05000263/LER-2007-001, Regarding Reactor Scram Due to Turbine Control Valve Housing Support Failure
| ML070790244 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/12/2007 |
| From: | Conway J Nuclear Management Co |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| L-MT-07-012 LER 07-001-00 | |
| Download: ML070790244 (5) | |
| 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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iv), System Actuation |
| 2632007001R00 - NRC Website | |
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'c ý NMC Committed to Nuclear Excellence Monticello Nuclear Generating Plant Operated by Nuclear Management Company, LLC March 12, 2007 L-MT-07-012 10 CFR Part 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket No. 50-263 License No. DPR-22 LER 2007-001, "Reactor Scram due to Turbine Control Valve Housing Support Failure" A Licensee Event Report for this occurrence is attached.
This letter contains no new commitments and no revisions to existing commitments.
John T. Conway Site Vice President, Monticello Nuclear Generating Plant Nuclear Management Company, LLC Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC 2807 West County Road 75 9 Monticello, Minnesota 55362-9637 Telephone: 763-295-5151 9 Fax: 763-295-1454
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2007 (6-2004)
COMMISSION
, the NRC may not conduct or sponsor, and a digits/characters for each block) person is not required to respond to, the information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Monticello Nuclear Generating Plant 05000263 1 of 4 TITLE (4) Reactor Scram due to Turbine Control Valve Housing Support Failure EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
MO DAY YEAR YEAR SEQUENTIAL REV MO DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO 05000 FACILITY NAME DOCKET NUMBER 01 10 2007 2007 001 00 03 12 2007 05000 OPERATING 1
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check 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 088 20.2201 (d.
20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LEVEL (10) 20.2203(a)(1) 50.36(c)(1)(i)(A)
X 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 Specify in Abstract below or 20.2203(a)(2)(iii) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C) in NRC Form 366A 20.2203(a)(2)(iv) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D) 20.2203(a)(2)(v) 50.73(a)(2)(i)(B) 50.73(a)(2)(vii) 20.2203(a)(2)(vi) 50.73(a)(2)(i)(C) 50.73(a)(2)(viii)(A) 20.2203(a)(3)(i) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(B)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Area Code)
Ron Baumer
[ 763-295-1357 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
MANU-REPORTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT F
U TECAUSE SYSTEM COMPONENT FACTURIER TO EPIX I FACTURER TO EPIX SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR INO SUBMISSION YES (if yes, complete EXPECTED SUBMISSION DATE).
IX INO DATE (15)
ABSTRACT On January 10, 2007, the turbine control valves [JJ] ramped from approximately 50 percent open to 100 percent open in two seconds without a demand from the pressure control system.
A Group 1 isolation and automatic scram ensued. Review of plant conditions immediately after the scram disclosed that the main turbine [TRB] control valves [SCV] did not close following the scram. During a post scram plant walk down it was identified that the supports for the turbine control valve enclosure [SPT] had failed, allowing the enclosure to transition downward approximately six inches.
The root cause of this event was determined to be latent shortcomings in the design and construction of the turbine control valve enclosure support system during initial construction.
Corrective actions include modifying and strengthening the enclosure support system, performing inspections of associated piping and systems, and performing an extent of condition review.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6 PAGE (3)
SEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBE NUMBER 2 of 4 2007 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Description
At approximately 1528 on January 10, 2007, with the unit stable in Mode 1 at approximately 88 percent power, all four turbine control valves [SCV] ramped from approximately 50 percent open to 100 percent open in approximately two seconds with no demand to do so from the pressure control system. The resultant decrease in reactor pressure resulted in a group 1 isolation on low steam line pressure with the MODE switch in RUN, and associated reactor scram. The closure of the Main Steam [SB] Isolation Valves and reactor scram caused reactor pressure to increase and reactor water level to decrease to minus eight inches. The low reactor water level also resulted in a Group II Emergency Safety Feature actuation as designed. All control rods [AA] inserted and all safety systems responded as expected.
During the reactor water level transient, the turbine [JJ] and main feed [SJ] pumps [P] tripped at 48 inches as designed. Maximum water level reached was 78 inches (30 inches below the main steam lines). Control of plant pressure was initially established by manually operating the safety relief valves (SRVs) [RV]. At no point was automatic operation of the SRVs required or challenged. Subsequently, reactor pressure control was established using the high pressure coolant injection system (HPCI) [BJ]. Reactor makeup was established using normal feedwater.
