ML15244A660
| ML15244A660 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/01/1991 |
| From: | Wiens L Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9108050288 | |
| Download: ML15244A660 (21) | |
Text
August 1, 1991 Docket Nos. 50-269, 50-270 and 50-287 LICENSEE: Duke Power Company FACILITY: Oconee Nuclear Station, Units 1, 2, and 3
SUBJECT:
SUMMARY
OF JULY 2, 1991, MEETING WITH DUKE POWER CONCERNING PRESSURIZER SAFETY VALVES LIFT SETPOINTS On July 2, 1991, the NRC staff (NRR, AEOD and Region II) met with representa tives of Duke Power Company (DPC) at the DPC corporate offices in Charlotte, North Carolina, to discuss a recent analysis of pressurizer safety valve (PSV) test results. The PSVs for both the Oconee and Catawba plants are sent to an outside laboratory for lift pressure setpoint testing. Many of the as-found setpoints have been outside +/-1% of the nominal lift pressure for these valves. As a result, DPC initiated an investigation to determine the cause for the changes in setpoint.
There were a number of potential causes for the setpoint changes jointly identified by DPC, the valve manufacturer, and the testing laboratory. The primary cause, however, appeared to be related to the procedure used for repairing minor seat leakage, which was done following setpoint testing.
Corrective actions have been implemented to correct the deficiencies identified in this process. Evaluation of this action and of other corrective actions will be evaluated by DPC after the next set of safety valves are tested. DPC agreed to submit a voluntary LER to report the results of the test data and their investigation. The NRC staff stated that since it appeared the informa tion would be of generic interest, the NRC may issue an Information Notice (IN) to inform other licensees of the lessons learned from DPC's investigation.
Meeting attendees are listed in Enclosure 1. Handouts distributed by DPC during the meeting are provided in Enclosure 2.
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Leonard A. Wiens, Project Manager Project Directorate 11-3 910B0502B 910801 Division of Reactor Projects - I/II PDR ADOCw oooo269 Office of Nuclear Reactor Regulation P
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Enclosures:
As stated cc w/enclosures:
See next page OFC
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ft REG(4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 August 1, 1991 Docket Nos. 50-269, 50-270 and 50-287 LICENSEE: Duke Power Company FACILITY: Oconee Nuclear Station, Units 1, 2, and 3
SUBJECT:
SUMMARY
OF JULY 2, 1991, MEETING WITH DUKE POWER CONCERNING PRESSURIZER SAFETY VALVES LIFT SETPOINTS On July 2, 1991, the NRC staff (NRR, AEOD and Region II) met with representa tives of Duke Power Company (DPC) at the DPC corporate offices in Charlotte, North Carolina, to discuss a recent analysis of pressurizer safety valve (PSV) test results. The PSVs for both the Oconee and Catawba plants are sent to an outside laboratory for lift pressure setpoint testing. Many of the as-found setpoints have been outside +/-1% of the nominal lift pressure for these valves. As a result, DPC initiated an investigation to determine the cause for the changes in setpoint.
There were a number of potential causes for the setpoint changes jointly identified by DPC, the valve manufacturer, and the testing laboratory. The primary cause, however, appeared to be related to the procedure used for repairing minor seat leakage, which was done following setpoint testing.
Corrective actions have been implemented to correct the deficiencies identified in this process. Evaluation of this action and of other corrective actions will be evaluated by DPC after the next set of safety valves are tested. DPC agreed to submit a voluntary LER to report the results of the test data and their investigation. The NRC staff stated that since it appeared the informa tion would be of generic interest, the NRC may issue an Information Notice (IN) to inform other licensees of the lessons learned from DPC's investigation.
Meeting attendees are listed in Enclosure 1. Handouts distributed by DPC during the meeting are provided in Enclosure 2.
Leonard A. Wiens, Project Manager Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/enclosures:
See next page
Oconee Nuclear Station Units Nos. 1, 2 and 3 Duke Power Company cc:
Mr. A.V. Carr, Esq.
Mr. Stephen Benesole Duke Power Company Duke Power Company 422 South Church Street Post Office Box 1007 Charlotte, North Carolina 28242-0001 Charlotte, North Carolina 28201-1007 J. Michael McGarry, III, Esq.
Winston and Strawn Mr. Alan R. Herdt, Chief 1400 L Street, N.W.
