IR 05000369/1986039
| ML20234C614 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 03/11/1987 |
| From: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Tucker H DUKE POWER CO. |
| Shared Package | |
| ML20234C617 | List: |
| References | |
| NUDOCS 8707060569 | |
| Download: ML20234C614 (32) | |
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March 11, 1937
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Docket Nos. 50-369, 50-370 License Nos. NPF-9, NPF-17 Duke Power Company v' ATTN: Mr. H. B. Tucker, Vice President Nuclear Production Department 422 South Church Street Charlotte, NC 28242 Gentlemen:
SUBJECT: MEETING SUMMARY OF ENFORCEMENT CONFERENCE - WRC INSPECTION REPORT NOS. 50-369/86-39 AND 50-370/86-39 REGARDING NRC INSPECTION REPORT NOS. 50-369/85-38 AND 50-370/85-39 This refers to a meeting held on December 8,1986, at the Nuclear Regulatory i'
Commission's (NRC) Region II Office in Atlanta, Georgia.
The subject of this meeting was contained in Inspection Report Nos. 50-369/85-38 and 50-370/85-39 and concerned the operability of the Nuclear Service Water (RN) system at the McGuire Nuclear Station.
Enclosed is a meeting summary including a list of attendees, a summary of issues discussed, an outline of the information presented by Duke Power Company regarding specific issues discussed, and correspondence from NRR rendering the NRC's technical position on the operability of the RN system.
In accordance with Section 2.790 of NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and its enclosure will be placed in the NRC Public Document Room.
Should you have any questions concerning these matters, we will be pleased to discuss them.
Sincerely, 9 7 7 1 ~ " 3 3'A
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J. Nelson Grace Regional Administrator Enclosure:
Meeting Summary w/5 attachments cc w/ encl:
s'T. L. McConnell, Station Manager i
bcc w/ encl:
(See Page 2)
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Duke Power Company
March 11, 1987 bec w/ enc 1:
VD. Hood, NRR G. R. Jenkins, RII R. B. Moore, RII l
vhRC Resident Inspector i
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Document Control Desk State of North Carolina I
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ATTACHMENT 1 List of Attendees at McGuire Enforcement Conference December 8, 1986 Name Organization
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T. A. Peebles NRC, Region II W. T. Orders Senior Resident Inspector, McGuire
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V. L. Brownlee NRC, Region II Neal Rutherford Duke Power Company Hal B. Tucker Duke Power Company Tony L. McConnell Duke Power Company, MNS Bruce H. Hamilton Duke Power Company, MNS Robert Gill Duke Power Company William J. Kronenwetter Duke Power Company G._R. Jenkins NRC, Region II Caudie Julian NRC, Region II'
Al Gibson NRC, Region II M. L. Ernst NRC, Region II Wayne Revels Duke Power Company William M. Suslick Duke Power Company, MNS Edward 0. McCraw Duke Power Company, Compliance Peter B. Moore NRC, Region II Bruno Uryc NRC, Region II Howard J. Wong NRC, Sr. Enforcement Specialist, IE Ed Yachimiak NRC, Sr. Enforcement Specialist, IE
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ATTACHMENT 2
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DISCUSS THE JANUARY RESULTED IN THE CAVITATION OF THE NU PUMPS AND THE RESULTANT NUCLEAR SERVICE WATER SYSTEM LOW FLOW CONDITION, JANUARY 27, 1986 FLOW BALANCE - THIS TEST WAS BEING PERFORMED TO CONFlRM THE OPERABILITY OF TRAIN 1A.
THE NRC HAD RAISED CONCERNS REGARDING CORROSION AND FOULING AFFECTING THE NUCLEAR SERVICE WATER SYSTEM FLOW BALANCE.
TEST CONDITIONS
NUCLEAR SERVICE WATER TRAIN 1A SUPPLYING LOCA LOADS WITH SUCTION FROM THE SNSWP
NUCLEAR SERVICE WATER TRAIN 2A SECURED FSAR FLOW RATES FOR COMPONENT COOLING (8000
GPM) AND CONTAINMENT SPRAY (5000 GPM)
TEST RESULTS
THE FLOW BALANCE FAILED AND THE 1A TRAIN WAS NOT DECLARED OPERABLE.
- WHILE VALVING THE LAST MAJOR LOAD THE PUMP DISCHARGE PRESSURE AND FLOW RATE SUDDENLY DECREASED.
- THE OPERATORS IMMEDIATELY RECOGNIZED PUMP WAS EXPERIENCING AN NPSH PROBLEM AND TERMINATED THE TRANSIENT BY THROTTLING BACK ON FLOW TO THE LAST MAJOR LOAD.
JANUARY 28, 1986 FLOW BALANCE - THIS TEST WAS BEING PERFORMED TO REINSTATE THE 1A TRAIN TO OPERABLE STATUS.
ADDITIONALLY, WE HAD REAll2ED THAT A VALID FLOW BALANCE MUST BE PERFORMED WITH THE COMPANION UNIT 2 TRAIN IN SERVICE SUPPLYING BLACKOUT LOADS.
