05000323/LER-2009-002
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. | |
| Event date: | 08-26-2009 |
|---|---|
| Report date: | 01-19-2010 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3232009002R01 - NRC Website | |
I. Plant Conditions
Unit 2 was in Mode 1 (Power Operation) at approximately 100 percent reactor power with normal operating reactor coolant temperature and pressure.
II. Description of Problem
A.� Background The Diablo Canyon Power Plants (DCPP) Units 1 and 2 are Pressurized Water Reactors (PWR) with four Reactor Coolant Loops (RCL)[AB] to circulate reactor coolant to each of the four steam generators (SG)[SG].
Each SG is a vertical U-tube design provided by the Nuclear Steam Supply System (NSSS) vendor, Westinghouse. Each SG has five main steam safety valves (MSSVs) for a total of twenty MSSVs, each sized in accordance with the DCPP design and ASME Code requirements.
The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs alsb provide protection against overpressurizing the reactor coolant pressure boundary (RCPB) by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the preferred heat sink, provided by the main turbine condenser and circulating water system, is not available.
Technical Specification (TS) 3.7.1, "Main Steam Safety Valves (MSSVs)," requires
- five MSSVs per SG to be operable in Modes 1, 2, and 3. If one or more MSSVs are inoperable, action to reduce the power range high neutron flux trip setpoint per Table 3.7.1-1 must be taken within 4-hours.
The MSSVs are located on each main steam header, outside the primary containment structure, upstream of the main steam isolation valves (MSIVs). The MSSVs have sufficient capacity to limit the secondary system pressure to less than 110 percent of the SG design pressure.
The MSSV design includes staggered setpoints, according to TS Table 3.7.1-2, so that only the needed MSSVs will actuate. MS-2-RV-224 has the highest required "as found" lift setting at 1115 psig (plus or minus 3 percent), and RV-11 has the lowest at 1065 psig (plus 3 percent, minus 2 percent) for SG 2-3. Staggered setpoints reduce the potential for valve chattering due to steam pressure that is insufficient to fully open all valves during an overpressure event.
11111■11■11 LICENSEE EVENT,REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) The operability of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances, to relieve SG overpressure, and reseat when pressure has been reduced. The operability of the MSSVs is verified by periodic surveillance testing in accordance with the I nservice Testing Program (1ST) that verifies the "as-left" lift setting is within plus or minus 1 percent for two successive lifts in accordance with Code requirements.
Maintenance Procedure (MP) M-4.18A, "Verification of Main Steam Safety Valves Lift Point Using Fermanite's Trevitest Equipment," performs'a semi-automated assisted valve lift methodology. The Trevitest System places an external pulling load on the valve stem in addition to the steam seat forces. When the total lift force overcomes the spring force, the valve lifts off its seat. Once the valve lifts off its seat, the added lift force is quickly released, closing the valve, and the lift setpoint is recorded.
MS-2-RV-224 was last replaced with a refurbished relief valve in March 1998. Testing is performed every other fuel cycle, and is normally performed as preoutage work. MS-2-RV-224 was last tested per TS Surveillance Requirement (SR) 3.7.1.1, in 2006 and the as-found and as-left setpoints were within specification with no adjustments required.
B.� Event Description On August 26, 2009, at 11:25 PDT, SG 2-3 surveillance testing performed in accordance MP M-4.18A identified that the RV-224 lift point was out of tolerance, with the setpoint "as found" 7 percent low.
On August 26, 2009, at 11:36 PDT, plant operators declared RV-224 inoperable and entered TS 3.7.1, Action A.1, due to the low as found lift.
On August 26, 2009, at 12:05 PDT, Engineering personnel performing Trevitesting informed the control operator (CO)/Shift Foreman (SFM) that RV-224 was adjusted within tolerance. (TS Sheet 2-TS-09-0706) On August 26, 2009, at 12:45 PDT, Engineering performing Trevitesting informed the CO/SFM that RV-224 had a crack in it's spring. Plant operators declared RV-224 inoperable due to the observed degradation.
(TS Sheet 2-TS-09-0711 for SG 2-3 safety valve RV-224) On August 26, 2009, at 13:29 PDT, plant operators commenced ramping the Unit from approximately 100 percent power to 80 percent power to LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Diablo Canyon Unit 2 0 5 0 0 0 3 2 3 2009 0 0 2 0 1 4 OF reset the Nuclear Instrument (NI) system power range high neutron flux trip setpoint in accordance with TS 3.7.1, Table 3.7.1-1.
