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{{#Wiki_filter:r Regional Administrator U.S. Nuclear Regulatory Commission 1 475 Allendale King of Prussia, PA 19406-1415 Gentlemen:
{{#Wiki_filter:r JAN 18 1999 LR-N990036 Regional Administrator U.S. Nuclear Regulatory Commission Regio~ 1 475 Allendale Ro~d King of Prussia, PA 19406-1415 Gentlemen:
JAN 18 1999 LR-N990036 LICENSEE EVENT REPORT 311/98-016-00 SALEM GENERATING STATION-UNIT2 FACILITY OPERA TING LICENSE NO DPR 75 DOCKET NO. 50-311 This Licensee Event Report entitled "ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE" is being submitted in accordance with the requirements of 1OCFR50.73 (a)(2)(ii)(B).
LICENSEE EVENT REPORT 311/98-016-00 SALEM GENERATING STATION-UNIT2 FACILITY OPERA TING LICENSE NO DPR 75 DOCKET NO. 50-311 This Licensee Event Report entitled "ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE" is being submitted in accordance with the requirements of
Attachment
* 10CFR50.73 (a)(2)(ii)(B).
/JCN Sincerely, A.C. Bakken, Ill General Salem Operations c U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Distribution:
Sincerely, A.C. Bakken, Ill General Manager-Salem Operations Attachment
LER File 3.7 9901290173 990118' ADOCK 0500031l PDR * 
  /JCN c     U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Distribution:
\. NRC FORM 366 [j.S. NCCLEAR REGULA (6-1998) LICENSEE EVENT REPORT (LER) (See reverse for required number of digits/characters for each block) FACILITY NAME (1) . COMMISSION SALEM UNIT 2 TITLE(4) ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE. APPROVED B B NO. 3150-0104 EXPIRES 06/30/2001 Estimated burden per response to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into the licensing process and fed back to industry.
LER File 3.7 9901290173 990118'
Forward comments regarding burden estimate to the Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Paperwork Reduction Project (3150-0104), Office of Management and Budget. Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
    ~DR ADOCK 0500031l PDR
DOCKET NUMBER (2) 05000311 PAGE (3) 1 OF 3 E ENT DAT (5) REP R ATE(7) MONTH DAY YEAR YEAR MONTH DAY YEAR 12 19 98 98 -016 00 01 99 OPERATING CAUSE SYSTEM YES (If yes, complete EXPECTED SUBMISSION DATE) .. ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) FACILITY NAME EXPECTED DOCKET NUMBER DOCKET NlJl>:IBER MONTH "DAY
 
