Text

Summary of 820324 Meeting W/Util,B&W & ORNL in Bethesda,Md Re Pressurized Thermal Shock Issue.Agenda,List of Attendees & Handouts Encl
	ML16162A350
Person / Time
Site:	Oconee
Issue date:	03/31/1982
From:	Vissing G; Office of Nuclear Reactor Regulation
To:	; Office of Nuclear Reactor Regulation
References
	TASK-A-49, TASK-OR NUDOCS 8204150012
	Download: ML16162A350 (59)
	v • d • e

FRE UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 31, 1982 Docket No.

50-269 LICENSEE:

Duke Power Company (DPC)

FACILITY:

OCONEE UNIT NO. 1

SUBJECT:

SUMMARY

OF MEETING WITH DPC CONCERNING THE, PRESSUR L

SHOCK (PTS) ISSUE FOR OCONEE UNIT NO.-l Introduction This meeting was held in Bethesda, Maryland, March 24, 1982, at the request of the NRC staff to discuss DPC's "150 day" response to our letter of August 21, 1981 concerning PTS issue for Oconee 1.

We have previously pro vided the requested participants with an agenda and our concerns related to DPC's "150 day" response (Enclosure 1).

The meeting followed an agenda (Enclosure 2) which deviated from the published agenda. The attendees of the meeting are identified in Enclosure

3.

The material for licensee's presentations and discussions which responded to our concerns are included in Enclosure 4.

Summary of Discussions The licensee's discussions regarding our concerns includes the followin highlights Overcooling Transient Analysis The overcooling transient analysis was based on an integrated and plant specific approach making use of data from the operating history and previous.

analyses such as the Failure Mode Effects Analysis for the ICS.. The transients selected for analysis were not limited to single failures.

Multiple failures were considered.

Analyses were conducted assuming reactor coolant pumps were tripped off and left on.

Existing plant pro cedures provide means to prevent PTS.

The operator has a large number of alarms which indicate an overcooling transient and the mitigating action required by the operator is conventional.

The procedures do not require a time frame. However, the analyses were based upon a time frame that precluded the RCS temperature going below 3000.

The analysis did not take credit for throttling HPI. However, this is an action which the operator would be expected to take.

The model used for the overcooling transient was the RETRAN Code which handles repressurization. The vent valves were not considered in the analysis; however, in actuality they would assist the mixing assumptions.

Of the 13 cases selected, the turbine bypass valve failure, pumps on/off, and main steam-line break, pumps on/off, were the worst cases and were selected for the fracture mechanics analysis.

8204150012 820331 PDR ADOCK 05000269

_PDR

50-269

-2 Small Break Loss of Coolant Analysis (SBLOCA)

The factors considered in the SBLOCA were break size, break location, HPI capacity, decay heat level, steam generator heat sink available, RCP trip, and operator action. The worst case considered was a code safety valve failure to close, maximum HPI flow, RC pump off, operator action to immediately trip RC pumps and throttle HPI to limit core outlet subcooling to 1000 (at 93 min.).

The CRAFT 2 Code, assuming 8.nodes, was the model used. It was indicated that the code has been bench marked by three LOFT tests.

The mixing analysis considered favorably with results of the CREARE one ftests.

The mixing assumption was required since the critical longitudinal weld was not directly under the nozzle.

Fracture Mechanics Methods The Fracture Mechanics Analysis used the data from the thermal hydraulic analyses of the selected transients to determine the EFPY remaining for the critical welds. The analyses was based on a crack arresting within 1/4 wall thickness. Table 8.4-1 of the Oconee 1 report provides the EFPY remaining for each weld for each transient. Warm prestressing (WP) was considered effective only for the SBLOCA transient. Without taking credit for WP, the critical weld would have 16 EFPY remaining.

Operator Action and Training The more severe transients raqu res the operator to act.

For a MSLB the analysis was based on operator taking action (isolate feedwater) within 5 minutes.

No analysis was performed for longer operator time.

It was indicated that this action was conventional and could be expected to take 2 minutes after initiation of the event.

For the SBLOCA the operator was assumed to isolate feedwater at 20 minutes and throttle HPI at 93 minutes.

