05000311/LER-2005-001
| Docket Number Sequential Revmonth Day Year Year Month Day Yearnumber No. | |
| Event date: | 03-24-2005 |
|---|---|
| Report date: | 5-23-2005 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3112005001R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
Westinghouse - Pressurized Water Reactor Charging/Safety Injection System / Manual Discharge Valve— {CB/ V} * * Energy Industry Identification System (EIIS) codes and component function identifier codes appear in the text as {SS/CCC}.
IDENTIFICATION OF OCCURRENCE
Event Date: March 24, 2005 Discovery Date: March 24, 2005
CONDITIONS PRIOR TO OCCURRENCE
Salem Unit 2 was in Mode 1 (POWER OPERATION) at approximately 100% power at the time of the event. No structures, systems or components were inoperable at the time of the occurrence that contributed to the event.
DESCRIPTION OF OCCURRENCE
The 23 Chemical and Volume Control system positive displacement pump (PDP) had recently been tagged out for planned maintenance. This work required using a freeze seal on the discharge line to compensate for the known leaking manual pump isolation valve 2CV64 {CB/ V} and to prevent potential leakage to the Refueling Water Storage Tank. On March 24, 2005 day shift, tags were released to perform a fill and vent of the pump. The fill and vent procedure directed opening the CVCS unit cross tie valve 2CV462.
On March 24, 2005 at 2015 hours0.0233 days <br />0.56 hours <br />0.00333 weeks <br />7.667075e-4 months <br />, during routine surveillance to determine Reactor Coolant System (RCS) leakage, a licensed Reactor Operator (RO) observed a decreasing Volume Control Tank (VCT) level corresponding to a RCS leakage of approximately 0.5 gpm. After expanding the time scale on the plant computer, it was determined that the initiation time for the indicated leak corresponded to the opening of the Chemical and Volume Control System cross tie valve 2CV462 for the fill and vent procedure for restoration of the 23 CVCS Positive Displacement pump (PDP) which was still in progress.
An abbreviated one-hour RCS leak rate determined that the leakage was approximately 0.34 gpm, which was within Technical Specification 3.4.7.2 limits for unidentified leakage. The Primary Operator was dispatched to 23 PDP pump to investigate for obvious signs of leakage; none were found. The PDP suction pressure was determined to be 33 psig as read on a local pressure gage; the PDP discharge pressure indicated approximately 0 psig as read on a local and remote gage. When the operator closed valve 2CV462, the VCT level stabilized, indicating leakage was isolated. A subsequent full 3-hour leak rate resulted in an RCS leak rate of 0.055 gpm, which was consistent with previous leak rate determinations.
DESCRIPTION OF OCCURRENCE (contd) The leakage exceeded the Updated Final Safety Analysis Report (UFSAR) Section 6.3.2.11 limit for emergency core cooling system (ECCS) leakage outside the containment (3800 cc/hr). Because this leakage could not have been isolated from the recirculation flow path during the recirculation phase of the design basis Loss of Coolant Accident (LOCA), the assumptions made in the dose analysis calculations were exceeded.
The dose analysis assumption for ECCS leakage outside containment ensures that following a LOCA the radioactive releases will remain within the requirements of 10CFR100 for offsite releases and 10CFR50 Appendix A General Design Criterion 19 (GDC-19) for exposure to Control Room Operators.
Although there is sufficient margin between the current dose analysis and 10CFR100 limits, the GDC-19 limits for exposure to the Control Room Operators with the identified leakage could not be demonstrated. Accordingly, an 8-hour notification was made to the NRC. This event is reportable in accordance with 10CFR50.73(a)(2)(v), "any event or condition that could have prevented the fulfillment of the safety function of structures or system that are needed to: (C) control the release of radioactive material; or (D) mitigate the consequences of an accident.
CAUSE OF OCCURRENCE
The failure to perform an in-depth review of the known 2CV64 leakage and its impact on leakage outside containment resulted in the discharge of RCS water to the RWST.
PREVIOUS OCCURRENCES
Salem and Hope Creek Generating Station LERs for years 2002 through 2005 were reviewed for similar occurrences of ECCS leakage outside containment.
Analysis Limits (11 RHR Heat Exchanger). The apparent cause was determined to be inadequate torquing of the flange during a previous refueling outage.
Analysis Limits (23 Charging Pump). The apparent cause for this event was excessive leakage due to the failure of the 2CV64 to provide full isolation of the pump during maintenance. The valve repair required a unit outage. Valve 2CV64 was repaired during refueling outage 2R14 (March/April 2005).
The corrective actions taken were appropriate and specific for these events; but they would not have been expected to prevent this occurrence.
SAFETY CONSEQUENCES AND IMPLICATIONS
There were no safety consequences associated with this event.
As stated above, the additional leakage would not have exceeded the limits of 10CFR100 for offsite releases. However, the limits of GDC-19 for exposure of the Control Room Operators could have been exceeded had a LOCA occurred while this leakage existed as determined by a review of the LOCA dose analysis.
The LOCA dose analysis calculation is a conservative model used to determine the effect of the radioactive release to the control room operators. This model does not assume any compensatory measures are taken by the operators to reduce their exposure to the radioactive release beyond the control room emergency air conditioning system aligning to its post-accident configuration.
However, during an emergency condition (LOCA), Emergency Preparedness procedures call for providing guidance to emergency response personnel for administration of potassium iodide (KI) during an emergency at Hope Creek or Salem Generating Station. Appendix B of NEI 99-03, Control Room Habitability Assessment Guidance, Revision 1, indicates that a factor of 10 reduction in thyroid dose may be credited due to the administration of potassium iodide (KI). Therefore, with administration of KI, the thyroid dose guideline for control room personnel and other plant personnel would not have been exceeded in the high leakage conditions.
Based on the above, there was no impact to the health and safety of the public.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in the Nuclear Energy Institute (NEI) 99-02 did occur.
CORRECTIVE ACTIONS:
1. Valve 2CV64 was repaired during refueling outage 2R14 (March/April 2005).
2. Operations (OPS) Work Week Management will change the OPS work-week procedures to ensure ECCS leakage work/tags are properly identified in the schedule and reviewed.
3. The Inservice Testing (1ST) Program manager will incorporate 2CV462 and the applicable IST requirements to the IST manual. This also applies to the 1CV462.
4. Update the Charging Pump operating procedure pre job briefs to include this event as Operating Experience.
COMMITMENTS
This LER contains no commitments.