* REPORTABLE TO EPIX YEAR On 12/19/98, Operations personnel determined that there was leakage into the RHR sump that exceeded the maximum allowable ECCS leakage from sources outside Containment.
\. NRC FORM 366           [j.S. NCCLEAR REGULA         . COMMISSION                            APPROVED B            B NO. 3150-0104 EXPIRES 06/30/2001 (6-1998)
The leakage path was determined to be from a three quarter inch manual pump casing drain valve (21RH34) on the 21 residual heat removal loop. The valve was reseated upon discovery and ECCS leakage returned to within acceptable limits. The cause of the event was the leakage past the seat of the 21RH34 manual drain valve. Subsequent investigation was unable to determine how the .valve was positioned off of it's seat but it is believed to have occurred during the performance of a surveillance on the day before the event. This valve is not normally operated as part of the surveillance.
Estimated burden per response to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into LICENSEE EVENT REPORT (LER)                                                                 the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 F33), U.S.
The leak posed no actual safety consequences to the health and safety of the public or to the plant staff. The source of the leaking water was from the refueling water storage tank (RWST) . Estimated potential post accident offsite doses remain within lOCFRlOO limits. Estimated post accident onsite beta skin dose, whole body and thyroid doses are estimated to remain within the GDC 19 limits, based on conservative estimates.
Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the (See reverse for required number of                                                          Paperwork Reduction Project (3150-0104), Office of Management and digits/characters for each block)                                                            Budget. Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
This event is reportable pursuant to 10CFR50. 73 (a) 2 (ii) "any event or condition  
FACILITY NAME (1)                                                                           DOCKET NUMBER (2)                                PAGE (3)
...... that resulted in the nuclear power plant being: ...... (B) in a condition that was outside the design basis of the plantn.
SALEM UNIT 2                                                      05000311                                        1 OF 3 TITLE(4)
NRC FORM 366A (6-1998) U.S. NUCLEAR REGULA TORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (I) SALEM UNIT2 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17) PLANT AND SYSTEM IDENTIFICATION DOCKET(2)
ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE.
NUMBER (2) 05000311 Westinghouse
E ENT DAT (5)                                                   REP R       ATE(7)
-Pressurized Water Reactor Residual He.at Removal/Drain Valve {BP/V}* LER NUMBER (6) PAGE (3) YEAR I SEQUENTIAL I REVISION 2 NUMBER NUMBER OF 3 98 0 1 6 00
MONTH       DAY     YEAR     YEAR                               MONTH     DAY         YEAR                                                 DOCKET NUMBER 12       19       98       98 -016               00         01                   99         FACILITY NAME                            DOCKET NlJl>:IBER OPERATING CAUSE         SYSTEM                                                                                                                                 REPORTABLE TO EPIX EXPECTED              MONTH        "DAY          YEAR YES (If yes, complete EXPECTED SUBMISSION DATE) ..
* Energy Industry Identification System {EIIS} codes and component function identifier codes appear as (SS/CCC) CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 2 was operating at 100% Power;
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
* DESCRIPTION OF OCCURRENCE On 12/21/98 Operations personnel identified an increase in the leakage into the Residual Heat Removal System sump which was excessive.
On 12/19/98, Operations personnel determined that there was leakage into the RHR sump that exceeded the maximum allowable ECCS leakage from sources outside Containment. The leakage path was determined to be from a three quarter inch manual pump casing drain valve (21RH34) on the 21 residual heat removal loop.                                                                           The valve was reseated upon discovery and ECCS leakage returned to within acceptable limits.
An investigation was initiated to identify the source of leakage. The leakage was determined to be approximately 0.5 gpm through a manual pump casing drain valve (21RH34)'.
The cause of the event was the leakage past the seat of the 21RH34 manual drain valve.           Subsequent investigation was unable to determine how the .valve was positioned off of it's seat but it is believed to have occurred during the performance of a surveillance on the day before the event. This valve is not normally operated as part of the surveillance. The leak posed no actual safety consequences to the health and safety of the public or to the plant staff.                                                                               The source of the leaking water was from the refueling water storage tank (RWST) .
The leakage into the RHR sump in the Auxiliary Building exceeded the maximum allowable ECCS Leakage as defined by UFSAR Section 6.3.2.11.
Estimated potential post accident offsite doses remain within 10CFRlOO limits.
UFSAR Section 6.3.2.11 addresses leakage during the recirculation phase 6f an accident, and states that "The total leakage resulting from all sources is about 3800 cc/hour as described in UFSAR Section 15.4.1. Recirculation loop leakage sources are summarized in Table 6.3-12. Leakage is monitored by procedure to ensure this leak rate is not exceeded." During the investigation into the source of the leakage, it was determined that the drain valve (21RH34) was open approximately one-third turn from full closed. Upon closure of the valve the leakage was reduced by about 0.5 gpm. CAUSE OF OCCURRENCE The cause of the event was attributed to leakage past the seat of the 21RH34 RHR drain valve. The manner in which the valve came to be unseated can not be determined.
Estimated post accident onsite beta skin dose, whole body and thyroid doses are estimated to remain within the GDC 19 limits, based on conservative estimates. This event is reportable pursuant to 10CFR50. 73 (a) 2 (ii) "any event or condition ...... that resulted in the nuclear power plant being: ...... (B) in a condition that was outside the design basis of the plantn.
However, the drain valve may have been inadvertently bumped during the conduct of the RHR Pump 4.0.5P In Service Test (IST) Surveillance which was performed on the day before. NRC FORM 366A (6-1998)
 
