ML20010A335
| ML20010A335 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/31/1980 |
| From: | Jeffries J, Schmidt W ELECTRIC POWER RESEARCH INSTITUTE, MPR ASSOCIATES, INC. |
| To: | |
| Shared Package | |
| ML20010A327 | List: |
| References | |
| EPRI-NP-80-13-LD, EPRI-NP-80-130L, NUDOCS 8108110325 | |
| Download: ML20010A335 (102) | |
Text
{{#Wiki_filter:. Examination and Test of Crystal gl River Unit No. 3 Power-Operated Relief and Safety Valves '"",=S?'ni P PWR Safety and Relief Valve Test Program $*e$e"r*08$ Limited Distribution Keywords: Copy Number Relief Valves Over Pressure Protection System Safety Valves PWR Transient Prepared by MPR Associates, Inc. Washington, D.C. 1 i l ELECTRIC POWER RESEARCH INSTITUTE hbR O O O2 P PDR
Examination and Test of Crystal River Unit No. 3 Power Operated Relief and Safety Valves PWR Safety and Relief Valve Test Program t N P-80-13-LD Research Project V102 31 Interim Report, December 1980 Prepared by l MPR ASSOCIATES, INC. 1140 Connecticut Avenuc, N.W. Washington, D.C. 2066 Principal Investigator W. R. Schmidt l l Preoared for Participating PWR Utikties and Electric Power Research Institute 3412 Hillview Avenue Palo Alto, California 94304 EPRI Project Manager J. D. E. Jeffries Nuclear Power Division
1 f EPRI PERSPECTIVE
- PROJECT DESCRIPTION The project described by this report is part of the overall Pressurized hater Reactor (PWR) Safety and Relief Valve Test Program and focuses on the examination and testing of valves that actuated under transient conditions in a nuclear unit.
On February 26, 1980, Crystal River Unit No. 3 experienced an i electrical system malfunction that led to power operated relief valve (PORV) actuation, a reactor trip, initiation of high-pressure safety injection, and eventually activation of a pressurizer safety valve. Analysis of these transient and resultant conditions shows that the primary system was indeed
" solid," i.e., no steam bubble in the pressurizer, and that the safety valve discharged water for approximately two hours.
Actuation of the pressurizer safety valves under an-; condition is rare; and, according to a rather extensive literature search, this, event was the first where liquid flowed through a safety valve. This project examined the internal portions of 3 the affected valves and " pop" tested the safety valves to determine if the setpoints in the plant were reproducible. iii , 1
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PROJECT OBJECTIVES The main objective, of course, was to determine if there were any deleterious effects to the safety valve due to the dis-charge of liquid effluent. The safety valve that opened during the transient actuated approximately 100 psi below its normal setpoint; we wanted to determine if this point of actuation was reproducible under controlled conditions. During the transient, primary system pressure never reached 2475 psig, the lower limit for safety valve setpoints; it was an objective of this program to find out where the second safety valve would have actuated had primary system pressure continued to rise. Another objective was to determine the condition of the PORV via detailed examination and an assess-ment of the discharge piping. PROJECT RESULTS The objectives of the project were successfully met in all aspects. There were no signs of damage to any of the valver ateributable to the transient, specifically the liquid dis-charge. The safety val /e that actuated performed under con-trolled conditions as it did in the plant. The safety valve that did not actuate would have, had the primary pressure increased to around 2470 psig. No evidence of damage was found during the inspection of the discharge piping. iv l
6 . . The cooperation of Florida Power Corporation, particularly the staff of Crystal River Unit No. 3, is gratefully acknowledged. J. D. E. Jeffries, Project Manager Nuclear Power Division l v
4 ABSTRACT I I The PWR Safety and Relief Valve-(S/RV) Test Program was developed \ to respc"'d to NRC requirements placed on the utility industry via Sectior. _.1.2, NUREG 0578. The Program is focused on determining the operability of safety and relief valves found 6n the primary side of PWRs by full scale prototypical testing. In addition to , testing valves under controlled conditions, the Program established I an objective to utilize applicable nuclear plant operating events e- to assist in evaluating S/RV performance. On February 26, 1980 a transient occurred at Florida Power Corporation's Crystal River Unit 3. This event caused the power operated relief valve to open; l and later, led to lifting a safety valve. This report describes l the transient and the subsequent valve actuations. Moreover, it provides the results of a comprehensive examination of the valves and the associated overpressure protection piping. L f l vii _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ u
l TABLE OF CONTENTS t I. Introduction II. Summary and Conclusions III. Discussion l IV. References V. Appendices A. Inspection and Test Guidelines for Crystal River-3 Safety. Valves B. Test Procedure for Steam Set Pressure and Leakaga Testing of Sprir.g-Operated Safety Valves f C. Dresser Pressurizer Safety Valve Inspection, Refurbishment and Test Report i. i
I. INTRODUCTION 1 In February 1980, an abnormal transient occurred at the ; 1 I Crystal River Unit No. 3 (CR-3) Nuclear Generating Station. The cause of this transient was an electrical system malfunction which opened the power operated relief valve (PORV) on t.e . pressurizer and held it open electrically for approximately 5 to 7 minutes until the PORV-was isolated by closing the upstream block valve. A reactor trip, system blowdown to about 1350 psig, and automatic initiation of high pressure injection into the reactor coolant system occurred within the first few minutes of the transient. The addition of high pressure injection flow filled the pressurizer, increased reactor coolant system pressure, and eventually resulted in actuation of one of the two spring-loaded safety valves installed on the pressurizer. This safety valve subsequently cycled open and closed over a period of up to 2 hours. During the course of the transient, the PORV opened and discharged saturated steam (and possibly water) and the safety valve opened on steam and/or water and subsequently discharged solid water. The transient was terminated without damage or other adverse effects. However, because of the fact that the safety valve (and possibly the PORV) discharged water flows for which they were not designed, a program was undertaken by EPRI to
i I evaluate the effect of these conditions on the valves and connecting piping. ). The purpose of this report is to present the results of these j investigations. - Specifically, this report describes (1) the safety and relief valves and overpressure protection system installed at' Crystal River Unit No. 3, (2) the transient which occurred at Crystal River Unit No. 3,_and (3) the results of j examinations and tests.of the Crystal River Unit No. 3 PORV, safety valves and discharge piping. The cooperation and the assistance of representatives of j Florida Power Corporation, Babcock and Wilcox, Dresser Industries, and Wyle Labs, who participated in this effort are gratefully acknowledged. i I i l t s 1 f o h i k i ). !
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SUMMARY
AND CONCLU3 IONS During the February 26, 1980, transient at Crystal River Unit No. 3 (CR-3), the PORV was opened and held open as a result I of an electrical system malfunction. The PORV remained open for approximately 5 to 7 minutes at which time it was isolated by closing the upstream block valve. The PORV was actuated at a system pressure of approximately 2200 psig on saturated steam, discharged steam (and possibly water) and was isolated at a reactor coolant pressure of approximately 1350 psig. During sabsequent re-pressurization of the reactor coolant system by means of the high pressure injection pumps, one of the two installed safety valves, RCV-8, opened by self-actuation at a reactor coolant system pressure of about 2400 psig. This pressure is approximately 100 psi, or 4 percent below its specified set pressure. Safety valve RCV-8 subsequently cycled open and closed a number of times (or remained partially open) between 2400 and 2300 psig over the next 2 hours until the revetor coolant system pressure was reduced below this pressure range. Since the pressurizer was solid (i.e., filled with water) during this period, it is believed that RCV-8 discharged water for a significant period of time. The other safety valve, RCV-9, did not open during this transiert. The evaluation of the effect of those transient flow conditions on the CR-3 PORV and safety valves consisted of the following:
l Inspection of the valves, piping and pipe supports at CR-3 after the transient. Disassembly and inspection of the PORV and safety valves for signs of abnormal wear or other distress. Steam testing of safety valves RCV-8 and RCV-9 at Wyle Laboratories. Post test inspection of safety valves RCV-8 and RCV-9. The main results and conclusions of these investigations and tests can be summarized as follows:
- 1. PORV and Safety Valve Discharge Piping and Sup, orts Visual examination of the discharge piping for the PORV and adjacent spring-loaded safety valves RCV-8 and RCV-9 revealed no damage or deformation in the piping or piping supports.
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- 2. PORV Visual examination and electrical actuation tests of the PORV solenoid operator prior to removal of the PORV from the system confirmed that the solenoid operator and linkages to the valve pilot operated properly when energized and de-energized reper'edly from the control room.
- Examination of the PORV after disassembly by Babcock and Wilcox, Dresser and EPRI representatives indicated that the PORV assembly and internal parts were normal in every respect and showed no evidence of malfunction or damage as a result of the February 1980 transient or previous CR-3 operation.
- 3. Safety Valves RCV-8 and RCV-9 Inspection of safety valves RCV-8 and RCV-9 after disassembly indicated that RCV-8 had been leaking for some time in service, but there was no evidence of damage or abnormal wear due to the February 1980 t transient. Specifically, RCV-8 showed no evidence of instability or excessive loads due to the water discharge which occurred during the February transient.