Throttling of HPI for MSLB was not assumed in the analysis; however, the operator would be expected to do this.

PTS is considered directly in operator training and requalification training.

Operators are examined considering PTS and gains experience in handling events on the simulator. The shift technical advisor-also provides on shift training on PTS.

Future training will be implemented for the new procedures as a result of ATOG development. This will include specific training on PTS. The Oconee site will have a plant specific simulator operational by late 1982.

The licensee's approach is to assure that the operator will take.action at the time required. Analysis were not performed assuming operator did not take action or was delayed in taking action.

5 0- 69 3

Probability Analysis The frequencies of PTS events were determined from fault tree models of sequences considered.

The resultant probabilities represented what was considered to be the major PTS events.

The severe PTS events were in the range of 10 to 10-.

With the conditional probability of vessel failure (a model which has.not beeh developed yet), PTS was not considered a significant contributor to risk.

The Oconee probabilistic analysis presents the first analysis of the kind by any of the eight licensees who received the August 21, 1981 letter concerning PTS.

Conclusion DPC considered that the presentation and presentation material addressed the staff concerns.,The staff noted that the 150 day response did provide a note worthy and commendable probabilityassessment of transients scenerios contributing to PTS.

However, the staff was concerned that there are multiple failure events not yet identified that may be major contributors to the total probability of PTS.

'The staff has contract work underway under the long term program.for looking at the event tree approach to the probability assessment of PTS.

The staff also concluded:

1. The operator action issue is our major concern. The sensiti-vity of transients to time assumed for operator action (i.e., if the operator isolates feedwater at.10 or 20 minutes later than assumed or restarted a RCP later than assumed, what are the resultant P/T transients?).is a major concern.

2.

The staff would be discussing the fracture mechanics.analysis further with B&W to gain an understanding of the model used in the analysis.

3.

The staff would be discussing with B&W the damage function related to the flux within the vessel thickness.

.4. The staff will continue to be concerned about identi'fying the most bounding transient event.

Originai signed Y Guy S. Vissing, Project Manager (PTS)

Operating Reactors Branch #4 Division of Licensing

Enclosures:

1 D

slihed Aganda wvinal4 an erd --n e_'

OFFICE.

2.

Meetin 9 Aqenda 34:.. ticesee' Prnsen eteifMg Attei ile DATEI.....

~.. W/.eneos-ures-S E.eX t**page NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960

ORB#4:DL MEETING

SUMMARY

DISTRIBUTION Licensee: Mr. William 0. Parker, Jr. -

Vice President, Steam Production at Duke Power Company

Copies also sent to those people on service (cc) list for subject plant(s).

HDenton/ECase Docket File DEisenhut NRC PDR GLainas L PDR RVollmer ORB#4 Rdg WHazelton RMattson JStolz TSpeis Project Manager-GVissing, PWagner TMurley Licensing Assistant -RIngram HThompson OELD SHanauer Heltemes, AEOD LShao IE RBernero SShowe (PWR)

TMarsh Meeting Sunnary File-ORB#4 JAustin RFraley, ACRS-10 MVagi ns Program Support Branch DZiemann Clohnson ORAB, Rm.

542 EAbbott BGrimes, DEP RJohnson SSchwartz, DEP..

EGoodwin SRamos, EPDB TNovak FPagano, EPLB JClifford GZech NRC

Participants:

JRoe RWKlecker 7 EIgne RWoods Ve SJBhatt JBuzy CSerpan FManning v-PNRandal 1 BEll iot WJohnston REJohnson CMorris BDLiaw ASpano EDThrom

/LLois ipellet AOxfurth JMazetis

.. DLBasdekas-BClayton TDunni ng FSchroeder CRossi JStrosnider Duke Power Company / NRC/ B&W Meeting Concerning Pressurized Thermal Shock March 24, 1982 P-118 8:30 a.m.

