ML20198E762
| ML20198E762 | |
| Person / Time | |
|---|---|
| Site: | Indian Point, Seabrook, Waterford, 05000000 |
| Issue date: | 05/08/1986 |
| From: | Tarnoff D Office of Nuclear Reactor Regulation |
| To: | Neighbors D, Weiss E, Joshua Wilson, Wilson T NRC |
| Shared Package | |
| ML20151L176 | List: |
| References | |
| FOIA-86-266 NUDOCS 8605280241 | |
| Download: ML20198E762 (22) | |
Text
{{#Wiki_filter:. _ _ _ _ _ q v, [('3 k UNIT ED STATES , J'} e g NUCLEAR REGULATORY COMMISS:ON .j WASHINGTON, D. C. 20555 %,...../ NOTE T0: E. Weiss D. Neighbors J. Wilson R. Bosnak H. Rood P. Moriette FROM: ~ D. Tarnoff, ORAB:DL
SUBJECT:
OPERATING REACTORS EVENTS BRIEFING The next NRR Operating Reactor Events Briefing is scheduled for Tuesday, July 16,1935, at 3:00 p.m. in Conference Room P-422. Direct participants to the presentation will find guidelines in the enclosure. The tentative agenda for the meeting is shown below. , Plant Subject Presenter Indian Point 3 . Steam Generator Inspection Update D. Neighbors Seabrook Crosby Relief Valve Problem IE Waterford Plant Startup Experience J.' Wilson Mojave Steam Line Failure R. Bosnak Combustion Engineering Re-evaluation of C-E H. Rood large Break LOCA Model Paluel (France) Internals Vibration Problems P. Moriette 9 T a / Daniele Tarnoff, x29526 cc: G. Edison K. Seyfrit R. Hernan J. Hannon G. Holahan C. Thomas E. Rossi B. Sheron R.,Baer G. Knighton M. Srinivasan S. Varga B. D. Liaw F. Cherney G. Lanik 8605200241 860508 ~ PDR FOIA MURPHYB6-266 PDR
m, L ENCLOSURE GUIDELINES FOR PRESENTERS 9 l l Each presenter should plan to attend the dry run, which is scheduled for l Tuesday morning, July 16, at 10:00 a.m. in room 550. You should provide ~ -Daniele Tarnoff with your summaries no later than noon on July 15. -It is imperative that the summa'ry should be no longer than one page in the ToTTowing format: Plant Name, captioned title, event date, presenter's name Plant status prior to or during the event (i.e., plant operating at full power; mode 5 for past 6 months) Safety significance and/or briefing significance (i.e., why are we presenting the event) Major points of the sequence of events and/or findings Licensee corrective action
- Generic implication NRC followup action Some examples of bri,efing summaries and simplified diagrams are enclosed.
If you decide to have your summary typed, please make sure that one of the enclosed examples is used as a model for spacing, letter size, etc., and that 1 the document name under which it is entered on the 5520 is telephoned to Debbie Miller (x27415). The Office Director has specifically requested that summaries address fundamental issues of safety significance and generic applicability, and that the briefing for each event run no longer than 10 minutes, including a question or two. Your cooperation is appreciated. m J O e e
GINNA - POST - LOCA CHARC0AL FILTERS POTENTIALLY INOPERABLE MAY 6, 1985 (W. SWENSON, NRR) i PROBLEM - PORTIONS OF CHARC0AL F!LTER DISCHAPGE DUCTS MAY FLOOD FOLLOWING A LOCA SAFETY SIGNIFICANCE - POTENTIAL LOSS OF 2 CF 4 FAN COOLER UNITS AND BOTH CHARCOAL FILTEPS UNDER ACCIDENT CONDIT!0tlS, PROBLEM DISCOVERED BY LICENSEE ANALYSIS, VERIFIED BY ~ CONTAINMENT ENTRY. s DESIGN DEFICIENCY HAS EXISTED FOR LIFE OF PLANT. CORRECTIVE ACTIONS - OPEN MANWAY AND PIN OPEN PRESSURE RELIEF DAMPERS IN DISCHAPGE DUCT. MODIFICATION OF DUCTWORK IS BEING CONSIDEPED AS A PERMANENT FIX. c ~ MAY BE' PROPOSED TO AEOD AS AN ABNORMAL OCCURRENCE 00 "0THER EVENT OF INTEREST." 4 e 9 O 9 r, _.. g --,r
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i ( SURRY 2 - S/G WELD INDICATION-MARCH 20, 1985 (D0N NEIGHBORS) l PLANT IN REFUELING STATUS WELD INSPECTION OF SG-A REVEALED SURFACE AND SUB-SURFACE INDICATIONS IN UPPER TRANSITION CONE GIRTH WELD. INDICATIONS APPEAR TO BE 1/8" DEEP FOR FULL CIRCUMFERENCF CORRECTIVE ACTION - GRIND OUT ~ PARTIAL INSPECTION OF B & C STEAM GENERATORS SHOWS SOME SIMILAR INDICATION REG. 11 HAS INSPECTOR ON SITE NOTE THIS APPEARS TO BE SIMILAR TO WELD CRACKS ON INDIAN ~ POINT 3 STEAM GENERATOR UPPER SHELL TO TRANSITION GIRTH WELD, THAT WERE IDENTIFIED IN MARCH 1982. ~^ IE NOTICE 82-37 ISSUED IN SEPTEMBER 1982 IN RESPONSE TO IE NOTICE 82-37, SURRY 2 DETECTED ~ POTENTIAL WELD PROBLEMS BY ULTRASONIC TESTS IN AucuST 1983. l .,.--..n,. .-.,n., ..,_.n,.