TEST CONDITIONS
NUCLEAR SERVICE WATER TRAIN 1A SUPPLYING LOCA LOADS WITH SUCTION FROM THE SNSWP
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NUCLEAR SERVICE WATER TRAIN 2A SUPPLYING B/O
LOADS WITH SUCTION FROM THE SNSWP (TRAINS 1A AND 2A SHARE A COMMON SUCTION LINE)
NEWLY APPROVED ACCIDENT ANALYSIS FLOW RATES
i FOR COMPONENT COOLING (6000 GPM) AND CONTAINMENT (3800 GPM).
THESE NEW FLOW RATES HAD BEEN REQUESTED FROM WESTINGHOUSE DURING THE PREVIOUS WEEK AND HAD JUST BECOME AVAILABLE.
TEST RESULTS THE FLOW BALANCE WAS SUCCESSFUL AND THE'1A,
TRAIN WAS DECLARED OPERABLE.
- ALTHOUGH THE CAVITATION INCIDENT DID NOT REPEAT ITSELF UNDER FLOW BALANCE CONDITIONS, THE TEST COORDINATOR INCREASED 1A PUMP
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DISCHARGE FLOW INTENTIONALLY TO TEST FOR NPSH PROBLEMS.
PROBLEMS INDICATING INSUFFICIENT NPSH WERE NOTED AT A FLOW ~ 500 GPM ABOVE THAT REQUIRED FOR THE FLOW BALANCE.
CONCLUSIONS IN RETROSPECT, THE JANUARY 27, 1986 CAVITATION
INCIDENT INDICATED THAT THE 1A TRAIN OF THE NUCLEAR SERVICE WATER SYSTEM, UNDER WORST CASE ACCIDENT CONDITIONS AND USING THE FSAR VALUES FOR HEAT EXCHANGER FLOW RATES, HAD BEEN
" TECHNICALLY INOPERABLE".
THAT IS TO SAY THE 1A PUMP WOULD HAVE EXPERIENCED CAVITATION DUE TO INSUFFICIENT NPSH WHICH, IF NOT CORRECTED WOULD HAVE LED TO EVENTUAL PUMP FAILURE.
TRAINS 1B, 2A AND 2B MAY HAVE BEEN TECHNICALLY
INOPERABLE UNDER THESE SAME CONDITIONS.
TEST RESULTS INDICATE THAT THE 1A NUCLEAR SERVICE WATER PUMP HAS THE WORST PROBLEM WITH NPSH.
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IT IS NOT CLEAR THAT THE SAME PROBLEM WOULD OCCUR WITH THE OTHER~3 PUMPS.
THE PLANT OPERATORS, USING THE MULTlPLE
INDICATIONS AVAILABLE TO THEM, WERE ABLE TO QUICKLY DIAGNOSE THE NPSH PROBLEM.
SINCE THE FAILURE OF THE PUMP WOULD HAVE TAKEN HOURS (IF i
NOT DAYS) TO OCCUR, WE ARE CONFIDENT THE PLANT OPERATORS WOULD HAVE RECOGNIZED THIS CONDITION IN TIME TO AVERT A FAILURE OF THE PUMP.
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DISCUSS WHY THE MCGUIRE UNITS 1 AND 2 NUCLEAR SERVICE WATER SYSTEMS WERE NOT TESTED IN THE DESIGN BASIS ACCIDENT CONFIGURATION PRIOR TO JANUARY 27, 1986.
PREOPERATIONAL TEST DEVELOPMENT
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THE UNIT 1 TEST WAS WRITTEI TO MEET THE
REQUIREMENTS OF THE FSAR CFaPTER 14 TEST ABSTRACT.
THIS ABSTRACT DID NOT SPECIFY TESTING THE NUCLEAR SERVICE WATER SYSTEM UNDER DESIGN BASIS ACCIDENT CONDITIONS.
THE DESIGN BASIS ACCIDENT CONDITION, AS ESTABLISHED
IN ANOTHER PORTION OF THE FSAR, IS BOTH UNITS 1 AND 2 TRAINS A AND B ALIGNED TO THE SNSWP WITH ONE UNIT SUPPLYING ITS LOCA LOADS AND THE OTHER UNIT SUPPLYlNG ITS B/O LOADS.
AT THE TIME THE UNIT 1 NUCLEAR SERVICE WATER SYSTEM
PREOPERATIONAL TEST WAS PERFORMED THE UNIT 2 SYSTEM WAS UNDER CONSTRUCTION AND UNAVAILABLE FOR OPERATION.
THE UNIT 2 TEST WAS WRITTEN TO MEET THE
REQUIREMENTS OF THE FSAR CHAPTER 14 TEST ABSTRACT AND USED THE UNIT 1 COMPLETED TEST PROCEDURE AS A GUIDE.
CONCLUSION IN RETROSPECT, THE APPROPRIATE TESTING SHOULD HAVE
BEEN PERFORMED WHEN THE UNIT 2 NUCLEAR SERVICE WATER SYSTEM BECAME AVAILABLE TO TEST.
TESTING PRIOR TO THIS WOULD NOT HAVE BEEN POSSIBLE.
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DISCUSS WHY PREOPERATIONAL NUCLEAR SERVICE WATER COMPONENT THROTTLE SETTINGS WERE INCORRECTLY TRANSLATED TO OPERATING PROCEDURES.
