator did not determine the rate at which steam generator water levels were decreasing.

Steam generator narrow range water levels are logged every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The Unit 1 Reactor Operator logged 14 Steam Generator Narrow Range Level at 32% at 0730. At 1330, while performing his logs, he noted 14 level to be 12% based on one of three narrow range level instruments. *The other two narrow range level instruments were at 9%. When level reached 9 %, the autostart of the auxiliary feedwater pumps occurred as designed.

A contributing cause* of this event was ineffective shift turnover and poor team communication.

Although the additional steam demand on the steam generators was discussed during the individual watch turnovers, the subject was never discussed at the Pre-watch Shift Brief nor did the Unit 1 Control Room Crew discuss any increased monitoring of steam generator water levels. The operating crew also did not discuss what steam generator water level should be maintained. The Operations Superintendent was not informed that the MS18 valves were opened an additional 2%.

NRC FORM 366A (4*95)

1 NRC r-"ORM 366A (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET NUMBER (2)

U.S. NUCLEAR REGULATORY COMMISSION LER NUMBER (6) PAGE (3)

FACILITY NAME (1)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SALEM UNIT1 05000272 98 -- 006 -- 00 4 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

PRIOR SIMILAR OCCURRENCES A review of LERs issued in the past two years did not identify any similar occurrences regarding the autostart of the auxiliary feedwater pumps SAFETY CONSEQUENCES AND IMPLICATIONS The Auxiliary Feedwater pump automatic start on Lo-Lo steam generator level functioned as designed. While steam generator level was dropping, the rate of decrease was relatively slow. The operators had ample time to react to the event prior to level in the steam generator reaching the extent (reaching hot and dry conditions) where damage could occur. Level in the other three steam generator levels remained above the autostart initiation level. At the time of the event core cooling was being provided by 11 Residual Heat Removal Loop. The steam generators were not being relied on for decay heat removal. Therefore, the health and safety of the public were not affected.

CORRECTIVE ACTIONS

1. The operators promptly established feedwater to all steam generators and restored proper water levels.
2. All steam generator water level narrow range chart recorders and program water level deviation console alarms were returned to operable status on February 21 , 1998.
3. The Reactor Operator and Plant Operator at the time of the event have discussed this event and lessons learned with all operating crews.
4. All on shift Operations Superintendents have discussed this event and have led a discussion with their respective crews regarding their responsibilities and the importance of safe operations.
5. All personnel involved have been held accountable in accordance with PSE&G's procedures and policies.

NRC FORM 366A (4-95)

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