Duke Power / B&W Presentation Introduction Program Overyiw:'*

Systems Analysis

Overcooling Transient Analysis

SBLOCA Transient Analysis Fracture Mechanics Analysis

Thermal Analysis

Material Properties/Fluence LEFM Discussions of Operator Actions/Training

Operator Actions

Current Operator Guidance and Training

Future Training

Future Procedures Discussion of Frequency of Occurrences Summary Break Identification of Future Actions

-w ATTENDANCE LIST FOR MEETING WITH DUKE POWER COMPANY CONCERNING PRESSURIZED THERMAL SHOCK ISSUE FOR OCONEE UNIT 1 MARCH 24, 1982 NRC ORNL B&W IT.W. Klecker RT. Stone

.Th. Morgan S. J. Bhatt L. Redd E. Igne F. Walters F. Manning Duke Power Co.

B. J. Short R. E. Johnson P. M. Abraham P. C. Wagner G. Swindlehurst E.D. Throm K. Canady J0 Mazetil N. Popelm Murphy F. Schroeder S. Lewis P. N. Randall R., Gill C.

orrs

-GPU Nuclear.*

L.

Lois D. 1. Basdekas

-J eeesi-R0. Woods.:

J. Buzy

.B. Elliot B. D. Liaw J. Pellet B. Clayton G. Vissing

~.

Duke Power Company / NRC-/ B&W Meeting Concerning Pressurized Thermal Shock March 24, 1982 P-118 8:30 a.m.

Duke Power / B&W Presentation' Introduction Program Overview Systems' Analysis

Overcooling Transient Analysis

SBLOCATransient Analysis Fracture Mechanics Analysis

Thermal Analysis

Material Properties/Fluence

LEFMi.-

tj-C Discussions of Operator Actions/Training Operator Actions Current Operator Guidance and Training

Future Training

Future Procedures Discussion of Frequency of Occurrences Summary Break Identification of Future Actions

REACTOR VESSEL INTEGRITY CHRONOLOGY OF EVENTS LATE 1960's VERIFICATION OF EFFECT OF COPPER, PHOSPHOROUS ON WELD TOUGHNESS 1973 10 CFR 50 APPENDIX G 1974 SURVEILLANCE PROGRAMS IMPLEMENTED 1976 INITIAL DISCUSSIONS WITHIN B&W AND OWN ER SBOUT REACTOR VESSEL MATERIALS 1977 IMPLEMENTATION OF B&W OWNERS REACTOR VESSEL MATERIALS PROGRAM 1979 THREE MILE ISLAND EVENTI INITIAL REQUEST BY STAFF TO EVALUATE THERMAL SHOCK CONCERN WITH SBLOCA AND EXTENDED HPI COOLING.

1981 BAW-1648 SUBMITTED BY DUKE IN RESPONSE To ACTION PLAN ITEM II.K.2.13 N

INITIAL NRC/INDUSTRY MEETINGS REGARDING

- -,PRESSURIZED THERMAL SHOCK BAW-1511P-REPORT OF VESSEL MATERIALS PROGRAM SUBMITTED OCONEE 1 REACTOR VESSEL ISI UTILIZING RG 1.150 ORNL PHASE I EFFORT ON OCONEE 1 1982 OCONEE 1 PLANT SPECIFIC EVALUATION SUBMITTED OCONEE 2 REACTOR VESSEL ISI

SYSTEMS ANALYSIS

OVERCOOLING TRANSIENT ANALYSIS

SBLOCA TRANSIENT ANALYSIS FRACTURE MEC HAN CSANALYSIS

THERMAL ANALYSIS

MATERIAL PROPERTIES/FLUENCE LEFM

~9 DISCUSSIONS OF OPERATOR ACTIONS/TRAINING

OPERATOR ACTIONS

CURRENT OPERATOR GUIDANCE AND TRAINING

FUTURE TRAINING

FUTURE PROCEDURESm-,.

DISCUSSION OF FREQUENCY OF OCCURRENCES

SUMMARY

LEFM RESULTS FOR THERMAL SHOCK ANALYSES OF OCONEE 1 REACTOR VESSEL NEUTRON FLUENCE.