Project Branch #3 Washington, D.C. 20005 U.S. Nuclear Regulatory Commission 101 Marietta Street, NW, Suite 2900 Mr. Robert B. Borsum Atlanta, Georgia 30323 Babcock & Wilcox Nuclear Power Division Ms. Karen E. Long Suite 525 Assistant Attorney General 1700 Rockville Pike N. C. Department of Justice Rockville, Maryland 20852 P.O. Box 629 Raleigh, North Carolina 27602 Manager, LIS NUS Corporation Mr. R.L. Gill, Jr.
2650 McCormick Drive, 3 Floor Nuclear Production Department Clearwater, Florida 34619-1035 Duke Power Company P.O. Box 1007 Senior Resident Inspector Charlotte, North Carolina 28201-1007 U.S. Nuclear Regulatory Commission Route 2, Box 610 Mr. M. S. Tuckman Seneca, South Carolina 29678 Vice President Nuclear Operations Regional Administrator, Region II Duke Power Company U.S. Nuclear Regulatory Commission P. 0. Box 1007 101 Marietta Street, N.W., Suite 2900 Charlotte, NC 28201-1007 Atlanta, Georgia 30323 Mr. Heyward G. Shealy, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 Office of Intergovernmental Relations 116 West Jones Street Raleigh, North Carolina 27603 County Supervisor of Oconee County Walhalla, South Carolina 29621
ENCLOSURE 1 SAFETY VALVE MEETING ATTENDEES NAMEORGANIZATION Len Wiens NRC/NRR G. A. Belisle NRC/Rh W. T. Orders NRC/SRI/CAT M. H. Hazeltine NPD/GO Gary Hammer NRC/NRR/DET/EMEB Gregg Swindlehurst Duke/Design Engineering Mary S. Wegner NRC/AEOD Ruth Oakley Duke/Oconee Engineering Edward B. Kulesa Duke/Catawba Engineering Stephen D. Hart Duke/Nuci. Maint Stephen G. Benesole Duke/Reg. Compliance Edward H. Girard NRC/Rh Francis Jape NRC/RII/DRS Charlie Boyd Duke/Oconee Engineering Todd Cooper NRC/McGuire RI Terry Edward Duke/Catawba Engineering Luellen Jones DPC/Conpliance Lee Hartzell DPC/Compliance Kelly Bishop DPC/MES/CNS
ENCLOSURE 2 NRC Meeting on Pressurizer Code Safety Valves Introduction Steve Benesole Historical Background Ed Kulesa Discussion of Findings, Evaluations, Steve Hart and Corrective Actions Evaluation of Potential Consequences Greg Swindlehurst Summary Steve Benesole
DUKE EXPERIENCE WITH PRESSURIZER SAFETY VALVES
- OCONEE AND CATAWBA HAVE DRESSER PRESSURIZER SAFETY VALVES (PSV'S), McGUIRE HAS CROSBY.
- THE CATAWBA AND OCONEE PSV'S ARE TESTED WITH STEAM.
DUKE DOES NOT HAVE STEAM TESTING CAPABILITY; WYLE LABS WAS CHOSEN TO DO THIS TESTING FOR DUKE. McGUIRE HAS LOOP SEALS AND THEIR PSV'S ARE TESTED ON SITE WITH WATER/N 2.
- DRESSER HAS SUPPLIED PSV'S TO APPROXIMATELY 1/3 OF THE NUCLEAR UTILITIES IN THE UNITED STATES.
- WYLE LABS TESTS PSV'S FOR MOST OF THESE UTILITIES.
- EARLY OPERATING EXPERIENCE AT OCONEE SHOWED THAT PSV SEAT LEAKAGE WOULD RESULT IN PLANT SHUTDOWNS.
- DUKE (AS WELL AS MANY OTHER UTILITIES) STRUCTURED PSV TEST PROGRAMS TO GUARD AGAINST SEAT LEAKAGE.
- HENCE, THE DEVELOPMENT OF THE "JACK AND LAP" (J&L)
PROCESS; BOTH WYLE AND DRESSER APPROVED THE J&L PROCEDURE AS AN ACCEPTABLE REPAIR METHOD FOR PSV'S.
- J&L CONSISTENTLY PROVIDED "BUBBLE TIGHT" PSV'S.