On August 26, 2009, at 14:36 PDT, technical maintenance (TM) personnel commenced resetting the power range high flux reactor trip setpoint to 87 percent power On August 26, 2009, at 14:37 PDT, plant operators stabilized reactor power ramp at approximately 80 percent.
On August 26, 2009, at 16:06 PDT, TM personnel completed resetting the power range high flux reactor trip setpoints from 109 percent power to 87 percent reactor power completing TS Action 3.7.1.A.1.
On August 26, 2009, at 16:27 PDT, maintenance personnel placed a gag on the MSSV to preclude inappropriate opening of the valve.
C. Status of Inoperable Structures, Systems, or Components that Contributed to the Event None.
D. Other Systems or Secondary Functions Affected No additional safety systems were adversely affected by this event.
E. Method of Discovery During Trevitesting on MS-2-RV-224, an area of degradation was noted on the valve spring while making adjustments to spring load in accordance with MP M-4.18A.
F. Operator Actions Plant operators declared the MSSV inoperable, reduced reactor power and initiated actions to reset the NI system setpoints in accordance with TS Action 3.7.1.A.1.
G. Safety System Responses None.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) III.� Cause of the Problem
A. Immediate Cause
Preliminary ISI evaluation of the cracked spring concluded that the most likely failure mechanism is corrosion pitting of the spring. This is based upon high strength spring steel's low tolerance for surface flaws such as pitting, such that once a corrosion pit achieves a certain size, a spring will crack rather quickly through the thickness (cross section) of the spring.
It appears that corrosion between the spring and the bottom spring holder provided the environment for the formation of the pitting of the spring surface. Once the pit reached a critical size, the spring cracked through the cross section. Examination of the opening of the crack showed a fractured surface that is corroded, indicating that the spring has been cracked for some time.
B. Cause
A cause evaluation found the spring material to be in accordance with the original design specification. The evaluation of the fracture was hampered by the excessive amount of corrosion product on the fracture surface.
This amount of corrosion was a strong indicator that the spring had been cracked for some time and the crack had propagated over a large percentage of the spring cross section before it failed due to overload. -It was determined that the crack initiated at the edge of the spring that interfaces with the bottom spring washer; however, the, exact initiation site could not be determined.
The cause evaluation concluded that the original spring coating was inadequately maintained, which contributed to the environmentally induced corrosion initiated cracking and subsequent fracture of the valve spring.
IV.� Assessment of Safety Consequences There were no safety consequences as a result of this event.
The Unit 2 reactor was maintained in Mode 1, with TS-required equipment operable, while reactor power was decreased to allow reset of the NI system setpoints in accordance with TS Action 3.7.1.A.1.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PG&E presumes RV-224 would have lifted low, outside the TS 3.7.1 allowable lift setting limits prior to discovery, for a period longer than allowed by the TS Action.
Although the valve was capable of performing its safety function of steam pressure relief, it was inoperable, and could have resulted in an out-of-sequence MSSV valve lift. The remaining 19 MSSVs were available and capable of relieving steam pressure as necessary to protect the unit from overpressurization. An out-of-sequence MSSV lift could result in additional cooldown following a reactor trip; however, this condition is bound by a failed open MSSV, a previously analyzed condition in Final Safety Analysis Report (FSAR) Update, Section 15.2.14, "Accidental Depressurization of the Main Steam System." Therefore, the MSSV pressure relief system was capable of performing its safety function and this condition was not a safety system functional failure.
This event is not considered risk significant and it did not adversely affect the health and safety of the public.
V.C Corrective Actions
A. Immediate Corrective Actions
Plant operators entered TS 3.7.1, maintained Unit 2 in. Mode 1, and decreased reactor power to allow reset of the NI system setpoints in accordance with TS Action 3.7.1.A.1 Visual inspections were performed on all Unit 1 and Unit 2 MSSV springs. No linear indications were observed other than RV-224.
B. Corrective Actions to Prevent Recurrence (CAPR) The RV-224 valve spring was removed and sent offsite for detailed material analysis of the failure.
The RV-224 valve body will be shipped to an independent valve facility for replacement of the spring, valve refurbishment, and steam testing. Upon satisfactory completion of the offsite refurbishment and testing, the valve will be returned to warehouse spares for future use.
The MSSV Springs will be replaced on a six refueling outage cycle (6R) frequency. The new springs will have a corrosion protection coating.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Diablo Canyon Unit 2 0 5 0 0 0 3 2 3 2009 0 0 2 0 1 7 OF VI.�Additional Information
A. Failed Components
MSSV manufactured by Dresser Industries, a 6 inch ASME Code valve.
Model: 3707RAX621.
B. Previous Similar Events
None.
Industry Reports
None.