INSIDE SUFC TRANSIENT LIMITING WELD(S)

N/CM2 EFPY SMUD EVENT SA-1430(HW) 1.09 E 19 25 SBLOCA SA-1229(CW) 9.35 E 18 32 (WPS)

SA-1585(CW) 1q23 E 19 32 (WPS OVERCOOLING SA-1430(LW) 1.09 E 19 25 (CASE 9)

OVERCOOLING TRANSIENT ANALYSES

INTEGRATED AND PLANT SPECIFIC APPROACH

,B9WNSSS.-DES IGN.'CHARACTERIC

TRANSIENT SELECTION MODELING eANALYSISRESULTS

INTEGRATED AND PLANT SPECIFIC APPROACH

PLANT SPECIFIC DESIGN AND SYSTEM PERFORMANCE

THOROUGH REVIEW OF OPERATING HISTORY

IDENTIFICATION OF PRECURSOR EVENTS AND OVERCOOLING f SCENARIOS

EVALUATION OF PLANT PROCEDURES AND OPERATOR ACTIONS

CONSIDERATION OF THE LIKELIHOOD OF ASSUMED FAILURES

REALISTIC ANALYSES BASED ON CONSERVATIVE ASSUMPTIONS od-v o

B&W NSSS DESIGN CHARACTERISTICS

STEAM GENERATOR INVENTORY / PLANT RESPONSE

REACTOR VESSEL INTERNALS VENT VALVES TURBINE BYPASS SYSTEM CAPACITY--,-',

-7~~~.

-7 7 ~ ~~~~

41.

7-

MODELING RETRANO2-MODOO1.

NODALIZATION UTILIZED BENCHMARKING EFFORTS PHENOMENA OF INTEREST

mixING

NON EQUILIBRIUM PRESSURI ZER 0 SYSTEM REPRESSURIZATION

VOIDING

LOSS OF NATURAL CIRCULATION HEAT TRANSFER

VENT VALVES-.7 BOUNDARY CONDITIONS FOR LEFM ANALYSIS

MINIMUM COLD LEG NOZZLE TEMPERATURE."Ir.

REACTOR VESSEL DOWNCOMER PRESSURE

TRANSIENT SELECTION THREE CATEGORIES OF INITIATING EVENTS

STEAM GENERATOR OVERFEED (3)

TURBINE BYPASS SYSTEM FAILURES (7)

STEAM LINE BREAKS (3)

INITIAL SYSTEM CONDITIONS FOLLOWING REACTOR TRIP FROM FULL POWER FROM HOT SHUTDOWN EQUIPMENT FAILURE ASSUMPT IONS

FAILURES IN CONTROL SYSTEMS

MULTIPLE FAILURES CONSIDERED OPERATOR ACTION

BASED ON CURRENT OPERATING PROCEDURES

CONVENTIONAL RESPONSE MITIGATE OVERCOOLING

REACTOR COOLANT PUMPS ON / OFF PLANT ALARMS ALERT THE OPERATOR TO INITIATE PROMPT AND POSITIVE ACTION' TIMEFRAME SELECTED TO.PRECLUDE EXTENDED COOLDOWN BELOW 300 F

MORE SEVERE TRANSIENTS RESULT IN MORE RAPID AND c CLEARER INDICATIONS WHICH JUSTIFY EARLIER OPERATOR ACTION

OVERCOOLING TRANSIENTS DO NOT INVOLVE A LOSS OF INVENTORY AND ARE EASILY MITIGATED

NO CREDIT FOR OPERATOR THROTTLING THE HPIS TO LIMIT REPRESSURIZATION

ANALYSIS RESULTS FOUR WORST CASES SELECTED FOR LEFM ANALYSIS

TURBINE BYPASS FAI URE W/WO RC PUMPS

STEAM LINE BREAK W/WO RC PUMPS CONSISTENT. RESULTS DETERMINED IMPACT OF:

FAILURE ASSUMPTIONS

REACTOR COOLANT PUMPS ON / OFF

OPERATOR ACTION REALISTIC ANALYSES BASED ON CONSERVATIVE ASSUMPTIONS NO OPERATOR THROTTLING OF HPI -

FULL SYSTEM REPRESSURIZATION OPERATING HISTORY DOES NOT SHOW FREQUENT OR SEVERE' OVERCOOLING EVENTS HAVE OCCURRED:

