Junk Plant Hatch Safety Relief Valves: Why Did They Cancel Extending The Testing Frequency?

Update Oct 16: (made wording better) 

???

Thinking of a 2.206. 

Technical Letter Report
 

Three-Stage Target Rock Safety Relief Valve Performance and Reliability in Long-Term Station Blackout Accident Scenario for State of the Art Reactor Consequence (SOARCA) Program 

As mentioned above, in a LTSBO for a BWR 4 with Mark I containment, SRVs are assumed operable at the start of the event to provide RCS over-pressure control.  Analyses have shown [1] that, in the first ten hours of this transient, the SRVs will cycle open and closed approximately 440 times.  During each cycle, the valves open approximately every 45 seconds, and remain open for approximately four to six seconds.  After ten hours of the LTSBO, thermal hydraulic calculations show that the heat removal capacity of the wetwell is effectively exhausted (since the residual heat removal system is unavailable), and RCS temperature will rise.  RCS pressure will also continue to increase, resulting in additional open demands on the SRVs with increasing steam temperature.

An alternative basis for the SRV stochastic failure probability is proposed.  An industry representative has stated that the valve manufacturer warrants the valves for hundreds o

cycles at operating temperature and pressure conditions.  Note that these conditions are to be expected throughout the initial time period of a LTSBO, up to approximately 10 hours during which the valve will be demanded to cycle approximately 400 times.  The valve manufacturer has stated that laboratory tests have been performed to substantiate the conditions included in the valve warranty, i.e. that the valve can reliably open and close for hundreds of cycles.  The industry representative also stated that fossil-fuel plant experience with the same types of valves can be used to validate these assertions.

*** It is ridiculous Hatch would even think of extending their SRV testing interval on their very troublesome and unreliable SRVs going back a decade or more. They switched back and forth between stage 2 and 3 so many times I can keep up. They are a lemon designed valve and they should be replaced.

What a "crock of shit" with the drywell ventilation air blowing on the SRVs. This is probrably the third roll out of the 3 stage SRVs. The first two they had withdraw the valves from the plant due to being defective. They replaced the 2 stage with the 3 stage because of set point drift and leakage. No LERs in the past indicating ventilation air damaging on the valves in both plants. These excuses are highly suspect. At best, with this kind of unexpected failure, you can make the case the 3 stage is a too delicate a valve for the normal and emergency operation of the plant.

On the positive side, setpoint drift failures are a lot less in both plants since they installed all the 3 stage valves. The rub is, these valves haven't been in the plant for very long time. Very little aging in these young valves. By the way, the 3 stage valves aren't new valves. They are refurbished and have so called updated components in the valves. I suspect when these valves have been in the plant for two or more cycle they will fail at much higher rates.  

This LER below is probably the reason for canceling the testing extension.

The new SRV LER on unit 1 troubles was caused by ventilation air on the SRV causing abutment issues.  

Safety Relief Valves' As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria

On June 20, 2018, Unit 1 was at 100 percent rated thermal power (RTP) when "as-found" testing results of the 3-stage main steam safety relief valves (SRVs) indicated two of the eleven Unit 1 SRVs had experienced a setpoint drift during the previous operating cycle which resulted in their failure to meet the Technical Specification (TS) opening setpoint pressure of 1150 +/34.5 psig as required by TS Surveillance Requirement (SR) 3.4.3.1. The test results showed that the two SRVs were slightly out of spec tow due to setpoint drift.
The SRV pilots were disassembled and inspected to investigate the reason for the setpoint drift. Based on inspection results, the drift in setpoint was due to low abutment gap and low abutment pressure. Due to their location, drywall ventilation blowing on these two safety relief valves caused them to undergo a cyclic heating and cooling every 12 hours during the Unit's 2-year operating cycle. These temperature gradients across the valve internals caused a relaxation of the setpoint spring and bellows assembly.
NRC...

The below LER is in unit 2. It came out last year...it is their newest LER on SRV issues. It is way suspicious abutment issues are in both plants. Abutment issues are happening all over the industry beginning at Pilgrim and so called test stand damage. So loose manufacturing tolerances are behind this plant's the SRV abutment problem, but different than unit 1. So the strategy is to come up with a new failure mechanism every time they have tech spec SRV drift point test failing. All the NRC cares about is if they find the cause of the failure and it isn't repeated in the future. That is way they come up with a new angle every time they fail a tech spec lift test in the abutment issues. The go shopping for a new failure mechanism every time they submit a LER. I doubt the NRC inspectors have the time to verify deeply the failure. I am sure management just tell the local inspectors to just trust the licensee. I am sure the licensee knows how to game the system of trust thus the shifting failure mechanism to minimize a deeper investigation.     

I am appalled Hatch and the NRC tolerate the "most likely" acceptable standard for failure mechanisms for Unit 2. The cause of the defect and its fix is suppose to be bullet proof. It should have gone though a comprehensive engineering evaluation on the cause and then put them on a testing stand to beat the crap out of the valves in its supposed environment. If new issues are discovered on the testing stand, then a redesign of the defective and poorly engineered component happens. Then back to the stand. And do this over and over again until the design is perfect. 

(Updated) Edwin I. Hatch Nuclear Plant license Event Report 2017-(){)4-()() Safety Relief Valves' As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria

On June 30 2017, with Unit 2 at 100 percent rated thermal power {RTP), "as-found" testing of the 3-stage main steam safety relief valves (SRVs) (EBS Code RV) showed that two of the eleven main steam SRVs that were tested had experienced a drift in pressure lilt setpoint during the previous operating cycle such that the allowable technical specification (TS) surveillance requirement (SR) 3.4.3.1 limit of 1150 +/-34.5 psig had been exceeded. Below is a table illustrating the Unit 2 SRVs that failed as found testing results alter being removed tom service during the Spring 2017 refueling outage. 

MPL 2821-F013C 2821-F013E 
Event cause Analysis 
Drift -39 psig -49 psig 

The SRV pilots were disassembled and inspected while investigating the reason for the drift It was found that the abutment gap closed prematurely during testing using a linear variable differential transformer to measure pilot stroke distance. The pre-mature abutment gap closure is most likely due to loose manufacturing tolerances leading to SRV setpoint drift. 

Something to do with the BWR SRV owners group formed on investigating the poor quality of the SRVS?  

October 15, 2018

SUBJECT: EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 PROPOSED ALTERNATIVE RR-V-12 REGARDING MAIN STEAM SAFETY RELIEF VALVE TESTING (EPID L-2018-LLR-0054)

Dear Ms. Gayheart 

By letter dated April 9, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 18099A146), the Southern Nuclear Operating Company, Inc. (SNC or licensee) submitted an alternative request for the Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2. The proposed alternative RR-V-12 would extend the frequency for testing all main steam safety relief valves for each unit from once every 5 years to once every 3 refueling cycles (i.e., 6 years), as allowed by American Society of Mechanical Engineers (ASME) Code Case OMN-17. Subsequently, by letter dated July 18, 2018 (ADAMS Accession No. ML 18199A588), SNC withdrew the alternative request. The U.S. Nuclear Regulatory Commission (NRC) staff had recently approved the code case for unconditional use in Regulatory Guide 1.192. 

The purpose of this letter is to advise you that the NRC has received your request to withdraw the application dated April 9, 2018, and upon receiving the withdrawal letter, the NRC staff ceased its review of the above-cited application. 

If you have any questions regarding this matter, I may be reached at 301-415-4032 or via email at Randy.Hall@nrc.gov.
Docket Nos.: 50-321 and...