Is this a grade I or II valve and its component?
Generally a grade II is a commercial grade. A grade I is a full nuclear safety grade valve. They have been putting grade II as a result of secret deregulation...
Yep, this is a update from the first one on April 2017. We really don't know all of the recent secret failures or degradation with these components. Just like this guy, nobody outside the plant knew this guy was leaking. You really can't ascertain how unreliable these are. Are the walking dead plants allowed secret deregulations.
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: (04-2017) 0313112020
~ ... ,.,"~ Estimated burden per response to comply with this mandatory collection request: 80 hours. l i\ LICENSEE EVENT REPORT (LER) Reported lessons learned are incorporated into the licensing process and fed back to industry . • ! ~ .f" (See Page 2 for required number of digits/characters for each block) Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. ..... .~ Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects .
(See NUREG-1022, R.3 for instruction and guidance for completing this form
Resource@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 means htti;1://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3!J used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information
1. FACILITY NAME 2. DOCKET NUMBER .PAGE Pilgrim Nuclear Power Station 05000293 1 OF4
4. TITLE Supplement to Potential lnoperability of Safety Relief Valve 3A 5. EVENT DATE 6. LER NUMBER 7. REPORT DATE 8. OTHER FACILITIES INVOLVED
SEQUENTIAL
FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR REV MONTH DAY YEAR N/A NUMBER NO. 05000 N/A FACILITY NAME DOCKET NUMBER 04 24 2017 2017 - 007 01 12 20 2017 N/A 05000 N/A
9. OPERATING MODE 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
N
D 20.2201(b) D 20.2203(a)(3)(i) D 50.73(a)(2)(ii)(A) D 50.73(a)(2)(viii)(A) D 20.2201(d) D 20.2203(a)(3)(ii) D 50.73(a)(2)(ii)(B) D 50.73(a)(2)(viii)(B) D 20.2203(a)(1) D 20.2203(a)(4) D so.73(a)(2)(iii) D 50.73(a)(2)(ix)(A) D 20.2203(a)(2)(i) D 50.36(c)(1)(i)(A) D 50.73(a)(2)(iv)(A) D 50.73(a)(2)(x) 10. POWER LEVEL D 20.2203(a)(2)(ii) D 50.36(c)(1)(ii)(A) D 50.73(a)(2)(v)(A) D 73.71(a)(4) D 20.2203(a)(2)(iii) D so.36(c)(2) D 50.73(a)(2)(v)(B) D 73.71(a)(s) 0 D 20.2203(a)(2)(iv) D so.46(a)(3)(ii) D 50.73(a)(2)(v)(C) D 73.77(a)(1) D 20.2203(a)(2)(v) D 50.73(a)(2)(i)(A) D 50.73(a)(2)(v)(D) D 73.77(a)(2)(i) D 20.2203(a)(2)(vi) D 50.73(a)(2)(i)(B) D 50.73(a)(2)(vii) D 73.77(a)(2)(ii) D 50.73(a)(2)(i)(C) ~ OTHER Specify in Abstract below or in NRG Form 366A
12. LICENSEE CONTACT FOR THIS LER
LICENSEE CONTACT
11
LEPHONE NUMBER {Include Area Code) !Mr. Everett P. Perkins, Jr.- Regulatory Assurance Manager 1508-830-8323 13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT
CAUSE SYSTEM COMPONENT
MANU- REPORTABLE
CAUSE SYSTEM COMPONENT
MANU- REPORTABLE FACTURER TOEPIX FACTURER TOEPIX B SB RV T020 y
14. SUPPLEMENTAL REPORT EXPECTED 15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE) ~ NO SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) On April 24, 2017, during Refueling Outage 21 while performing testing on the Pilgrim Nuclear Power Station (PNPS) Safety/Relief Valves, a high resistance was measured across the solenoid pilot valve coil of SV203-3A. !This solenoid pilot valve was replaced during Refueling Outage 21. After the solenoid pilot valve was removed it was transported to an offsite vendor for additional testing.
rrhis LER supplement is being submitted to provide the NRC with additional information. PNPS continued to follow the testing performed by our offsite vendor and now has additional information to provide the NRC.