DESCRIPTION OF THE PROBLEM FOLLOWING THE UNIT 1 NUCLEAR SERVICE WATER TEST THE
" AS LEFT" MANUAL THROTTLE VALVE SETTINGS FOR THE SMALL HEAT EXCHANGERS WERE FORMALLY COMMUNICATED TO THE OPERATIONS PROCEDURAL GROUP.
THESE SETTINGS WERE INCORPORATED IN THE UNIT 1 OPERATING PROCEDURES.
FOLLOWING THE UNIT 2 NUCLEAR SERVICE WATER
PREOPERATIONAL TEST THE "AS LEFT" MANUAL THROTTLE VALVE SETTINGS WERE NOT COMMUNICATED TO THE OPERATIONS PROCEDURE GROUP.
THE OPERATING PROCEDURES WERE NOT CHANGED TO INCORPORATE THE PROPER SETTINGS.
THE PROCEDURES WERE LEFT AS ORIGINALLY WRITTEN PRIOR TO SYSTEM STARTUP SIMPLY REQUIRING THE VALVES TO BE " THROTTLED".
(IT SHOULD BE NOTED THAT ONLY THE MANUAL THROTTLE VALVES ON UNIT 2 SMALL HEAT EXCHANGERS WERE AFFECTED).
THE UNIT 2 MANUAL THROTTLE VALVES WERE FOUND OPEN
DURING THE FLOW BALANCES IN 1986.
CURRENT SITUATION
ON BOTH UNITS WE NOW INTENTIONALLY FLOW BALANCE THE NUCLEAR SERVICE WATER SYSTEM VITH THE MANUAL THROTTLE VALVES FOR THE SMALL HEAT EXCHANGERS FULLY OPENED.
WE DO THIS BECAUSE IT IS THE MOST METHOD OF FLOW BALANCING AND IT IS EASIEST TO
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CONCLUSION THE UNIT 2 NUCLEAR SERVICE WATER MANUAL THROTTLE
VALVE "AS LEFT" SETTINGS ON SMALL HEAT EXCHANGERS WERE INADVERTENTLY NOT INCLUDED IN UNIT 2 OPERATING PROCEDURES DUE TO ADMINISTRATIVE OVERSIGHT.
SINCE THE VALVES WERE ALL FOUND FULLY OPENED THERE
WAS NO EFFECT ON OPERABILITY.
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DISCUSS WHY THE PREOPERATIONAL TESTS DlD NOT VERIFY REQUISITE NUCLEAR SERVICE WATER FLOW TO UNITS 1 AND'2 CONTROL ROOM AIR CONDITIONERS.
UNIT 1 PREOPERATIONAL TEST COMPLETED 7/25/79 DURING PERFORMANCE OF THE TEST, THE NUCLEAR SERVICE
WATER FLOW TO THE CONTROL ROOM AIR CONDITIONING COULD NOT BE ACCURATELY DETERMINED DUE TO DESIGN PROBLEMS WITH THE INSTALLED INSTRUMENTATION.
- BASED ON THE TEST COORDINATORS ENGINEERING JUDGEMENT THAT FLOW TO THIS PARTICULAR HEAT EXCHANGER WAS ADEQUATE, THE FACT THAT MANY FLOWS TO SMALLER LOADS ON THE NUCLEAR SERVICE WATER SYSTEM WERE NOT MEASURED, AND HIS BELIEF THAT THE TEST COULD BE PERFORMED CONSERVATIVELY BY FAILING THE AIR OPERATED THROTTLE VALVE OPEN, A PROCEDURE CHANGE TO DELETE THE REQUIREMENT TO VERIFY A FLOW OF 789 GPM TO THE AIR CONDITIONERS WAS APPROVED.
UNIT 2 PREOPERATIONAL TEST COMPLETED 11/12/82 THE TEST COORDINATOR OF THIS TEST WAS AWARE OF THE
FLOV MEASUREMENT PROBLEM AND BASED ON THIS, PLUS THE UNIT 1 APPROVED PROCEDURE CHANGE, DID NOT INCLUDE THE REQUIREMENT IN THE UNIT 2 TEST.
CURRENT STATUS
THE "B" TRAIN FLOW ORIFICE HAS BEEN MOVED TO A MORE SUITABLE LOCATION AND IS NOW ACCURATE.
MEASUREMENTS HAVE CONFIRMED THE EARLIER ASSUMPTION THAT THE FLOW IS SUFFICIENT.
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"A" TRAIN FLOW HAS BEEN MEASURED USING ULTRASONICS AND HAS BEEN VERIFIED TO BE SUFFICIENT.
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RELOCATION OF THIS FLOW ORIFICE IS PLANNED FOR THE NEAR FUTURE.
(BOTH UNITS SHARE THE SAME AIR CONDITIONING UNITS)
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THE TEST COORDINATORS ENGINEERING JUDGEMENT HAS BEEN PROVEN TO BE VALID BY SUBSEQUENT MEASUREMENTS.
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THE FAILURE TO MEASURE THIS FLOW DURING PREOPERATlONAL TESTING DOES NOT AFFECT OPERABILITY.