- NUCLEAR INDUSTRY RECOGNIZED THERE WAS DIFFICULTLY IN ACHIEVING THE DESIRED +/- 1% REPEATABILITY FOR PSV'S.
60% OF CATAWBA'S AS-FOUND TESTS EXCEEDED +/- 1%.
70% OF OCONEE'S AS-FOUND TESTS EXCEEDED +/-1%.
42 % OF INDUSTRY AS-FOUND TESTS EXCEEDED + /- 1 %
(FROM NPRDS SEARCH); SOME UTILITIES AS HIGH AS 75%.
- A MULTITUDE OF INFORMATION NOTICES AND OEP ITEMS WERE ORIGINATED CONCERNING PSV SETPOINT DRIFT.
- MOST OF THESE NOTIFICATIONS IMPLIED ERROR WAS DUE TO DIFFERENCES BETWEEN INSTALLED AND TESTED CONDITIONS; EG. TEMPERATURE EFFECTS, FLANGE LOADING, ETC.
-INDUSTRY CONCLUDED +/- 1% WAS SOMEWHAT UNREALISTIC; ASSUMED +/- 3% WAS A MORE REASONABLE VALUE.
- MOST UTILITIES WERE ABLE TO PERFORM SAFETY ANALYSES WITH +- 3% WITHOUT EXCEEDING OVERPRESSURE LIMITS.
- INCIDENTS OF DEVIATIONS BEYOND +/- 3 % SUGGESTED OTHER FACTORS COULD BE CONTRIBUTING TO SETPOINT VARIANCES.
- 2 OF 3 VALVES REMOVED FROM CATAWBA WERE -
6% HIGH.
- OCONEE ALSO HAD ONE VALVE AT 7.4% HIGH.
- INITIAL INVESTIGATION TARGETED J&L PROCEDURE.
- DRESSER NOR WYLE CONTROLLED THE BEFORE AND AFTER GAP MEASUREMENTS; DRESSER TIGHTENED GAP TOLERANCES TO +/- 0.0035" IN RESPONSE TO DUKE'S CONCERNS.
- FIRST PSV TESTED TO NEW J&L PROCEDURE PRODUCED A DEVIATION OF 2.5% HIGH.
- CATAWBA RETURNED A PSV WHICH HAD NOT BEEN INSTALLED, AND IT TESTED AT 6% HIGH.
- JOINT MEETING BETWEEN DUKE'S ENGINEERING STAFF, DRESSER VALVE ENGINEERS, AND WYLE LABS SHOWED THAT THE EXISTING TEST PRACTICES COULD BE INTRODUCING ERROR.
- PRIMARY EMPHASIS OF TEST PROGRAM HAD BEEN ON LEAK TIGHTNESS; SETPOINT REPEATABILITY WAS SECONDARY.
- DISCUSSIONS PRODUCED NINE (9) POTENTIAL CAUSES FOR SETPOINT DRIFT; TOP FOUR (4) THOUGHT TO BE PROBABLE.
July 2, 1991 DUKE POWER COMPANY POTENTIAL ROOT CAUSES OF PRESSURIZER SAFETY VALVE SETPOINT DRIFT
- 1) Leaking valve during setpoint test If valves are leaking during the steam setpoint verification test, the valve's huddle chamber could become pressurized, effectively increasing the valve seat area. This would cause a lower apparent setpoint and the compression screw would be adjusted to further compress the spring.
This would give an artificially high setpoint on the valve if the leak was later repaired as is done with the jack and lap process. If further testing on steam is not performed then the repaired valve will have a higher setpoint than desired.
- 2) The jack and lap process The jack and lap process is a partial disassembly of the valve while maintaining spring compression.
This process is used to polish valve seats, often in preparation for a final gaseous nitrogen leakage test.
This disassembly process obviously can introduce some small error into the as-tested setpoint since it is very difficult to reassemble the valve exactly as it was found.
- 3) Setpoint trending control When performing setpoint verification testing, particularly after valve rebuild, performance stabilization is necessary to validate a true setpoint. If setpoints are trending in one direction some stabilization or "turn around" is desired before a test is considered valid. Test controls should be tight enough to insure that trending will not continue and that the valve's true setpoint has been achieved.
- 4) Ring adjustments after a setpoint test If a valve undergoes ring adjustments, particularly on the lower ring, its transition from simmer to pop is changed and its performance under leaking conditions also is changed.