PNPS stated at the time that this event was reportable under 10 CFR 50. 73(a)(2)(i)(B), as a condition prohibited by Technical Specifications and also, potentially reportable under 50.73(a)(2)(v)(B) and 50.73(a)(2)(v)(D), a condition that could have prevented fulfillment of a safety function needed to remove residual heat arid mitigate the consequences of an accident. However, additional information provided by our offsite vendor and an engineering evaluation, support the conclusion that there was never a loss of safety ~unction regarding SV203-3A. Therefore, this event was not reportable under 10 CFR 50. 73(a)(2)(i)(B) nor under 10 CFR 50. 73 (a)(2)(v)(B) or (D).
rrhis event posed no threat to public health and safety.
NRC FORM 366 (11-2015)
NRC FORM 366A (04-2017)
u.s_ NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 3/31/2020
LICENSEE EVENT REPORT (LER) CONTINUATION SHEET
(See NUREG-1022, R.3 for instruction and guidance for cqmpleting this form http://www.nrc.gov/readinq-rm/doc-collections/nuregs/staff/sr1022/r3D
Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@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 means used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
1. FACILITYNAME 2 DOCKET NUMBER 3. LERNUMBER
Pilgrim Nuclear Power Station
NARRATIVE BACKGROUND
05000-293
YEAR
2017
SEQUENTIAL NUMBER
- 007
REV NO.
- 01
The 2-stage pilot operated safety relief valve consists of two principle assemblies: a pilot valve section (top works) and the main valve section. The pilot valve section (first stage) is the pressure sensing and control element and the main valve (second stage) provides the pressure relief function. The first stage consists of a pilot-stabilizer disc assembly. The pilot is the pressure sensing member to which the stabilizer disc movement is coupled. Though not mechanically connected, a light spring keeps the stabilizer in contact with the pilot. A pilot preload spring permits set point adjustment of the valve and provides pilot seating force. The solenoid-operated pilot valve controls the pneumatic pressure applied to a diaphragm actuator which controls the relief valve directly. An accumulator is included with the control equipment for each relief valve to store pneumatic energy for relief valve operation. The second or main stage consists essentially of a large piston which includes the main valve disc, the main valve chamber, and a preload spring.
PNPS has four safety relief valves. Each of the four relief valves is equipped with an accumulator and check valve arrangement. These accumulators are provided to assure that the valves can be held open following failure of the nitrogen supply to the accumulators, and are sized to contain sufficient nitrogen for a minimum of 20 valve operations for each safety relief valve. Bottled gas can be used to manually recharge the accumulators associated with two safety relief valves. This capability was installed to address a potential loss of normal nitrogen supply to the accumulators which was identified during seismic reviews.
EVENT DESCRIPTION
On April 24, 2017, while performing testing on the Pilgrim Nuclear Power Station (PNPS) safety relief valves a high resistance was measured across the solenoid valve coil circuit of SV203-3A.
CAUSE OF THE EVENT
The degradation mechanism has been determined to be the solenoid pilot valve coil with high electrical resistance. Per input from our offsite vendor, corrosion of the SV203-3A crimp connections inside the coil created the high resistance indicated by a 9 VDC multimeter.
CORRECTIVE ACTIONS
Removed and replaced solenoid pilot valve assembly for SV203-3A.
NRC FORM 366 (11-2015) Page 2 of 4
NRC FORM 366A (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 3/31/2020
. . "t"'i;, ...... "/
LICENSEE EVENT REPORT (LER) CONTINUATION SHEET
(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nureqs/staff/sr1022/r3l)
Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@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 means used to impose an information collection does not display a currently valid 0MB control number, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
1. FACILITYNAME 2. DOCKET NUMBER 3. LERNUMBER
YEAR
Pilgrim Nuclear Power Station 05000-293
2017
ADDITIONAL INFORMATION FROM THE OFFSITE VENDOR
SEQUENTIAL NUMBER
-007
REV NO.