OVERCOOLING TRANSIENTS ANALYZED ARE MORE SEVERE THAN THOSE THAT HAVE OCCURRED OR ARE EXPECTED TO OCCUR

SBLOCA TRANSIENT SELECTION BREAK SIZE -

LARGE BREAKS RESULT IN COMPLETE DEPRESSURIZING, SMALL BREAKS RESULT IN HIGHER TEMPERATURES. BREAK SIZE, 2

.023 FT,,IS IN THE RANGE THAT RESULTS IN THE COLDEST DOWNCOMER TEMPERATURE WITHOUT COMPLETELY DEPRESSURIZING THE RCS.

BREAK LOCATION - BREAK TOP OF PRESSURIZER OR HOT-LEG-BREA IN COLD LEG IS LESS SEVERE BECAUSE: 1) HPI FLOWS OUT BREAK, AN122NGREASEl VEN VALVE FLOW THROUGH COLD LEG BREAK.

HPI CAPACITY - MAXIMUM HPI FLOW FOR OCONEE CLASS PLANTS IS 3 HPI PUMPS.

DECAY HEAT LEVEL - 400 EFPD WITH 1.0 ANS STANDARD DECAY CURVE SG HEAT SINK AVAILABILITY - SINCE NATURAL CIRCULATION WAS NOT MAINTAINED, S.G. HEAT SINK IS NOT IMPORTANT.

RC PUMP -

PUMPS TRIPPED EARLY TO ACCELERATE.LOSS OF FORCED CIRCULATION, OPERATOR ACTIONS - TRIP RC PUMPS (IMMEDIATELY) AND THROTTLE HPI FLOW TO LIMIT CORE OUTLET SUBCOOLING TO 1000F (AT-93 MINUTES).

SBLOCA TRANSIENT

.023 FT.

BREAK (ONE PZR CODE SAFETY)

LOCATION OF BREAK - TOP OF PZR

INITIAL CONDITIONS - 2568 MWT (100% POWER)

DECAY HEAT -

400 EFPD - 1.0 x ANS HPI TEMPERATURE - 50

-5 7'

FEEDWATER TERMINATED AT TIME ZERO

RC PUMPS TRIPPED AT TIME ZERO

NATURAL CIRCULATION LOST AT - 8 MINUTES

SBLOCA TRANSIENT ANALYSIS CODE USED:

CRAFT 2 - 8 NODE MODEL DEVELOPED FROM EXISTING CRAFT 2 22 NODE MODEL:

REFERENCE - BAW-10092, APRIL 1975.

BENCHMARK:

CRAFT 2-22 NODE MODEL CRAFT 2-8 NODE MODEL LOFT TESTS BAW-1648 L3-1

_ SUBMITTED TO NRC WITH 22 NODE MODEL R

L3-6 AND HAS BEEN SBLOCA TRANSIENT WITH S-07-1OD REVIEWED BY STAFF.

STUCK OPEN PORV THESE BENCHMARK RESULTS HAVE PROVIDED AN ADEQUATELY VERIFIED AND APPROVED CODE FOR USE IN THERMAL SHOCK ANALYSIS.

SPECIFIC ITEMS OF INTEREST:

STRUCTURAL METAL INCLUDED e NO REVERSE HEAT TRANSFER MODELED

PRIMARY/SECONDARY HTC CONSTANT:

5000 B-TU/HR FT2 OF s THERMAL EQUILIBRIUM MODEL:

HOMOGENOUS MIXING o VOIDING IN HOT LEG & HEAT WILL BE CONDENSED BY MODEL s TRANSIENT ANALYZED CONSERVATIVE

CRAFT RESULTS USED IN THE"MIXING ANALYSIS AND FRACTURE MECHANICS ANALYSIS ARE:

a VENT VALVE FLOW VENT VALVE TEMPERATURE e

SYSTEM PRESSURE COLD LEG FLOW

F igure 3-20 RCS PRESSURE VS TEMPERATURE FOR SB.LOCA TRANS IENT 14 2200 A

2000 I

1800 600 1400

9.

