ML20212C794
| ML20212C794 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/29/1987 |
| From: | Zech L NRC COMMISSION (OCM) |
| To: | Markey E HOUSE OF REP., ENERGY & COMMERCE |
| Shared Package | |
| ML20209D110 | List:
|
| References | |
| NUDOCS 8703030866 | |
| Download: ML20212C794 (28) | |
Text
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UNITE 3 STATES NUCLEAR REGULATORY COMMISSION Y
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\\*,*****sl wAsHlWTON, D. C. 20666 3
f CHAIRMAN January 29, 1987 The Honorable Edward J. Markev Committee on Energy and Commerce U.S. House of Representatives Washington, D.C.
20515
Dear Concressman Markey:
I am responding to your letter of December 18, 1986 concerning the recent tragic accident at the Surry Power Station.
The NRC has dispatched an augmented inspection team to the site to review what occurred during the accident and to determine its Cause.
Answers to most of your questions are enclosed.
The answer to Question 8a includes confidential and proprietary information and is being submitted in a separate letter.
Answers to parts of Questions 2 and 6 and to all of Question 12 will be submitted as soon as possible.
Sincerely, b
O a0&da 6v.
\\,
Lando W.
h, r,
Enclosure:
As Stated 0703Oh0066070226 0 mms NRCC HESPONDENCE PDR P
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7 AUESTION 1.
PROVIDE THE SUBCOMMITTEE WITH A COMPLETE DESCRIPTION OF THE ACCIDENT.
INCLUDE IN THIS DESCRIPTION A DETAILED CHRONOLOGY OF. EVENTS AND THE FEDERAL, STATE, AND LOCAL OFFICIALS WHO WERE CONTACTED AND BY WHOM.
ANSWER.
THE ANSWER TO THIS OUESTION IS PRESENTED IN ATTACHMENT 1 TO THIS ENCLOSURE.
F QUESTION 2.
P(EASEPROVIDEANOPERATIONALHISTORYOFTHESURRY FACILITY.
INCLUDE IN THIS HISTORY A CHRONOLOGY OF ALL NOTICES OF UNUSUAL EVENTS AND ALL ENFORCEMENT ACTIONS TAKEN AGAINST THE FACILITY.
ANSWER.
ON JUNE 25, 1968, CONSTRUCTION PERMITS WERE ISSUED FOR UNITS 1 AND 2 AT THE SURRY POWER STATION.
AN OPER/ TING LICENSE FOR UNIT l'WAS GRANTED ON MAY 25, 1972, AND COMMERCIAL OPERATION WAS ACHIEVED ON DECEMBER 22 0F THAT YEAR.
FOR UNIT 2, AN OPERATING LICENSE WAS GRANTED ON JANUARY 29, 1973, AND COMMERCIAL OPERATION WAS ACHIEVED ON MAY 1 0F THAT YEAR.
TO DATE, THE MAJOR REPAIR EFFORT AT SURRY HAS BEEN REPLACEMENT OF THE STEAM GENERATORS IN BOTH UNITS.
THAT WORK WAS STARTED IN EARLY 1979 AND WAS C0HPLETED IN MID-1980.
A HISTORY OF ESCALATED ENFORCEMENT ACTIONS IS PRESENTED IN ATTACHMENT 7.
A MORE DETAILED OPERATING HISTORY, INCLUDING A HISTORY OF UNUSUAL EVENTS, WILL BE PROVIDED LATER.
QUESTION 3.
HOW MANY PEOPLE HAVE BEEN INJURED ONSITE.AT SURRY?
PROVIDE THE SUBCOMMITTEE WITH A LIST OF EACH ACCIDENT AT THE SURRY FACILITY WHERE A WORKER, EITHER AN EMPLOYEE OR A CONTRACTOR, HAS BEEN INJURED AND HOSPITALIZED, GIVING THE DATE OF THE ACCIDENT, THE NATURE OF THE INJURY AND THE CAUSE OF THE ACCIDENT.
ANSWER.
UNTIL FEBRUARY 29, 1980, WHEN 10 CFR 50.72 WAS INCORPORATE 0 IN THE REGULATIONS, LICENSEES WERE NOT REQUIRED TO PEPORT TO-NRC INJURIES l
OR FATALITIES OF PERSONS ONSITE.
CURRENTLY 10 CFR 50.72 REQUIRES 1
REPORTING OF ANY EVENT REQUIRING TRANSPORT OF A CONTAMINATED PERSON TO AN OFFSITE MEDICAL FACILITY FOR TREATMENT OR ANY EVENT RELATED TO SAFETY OF ONSITE PERSONNEL FOR WHICH A NEWS RELEASE IS PLANNED OR NOTIFICATION TO OTHER GOVERNMENT AGENCIES WILL RE MADE.
SINCE FEBRUARY 29, 1980 SUCH REPORTS HAVE BEEN RECEIVED BY TELEPHONE IN THE NRC'S OPERATIONS CENTER.
EARLY REPORTS WERE NOT COMPUTERIZED AND RETRIEVAL OF SPECIFIC RIPORTS IS DIFFICULT.
SINCE AUGUST, 1982, 50.72 REPORTS HAVE BEEN STORED IN COMPUTERS.
THAT DATA BASE HAS BEEN SEARCHED FOR REPORTS OF INJURIES.
THE DATA BASE CONTAINS TWO REPORTS (ATTACHMENT 10).
ONE IS FOR THE ACCI9ENT ON DECEMBER 9, 1986 AND THE OTHER IS FOR AN INJURY SUSTAINED ON APRIL 1, 1983.
c
4 5
QUESTION 4 HOW MANY PEOPLE HAVE DIED OF INJURIES RECEIVED ONSITE AT SURRY?
PROVIDE THE SUBCOMMITTEE WITH'A LIST OF EACH ACCIDENT AT THE SURRY FACILITY WHERE '
A WORKER, EITHER AN EMPLOYEE OR A CONTRACTOR, HAS BEEN KILLED, GIVING-THE DATE AND CAUSE OF THE ACCIDENT ANSWER.
A TOTAL OF NINE PERSONS HAVE DIED FROM INJURIES RECEIVED AT SURRY
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STATION AS A RESULT OF THE FOLLOWING ACCIDENTS:
(A)
STEAM VALVE MALFUNCTION IN 1972.
TWO TECHNICIANS WERE WORKING IN THE VALVE ROOM, NEAR THE SAFETY RELIEF VALVES WHEN THEY OPENED.
THE ROOM WAS FILLED WITH STEAM AND THE TECHNI-CIANS WERE FATALLY BURNED.
(B)
STEAM PIPE BURST ON OCTOBER 15, 1983.
A TRAINEE WAS OPERATING A HIGH PRESSURE VALVE i
WHEN RUPTURE OF A 24-INCH PIPE OCCURRED.
