United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 82-41: Failure of Safety/Relief Valves to Open at a BWR

                                                       SSINS NO.: 6835 
                                                       IN 82-41 

                               UNITED STATES 
                       NUCLEAR REGULATORY COMMISSION 
                    OFFICE OF INSPECTION AND ENFORCEMENT 
                          WASHINGTON, D. C. 20555 

                              October 22, 1982 

Information Notice No. 82-41:   FAILURE OF SAFETY/RELIEF VALVES TO OPEN 
                                   AT A BWR 

Addressees: 

All nuclear power reactor facilities holding an operating license (OL) or 
construction permit (CP). 

Purpose: 

This information notice is provided as a notification of a potentially 
significant problem pertaining to Target Rock two-stage safety/relief 
valves. It is expected that recipients will review the information for 
applicability to their facilities. No specific action or response is 
required at this time. 

Description of Circumstances: 

On July 3, 1982, Georgia Power Company's Hatch Unit 1 was operating at 100% 
power when a spurious high-pressure signal caused a reactor scram. The 
variation in pressure with time is shown in Figure 1. The main turbine had 
not tripped when a Group 1 isolation* occurred. High-pressure coolant 
injection (HPCI) and reactor core isolation cooling (RCIC) auto-started and 
injected and the recirculation pumps tripped. The main turbine was then 
manually tripped. When vessel water level recovered and reached the high 
water level trip set point, HPCI, RCIC, and the feedwater pump turbines 
tripped. 

Gradual vessel repressurization continued beyond the high-pressure scram 
setpoint on a 0.5 psi/sec ramp without relief valve actuation. About 1180 
psig, three safety/relief valves (SRVs) automatically actuated, relieving 
vessel pressure rapidly. Upon the SRVs' closure, the main steam isolation 
valves were manually reopened and the reactor was cooled and depressurized 
to cold shutdown. During cooling and depressurizing, the remaining eight 
SRVs were manually actuated and functioned properly. 

The SRVs installed on Hatch 1 are the two-stage Target Rock model number 
7567F (see Figure 2). All three SRVs that opened automatically were located 
on the same steam line and were the only valves on that line. Their 
setpoints were 1080, 1080, and 1090 psi. The remaining eight SRVs were set 
at 1080, 

*Closure of main steam isolation valves, main steam drain isolation valves, 
and recirculation loop sample isolation valves. 

8208190239 
.

                                                      IN 82-41  
                                                      October 22, 1982  
                                                      Page 2 of 3 

1090, or 1100 psi. All had been refurbished and steam set at Wyle Labs 
during the previous refueling outage and had most recently been actuated in 
August of 1981. 

Following the July 3, 1982 event, the top works or pilot section (see Figure
3) of all the SRVs were removed and sent to Wyle Labs, where they were 
tested in the as-received condition. Six passed their first test, four 
passed on retest, and the final valve passed on the second retest -- all 
without setpoint spring adjustment. The average first actuation pressure was 
0.9% above nameplate with the highest pressure required being 4.1% above 
nameplate. No abnormal leakage characteristics were observed for any of the 
valves. No apparent mechanical failure was found in the top works at Wyle 
Labs or the valve bodies inspected at Hatch. 

Three additional licensees--TVA, Northeast Nuclear Energy Company, and 
Boston Edison--had reported that two-stage Target Rock valves, tested in the
as-received condition at Wyle Labs, failed to actuate within 1% of the 
setpoint. (Reference LER 50-259/81-25, 50-296/81-74, 50-293/81-62, 
50-260/82-27). (The excessive leakage and the damaged internals of the 
Pilgrim valves may present quite a different problem from that of Hatch, 
Browns Ferry, or Millstone.) The Hatch 1 event of July 3, 1982 was 
potentially the most significant in terms of both (1) the fraction of valves
that failed to open at their setpoint, and (2) the pressure above setpoint 
required to open the valves. 

The General Electric Company (GE) and the Target Rock Company have joined 
Georgia Power in attempting to determine the cause of the failure of the 
valves to actuate. A GE analysis suggests that the most likely cause of the 
high actuation pressure is some combination of friction in the labyrinth 
seal area and/or sticking of the pilot disc in its seat. The slow 
repressurization ramp and the extended period during which the valves were 
not actuated are also considered possible contributors to the incident. 

To define the problem and to improve the probability of actuation of the 
SRVs, Georgia Power has instituted a program at Hatch whereby nine of the 
eleven Unit 1 valves will be exercised regularly. Two valves will not be 
exercised and will be utilized for possible future testing. Unit 2 valves 
will be subjected to a similar program. Also, Georgia Power has arranged 
with GE and with cooperating licensees for screening tests to be done on 
additional SRVs at Wyle Labs. Valves which are pressurized at the 0.5 psi 
ramp to 103% of nameplate rating without actuating are to be candidates for 
diagnostic  testing to determine the magnitude of forces in the disc-to-seat
interface and the labyrinth seal area. Further, examination of interior 
surfaces will be conducted to locate any physical damage. Two such 
candidates were found in the recent testing of three SRVs belonging to 
Northeast Nuclear Energy Company's Millstone Unit 1. 
.

                                                           IN 82-41  
                                                           October 22, 1982 
                                                           Page 3 of 3 

If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate NRC Regional Office, or this office. 



                                   Edward L. Jordan, Director 
                                   Division of Engineering and  
                                     Quality Assurance  
                                   Office of Inspection and Enforcement 

Technical Contact:  M. S. Wegner, IE 
                    301-492-4511 

Attachments: 
1. Appendix A: Figures 1 through 3 
2. List of Recently Issued IE Information Notices  

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