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, = OPERATING REACTORS EVENTS BRIEFING (85-12) INDIAN POINT UNIT 3 - STEAM GENERATOR WELD INDICATIONS SEABROOK MAIN STEAM SAFETY VALVE TEST FAILURE OCONEE UNIT 2 EXTENDED BLOW DOWN FROM MAIN STEAM SAFETY VALVES WATERFORD UNIT 3 PLANT. TRIPS JULY 4-7, 1985 COMBUSTION ENGINEERING LOCA ANALYSIS ERROR M0JAVE GENERATING REHEAT LINE FAILURE STATION PALUEL UNITS 1, 2 IN-CORE INSTRUMENTATION TUBE VIBRATION PROBLEMS WATERFORD/WOLFCREEK _ <a TA /t ( U F E M F /F R 1 F A'YE BYRON / CATAWBA 6 " F # # ' Y # # ) e e
b INDI AN POINT 3 - SG WELD INDICATIONS- ()? bA1EE JULY 16, 1985 C&QN' NEIGHBORS, NRR) ~ P,
- PLANT IN REFUELING STATUS T.S. REQUIRES INSPECTIONS OF SG TRANSITION ZONE UPPER GIRTH WELDS INDICATIONS FOUND BY UT:
SG 31 - 1 SG 32 - 2 SG 33 - 0 SG 34 - 23 i SG-34 HAD WELD REPAIR IN 1983 l MT ON SG-34 SHOWED CLEAN ON 16 0F 23 INDICATIONS REMAINING 7 WELDS ON 34, AND 3 ON 31 AND 33 MAY NOT EXCEED CODE l LICENSEE STILL INSPECTING AND EVALUATING M 4 Y A 2.SQ L L/ f 6Y R& A ct U A F M S C /Y A W / E C NRR HAS LEAD (SINCE 7/15/85) IE DEVELOPING INFORMATION NOTICE
Upper Head ( __, r Support Cylinder 5]. = 2 't Peerle,ss Separator ip Upper Shell Deck Plata { i: Rib Support Svirl Vanc Cylinder 7 D Downcomer Barrel . Z!!: b. reedvater min o - Transition Cone Respper Flow Place e-Tube x Anti-Vibration Bars [f g.. ll I-( I Louer Shell 1 l U 1 ~ u t Tube Supports e f ll l a H Otayrod Spacer 5tayrod , jg ], g f Il[ f 1 l) 8 14 Q X h % % W g Tube Sheet Partition Plata ' Support Pad Channel Esaf. ( . FIGURE 2 3-1 SERIES 51 STEAM GDERATOR
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SEABROOK - CROSBY MAIN STEAM SAFETY VALVE FLOW DEFICIENCY - DECEMBER 1984 (G. HAMMER, NRR) PROBLEM - FULL FLOW TEST RESULTS INDICATE SPRING-ACTUATED MAIN STEAM SAFETY VALVES MAY NOT ACHIEVE RATED FLOW
- CAPACITY, SAFETY SIGNIFICANCE - POSSIBLE INADEQUATE OVERPRESSURE PROTECTIONOFSECONDARYCCCLINCkYSTEMINPWRs 05'uc, rgest vAtyg3 LA5 WYLEfTESTSRESULTt4 INADEQUATE LIFT OF VALVE DISK (ABOUT 50%) WITH THE VENDOR (CROSBY) RECOMMENDED RING SETTING ADJUSTMENTS.