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DISCUSS WHY DUKE POWER COMPANY DID NOT PERFORM A FORMAL ENGINEERING EVALUATION OF THE 12/17/85, NUCLEAR SERVICE WATER SYSTEM FLOW TEST RESULTS UNTIL JANUARY, 1986.
THE 12/17/85 RN FLOW BALANCE TARGET ACTUAL COMPONENT COOLING WATER 8000 8000 CONTAINMENT SPRAY
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5000 4800x D/G COOLING WATER 900 900 CONTROL ROOM CHILLER 789 707x CHARGING PUMP OIL COOLER
15x SAFETY INJECTION PUMP OIL COOLER
21 SPENT FUEL POOL PUMP AHU
15x CONTAINMENT SPRAY PUMP AHU
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THIS WAS A "FIRST OF A XIND" TEST FOR MCGUIRE.
THOSE INVOLVED DID NOT FEEL AN URGENT NEED TO FORMALLY EVALUATE THESE RESULTS BUT WERE MORE INTERESTED IN CONFIRMING THE TEST METHOD.
- AN INFORMAL EVALUATION WAS PERFORMED IMMEDIATELY BY STATION, DESIGN ENGINEERIN3 AND THE G.O.
LICENSING PERSONNEL AND RESULTED IN THE CONCLUSION THAT THE HEAT EXCHANGERS WERE " DEGRADED" BUT NOT INOPERABLE.
THIS INFORMAL EVALUATION WAS MAINLY BASED UPON THE DESIGN ENGINEER'S UNDERSTANDING OF THE EFFECT OF DECREASED COOLING WATER FLOW ON HEAT EXCHANGER PERFORMANCE.
IN ADDITION, A WESTINGHOUSE ANALYSIS WAS IN HAND THAT CONCLUDED 4800 GPM TO THE CONTAINMENT SPRAY HEAT EXCHANGER WAS ACCEPTABLE.
- THESE CONCLUSIONS WERE LATER SHOWN CORRECT WHEN A FORMAL ENGINEERING EVALUATION WAS REQUESTED BY THE STATION ON THE ADVISE OF THE NRC RESIDENT INSPECTOR.
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THE NEED FOR FORMAL JCO (JUSTlFICATION FOR
CONTINUED OPERATION) WAS NOT RECOGNIZED BY.THE PERSONS INVOLVED WITH THE TEST.
THIS REQUIREMENT IS MUCH BETTER ESTABLISHED NOW AND THE METHOD OF
OBTAINING A JCO HAS BEEN WIDELY COMMUNICATED.
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A FORMAL ENGINEERING EVALUATION SHOULD HAVE BEEN PERFORMED IMMEDIATELY AFTER THE 12/17/85 TEST.
- LATER CALCULATIONS VERIFY THAT THE SYSTEM WAS OPERABLE AT THE TIME OF THIS TEST.
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'HE' GOVERNING PROCEDURE, PT/1/A/4403/04, LACKED R QUANTITATIVE ACCEPTANCE CRITERIA.
NUCLEAR SERVICE ON OF THE PROCEDURE t]ONTAINMENT SPRAY 1ANCE OF THE PT/1/A/4403/04 WAS NEVER INTENDED TO HAVE AN ACCEPTANCE CRITERIA.
PT/1/A/4403/04 WAS WRITTEN TO COLLECT DATA REQUIRED NG BY IE BULLETIN 81-03 WHICH ADDRESSES THE POTENTIAL FOULING OF SAFETY RELATED HEAT EXCHANGERS BY ASIATIC CLAMS AND SHELL DEBRIS.
TAINMENT SPRAY TWO HEAT EXCHANGERS WERE MONITORED ON A QUARTERLY OLATION VALVE BASIS AT AN ESTABLISHED FLOW RATE AND THE DELTA P I OF THE MAS RECORDED AND TRENDED.
A STAFF PERSON REVIEWED S QUESTIONABLE.
THESE TRENDS FOR SIGNS OF ASIATIC CLAMS OR SHELL RECORDED BUT WE DEBRIS.
(LATER THAT DAY ALTHOUGH NOT REQUIRED, WE ELECTED TO USE A THIS VALVE).
PROCEDURE TO GATHER THIS DATA.
THE USE OF A
.S NOT ALIGNED PROCEDURE FACILITATED SCHEDULING.
- XACT ALIGNMENT IE NUCLEAR NT FOR ACCEPTANCE CRITERIA
.OW (42 PSIG) AND
.DERS WERE BEING A PROCEDURE USED MERELY TO GATHER DATA IS NOT
" TESTING" AND THEREFORE AN ACCEPTANCE CRITERIA IS NEITHER NEEDED NOR REQUIRED.
10CFR50 APPENDIX B CRITERIA V REQUIRES THAT PROCEDURES INCLUDE APPROPRIATE QUANTITATIVE AND CONTROL ROOM QUALITATIVE ACCEPTANCE CRITERIA FOR DETERMINING 100 GPM THROUGH THAT IMPORTANT ACTIVITIES HAVE BEEN SATISFACTORILY R.