THE TRAINEE WAS STRUCK BY 360 DEGREE F STEAM ANO WAS PRONOUNCED DEAD AT THE SCENE.
OUESTION 4 (CONTINUED) (C)
ELECTROCUTIONS ON AUGUST 10, 1984.
A CONTRACT WORKER WAS ELECTROCUTED WHEN HE DRILLED INTO A 4160 VOLT AC STATION TRANSFORMER LINE.
DURING RESCUE OPERATIONS, A SECOND CONTRACT WORKER CAME IN CONTACT WITH THE DRILL WHICH WAS STILL IMBEDDED IN THE 4160 VOLT AC LINE AND WAS ALSO ELECTROCUTED.
(D)
PIPE BURST ON DECEMBER 9, 1986.
DETAILS ARE IN THE RESPONSE TO QUESTION 1.
TO DATE, FOUR FATALITIES HAVE OCCURRED AS A RESULT OF BURNS.
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QUESTIGN 5.
HOW MANY PEOPLE HAVE DIED AT ALL OPERATING NUCLEAR POWER PLANTS SINCE 1972?
PROVIDE THE SUBCOMMITTEE WITH A LIST OF EACH ACCIDENT, THE OPERATING PLANT, THE DATE AND CAUSE OF THE ACCIDENT.
ANSWER.
FOR THE PERIOD FROM THE 1972 TO DATE, THE STAFF IS AWARE OF ACCIDENTS AT OPERATING HUCLEAR POWER PLANTS IN WHICH 31 PERSONS HAVE SUSTAINED INJURIES THAT HAVE RESULTED IN DEATH.
THESE ACCIDENTS AND THE OPERATING PLANTS AT WHICH THE INJURIES OCCURRED ARE LISTED IN ATTACHMENT 2.
THE TABLE WAS CONSTRUCTED BY THE STAFF WITH ASSISTANCE FROM THE OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION TO WHOM DEATHS AND INJURIES IN INDUSTRIAL ACCI-DENTS ARE REGULARLY REPORTED.
00ESTION 6.
PROVIDE THE SUBCOMMITTEE WITH A RECENT UPDATE ON THE FINDINGS OF THE AUGMENTED INSPECTION TEAM'S INVESTI-GATION OF THE ACCIDENT.
IN ADDITION, PROVIDE THE SUBCOMMITTEE WITH A DRAFT COPY OF THIS REPORT ~AS SOON AS IT HAS BEEN COMPLETED.
ANSWER.
A RECENT UPDATE OF THE AIT'S FINDINGS IS PRESENTED IN ATTACHMENT 3.
THE AIT PLANS TO ISSUE ITS REPORT BY FEBRilARY 6, 1987, A COPY WILL BE PROVIDED TO THE SUBCOMMITTEE.
l
. QUESTION 7.
IS THE COMMISSION AWARE OF A STEAM PIPE RUPTURING AT ANY OTHER NUCLEAR FACILITY THAT HAS CAUSED SERIOUS INJURY, FATALITY, OR INTERFERED WITH THE OPERATION OF THE FACILITY.-
IF YES, PROVIDE THE SUBCOMMITTEE WITH A LIST OF EACH OCCURRENCE.
' ANSWER.
NRC IS NOT AWARE OF ANY EVENT AT ANY OTHER NUCLEAR FACILITY-INVOLVING RUPTURE OF STEAM, FEEDWATER, OR CONDENSATE PIPING WHICH CAUSED FATALITIES.
NRC IS AWARE OF AN EVENT-AT TROJAN ON MARCH 9, 1985, WHICH RESULTED IN SERIOUS BURNS TO ONE PERSON THAT REQUIRED HOSPITALIZATION FOR THREE WEEKS.
THE RUPTURE OCCURRED IN AN ERODED SECTION OF A 4-INCH HEATER DRAIN PIPE.
FOR THE PERIOD FROM THE BEGINNING OF 1976 THROUGH EARLY. 1984, 40 LEAKS OR RUPTURES OF STEAM AND FEEDWATER PIPING OCCURRED IN OPERATING NUCLEAR FACILITIES WHICH, TO VARYING EXTENT, INTERFERED i
WITH OPERATION.
THESE EVENTS ARE LISTED IN TABLES A4.1 AND A4.?
0F A STUDY ON EROSION (ATTACHMENT 4) WHICH WAS PREPARED BY NRC'S OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA.-
IN ADDITION TO THESE EVENTS, NRC IS AWARE OF AN EVENT THAT OCCURRED ON-MARCH 16, 1985, AT HADDAM NECK.
A BREAK OF 1/2 INCH BY 2-1/4 INCHES IN THE WALL OF AN'8-INCH.FEEDWATER HEATER LINE WAS CAUSED BY EROSION.
QUESTION 8.
WAS THE COMMISSION MADE AWARE OF STEAM PIPE RUPTURING AT ANY GENERATING STATION, EITHER C0AL-OR OIL-FIRED, IN THE PAST TWO YEARS?
IF THE ANSWER IS YES, PLEASE ANSWER THE FOLLOWING OUESTIONS:
ANSWER.
NRC IS AWARE OF PIPE RUPTURE INCIDENTS AT THE M0HAVE GENERATING STATION AND AT THE MONR0E POWER PLANT.
OUESTION 8A.
PROVIDE THE SU8 COMMITTEE WITH THE DETAILS OF THE INCIDENT.
COMPARE THE RUPTURE IN THAT INCIDENT WITH THE RUPTURE OF THE PIPE AT THE SURRY PLANT.
WAS THE RUPTURE ALONG THE WALL 0F THE PIPE?
WHAT WAS THE ORIGINAL THICKNESS OF THE PIPE?
NHAT WAS THE PRES-SURE OF THE STEAM?
ANSWER THE RESPONSE TO THIS QUESTION IS BASED ON INFORMATION TAKEN FROM VARIOUS DOCUMENTS FOR SURRY INCLUDING INFORMATION NOTICE 86-106, FROM SECY-85-275 FOR THE M0HAVE STATION, AND FROM A ORAFT DOCUMENT PREPARED BY THE NRC PIPING REVIEW COMMITTEE FOR THE MONR0E PLANT.
THE INFORMATION CONCERNING THE M0HAVE STATION IS PROPRIETARY, AND
QUESTION 8.
(CONTINUED)
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THE INFORMATION CONCERNING THE MONR0E PLANT WAS PROVIDED TO THE NRC WITH THE UNDERSTANDING THAT IT WOULD NOT BE PUBLICLY DISCLOSED WITHOUT PRIOR NOTICE.
THE RESPONSE TO THIS QUESTION IS BEING PROVIDED UNDER SEPARATE COVER.
QUESTION 88.
DID THE COMMISSION NOTIFY LICENSEES OF THE POTENTIAL DANGER OF A SIMILAR ACCIDENT OCCURRING AT A NUCLEAR FACILITY?