TESTS WERE CONDUCTED TO DETERMINE ADEQUACY OF DISCHARGE PIPING. CORRECTIVE ACTION - RINGS READJUSTED, OBTAINED FULL LIFT ON SEABROOK VALVES GENERIC IMPLICATION - SEABROOK VALVES AND DISCHARGE PIPING SIMILAR TO OTHER PWRs. FULL FLOW TESTS NOT NORMALLY RUN TO ADJUST RINGS, NRC FOLLOWUP ACTION: (1) DEVELOPING IE INFORMATION NOTICE (2) STAFF MAY PURSUE AS A GENERIC ISSUE (3) DISCUSSIONS WITH CROSBY BY REGION 1 AND NRR REGARDING ADEQUACY OF VENDOR GUIDANCE AND SRV RING SETTINGS,
OCONEE 2 - EXTENDED BLOWDOWN FROM MAIN STEAM SAFETY VALVES JULY 11,1985 (H, NICOLARASWAA OCONEE UNIT 2 REACTOR TRIP. FROM 94% POWER CAUSED BY PERSONNEL ERROR TWO MAIN STEAM SAFETY VALVES DID NOT RESEAT AT SETPOINT - EXTENDEDBLOWDOWN-TOABOUTp) PSI TO RESEAT VALVES, OPERATORS REDUCED STEAM PRESSURE THROUGH TURBINE BYPASS VALVES. FAILURE OF CROSBY MAIN STEAM SAFETY VALVES TO PROPERLY RESEAT HAS ALSO REPEATEDLY OCCURRED AT OCONEE UNIT 1 IMPROPER RING SETTING IS A LIKELY CAUSE 0F EXCESS BLOWDOWN, BUT NOT CONFIRMED, DUKE POWER COMMITTED CORRECTIVE ACTIONS TO REGION II 3 low *#M g giug.
SUMMARY
OFPLANTS.REPORTINGPROBLEM[~io LAst yuts; PLANT KNOWN # OF EVENTS '0CONEE 1 7 OCONEE 2 1 TROJAN 1 SALEM 1 t
WATERFORD 3 - PLANT TRIPS JULY 4-7, 1985 (J. WILSON, NRR) WATERFORD 3 EXPERIENCED FOUR REACTOR TRIPS IN LESS THAN THREE DAYS DURING A PORTION OF THIS TIME, THE EFW TURBINE-DRIVEN PUMP WAS UNAVAILABLE DUE TO INADVERTENT BUMPING OF THE MECHANICAL OVERSPEED TRIP LATCH JULY 4 AT 0950 HOURS - 100% PWR - LOW LEVEL-HIGH VIBRATION ON "A" MAIN FEEDWATER PUMP JULY 4 AT 2217 HOURS - 6% PWR - CPC AUXILIARY TRIP ON AXIAL SHAPE INDEX - XE OSCILLATIONS JULY 5"AT 2219 HOURS - 60% PWR - HIGH SG LEVEL DUE TO. OVERFEEDING SG WHILE IN MANUAL CONTROL WITH ONE MAIN FEEDWATER PUMP RUNNING JULY 6 AT 0915 HOURS - TERRY TURBINE OVERSPEED LATCH WAS FOUND TO BE TRIPPED JULY 7 AT 0121 HOURS - LOW SG LEVEL - LOSS OF MAIN FEEDWATER PUMPS ON LOW SUCTION WHILE AN OPERATOR WAS ATTEMPTING TO BACKWASH A CONDENSATE POLISHING SYSTEM FILTER LP&L CORRECTIVE ACTIONS: REMOVING TRIP ON MAIN FEEDWATER PUMP VIBRATION - ALARM ONLY REVISE OPERATING PROCEDURES TRAINING, NIGHT ORDERS
e e CE LOCA ANALYSIS ERROR JULY 2,1985 (H. R00Dh/AR ~ NON-CONSERVATIVE ERROR FOUND IN CE LARGE-BREAK LOCA MODEL CENTER PEAK AXIAL POWER SHAPE YIELDS 34*F HIGHER PEAK CLAD TEMPERATURE (PAT) THAN PREVIOUSLY ASSUMED TOP-PEAKED SHAPE. c. FOR THREE CE PLANTS THAT AR T CYCLE THIS WOULD YIELD A PCT IN EXCESS OF THE 2200*F LIMIT OF 10 CFR 50.46. PLANTS ARE: PALO VERDE 1 SAN ONOFRE 3 WATERFORD 3 BASED ON CE REANALYSIS, OTHER FACTORS IN LARGE-BREAK LOCA MODEL WILL REDUCE PCT TO BELOW 2200*F. LETTERS FROM THESE 3 LICENSEES BEING SUBMITTED GIVING BASIS FOR CONTINUED OPERATION. Q usMr pd on.*.'u f e00\\ OTHER CE LICENSEES BEYOND CYCLE 1 AND HIGHER PCT DOES NOT f REACH 2200*F LIMIT. 2 8 I e a .----__4 ~, _ - - - -.