- 5) Temperature Effects Wyle heats the valve body and bonnet by using the test media steam at saturation conditions and electric heaters strapped to the outside of the valve. The temperature is 'controlled by monitoring three thermocouples strapped to the body & bonnet flanges and, therefore, is reasonably re peatable within the temperature guidelines of their procedure. Valve heat up (or soak time) and the number of actuations could effect the actual
internal temperatures, however, and could be different from the external monitored temperatures.
- 6) Spring Performance Springs may be effected by temperature or aging or complete relaxation during rebuild or varying friction factors between the spring and spring washer and spindle. This could result in erratic valve setpoint changes.
- 7) Seat Adhesion Effects A perfectly lapped seat or seat corrosion could contribute to a high as-found setpoint.
- 8) Transportation/Handling Effects Excessive Valve shock or acceleration could impact valve performance leading to erratic as-found tests.
- 9) Test Process Effects Steam quality, pressurization rates, valve installation (torquing)effects, equipment calibration errors, and personnel errors could impact the test results at Wyle.
DUKE POWER COMPANY ACTION TAKEN TO RESOLVE PRESSURIZER SAFETY VALVE SETPOINT DRIFT Potential Root Cause Action
- 1) Leaking valve during A) Wyle procedure enhancements-below setpoint test are now in process.
- i. Valves leaking during or after the steam setpoint verification test must be repaired and retested on steam.
ii. Stricter leakage acceptance criteria will allow no fogging at 93% of set pressure.
- 2) The 'jack & lap' process A) Dresser procedure enhancement requires the body/bonnet gap to be re-established within.0035 inches of the as-found position. This item is now complete.
B) Wyle procedure enhancement will require a steam setpoint verification test after the 'jack'and lap' process.
- 3) Setpoint trending control A) Wyle procedure enhancement will require all three setpoint verification pops to be within 10 psi if trending occurs with no turnaround.
- 4) Ring adjustments after A) Wyle procedure enhancement will setpoint test not allow ring adjustments after the final setpoint verification test.
B) Dresser has already implemented changes to their procedures to better control the method of measuring ring settings. The ability to maintain consistent ring settings will remove this as a source of uncertainty in our investigative efforts.
- 5) Temperature Effects A) Wylie procedure enhancements listed below are now in process:
- i.
Use two thermocouples on each valve/bonnet flange location to ensure even heating and, therefore, alignment.
ii. Increase the minimum hold time between actuations to allow improved temperature stabilization.
B) Dresser will further investigate the temperature effect on spring performance.
C) Duke has completed a preliminary investigation of spring temperature versus flange temperature and found that the variances were minimal.
The effect on set point will be better understood when the results in B) above are available.
- 6) Spring Performance A) Dresser will conduct a spring performance test on a Catawba pressurizer valve spring.
B) Dresser will investigate the need for establishing stricter controls on the spring/washer sliding surfaces.
C) Duke will require a complete disassembly, inspection, and cleaning every RFO to control the possible effects of corrosion on valve performance.
D) Duke has confirmed with the spring manufacturer that these tungsten steel springs are rated for 800 F and are not prone to take a set or relax at our service conditions.
- 7) Seat Adhesion Effects A) Dresser controls this.problem through their material selection (347SS nozzle and Inconel 750 disc) and the seat lapping procedure which specifies use of a cast iron lap and 800 grit polishing compound.
- 8) Transportation/Handling Effects A) Dresser sited several cases in which valves had been dropped or were returned after many years and still had as-found setpoints within +1%.
B) Wyle reiterated Dresser's position so no further action is planned.
- 9) Test Process Effects A) Wylie procedure enhancements below are currently in process:
- i. A torquing guideline will be used when installing the valve on the test header to ensure even loading and alignment.
ii. The pressurization rate will be standardized to reduce its effect on setpoint variations.
- 10)
Other A)
Duke plans to investigate safety valve setpoint repeatability under the new procedure enhancements by retesting two valves two months after the initial test.
