05000275/LER-2002-005
| Event date: | |
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| 2752002005R00 - NRC Website | |
I. Plant Conditions
Unit 1 has operated in various plant modes with the described condition.
II. Description of Problem
A. Background
Technical Specification (TS) 3.4.10, "Pressurizer Safety Valves," requires that three pressurizer safety valves (PSVs)[AB][RV] shall be operable with a lift setting greater than 2460 psig and less than 2510 psig corresponding to ambient conditions of the valve at nominal operating temperature and pressure. This upper and lower pressure limit is based on a nominal pressure of 2485 psig with an upper and lower tolerance limit of ± 1%.
Surveillance Test Procedure (STP) M-77, "Safety and Relief Valve Testing," verifies the PSVs lift setting in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section Xl. The initial lift setting is evaluated for TS compliance. STP M-77 requires that the valves lift within the required tolerance in order to declare them operable.
STP M-77 test methodology obtains the as-found lift setting by placing the PSVs in an environmentally controlled enclosure and heating the ambient air to the temperature conditions typical at Diablo Canyon Power Plant (DCPP). The loop seal is also heated to simulate the piping temperature conditions at DCPP. Testing is accomplished by the addition of steam at a defined ramp rate. Steam is added until physical evidence of stem movement is visible on the remote data acquisition display screen. The data is then reviewed to ascertain "first discernible stem movement" and the pressure at which it took place.
B. Event Description
Following the Unit 2 ninth refueling outage in October 1999, the PSVs lift settings were set and verified to be within the range required by TS 3.4.10.
The PSVs were then returned to warehouse stock. During the Unit 1 tenth refueling outage in October 2000, these three PSVs were placed in service and declared operable without any additional adjustment of the lift settings until the PSVs were tested following removal in April 2002.
On June 10, 2002, two of the three Unit 1 PSVs were identified with lift setting outside TS 3.4.10 requirements. One PSV was found with a lift setting less than 2460 psig and one PSV was found with a lift setting of LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (6) OF 2002 5 3 0 0 5 0 5 7 0 0 0 5 0 Diablo Canyon Unit 1 0 2 greater than 2510 psig. The valves were found to lift 1.9 percent low and 2.6 percent high, respectively during offsite testing.
PSV lift setting repeatability has been recognized as an industry-wide problem. PG&E has participated in extensive investigative test programs, both jointly with the nuclear steam supply system vendor, Westinghouse Owners Group, and independently. The results of the industry investigations are documented in WCAP — 12910, "Pressurizer Safety Valve Set Pressure.
C. Inoperable Structures, Systems, or Components that Contributed to the
Event
None.
D. Other Systems or Secondary Functions Affected
None.
E. Method of Discovery
This condition was discovered by PG&E while performing a routine scheduled surveillance test in accordance with STP M-77.
F. Operator Actions
None.
G. Safety System Responses
None.
III. � Cause of the Problem
A. Immediate Cause
Two of three PSVs did not lift within the TS 3.4.10 tolerance.
B. Root Cause
The cause of the lift setting change has been determined to be random lift setting spread.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (I) PAGE (3) DOCKET NUMBER (2) LER NUMBER (6) OF 2002 5 0 4 0 5 0 5 7 2 0 0 0 5 0 Diablo Canyon Unit 1 0 C. � Contributory Cause None.
IV. � Assessment of Safety Consequences The limiting event for evaluating the lift setting is the loss of load analysis that requires the maximum reactor coolant system (RCS) pressure of 2750 psia not be exceeded. The RETRAN computer model was run to determine if the RCS pressure would exceed 110 percent of ASME design acceptance criteria, or 2750 psia. The analysis confirmed that the as-found set points would have maintained adequate RCS pressure relief capacity, such that the plant remained bounded by the limiting loss of load analysis provided in Final Safety Analysis Report Update, Section 15.2.7, "Loss of External Electrical Load and/or Turbine Trip." Based on this information, PG&E used the NRC's significance determination process and believes the condition had low risk significance.
Therefore, the event:
- Is of very low risk significance
- was not a Safety System Functional Failure; and
- did not adversely affect the health and safety of the public.
V. � Corrective Actions
A. Immediate Corrective Actions
The valves were disassembled, inspected, reset within tolerance, and returned to warehouse stock.
B. Corrective Actions to Prevent Recurrence
No corrective action to prevent recurrence was required because this inherent characteristic of the valve is within the analysis basis of DCPP.
VI. � Additional Information A. � Failed Components None.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (I) PAGE (3) DOCKET NUMBER (2) LER NUMBER (6) OF 2002 5 5 5 7 2 0 0 0 5 0 Diablo Canyon Unit 1 0 0 5 0 0 B. P Previous Similar Events November 7, 1995, regarding PSVs found outside TS limits during the Unit 1 sixth refueling outage. The root cause of this event was determined to be random lift setting spread. No corrective action to prevent recurrence was required because this inherent characteristic of the valve was within the analysis basis of DCPP. However, a prudent action to replace the PSV upper spring washer was recommended. The implementation of this prudent action has been deferred until NRC concerns regarding valve performance can be aeceptably resolved.
seventh refueling outage. The root cause of this event was determined to be random lift setting spread. No corrective action to prevent recurrence was required because this inherent characteristic of the valve was within the analysis basis of DCPP. However, a prudent action to replace the PSV upper spring washer was recommended. The implementation of this prudent action has been deferred until NRC concerns regarding valve performance can be acceptably resolved.
tenth refueling outage. The root cause of this event was determined to be random lift setting spread. No corrective action to prevent recurrence was required because this inherent characteristic of the valve was within the analysis basis of DCPP.