IF YES, PROVIDE THE SUBCOMMITTEE WITH DOCUMENTATION.
IF NO, WHY NOT?
ANSWER.
NRC DID NOT NOTIFY LICENSEES OF THE EVENTS AT M0HAVE AND MONROE BECAUSE OF THE DIFFERENCE IN CONSTRUCTION OF THE PIPE COMPONENTS, IN THE MATERIALS OF CONSTRUCTION, IN OPERATING CONDITIONS, AND IN PROBABLE PHENOMENA CAUSING FAILURE.
QUESTION 8C.
HAS THE COMMISSION NOTIFIED LICENSEES REGARDING THE INTEGRITY OF PIPES CARRYING HIGH-PRESSURE STEAM OR WATER?
PROVIDE THE SURCOMMITTEE WITH ALL OFFICE OF INSPECTION AND ENFORCEMENT PRELIMINARY NOTIFICA-TIONS OR INFORMATION NOTICES AND BULLETINS RELATING TO PIPE INSPECTION, PIPE EXAMINATION, AND INTEGRITY.
QUESTION 8.
(CONTINUED) ANSWER.
YES, AN ATOMIC ENERGY COMMISSION LETTER TO ALL LICENSEES DATED DECEMBER 18, 1972, ADDRESSED THE COMMISSION'S CONCERN FOR POSTULA-TED FAILURE OF HIGH ENERGY PIPING OUTSIDE OF CONTAINMENT AND THE POTENTIAL FOR DAMAGE.TO SAFETY-RELATED SYSTEMS.
VIRGINIA ELECTRIC AND POWER COMPANY'S RESPONSE FOR THE SURRY POWER STATION IS CON-TAINED IN ATTACHMENT 8.
INFORMATION NOTICES AND BULLETINS FOR 1983 TO PRESENT, AND PRELIMINARY NOTIFICATIONS FOR 1984 TO PRESENT THAT RELATE TO PIPE INSPECTION, EXAMINATION, AND INTEGRITY, ARE PRESENTED IN ATTACHMENT 11.
ATTACHMENT 11 ALSO INCLUDES A PRE-LIMINARY NOTIFICATION ISSUED IN 1983 WHICH DESCRIBES A FATAL ACCIDENT AT SURRY.
INFORMATION NOTICES AND BULLETINS ARE ADDRESSED TO LICENSEES FOR INFORMATION OR ACTION, AS APPROPRIATE.
PRELIMINARY NOTICES ARE IISED WITHIN NRC TO EXPEDITE THE FLOW OF INFORMATION
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00ESTION 9.
THIS ACCIDENT APPEAR (S) TO BE HIGHLY llNUSUAL SINCE THE PIPE-DID NOT RUPTURE AT THE WELD RUT ~IT HAS BEEN REPORTED THAT THE WALL OF THE PIPE WAS RUPTURED.
ANSWER.
ROUGHLY 15% OF THE FRACTURE OCCURRED CLOSE TO A CIRCUMFERENTIAL WELD AS SHOWN IN ATTACHMENT 5.
QUESTION 9A.
DO THE COMMISSION'S INSPECTIONS INCLUDE THE INTEGRITY OF THE WALL OF A PIPE CARRYING WATER OR STEAM, RADI0 ACTIVE OR NON-RADI0 ACTIVE?
ANSWER.
IN THE APPLICATION FOR A CONSTRUCTION PERMIT, LICENSEES COMMIT TO CONSTRUCTING PIPING SYSTEMS IN ACCORDANCE WITH INDUSTRY CODES AND STANDARDS AND NRC REGULATIONS.
THE INDIVIDUAL PIPING COMPONENTS AND COMPLETED SYSTEMS ARE REQUIRED TO BE HYDR 0 STATICALLY TESTED TO PRESSURES EXCEEDING THE DESIGN PRESSURE.
THE CODE ALSO RE-QUIRES THAT DESIGNERS PROVIDE AN ALLOWANCE FOR CORROSION.
THESE HYDROSTATIC TESTS PROVIDE A PHYSICAL TEST OF THE PIPING SYSTEMS INCLUDING WELDS, FITTINGS, AND PIPE INTEGRITY.
HYDROSTATIC TESTS ARE REVIEWED, INSPECTED, AND MONITORE9 BY NRC INSPECTION l
l
OVESTION 9.
(CONTINUED) PRIOR TO LICENSING.
NRC INSPECTION PROGRAMS, HOWEVER, DO NOT INCLUDE DIRECT EXAMINATION OR MEASUREMENT BY NRC INSPECTORS OF THE INTEGRITY OR THICKNESS OF THE WALLS OF PIPES CARRYING WATER OR STEAM WHICH IS EITHER RADI0 ACTIVE OR NON-RADI0 ACTIVE.
QUESTION 9B.
DOES THE COMMISSION'S PROBABILISTIC RISK ASSESSMENT EXAMINE THE POSSIRILITY OF THE WALL OF A PIPE RUPTURING?
IF YES, PROVIDE THE SUBCOMMITTEE WITH THAT ANALYSIS.
ANSWER.
YES, NRC PROBA8ILISTIC RISK ASSESSMENTS (PRA) CONSIDER THE PROBABILITY OF PIPE RUPTURE AND THE CONSEQUENCES OF THAT KIND OF AN EVENT.
THE PRA FOR AN OPERATING POWER REACTOR IS EXTENSIVE AND DOCUMENTATION OF THE PRA REQUIRES MANY VOLUMES.
A DISCUSSION OF PIPE RUPTURE AND ITS RELATIONSHIP TO PRA IS PROVIDED IN ATTACH-MENT 6.
FURTHER, LICENSEES ARE REQUIRED TO ANALYZE THE CONSEQUEN-CES OF HIGH ENERGY PIPING FAILURES OUTSIDE OF CONTAINMENT INDEPEN-DENT OF THEIR PERCEIVED LIKELIHOOD.
A DESCRIPTION OF THE ANALYSIS FOR SURRY FROM THE FINAL SAFETY ANALYSIS REPORT IS PROVIDED IN ATTACHMENTS 8 and 9.
QUESTION 10.
DESCRIBE THE EXISTING NRC REGULATIONS ABOUT PIPES CARRYING RADI0 ACTIVE WATER AND STEAM AND NON-RADI0 ACTIVE WATER AND STEAM.
INCLUDE IN THIS DESCRIPTION ANY REQUIREMENTS FOR INSPECTIONS OF EACH OF THESE CLASS OF PIPES.
ANSWER.
GENERAL DESIGN CRITERION 1 (GDC-1) IN 10 CFR 50, APPENDIX A, OUALITY STANDARDS AND RECORDS, STATES THAT STRUCTURES, SYSTEMS, AND COMPONENTS IMPORTANT TO SAFETY SHALL BE DESIGNED, FABRICATED, ERECTED, AND TESTED TO QUALITY STANDARDS COMMENSURATE WITH THE IMPORTANCE OF THE SAFETY FUNCTIONS TO BE PERFORMED.
