05000275/LER-2015-001, Regarding Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld
| ML15061A548 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/02/2015 |
| From: | Welsch J Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-15-033 LER 15-001-00 | |
| Download: ML15061A548 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 2752015001R00 - NRC Website | |
text
Pacific Gas and Electric Company March 2, 2015 PG&E Letter DCL-15-033 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 James M. Welsch Site Vice President 10 CFR 50.73 Licensee Event Report 2015-001-00. Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld
Dear Commissioners and Staff:
Diablo Canyon Power Plant Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 805.545.3242 Internal: 691.3242 Fax: 805.545.4884 Pacific Gas and Electric Company (PG&E) submits the enclosed Licensee Event Report (LER) regarding an event or condition that could have prevented fulfillment of a safety function when both trains of the residual heat removal system were declared inoperable due to a circumferential crack on a socket weld. PG&E is submitting this LER in accordance with 10 CFR 50.73(a)(2)(v), and 50.73(a)(2)(vii).
This is the initial LER submittal. PG&E will submit a supplemental LER describing the event cause and corrective actions no later than May 8, 2015.
PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this report.
This event did not adversely affect the health and safety of the public.
Sincerely,
~~
~e~ M. Welsch APHS/6470/50680117 Enclosure cc\\enc:
Marc L. Dapas, NRC Region IV Administrator Thomas R. Hipschman, NRC Senior Resident Inspector Siva P. Lingam, NRR Project Manager IN PO Diablo Distribution A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Diablo Canyon
- Palo Verde
- Wolf Creek
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
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, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
YEAR 2015
- 6. LER NUMBER I
SEQUENTIAL I NUMBER 001 REV NO.
00 2
- 3. PAGE OF 4
10 CFR 50.73(a)(2)(v)(B&D), "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: Remove Residual Heat and Mitigate the consequences of an accident" 10 CFR 50. 73(a)(2)(vii), "Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to: Remove Residual Heat, and Mitigate the consequences of an accident"
II. Plant Conditions
At the time of the event, Diablo Canyon Power Plant (DCPP) Unit 1 was in Mode 3 (Hot Standby) at normal operating reactor coolant temperature and pressure conditions.
Ill. Problem Description A. Background The function of the emergency core cooling system (ECCS) is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:
a) loss-of-coolant accident, non-isolable coolant leakage greater than the capability of the normal charging system [CB]
b) rod ejection accident c) loss-of-secondary-coolant accident, including uncontrolled steam release or loss of feedwater and d) steam generator tube rupture. *.
The addition of negative reactivity is designed primarily for the loss-of-secondary-coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant reactor power. The ECCS consists of three separate subsystems: centrifugal charging (high head) [BQ],
safety injection (intermediate head), and residual heat removal (RHR) (low head) [BP]. Each subsystem consists of 2 redundant, 100 percent capacity trains.
The design function of relief valve [RV] RHR-1-RV-8708 is to protect the RHR discharge piping from exceeding its design pressure rating. The inlet pipe to the valve is connected to a 12-inch RHR header line, which provides a flow path for injection to reactor coolant system (RCS) [AB] hot legs 1 and 2. This line is occasionally used to fill the reactor cavity during refueling outages. The normal flow path for shutdown cooling (Modes 4 and 5) does not use this line.
B. Event Description
On December 31, 2014, while performing a walkdown as part of a surveillance test procedure, plant personnel identified an accumulation of boric acid on the inlet pipe to relief valve RHR-1-RV-8708. The problem was entered into the corrective action program. Subsequent cleanup of the boric acid accumulation revealed active seepage of 30 drops per minute (dpm). A visual inspection identified a circumferential crack on the socket weld.
The active boric acid leak was located on the common header from the RHR pump [P] discharge to RCS hot legs 1 and 2. The active boric acid seepage could not be isolated. Both trains of the RHR system were declared inoperable and Technical Specification (TS) 3.0.3 was entered on December 31, 2014, at 1105 PST. DCPP made an 8-hour notification to the NRC regarding an event or condition that could have prevented fulfillment of a safety function (NRC Event Notification 50711 ).
C. Status of Inoperable Structure, Systems, or Components That Contributed to the Event None.
D. Other Systems or Secondary Functions Affected
None.
E. Method of Discovery
On December 31, 2014, while performing a walkdown as part of a surveillance test procedure, plant personnel identified an accumulation of boric acid on the inlet pipe to relief valve RHR-1-RV-8708.
Subsequent cleanup of the boric acid accumulation revealed active seepage of 30 dpm. A visual inspection identified that the source of the leak was a circumferential crack on the socket weld.
F. Operator Actions
Operators declared both RHR trains inoperable. TS 3.0.3 was entered on December 31, 2014, at 1105 PST and was exited at 2256 PST, when the plant entered Mode 4. Additionally, in accordance with TS 3.6.3.C, the associated containment penetration flow path was isolated.
G. Safety System Responses None.
IV. Cause of the Problem Pacific Gas and Electric (PG&E) is conducting a root cause evaluation (RCE) and will submit a supplemental Licensee Event Report (LER) documenting the results of this investigation once it is complete.
V. Assessment of Safety Consequences
DCPP assessed the Unit 1 risk significance of seepage in the vent line of the RHR supply line to the hot leg 1 and 2 supply lines. This assessment concluded that the incremental conditional core damage probability was less than 1.0E-06 per year. Therefore, this event is not considered risk significant and did not adversely affect the health and safety of the public.
VI. Corrective Actions
A. Immediate Actions
- 1. Performed piping repair on cracked socket weld.
- 2. Installed permanent pipe support.
B. Other Corrective Actions Once the associated RCE is complete, PG&E will take corrective actions as necessary, and describe the additional corrective actions in the supplemental LER.
VII.
Additionallnformation A. Failed Components DCPP discovered a circumferential crack on the socket weld on the inlet to relief valve RHR-1-RV-8708.
B. Previous Similar Events
"Unit 1 Licensee Event Report 2013-005-00, Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld," dated August 22, 2013, is similar to this event.
A. Industry Reports None.