\ .. . 0 NRC FORM 366A (6-1998) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) SALEM UNIT 2 TEXT (If more space is required.
NRC FORM 366A               U.S. NUCLEAR REGULA TORY COMMISSION (6-1998)
use additional copies ofNRC Fonn 366A) (17) PRIOR SIMILAR OCCURRENCES DOCKET(2)
LICENSEE EVENT REPORT (LER)
NUMBER(2) 05000311 LER NUMBER (6) PAGE (3) YEAR I SEQUENTIAL I REVISION 3 NUMBER NUMBER OF 3 98 0 1 6 00 1996, 1997 and 1998 LERs were reviewed for similar occurrences.
TEXT CONTINUATION FACILITY NAME (I)                                 DOCKET(2)      LER NUMBER (6)           PAGE (3)
LER 311/98-010-00 reported excess leakage through a Boron Injection Tank sample valve which exceeded the UFSAR limit. SAFETY CONSEQUENCES AND IMPLICATIONS The source of the leaking water was the RHR system that is maintained full with. water from the refueling water storage tank (RWST). Therefore, during the period of leakage, there were no safety consequences to the health and safety of the public or to the plant staff. In an effort to assess the potential safety consequences, dose were conducted.
NUMBER (2)
If the system had been filled with reactor coolant at the Technical Specification activity limit the consequences would have remained minimal. The estimates were re-performed for the accident case using the actual leakage through 21RH34 (0.5 gpm) determined by the RCS leak rate program in addition to the design basis leakage. The analysis assumes that RHR sump vent is filtered once the ventilation system begins to operate. Offsite accident doses for whole body, thyroid and beta skin for the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) while greater than the design basis, were well within the Part 100 limits. Potential onsite whole body, thyroid and beta skin doses were estimated to remain within the General Design Criterion (GDC) 19 limit. The potential consequences of the event depend on the core damage frequency, which for Salem is 4.45xE-5/year.
SALEM UNIT2                                      05000311    YEAR I SEQUENTIAL NUMBER IREVISION NUMBER 2  OF     3 98   0   1   6     00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)
While it is impossible to exactly identify a time when the 21RH34 leakage became a significant of the RCS unidentified leakage, it appears that the valve positioned off of its seat during the conduct of surveillance conducted on the prior day. If the period of the leakage is increased to 24 hours, the core damage frequency for the period is approximately 1.23 E-7. Therefore, the potential consequences are insignificant.
PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Residual He.at Removal/Drain Valve {BP/V}*
* Energy Industry Identification System {EIIS} codes and component function identifier codes appear as (SS/CCC)
CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 2 was operating at 100% Power; DESCRIPTION OF OCCURRENCE
* On 12/21/98 Operations personnel identified an increase in the leakage into the Residual Heat Removal System sump which was excessive. An investigation was initiated to identify the source of leakage. The leakage was determined to be approximately 0.5 gpm through a manual pump casing drain valve (21RH34)'.           The leakage into the RHR sump in the Auxiliary Building exceeded the maximum allowable ECCS Leakage as defined by UFSAR Section 6.3.2.11.
UFSAR Section 6.3.2.11 addresses leakage during the recirculation phase 6f an accident, and states that "The total leakage resulting from all sources is about 3800 cc/hour as described in UFSAR Section 15.4.1. Recirculation loop leakage sources are summarized in Table 6.3-12.                                         Leakage is monitored by procedure to ensure this leak rate is not exceeded."
During the investigation into the source of the leakage, it was determined that the drain valve (21RH34) was open approximately one-third turn from full closed. Upon closure of the valve the leakage was reduced by about 0.5 gpm.
CAUSE OF OCCURRENCE The cause of the event was attributed to leakage past the seat of the 21RH34 RHR drain valve. The manner in which the valve came to be unseated can not be determined.                   However, the drain valve may have been inadvertently bumped during the conduct of the RHR Pump 4.0.5P In Service Test (IST) Surveillance which was performed on the day before.
NRC FORM 366A (6-1998)
 