M0HAVE GENERATING STATION - REHEAT LINE FAILURE JULY 9, 1985 (R. B0SNAK, NRR) FAILURE OCCURRED JUNE 9, 1985 WHEN A 30" REHEAT LINE SUDDENLY SPLIT LONGITUDINALLY ] 1 FRACTURE WAS FISH MOUTH RUPTURE APPR0XIMATELY 20' x 6' FIG 1A SB SAFETY SIGNIFICANCE FOSSIL PLANTS OF S'IMILAR VINTAGE NUCLEAR PLANTS 4 l I 1 O 4 e P
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..e l i Figure 1 (b) Photo of Mohave Pipe Rupture Suited persons are NRC Pipe Review Comittee Members and Consultants ,i l l ) l ,Q.7 '* f !u a a. w. .k, } wy. L. '. )8 3.'" ? *
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~ REHEAT LINE - VITAL STATISTICS 4 DESIGNED TO B31.1 CODE FOR STEAM CONDITIONS OF 1000*F AND 600 PSIG CO STRUCTION LATE 1960 MMENCED OPERATIO FAILURE IN A HORIZONTAL SP0OL 30"-DIAMETER ROLLED AND WELDED OF A-378 C PLATE (1 1/4 CR-1/2 M0) TO MEET A-155 WELDED PIPE COMPARISON WITH LWR PIPING MATERIAL NOT USUALLY USED IN LWR ~ UPPER TEMPERATURE NOT IN CREEP RUPTURE AND CREEP FATIGUE RANGE IN LWR FABRICATION CONTROLS INCLUDING NDE SUPERIOR IN LWR LEAK DETECTION REQUIREMENTS IN LWR O INSERVICE INSPECTION IN LWR FAILURE ANALYSIS ~ RESULTS EXPECTED FROM SCE BY EARLY AUGUST ~
~ PALUEL 1 8 2. IN-CORE INSTRUMENTATION TUBE VIBRATION PROBLEMS MARCH 29, 1985 (P. MORIETTE, NRR) INITIAL EVENT: MARCH 29, 1985, PALUEL 1 IN COLD SHUIDOWN, LEAK DETECTED ON ONE THIMBLE TUBE, WHILE LEAK TESTING IN-CORE INSTRUMENTATION SYSTEM. SUBSEQUENT FINDINGS: APRIL 5: MECHANICAL WEAR (WITHOUT LEAK) ON 4 OTHER
- THIMBLES, APRIL 16: A PROBE CANNOT BE COMPLETELY INSERTED IN ONE THIMBLE (PALUEL 1),
MAY-JUNE: 2 LEAKS ON PALUEL 2, ANOTHER LEAK'0N PALUEL 1 SAFETY SIGNIFICANCE: REACTOR COOLANT LEAKS, OR: NO FLUX
- MAPS, POSSIBILITY OF MIGRANT OBJECTS, MAJOR POINTS:
DEFECTS (OR LEAKS) LOCATED AT DISCONTINUITY IN GUIDING STRUCTURE CAUSE THOUGHT TO BE HYDRAULIC EXCITATION DUE TO TURBULENCE N THE CORE SUPPORT PLATE - BOTTOM OF FUEL ASSEMBLY REGION. DIFFERENCES (FROM 900MWE SERIES) IN LOWER INTERNALS DESIGN AND MEASURED FLOW PARAMETERS SUPPORT THIS HYPOTHESIS, LOWER INTERNALS W DESIGN, CORE INSTRU-MENTATION SYSTEM (0UTSIDE VESSEL) FRAMATOMfDESIGN, GENERIC IMPLIC'ATIONS: ALL 1300MWE SERIES REi: TORS.AFFECTED IN FRANCE 1
~ ~ ~ LICENSEE CORRECTIVE ACTIONS: SHORT TERM: JUSTIFY OPERATION JU5TiFY GFERATiON-WITHOUT IN-CORE INSTRUMENTATION FOR 1 1/2 MONTH. LONG TERM: MODIFY THIMBLE GUIDING PIECES ON TOP OF CORE SUPPORT PLATE FOR BETTER PROTECTION, REDUCE TURBULENT FLOW AROUND THIMBLES. ONLY AFFECTED US FACILITY: SOUTH TEXAS PROJECT 1 a 2 6 e e e S see
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UNPLANNED REACTOR TRIPS
- AVERAGE WEEKLY TRIP FREQUENCY FOR PAST 6 WEEKS IS APPR0XIMATELY 10 TRIPS /WE # #ilCH IS NEAR AVERAGE BREAKDOWN OF REPORTED CAUSES AUTOMATIC EQUIPMENT FAILURES 46%
PERSONNEL ACTIVITIES 46% i MANUAL 8% ' BASED ON 10 CFR 50.72 REPORTS FOR PLANTS WITH LICENSES FOR FULL POWER OPERATION \\ l -.}}