05000220/LER-2004-001, Re Manual Reactor Scram and Cooldown Rate Exceeding Technical Specification Limits Due to Electrometric Relief Valve Failure to Close
| ML041950181 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 07/01/2004 |
| From: | Hopkins L Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP1L 1848 LER 04-001-00 | |
| Download: ML041950181 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(iv)(A), System Actuation 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)(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)(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) |
| 2202004001R00 - NRC Website | |
text
Constellation Energy-PO. Box 63 Nine Mile Point Nuclear Station Lycoming, New York 13093 July 1, 2004 NMPIL 1848 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
Nine Mile Point Unit 1 Docket No. 50-220; DPR-63 Licensee Event Report 04-001, "Manual Reactor Scram and Cooldown Rate Exceeding Technical Specification Limits Due to Electromatic'Relief Valve Failure to Close" Gentlemen:
In accordance with 10 CFR 50.73(a)(2)(i)(A), 10 CFR 50.73(a)(2)(i)(B), and 10 CFR 50.73(a)(2)(iv)(A), we are submitting Licensee Event Report 04-001, "Manual Reactor Scram and Cooldown Rate Exceeding Technical Specification Limits Due to Electromatic Relief Valve Failure to Close."
Very truly yours, Lawrence A. opkins Plant General Manager LAHIJRH/jm Attachment cc:
Mr. H. J. Miller, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector
'l.
I NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004 (1.200 Estimated burden per response to comply with this mandatory Information collection request: SC hours. Reported lessons learned are Incorporated Into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records Management Branch (T-C E6). U.S. Nuclear Regulatory Commission. Washington, DC 20555-0001, or by Internet e-mail LICENSEE EVENT REPORT (LER) to ba s 1 nrc. gov. and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0104). Office of Management and Budget, Washington. DC 20503. If a (See reverse for required number of means used to Impose Information collection does not display a currently valid OMB control digits/characters foreach block) number, the NRC may not conduct or sponsor, and a person Is not required to respond to, the
, sinformation collection.
FACILITY NAME (1)
DOCKET NUMBER (2) l PAGE (3)
Nine Mile Point, Unit 1 05000220 1
OF 5
TITLE (4)
Manual Reactor Scram and Cooldown Rate Exceeding Technical Specification Limits Due to Electromatic Relief Valve Failure to Close EVENT DATE (S)
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 f
05000 05 02 2004 2004 -
001 -
00 07 01 2004 FACILITY NAME DOCKET NUMBER 05000 OPERATING 1
THISREPORT IS SUBMITTED PURSUANT TO HE REQUIREMENTS OF 10CFR 3:(Checkall that apply) (11)._
MODE (9)
POWER 20.2201(b) l 20.2203(a)(3)(ii) ll50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)
LEVEL (10) 1 20.2201(d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x) 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 20.2203(a)(2)(iii) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C) 20.2203(a)(2)(iv)
X 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D) 20.2203(a)(2)(v)
X 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)
I_-
20.2203(a)(3)(i) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(B)
_i LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Area Code)
M. Steven Leonard, General Supervisor Licensing 315-349-4039 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE r
CAUSE
SYSTEM COMPONENT I MANU-I REPORTABLE I
F FACTURER I
TO EPIX E
I FACTURER S TO EPIX A
SB PSV Dresser Valve Y
lI l
I SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED SUBMISSION DATE (15) l DAY YEAR F
VYES (If yes, complete EXPECTED SUBMISSION DATE).
XUNO I
- - I I
'-I I-ABSTRACT (Limit to 1400 spaces, I.e., approximately 15 single-spaced typewritten lines) (16)
On May 2, 2004, at approximately 0209 hours0.00242 days <br />0.0581 hours <br />3.455688e-4 weeks <br />7.95245e-5 months <br />, while conducting post-maintenance testing following plant startup, Electromatic Relief Valve (ERV) 123 would not reclose following remote manual actuation. Efforts to close the ERV failed and at 0217 hours0.00251 days <br />0.0603 hours <br />3.587963e-4 weeks <br />8.25685e-5 months <br /> the plant was manually scrammed. High Pressure Coolant Injection initiated on low Reactor Water Level due to the level transient associated with the scram as expected. Reactor cooldown rate exceeded Technical Specification limits for the first hour following the scram. An engineering evaluation concluded that 10 CFR 50 Appendix G requirements were not violated and that structural integrity of the Reactor Pressure Vessel (RPV) was not compromised.
The root cause for the ERV failure to close was inadequate use of procedures." The failure resulted from improper assembly of the pilot valve following maintenance. An extra gasket erroneously installed in the pilot valve assembly allowed bypass leakage around the pilot valve causing the main valve to remain open after the pilot valve was closed.
The excessive cooldown rate resulted from the stuck open ERV, which limited the ability to control the rate of RPV depressurization and cooldown.