\..
.0 NRC FORM 366A               U.S. NUCLEAR REGULATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)                                 DOCKET(2)      LER NUMBER (6)           PAGE (3)
NUMBER(2)
SALEM UNIT 2                                      05000311    YEAR   SEQUENTIAL IREVISION 3 OF      3 I NUMBER     NUMBER 98   0 1     6     00 TEXT (If more space is required. use additional copies ofNRC Fonn 366A) (17)
PRIOR SIMILAR OCCURRENCES 1996, 1997 and 1998 LERs were reviewed for similar occurrences. LER 311/98-010-00 reported excess leakage through a Boron Injection Tank sample valve which exceeded the UFSAR limit.
SAFETY CONSEQUENCES AND IMPLICATIONS The source of the leaking water was the RHR system that is maintained full with. water from the refueling water storage tank (RWST). Therefore, during the period of leakage, there were no safety consequences to the health and safety of the public or to the plant staff.
In an effort to assess the potential safety consequences, dose estimat~s were conducted. If the system had been filled with reactor coolant at the Technical Specification activity limit the consequences would have remained minimal.           The estimates were re-performed for the accident case using the actual leakage through 21RH34 (0.5 gpm) determined by the RCS leak rate program in addition to the design basis leakage.                                       The analysis assumes that RHR sump vent is filtered once the ventilation system begins to operate.
Offsite accident doses for whole body, thyroid and beta skin for the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) while greater than the design basis, were well within the Part 100 limits.
Potential onsite whole body, thyroid and beta skin doses were estimated to remain within the General Design Criterion (GDC) 19 limit.
The potential consequences of the event depend on the core damage frequency, which for Salem is 4.45xE-5/year. While it is impossible to exactly identify a time when the 21RH34 leakage became a significant co~ponent of the RCS unidentified leakage, it appears that the valve positioned off of its seat during the conduct of surveillance conducted on the prior day. If the period of the leakage is increased to 24 hours, the core damage frequency for the period is approximately 1.23 E-7. Therefore, the potential consequences are insignificant.
CORRECTIVE ACTIONS The 21RH34 valve was reseated, isolating the leak and restoring ECCS leakage from systems outside containment to within the USFAR design limits. The plant staff will be briefed on this event.}}
CORRECTIVE ACTIONS The 21RH34 valve was reseated, isolating the leak and restoring ECCS leakage from systems outside containment to within the USFAR design limits. The plant staff will be briefed on this event.}}

Latest revision as of 04:49, 3 February 2020

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
ML18106B040
Person / Time
Site: Salem PSEG icon.png
Issue date: 01/18/1999
From: Bakken A, Nagle J
Public Service Enterprise Group
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
LER-98-016, LER-98-16, LR-N990036, NUDOCS 9901290173
Download: ML18106B040 (4)
v • d • e


Text

r JAN 18 1999 LR-N990036 Regional Administrator U.S. Nuclear Regulatory Commission Regio~ 1 475 Allendale Ro~d King of Prussia, PA 19406-1415 Gentlemen:

LICENSEE EVENT REPORT 311/98-016-00 SALEM GENERATING STATION-UNIT2 FACILITY OPERA TING LICENSE NO DPR 75 DOCKET NO. 50-311 This Licensee Event Report entitled "ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE" is being submitted in accordance with the requirements of

Sincerely, A.C. Bakken, Ill General Manager-Salem Operations Attachment

/JCN c U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Distribution:

LER File 3.7 9901290173 990118'

~DR ADOCK 0500031l PDR

\. NRC FORM 366 [j.S. NCCLEAR REGULA . COMMISSION APPROVED B B NO. 3150-0104 EXPIRES 06/30/2001 (6-1998)

Estimated burden per response to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into LICENSEE EVENT REPORT (LER) the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 F33), U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the (See reverse for required number of Paperwork Reduction Project (3150-0104), Office of Management and digits/characters for each block) Budget. Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

SALEM UNIT 2 05000311 1 OF 3 TITLE(4)

ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE.

E ENT DAT (5) REP R ATE(7)

MONTH DAY YEAR YEAR MONTH DAY YEAR DOCKET NUMBER 12 19 98 98 -016 00 01 99 FACILITY NAME DOCKET NlJl>:IBER OPERATING CAUSE SYSTEM REPORTABLE TO EPIX EXPECTED MONTH "DAY YEAR YES (If yes, complete EXPECTED SUBMISSION DATE) ..

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

On 12/19/98, Operations personnel determined that there was leakage into the RHR sump that exceeded the maximum allowable ECCS leakage from sources outside Containment. The leakage path was determined to be from a three quarter inch manual pump casing drain valve (21RH34) on the 21 residual heat removal loop. The valve was reseated upon discovery and ECCS leakage returned to within acceptable limits.

The cause of the event was the leakage past the seat of the 21RH34 manual drain valve. Subsequent investigation was unable to determine how the .valve was positioned off of it's seat but it is believed to have occurred during the performance of a surveillance on the day before the event. This valve is not normally operated as part of the surveillance. The leak posed no actual safety consequences to the health and safety of the public or to the plant staff. The source of the leaking water was from the refueling water storage tank (RWST) .