Steam tests at Wyle Laboratories showed that safety valve RCV-9 had a lift pressure within approximately 1 percent of the specified set pressure as received by Wyle. Safety valve RCV-8 exhibited a lift pressure on tests approximately 100 -psi low in the same manner that it operated during the February 1980 transient at Crystal River Unit No. 3.
- While the reason for the low set pressure was not I specifically identified, it does not appear that l it was in any way related to the February transient.
Instead, it is likely that the valve operated at a low lift pressure as a result of either improper adjustment of the set pressure initially, the effect of leakage through the valve, or both. Valve RCV-8 had a history of seat leakage in the plant, and also leaked during the steam tests at Wyle. In summary, the steam and water discharge through the CR-3 PORV and safety valve RCV-8 during the February 1980 transient , did not result in damage, abnormal wear, unstable operation, , or other distress. l l
III. DISCUSSION The CR-3 overpressure protection system, the transient which occurred at Crystal River in February 1980 and results of subsequent examinations and tests of the CR-3 PORV and safety valves are described below. A. Crystal River-3 Overpressure Protection System The Crystal River-3 nuclear generating station is a pressurized water reactor designed by Babcock and Wilcox and owned and operated by Florida Power Corporation. The reactor system includes a single power operated relief valve (PORV) and two spring-loaded safety valves which are mounted on top of the pressurizer. The safety valves are designated RCV-8 and RCV-9; the PORV is designated RCV-10. i The PORV is provided with a remotely operable steam isola-tion (block) valve. The PORV and safety valves discharge to a drain tank inside containment. The arrangement of these valves and discharge piping at the top of the pressurizer are shown in Figure 1. The specific types and characteristics of the PORV and safety valves are as follows:
- 1. -PORV Manufacturer - Dresser Industries Model No. - 31533VX-3 Electroma cic Relief Valve Size 1/2" x 4"
Orifice Size 5/16" Capacity - 150,000 lbs/hr of saturated steam Set Pressure - 2450 psig A cross-section of the Dresser Electromatic PORV is shown in Figure 2.
- 2. Safety Valves Manufacturer - Dresser Industries Model 1/2 - 31 39A Size 1/2" x 5" Orifice Size - 2.545" Capacity - 318,000 lbs/hr of saturated steam Set Pressure - 2500 psig i
Specified Blow-down - 2% to 4% A cross-section of the Dresser safety valves is shown ( in Figure 3. B. Description of February 1980 Transient i ) l A reactor coolant system transient occurred at CR-3 on February 26, 1980, as a result of an instrument and control system electrical malfunction. This malfunction caused a reduction in feedwater flow to the steam generators, a slight increase in reactor power, and a t resulting increase in the reactor coolant system pressure. l The electrical system malfunction also produced a signal which opened the PORV and held its pilot valve operator open. Under these conditions, the reactor coolant system pressure continued to rise for approximately 23 seconds,. at which time both reactor and turbine trips were initiated automatically at the overpressure trip point of 2300 psig. The reactor coolant system pressure peaked at 2320 psig and as a result of the reactor trip and the open PORV, dropped to 1500 psig within 3 minutes of the beginning of the transient. At this point, two additional high pressure injection pumps automatically initiated injection into t' eactor coolant system (one high pressure injection pump was already operating for normal system make-up) and provides e. total injection flow into the system of about 1100 gpm. At this point in time, the plant operators f secured the reactor coolant pumps and isolatad the open PORV using the upstream isolation valve. It is estimated-that the PORV was isolated at about 5 to 7 minutes after i I initiation of the transient. With the high pressure injection pumps still on and the PORV isolated, the reactor coolant system pressure , increased. After approximately 9-1/2 minutes into the j transient, safety valve RCV-8 opened at a pressure of about 2400 psig. For the next 20 minutes, it appears that I RCV-8 remained open and then reclosed at approximately 2300 psi . A portion of the transient is shown in Figure 4. ! Subsequen _y, the operators throttled the high pressure-i ! injection flow rate and opened let-down system isolation r
valves to permit let-down flow from the reactor system; however, the recorded system pressure traces indicate that safety valve RCV-8 cycled open and closed a number of times (or perhaps remained in a partially open condition) over the next two hours. The reactor' coolant system pressure during this period is shown in Figure 5. (Note that power to the strip chart recorder was lost during , the first 20 minutes of the transient. Figure 4 should be referred to for preasures during this time period.) During this two hour period, the pressurizer was solid. E"entually, the operators re-established a steam bubble
#.n the pressurizer and pressure was reduced to the 1800 -
l 2000 psi range. At this time, RCV " remained closed. At no time during the transient did RCV-9 open. During the course of the transient, approximately 43,000 gallons of reactor coolant were dumped into the reactor building via the reactor coolant drain tank (RCDT) overflow, i The fluid conditions to which the PORV and safety valve were subjected during this transient are as follows: , 1
- 1. PORV - The PORV was opened by the electrical signal due to the instrumentation and control system malfunction about 2200 rsig and was isolated by the upstream block valve at approximately 1350 psig. It is most likely that the PORV opened on and discharged 1
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saturated steam during the approximately 5 to 7 minutes that it was discharging. This is consistent with the pressure trace given in Figure 4. However, there is a possibility that the pressurizer could have filled solid while the PORV was still open (at about 6 minutes into the transient) and that the PORV was isolated at about 9 minutes into uhe transient. In thi eventuality, the PORV could have discharged saturated or s?.ightly j sub-cooled water.
. Safety Valve RCV Safety vs.lve RCV-8 opened initially at 2400 psig, which is about 100 psi below its initial set pressure. A reasonable itterpretation of the pressure trace given in Figura i s that after PORV isolation, at about 6 minutes af transient initiation, the pressurizer filled solid at about 9 minutes. Safety valve RCV-8 opened after about 9-1/2 minutec. In this j case, RCV-8 would have opened on water and disch/rsed i
i saturated and sub-cooled water over the next two hours. j It is possible that a small amount of steam was present i in the pressurizer nozzle under safety valve RCV-8 l l during its initial " pop." It is very likely, however, l l that the second and any subsequent " pops" of RCV-8 e occurred with only water in the pressurizer and safety valve nozzle and that RCV-8 discharged water for [ j essentially all of the time it was open. l l l r l i i i l
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- 3. Safety Valve RCV Valve tail pipe thermocouple measurements during this transient indicate that safety valve RCV-9 did not open or leak during the i
event. These measurements are considered valid since individual discharge lines run from each valve to the RCDT. C. On-Site Examinations and Tests In April ]v80, PORV RCV-10 and safety valves RCV-8 and RCV-9 were removed from the CR-3 plant. During this operation, a number of inspections and tests were performed by representatives of Florida Power Corporation, Babcock and Wilcox, Dresser and EPRI. The results of these examinations and tests are as follows:
- 1. Visual examination of the PORV and its operator, including the solenoid linkages, showed no abnormal conditions or evidence of damage. The operator linkages were in the normal, closed position.
- 2. Electrical actuation tests were made which confirmed that the solenoid and linkage to the PORV pilot operated properly when energized and de-energized ,
repeatedly from the control room.
- 3. Visual examinations of the discharge piping for the PORV and the adjacent spring-loaded safety valves (RCV-8 and -9) indicated no damage or deformation to the piping or supports. A portion of this piping and supports is shown in Figure 1.
- 4. There was no evidence of cold spring in the_ discharge piping when the valves were removed.
The PORV was removed to the radioactive machine shop for disassembly at CR-3. Examination of the valve inlet and outlet ports prior to disassembly showed no. signs of leakage in the form of steam cutting or boric acid crystal accumulation. The following observations were made during PORV disassembly and subsequent examination of valve internal parts.
- 1. The valve pilot guide (Part 12) was discolored on its OD, but showed no signs of damage. The ID of the pilot disc guide had slight circumferential wear marks at the top of the guide. These marks were visible, but could not be identified by feel, and are considered normal wear marks. The mating surfaces of the pilot disc showed no wear of any kind.
- 2. The pilot valve bellows (Part 14A) were intact and appeared undamaged.
- 3. The pilot disc (Part 10) had a 1/8-inch long discolor-ation on its seating surface. The discolored area could not be felt by hand, but could be the result of slight pilot leakage at some time in the past.
- 4. The main disc retaining plate and locking devices were in place and undamaged. It was noted that the central
screw in the retainer locking screw was twisted off. It'is apparent that this had happened on a prior disassembly of the valve, as the intact part of the screw was staked in place and the lock wire had been attached to a rib in the valve cage.
- 5. The main disc (Part 3) and seat showed no evidence of leakage.
- 6. There was no metal upsetting on the disc, the seat or the disc stop, indicating the valve had not chattered or been subjected to excessive opening or closing loads.
- 7. The main disc guide (Part 5) showed no significant scoring or wear.