4600 GPM WAS ACCOMPLISHED.
THE IMPORTANT ACTIVITY IN THIS CASE WAS MERELY THAT THE DATA WAS TAKEN AND REVIEWED BY
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iT ON MORE FLOW.
ATOR COULD NOT N
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(REMEMBER AN ACCEPTANCE CRITERIA FOR PT/1/A/4403/04 WAS NOT LENT CONDITIONS.)
REQUIRED.
IED THROUGH THIS OF THIS TEST, IED THROUGH THIS
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HEAT EXCHANGER ON THE FOLLOWING DAY BY ADJUSTING SYSTEM ALIGNMENTS.
CONCLUSIONS THE AS FOUND CONDITION OF 800 GPM FLOW TO THE
CONTAINMENT SPRAY HEAT EXCHANGER IS SIGNIFICANT ONLY IN THAT THE THROTTLE VALVE WAS CLOSED MORE THAN REQUIRED TO OBTAIN THE REQUIRED ACCIDENT FLOW.
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SINCE THE CONTAINMENT SPRAY HEAT EXCHANGERS ARE
VALVED IN BY THE OPERATORS AFTER THE INITIAL AUTOMATIC ACTIONS HAVE TAKEN PLACE, AND ALL OPERATORS ARE AWARE-(AS VERIFIED BY THE NRC INSPECTOR) OF THE REQUIRED ACCIDENT FLOWRATE WE ARE CONFIDENT THE VALVE POSITION WOULD NOT HAVE LED TO A PROBLEM.
THE STATION COULD HAVE ACHIEVED 5000 GPM IF
AllGNMENTS WERE MODIFIED TO ACCOMPLISH THIS.
THE 4600 GPM FLOW HAS NO SIGNIFICANCE IN ITSELF SINCE WE WERE NOT TESTING TO VERIFY THE FLOW BALANCE.
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CURRENT STATUS OF THE PERIODIC TESTING AND MAINTENANCE PROGRAM i
WHICH WAS PROPOSED IN MARCH, 1986 AS A RESULT OF PROBLEMS UNCOVERED DURING THE EVALUATION OF THE NUCLEAR SERVICE WATER SYSTEM.
PROGRAM AND SYSTEM STATUS
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THE NUCLEAR SERVICE WATER PROGRAM FOR PERIODfC TESTING AND MAINTENANCE HAS BEEN FULLY IMPLEMENTED SINCE APRIL, 1986.
- THE PROGRAM IS PERFORMING AS DESIGNED AND SEVERAL DEGRADED COMPONENTS HAVE BEEN DISCOVERED.
AFTER DISCOVERY, THE COMPONENTS HAVE BEEN PROMPTLY CORRECTED AND THE TECHNICAL SPECIFICATION ACTION ITEMS HAVE BEEN COMPLIED WITH.
- MOST PROBLEMS HAVE ARISEN WITH COMPONENTS THAT ARE NORMALLY IN SERVICE, SUCH AS; COMPONENT COOLING AND CERTAIN MOTOR COOLERS.
SOME CASES OF FOULING HAVE BEEN NOTED ON NORMALLY INACTIVE EQUIPMENT AND THESE CAN BE ATTRIBUTED TO POOR EQUIPMENT ISOLATION.
- WE HAVE NOTED THAT DEGRADATION IS SLOW DURING THE WINTER, SPRING AND SUMMER MONTHS, AND THAT IT INCREASES IN THE FALL.
FIFTY-SEVEN (57) TESTS HAVE BEEN PERFORMED WITH
FORTY-ElGHT (48) HAVING ACCEPTABLE RESULTS.
NINE (9) TRAIN FLOW BALANCES HAVE BEEN PERFORMED
AND ALL HAVE BEEN SUCCESSFUL.
(IN FACT, ALL FLOW BALANCE PERFORMED SINCE THE LOWER FLOW REQUIREMENTS OF 1/28/86 WERE IMPLEMENTED HAVE BEEN SUCCESSFUL).
IN ADDITION TO THE NINE (9) HEAT EXCHANGERS
IDENTIFIED FOR CLEANING BY TESTING, FIVE (5) HEAT EXCHANGERS HAVE BEEN CLEANED AS A RESULT OF PERIODIC MAINTENANCE REQUIREMENTS NOW IN PLACE.
PERIODIC MAINTENANCE IS ESTABLISHED FOR ALL HEAT EXCHANGERS THAT CAN NOT BE TESTED WITH THE INTERVAL BASED ON THE AS-FOUND FOULING.
CURRENT PROGRAM DESCRIPTION (SEE ATTACHED _ DESCRIPTION)
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SOME MINOR PROGRAM CHANGES HAVE BEEN MADE BASED ON f
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OUR EXPERIENCE SINCE APRIL, 1986.
THESE CHANGES ADDRESSED DIFFICULTIES IN PERFORMING CERTAIN TESTS OR SCHEDULING RESTRAINTS DUE TO EQUIPMENT AVAILABILITY.
NONE OF THESE CHANGES AFFECTS THE QUALITY OF THE PROGRAM.
PENDING CHANGES TO THE PROGRAM
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AS PLANNED MODIFICATIONS ARE COMPLETED, MORE
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COMPONENTS WILL BE ADDED TO THE TESTING PROGRAM AND I
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ELIMINATED FROM THE PERIODIC CLEANING PROGRAM.