The ERV was disassembled and reassembled correctly. Subsequent testing demonstrated satisfactory valve performance.
NRC FORM 366 (1.2001)
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1,.. -
NRC FORM 361A U.S. NUCLEA REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
NUMBER (2)
Nine Mile Point, Unit 1 05000220 YEAR SEQUENTIAL REVISIO 2
OF 5
_ NUMBER NUMBER 2004 001 00
NARRATIVE
(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 Form 366A) (17)
III. Analysis of Event
Upon determination that ERV 123 could not be reclosed, the reactor was shutdown by means of a manual scram to minimize uncontrolled heat addition to the suppression pool. Following the scram, plant equipment performed as designed with the exception of the stuck open ERV. Inability to close the ERV limited ability to control plant depressurization and cooldown rate. Actions were taken to minimize the cooldown rate, including promptly closing MSIVs. Despite these actions, TS cooldown rate limitations were exceeded during the first hour following the scram.
T.Jhe cooldown injthe first hour-was 221.9.degrees F.The subsequent cooldown in the following 1 and 1/2 hour period was approximately 60 degrees F. Temperature was then stabilized and held constant for approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> prior to completing the cooldown to cold shutdown conditions. This event is considered a blowdown event which was bounded by the design basis 300 degree emergency cooldown analyses. The design basis allows for 10 emergency cooldown events at the maximum cooldown rate of 300 degrees per hour.
Operator action to insert a manual scram in a timely fashion limited suppression pool maximum temperature during the event to 104 degrees F, less than the TS maximum of 110 degrees F, assuring operability of the pressure suppression function of primary containment throughout the event.
Engineering evaluations conducted prior to unit restart were based upon conservative cooldown rate assumptions.
These evaluations determined that the stuck open ERV event did not create conditions outside the design basis of the plant and that no supplemental inspections or reviews were required prior to plant startup.
Evaluations included the following:
An ASME Xl, Appendix E evaluation was conducted which determined that Appendix E criteria were met and that structural adequacy was assured for all vessel regions throughout the event. In addition, the actual event cooldown was compared to a core-not-critical Pressure and Temperature Limit curve (P/T curve) derived based upon an assumed 225 degrees F in one hour temperature change and based upon existing NMP1 PIT curve methods. The vessel stresses due to the event were shown to remain acceptable from a fracture mechanics standpoint.
The transient thermal stresses created from this event were reviewed and determined to be bounded by the RPV stress, fatigue and fracture analysis 300 degree per hour emergency cooldown normal/upset stress and fatigue analyses. As such, no supplemental inspection-orreview.requirements apply.
An evaluation concluded that no thermal stratification between the lower plenum and the beltline occurred during the event.
The core shroud repair tie rod assemblies design basis analyses were reviewed and determined to be acceptable for this event.
A thermal-mechanical assessment of the fuel concluded that the impact of the event on the fuel was negligible. In addition, a review of fuel cladding TS Limiting Safety System Settings and supporting Bases indicated that there was no impact on fuel reliability due to the rapid cooldown.
A review determined that the event is bounded by the Upper Shelf Energy equivalent margin analysis.
A review of various recirculation piping flaw and fatigue evaluations was performed. The review concluded that the transient thermal stresses created from this event had no impact on the conclusions of the previous calculations.
Based upon satisfactory performance of plant systems and based upon engineering evaluations of the effects of the event, the event did not pose a threat to the health and safety of plant personnel or the public.
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I NROFCRM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
.NUMBER (2)
Nine Mile Point, Unit 1 05000220 YEAR SEQUENTIAL I REVISION 5
OF 5
NUMBER NUMBER 2004 001 00 I-NAiHA IWt (if more space Is required, use additional copies or -C Fi-on ar6mA) (1 7)
IV. Corrective Actions
ERV 123 was disassembled and repaired. Post maintenance testing and subsequent performance were satisfactory.
i
V. Additional Information
- - A. -Failed.Components:---
~~
Electromatic Relief Valve 123 Dresser Industries Model 6-1525VX-3-NCO60
B. Previous similar events
LER 2000-004 reported lifting and failure to close of ERV 111. ERV failure was caused by a bent pilot valve stem.
The event occurred during plant startup with reactor power at less than one percent thermal power and with reactor pressure at approximately 30 psig. Although this event also involved an ERV failure to close, the cause and circumstances differ. Corrective actions from LER 2000-004 would not have prevented the current event.
C. Identification of components referred to in this Licensee Event Report:
Components IEEE 805 System ID IEEE 803A Function Electromatic Relief Valve SB PSV High Pressure Coolant Injection BJ N/A Reactor Pressure Vessel SB RPV Suppression Pool NH N/A Main Steam Isolation Valves SB ISV Shutdown Cooling System BO N/A Reactor Protection System JC N/A
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