ML16162A417

Summary of 830111 Meeting W/B&W Owners Group in Bethesda,Md Re Pressurized Thermal Shock Program

ML16162A417
Person / Time
Site:	Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, 05000000, Crane
Issue date:	01/17/1983
From:	Vissing G Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8301250689
Download: ML16162A417 (89)

Text

January 17, 1983 Docket Nos.

50-313, 50-302, 50-346, 50-312, 50-289, 50-269/270/287

SUMMARY

OF MEETING WITH THE BABCOCK & WILCOX OWNERS GROUP (B&WOG)

CONCERNING THE B&WOG PRESSURIZED THERMAL SHOCK (PTS)

PROGRAM January 11, 1983 Introduction The meeting was held in Bethesda, Maryland on January 11, 1983, at the request of the R&WOG to discuss the subject of the agenda Enclosure 1. The meeting was a followup to the meeting of December 20, 1982, with Mr. Denton to discuss the B&WOG Pressurized Thennal Shock Program and the applicability of the staff's screening criterion to B&W reactor plants. The attendees of the meeting are identifed in Enclosure 2. The material for the B&WOG presentation is included in Enclosure 3.

Discussion The B&WOG indicated that several design features of B&W plants reduces the sensitivity of R&W plants to PTS events. These features included the operation of the vent valves during a transient, the relatively low inventory of the steam-generators and. the low vessel fluences.

The analysis by the COMEX code shows that the vent valve fluid mixes with the HPI in the cold leg to heat the water into the down comer from 50 to 240 during a small break LOCA event.

The staff expressed the concern that this condition has not been verified by experiment or test.

The small inventory of the OTSG results in a fast acting blow dovin during a main steam line break (MSLB) event. Thus, the severity of MSLB event is less for the B&W plant than with the same event for a plant with U tube steam generators.

The vessel fluences are lower because of a large gap between the core and vessel wall.

After submitting RAW 1648 in response to NUREG-0737 II.K.2.13 the B&WOG decided to submit more realistic plant specific PTS analyses. Oconee 1 was the first which provided the ground work and the first probabilistic approach. The TMI-1 report was similar to the Oconee 1 report with more updated discussion concerning the mixing assumptions. The SBLOCA in the region of the size of the code safety valve (.023 sq. ft.) was determined to be the most severe (for PTS concern) transient. This would minimize the vent valve flow and natural circulation would he lost at 8-10'min. into the transient. However, natural circulation within the vessel would be maintained. The operator action would trip the RCP (immediately) and throttle HPI flow at,100 subcooling (93 min. into the transient).

The failure (fail open) of the turbine bypass valves (both OTSGs) was determined to be the most severe overcooling transient.

OF I E I............

SURNAME 9301250689 830117 PDR ADOCK 05000269 DATE*

p PDR NRC FORM 318 (1080) NRCM 0240 OFFICIAL RECORD COPY USso: 1s81--33s-96

B&W Meeting Summary

-2 Plant modification which have been or will be made and which reduces the sensitivity of B&W plants to PTS events include improved ICS/NNI power system reliability (as a result of the 1978 Rancho Seco "light bulb" event and the 1980 CR-3 event),

upgraded auxiliary feedwater systems, anticipatory trips, inproved PORV reliability and improved system to reduce overfeed and excess heat removal transient.

As a requirement of orders resulting from TMI-2 action, a failure made and effects analyses of each plant ICS has been provided the staff.

As a result of the B&WOG Integrated Reactor Vessel Material Surveillance Program initial fluence predictions for B&W vessels has decreased approximately 40%.

Further decreases are expected for the low leakage fuel cycles.

The B&WOG concludes that there is no statistical difference between B&W operating plant events and the total industry events. Therefore the screening criterion should be valid for B&W plants. The original Oconee 1 report did not relate the PRA studies to the cummulative frequency versus T.

New data presented in the meeting and in the Duke Power Company letter dated December 23, 1982 indicate good agreement with the staff's studies. Therefore, Duke concludes that the screening criterion is valid for B&W plants. However, the staff doesn't understand what was done to develop the T vs. frequence conclusions.

Conclusion The B&WOG conclude that the B&W plants operating experience, plant design features, the plant specific evaluations and the Oconee probabilisitic analyses support the contention that the screening criteria is valid for B&W plants.

The staff considered additional meetings are necessary to give understanding of the probabilistic work, of the mixing-phenomen and data concerning the vent valves and the overall completeness of the PTS evaluations related to the B&W plants. The staff would establish priorities and schedules for the additional meetings to resolve the above concerns.

riina signed by.

Guy Vissing, Project Manager Operating Reactors Branch #-4 Division of Licensing

Enclosures:

1. Agenda

2. Attendee List

3. Presentation OFFICEO

.. R,

R.

.p.......

F~G ng;cf SURNAMEb... 4 1/./.83..........................................................

