05000282/LER-2003-001, Residual Heat Removal Valve CV-31236 Positioner Linkage Found Broken
| ML031530373 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/23/2003 |
| From: | Solymossy J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-03-048 LER 03-001-00 | |
| Download: ML031530373 (11) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
| 2822003001R00 - NRC Website | |
text
cI NMCe
.CommlKtedfo NudjearExcellncef Praire Island Nuclear Generatng Plant Operated by Nuclear Management Company, LLC L-PI-03-048 10 CFR 50.73 May 23, 2003 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET 50-282 LICENSE No. DPR-42 LER 1-03-01, RESIDUAL HEAT REMOVAL VALVE CV-31236 POSITIONER LINKAGE FOUND BROKEN The Licensee Event Report for this occurrence is attached. This letter and the attached report contain no new NRC commitments and no revisions to existing commitments.
Please contact u require additional information related to this event.
Josep M. Solymossy Site e-President, P rie IsI n Nuclear Generating Plant CC Regional Administrator, USNRC, Region IlIl Project Manager, Prairie Island Nuclear Generating Plant, USNRC, NRR NRC Resident Inspector - Prairie Island Nuclear Generating Plant Glenn Wilson, State of Minnesota Attachment 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2001 (1-2001)
COMMISSION
, 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) 05000 282 Praire Island Nuclear Generating Plant Unit 1 1 OF 10 TITLE (4)
Residual Heat Removal Valve CV-31236 Positioner Linkage Found Broken EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
SEQUENTIAL REV l
FACILITY NAME DOCKET NUMBER MO DAY YEAR YEAR NUMBER NO MO DAY YEAR 03 24 03 03 01 00 05 23 03 FACILITY NAME DOCKET NUMBER OPERATING 1
THIS REPORT IS SUBMIlTED PURSUANT TO THE REQUIREMENTS OF 10 CF lj(Ceck all that apply) (11)
MODE (9) 20.2201 (b) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)
POWER 100 20.2201 (d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LEVEL (10) 100 20.2203(a)(1) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 73.71(a)(4) 20.2203(a)(2)(i) 50.36(c)(1 )(ii)(A) 50.73(a)(2)(v)(A) 73.71 (a)(5) 20.2203(a)(2)(ii) 50.36(c)(2) 50.73(a)(2)(v)(B)
OTHER 20.2203(a)(2)(iii) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C)
Specify In Abstract below or In NRC 20.2203(a)(2)(iv) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D)
(if more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Forn 366A) (17) outlet control valve is used to control the cooldown rate of the reactor coolant system9 (RCS) by controlling the flow rate through Train B RHR heat exchanger` 0. During normal operation, the valve is in the open position to allow full flow through Train B RHR heat exchanger for low head injection. The line from Train B RHR heat exchanger to Train B SI pump11 for high head recirculation taps off upstream of the Train B RHR heat exchanger outlet control valve so the valve is not required for high head recirculation. When this valve is not full open, a light12 illuminates on the "SI Not Ready" panel13 in the control room, which indicates that the system is not in its safeguards configuration for plant operation in Modes 1, 2, and 3.
Initiating Events Requiring the Train B RHR Heat Exchanger Outlet Control Valve Function Since the RHR supply to the SI pump suction is upstream of the Train B RHR heat exchanger outlet control valve, this valve is only required to perform its safeguards function during transients and accidents that require RHR low head injection. The Train B RHR heat exchanger outlet control valve does not provide any function for high head recirculation so these initiating events are not affected by this valve's failure.
Because of the functions provided by the Train B RHR heat exchanger outlet control valve, it is required for the RHR system to mitigate the consequences of: (1) loss of coolant accidents (LOCAs) (including large break and small break LOCAs); and (2) steam generator14 tube rupture (SGTR) accidents.
Following a small break LOCA initiating event with successful SI system15 injection, the Control Room Operator can cooldown and depressurize the RCS to temperatures and pressures where the RHR system can be used to remove decay heat, thereby precluding the need for high head recirculation. The Train B RHR heat exchanger outlet control valve provides a function in this case as it can be used to control the RCS cooldown rate through Train B RHR heat exchanger. If SI system injection should fail, the RCS can be depressurized such that low head injection together with low head recirculation can prevent core damage. In this case, the Train B RHR heat exchanger outlet control valve allows full low head RHR flow to the reactor vessel16.