I Photographs of selected parts as-disassembled are shown in Figures 6 through 10. i It was concluded by Babcock and Wilcox, Dresser Industries and EPRI representatives that the PORV assembly and internal parts were normal in every respect and showed no evidence of malfunction or damage as a result of the February 1980 transient or previous operation. The safety valves (RCV-8 and -9) were removed for testing, disassembly and refurbishment at Wyle Laboratories, Huntsville, Alabama, and were not examined at CR-3. However, it is signficant that Florida Power Corporation representa-tives indicated that safety valve RCV-8 had a history of
1 1 leakage problems since its installation and had been refurbished by Florida Power Corporation maintenance personnel one or more times at CR-3 during the previous years of operation. Specifically, valve RCV-8 was reportedly leaking prior to the CR-3 transient in February 1980. D. Laboratory Examinations and Tests of Safety Valves RCV-8 and RCV-9 The inspections and tests of safety valves RCV-8 and RCV-9 were performed at Wyle Laboratories in accordance with the Inspection and Test Guidelines given in Appendix A. These guidelines require the following: Disassembly and inspection of RCV-8 as received at Wyle. A special disassembly procedure was specified to remove the valve bonnet assembly without removing the main spring or changing the main spring preload. Reassembly and test of RCV-8 without refurbishment or change in any settings. The test was performed at rated temperature and pressure on steam, but at reduced capacity and valve lift. Complete disassembly and reinspection of RCV-8 after steam test. Refurbishment and re-certification test of RCV-8. Test of RCV-9 as received at Wyle. 1
Disassembly and inspection of RCV-9 after steam test. Refurbishment, reassembly and re-certification test of RCV-9. The main results of these inspections and tests are summarized below.
- 1. Inspection of RCV-8 Safety valve RCV-8 was examined and photographed as received, was cleaned sufficiently to permit handling and was disassembled for further examination. Results of these inspections indicated the following:
t The internal surfaces of the valve showed evidence of leakage. A black deposit, believed to be mag-netite, was apparent over the majority of the internal surfaces of the valve. Photographs of RCV-8 prior to disassembly are presented in Figures 11 through 13.
- The valve seat and disc mating surfaces were l
extensively steam cut, bteam cutting resu~ted in several hundred radial marks across the sec ing faces with depths of several mils. The lower surface of the disc holder (Part 11) adjacent to the disc also showed erosion damage. Liquid penetrant examination of the seating surfaces of the seat and disc and the lower surface of the disc holder showed fine radial cracks 30-60 mils deep, scattered around the lower surface of the disc holder. These cracks were not visible to the unaided eye. The seat and disc were steam cut, but not cracked. Inspection of c'he adjusting rings, inlet nozzle, back pressure balancing bellows, disc guide and spindle revealed no evidence of damage, upset l metal, galling or scoring. In particular, the
- mating surfaces which guide the spindle and disc assembly and the back stop area which limits the stroke of the valve showed no evidence of excessive loads, unstable operation or banging of internal parts.
Two abnormal conditicns were found. First, the anti-rotation pin (identified as Part 8d in Figure 3) was found dislodged inside the bonnet cavity. This pin was bent and had become disen-gaged from the upper spring washer and fallen into the valve bonnet cavity. This anti-rotation pin serves no functional purpose in the valve. Rather, its purpose is to prevent rotation of the upper spring washer (Part 8 c) during adjustment of the set pressure comprescion screw (Part 9). Since the upper spring washer foes not move during valve operation, the defonaation of the pin had to have occurred due to improper alignment of the pin and the mating slot in the bonnet during the initial assembly and compression of the main spring. According to the Dresser representative (Mr. Bolgeri, this problem has occurred in a number of applications and this feature of the valve design has been changed in later valves. Binding of this anti-rotation pin on the bonnet during set pressure adjustment could lead to misalignment of parts and thereby promote leakage (which obviously did occur), but it would not appear that this pin would have any effect on valve operation (subsequent tests of the valve without the pin confirmed this). Second, the clearances between the disc holder (Part 11), disc (Part 5) and disc nut (Part 6B) were incorrect. Figure 14 shows the bellows asscmbly and indicates the clearances that are specified between the disc, disc holder and disc nut to permit slight movement or rocking of the disc on top of the disc nut in order that the disc can self-align with the seat. Inspection of this assembly showed that insufficient clearance was available to permit any rocking or self-aligning of the disc. The inspection showed that assembly of the disc holder on the disc nut firmly retained the disc in the assembly. As a result, the disc was not able to self-align readily on the seat. The differences between the specified and actual clearances could be due to initial assembly errors or deformation of the parts in service. This dimensional discrepancy could explain the observed valve leakage history, but would have no effect on the operation of the valve or its response to steam or water flow. Results of dimensional measurements of this assembly are shown in Figure 14 and explain why there was insufficient clearance to prevent motion of the disc in the assembly. I,
- The set pressure compression screw on valve RCV-8 did not have the standard lock wire and Dresser
- seal normally installed to certify that the valve has been adjusted. Also, a piece of wire similar to bailing wire was installed to lock the adjusting ring pins. No Dresser certification seal was used.
Photographs of RCV-8 parts after disassembly are given in Figures 15 through 18. Following the inspection of the parts of RCV-8, the valve was assembled without refurbishment in the same condition as received by Wyle. As discussed above, care was taken not to disturb the spring preload (set pressure) or adjusting rings.
- 2. Test of RCV-8 Reassembled valve RCV-8 was installed in the Wyle low capacity test facility for set pressure tests. The t
l procedure used for these tests is the standard Wyle procedure for set pressure verification of-Dresser safety valves and is included in Appendix B. In these tests, the safety valve is installed in the test facility,
'is allowed to heat up due to steam at the valve inlet at approximately 2200 psi and 650*F until temperatures recorded at the valve inlet, valve body, and the valve bonnet reach temperatures specified in the procedure and stabilize. This procedure takes 3 to 4 hours.
Following temperature stabilization, valve RCV-8 was popped three times. The lift pressures were 2392 psig, i 2388 psig and 2388 psig. Opening and closing of the
valve was normal during these tests with the exception that simmering of the valves occurred 1-2 seconds before each lift, due to the leakage from the valve. Measured valve lift and pressure traces are shown in Figures 19, 20, and 21. The leakage rate over the 4-hour period that the valve was installed on the test loop at approximately 2200 psi was between 1 and 1-1/2 gpm. The name plate set pressure for this valve is 2500 psig. It was noted thu the lift pressure of about 2390 psig is approximately the same pressure at which this valve lifted during the transient at CR-3 in February 1980. The reason for the depressed lift pressure could not be determined as a result of these inspections, however, the most likely explanations are:
- a. The valve liftec prematurely because of the effect of the leakage of steam through the valve, or
- b. The valve was not adjusted properly initially.
In any case, the repeatable performance of the valve at CR-3 and in the test loop indicates that the lower f than specified lift pressure was not related to the steam-water transient at CR-3.
- 3. Post-Test Inspection of RCV-8 Following completion of the performance test of RCV-8 it was completely disassembled and inspected. This inspection showed no change in the appearance of any l
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parts as compared to their condition prior to test. Photographs of RCV-8 parts following steam testing are presented in Figures 22 through 30. In summary, there was no evidence of any damage due to the test or previous operation in the plant, except for the steam cutting and erosion of the seat, disc and adjacent surfaces due to the 1 to 1-1/2 gpm leak rate. Following these inspections, RCV-8 was completely refurbished and reassembled for set pressure adjustment and verification test.
- 4. Steam Test of RCV-9 Safety valve RCV-9 was inspected and photographed as received and was installed on the test loop for set pressure test. Photographs of RCV-9 e9 received are shown in Figures 31 through 73. *;he test procedure and facility used were the same as for RCV-8. The tests of RCV-9 resulted in lift pressures of 2472 psig, 2466 psig and 2466 psig. There was no steam leakage before, during or after these tests. Operation of the valve was normal in every respect.
- 5. Inspection of RCV-9 Safety valve RCV-9 was removed from the test facility, disassembled and inspected. Results of these inspec-tions showed no evidence of distress, leakage or abnormal operation. The one problem noted was that, similar to the situation in the case of valve RCV-8, the anti-rotation pin (Part 8D) was missing from the upper spring washer (Part 8 C) . While this pin could not be located, it was apparent from the elongation of the hole in the upper spring washer that this pin had also been bent and dislodged at some point, probably a during assembly of the valve. The lack of this pin obviously had no effect on the performance of RCV-9.
Following these inspections, valve RCV-9 was decon-taminated, lightly lapped and reassembled for set pressure adjustment and verification tests. Photographs of RCV-9 parts after disassembly are shown in Figures 34 through 37.
- 6. Decontamination of RCV-8 and RCV-9 During disassembly, inspection and reassembly work on valves RCV-8 and 9, all parts were decontaminated by wiping and, in the case of the valve internal parts and trim, by the use of an ultrasonic bath. Following cleaning of the valves, the valves retained sufficient contamination to preclude unrestricted chipment. For this reason, refurbished and re-certified valves RCV-8 and RCV-9 were returned to Florida Power Corporation for use as plant spares.