DATE)

_ 1_

NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY.-

USGPO: 1981-335-960

ORB#4:DL MEETING

SUMMARY

DISTRIBUTION Licensee:

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

Docket File NRC PDR L PDR ORB#4 Rdg GLainas JStolz Project Manager -GVissing Licensing Assistant-RIngram OELD

Heltemes, AEOD IE SShowe (PWR) or CThayer (BWR),

IE Meeting Summary File-ORB#4 RFraiey, ACRS-10 Program Support Branch ORAB, Rm. 542 BGrimes, DEP SSchwartz, DEP SRamos, EPDB FPagano, EPLB Meeting Participants Fm.

NRC:

CBerlinger KKniel WJohnston RRantala FSchroeder RWoods DBasdekas LLois DLieno JStrosnider CJohnson EThrom RKlecker JClifford SPawlicki

THERMAL SHOCK ISSUE DISTRIBUTION H. Denton/E. Case D. Eisenhut R. Vollmer W. Hazelton R. Mattson T. Speis T,. Murley H. Thompson L. Shao R. Bernero AEOD E. Igne G. Knighton J. Austin J. Buzy J. Milhoan M. Vagins D. Ziemann D. Garner R. Johnson E. Goodwin T. Novak J. Clifford N. Randall S. Chesnut A. Rubin C. Morris C. Serpan L.. Shotkin A. Spano M. Virgilio T. Dunning C. Rossi S. J. Bhatt R. Senseney A. Thadani S. Isreal NRC PDR Felix Litton ATTENDANCE LIST FOR lEETING WITH B&WOG CONCERNING PTS ISSUE-JAN. 11, 1983 NAME ORGANIZATION C. Berlinger NRC/CPB K. Kniel NRC/GIB E. Davidson FPC W. Johnston NRC/DE G. Lainas NRC/DL R. Rantala NRC/MTEB F. Schroeder NRC/DST R. Woods NRC/DST G. Vissing NRC/DL P. Tremblay ACRS T. Myers TECo J. Bohart B&W B. Short B&W P. Abraham Duke Power L., Gibson Consumers Power Co.

D. Howard A P&L M. Snow AP&L H. Feinroth Gencon Corp.

D. Basdekas NRC/RES L. Lois NRC/CPB D. Lieno NRC/CPB C. Whitmarsh B&W C. Hudson B&W T. Cogburn SMUD R. Gradomski TECo M. Foust TECo C. Hendrix Duke A. Lowe B&W J. Strosnider NRC/RES C. Johnson NRC D. Spond AP&L H. Slager Consumers Power Co.

J. Pegram B&W K. Yoon B&W E. Throm NRC/DSI E. Wallace GPUN R. Gill Duke A. Rochino GPUN J. Delezenski GPUN R. Klecker NRC/MTEB J. Clifford, NRC/PSRB S. Pawlicki NRC/QE L. Conner NRC Calendar R. Ganthner B&W F. Walters B &W AGENDA B&W's OWNERS GROUP MEETING WITH NRC STAFF ON PRESSURIZED THERMAL SHOCK JANUARY 11, 1983

• BACKGROUND & PURPOSE ED WALLACE

• DESIGN FEATURES OF B&W PLANTS FRANK WALTERS e

B&W OG DATA SUPPORTING NRC SCREENING CRITERIA EVALUATION OF B&W OPERATING FRANK WALTERS EXPERIENCE OCONEE PRA BOB GILL PLANT SPECIFIC EVALUATIONS BOB GILL/

JERRY DELEZENSKI

• B&W OG PTS PROGRAM FRANK WALTERS LEE ROCHJINO

SUMMARY

• DISCUSSION AGENDA B&W's OWNERS GROUP MIEETING WITH NRC STAFF ON PRESSURIZED THERMAL SHOCK JANUARY 11, 1983 S BACKGROUND & PURPOSE ED WALLACE

1. DESIGN FEATURES OF B&W PLANTS FRANK WALTERS

• B&W OG DATA SUPPORTING NRC SCREENING CRITERIA EVALUATION OF B&W OPERATING FRANK WALTERS EXPERIENCE OCONEE PRA BOB GILL PLANT SPECIFIC EVALUATIONS BOB GILL/

JERRY DELEZENSKI e

B&W OG PTS PROGRAM FRANK WALTERS LEE ROCHINO e

SUMMARY

• DISCUSSION

PURPOSE OF TODAY's B&W OG -

NRC STAFF MEETING UPDATE NRC STAFF AND MANAGEMENT OF B&W OG APPROACH TO PTS, IMPROVE DIALOGUE BETWEEN NRC STAFF AND B&W OG ON PTS.

q B&W OG POSITION

SUMMARY

SCREENING CRITERIA IS CONSERVATIVELY VALID B&W OG HAS PROGRAMS THAT ASSURE THE CONTINUED INTEGRITY OF THE BW RV AND ARE RESPONSIVE TO STAFF'S CONCERNS