- 01
After performing multiple tests including destructive examinations of the SV203-3A solenoid valve coil the offsite vendor provided the following conclusions: Degradation of SV203-3A was limited to corrosion at the copper crimp connectors used to attach coil lead wires to the coil winding wire; the construction of the crimp connectors in SV203-3A included an insulating PVC sleeve material; although installation of the coil winding wires in the SV203-3A crimp connectors are inconsistent with industry practices, the coil winding wires were securely joined to the stranded lead wires at the stranded lead wire end of the crimp; the PVC sleeve material on the SV203-3A crimp connectors likely released chlorine, resulting in corrosion of the crimp connectors and wires; corrosion of the SV203-3A crimp connections created .the high resistance indicated by a 9 VDC multimeter prior to application of higher voltage; application of voltage as low as 30 VDC was sufficient to overcome the corrosion product layer allowing the SV203-3A valve to actuate with no nitrogen pressure applied to the inlet port; application of higher voltages up to and including 125 VDC during electrical testing disturbed the corrosion layer sufficiently to allow a 9 VDC powered multimeter to measure SV203-3A coil resistance values representative of actual service conditions with respect to voltage. The valve operated as intended throughout the electrical testing despite the corrosion product accumulation on the crimp connectors; and subsequent to electrical testing, coil resistance values remained within acceptable limits.
SAFETY CONSEQUENCES
There are no consequences to the general safety of the public, nuclear safety, industrial safety and radiological safety from this event. The original concern was that there was a potential inoperability of the Automatic Depressurization System (ADS) which provides a means to rapidly depressurize the primary system to a pressure where low-pressure systems can provide makeup for core cooling in the event of a small or medium break Loss of Coolant Accident. An Engineering evaluation determined that the safety relief valve was fully operable at all times and rem~ined available and capable of performing its intended safety function.
The engineering evaluation that was performed concluded that this event did not constitute a Safety System Functional Failure. (Reference NEI 99-02, Revision 7, Regulatory Assessment Performance Indicator Guideline, Section 2.2, Mitigating Systems Cornerstone, Safety System Functional Failures, Clarifying Notes, Engineering Analyses.) As such, this event will not be reported in the NRC Performance Indicator for Safety System Functional Failures since an engineering evaluation was performed which determined that the system was capable of performing its safety function.
No actions to reduce the frequency or consequence are necessary.
NRC FORM 366 (11-2015) Page 3 of 4
NRC FORM 366A (04-2017)
~'11.IU.!lr.,,,, l i -:.t ; ~ ...... ..,
U.S. NUCLEAR REGULATORY COMMISSION
LICENSEE EVENT REPORT (LER) CONTINUATION SHEET
(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3D
APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 3/31/2020
Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@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 means used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
1. FACILITY NAME 2. DOCKET NUMBER 3. LERNUMBER
YEAR
Pilgrim Nuclear Power Station 05000-293
2017
REPORT ABILITY
SEQUENTIAL NUMBER
- 007
REV NO.
- 01
PNPS believed at the time of the event that it was reportable under 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by Technical Specifications and also, potentially reportable under 50.73(a)(2)(v)(B) and 50.73(a)(2)(v)(D), a condition that could have prevented fulfillment of a safety function needed to remove residual heat and mitigate the consequences of an accident. However, additional information provided by our offsite vendor and an engineering evaluation, support the conclusion that there was never a loss of safety function regarding SV203-3A. Therefore, this event was not reportable under 1 O 50.73(a)(2)(i)(B) nor under 10 CFR 50.73 (a)(2)(v)(B) cir (D).
PREVIOUS EVENTS
LER 2015-002-00, Main Steam Safety Relief Valves Determined to be Inoperable. Following Evaluation
LER 2013-002-00 and -01, SRV-38 Safety Relief Valve Declared Inoperable Due to Leakage and Setpoint Drift
LER 2011-007-00, Safety Relief Valve Declared Inoperable Due to Leakage
REFERENCES
CR-PNP-2017-5067
CR-PNP-2017-5386
CR-PNP-2017-6183
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