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WE ARE NOW MODIFYING THE TESTING PROGRAM TO INCORPORATE SOME OF THE ASPECTS OF THE IWV/lWP PROGRAM.
" ALERT" RANGES WILL BE ESTABLISHED TO COMPLIMENT THE ALREADY EXISTING " ACTION" RANGES.
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THIS IS OUR PROGRAM AND WE ARE PROUD OF IT.
WE ARE CONFIDENT THIS PROGRAM ENSURES THE CONTINUED
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OPERABILITY OF OUR NUCLEAR SERVICE WATER SYSTEM.
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MODIFICATIONS TO THE NUCLEAR SERVICE WATER SYSTEM ALREADY COMPLETED:
ALL 4 PUMP DISCHARGE FLOW ELEMENTS WERE REPLACED WITH
STAINLESS STEEL SPOOL PIECES CONTAINING CAllBRATED ORIFICE PLATE.
LARGE TAPS WERE ADDED ON THE RN SYSTEM AT THE
CONTAINMENT SPRAY (NS) HEAT EXCHANGERS FOR DRAINING, CHEMICAL CLEANING AND INSPECTION.
SOME ADDITIONAL SYSTEM DRAINS HAVE BEEN ADDED, MORE ARE
SCHEDULED FOR 1987 AND OTHERS ARE BEING EVALUATED.
INSULATION ON THE COMPONENT COOLING (KC) HEAT EXCHANGERS
AND PIPING 15 CURRENTLY BEING COMPLETED, ALLOWING INCREASED RN FLOW RATES TO REDUCE RN PUMP WEAR.
MODIFICATIONS TO THE NUCLEAR SERVICE WATER SYSTEM SCHEDULED FOR 1987:
ALL 4 PUMPS WILL. RECEIVE STAINLESS STEEL IMPELLERS AND
MECHAhlCAL SEALS.
CURRENT IMPELLERS ARE LEADED BRONZE.
NEW IMPELLERS WILL WITHSTAND CAVITATION BETTER THAN OLD ONES.
NS WET LAYUP SYSTEM WILL BEGIN ON ALL 4 HEAT EXCHANGERS,
NEW DOWNSTREAM ISOLATION VALVES AND LIFT SYSTEM ARE REQUESTED.
STAINLESS STEEL SPOOL PIECES CONTAINING FLOW ELEMENTS
FOR UNIT 1 TRAIN A FLOW THROUGH KC AND NS HEAT
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EXCHANGERS WILL BE INSTALLED.
WE HOPE TO IMPROVE FLOW BALANCE RESULTS ON TRAIN A.
OTHERS WILL BE CONSIDERED BASED ON THE RESULTS OF THESE.
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ON UNIT 1, BYPASS LINES AROUND THE PUMP DISCHARGE CHECK VALVES WILL BE INSTALLED TO FACILITATE SYSTEM DRAINING.
(UNIT 2 WAS DONE EARLIER).
- FOR BOTH UNITS, SMALL DIAMETER RN PIPING (2" AND UNDER)
WILL BE REPLACED WITH STAINLESS STEEL FOR THESE MOTOR COOLERS:
RN PUMPS, AUXILIARY FEEDWATER (CA) PUMPS, SAFETY INJECTION (NI) PUMPS AND CENTRIFUGAL CHARGING (NV) PUMPS.
PROVISIONS FOR DP MEASUREMENT AND CHEMICAL CLEANUP OF PIPING AND HEAT EXCHANGER HAVE BEEN INCLUDED.
TAPS WILL BE ADDED TO ALLOW DP MEASUREMENTS ON THE VC/YC
CONDENSER AND DIESEL GENERATOR COOLING WATER (KD) HEAT EXCHANGERS.
MODIFICATIONS TO THE NUCLEAR SERVICE WATER SYSTEM SCHEDULED FOR 1988:
NS WET LAYUP SYSTEM WILL BE COMPLETED, RECIRCULATION
PUMP AND CHEMICAL FEEDER WILL BE INSTALLED.
FOR BOTH UNITS, SMALL DIAMETER RN PIPING 2" AND UNDER
WILL BE REPLACED WITH STAINLESS STEEL FOR THESE MOTOR COOLERS:
KC PUMPS, RESIDUAL HEAT REMOVAL (ND), NS AND FUEL POOL COOLING (KD) AIR HANDLING UNITS.
PROVISIONS FOR DP MEASUREMENT AND CHEMICAL CLEANUP HAVE BEEN INCLUDED.
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ATTACHMFNT 3 10CFR50.59 EVALUATION OF A TRAIN CROSSTIE In early October '85 ficGuire Nuclear Station notified Design that the 1A Nuclear Service Water pump had failed its quarterly performance test.
After changing the impeller and repeated performance tests, we suspected at that time erroneous flow measurements.
This has since been confirmed with the installation of stainless steel spool pieces upstream and downstream of the flow element. However at that time we had to assume the pump was degraded.
We noted that combined LOCA and non-LOCA flow requirements of 21,957 gpm could be supplied with 1A and 2A pumps running in parallel.