PROGRAMS RELATED TO PTS MARGIN ASSESSMENT PROGRAMS TRANSIENT SELECTION & EVALUATION MATERIALS FLAW CHARACTERIZATION (ISI)

MIXING INVESTIGATIONS FRACTURE MECHANICS RISK REDUCTION PROGRAMS PLANT MODIFICATIONS TRAINING & PROCEDURES FLUENCE REDUCTION TAP

SUMMARY

OF ACTIONS TAKEN BY THE B&W OG e

MATERIAL PROPERTY DEFINITIONS GENERIC & PLANT SPECIFIc,ANALYSES TRAINING a OPERATING PROCEDURES REALISTIC MIXING PROGRAMS ISI/NDE PLANT MODIFICATIONS FLUX REDUCTIONS

REACTOR VESSEL INTEGRITY CHRONOLOGY OF EVENTS LATE

• 1960's VERIFICATION OF EFFECT OF COPPER, PHOSPHOROUS ON WELD TOUGHNESS.

• 1973 10 CFR 50, APPENDICES G AND H; CV USE 50 FT-LB LIMIT ESTABLISHED.

• 1974 SURVEILLANCE PROGRAMS IMPLEMENTED.

• 1976 INITIAL DISCUSSIONS WITHIN B&W AND OWNERS ABOUT REACTOR VESSEL MATERIALS.

• 1977 IMPLEMENTATION OF B8 OWNERS REACTOR VESSEL MATERIALS PROGRAM.

• 1979

• THREE MILE ISLAND EVENT.

INITIAL REQUEST BY STAFF TO EVALUATE'THERMAL SHOCK CONCERN WITH SB LOCA AND EXTENDED HPI COOLING (B&W NSSS ONLY).

NRC REVISED IEB 79-05 RELFECTING COMMENTS BY UTILITIES ABOUT NEED FOR HPI FLOW AND CONCERN FOR RV INTEGRITY.

'INITIAL SB LOCA OPERATOR-GUIDELINES CON TAIN GUIDANCE REFLECTING CONCERN FOR REACTOR VESSEL INTEGRITY.

• 1980 MAY MEETING BETWEFN B&W OWNERS AND NRC STAFF TO DISCUSS PRELIMINARY RESULTS OF BOUNDING CALCULATIONS.

AUGUST DRAFT ATOG DOCUMENTS-REFLECT OPERATOR GUIDANCE FOR MAINTAINING RV INTEGRITY AS WELL AS ADEQUATE CORE COOLING.

SEPT-Nov B&W OWNERS GRUOP INITIATED DISCUSSIONS WITH CREARE IN ATTEMPT TO DEVELOP MIXING DATA AND MIXING CODES.

NUREG-0737, II.K.2.13 ANALYSES FROM CE, WESTINGHOUSE REQUIRED BY 01/01/82.

a 1981 JANUARY BAW-1648 SUBMITTED IN RESPONSE TO ACTION PLAN ITEM II.K,2,13, AS A RESULT OF THIS

ANALYSIS, OWNERS GROUP DETERMINED THAT GENERIC ANALYSES WERE OVERLY CONSERVATIVE

-- PLANS INITIATED FOR PLANT SPECIFIC EVALUATIONS IN A TIMELY MANNER, MARCH

.INITIAL NRC/INDUSTRY MEETINGS REGARDING PRESSURIZED THERMAL SHOCK.

,BAW-1511P -

REPORT OF VESSEL MATERIALS PROGRAM SUBMITTED (PURSUANT TO 10 CFR 50, APPENDIX G).

MAY SUBMITTALS REFLECTING STATUS OF OWNERS GROUP ACTIVITIES.

• 1981 JULY OWNERS GROUP REETING TO DISCUSS PLANS RELATIVE TO PTS--STAFF SEEKS GENERIC LIMITS--OWNERS PLAN PLANT SPECIFIC EVALUATIONS.

AUGUST OCONEE 1 REACTOR VESSEL 10-YEAR ISI UTILIZING R.G. 1.150.

SUMMER-FALL INDUSTRY EFFORT TO DEVELOP MIXING DATA AND MIXING CODES.

SEPTEMBER MEETING PROVIDING OWNERS GROUP PLAN TO RESOLVE PTS ISSUE; PLANT SPECIFIC SUB MITTALS FOR LEAD PLANTS; FOLLOWED BY PLANT SPECIFIC SUBMITTALS FOR REMAINING PLANTS.

OCTOBER ORNL PHASE I EFFORT ON OCONEE 1.

• 1982 JANUARY

-OCONEE 1 PLANT SPECIFIC EVALUATION SUBMITTED.

.ORNL PHASE 11 EFFORT STARTS.

JAN-FEB.

OCONEE 2 REACTOR VESSEL 10-YEAR ISIs MARCH MEETING WITH NRC STAFF TO DISCUSS OCONEE 1 SUBMITTAL. STAFF NOTES OCONEE PROBABILISTIC ANALYSIS ON PTS WAS FIRST OF A KIND.