Pertinent data from the hyle Laboratolies report of the testing of RCV-9 and RCV-9 are included in Appendix C. 1
i The results of the inspections and tests of RCV-8 and RCV-9, as summarized above, indicated the following:
- a. Inspection of valves RCV-8 and RCV-9 revealed no evidence of damage or abnormal wear due to the I plant transient or shop tests. Specifically, RCV-8 t
showed no evidence of instability c excessive leads due to the steam or water discharge which occurred during the February 1980 transient. ( b. Safety valve RCV-9 had a lift pressure approxi-mately within tolerance as received by Wyle. Valve RCV-8 exhibited a lift pressure approxi-l mately 100 psi low in the same manner that it i operated during the February 1980 transient at CR-3. While the reason for the low set pressure was not specifically identified, it does not appear that it was in any way related to the circumstances of the February 1980 transient. Instead, it is likely that the valve operated at a low lift pressure as a result of either improper l adjustment of the set pressure initially, the effect of the leakage through the valve, or both.
- c. Contamination levels of the valves after decontam-ination by Wyle are still such that the valves are not suitable for unrestricted shipment. As a result, it will probably not be possible to test valve RCV-8 or RCV-9 in the EPRI full flow safety and relief valve test program.
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QTY. NOMENCLATURE NO 1 1 MAIN BASE PILOT BASE ASSEM. (WELDED lNTEGRAL ASSEM.) 1A 1 INLET FLANGE IB 1 OUTLET FLANGE IC 1 CAGE ID 1 TUBEINSERT lE 8 MAIN BASE INLET STUD IF 1 PILOT BASE IG 4 PILOT BASE STUD
- 2 8 INLET STUD NUT 3 1 MAIN DISC 3A 1 PISTON RING 4 1 M AIN DISC SPRING S 1 GUIDE 6 1 GUIDEGASKET 7 1 GUIDE 4ETAINER PLUG 8 1 RETAINER PLUG CAP SCREW 8A 1 CAP SCREW LOCKWASHER 8B 1 LOCK SCREW 8C 1 LOCK SCREW LOCKWASHER 9 1 SEAL WIRE 10 1 PILOT DISC 11 1 PILOT DISC SPRING 12 1 SEAT BUSHING 12A 1 LOWER GASKET 128 2 UPPER GASKET 13 1 LOWER U iNDLE 14 1 BELLOY.SASSEM.
(WELDED, INTEGRAL ASSEM.) 14A 1 BELLOWS 14B FLANGE 14C 1 PISTON 15 1 UPPER SPINDLE 16 4 PILOT STUD NUT 17 1 SOLEN 0ID BRACKET 18 1 LEVER 19 1 LEVER PIN ASSEM. 19A 1 SHOULDER SCREW 198 1 NUT Figure 2 (Cont'd)
o QTY. NOMENCLATURE 19C 1 BRACKET BUSHING 19D 2 LEVER BUSHING 19E 1 COTTER PIN 20 1 ADJUSTING SCREW 20A 1 LOCKNUT
'.1 1 BRACKET PLATE a i 4 BRACKET PLATE CAP SCREW 22A 4 LOCKWASHER 23 1 SOLEN 0ID 24 4 SOLEN 0ID CAP SCREW 24A 4 LOCKWASHER 25 1 PLUNGER HEAD 26 1 LEFT HAND SPRING GUIDE 27 1 RIGHT HAND SPRING GUIDE 28 2 PLUNGER SPRING 29 2 PLAIN SPRING WASHER 30 2 SPRING COTTER PIN 31 2 GUIDE BRACKET 32 1 GUIDE BRACKET BOLT 32A 1 LOCKWASHER 328 1 NUT 33 1 SWITCH 34 2 SWITCH MACHINE SCREW 34A 2 LOCKWASHER 35 3 SPRING GUIDE CAP SCREW 36 1 SPECIAL SPRING GUIDE SCREW 37 4 SPRING GUIDE NUT-37A 4 LOCKWASHER
- 38. 1 BRACKET COVER ASSEM.
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TYPE 21/2 31739A 1 -W I?_cv 6 f t) f0' .3 .8 3-31739A1 31749A 1 4 5-117 70 3 31759A1 1 l'i5. 7 17/8 0 4 31749A1 31759A1 45 117 ~fd'
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TYPE 31la 31739A.I 3 31739A.I 3 31759A l 4 31749A.) RIF NO QIf NOMINCL AIURE I BASE ASSEMBLY 1A 1 BASE 18 1 OUTLET FL ANGE IC 1 N0ZZLE ID 1 C Scal or E,asket IE 4 WEB 2 1 BONNET 2A 1 BONNET PLUG 3 12 BONNET STUD 4 12 BONNET STUD NUT 5 1 DISC 6 BELLOWS ASSEMBLY 6A 1 BELLOWS 6B 1 DISC NUT 6C 1 FLANGE GD 1 FLANGE ADAPTOR 7 i SPINDLE 8 SPRING & WASHER ASSEMBLY 8A 1 SPRING 8B 1 BOTTOM SPRING WASHER 8C 1 TOP SPRING WASHER 8D 1 PIN 9 1 J
' 0MPRESS10N SCREW 10 1 COMPRESSION SCREW NUT 11 1 D!SC HOLDER 12 1 GUIDE 13 2 GulDE GASKET 14 1 SUPPORT PLATE 15 2 SUPPORT PLATE GASKET 16 1 WASHER RETAINER 17 1 FLOATING WASHER 18 1 RETAINER CAP SCREW 19 1 LIFT STOP 20 1 LIFT STOP COTTER PIN 21 1 DISC COL L AR 22 1 DISC COLL AR C0I1ER PIN 23 1 UPPER ADJUSTING RING 24 1 UPPER ADJUSTING RING PIN Figure 3 (Cont'd)
25 1 MIDDLE ADJUSTING RING 26 1 MIDDLE ADJUSTING RING PIN 27 1 LOWER ADJUSTING RING 28 1 LOWER ADJUSTING R.ING PIN 29 4 PIN GASKET 30 1 CAP 31 1 CAP GASKET 32 6 C'AP ST UD 33 6 CAP ST UD NUT 34 1 LEVER 35 1 LIFTING FORK 36 1 LEVER NUT 37 1 LEVER SHAFT 38 1 PACKING 39 1 PACKING NUT 40 1 COLLAR 41 1 RET AINING RING 42 1 RELEASE NUT 43 1 RELEASE NUT COTTER PIN 44 1 GAG 45 1 GAG PLUG 46 1 GAG PLUG GASKET Figtire 3 (Cont'd)
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IV. REFERENCES
- 1. Babcock and Wilcox, Nuclear Power Generator Division.
Transient Assessment Report, Reactor Trip at Crystal River - 3 Nuclear _ Station on February 26, 1980 (Preliminary Report), Lynchburg, Va., March 9, 1980
- 2. Institute of Nuclear Pcwer Operations and Nuclear Safety Analysis Center. Analysis and Evaluation of Crystal River - Unit 3 Incident, March 1980.
- 3. Wyle Laboratories, Scientific Services and Systems Group. Dresser Pressurizer Safety Valve Inspection, Refurbishment and Test Report, Huntsville, Alabama, 1980
- 4. Report of MPR Visit to Crystal River Unit No. 3 on April 23 and 24, 1980.
- 5. Report of MPR Visit to Wyle Laboratories on July 16 and 17, 1980.
l i V. APPENDICES 1 A. EPRI. Inspection and Test Guidelines for Crystal River-3 Safety Valves, 1980, pp. 124.
- B. Wyle Laboratories. Test Procedure for Steam Set Pressure and Leakage Testing of Spring-Operated Safety Valves, February 1980, pp. 1-14.
C. Wyle Laboratories. Dresser Pressurizer, Safety Valve Inspection, Refurbishment and Test Report, -, j 1980, title page through page 29. 4 Y 9 .l
- v.en, , , - ~ - -~----n, -- - - ~- - , e
ELECTRIC POWER RESEARCH INSTITUTE INSPECTION AND TEST GUIDELINES FOR CRYSTAL RIVER -3 SAFETY VALVES l
1.0 Purpose
The purpose of this document is to specifiy and define the require-ments associated with tests and inspections of Crystal River Unit 3 (CR-3) safety valves RCV-8 and RCV-9. These valves were installed and functional during a system transient that took pla'ce at CR-3 February 26, 1980. During this transient, RCV-8 opened and closed, perhaps several times, and disc.harged steam and water. RCV-9 apparently did not open during the transient.
2.0 Objectives
The objective cf these inspectiois and tests is to determine the present condition of the valves a::i to identify any effects of the transient - and specifically, the effects of steam and water flow through RCV on the integrity and operability of the valves. It is also desired tc determine the reason that valve RCV-8 opened at an indicated system pressure of about 2400 psig-in lieu of the specified valve set pressure of 2500 psig.