B

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MAJOR ASSUMPTIONS 0

DENSITY DRIVEN PLUME CAN BE APPROXIMATED BY MIXING LAYER AND JET.

0 TWO DIMENSIONAL 0

NO WALL INFLUENCE O

INITIAL WIDTH OF JET IS THE DIAMETER OF THE COLD LEG PIPE.

O SINGLE PHASE FLOW

DENSITY DRIVEN PLUME U = Use 2/ 2 e = T e E /2 Us =RtH Rt=

1 4 cop t

14 epC=

14 X T=RH MIXING LAYER U = Uj (1+erf ( )

= 13.5 X

= 7.67 X z

FULLY DEVELOPED JET,-.

U

- Uj 5d 2

cosh AT

=

U AT u

max

REFERENCES

1.

TENNEKES, H., & LUMLEY, J. L.

A FIRST COURSE IN TURBULENCE, MIT PRESS 2

SCHLICHTING H BOUNDARY LAYER THEORY, PERGAMON PRESS 1955 3

GRELLA, J.,

& FAETH, G M.

MEASUREMENTS N -A TWO-DIMENSIONAL THERMAL PLUME ALONG A VERTICAL ADIABATIC WALL" JOURNAL OF,-

FLUID MECHANICS (1975)

VOL 71

4.

ROTHE, P. H., MARSCHER, VD., BLOCK, J. A.

DATA EPORT FLUID AN THERMAL MISINGIN* A'MODEL,,.--

COLD LEG AND DOWNCOMER WITH VENT VALVE FLO EPRI NP -2227 CREARE INCORPORATED DECEMBER 1981

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60,..09

FRACTURE MECHANICS METHODS FLAw AsSUMPTIONS MATERIAL PROPERTIES THERMAL/STREss/LEFM ANAL YS I OF~

MATERIAL PROPERTIES ASME SECTION'III MATERIAL PROPERTIES TEMPERATURE DEPENDENT CARBON STEEL CONDUCTIVITY TEMPERATURE EPENDENT STAINLESS STEEL.

CONDUCTIVITY

- TEMPERATURE DEPENDENT CARBON STEEL SPECIFIC HEAT TEMPERATURE DEPENDENT STAINLESS STEEL SPECIFIC HEAT.

TEMPERATURE EPENDENT FRACTURE TOUGHNESS FLUENCE PER B&W RVSP

-CHEMISTRY PERB& OWNERS GROUP

MATERIAL PROPERTLE ASME SECTION II MATERIAL PROPERTIES TEMPERATURE DEPENDENT CARSON STEEL

CONDUCTIVITY TEMPERATURE DEPENDENT MAINLES TEE CONDUCTIVITY

_ TEMPRATURE EPE DEN SPECIFIC HEAT TEmRATURE DEPENDENT SINLESS TEEL SPacIFIc HEAT.

Tb F

TURE.--oGHNES

..................TEERATURE" DEPENDNRcTETOGES FuXENCE PER B&W EMI STRY PER B&l 01-Es GROUP

THERMAL/STRFSS/LEFM ANALYSIS THERMAL ANALYSIS FINITE ELEMENT CALCULATION.

CLAD ON INNER VESSEL SURFACE ADIABATIC OUTER VESSEL SURFACE DETERMINE THROUGH WALL TEMPERATURE DISTRIBUTION CALCULATION PERFORMED AT:300 SECONDS (MIN) INTERVALS FOR 3 HOURS

.,STRESS ANALYSIS SIF's THERMAL STRESS FOR HOLLOW CYLINDER PRESSURE STRESS FOR PRESSURIZED CYLINDER WITH INTERNAL PRESSURE RESIDUAL STRESS DETERMINED AS A FUNCTION OF WALL THICKNESS 7

Z DEAD-WEIGHT STRESS FOR CIRCUMFERENTIAL WELDS EXCLUDES CONSIDERATION OF THE CLAD EXCEPT FOR THERMAL FRACTURE MECHANICS LINEAR ELASTIC FRACTURE MECHANICS CONSIDERS TEMPERATURE, STRESS AND FLUENCE DISTRIBUTION THROUGH.THE WALL THICKNESS CALCULATES K1, KIC, KIA EVERY 300 SECONDS (MIN)