APRIL SUPPLMENTAL INFORMATION ON TRANSIENT RESPONSES WITH DELAYED OPERATOR ACTIONS

PROVIDED,

o 1982 JUNE

,0CONEE3 REACTOR VESSEL 10-YEAR ISI,

• TMI-1 PLANT SPECIFIC EVALUATION SUBMITTED.

NOVEMBER ANO-1 REACTOR VESSEL 10-YEAR ISI,

DESIGN FEATURES OF B&W: PLANTS LOCATION OF INTERNAL VENT VALVES HEATING OF HPI FLUID IN COLD LEG OTSG SECONDARY SIDE INVENTORY LOW VESSEL FLUENCES

RCS Flow During Total Loss of Feedwater Event With a Small Break in Pressurizer VALVE HOT LEG PIP TO QUENCH TANK COLD LEG PIPE So 40 VEN VPRESSURIZER VENT VALVE CORE SURGE LINE di REACTOR VESSEL DOWNCOMER HPI INJECTION FROM BWST STEAM GENERATOR

Figure 12.

Velocity Profile for SBLOCA Transient VENT VALVE FLOW.

RCSEfLOW ASSUMEO TO BE ZERO

'IIt dS 4-4 SP

.FLOW

-COLO LEG PIPING VO VESSEL DOINCOMER

Figure 13. Temperature Profile for SBLOCA Transient 324 0F VENT VALVE 8S 320Vf 3026f 284F2 248.O 2120.

140.7 230 COD LG P 88HPI NOZZLE SWST TEMPERATURE =50.F VESSEL 00 NCOMER 266.F 302.F

Figure 7-3.

Longitudinal Weld Locations to Azimuthal Fluence Profile OUTLET NOZZLE VESSEL (2) w y

INLET NOZZLE FLUENCE NORMALIZED TO PEAK FLUENCE LOCATION

12.

00 SA-IS2

• l WF-8 WELDWELD

.93.9

.0.92

.84 *7

.70

.71

.7

.69 CORE REGION BaY 7-6 Babcock &Wilcok

BW PLANT OPERATING EXPERIENCE SUPPORT SCREENING CRITERIA FINAL TEMP RT(NDT)

NRC B&W NRC BW TMI-I, II 225 400 209

>360 CR-3 250

> 400 250 305 RANCHO SECO 285 285 295 295 NRC DATA TAKEN FROM TABLE 3,1

REASONS FOR CHANGES IN CHARACTERIZATION OF TMI-II TFINAL 0

COLD LEG RTD, BECAUSE OF LOCATION PROVIDES IMPROPER INFORMATION NRC CONCLUDED THIS IS SECTION 2 OF PTS DOCUMENT RTD PROVIDES USEFUL INFORMATION DURING FORCED FLOW ONLY 0

CORE VOIDING AND STEAM CONDITIONS EXISTED IN UPPER VESSEL DOWNCOMER AND COLD LEG BECAUSE OF VENT VALVE OPENING 0

CONSERVATIVE CONDENSING STRATEFIED FLOW CALCULATIONS PREDICTS FLUID TEMPERATURES OF 400-500aF 0

HPI FLOW WAS MINIMAL AND PRIMARY WAS DECOUPLED FROM SECONDARY REASONS FOR CHANGES IN CHARACTERIZATION OF CR-3 TFINAL 0

POSSIBLE TEMPERATURE AS LOW As 250OF FOR 10-15 MINUTES WHEN ONE LOOP N/C WAS LOST 0

VESSEL NATURAL CIRCULATION CONTINUED THROUGH VENT VALVE FLOW PATH 0

TCOLD READ 510aF BEFORE LOSS OF N/C AND 490 aF IMMEDIATELY AFTER N/C RE-ESTABLISHED 0

CORE OUTLET TFMPERATURE WAS 500 0 F DURING ENTIRE PERIOD OF LOSS OF ONE LOOP NATURAL CIRCULATION

CONCLUSIONS FROM B&W PLANT OPERATING EXPERIENCE 0

BASED ON FREQUENCY OF OCCURRENCE THERE IS NO DIFFERENCE BETWEEN B&W OPERATING EXPERIENCE AND THE REST OF THE INDUSTRY.

O THUS, NRC RECOMMENDED SCREENING CRITERIA IS CONSERVATIVE FOR B&W PLANTS 0

SEVERITY OF PTS PRECURSOR TRANSIENTS HAS BEEN REDUCED SINCE CR-3 EVENT BECAUSE OF OPERATOR TRAINING AND PLANT MODIFICATIONS

OPHR ATI NG EXPER IENCES (90% CONFIDENCE INTERVALS)

ClACK EXTENSION WITHOUT ARREST E1f B&W CONFIDENCE INTERVAL 1201 -

CC W

CONFIDENCE INTERVAL ED CE CONFIDENCE INTERVAL OD 11n-E=ALL CONPIDENCE INTERVAL DM B&W MEAN FREQUENCY

• W MEAN FR UENCY 100 -

A CE MEAN FRE ENCY IV "ALL"MEAN FREQUENCY 930 70 60o 50 -1 1 40 30 200 225 250 275 300 325 350 375 400 CRITICAL RTNDT (F) e4

-7

RC Vessel D/C Temperature 600" TMI-2 O CR-3 O Rancho Seco 500 400 300 Conservative Calculation 200.