Estimated potential post accident offsite doses remain within 10CFRlOO limits.

Estimated post accident onsite beta skin dose, whole body and thyroid doses are estimated to remain within the GDC 19 limits, based on conservative estimates. This event is reportable pursuant to 10CFR50. 73 (a) 2 (ii) "any event or condition ...... that resulted in the nuclear power plant being: ...... (B) in a condition that was outside the design basis of the plantn.

NRC FORM 366A U.S. NUCLEAR REGULA TORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (I) DOCKET(2) LER NUMBER (6) PAGE (3)

NUMBER (2)

SALEM UNIT2 05000311 YEAR I SEQUENTIAL NUMBER IREVISION NUMBER 2 OF 3 98 0 1 6 00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Residual He.at Removal/Drain Valve {BP/V}*

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

CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 2 was operating at 100% Power; DESCRIPTION OF OCCURRENCE

  • On 12/21/98 Operations personnel identified an increase in the leakage into the Residual Heat Removal System sump which was excessive. An investigation was initiated to identify the source of leakage. The leakage was determined to be approximately 0.5 gpm through a manual pump casing drain valve (21RH34)'. The leakage into the RHR sump in the Auxiliary Building exceeded the maximum allowable ECCS Leakage as defined by UFSAR Section 6.3.2.11.

UFSAR Section 6.3.2.11 addresses leakage during the recirculation phase 6f an accident, and states that "The total leakage resulting from all sources is about 3800 cc/hour as described in UFSAR Section 15.4.1. Recirculation loop leakage sources are summarized in Table 6.3-12. Leakage is monitored by procedure to ensure this leak rate is not exceeded."

During the investigation into the source of the leakage, it was determined that the drain valve (21RH34) was open approximately one-third turn from full closed. Upon closure of the valve the leakage was reduced by about 0.5 gpm.

CAUSE OF OCCURRENCE The cause of the event was attributed to leakage past the seat of the 21RH34 RHR drain valve. The manner in which the valve came to be unseated can not be determined. However, the drain valve may have been inadvertently bumped during the conduct of the RHR Pump 4.0.5P In Service Test (IST) Surveillance which was performed on the day before.

NRC FORM 366A (6-1998)

\..

.0 NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET(2) LER NUMBER (6) PAGE (3)

NUMBER(2)

SALEM UNIT 2 05000311 YEAR SEQUENTIAL IREVISION 3 OF 3 I NUMBER NUMBER 98 0 1 6 00 TEXT (If more space is required. use additional copies ofNRC Fonn 366A) (17)

PRIOR SIMILAR OCCURRENCES 1996, 1997 and 1998 LERs were reviewed for similar occurrences. LER 311/98-010-00 reported excess leakage through a Boron Injection Tank sample valve which exceeded the UFSAR limit.

SAFETY CONSEQUENCES AND IMPLICATIONS The source of the leaking water was the RHR system that is maintained full with. water from the refueling water storage tank (RWST). Therefore, during the period of leakage, there were no safety consequences to the health and safety of the public or to the plant staff.

In an effort to assess the potential safety consequences, dose estimat~s were conducted. If the system had been filled with reactor coolant at the Technical Specification activity limit the consequences would have remained minimal. The estimates were re-performed for the accident case using the actual leakage through 21RH34 (0.5 gpm) determined by the RCS leak rate program in addition to the design basis leakage. The analysis assumes that RHR sump vent is filtered once the ventilation system begins to operate.

Offsite accident doses for whole body, thyroid and beta skin for the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) while greater than the design basis, were well within the Part 100 limits.

Potential onsite whole body, thyroid and beta skin doses were estimated to remain within the General Design Criterion (GDC) 19 limit.

The potential consequences of the event depend on the core damage frequency, which for Salem is 4.45xE-5/year. While it is impossible to exactly identify a time when the 21RH34 leakage became a significant co~ponent of the RCS unidentified leakage, it appears that the valve positioned off of its seat during the conduct of surveillance conducted on the prior day. If the period of the leakage is increased to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the core damage frequency for the period is approximately 1.23 E-7. Therefore, the potential consequences are insignificant.

CORRECTIVE ACTIONS The 21RH34 valve was reseated, isolating the leak and restoring ECCS leakage from systems outside containment to within the USFAR design limits. The plant staff will be briefed on this event.

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