IN ORDER TO PROVIDE GUIDANCE IN THE IMPLEMENTATION OF GOC-1 THE COMMISSION IN REGULATORY GUIDE 1.26, " QUALITY GROUP CLASSIFICATION AND STANDARDS," DESCRIBES AN ACCEPTABLE METHOD FOR DETERMINING QUALITY STANDARDS FOR QUALITY GROUP B, C, AND D, WATER-AND STEAM-CONTAINING COMPONENTS IMPORTANT TO SAFETY OF WATER-COOLED NUCLEAR PONER PLANTS.
THE INITIAL PORTION OF THIS CLASSIFICATION SYSTEMS IS DESCRIBED IN 10 CFP 50.55A, CODES AND STANDARDS, WHICH REQUIRES THAT COMPONENTS OF THE REACTOR COOLANT PRESSURE BOUNDARY BE DESIGNED, FABRICATED, ERECTED, AND TESTED TO THE HIGHEST AVAILABLE NATIONAL STANDARDS; THIS CORRESPONDS TO THE QUALITY STANDARD REQUIRED FOR QUALITY GROUP A 0F THE NRC SYSTEM.
IN ADDITION, REGULATORY GUIDE 1.143 PROVIDES DESIGN GUIDANCE FOR RADI0 ACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND
QUESTION 10.
(CONTINUED) COMPONENTS INSTALLED IN LIGHT-WATER COOLED NUCLEAR POWER PLANTS.
THE IMPLEMENTATION OF GDC-1 AND R.G. 1.26 IS DESCRIBED IN STANDARD REVIEW PLAN 3.2.2, SYSTEM QUALITY GROUP CLASSIFICATION.
10 CFR 50.55A ENDORSES SECTION XI 0F THE ASME BOILER AND PRESSURE VESSEL CODE WHICH REQUIRES LICENSEES TO PERIODICALLY EXAMINE ON A SAMPLING BASIS THE INTEGRITY OF WELDS.AND THE HEAT-AFFECTED ZONES IN REACTOR COOLANT SYSTEM PIPING AND CONNECTED PIPING AND IN MAIN STEAM AND FEEDWATER PIPING OUT TO THE OUTERMOST CONTAINMENT ISOLATION VALVE.
NORMALLY, THE SAMPLE INCLUDES 15 TO 25% OF THE WELDS AND HEAT-AFFECTED ZONES DEPENDING ON THE DIAMETER OF THE PIPING WITH LARGER DIAMETER PIPING RECEIVING MORE INSPECTION.
GENERALLY, EXAMINATION OF THIS SAMPLE IS REPEATED ONCE EVERY 10 YEARS.
BASED ON THE RESULTS OBTAINED AND ON NRC'S REGULATIONS OR THE ASME PIPING AND PRESSURE VESSEL CODE, LICENSEES MAY BE REQUIRED TO PERFORM SUPPLEMENTAL OR ADDITIONAL INSPECTIONS.
THE REACTOR COOLANT SYSTEM CONTAINS RADI0 ACTIVE WATER AND, FOR dOILING WATER REACTORS, STEAM.
THE MAIN STEAM AND FEEDWATER PIPING CONTAIN NON-RADI0 ACTIVE STEAM AND WATER.
THE PIPE RUPTURE AT SURRY OCCURRED BEYOND THE OUTERMOST ISOLATION VALVE OF THE FEEDWATER SY3 TEM IN AN AREA WHERE EXAMINATIONS WERE NOT REQUIRED.
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QUESTION 10.
(CONTINUED) GENERAL DESIGN CRITERION 4 IN 10 CFR 50, APPENDIX A, ENVIRONMENTAL AND MISSILE DESIGN BASES, SETS FORTH REQUIREMENTS THAT ADDRESS PROTECTION AGAINST THE CONSEQUENCES OF PIPING FAILURES.
THE IMPLEMENTATION OF GDC-4 IS DESCRIBED IN SRP 3.6.2, DETERMINATION OF RUPTURE LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED.WITH THE POSTULATED RUPTURE OF' PIPING.
IN ACCORDANCE WITH THIS SRP BREAKS WOULD NORMALLY BE POSTULATED IN SAFETY-RELATED SYSTEMS EXCEPT WHERE A BREAK IN A NONSAFETY-RELATED SYSTEM COULD IMPACT A SAFETY-RELATED COMPONENT.
THE BREAK AT SURRY DOES NOT FALL WITHIN THE SCOPE OF THESE CRITERIA.
OUESTION 11.
HOW DOES THE NRC VIEW THE RECORD OF'THE SURRY FACILITY?
ARE THE NUMBER OF ACCIDENTS, FATAL AND NEAR FATAL, AB0VE OR BELOW AVERAGE?
DOES THE COMMISSION VIEW THIS RECORD OF ACCIDENTS AS ACCEPTABLE?
DOES THE COMMISSION PLAN A SPECIAL EFFORT TO ARREST THIS ALARMING PATTERN?
ANSWER.
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BASED ON OSHA DATA, THE TOTAL NUMBER OF FATALITIES AT THE SURRY POWER PLANT EXCEEDS THOSE AT ANY OTHER NUCLEAR POWER PLANT..THE NRC IS CONCERNED ABOUT ANY EVENT AT A PLANT THAT RESULTS IN A FATALITY OR SERIOUS INJURY.
THE FATALITIES AT SURRY WERE NON-NUCLEAR IN NATURE AND ARE CONSIDERED TO BE INDUSTRIAL TYPE ACCI-DENTS WHICH ARE REPORTED TO AND REVIEWED BY THE OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION.
HOWEVER, THE NRC IS CONCERNED DVER THE OCCURRENCE OF NON-NUCLEAR ACCIDENTS AT SURRY AND THE IMPLICATIONS OF THESE ACCIDENTS ON NUCLEAR-RELATED OPERATIONS AND WORKER SAFETY.
IN RESPONSE TO THE DECEMBER 1986 STEAM PIPE RUPTURE, THE NRC DISPATCHED A SPECIAL AUGMENTED INSPECTION TEAM TO REVIEW THE CIRCUMSTANCES AND SE0llENCE OF EVENTS OF THE FATAL STEAM PIPE RUPTURE AT SURRY AND THE NRC IS STUDYING THE ACCIDENT TO IDENTIFY THE LESSONS TO BE LEARNED FROM THIS ACCIDENT.
IN THE INTERIM, THE NRC ISSUED AN INFORMATION NOTICE, IN-86-106, FEEDWATER LINE BREAK, TO LICENSEES PROVIDING EARLY RESULTS OF THE REVIEW 0F THE ACCIDENT.
QUESTION 12.