I 0

15 30 45 60 75 90 105 3.5 HR TimeMin.

CTJMULATI VE F-REQ UENC Y (R-Y/' 1000)

Coto

-cr r)-I-m rnr rz z

(P

-1~~1t1 rnl Nyrn

PRA STUDY OF OCONEE RV PTS SEQUENCES PERFORMED AS PART OF 150 DAY REPORT (DPC-RS-1001)

TO COMPLEMENT THE DETERMINISTIC ANALYSIS ANALYSIS DOCUMENTED IN SECTION 9 OF REPORT REPORT SUBMITTED TO NRC ON 1-15-82

'DISCUSSED WITH NRC STAFF AT THE 3-24-82 MEETING

PURPOSE OF PTS PRA:

TO OBTAIN UNDERSTANDING OF THE FREQUENCIES OF ALL POSSIBLE PTS ACCIDENT SEQUENCES WHICH POTENTIALLY COMPROMISE RV INTEGRITY APPROACH:

• GROUPED EVENTS SIMILAR IN SYSTEM RESPONSE AND CONSEQUENCES INTO CLASSES -SBLOCA, AND NON-LOCA OVERCOOLING EVENTS

• IDENTIFIED APPLICABLE INITIATING EVENTS, SYSTEM FAILURES, AND OPERATOR ERRORS NECESSARY TO ACHIEVE SEVERE PTS CONDITIONS

• DEVELOPED LOGIC MODELS (FAULT TREES) OF SEQUENCES

• INITIATING EVENTS, SYSTEM LOGIC MODELS, AND BASIC EVENT.DATA DERIVED FROM OCONEE PRA PROGRAM

FEATURES:

• CONSIDERED "PIPE BREAK" TYPE AND TRANSIENT INDUCED SMALL BREAK LOCA 'S

• CONSIDERED OVERCOOLING TRANSIENTS RESULTING FROM STEAM LINE BREAK, STEAM RELIEF VALVE FAILURES, AND EXCESSIVE FEEDWATER

• 9 MAJOR SEQUENCE CATEGORIES CONTAINING SEVERAL THOUSAND CUT SETS QUANTIFIED

• CUMULATIVE FREQUENCY OF SEVERE PTS EVENTS CALCULATED TO BE-5.7 x 10-4/RY (MEAN VALUE)

OTHER FEATURES:

• QUANTIFIED THE FREQUENCIES OF 13

,,SEQUENCES WHOSE TRANSIENT RESPONSE WAS ANALYZED IN DETAIL

• RESULTS WERE NOT PRESENTED IN THE FORM OF "TF vs CUMULATIVE FREQUENCY

TEMPERATURE RANGES AND FREQUENCIES OF PTS ACCIDENT SEQUENCES OBTAINED FROM PRA TECHNIQUES FINAL FREQUENCY CUMULATIVE FREQUENCY VESSEL FLUID (MEAN VALUE (MEAN VALUE TEMPERATURE, OF PER R-Y)

PER R-Y) 200-300 5.1 x 1-4 5.1 x10 300-350 2,4 x 10-2,9 x 10 350-U00 1.3 x 10-2 1,6 x 10-2 400-450 1.6 x 102 3.2 10-2

FREQUENCY BASED ON PRA STUDIES FINAL FLUID TEMPERATURE LEGEND NRC STAFF PRA WEST INGHOUSE PRA CL 10 5*10 5o oo0 150 200 75o 3oo 35o 4oo 450o 5oo TE.MPERATURE.(DEG F)

FIGURE 8-1

OCONEE PRA RESULTS CONFIRM TRANSIENT SELECTION RESULTS COMPARE TO W & STAFF RESULTS RESULTS CONFIRM THAT SCREENING CRITERIA IS CONSERVATIVE & THUS VALID

REACTOR VESSEL PRESSURE BOUNDARY ANALYSiS LINEAR ALLOWABLE.

