05000440/LER-2013-003, Shutdown Required by Technical Specifications Due to RCS Pressure Boundary Leakage
| ML13231A184 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/14/2013 |
| From: | Kaminskas V FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-13-238 LER 13-003-00 | |
| Download: ML13231A184 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 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)(i)(B), Prohibited by Technical Specifications |
| 4402013003R00 - NRC Website | |
text
FENOCQ FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant PO. Box 97 10 Center Road Perry Ohio 44081 Vito A. Karninskas Vice President 440-280-5382 Fax: 440-280-8029 August 14, 2013 L-1 3-238 10 CFR 50.73(a)(2)(i)(A) 10 CFR 50.73(a)(2(ii)A) 10 CFR 50.73(a)(2)(iv)(A)
ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Licensee Event Report Submittal Enclosed is Licensee Event Report (LER) 2013-003, "Shutdown Required by Technical Specification Due to RCS Pressure Boundary Leakage." There are no regulatory commitments contained in this submittal.
If there are any questions or if additional information is required, please contact Mr. Thomas Veitch, Manager-Regulatory Compliance, at (440) 280-5188.
Sincerely,
Enclosure:
LER 2013-003 cc:
NRC Project Manager NRC Resident Inspector NRC Region III I
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 10/31/2013 (10-2010)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Perry Nuclear Power Plant, Unit 1 05000-440 1 OF 4
- 4. TITLE Shutdown Required by Technical Specifications due to RCS Pressure Boundary Leakage
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MOT DY Y
EA EQETIAL[REV MOT DA FACILITY NAME IDOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 06 15 2013 2013
- - 003
- - 00 08
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
[- 20.2201(b)
LI 20.2203(a)(3)(i)
[E 50.73(a)(2)(i)(C)
FI 50.73(a)(2)(vii)
El 20.2201(d)
El 20.2203(a)(3)(ii)
Z 50.73(a)(2)(ii)(A)
EL 50.73(a)(2)(viii)(A)
E] 20.2203(a)(1)
L] 20.2203(a)(4) 17 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
E] 20.2203(a)(2)(i)
EL 50.36(c)(1)(i)(A)
EL 50.73(a)(2)(iii) 0l 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E] 20.2203(a)(2)(ii)
[E 50.36(c)(1)(ii)(A)
Z 50.73(a)(2)(iv)(A)
E] 50.73(a)(2)(x)
[E] 20.2203(a)(2)(iii)
[E 50.36(c)(2) 17 50.73(a)(2)(v)(A)
El 73.71(a)(4) 008 E] 20.2203(a)(2)(iv)
[: 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71 (a)(5)
E] 20.2203(a)(2)(v)
Z 50.73(a)(2)(i)(A)
E] 50.73(a)(2)(v)(C)
El OTHER Specify in Abstract below El 20.2203(a)(2)(vi)
[E 50.73(a)(2)(i)(B) lj 50.73(a)(2)(v)(D) orin At 0313 hours0.00362 days <br />0.0869 hours <br />5.175265e-4 weeks <br />1.190965e-4 months <br />, the plant entered Mode 2 when the operators placed the mode switch in the Startup/Hot Standby position.
The type of plant shutdown conducted was a rapid soft shutdown where the operators insert control rods manually. At 0340 hours0.00394 days <br />0.0944 hours <br />5.621693e-4 weeks <br />1.2937e-4 months <br />, the Unit Supervisor declared the reactor subcritical based on in-core nuclear instrumentation readings. The operators continued to insert control rods in accordance with the established shutdown sequence. At 0353 hours0.00409 days <br />0.0981 hours <br />5.83664e-4 weeks <br />1.343165e-4 months <br />, the rod control and information system (RC&IS) [AA] malfunctioned preventing normal control rod movement. The operators entered the off-normal instruction for inability to move control rods. At 0403 hours0.00466 days <br />0.112 hours <br />6.66336e-4 weeks <br />1.533415e-4 months <br />, a manual reactor scram was inserted to complete the shutdown in accordance with normal operating procedures and the evolution specific reactivity plan termination criteria. All withdrawn control rods at the time fully inserted and there were no complications experienced in the scram. Mode 3 was entered at 0403 hours0.00466 days <br />0.112 hours <br />6.66336e-4 weeks <br />1.533415e-4 months <br />. The scram was reset at 0410 hours0.00475 days <br />0.114 hours <br />6.779101e-4 weeks <br />1.56005e-4 months <br />.