We confirmed this by first plotting the degraded pump curve next to the vendor pump curve.
Since pump 2A was not degraded, we used the vendor curve to represent pump 2A.
The combined pump curve for two pumps operating in parallel can be plotted by adding flows at the same total dynamic head. We ncted that 15040 gpm at 158 feet of head was required for design basis accident.
The non-LOCA flow requirement is 6,917 gpm.
Therefore the combined LOCA and non-LOCA requirements is 21,957 gpm at 158 feet. As can be seen, this point is below the combined pump curve.
Therefore even though pump 1A was suspected of being degraded, pumps 1A and 2A operating in parallel would provide all design basis accident flows.
However, in order to do this we had to make sure that this would satisfy all safety concerns. There is normally two independent trains.
Each of the A and B suction trains split into separate Unit 1 and Unit 2 headers.
The headers then combine into common discharge headers. What we proposed to do was to combine the 1A and 2A pump discharges together.
As can be seen, there still remained two independent Nuclear Service Water Trains.
This alignment can take any single failure and provide all required flows.
If an A train pump failed, the combined A train could possibly be lost; however, the B trains would still provide all required flows.
This met the requirement of Technical Specifications 3.7.4.
We had to satisfy other safety concerns as listed in 10CFR50.59.
The three major concerns were:
1.
Woula the consequences of an accident previc,usly evaluated be increased?
- No, Nuclear Service Water System would supply all required flows during a design basis accident with any single failure.
2.
Would there exist an accident or malfunction of equipment of a differcnt type than previously analyzed?
- No, the single failure of unit shared Nuclear Service Water System had been evaluated since units share common suction and discharge lines.
_
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,
-
.
3.
Would the margin of ssfety as defined in the basis for any technical specification be reduced.
- No, margin of safety had not been reduced since two independent trains of Nuclear Service Water still existed.
Conclusinns:
Attached to your handout is the Safety Analysis performed in early October 1985. We conclude that this alignment and Safety Evaluation did meet the criteria of 10CFR50.59.
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50.59 EVALUATION
1.
WOULD THE CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED BE INCREASED
-
e NUCLEAR SERVICE SYSTEM WOULD SUPPLY ALL REQUIRED FLOWS DURING A DESIGN BASIS ACCIDENT WITH ANY SINGLE FAILURE.
2.
WUULD THERE EXIST AN ACCIDENT OR MALFUNCTION OF EQUIPMENT OF A DIFFERENT TYPE THAN PREVIOUSLY ANALYZED e IHE SINGLE FAILURE OF UNIT SHARED NUCLEAR l
SERVICE bYSTEM HAD BEEN EVALUATED SINCE
!
UNITS SHARE COMMON SUCTION AND DISCHARGE j
LINES.
3.
WOULD THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY
,
TECHNICAL SPECIFICATION BE REDUCED l
e MARGIN OF SAFETY HAD.NOT BEEN REDUCED i
SINCE TWO INDEPENDENT TRAINS OF NUCLEAR I
SERVICE WATER STILL EXISTED.
.
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_______________m.___.______________________________
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Form 34634 (5-85)
DUKE POWER COMPAhT NUCLEAR SAFETY EVALUATION CHECKLIST
!
(1) STATION:
//I ( b / / G.
UNIT: 1
/
3 OTHER:
(2) EVALUATION APPLICABLE TO (DESCRIPTION AND NUMBER OF NSM, PROCEDURE, I.
PROCEDURE CHANGE, OR TEST / EXPERIMENT): CP/1/ A /(M oc/O(-,
AJucleN Serme v/Ah r 5,f s u m (3) SAFETY EVALUATION - PART A j
The item to which this evaluation is applicable represents:
Yes No A change to the station or procedures as described in the FSAR:
oratestorexperimentnotdesertgdin the FSAR? Affected FSAR Section(s) are:.GM EdL,2 If the answer to the above is "Yes," identify the affected section(s) of j
the FSAR. Attach additional sheets as necessary.
j (
(4) SAFETY EVALUATION PART B
{
/
Yes No Will this item require a change to the station Technical Specifications? Affected Tech. Specs.
Sections (s) are:
If the answer to the above is "Yes," identify the specification (s)
affected and/or attach the applicable page(s) with the change (s) indicated.
Tech. Spec. changes require NSRB and NRC approval prior to use. _
(5) SAFETY EVALUATION - PART C As a result of the item to which this evaluation is applicable:
Yes No Will the probability of an accident previously y
evaluate in the FSAR be inc e sed? Explatn: d/./>dT (M
W Luuk4 5 d &
%erbA/ 5 4+h %%s & D W ew W' W J l' W 4-~< K va.
Yes No WillQthe consequencks of an accident previouslyh % a evaluated in the FSAR be i.ncreased? Explatn:
nA wit [ N e d M
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_ _ _ _ _
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Form 34634 (5-84)
'
j DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST Yes No May the possibility of an accident which is different than already evaluated in the FSAR be created?
Explain: 6/c_.