ASSUMEDELASTIC OPERATING MECHANICS CONDITIONS

PRtS3URIZE 0 THERMA SHOCK AALYTICAA. PROGRAM TRANSeANT ANALYSES HEAT TRANSFER/

VESSEL BAATIMALS FRACTURE MECHANICS MIXING SGLOCA 2

"IESSURE (Ii-INITIATIN4G TEMWtaATUste (tt EVENT k2ftTf RC5 FLOW 1r0 LU40 ANALYSIS HIM FLOW Ws "xIOG I

VESSEL WALL THERMAL GRADIENT LINEAR a

OVERCOOLING 3

M RtL EASCUE-RESULTS INITATIG TANSENTPROPERTIES MECHANICS EVENT TEMPERATURE I0)

ANALYSIS RC SSUW (tt)

FLUEN CE FREOUENCIES OF POSTULATED EVENT SEOUENCES

LEFM RESULTS FOR THERMAL SHOCK ANALYSES OF OCONEE 1 REACTOR VESSEL NEUTRON FLUENCE EOL INSIDE SURFACE TRANSIENT LIMITING WELD(S)

N/cM2 EFPY SMUD EVENT SA-1430(LW) 1.09 E 19 25 SBLOCA SA-1229(CW) 9.35 E 18 32 (WPS)

SA-1585(CW) 1.23 E 19 32 (WPS)

OVERCOOLING SA-1430(LW) 1.09 E 19 25 (CASE 9)

Figure 2.1-1 Qconee Nuclear Station Reactor Coolant System

__W Lag OnMovies Tap so,***

auses"Lag meav w2 asnsas"O Som amage seen msng aem o2 s

2-46

Figure 2.1-2 High Pressure Inlectior System Nozzle Geometry BOT LES PIPE COLD LEG PIPE STEAM GENERATOl PLAN VIEW MISM PRESSURE INJECTION NOZZLE CIF AOZZLV REACTOR VESSEL ELEVATION VIEW 2-47

REACTOR COOLANT TEMPERATURE (OF) 100 3W W

2400-1 702400

1.

With Retor Coont Pu mps off operate in Rgin II only.

2200-

2. With Reacwor Coolant

-2200 m()

onopeate m I

1 or II1.

REGION I EGION II

3.

Wth Reactor Cootenq 2000-reture sheI be

-2000 d iiiiiet ebsraging aefe (W) highest incor

4. Mano nin the Reoo

-10 Cooonet 509F auIso oed mka p nw e over the Srite Frggatr Umot L Wh Rerter Coolent 1000-

^wnvs off. the temer.

.- 1800 atwuo oeat be kept within egonit. mal HPI ptew 1400 Wr W

o a

1200 100D-

-1000 c

0 0

800 UNACCEPTABLE UNACCEPTABLE 00 IWO-

-aco 400

-400 20-

-200 100 200 300 400 o

so 7W REACTOR COOLANT TEMPERATURE (OF)

FIgure 3.1-4

SELECTION OF OVERCOOLING TRANSIENTS

• SBLOCA

• SENSITIVITY CALCULATIONS FROM PREVIOUS GENERIC EFFORT

• PLANT SPECIFIC HPI

• CONSERVATIVE MIXING ASSUMPTIONS

• SECONDARY SIDE OVERCOOLING TRANSIENTS

• REVIEW OF OPERATING EXPERIENCE

• THIRTEEN CASES SELECTED FOR DETAILED THERMAL HYDRAULIC EVALUATION

• CASES CONSIDERED TO BE BOUNDING BASED ON ASSUMPTIONS INCORPORATED INTO EACH ANALYSIS

• UTILIZED PRESENT PLANT DESIGN AND OPERATOR PROCEDURES

• FREQUENCY OF PTS EVENTS

• FREQUENCY OF CASES CONSIDERED

• FREQUENCY OF EVENTS MORE SEVERE THAN THOSE CONSIDERED

SMALL BREAK LOCA MECHANISM COMPLETE LOFW STUCK OPEN PZR SAFETY VALVE HPI INITIATION/RCP TRIP/ESAS INTERRUPTION OF NATURAL CIRCULATION HPI COOLING CONTINUES UNTIL 1000F SUBCOOLED SEVERITY WORST HOT LEG SIDE MECHANISTIC BREAK MORE SEVERE THAN PORV LESS VENT VALVE FLOW THAN COLD LEG BREAK MORE HPI ENTERS DOWNCOMER CONTINUED HPI COOLING PROVIDES THERMAL SHOCK

MIN RCS UNIT DATE TEMP OF COMMENT 1

05/05/73 500 No OVERCOOLING RESULTED 2

01/04/74 422 COOLDOWN RATE 1400 F/HR 2

07/11/74 515 No OVERCOOLING RESULTED 2

09/10/74 NORMAL No OVERCOOLING RESULTED 2

09/17/74 547 No OVERCOOLING RESULTED 2

03/07/75

-540 No OVERCOOLING RESULTED 3

L/30/75 540 No OVERCOOLING RESULTED 3

05/25/75 485 No OVERCOOLING RESULTED 3

06/13/75 510 No OVERCOOLING RESULTED 3

07/13/75 NORMAL No OVERCOOLING RESULTED 1

08/14/76 NORMAL No OVERCOOLING RESULTED 1

12/14/78 500 No OVERCOOLING RESULTED 3

11/10/79 420 COOLDOWN RATE 1150F/HR 2

01/30/80 540 No OVERCOOLING RESULTED 3

03/14/80 546 No OVERCOOLING RESULTED 1

05/04/81 NORMAL No OVERCOOLING RESULTED

Figure 4.2-2 MIXING OCCURRING BELOW VESSEL INLET NOZZLE VENT VALVES VESSEL COLD LEG 5500F 5500F 500F 550OF 5200F 500F 630F Y

4-13

Figure5.3-1 OCONEE I INSIDE SURFACE OF REACTOR VESSEL VELD LOCATIONS x

y z

0 90 180 270 360 11.V. FLANGE 223.5 REF.