3.0 Prerequisites
The inspections and tests are considered part of- the PWR Safety and Relief Valve Test Program. Therefore EPRI will direct the various parts of this project either through an EPRI staff member or other, authorized agent at the Wyle facilities. All work accomplished during this effort will be done by qualified Wyle or Dresser Industries personnel in accordance with standard approved procedures. Na valve modifications or refurbishment will be done unless approved by EPRI and under the direct supervision of a qualified Dresser Industries representative. Qualified Wyle personnel may alter or modify a valve if approved by EPRI and authorized in writing by Dresser Industries.
4.0 References 4.1 Technical Service Agreement with Wyle Laboratories. 4.2 Wyle Quotation No. 544/4002/CP. 4.3 Leti.cr .L O.E. Jeffries (EPRI) to Ray Scates (Wyle) dated March 24, 1980. - t 4.4 Wyle Test Procedure No. 1009 for Steam Set Pressure and Leakage l Testing of Spring Operated Safety Valves. ' 4.5 Wyle Laboratories Quality Assurance Fclhier :r.J Procedures Manual. 4.6 Instruction Manuals for Dresser Spring Operated Safety Valves Model 31739A. 4.7 Dresser Special Procedures for Disassembly of Spring Operated Safety Valves (to ensure original set pressure and spring preload). 5.0 Time Required (To be estimated by Wyle later). 6.0 Overview 6.1 Inspection - both safety valves, RCV-8 and RCV-9, will be inspected prior to having any work performed. Exterior con-dition will be noted and photographs will be taken. The valves will then be disassembled and the internal components will be examined and critical dimensions checked. Extensive photographs will be made of the disassembled valves. All facets of the inspection will be documented. RCV-8 will be reassembled as close to the "as received" condition as possible. (Note: a special procedure is to be used in the disassembly of RCV a procedure that will maintain the original spring preload and set pressure adjustment). RCV-9 will be refurbished as required to restore it to a condition where it can be returned to CR-3 as a spare pressurizer safety valve. 6.2 Tests - RCV-9 will be tested for leakage and set pressure before disassembly and inspection. Both RCV-8 and RCV-9 will
be checked for set pressure and leakage after refurbishment / reassembly. NOTE: Be careful not to change set pressures or blowdown ring adjustments during the inspection or photography sequence. l 7.0 Inspections and Tests 7.1 Pre-disassembly Inspections - On receipt the valves are to be removed from their packaging, all protective covering is to be removed and the valves set up for inspections and photographs. Placards are to be made for each valve identifying the valve and orientation. These placards are to be utilized to describe the photographs as pictures are taken of the front, both sides, back, and inlet & outlet ports (looking into the valves). Close-up photographs are to be taken of any abnormalities and/ or deposits of foreign material in the valve ports. 7.2 Test of RCV Following completion of the inspections outlined in 7.1 above, but without valve disassembly, valve RCV-9 is to be installed in a steam test facility for set pressure verification tests. Set pressure (and, if possible, blowdown) should be measured along with seat leakage (before and after each pop) in accordance with reference 4.4 or approved variation thereof. All results are to be documented in accordance.with reference 4.5. 7.3 Disassembly and Inspection of RCV Valve RCV-8 is to be dis-assembled in accordance with a special disassembly procedure (by Dresser) in which the valve spring, set pressure adjusting nut and yoke assembly are removed without changing set pressure adjustment or the main spring preload. Special tools and/or fixtures may be required for this operation. It is the intent of this procedure that the same set pressure of the valve be retained during valve disassembly and inspection so that the valve can be reassembled in the same configuration (and with the same spring preload) as-received at Wyle. Visual examination shall be made and documented (by written notation and photographs) of each part removed at each major stage of disassembly. Positions
of blowdown adjusting rings shall be noted for subsequent reassembly. Of particular interest are any signs of distress (galling, scoring, pitting, or unusual wear marks) and any signs of binding, chattering or leakage. Close-up photographs shall be taken of any such abnormalities. Following disassembly, the valve ports shall be laid out on a suitable surface (without any cleaning or decontamination) labelled and photographed. 7.4 Disassembly and Inspection of RCV Valve RCV-9 shall be disassembled in accordance with standard Technical Manual pro-cedures and inspected in the same manaer as valve RCV-8,- specified in paragraph 7.3 above. Note: The special pro-cedure utilized to disassemble the valve maintaining spring preload, etc. , is not required for RCV-9. After com-pletion of inspectien, valve RCV-9 shall be (urbished and reassembled under the direction of the Dre ser representativ . 7.5 Re-assembly of RCV Unless valve RCV-8 is determined to be in a damaged condition by Dresser and EPRI representatives, it is to be re-assembled without re-work or seat reconditioning. Photographs, measurements and match marks should be used (as appropriate) to assemble and set RCV-8 as closely as possible in the same configuration as received. 7.6 Final Testing of Both Valves - Both RCV-8 and RCV-9 are to be installed in the steam test facility and have set pressure verification and seat leakage tests in accordance with reference 4.4. 7.7 Results of all inspections and tests will be evaluated by EPRI, Dresser, Wyle, and B&W representatives and the need for any additional inspections or tests detemined. If additional actions are considered necessary, supplemental procedures will be prepared. After completion of all inspections and tests, valves RCV-8 and RCV-9 shall be protected from the environment by storage in a dry, water proof area. RCV-9 s I be prepared for transportation and return to Florida Power Corporation's Crystal River - Unit 3. 8.0 Final Report A report of all inspections and +,ests, including photographs, shall be prepared and transmitted to EPRI; 5 copies are required.
4 APPENDIX B Page No. 68 T 4 Test ReTmt No9 &097 0E6URE TEST PROCEDURE NO. 1009 WYLE LABORATDRIS SC:ENTIFiC SERviCFS AND SYSTEMS GROUP 8 0 8041008 a MUNT5viLLE ALA8AMA 35007 DATE.* February 22, 1980 TWX 18101726-m . TELEPHON( 12%l $374411 TEST PROCEDURE FOR STEAM SET PRESSURE AND LEAKAGE TESTING OF SPRING-0PERATEb SAFETY VALVES APPROVED BY: APPROVED BY -' FOR: ( PROJECT MANAGER:Q - APPROVED BY: APPROVED BY FOR:
~p QUAUTY ENGINEER- -
7 APPROVED BY: PREPARED BY FOR: ' PROJECT ENGINEER- - m REVISIONS ,0,o ,0sA. , , _ ,,,, REv. NO. l OATE l # AGES AFFECTED l Sv AP P'L. CESCRIPTION OF CHANGES j l l I I I ! ; I ! '
! I I I I ! !
! i I
i i I i i COPYRIGHT SV WYLE A80RATORIES THE RIGHT TO REP 900U0E. COPY. EXHISIT. OR OTHERWISE UTILIZE ANY OF THE M ATER!A( CONTa ulTMOUT THE E\ PRESS PRLOR PERMISSION OF wvLE LA80RATORIES 15 PRCHIBITED. THE ACCEPTANCE OF A PURCHASE ORDER IN CON AECT;0N WITH THE M ATEP! AL CONTAINED HEREiN SMALL SE EGutvALENT TO EXPRESS PRIOR PERMISSION.
t*T9eN. W) TestReportNo.4}097-0 009 TEST PROCEDURE NO. PAGE NO. 1.0 PURPOSE The purpose of this test procedure is to present the methods and procedures used in handling and calibrating safety valves. 1
2.0 REFERENCES
2.1 ASME Boiler and Pressure Vessel Code, as applicable. 2.2 American National Standard ANSI /ASME N45.2-1977 " Quality Assurance Program Requirements for Nuclear Facilities". 2.3 Nuclear Regulatory Commission Regulation 10CFR50, Appendix B. 1 2.4 Nuclear Regulatory Commission Regulation 10CRF21. 2.5 American National Standard ANSI /ASME PTC 25.3-1976, " Safety and Relief Valves". 2.6 Customer's Purchase Order and Special Instructions. 3.0 TEST EQUIPMENT AND INSTRUMENTATION 3.1 All test equipment and instrumentation used for the performance of this test program complies with the requirements of Wyle Laboratories' , Quality Assurance Policies and Procedures Manual, which conforms to the applicable portions of ANSI-Nd5.2 and Military Soecification MIL-C-45662. Wyle Laboratories has the option to substitute ecuiv-alent test equipment in lieu of listed equipment, if recuired. 3.2 Calibration of Test Eouicment and System Calibration All test equipment is calibrated on a periodic basis and the calibration interval is displayed on a decal. This decal is affixed to the equip-ment indicating the last calibration date, the next calibration due date, accuracy, and by whom calibrated. In addition to individual component l calibration, prior to and immediately following test, an end-to-end l system calibration is performed on equipment used to establish valve set pressure.