PROVIDE THE SUBCOMMITTEE WITH ALL CORRESPONDENCE BETWEEN THE NRC AND VIRGINIA POWER REGARDING THE PIPING SYSTEM AND STEAM-RELATED ACCIDENTS AT SURRY SINCE 1972.
IN ADDITION, PROVIDE THE. SUBCOMMITTEE WITH ALL CORRESPONDENCE BETWEEN NRC AND VIRGINIA POWER AFTER THE REPORT OF THE ACCIDENT.
ANSWER.
THIS INFORMATION WILL BE PROVIDED AS SOON AS POSSIBLE.
1
. QUESTION 13.
ON JULY 23, 1986, I WROTE'TO THF COMMISSION AND REQUESTED AN OI INVESTIGATION INTO ALLEGATIONS CONCERNING THE FALSIFICATION OF WELDING CERTIFICATION PAPERS FOR WELDERS WORKING FOR POWER PLANT SPECIALIST, INC.
I RECEIVE 0 AN INTERIM RESPONSE INDICATING THAT AN INVESTIGATION HAD BEEN UNDERTAKEN.
PLEASE PROVIDE THE SU9 COMMITTEE WITH THE STATUS OF THIS INVESTIGATION AND A COPY OF THE DRAFT REPORT AND ANY OTHER DOCUMENTS GENERATED FROM THIS INVESTIGATION.
ANSWER.
THE OFFICE OF INVESTIGATIONS (OI) HAS BEEN CONDUCTING AN INVESTIGATION CONCERNING AN ALLEGATION OF FALSIFICATION OF WELDING CERTIFICATIONS FOR WELDERS WORKING FOR POWER PLANT SPECIALISTS, INC.
THE OFFICE IS CURRENTLY REVIEWING THE INVESTIGATIVE DATA COLLECTED TO DATE AND IS IN THE PROCESS OF EVALUATING THE NEED F0P ADDITIONAL CLARIFICATION OF REGULATORY REQUIREMENTS.
A COMPLETE DRAFT REPORT THEREFORE HAS NOT AS YET BEEN COMPILED.
01 PROVIDED A PERSONAL BRIEFING TO YOUR STAFF CONCERNING THE STATUS OF THE INVESTIGATION ON JANUARY 12, 1987.
ATTACHMENTS:
1.
DESCRIPTION OF THE SURRY 2 ACCIDENT 2.
FATALITIES AT 0PERATING NUCLEAR POWER PLANTS-3.
PRELIMINARY METALLURGICAL EVALUATION 4
EROSION IN NUCLEAR POWER PLANTS 5.
ISOMETRIC DRAWINGS OF FAILED PLANTS 6.
PIPE FAILURES AND PRA 7.
ENFORCEMENT HISTORY FOR SURRY 8.
EFFECTS OF PIPING SYSTEM BREAKS OUTSIDE CONTAINMENT
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9.
LOSS OF NORMAL FEEDWATER 10.
REPORTABLE EVENT NUMBER 07140' 11.
INFORMATION NOTICES, BULLETINS, AND PRELIstNARY NOTIFICATIONS i
4 DESCRIPTION OF THE SURRY 2 ACCIDENT 4
THE INFORMATION PROVIDED BELOW IS PRELIMINARY AND IS BASED UPON THE AUGMENTED INSPECTION TEAM'S INITIAL REVIEW CONDUCTED DECEMBER 9 THROUGH 12, 1986. A FINAL INSPECTION REPORT WILL BE ISSUED BY FEBRUARY 6, 1987.
OVERVIEW OF EVENT 4
ON DECEMBER 9, 1986, WITH BOTH UNITS OPERATING AT 100 PERCENT POWER, A UNIT 2 REACTOR TRIP FOLLOWED BY A MAIN FEEDWATER (MFW) LINE RUPTURE OCCURRED. UNIT 2 HAD COMPLETED A REFUELING OUTAGE AND RETURNED TO FULL POWER OPERATION ON DECEMBER 8, 1986.
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A LOW-LOW S/G LEVEL IN THE C STEAM GENERATOR (S/G) CAUSED A REACTOR TRIP AND
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START OF THE TWO MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS.
THE CONTROL ROOM OPERATORS NOTED THE S/G CODE SAFETY VALVES LIFTING AND REGULATED S/G PRESSURE THROUGH THE ATMOSPHERIC DUMP VALVES. APPR0XIMATELY 30 SECONDS AFTER THE TRIP THE UNIT'S ELECTRICAL BUSSES AUT0-TRANSFERRED TO 0FFSITE POWER. APPR0XIMATELY FIVE SECONDS LATER, A SMALL STEAM RELEASE NOISE WAS HEARD FOLLOWED BY A VERY LOUD NOISE.
A SHIFT SUPERVISOR WHO WAS IN THE TURBINE BUILDING NOTED A LARGE STEAM BREAK i
AND WENT TO THE CONTROL ROOM TO ADVISE THE CONTROL ROOM WATCH OF THE BREAK.
l ALL SECONDARY PUMPS WERE SECURED AND THE BREAK ISOLATED. WATER TO THE S/GS WAS SUPPLIED BY THE AUXILIARY FEEDWATER SYSTEM.
THE PRIMARY SYSTEM RESPONDED NORMALLY TO THE LOSS OF LOAD TRANSIENT.
PRIMARY COOLANT TEMPERATURE, PRIMARY PRESSURE, AND PRESSURIZER LEVEL WERE STABILIZED IN THE DESIRED BAND.
A NOTIFICATION OF UNUSUAL EVENT WAS DECLARED AT 1427 HOURS AND WAS UPGRADED TO AN ALERT IN ORDER TO ENSURE THAT STATION PERSONNEL ACCOUNTABILITY WAS EFFEC-TIVELY ACCOMPLISHED.
THE 18 INCH SUCTION LINE TO A MAIN FEEDWATER PUMP WAS FOUND TO HAVE RUPTURED AT THE ELB0W WHERE THE LINE CONNECTS TO THE 24 INCH CONDENSATE HEATER.
IN ADDITION, STATION HALON AND CARD 0X FIRE PROTECTION SYSTEMS ACTUATED BECAUSE I
0F WATER SHORT CIRCUITING CONTROL SYSTEMS IN THE AREA.
CONTROL ROOM HABITABILI-TY WAS A CONCERN PRIOR TO INITIATING CONTROL ROOM VENTILATION BECAUSE DOORS WERE BLOCKED OPEN TO ALLOW BETTER CONTROL ROOM ACCESS WITHOUT RECOGNIZING THAT I
E CARBON DIOXIDE HAD BEEN DISCHARGED IN THE AREAS ABOVE.
THE CO2 WAS APPARENTLY COMING INTO THE CONTROL ROOM FROM THE HALLWAY.
THE EMERGENCY WAS TERMINATED AT 1623 HOURS AFTER ACCOUNTABILITY HAD BEEN ESTABLISHED.