36" I.0.

MATING C.F. NOZZLE SURFACE 28" 1.0.-

I C'4 56.9" SA -1135.

166.71911 SA-1073 c

433.992" 202. 971"

.20. 53"1 SA-1 229 SA-1493 -

471. 105".

162. 186" 63.3"1 SA-1585 SA-1430 428.554" 536 4"

Figure 5.3-2 SB LOCA TRANSIENT WALL TEMPERATURE PROFILES FOR LOCATION OF SA-1493 WELD 600 550 500 2al CLAD 400 300 200 15 I

/4 T 1/2T 1.36 1.0 2.0 3.0 5.4 Distance thru vessel wall, inches OUTER SURiACE INNER SURFACE 5-18

Figure5.3-3 SO LOCA TRANSIENT VALL TEVPERATURE PROFILES FOR LOCATION OF SA-1229 tELD Go0 4.'5 A N.

CLAD 400

~he 52 IIN.

300 200 100 1/4T 1/2T I I J

-2.0 3 0.

DNSIDE istance tnru vessel wa l, incnes OUTSIDE SURFACE SURFACE

Figure 5.3-4 CASE 9, WALL TEMPERATURE PROFILE VERSUS TIME FOR OVERCOOLING TRANSIENT 000 24 til0 500 400 45 11lN.

300 CL AD

.200

.2" 1/4T I/ 2T

2. I 4.2 6.3 OUTER INNE R Oitne1huvsslwl, inches SURFACE SURFACE zI--

Table 6.6-1 MATERIAL PROPERTIES USED IN THE LEFM ANALYSIS OF THE OCONEE UNIT 1 REACTOR VESSEL 32 EFPY MATERIAL IDENTIFICATION(3)

CHEMISTRY(1 )

NEUTRON INITI FLUENCE RD (Inside Weld or Heat Cu P

Surface)

Number Type Location

-vt. %

wt.

n/cm2 eF AHR-54 SA508,CL2 Nozzle Belt 0.16

.006 1.97E18

(+60)

Forging SA-1135 circumfer-Nozzle Belt/

1.97E18

(+20) ential weld Upper Shell C2197 SA302B Upper Shell 0.15 0.008 9.35E18

(+40)

SA-1073 longitudi-Upper Shell 7.38E18

(+20) nal weld SA-1229 circuafer-Upper Cir-9.35E18

(+20) ential weld cumferential (61% I.D.)

SA-1493 longitudi-Middle 8.98E18

(+20) nal weld Shell C3278-1 SA-302B Middle 0.12 0.010 1.23E19

(+40)

Shell SA-1585 circumfer-Middle 1.23E19

(+20) ential weld Shell C2800-1 SA-302B Lower 0.11 0.012 1.23E19

(+40)

Shell SA-1430 longitudi-Lower 1.09E19

(+20) nal weld Shell (1) Chemistry per BAW-1511P, October 1980.

(Weld data is proprietary.)

(2) Estimated RTNDT values per BAW-10046A, Rev. 1, March 1976.

(3) Per BAW-1436, September 1977.

1-li

Figure 7.4-1 Location and Identification of aterials Used in Fabrication of Oconee Unit 1 Reactor Pressure Vessel ZV2861 (Nozzle Belt)

SA15261 Outlet SA1494 Notzles Only SAll1 5 0

C2197-2 (Inter mediate Shell)

"N* SAl229 -61% (ID)

APPROX W

25 39% (OD)

C3265-11 LOCATION Upper Shell C3278-1 OF e

ACTIVE FUEL SA1585 C2800-1 Lover shell meni'-

C2800-2

- -W112 122S34VA1 Dutchman

Figure 7.2-3 LONGITUDINAL WELD LOCATIONS TO AZIMUTHAL FLUENCE PROFILE OUTLET NOZZLE x

VESSEL INLET NOZZLE FLUENCE NORMALIZED TO PEAK FLUENCE LOCATION 120 190 SA-1430 WELD 220 SA-1O73 EELD

.3

.95 1LD.92 45' SA-1493 WELD

.70,.

.7

.7.6 CORE REGION

..Y

Table 8.4-1 LEFM RESULTS FOR THERMAL SHOCK ANALYSES OF OCONEE I REACTOR VESSEL TRANSIENT WELD I.D.