- 3.3 Measuremerts and Tolerances l
Unless specified otherwise, the maximum allowable tolerance on test condition measurements shall be as follows:
- Parameter Tolerance A. Tempera nre
- +/- 4 UF i B. Pressure:
Heise Gauge +/- 0.1% F. S. , Deadweight Tester or Digigage +/- 0.03% Pressure Transducer +/- 1.0 psi i WYLE LABORATORIES 50au '054 7 aev oct 79 Huntsvilla Facshty
Test Report No. 45097-0 TEST PROCEDURE NO. 1009 PAGE NO. 1 4.0 PERSONNEL CERTIFICATION Wyle certifies that all personnel assigned to the steam valve facility are qualified for the tasks assigned. Personnel certification is achieved through personnel education levels, vocational training, and practical experience as outlined in,, ANSI-N45.2. 5.0 STORAGE During any prolonged non-testing period, the test specimens shall be stored in a controlled storage area. The storage shall be maintained in accordance with good laboratory practices, i .e. , being properly protected from grease, oil, solvents, and any surface dirt that could influence the results of the test program. The storage area shall bein compliance with ANSI-N45.2.2, Level C. 6.0 REQUIREMENTS AND PROCEDURES 6.1 Set Pressure Test Procedure 6.1.1.1 A visual inspection shall beconducted for shipment damage and correct receipt of valves. Each valve shall beidentification coded by its serial number. Special attention shall be given to the inlet and outlet flanges. Any condition which would affect the steam tightness at these areas shall be criteria to prevent t.st until the flanges are repaired. 6.1.1.2 Each valve shall beinstalled in its normal operating position on a steam header. 6.1.1.3 The cap assembly and release nut of the safety relief valve shall be removed. A fixture to restrict the lift of the valve shall be installed. Nominal lift restriction values are as follows: I I I ! LIFT
;' RESTRICTION ,
VALVE TYPE MANUFACTURER SERVICE (IN) l 3700 Series Dresser Main Steam Safety 0.100 i ,.
; Style HA '
Crosby ! Main Steam Safety 0.050 Style HB Crosby 0.030 l lPressurizerSafety ; 31700 Series Dresser ' Pressurizer Safety i 0.030 WYLE LABORATORIES FORM 1054
- ilev oct 19 u,,m ...o., rv.ia, b., ii ,
Page No.'/I Test Report No. 45097-0 TEST PROCEDURE NO. l009 PAGE NO. 4 6.0 REQUIREMENTS AND PROCEDURES (Continued) 6.1.2 , Instrumentation 6.1.2.1 A linear variable-differential transformer (LVDT) shall beinstalled on the spindle to measure the lift of the disc. 6.1.2.2 The instrumentation used to determine set pressure (pressure trans-ducer, X-Y plotter) shall be subjected to an in-system calibration as required to achieve the desired accuracies prior to and subsequent to testing by use of a deadweight tester. 6.1.2.3 Thennoccuples and a pressure transducir shall be mounted to the test valve as shown in Figure 1. 6.1.3 Set Pressure Test 6.1.3.1 Saturated steam at 90% of the set pressure shall be applied to the valve inlet. The steam pressure shall bemaintained until the spring temperature is stabilized at 140 +/- 5 F or as otherwise directedgby the customer and the valve inlet neck temperagure is at least 350 F for Dresser 3700 and Crosby HA valves and 400 F for Dresser 31700 and Crosby HB valves. 6.1.3.2 A coiled steam line may be placed around the spring housing to assist temperature stabilization of the spring. Insulation. blankets may also be wrapped around the valve to assist in heating the valve. 6.1.3.3 After temperature stabilization of the valve spring for 15 minutes and with steam pressure maintained at 90% of set pressure, the steam pressure shall be increased until the valve disc lifts off its seat. The valve set pre.sure established at this point shall be recorded. The valve set pressure must be within its nameplate set pressure +/- 1%. A total of three consecutive valid set pressure tests shall be perfonned. 6.1.3.4 The spring stabilization temperature shall be maintained during the test by adjusting the insulation and/or the amount of steam through the coil. If the spring temperature rises above the tolerance, after removing the insulation and turning off the steam coil, testing may continue and data shall be marked " Natural temperature of spring '. 6.1.3.5 If the measured set pressure does not meet the criteria of paragraph 6.1.3.3, the valve shall be adjusted and checked again. 6.1.3.6 Wait times between valve actuations shall be a minimum of five minutes. ii.1.4 f.c.1k Che ck 6.1.4.1 At the completion of the set pressure , iibration, the valve shall be pressurized to 90% t/- 10 usi of nameplate set pressure. A leak check sha11 then be conducted. WYLE LABORATORIES Fopu tc54 7 % oci 79 N tJf if Systle Evsitty
Page No. 72 Test Report No. 45097-0 TEST PROCEDURE NO. 1009 PAGE NO. 5 6.0 REQUIREMENTS AND PROCEDURES (Continued) 6.1.4 Leak Check (Continued) 6.1.4.2 The leak check shall be performed as follows: A cold ( < 100UF) mirror about 2 inches square and 1/8 inch thick shall be passed around the disc to seat interface and the mirror inspected. If no moisture or a faint fogging is detected, it may be concluded that essentially zero steam leakage is ,present. If the mirror surface shows condensation droplets, the leakage is unacceptable and the valve must be repaired per Reference 2.6. If in doubt about whether the condensation is leakage or faint fogging, holding the mirror at the suspect area for a longer period of time will produce increased condensation droplets if steam leakage exists. The mirror shall not be allowed to heat up. 6.1.5 Documentation 6.1.5.1 Test Loo A test log shall be maintained and shall include a daily cescriotion of the testing perfomed and any pertinent information regarding , status of the test specimen.. The log s'1all be a complete chronological log including details of all test setups and calibration, specimen hand-ling and setup, installation, and test data sumaries. 6.1.5.2 Test Data The recordi~ng and ch' art paper shall be reviewed for accuracy and ,cuality after each test. The test data shall be clearly identified with the valve serial number, Wyle job number, test date, customer, chart speed, remarks, or any other pertinent information required for data analysis or data retrieval. 6.1.5.3 Test Report Three copies of a certification test report shall beissued subsequent to completion of calibration. The certification test report shall consist of: e A brief summary of the valve's history during the calibration process, and calibration data points.
- The_ set pressure recordings, e The instrumentation list indicatino the in-struments used and the instruments' calibration.
dates. A sample report is presented in Appendix I. WYLE LABORATORIES Foau 1o54 2 new oct 79 Hurtsydle Facihty
Page No. 73 Test Report No. 45097-0 TEST PROCEDURE NO. 1009 PAGE NO. O E-1i lI I f 1 l 1 I T = 3 (Band) NOTE: Band T3 to bonnet housing if spring ( ", is enclosed. \ / L 4 [ [ 8 r l T2 = (Band) NOTE: Install under inlet flange nut if banding 1s impractical. P), T) , (Steam) - Figure 1 Thermocouple and Pressure Transducer Locations WYLE LABORATORIES scaw50547 % oci79 i e..... w.ii,. r .. ... e ,
Page No. 74 Test Report No. 45097-0 VEST PROCEDURE NO. 1009 PAGE NO. 7 l APPENDIX I WYLE LABORATORIES' CERTIFICATION TEST REPORT SAMPLE WYLE LABORATORIES 50av 1054 7 % oct 79 s..new.no r 3coav
Page No. 75 g Tuo.- Revvi . Nu. C7-0 CERTIFICATION TEST REPORT WYLE LABORATORIES SCIENTIFIC sEAVICES ANO S'fSTEMS GROUP MUNTSVILLa. ALA6AMA I l REPORT NO. ?4976-1 ABC Corporation 'NYt.E JC8 NO. 14976 P.O. Box 2222 L Anytown, USA 22222 ! CUSTOMER P. O. No. 21192-22 MANUFACTURER Cresser Industries CONTRACT N/A CATE February 14, 1980 5 PAGE REPORT L SPECIMEN Main Steam Safety Valve 2. SET PRES 3URE: 1212 +/- 1.
- 3. PART NO. 3777 1 SERIAL NO. BK6528
- 5.