OCONEE NUCLEAR STATION INCREASED PRESSURIZER SAFETY VALVE DRIFT SAFETY EVALUATION
- PSVs PREVENT RCS OVERPRESSURIZATION DURING TRANSIENTS AND ACCIDENTS (FSAR CHAPTER 15) 2750 PSIG (CONDITIONS I-III) 3000 PSIG (CONDITION IV)
- LIMITING TRANSIENTS INVOLVE A MISMATCH BETWEEN PRIMARY HEAT SOURCE AND SECONDARY HEAT SINK, WHICH RESULTS IN A RAPID PRESSURIZER INSURGE
- UNCONTROLLED ROD WITHDRAWAL
- LOSS OF MAIN FEEDWATER ROD EJECTION
- THE IMPACT OF AN INCREASED PSV LIFT SETPOINT WOULD BE TO DECREASE THE MARGIN BETWEEN THE PEAK TRANSIENT RCS PRESSURE AND THE LIMIT
- THE EVALUATION PERFORMED INVOLVED REANALYSIS OF THE LIMITING ANALYSES TO DETERMINE THE IMPACT
- +6% PSV SETPOJNT DRIFT, +3% ACCUMULATION
- -4% PSV SETPOINT DRIFT
IMPACT OF +6%
DRIFT
- UNCONTROLLED ROD WITHDRAWAL ANALYSIS RESULTS FROM ZERO POWER FROM 25% POWER
- CONSERVATIVE INITIAL AND BOUNDARY CONDITIONS
- REACTOR PHYSICS PARAMETERS VALID FOR THE CURRENT FUEL CYCLES
- PEAK PRESSURE = 2735 PSIG (LIMIT = 2750 PSIG)
- LOSS OF MAIN FEEDWATER FROM 102% POWER CONSERVATIVE INITIAL AND BOUNDARY CONDITIONS
- PEAK PRESSURE = 2725 PSIG (LIMIT = 2750 PSIG)
- ROD EJECTION - SINCE THE ROD EJECTION IS A CONDITION IV EVENT, THE ACCEPTANCE CRITERION IS 3000 PSIG. SINCE THE PSVs WOULD BE FULL OPEN AT 2725 PSIG, IT CAN BE CONCLUDED THAT THERE WOULD BE LITTLE IMPACT ON THE PRESSURE RESPONSE, a FEED & BLEED CAPABILITY
- NORMALLY CREDIT PZR PORV FOR BLEED PATH
- W/O PORV, ONE PSV AT 3% DRIFT IS ACCEPTABLE BOTH PSVs AT 6% DRIFT, NO PORV, AND NO EFW IS OF VERY LOW PROBABILITY --
NO IMPACT
- SIMILAR TO ROD EJECTION, THE PSVs WOULD GO FULL OPEN DURING THE TRANSIENT, THEREFORE THERE WOULD BE NO IMPACT DUE TO THE SETPOINT DRIFT
IMPACT OF -4% DRIFT
- THE CONCERN IS THE POTENTIAL FOR LIFTING THE PSV PRIOR TO TRIPPING ON HIGH RCS PRESSURE
- THE MAXIMUM RCS PRESSURE AT THE TIME OF REACTOR TRIP WOULD BE 2355 + 30 = 2385 PSIG.
- IN ADDITION TO THE 1.5 Psi MARGIN, A SIGNIFICANT PRESSURE DROP ACROSS THE PRESSURIZER SURGE LINE WOULD EXIST DURING PRESSURIZATION TRANSIENTS, THIS WOULD ACCOMODATE ANY OVERSHOOT IN PRESSURE DUE TO RPS INSTRUMENTATION DELAYS
- -4% DRIFT CAN BE TOLERATFD WITHOUT A SIGNIFICANT IMPACT ON SAFETY NOTE THE TRANSIENT ANALYSES USED TO EVALUATE THE +6/-4%
PSV DRIFT HAVE ONLY BEEN EVALUATED FOR THE CURRENT FUEL CYCLES,
CATAWBA NUCLEAR STATION INCREASED PRESSURIZER SAFETY VALVE DRIFT SAFETY EVALUATION
- 2750 PSIA (CONDITIONS I-III)
- 3000 PSIA (CONDITION IV)
- LIMITING TRANSIENTS INVOLVE A MISMATCH BETWEEN PRIMARY HEAT SOURCE AND SECONDARY HEAT SINK, WHICH RESULTS IN A RAPID PRESSURIZER INSURGE
- UNCONTROLLED ROD WITHDRAWAL
- ROD EJECTION