Mode 4 was entered at 1358 hours0.0157 days <br />0.377 hours <br />0.00225 weeks <br />5.16719e-4 months <br />.
CAUSE OF EVENT
The root cause analysis of the RCS pressure boundary leakage remains under management review at the time of this report. A supplemental licensee event report will be issued to further address the causes and corrective actions once they are finalized.
Preliminarily, the results indicate that the weld crack and through wall leak at the inlet socket weld to 1 B33F0647B were caused by vibration induced high cycle fatigue socket weld failure.
Engineering evaluations of the Reactor Recirculation (RR) system design configuration performed in response to a similar socket weld failure in 1991 were inconclusive with respect to risk and probability of repeat failure. As a result, no modifications to RR system piping/valve configuration were made to ensure a similar socket weld appendage failure due to high cycle fatigue would not occur.
EVENT ANALYSIS
The initial reactor downpower to eight percent RTP and the rapid soft shutdown to zero percent RTP were performed in accordance with plant operating procedures and the reactivity plan. The RC&IS malfunction and the manual RPS actuation occurred after the reactor was shutdown. No plant parameters experienced in the shutdown process challenged the transients described in the Updated Safety Analysis Report Chapter 15, Accident Analysis.
The RR system provides a forced coolant flow through the core to remove heat from the fuel to allow operation at significantly higher power levels than would otherwise be possible. The system consists of two recirculation flow loops each consisting of a motor driven pump and a flow control valve. The RCS pressure boundary leak, as described, would not have prevented the RR system flow control valve B from performing its design function.
A qualitative probabilistic risk assessment (PRA) was performed for this event. While RCS pressure boundary leakage was present, this leakage did not require an immediate scram. The PRA model does not consider a controlled plant shutdown as an initiating event. Furthermore, the given condition did not make any PRA modeled equipment/functions unavailable. Therefore, the PRA assessment concludes there are no changes in core damage frequency (CDF) or large early release frequency (LERF). The delta CDF and delta LERF values remain well below the acceptable thresholds of 1.0E-06/yr and 1.OE-07/yr respectively as discussed in Regulatory Guide
1.174. Plant configurations with changes in CDF of less than 1.OE-06 and LERF of less than 1.OE-07 are not considered to be significant risk events. Based on the PRA results, the safety significance of this event is considered to be small.
CORRECTIVE ACTIONS
Visual (VT-2) and liquid penetrant (PT) inspections were performed on the other vent and drain appendages for the RRS flow control valves. Examinations of all welds on these valves were satisfactory with no indications of surface cracks.
A new vent valve assembly was fabricated and installed on 1 B33F0060B. The pipe section between 1B33F0060B and 1B33F0647B was replaced with schedule 160 pipe and the pipe weld was built up to the design "hour glass" dimensions.
The removed vent valve assembly containing the cracked weld has been sent off-site to a qualified testing facility for metallurgical failure analysis. The results are pending at the time of this report.
The design configuration options to eliminate cyclic fatigue failure of appendages from the RR system piping will be evaluated. The options include use of additional supports and other changes to modify the line stiffness to prevent the effects of vibration.
The flange connection for CRDM 30-15 is a mechanical joint and was reworked to eliminate any reactor water leakage. Corrective actions to repair the RCS pressure boundary leakage and the CRDM flange connection restored Drywell unidentified leakage back to less than the previous operating cycle leakage.
PREVIOUS SIMILAR EVENTS
A review of LERs and the corrective action database for the past three years did not identify any previous similar events or condition reports associated with RCS pressure boundary leakage. Two LERs were written for completion of shutdown required by TS. These include LER 2011-002-01, Condition Prohibited by Technical Specifications and Plant Shutdown due to Unit 1 Startup Transformer Issues, and LER 2010-003, Loss of Control Rod Drive Header Pressure Results in Manual RPS Actuation. None of the corrective actions for these LERs would have been reasonably expected to prevent the event documented in LER 2013-003.
COMMITMENTS
There are no regulatory commitments for these LERs contained in this report. Actions described in this document represent intended or planned actions, are described for the NRC's information, and are not regulatory commitments.