EIGHT INDIVIDUALS WERE INJURED DUE TO THE STEAM AND WATER.
FOUR OF THE l
INJURED SUBSEQUENTLY DIED. TWO 0F THE INJURED WERE TREATED AND RELEASED.
THE UNIT WAS PLACED IN COLD SHUTDOWN AT 0703 HOURS ON DECEMBER 10, 1986.
i SEQUENCE OF EVENTS INITIAL PLANT CONDITIONS 1
THE ONLY TWD MAJOR MAINTENANCE OR SURVEILLANCE EVOLUTIONS IN PROGRESS WERE:
THE TROUBLESHOOTING OF A B TRAIN UNDERFREQUENCY RELAY FOR A REACTOR COOLANT PUMP (RCP); AND THE TROUBLESHOOTING OF AN AUXILIARY INSTRUMENT AIR COMPRESSOR. THE FIRST ITEM HAD REQUIRED THE RACKING IN AND CLOSING OF THE B REACTOR TRIP BYPASS BREAKER. THE A AND B REACTOR TRIP BREAKERS WERE STILL CLOSED.
THE SECONO ITEM REQUIRED THE SHUTDOWN OF THE RUNNING AUXILIARY INSTRUMENT AIR COMPRESSOR AND THE ATTEMPTED START OF THE NON-RUNNING AUXILIARY INSTRUMENT AIR COMPRESSOR.
INSTRUMENT AIR WAS AT 78 PSIG INSTEAD OF 100 PSIG. SOME MINOR CONSTRUCTION ACTIVITY WAS OCCURRING IN THE VICINITY OF THE MAIN FEEDWATER PUMPS.
THE UNIT'S DATA GATHERING COMPUTER (PRODAC 250) WAS OUT OF SERVICE, BUT REACTOR TRIP INFORMATION WAS AVAILABLE FROM A SEQUENCE OF EVENTS ALARM PRINTER AND A NEWLY INSTALLED EMERGENCY RESPONSE FACILITY COMPUTER (ERFC).
THE ALARM PRINTER UPDATES ON A MILLISECOND BASIS JUST PRIOR TO AND FOLLOWING A REACTOR TRIP, BUT IS LIMITED IN SCOPE. THE ERFC UPDATES IN FIFTEEN SECOND INCREMENTS.
INTERVIEWS WITH THE SHIFT SUPERVISOR AND CONTROL ROOM OPERATORS WERE USED TO CORRELATE TIMES AND TO FILL IN GAPS OF THE EVENT.
PLANT CONDITIONS DURING THE EVENT THE FOLLOWING SEQUENCE OF EVENTS HAS BEEN DEVELOPED BY THE AIT BASED PRIMARILY UPON INFORMATION DERIVED FROM ERFC DATA. THE LICENSEE HAS DEVELOPED A SEQUENCE PRIMARILY BASED UPON THE SEQUENCE OF EVENTS COMPUTER AND CERTAIN TIME MOTION STUDIES OF PERSONNEL MOVEMENTS DURING THE EVENT.
IT SHOULD BE NOTED THAT THE TWD SEQUENCES MAY DIFFER BY APPROXIMATELY 80 SECONDS DUE TO DIFFERENT START TIMES OF THE COMPUTER CLOCKS.
THE AIT WILL CONSIDER THE ADDITIONAL SEQUENCE OF EVENTS INFORMATION IN DEVELOPING ITS FINAL INSPECTION REPORT.
14:20:00-14:21:00 THE FIRST INDICATION OF A PROBLEM OCCURRED AT 2:20 P.M. WHEN THE UNIT 2 CONTROL ROOM RECEIVED AN ANNUNCIATOR ALARM FOR THE B STEAM GENERATOR (S/G)
AS FEEDWATER FLOW WAS LESS THAN STEAM FLOW.
THIS INDICATION AND THE SUBSEQUENT ALARM ON A S/G INDICATED THAT THE C MAIN STEAM TRIP VALVE (MSTV) i HAD SPURIOUSLY CLOSED WHICH CAUSED INCREASED STEAM FLOW IN THE OTHER TWO LINES. l
14:21:(00-:15)
THE CLOSURE OF THE C MSTV CAUSED MAIN FEEDWATER (MFW) PRESSURE DOWNSTREAM OF THE C MFW FLD'W CONTROL VALVE (FCV) TO INCREASE FROM 865 PSIG TO 970 PSIG WITH A AND B PRESSURES OF 845 AND 83S PSIG.
THE OTHER MSTVS CLOSED SHORTLY AFTER DUE TO THE HIGH STEAM FLOW IN THOSE LINES CAUSED BY THE CONTINUING 100 PERCENT DEMAND OF THE MAIN TURBINE.
14:21:(:15 :30)
A LOW-LOW S/G LEVEL ANNUNCIATOR WAS RECEIVED FOR C S/G.
A REACTOR TRIP CAUSED BY LOW-LOW S/G LEVEL IN C S/G OCCURRED AT 14:21:22 (RT 00); THIS CAUSED THE START OF THE TWO MOTOR ORIVEN AUXILIARY FEEDWATER PUMPS AND A MAIN TURBINE TRIP.
AT RT +03 (3 SECONOS AFTER RT) THE CONTROL ROOM OPERATORS (CRO) MANUALLY TRIPPED THE REACTOR.
ONE CONTROL ROD (M-10)
INDICATED THAT IT HAD INSERTED ONLY TO 35 STEPS.
AT RT +04, THE CR0 NOTED THE S/G CODE SAFETY VALVES LIFTING AND TOOK THE S/G POWER OPERATED RELIEF VALVES (PORV) OUT OF MANUAL AND BEGAN TO REGULATE S/G PRESSURE THROUGH THIS ATMOSPHERIC DUMP MODE.
PRESSURE DOWNSTREAM OF THE MFW FCVS DECREASED TO 1008-1028 PSIG.
14:21(:30 -:45)
S/G PRESSURES WERE 1028, 1013, AND 1055 PSIG.
LOW-LOW LEVELS OCLURREd IN THE A AND B S/G WHICH CAUSED THE STEAM INLET VALVE TO THE TURBINE DRIVEN AUXILIARY FEEDWATER PUMP TO OPEN.
14:21(:45-14:22:00)
FOLLOWING THE REACTOR TRIP AND WITH THE PRIMARY TEMPERATURE AT 552 DEGREES F, THE THREE MFW FCVS AUTOMATICALLY CLOSED DUE TO A MAIN FEEDWATER ISOLATION SIGNAL.
THE MFW PUMP RECIRCULATION VALVES (FCV-FW-250A AND 250B) FOR A AND B PUMPS AUTO-0PENED AS REQUIRED.
PRESSURE DOWNSTREAM OF THE MFW FCVS INCREASED ON A TO 1059 AND DECREASED ON B AND C TO 812 AND 949 PSIG.