DETERMINES MINIMUM AND MAXIMUM FLAW INITIATION DEPTHS DETERMINES MAXIMUM FLAW AREST DEPTH ASME CODE UPPER SHELF LIMIT OF 200 KSI /IN

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SBLOCA Tra ls Wall Temperature Profiles for Locates of SA-1229 Weld CLAS 400 e.A o

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& CONTROL PAW ETERS CMPUES tWTN STRESS TN INTENSITYTAL FACTOR CALCULATE RTNDT IPRENT SREADIN TIM PRINT K IC KIA, ~IC IAU CACLTE SUBRUTIN ICRETICLINTP PRESSURE PRINTIN HI STORIE PLOTS

.END

EFPY,

-32 AS STARTING POINT FLAW ORIENTATION

-1)

LONG ITUD INAL 3i..33533Z-

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CtRCUMFERENTIAL

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7 ASPECT RATIO, Al 1/6:--

CHEMISTRY (BAV-1511P)

FLUENCE 'BENCHMARKED-BY SURVEILLNCE PROGRAM ATTENTUATION FACTORS (BAW-1485' 47.u UPPER SHELF TOUGHESS, KIn 200 KSI CRTERIA: ARREST WITHIN1/4 THICKNESS

Fluence in Oconee 1 R.V.

3.0 2,5 2.0 (2.13)

(1.73 Predited baedo

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~

(122),

0.5 Cycle

(.216)

Benchmarked to capsulewithdrawn in Cycle 1.

0 0

-4 8

12 16 20 24 28 32 EFFECTIVE FULL POWER YEARS (EFPR)

PRESSURE VS. TIME FROM SYSTEM ANALYSIS TEMPERATURE VS. TIME FROM F.E. SOLUTION I'-

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-SLOCA SH-12 29 I

11'11

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a r

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=

L

OPERATOR ACTION STEAMLINE BREAK ISOLATE FDW TO OTSG'S TRIP RCP'S IF ES CHANNELS 1 2 INITIATED DUE JO LOW PRESSURE, DETERMINE AFFECTED OTSG FEED UNAFFECTED OTSG RCS ESTABLISHED 500F SUBCOOLED THEN HPI THROTTLED RCP RESTARTED WHEN RCS 500F SUBCOOLED NORMAL COOLDOWN

OPERATOR TRAINING REQ[ALIF.ICATION 01/08/82 THROUGH 02/10/82

- DRAFT ATOG GUIDELINES

- PRESSURIZED THERMAL SHOCK EVALUATION SIMULATOR: TO GAIN EXPERIENCE IN HANDLING EVENTS ONSHIFT: REVIEW OF TRAINING PACKAGE

FUTURE OPERATOR TRAINING REOUALIFICATION 06/30/82 THROUGH 08/04/82 ALL LICENSED OPERATORS TO RECEIVE SPECIFIC TRAINING ON PTS

- SITE SPECIFIC SIMULATOR

- CHECK-OUT IN PROGRESS

- OPERATIONAL LATE 1982

FUTURE PROCEDURES SYMPTOM GUIDELINES 7

ATG 7

7

CHAPTER 9 FREQUENCIES OF PTS EVENTS PURPOSE

ACHIEVE AN UNDERSTANDING O RELATIVE FREQUENCIES OF PTS EVENTS

GAIN PERSPECTIVE ON RELATIONSHIP OF THERMAL SHOCK FREQUENCY TO OVERALL CORE-DAMAGE FREQUENCY AND, CONSEQUENTLY TO RISK SCOPE SMALL-BREAK LOCAs BOTH AS INITIATORS AND RESULTING.