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important to safety previ usi ev 1 ated in the JSAR
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hat 9 % mo L &M D~M Y l1Will the consWqu)e,nces o&
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important to. safety previously evaluated in the FSAR be increased? Explain: ONw M %,
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y the possibility of malfunctions ofJ equipm t important to safety different than any alreadv ('
created?
xp1 in: Y evgluated'ntheFSAR
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Technical Speci ication be r duced? Explain:
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Justification for t answers ove ( s or No) must be rov ed in the
above spaces (attach additional sheets as necessary).
An unreviewed safety question g involved if any answer to Part C above is
"Yes" and NRC authorization is required.
(6) Prepared by:
Date:
(7) Reviewed by:
Date:
(Qualif ted Reviewer)
(8) Page 2 of
-
Figure 4.2-16 (2 of 2)
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Fcrm 34811 (7 81)
(Formsrly SPD 1003 2)
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DUKE POWER COMPANY (1)
ID No:0/
44#0/0 5 _
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PROCEDURE MAJOR CHANGE Change No: 50
'
PROCESS RECORD
- Ps._...../ Restricted To (2) STATION: MM (3) PROCEDURE TITLE: OW (f/ hvM f Mb IVMk
'
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(4)
SECTION(S) 0F PROCEDURE AFFECTED: b h
st24 t,-
8b (5) DESCRIPTION OF CHANCE:
(Attach additional pages w "l0cifnecessaryl.)S'l "N hy& $0$sW eh $ bllddig rLJ t
/ded' W :
IRW-330?D %f l# Md Cronow Es*U)
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REASON FOR CHANGE: g g
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(7) PREPARED BY:
DATE:
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SAFETY EVALUATION This change:
Yes o
Represents a change to the station or procedures as described in the FSAR, or a test or experiment not described in the FSAR7
!JequiresachangetothestationTechnicalSpecifications?
Yes No Yes No t/ Involves an unreviewe saf ety question'hectist-
'
5ee W/A:lecl A)ac/ew s B Mu*hoa C If the answer to any of the above is "Y s",
attach a detailed explanation.
As appro *1 ate attach a cot::pleted " Nuclear Saf ety Evaluation Check List" fort.
/O/7/b By:
"
I Date:
(9)
REVIEWED BY:
DATE:
,
l Cross-Disciplinary Review By:
N/R:
(10) TEMDORARY APPROVAL (IF NECESSARY):
By:
(SRO) Date:
By:
Date:
(11) APPROVED BY:
DATE:
(12) MISCEL1.ANEOUS :
Reviewed / Approved By:
Date:
Reviewed / Approved By:
Date:
(13) Page 1 of
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ATTACHMENT 4 McGUIRE NUCLEAR STATION RN ENFORCEMENT CONFERENCE DECEMBER 8, 1986 OPERABILITY SUMMARY
(A)
THROUGHOUT THE ENTIRE PERIOD IN QUESTION, OCTOBER, 1985 THROUGH APRIL, 1986, WITH THE TECHNICAL INFORMATION WE
HAD AVAILABLE, WE ALWAYS BELIEVED THE SYSTEM WAS OPERABLE AT THAT POINT IN TIME.
IF THERE WERE QUESTIONS OF OPERABILITY ON THE SYSTEM OR INDIVIDUAL COMPONENTS, THEN WE TESTED, ANALYZED, OR INSPECTED TO ELIMINATE THE QUESTION.
l (B)
IF, DURING THIS PERIOD OF TIME, WE FOUND AN INDIVIDUAL COMPONENT INOPERABLE DUE TO TEST OR INSPECTION, WE DECLARED. l T INOPERABLE AND ENTERED THE APPROPRIATE TECH SPEC LCO.
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(C)
WE DID FIND A FEW SMALL HX'S PLUGGED SIGNIFICANTLY ENOUGH TO REMOVE THEM FROM SERVICE AND IMMEDIATELY CLEAN THEM.
OUR EFFORTS AT THAT POINT WERE TO QUICKLY TEST OR INSPECT AND CLEAN ALL THE RN HX'S.
ABSOLUTE OPERABILITY ON ALL THESE HX'S COULD NOT BE PROVED OR DISPROVED AT THAT POINT IN TIME.
PERFORMANCE TESTING AND NORMAL OPERATION OF THESE COMPONENTS DID NOT SHOW AN OVERHEATING PROBLEM.
THIS WOULD INDICATE THAT THESE HEAT EXCHANGERS ARE SIZED CONSERVATIVELY.
(D)
IN HINDSIGHT, WE FEEL THAT THE OVERALL SYSTEM WAS
" TECHNICALLY INOPERABLE" PRIOR TO THE CORRECTIVE ACTION PROGRAM BECAUSE OF THE LOW PUMP NPSH CONDITION WITH A LOCA ON ONE UNIT AND A B.O.
ON THE OTHER UNIT.
(IT IS NOTEWORTHY THAT THE PROBLEM WAS NOT CAUSED BY FOULING.)
WE DO FEEL, HOWEVER, THAT ALTHOUGH THE PUMPS DID CAVITATE UNDER THOSE CONDITIONS THAT THEY WOULD HAVE OPERATED'FOR A CONSIDERABLE LENGTH OF TIME, CERTAINLY LONG ENOUGH FOR OPERATOR CORRECTIVE ACTION TO TAKE PLACE.
- -. -
.