- THE IMPACT OF AN INCREASED PSV LIFT SETPOINT WOULD BE TO DECREASE THE MARGIN BETWEEN THE PEAK TRANSIENT RCS PRESSURE AND THE LIMIT
- C1C5 AND C2C4 HAVE BEEN EVALUATED TO ASSESS THEIR CURRENT CONDITIONS WITH RESPECT TO POSSIBLE PSV SETTINGS
C1C5 EVALUATION
- BASED ON THEAS-FOUND PSV SETPOINTS, TWO PSVs WITH +6%
DRIFT, AND ONE WITH NO DRIFT ARE ASSUMED, AL.ONG WITH 2% ACCUMULATION
- A CONTROL. ROD DROP TIME OF 2.2 SECONDS IS ASSUMED, THIS VAL.UE HAS BEEN SUBMITTED TO THE NRC AS A TECH SPEC CHANGE REQUEST
- CONSERVATIVE INITIAL AND BOUNDARY CONDITIONS
- REACTOR PHYSICS PARAMETERS BOUND CIC5
- PEAK PRESSURE = 2741 PSIA (TURBINE TRIP AND UCRW)
- ROD EJECTION - SINCE THE ROD EJECTION IS A CONDITION IV EVENT, THE ACCEPTANCE CRITERION IS 3000 Psi, SINCE THE PSVs WOULD BE FULL OPEN AT 2700 PSIA, IT CAN BE CONCLUDED THAT THERE WOULD BE LITTLE IMPACT ON THE PRFSSURE.RESPONSE (VALID FOR C2C4 ALSO).
- FEED & BLEED CAPABILITY - THIS SITUATION NORMALLY CREDITS PZR PORVs (3 OF THEM).
THERE IS NO IMPACT DUE TO PSV DRIFT IF THE PORVs ARE AVAILABLE FOR THIS BEYOND-DESIGN BASIS EVENT, THE SITUATION WITH NO PORVs IS LOW IN PROBABILITY (VAL.ID FOR C2C4 ALSO).
- ATWS - SIMILAR TO ROD EJECTION, THE PSVs WOULD GO FULL OPEN DURING THE TRANSIENT. THEREFORE THERE WOULD BE NO IMPACT DUE TO THE SETPOINT DRIFT (VALID FOR C2C4 ALSO),
C2C4 EVALUATION
- PSVs ARE ASSUMED TO ALL BE AT 6% DRIFT WITH 2% ACCUMULATION
- A CONTROL ROD DROP TIME OF 2.2 SECONDS IS ASSUMED.
- CONSERVATIVE INITIAL. AND BOUNDARY CONDITIONS
- REACTOR PHYSICS PARAMETERS BOUND C2C4 FOR THE CURRENT CORE BURNUP
- PEAK PRESSURE = 2741. PSIA CONCLUSIONS
- FOR CIC5.. THE IMPACT OF THE AS-FOUND SETPOINTS DID NOT PRESENT A SITUATION WHERE THE OVERPRESSURE LIMIT COULD HAVE BEEN VIOLATED IF THE LIMITING DESIGN BASIS TRANSIENT HAD OCCURRED.
- FOR C2C4, ASSUMING +6% DRIFT FOR ALL. PSVs, THE SITUATION DOES NOT PRESENT THE POTENTIAL FOR EXCEEDING THE OVER PRESSURE LIMIT FOR THE REMAINDER OF THE FUEL CYCLE,
NOZ-DISC HUDDLE CHAMBER SEAT INTERFACE SECONDARY ORIFICE LOWER ADJ.RING POINT DIAMETER SECTION A-A LIFT STOP DISC HOLDER UPPER ADJ.
RING VALVE OUTLET MIDDLE AOJ.RING A-A VALVE BODY BOWL LOWER AOJ.RING NOZZLE FIGURE 4 VALVE OPENING 21
44 6
413 2A 19 4139-415/
6-/
3-0 353A13l4 1
45127 322 23- -
21 220 4-3149A1
- 31739A-1 451/2 701/
A1155.7 1778.0 11
August 1, 1991 DISTRIBUTION:
Docket File NRC & Local PDRs FMiraglia, 12G18 JPartlow, 12G18 SVarga GLainas PDII-3 R/F Oconee R/F OGC, 15B18 EJordan, MNBB3701 ARCS (10), P-315 RLobel, 17G21 LWiens DMatthews LBerry GBelisle, RH WOrders, RH GHammer, 7E23 MWegner, MNBB9715 EGirard, RII Fape, RH TCooper, RH