FROM T RAN ENTS

SEVERE OVERCOOLING EVENTS, INCLUDING EXCESSIVE FEEDWATER AND INSUFFICIENT STEAM PRESSURE CONTROL

METHOD OF ANALYSIS

GROSS GROUPING OF SEQUENCE TYPES

CONSTRUCTION AND SOLUTION OF SEQUENC LEVEL FAULT-TREE MODELS;V

QUANTIFICATION OF MINIMAL CUT SETS TO OBTAIN SEQUENCE FREQUENCIES MODELING

SEQUENCE STRUCTURE BASED ON JUDGEMENT REGARDING SEVERE PTS EVENTS

MOST SYSTEM MODELS FROM OCONEE PRA:

SOME SIMPLE NEW MODELS

DATA SOURCES

INITIATING EVENTS INDUSTRY EXPERIENCE --

LOCAs, MSLB, FWLB

INDUSTRY EXPERIENCE (EPRI NP-801 1UPDATED WITH OCONEE EXPERIENCE -- MOST TRANSIENTS

OCONEE EXPERIENCE AND FAULT-TREE EVALUATION OCONEE-SPECIFIC INTITIATORS INCLUDING LOSS OF ICS POWER, LOSS OF INSTRUMENT AIR

BASIC EVENTS ASH-1 00OAND OTHER SOURCES UPDATED WITH OCONEE EXPERIENCE MECHANICAL COMPONENTS IEEE-500 AND WASH-1400 --' ELECTRICAL AND ELECTRONIC COMPONENTS

HUMAN RELIABILITY ANALYSIS

BASED ON OPRA HRA, PERFORMED BY OPRA HRA TEAM

MODELING BASED ON TAXONOMY UNIQUE TO OPRA7 UNAVAILABILITY ERROR

INADVERTENT ACTIONS S*OPERATOR INHIBITS.,

OPERATOR FAILS TO AC e DATA SOURCES

NUREG:CR-1278 FOR BASIC DATA TECHNIQUES FOR HANDLING STRESS LEVELS.

CONTROLROOM REVIEW AND INTERVIEWS WITH PLANT STAFF FOR PERFORMANCE SHAPING FACTORS

WASH-1400 AND IREP FOR COMPARISON OF VALUES OBTAINED

SIMULATOR TRAINING FOR MISDIAGNOSIS

SUMMARY

OF THERMAL SHOCK EVENT FREQUENCIES MEAN FREQUENCY SBLOCA W/ FAILURE TO THROTTLE HPI 415E-4 STUCK-OPEN RELIEF VALVE RESULTING 17E 6 FROM OMF ITH FAILURE TO THROTTLE HPI' STUCK-OPEN RELIEF VALVE: FOLLOW IN 7.5E-6 PRESSURIZATIO DU TO HPP WITH FAILURE TO THROTTLE HPF FAILURE OF PRESSURE CONTROL SYSTEM 6.3E-6 CAUSING CHALLENGE TO SRV, WITH FAILURE TO RECLOSE LARGE MSLB OR-FWLB WITH FAILURE TO OE ISOLATE BROKEN SG OR RESTART RCPs SMALL SLB OR STUCK-OPEN MSRV WITH 22E-4 FAILURE TO ISOLATE SG FAILURE OF TBVs TO RECLOSE, WITH 4.9E-5 FAILURE TO CLOSE BLOCK VALVES OR ISOLATE FEEDWATER SUSTAINED EXCESSIVE MFW FLOW 3.6E-5 EXCESSIVE EFW COMBINED WITH FAILURE 1.3E-6 OF STEAM PRESSURE CONTROL.

EVERE PTS EVNSINRNEOf,1 O17 WIHCONDITIONAL PROBAB ILI TY OF VESSEL FAILURE, NOT, SGIIATCONTRIBUTOR TO RISK

>t:.

t

'.'s-.r

-*NOOUTLIERS REQUIRING REMEDIAL ACTION IDENTIFIED

SUMMARY

-LATSPECIFIC OVERCOOL ING -,TRANS IENT, EVALUATION CONSERVATIVE ASSUMPTIONS OF OPERATOR ACTIONS

CONSERVATIVE MIXING CALCULATIONS:

PLANT SPECIFIC MATERIALS PROPERTIE

CONSERVATIVE POSTULATED FLAW SIZES

CONSERVATIVE CRACK ARREST CRITERIAA

ESTIMATED FREQUENCY OF OCCURRENCE OF SEVERE REACTOR VESSEL THERMAL SHOCK EVENTS IS SMALL

-t

'A C'0p7

ML16162A350

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