THE UNIT'S ELECTRICAL BUSSES AUTO-TRANSFERRED TO OFFSITE POWER AT RT +32, WHEN THE MAIN GENERATOR AUTO-TRIPPED ON REVERSE CURRENT, AS NORMAL.
4 FIVE SECONDS LATER AT RT +37, A SMALL STEAM RELEASE WAS SEEN AND HEARD IN THE VICINITY OF THE FIRST POINT HEADER STEAM-SIDE SAFETY RELIEF, WHICH IS i
NEAR THE MFW PUMPS.
l 14:22(:00-15)
PRESSURE BETWEEN THE MFW FCVS AND THE INSIDE CONTAINMENT CHECK VALVES DECREASED TO 445 PSIG. THE PRESSURE AT THE MFW PUMP'S DISCHARGE REACHED A PEAK OF 1290 PSIG..
THE NOISE OF A SMALL STEAM RELEASE WAS FOLLOWED AT APPR0XIMATELY RT +42 BY A VERY LOUD NOISE FROM THE VICINITY OF THE MFW PUMP SUCTION PIPING.
THE PRIMARY SYSTEM RESPONDED NORMALLY TO THE LOSS OF LOAD TRANSIENT.
PRIMARY COOLANT TEMPERATURE WAS STABILIZED AT 520 DEGREES F AND PRESSURIZER LEVEL WAS RECOVERED AS IT REACHED THE LOW LEVEL SETPOINT.
PRIMARY PRESSURE DECREASED FROM 2235 TO 2015 PSIG FOLLOWING THE REACTOR TRIP.
THE PROBABLE TIME FOR PIPING BREAK WAS RT +40-45.
THE STEAM BREAK WAS FOUND TO BE IN THE PIPING AT THE 24 INCH TO 18 INCH REDUCING TEE TOWARD THE SUCTION OF THE A MFW PUMP.
14:22(:15-30)
APPR0XIMATELY 18 SECONDS AFTER THE RUPTURE OCCURRED THE A MFW PUMP TRIPPED, 23 SECONDS LATER THE B MFW PUMP ALSO TRIPPED.
AN OPERATIONS SUPERVISOR WAS IN THE TURBINE BUILDING LOOKING AT THE CONSTRUCTION ACTIVITY AROUND THE MFW PUMPS.
HE NOTED THE LARGE STEAM BREAK, WENT TO THE CONTROL ROOM, AND ADVISED THEM 0F THE BREAK. THE SHIFT SUPERVISOR ORDERED THAT ALL SECONDARY PUMPS BE SECURED.
14:22(:30-45)
THE A MFW PUMP BREAKER INDICATION DID NOT DISPLAY ITS AUTO-0FF DISAGREEMENT LIGHT AS IT SHOULD HAVE. THE B MFW PUMP WAS FOUND AUTO-0FF WITH ITS YELLOW DISAGREEMENT LIGHT ON. THE HIGH PRESSURE HEATER DRAIN PUMP WAS RUNNING AND HAD TO BE TURNED OFF. AFTER ALL SECONDARY PUMPS WERE SECURED, THE NOISE STOPPED. THE ERFC AGREED ~WITH THE OPERATORS.0N THIS TIME FRAME.
14:27 AN UNUSUAL EVENT WAS DECLARED. COMMUNICATIONS WITH THE NRC OPERATIONS CENTER AND NRC REGION II WERE ESTABLISHED.
14:30 THE CR0 CHANGED NORMAL SUCTION OF THE CHARGING PUMPS TO THE REFUELING WATER STORAGE TANK.
14:34 THE CR0 SECURED B RCP TO ALLEVIATE ADDITIONAL HEAT INPUT TO THE PRIMARY.
PLANT CONDITIONS WERE STABLE WITH RCS TEMPERATURE BEING MAINTAINED AT APPR0XIMATELY 250 F WITH THE C SG PORV.. -.
14:40 AN ALERT WAS DECLARED TO ASSIST IN PERSONNEL ACCOUNTABILITY.
14:45 SECURED A RCP.
THE SHIFT SUPERVISOR NOTED THAT THE CONDENSER STILL HAD A VACUUM AND AS THERE WAS NO STEAM FOR THE MAIN TURBINE GLAND SEAL, OPENED THE VACUUM BREAKER.
15:06 THE CR0 SECURED B AUXILIARY MOTOR DRIVEN FEEDWATER PUMP.
15:14 THE CR0 BEGAN EMERGENCY B0 RATION TO COLD SHUTDOWN CONCENTRATION AS PART OF THE NORMAL POST TRIP PROCEDURES.
15:39 THE CR0 SECURED EMERGENCY B0 RATION.
16:25 THE ALERT WAS TERMINATED. ONE CONTROL R0D (M-10) HAD INDICATED THAT IT INSERTED ONLY TO 35 STEPS AND NOW IT WAS NOTED TO INDICATE FULLY INSERTED.
OFFICIALS CONTACTED NRC'S RESPONSE COORDINATION TEAM CONTACTED OFFICIALS AT THE DEPARTMENT OF ENERGY, FEDERAL EMERGENCY MANAGEMENT AGENCY, ENVIRONMENTAL PROTECTION AGENCY, DEPARTMENT OF HEALTH AND HUMAN SERVICES, AND THE DEPARTMENT OF AGRICULTURE.
STATE AND LOCAL OFFICIALS WERE CONTACTED BY THE LICENSEE. -
FATALITIES AT OPERATING NUCLEAR POWER PLANTS PLANT DATE NUMBER CAUSE Arkansas 01/84 1
Drowning in discharge canal Nuclear One Browns Ferry 2 03/28/85 1 Crane hook fell when cable in turbine building parted Crystal River 3 01/10/86 2 Drowning in intake structure Crystal River 04/79 1
Fall Oresden 3 12/85 1
Fall from ladder Dresden 3 06/84 1
Asphyxiation Ginna 02/79 1
Electrocuted by hand tool Hatch 08/02/86 1 Fall from ladder McGuire 2 02/28/85 1 Operator crushed between moving manipulator crane and electrical panel Peach Bottom 09/85 1
Drowning in canal Peach Bottom 05/76 1
Fall Quad Cities 1970s 1
Truck driver's clothing caught in drive shaft Quad Cities 1970s 1
Fall from top of torus Rancho Seco 06/84 2
Auxiliary steam flange disassembled without two valve protection Salem 2 08/80 1
Head injury caused by tools San Onofre 06/85 1
Machine shop accident Nuclear Generating Station San Onofre 09/80 1
Orowning in intake structure Nuclear Generating Station Surry 12/09/86 4 Burns from main feedwater line rupture Surry 08/10/84 2 Electrocutions caused by drilling into a 4160 Volt ac power line Surry 10/15/83 1
Burns from feedwater line rupture Surry 1972 2
Burns sustained when safety relief valves opened Turkey Point 07/85 1
Fall from roof of building Zion 10/85 1
Fall in new radioactive waste building Zion mid 1970 1 Crane accident 31 All plants 1
Asphyxiation 9
Burns from pipe ruptures or opening of valve 5
Drownings 3
Electrocutions 8
Falls 5
Other injuries involving tools and machinery 31
PRELIMINARY METALLURGICAL EVALUATION THE RUPTURE OF THE 18" A MAIN FEEDWATER PUMP (MFP) SUCTION PIPE OCCURRED ON A 90 DEGREE ELB0W AT A POINT ABOUT ONE FOOT FROM WHERE THE SUCTION PIPE JOINS THE CONDENSATE SUPPLY HEADER. THE PIPE FRACTURE WAS 360 DEGREES AND A SECTION ARPROXIMATELY 2'X4' ADJACENT TO THE FRACTURE WAS BLOWN OUT OF THE PIPE. ALL FRACTURE SURFACES APPEAR TO BE DUCTILE IN NATURE. THE LICENSEE HAS MAPPED THE THICKNESS OF THE BLOWN OUT SECTION. THE THICKNESS GENERALLY RANGES FROM.120 INCH TO.330 INCH. THE DRAWING NOMINAL THICKNESS IS.500 INCH.