9 EIIS System Identifier: AB 10 EIIS System Identifier: BP Component Identifier: HX EIIS System Identifier: BQ Component Identifier: P 12 EIIS System Identifier: IB Component Identifier: IL 13 EIIS System Identifier: IB 14 EIIS System Identifier: SB Component Identifier: SD is EIIS System Identifier: BQ 16 EIIS System Identifier: AB Component Identifier: RPV (If more space is required, use additional copies of NRC Forn 366A) (17)
For large break LOCAs, the functions provided by the Train B RHR heat exchanger outlet control valve are the only means to prevent core damage as low head injection together with low head recirculation are required to mitigate the accident.
Following a SGTR, SI system injection is needed to prevent core uncovery. SI pumps will be sequentially stopped and RHR shutdown cooling would be needed to provide decay heat removal while the RCS is depressurized to atmospheric pressure. In this case, the functions provided by the Train B RHR heat exchanger outlet control valve are required to mitigate core damage.
Deterministic Assessment In the "as-found" condition, the Train B RHR heat exchanger outlet control valve had an adjustable wrench lodged in the valve air-operated control mechanism and the feedback linkage was broken. The valve was open and could have performed its design basis safety function. The light on the "SI Not Ready" panel associated with this valve was not illuminated.
Tests on a valve mockup determined that the valve could be fully closed and fully opened in this configuration but could not be modulated.
On the basis of tests performed on a valve mockup, if the wrench became dislodged from the valve operator, the valve would go to the fully closed position without any Operator action. The operators may not be able to re-open the valve when the wrench becomes dislodged and the valve would not be able to perform its safety function.
The wrench appears to have been lodged in the Train B RHR heat exchanger outlet control valve operator for nearly three months. The magnitude of force required to dislodge the wrench from this Train B RHR heat exchanger outlet control valve is not known. Pictures of the wrench in the "as-found" position and mockup tests demonstrated that that the combination of interferences and geometry of the valve positioner mounting bracket served to cradle the wrench.
If the wrench became dislodged during normal plant operations, without provocation, the valve would go to the fully closed position due to dislodgement of the wrench and the Control Room Operators would be alerted to the valve closure by the "SI Not Ready" panel light illumination.
TS 3.5.2, Condition A would have been entered and the valve would be repaired on a timely basis.
For the purposes of discussion, it is postulated that likely causes for dislodging the wrench could include valve operations or an event which would shake the plant such as an earthquake. This valve is in an obscure location; so unobserved bumping by plant personnel resulting in wrench dislodgement is unlikely.
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Forn 366A) (17)
Deterministic Assessment Conclusions The RHR Train B control valve with broken feedback linkage and an adjustable wrench lodged in the valve air operator mechanism is considered an event of low safety significance. This conclusion is based on considerations presented above which include:
- 1. The redundant RHR train, Train A, was operable over 99.99 percent of the time that RHR Train B could be considered inoperable.
- 2.
If the valve closed due to dislodgement of the wrench, the Control Room Operators would have indication from the "SI Not Ready" panel.
- 3.
If wrench dislodgement was associated with an event during which the operators anticipate RHR flow to the reactor vessel, the Control Room Operators could also observe that there is no RHR Train B flow.
- 4.
Postulated events such as an earthquake that could cause significant plant shaking to dislodge the wrench and require RHR flow are very unlikely. An earthquake that could dislodge the wrench could also disable the instrument air system (not designed to withstand seismic events) which would re-open the valve.
Risk Significance Assessment Assumptions
- 1.
It was conservatively assumed for this risk analysis, that the Train B RHR heat exchanger outlet control valve was failed in the closed position.
- 2.
It was assumed that the valve could not be opened from the Control Room.
- 3.
Credit for local Operator action to fail air to the valve to open it was not credited.
- 4.
It was assumed that the Train B RHR heat exchanger outlet control valve was failed in the closed position for 90 days from December 4, 2002 until discovery on March 3, 2003.
- 5.
Large early release frequency (LERF) was not considered due to the insignificant impact of the RHR system on LERF results.
Risk Significance Initiating Events Impact Because of the functions provided by the Train B RHR heat exchanger outlet control valve, it is required for the RHR system to mitigate the consequences of: (1) loss of coolant accidents (LOCAs) (including large break, medium break, small break and reactor coolant pump (RCP) seal LOCAs); and (2) steam generator tube rupture (SGTR) accidents.
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Fomn 366A) (17)
FAILED COMPONENT IDENTIFICATION Train B RHR heat exchanger outlet control valve CV-31236.
PREVIOUS SIMILAR EVENTS
No previous LERs with similar events.