SUMMARY
On February 14, 1980, Valve Serial ' lumber BK6528 was calibrated fcr set crassure and tested for leakage with steam as the test medium. 7estina was in accordance wi-h Wyle Laboratories Test Precedure 1007, Revision S, dated January 12, 1980. The folicwing table lists four certification test runs and the leakage test results. Set Pressure Tes t Run Set P essure lift Valve Sc'ing Valve md o Stean Temo (OF )y Temo (OFI No. (osial (In) Temo ("F1 1217 0.106 137 360 559 2 3 1218 0.106 136 363 555 a 1215 0.1C6 138 366 561 5 1218 0.106 138 369 556 Leakace ~est Stean Pressure Ocs Tes: losic' Leakace 1090 Zero ama. .aes as=e m aaoaa ar e.megyse .cw .q u asenow.or ymn a w ,cusag nosca ,y
- or=esa,e== caraeans ,esaag aura **w.s eas==sr=3 ve .y=cus aswees ev er=e eveen COUNTY OF MActSCN I
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PAGE NO. 78 TEST REPORT NO. 45097-0 i (1 3 9 h i THIR PAGE INTENTIONALLY LEFT BLANK , 1 1 J 1 WYLE LABORATORIES Huntsvella Fv.tIItv
1 Test Report No. 45097-0 TEST FROCEDURE NO. 1009
! PAGE NO. II i
i 1 APPENDIX II 2 OPTIONAL TESTS WYLE LABORATORIES # # "" *I I9 Huntsville Facitety
Page Ro. 80 Test Report No. 45097-0 TEST PACCECURE RO. 1009 PAGE NO. I2 s 1.0 SCOPE 1.1 This addendum to Wyle Laboratories Test Procedure 1009 is to define testing to be acccmolished on pressurizer safety valves by direction of the custcmer.
2.0 REFERENCES
2.1 Dresser Engineering Instruction PT-66, steam testing and valve setting instructions for 31700 pressurizer safety valves. 2.2 Dresser Engineering Instruction PT 43, Pneumatic Backpressure Testing of 31700 Pressurizer Safety Valves. 3.0 ELEVATED TEMPERATURE TEST 3.1 The elevated temperature test shall be performed as stated in the preceding basic test procedure, except as stated in the following paragraphs. 3.2 Install themoccuples as shown on Figure 2-1. 3.3 Wrap valve in insulation blankets. A steam coil may be used to accelerate heat-up. The in-service temperatures to be approximated during this test are: 0 T2 - 465-485 F T3 - 200-225 F T4 - 175-185 F 3.4 Temperatures shall be considered stabilized when no changes greater than 100 F occur in a 1 hour interval. i 3.5 Perform leak test at 90% of set pressure as described in the basic procadure.
- 3.6 Verify set pressure a , described in the basic procedure.
l 3.7 Perfom leak test at 90% of set pressure as described in the basic procedure. WYLE LABORATORIES 50RM 10347 *ev oct 79 3.i ,.. iia r ioni,
Page No. 81 Test Report No. 45097-0 TEST PROCEDURE NO. _1009 PAGE NO. I3 4.0 BACKPkFSSURE TEST 4.1 Before set pressure testing, a backpressure test shall beperformed. 4.2 Apply 800 psig (Dresser Series 31700) or 500 psig (Crosby, Style HB) of gaseous nitrogen to the valve discharge port. Maintain pressure for 30 minutes minimum. 4.3 Install a pipe plug very loosely in the bonnet vent. 4.4 Brush or spray " Leak Test Solution" (soap solution) over the valve external surface, the valve body to bonnet interface, and the bonnet vent plug. 4.5 , " Bubble" evidence of leakage through the body wall shall be cause for rejection. Bubbling around ring pins, body drain, or bonnet body joint, shall require further tightening or gasket replacement, i or acceptance of leakage as determined by the cognizant Wyle engineer. l 1 4.6 Any bubbles around the bonnet vent plug is indicative of a leaking bellows or other leakage into the bonnet chamber. Leakage is not acceptable. 4.7 Bleed off test pressure. 4.8 Rinse the external surfaces with demineralized water. l WYLE LABORATORIES som 1054 7 % st ti Hunf $welle F 4ethty
iest Procedure flo. 1000 Page flo.14 Page No. 82 Test Report No. 45097-0 D ring Testing (Band)
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APPENDIX C l l I i Excerpts Frcs
- Dresser Pressurizer Safety Valve Inspection, Refurbishment and Test Report For Electric Power Research Institute Palo Alto, California a
1 i i I i 1 J
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For Crystal River Nuclear Power Plant Crystal River, Florida 4 t I i 1
PAGE NO. i TEST REPORT NO. 45097-0
- TABLE OF CONTENTS Page Number 1.0 PURPOSE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2.0 REFERENCES
1 3.0 TEST SPECIMEN DESCRIPTION . . . . . . . . . . . . . . . . . . . . . 1 4.0 TEST CONDITIONS, SYSTEMS AND EQUIPMENT .............. 1 5.0 TEST PROGRAM
SUMMARY
. . . . . . . ................ 1 5.1 Re cei vi ng I n fo rma ti o n . . . . . . . . . . . . . . . . . . . . . . . 2 5.2 Testing and Refurbishment, Valve FPC Tag No. RCV-9 ........ 2 5.3 Testing and Refurbishment, Valve FPC Tag No. RCV-8 ........ 2 5.4 Shipment Information . . . . . . . ................ 3 TABLE I RECEIVING INSPECTION
SUMMARY
. . ................ 4 TABLE II CHRONOLOGICAL
SUMMARY
OF TESTING ................ 5 APPENDIX I AS RECEIVED TEST REPORT FOR VALVE FPC 1AG NO. RCV-9 . . . . . 7 APPENDIX II AS RECEIVED TEST REPORT FOR VALVE FPC TAG NO. RCV-8 . . . . . 13 APPENDIX III CERTIFICATION TEST REPORT, VALVE FPC TAG NO. RCV-9 ..... 19 APPENDIX IV CERTIFICATION TEST REPORT, VALVE FPC TAG NO. RCV-8 ..... 25 APPENDIX V PHOTOGRAPHS . . . . . . . . . . . . . . . . . . . . . . . . . 31 APPENDIX VI DYE PENETRANT INSPECTION REPORTS .............. 61 APPENDIX VII INSTRUMENTATION EQUIPMENT SHEET . . . . . . . . . . . . . . . 65 APPENDIX VIII WLTP 1009 ......................... 67 EXHIBIT I E.P.R.I. INSPECTION AND TEST GUIDELINES . . . . . . . . . . . 83 EXHIBIT II DRESSER INDUSTRIES REPLACEMENT PART CERTIFICATIONS ..... 89 EXHIBIT III DRESSER INDUSTRIES ENGINEERING AND FIELD SERVICE REPORTS .. 105 WYLE LABORATORIES Hunt 0ve FPMTrb
TEST REPORT
,c,cer ,0. mu-0 E MMO SCIENTIFIC SERVICES AND SYSTEMS GROUP ouR JOB No. 45097 HUNTsVILLE. ALA8AMA YOuR P. O. NO. Il4I4 I
Electric Power Research Institute CONTRACT N/A P. O. Box 10412 Palo Alto, CA 94303 PAGE 1 of Il9 PAGE REPORT L J cATE August 5, 1980 1.0 PURPOSE The purpose of this report is to present the test procedures used and the test results obtained during the performance of an inspection, refurbishment and certification test program. The program was conducted to determine the conformance of two pressurizer safety valves to the require-ments specified in References 2.1 through 2.3.
2.0 REFERENCES
2.1 i Wyle Laboratories Test Procedure Number 1009, dated February 22, 1980, including Appendix II, Elevated Temperature Test. 2.2 EPRI Purchase Order Number 11414 2.3 EPRI Inspection and Test Guidelines (Refer to Exhibit I). 3.0 TEST SPECIMEN DESCRIPTION Two Dresser Pressurizer Safety Valves, Type Number 31739A, Serial Numbers BL8899, and BL8900, F.P.C. Tag Numbers RCV-9 and RCV-8, respectively. 4.0 TEST CONDITIONS, SYSTEMS AND EQUIPMENT Refer to References 2.1 through 2.3 and Appendix II. 5.0 TEST PROGRAM
SUMMARY
Two pressurizer safety valves, FPC Tag Numbers RCV-8 and RCV-9 were tested in the "as received" condition, inspected, decontaminated, refurbished and certification tested. The test program was conducted in accordance with the EPRI Inspection and Test Guidelines presented in Exhibit I. So^u'n'TEofu'I$i$n }* E U7 YE*0.E71TTCOO Larry Frazier TEST BY , Steam Services JWn amA
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Page No. 68 r flo. 4 Test RETE $T PRd097-0CEDURE TEST PROCEDURE NO. 1009 WYLE LABORAMS sC:ENiirC ssPviCEs amo svsites ca0cP DATE: February 22, 1980 P o een 'oos.auNrsvut AueAvA ssee7 fWx +010172122:s e ffLEPwCNE 12086 837411 TEST PROCEDURE FOR STEAM SET PRESSURE AND LEAKAGE TESTIftG OF SPRING-0PERATED SAFETY VALVES APMOVED BY: APPROVED BY 7 FOR: PROJECT MANAGER: i / ' b-APET.OVED BY: APPROVED BY p FOR: QUALITY ENGINEER: 7 AP7 ROVED BY: PREPARED BY FOR: j PROJECT ENGINEER-m REVISIONS ,o . , i . , R ., .,,. REv. NO. l OATE l PAGES AFFECTED l BY A P 8' L. DESCRIPTION OF CHANGES I I I I I I I i l I ! I I I COPvelGHT 8v wvl! .A80eATOstES. TME CIGHT TO REP 9000CE. COPY, EXHIBIT. OR OTHERwl5E UTILIZE ANY OF THE M ATERIAL CONTAINED HEREIN WITM0tlT THE E1 PRE 55 8'10R PE9MI55tCN OF WYLE LABORATORIES IS PROHIBITED. THE ACCEPTANCE CF A PURCHASE ORDER IN CONNECTION WITH YME Vaf t#!st CONTAINED MERES 4 SM ALL at EOU: VALENT TO EXPRESS PRIOR PERMIS5iCN.