Post scram review disclosed that the turbine control valves did not close on the turbine trip as designed. A post scram walk down revealed that supports [SPT] for the control valve enclosure had failed, allowing the enclosure to transition downward approximately six inches, causing deformation of adjoining support structures. No system breach occurred and there was no release of radioactivity.
Following plant stabilization, the decision was made to proceed to cold shutdown (Mode 4).
As uncertainty existed regarding the status of the main steam system (the control valve enclosure contacted insulation on a section of the main steam line and the control valves failed to close following the turbine trip), the decision was made to cool down using HPCI and the SRVs for pressure control, main feed for makeup, and the residual heat removal system [BO]. Mode 4 was achieved at 0613 on January 11, 2007.
Event Analysis
Pursuant to 1 OCFR 50.72 paragraphs (b)(2)(iv)(B) for the RPS actuation and paragraph (b)(3)(iv)(A) as an ESF actuation, an eight-hour event notification was made to the USNRC.
Per 10 CFR 50.73 (a)(2)(iv), a Licensee Event report is required for this event.
The event is not classified as a safety system functional failure.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 3 of 4 2007 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Safety Significance
At approximately 88 percent reactor power, all turbine control valves ramped full open. As reactor pressure dropped, reactor power lowered. The Group 1 isolation occurred at 840 psi as required. In response to the automatic scram at a MSIV position of 90 percent open with the MODE switch in RUN, all control rods inserted as designed. The minimum reactor water level reached during the transient was minus eight inches. No Emergency Core Cooling System (ECCS) actuation was required or occurred. As level lowered below nine inches, the Group 2 isolation occurred as expected. With the feed system in service, feedwater, in conjunction with decay heat, caused reactor water level to rise. At 48 inches the main turbine and the main feedwater pumps tripped as expected. The highest water level reached was 78 inches, approximately 30 inches below the main steam lines. Reactor pressure was controlled initially by manually operating the safety relief valves. No automatic operation of the safety relief valves was required. Water level was subsequently restored to normal levels, feedwater was restarted for makeup, and high pressure coolant injection started and operated in the pressure control mode to remove decay heat.
With the exception of the failure of the turbine control valves to close, all systems and components responded as expected to mitigate the transient and assure core cooling.
Automatic operation of the SRVs and emergency core cooling systems did not occur and were not required.
No safety systems or systems credited in the Probabilistic Risk Assessment (PRA) as capable of mitigating an accident were compromised as a result of the structural failure. As a result, there was no significant increase in the risk of a core damage accident or a release of radioactive material to the environment. The transient was well within the analyzed and anticipated frequency of a turbine trip initiator analyzed in the PRA model and bounded by the Pressure Regulator failure analysis in the USAR.
In conclusion, the safety significance in terms of reactor safety and radiological release to the environment as a result of the transient that occurred on January 10, 2007, was not significant.
Cause
The root cause of this event was determined to be latent shortcomings in the design and construction of the turbine control valve enclosure support system during initial construction.
The failure of the support system and associated vertical transition of the control valve enclosure caused the cam that actuates the servo for the control valves to rotate in a direction commanding the control valves to open. This motion overrode signals from the pressure regulating system to close the control valves. Because the cam motion was forced by theU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 4 of 4 2007 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) transition downward of the control valve enclosure, the control valves remained open despite the turbine trip signal.
Corrective Action
The following actions have been completed:
- 1. The control valve enclosure support system was redesigned assuring that stress levels remained well below code allowable values. The modified design was installed prior to startup.
- 2. Collateral damage caused by the failure of the control valve supports and the vertical transition of the enclosure was analyzed and repairs completed as required.
- 3. Inspections and tests of the turbine control system were performed and repairs affected as necessary to assure full system functionality prior to startup.
- 4. An extent of condition review and plant walkdown was completed which concluded that other similar support configurations were sufficiently robust such that reasonable assurance of their functionality existed.
Failed Component Identification None
Previous Similar Events
No previous similar events were identified.