THE THINNING APPEARS TO BE RELATIVELY UNIFORM EXCEPT FOR SOME SMALL LOCALIZED AREAS. THE THINNEST AREAS ARE LOCALIZED AREAS ABOUT 1/16 INCH IN THICKNESS.
THE THINNING APPEARS TO BE CAUSED BY EROSION / CORROSION WITH POSSIBLY SOME STRAINING.
IT HAS NOT YET BEEN DETERMINED HOW MUCH EACH PHENOMENON CONTRIBUTED TO THE EXTENT OF THE THINNING.
PRELIMINARILY, THE FAILURE APPEARS TO BE CAUSED BY OVERLOAD RECAUSE OF THE
~
THINNED PIPE. THE PRELIMINARY VISUAL INSPECTION OF THE FAILED PARTS INDICATES THAT THE WALL THINNING COULD BE CAUSED BY A CORROSION / EROSION j
MECHANISM. CORROSION PITTING DOES EXIST. THE PIPING CONFIGURATION AND THE CORROSION PATTERN ARE SIMILAR TO THE STEAM GENERATOR'd TUBE (EXCEPT FOR MUCH LARGER DIAMETER) CONFIGUPATION AND CORROSION PATTERN THAT HAS PREVIOUSLY BEEN IDENTIFIED AND WELL DOCUMENTED. THE LICENSEE ANALYSES. DETAILED ABOVE ARE IN PROCESS TO PROVIDE A BETTER UNDERSTANDING OF THE FAILURE MECHANISM.
PORTABLE MICR0 EXAMINATION ON THE SURFACE OF THE FRAGMENT INDICATES NO LINEARIZATION OF THE GRAIN STRUCTURE AT THE SURFACE OF THE METAL. THIS PRELIMINARY EXAMINATION INDICATES THAT THE PIPE SURFACE NEAR THE FRACTURE HAS NOT BEEN HIGHLY STRAINED AS WOULD RE EXPECTED WITH A HIGH STRESS EVENT SUCH AS A HIGH PRESSURE SPIKE.
FURTHER EXAMINATION WILL BE REQUIRED TO CONFIRM THIS.
THE PIPE IN QUESTION IS NOMINALLY 0.500" THICK WITH A CODE ALLOWABLE VALUE OF 0.360".
AT A SYSTEM PRESSURE OF 600 PSI (APPROXIMATELY 150 PSI AB0VE OPERATING SYSTEM PRESSURE) AND TEMPERATURE OF 370 DEGREES F, PRELIMINARY CALCULATIONS INDICATE THAT THE PIPE WOULD RUPTURE AT A WALL THICKNESS OF APPR0XIMATELY 0.090" AND WOULD YIELD AT A WALL THICKNESS OF APPROXIMATELY j
0.173".
THESE RESULTS TEND TO INDICATE TPAT WITH THE PIPE IN ITS DEGRADED CONDITION, SYSTEM PRESSURE WITH PRESSURE TRANSIENTS CAUSED BY NORMAL PLANT OPERATION COULD RUPTURE THE PIPE.
THE LICENSEE PLANS FURTHER ANALYSIS TO DETERMINE THE METALLURGICAL DETAILS OF FAILURE.
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- NUCLEAR REGULATORY COMMISSION UNITED STATES g
WASHINGTON, D. C. 20585
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June 1l. 1984 AEOD/E416 MEMORANDUM FOR: Karl V. Seyfrit, Chief Reactor Operations Analysis Branch Office for Analysis and Evaluation of Operational Data FRCN:
Earl J. Brown, Lead Engineer Engineering Systems Reactoi Operations Analysis Branch Office for Analysis and Evaluation of Operational Data
SUBJECT:
EROSION IN NUCLEAR POWER PLANTS The enclosed Engineering Evaluation Report is forwarded for your infonnation and further consideration.
The report concludes that the data base should only be considered representative of the types of degradation that can occur due to erosion rather than a complete list of events and that there are potential safety issues even though it does not seem feasible to identify a specific safety problem that requires immediate attention.
Areas in which constructive action may be possible are (1) recognize that certain sites or systems appear susceptible to erosion, (2) identification of specific plant equipment and physical configurations that appear susceptible to erosion, and (3) implenentation of monitoring programs to detect degradation of equip-ment (pumps, valves, heat exchanges, and piping).
Some areas appear to have potential safety implications and the report suggests that NRR consider and review the following:
A.
Water Systems 1.
Erosion events appear related to the specific water source with suspended solids (raw water, radwaste, etc.); the use of throttling devices such as valves and orifices, or a combination of the effects of water with suspended solids and a throttling device.
Service water systems appear to be candidates for monitoring to detect degradation.
2.
Erosion of J-tubes in steam generator feedwater headers may warrant consideration for possible monitoring and detection requirements if water hammer relates to a specific safety issue.
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Karl V. Seyfrit 3.
The emergency feedwater system at Ft. St. Vrain has had approximately 25 erosion events and appears worthy of review for possible safety implications.
B.
Steam Systems 1.
The current NRR review of MSIV leakage should be continued until it is satisfactorily resolved.
2.
A review of available data on leakage or rupture of steam piping appears to suggest that erosion could pose a personnel (worker) safety concern in contrast to a plant safety problem.
Some licensees have implemented monitoring programs to detect degraded piping.
It is appropriate to emphasize that a general data base search on erosion miss, as discussed in the report, some safety significant identified in Reference 12.
This suggests that a review of individual events may identify safety concerns that are not apparent from a broad overview such as this report.
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Earl J. Brown, Lead Engineer Engineering Systems
Enclosure:
As Stated 4
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