PAGE NO. 2 TEST REPORT NO. 45097-0 s 5.0 TEST PROGRAM
SUMMARY
(Continued) Testing was performed per Wyle Laboratories Test Procedure 1009, dated February 22, 1980, which is presented in Appendix VIII. The inspections were performed by personnel from EPRI, Dresser, Babcock and Wilcox, M.P.R. and Wyle. Decontamination and cleaning were performed by Wyle personnel. The valves were refurbished by Dresser personnel with Wyle assistance. Detailed discussions of the work performed during the test program are presented in Paragraphs 5.1 through 5.4 and Exhibit III. Representative photographs of the inspection and refurbishment of the valves are pre-sented in Appendix V. 5.1 Receiving Information The valves were received from Crystal River Nuclear Power Plant on June 24, 1980. An initial receiving inspection was performed and the results are detailed in Table I and Appendix V. 5.2 Testing and Refurbishment, Valve FPC Tag No. RCV-9 The valve, FPC Tag No. RCV-9,was initially tested in the "as received" condition for set pressure and leakage. The initial set pressure of the valve was 2472 psig,which was 3 psi below the nameplate set pressure range of 2475 to 2525 psig. The valve exhibited.zero leakage both before and after test. Complete test data is presented in Appendix I. Subsequent to its initial test, the valve was disassembled, decontaminated. inspected and refurbished. There were no new parts,other than gaskets, required for refurbishment. The valve was successfully retested after refurbishment. A Certification Test Report is presented in Appendix III. 5.3 Testing and Refurbishment, Valve FPC Tag No. RCV-8 The valve, FPC Tag No. RCV-8,was disassembled for inspection with the spring compression retained. The major anomalies noted during the inspec-tion were: 1) The nozzle seat and disc exhibited gross steam cutting,
- 2) The anti-rotating pin had been broken from the top spring washer and was laying on top of the support plate, 3) The disc holder exhibited considerable surface wear and numerous linear indications when examined by dye penetrant, 4) There was no movement or " rock" in the disc. After inspection, the valve was reassembled without repair for an "as receiven" set pressure and leakage test.
VlYLE LABORATORIES Huntsville Facuity
PAGE NO. 3 TEST REPORT NO. 45097-0 5.3 Testing and Refurbishment, Valve FpC Tag No. RCV-8 (Continued) The valve's initial set pressure was 2392 psig, which was approximately 100 psi low. The valve exhibited a significant leakage rate throughout the test of approximately 1.1 gpm (550 lb/hr), but still actuated and performed well. Complete test data is presented in Appendix II. Subsequent to testing, the valve was disassembled, decontaminated, inspected and refurbished. Replacement parts required for refurbishment were a new disc, disc holder, bellows, and gaskets. Refer to Exhibit II for part certification. The valve was successfully retested after refurbishment. A certification Test Report is presented in Appendix IV. 5.4 Sgment Infonnation Both valves were shipped to the Crystal River Nuclear Power Plant on July 29, 1980. WYLE LABORATORIES Huntsvu6e FacHity
PAGE NO. 4 TEST REPORT NO. 45097-0 TABLE I RECEIVING INSPECTION
SUMMARY
Valve Remarks g, RCV-8 1. Valve cap not wired and sealed.
- 2. Some small deposits of boric acid crystal on valve inlet flange.
- 3. Adjusting ring pins not sealed and not properly wired.
- 4. One non-standard nut on valve cap.
- 5. No cap gasket.
- 6. One non-standard nut on valve bonnet flange.
- 7. Ring pins appear to have been leaking.
RCV-9 1. Valve cap not wired or sealed.
- 2. Boric acid crystals present on inlet flange and internals of valve (primarily in outlet side).
VWYLE LABORATORIEU HuntsvHis Fac hty
PAGE NO. 5 TEST REPORT NO. 45097-0 TABLE II CHRONOLOGICAL
SUMMARY
OF TESTING Valve Tag Description Date No. of Test Pass / Fail Remarks 7-16-80 RCV-9 As Received N/A Information Test Only 7-16-80 RCV-8 As Received N/A T.nformation Test Only 7-18-80 RCV-9 Certification P None Test 7-18-80 RCV-8 Certification P None Test 4 WYLE LABORATORIES Huntsville Facility L..
PAGE NO. 6 TEST REPORT NO. 45097-0 i THIS PAGE INTENTIONAL.LY LEFT BLANK l 4 W'YLE LABORATORIES HuntsvHis Facility
PAGE NO. 7 TEST REPORT NO. 45097-0 APPENDIX I AS RECEIVED TEST REPORT FOR VALVE FPC TAG NO. RCV-9 / WYLE LABORATORIES Huntsville Facility
Page No. 8 Test Report No. 45097-0 CERTIFICATION TEST REPORT WYLE LABORATORIES SCIENTIFIC SERVICES ANO SYSTEMS GROUP HUNTSVILLE ALABAMA f- l REPORT NO. 45097-1 Electric Power Research Institute 45097-0 P. O. Box 10412 wYLE JOB NO. Palo Alto, CA 94303 N/A MANUFACTURER Dresser Industries CONTRACT July 16, 1980 4 PAGE REPORT DATE
- 2. SET PRESSURE: 2500 +/- 1% psig
- 1. SPECIMEN Pressurizer Safety Valve
- 3. PART NO. 31739A 4 SERIAL NO. BL8399 FPC Tag No. RCV-9
- 5.
SUMMARY
Valve Serial Number BL8899 was tested in the "As Received" condition fer set pressure and leakage with steam as the test medium per Wyle Laboratories Test Procedure 1009, dated February 22, 1980, with Elevated Temperature Test per Appendix II. The test results are presented below: Set Pressure Test Set Steam Inlet Flange Lower Bonnet Upper Bonnet Tetop Temp Temp Run Prassure Disc Temp No. (psig) Lift (OF) (OF) (OF) (oF) 476 244 185 1 2472 0.031 651 0.032 651 480 245 184 2 2456 0.031 651 477 246 182 3 2466 Seat Leakage Test Pretest leakage at 2250 psig = zero Post-Test leakage at 2220 psig = zero C."*,.7M6".' '.~ '" '.""" ""' 7 "*'.*.*' ~ STATE OF ALABAMA { " COUNTY OF MADISON f Larry E. Frazier , ,,,,, ,,,, ,,,,,,,
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PAGE NO.12 TEST REPORT NO. 45097-0 THIS PAGE INTENTIONALLY LEFT BLANK WYLE I.ABORATORIES Huntsville FaCality
PAGE NO. l3 TEST REPORT NO. 45097-0 APPENDIX II AS RECEIVED TEST REPORT 4 FOR VALVE FPC TAG NO. RCV-8 6 WYLE LABORATORIES Huntsystle Facati'y
Page No. 14 Test Report No. 45097-0 CERTIFICATION TEST REPORT WY1.E LABORATORIES SCIENTIFIC SERVICES AND SYSTEMS GROUP HUNTSVILLE. ALA8AMA f- 1 REPORT NO. 45097-2 Electric Pcwer Research Institute P. O. Box 10412 WYLE JOB NO. 45097-0 Palo Alto, CA 94303 MANUFACTURER Dresser Industries CONTRACT N/A DATE July 16. 1980 4 PAGE REPORT
- 1. SPECIMEN Pressurizer Safety Valve 2. SET PRESSURE:2500 + 1% psig
- 3. PART NO. 31739A 4. SERIAL NO. BL8900 FPC Tag. No. RCV-8
- 5.
SUMMARY
Valve Serial Number BL8900 was tested for ';et pressure and leakage with steam as the test medium per Wyle Laboratories Test Procedure 1009, dated February 22. 1980, with Elevated Temperature Test per Appendix II. Prior to test, the valve wa disassembled for inspection with spring compression retained and reassembled for test The test results are presented below: Set Pressure Test Set Steam Inlet Flange Lower Bonnet Upper Bonnet Run Pressure Disc Temp Temp Temp Temp No. (psig) Lift (OF) (OF) (OF) (OF) 1 2392 0.037 645 480 212 195 2 2388 0.038 645 481 21 3 196 3 2388 0.038 645 481 213 195 Seat Leakage Test Pretest Leakage at 2160 psig = gross Post-Test Leakage at 2150 psig = gross By measuring condensate collected during test, valve leakage was 1.0-1.2 GPM or approximately 550 lb/hr.
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- STA TE OF ALABAMA }
COUNTY OF MADISON 3 Larry E. Frazier . .... . .,. Steam services
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