SMUD OVERCOOLJING CASE OR HEAT EVENT SBLOCA 9

T1 1

12 AHR 54 32 32 32 32 32 32 SA-1135 32 32 32 32 32 32 C2197-2 32 32 32 32 32 32 SA-1073 30 32 30 32 32 32 SA-1229 32 32(3) 32 32 32 32 SA-1493 30 32 31 32 32 32 C3278-1 32 32 32 32 32 32 SA-1585 32 32(3) 32 32 32 32 C2800-1 32 32 32 32 32 32 SA-1430 25 32 25 32 32 32 Vessel Inlet Nozzle 32 32 32 NA (1) Acceptance criteria is crack-arrest within 1/4T and no credit for WPS, except where otherwise noted.

(2) Refer to Tables 6.6-1 and 6.6-2 for location.

(3) Results using war= prestressing (acceptance criteria is crack arrest within 1/2T).

• Since Case 9 has a larger vessel temperature gradient and lower flaw tip temperature, it is judged that Case 9 bounds Cases 10 and 11.

NA Not analyzeds

RESULTS OF OVERCOOLING TRANSIENTS ANALYSES

• REALISTIC YET STILL CONSERVATIVE RESULTS o TRANSIENTS EVALUATED ARE MORE SEVERE THAN OPERATING EXPERIENCE o TRANSIENTS EVALUATED ARE MORE SEVERE THAN CREDIBLE TRANSIENTS WHICH MIGHT OCCUR IN FUTURE REVIEWED WITH NRC STAFF MARCH 31, 1982 EVALUATION OF SENSITIVITY OF OPERATOR RESPONSE TIME SUBMITTED APRIL 30, 1982

• TEND TO SUPPORT VALIDITY OF STAFF'S SCREENING CRITERIA

SUMMARY

PLANT SPECIFIC.OVERCOOLING TRANSIENT EVALUATION CONSERVATIVE ASSUMPTIONS OF OPERATOR ACTIONS CONSERVATIVE MIXING CALCULATIONS PLANT SPECIFIC MATERIALS PROPERTIES CONSERVATIVE POSTULATED FLAW SIZES CONSERVATIVE CRACK ARREST CRITERIA ESTIMATED FREQUENCY OF OCCURRENCE OF SEVERE REACTOR VESSEL THERMAL SHOCK EVENTS IS SMALL

EVOLUTJON nF THE TMTJ-1 PTS ANALYSES DECISION MADE TO USE Til-1 REPORT AS NEXT STEP IN EVOLUTION OF B&W OG PTS PROGRAM USING:

1. INCREASE IN NUMBER OF TMI.-1 PLANT SPECIFIC PARAMETERS

2.

TRANSIENT SELECTION CRITERIA

• REALISTIC DEFINED MECHANISM OF OCCURRENCE s REASONABLE OCCURRENCE PROBABILITY SINGLE/MULTIPLE FAILUPES

• SEVERE ENOUGH TO CHALLENGE VESSEL

3.

REALISTIC COMMIX IA MIxi Nt.

PLj a BENCHMARKED AGAINST AR FLOW TESTS

4. No WARM PRESTRESS, NO CRACK INITIATION RESULT:

32 EFPY*

RTNDT =35, F*

(CRITICAL LONG WELD)

BASED ON CURRENT 12 MONTH FUEL CYCLE

J1IXING ANALYSES EMPLOYED IN BW/OG THERMAL SHOCK ANALYSIS JET TURBULENT MODEL

- DEVELOPING MIXING LAYER AND FULLY DEVELOPED JET DETERMINE 2-D THERMAL PROFILE IN VESSEL D/C ASSUMED NO MIXING OCCURS IN COLD LEG PROVIDES CONSERVATIVE TEMPERATURE PROFILE AS SEEN IN COMPARISON WITH CREARE DATA FLOW 2-D CODE ANALYSIS OF ACTUAL GEOMETRY AND SB LOCA TRANSIENT IN 2-D GEOliETRIES PROVIDED TEMPERATURE PROFILES IN COLD LEG A D VESSEL DOWJCOMER

- DETERMINED THAT SIGNIFICANT MIIXING OCCURS IN COLD LEG

- BENCHIMARKED AGAINST CREARE 1/5 SCALE MODEL DATA WITH GOOD PREDICTION OF THE I IITED TEMPERATURE DATA

flIXING ANALYSES EMPLOYED IN BW/OG THERMAL SHOCK ANALYSIS (CONTINUED)

COMMIX-1A -

3D T/H CODE

- ANALYSIS OF ACTUAL GEOMETRY IN THREE DIMENSIONS TO PROVIDE TEMPERATURE PROFILES IN COLD LEG AND DOWNCOMER

- DETERMINED THAT SIGNIFICANT MIXING OCCURS IN COLD LEG

- QUALIFIED FOR.PTS EVALUATIONS VIA BENCHMARKS WITH ANALYTICAL SOLUTIONS o BW/ARC MIXING TESTS EPRI EVALUATIONS OF PTS HAVE USED COMMIX

- PRESENTLY BEING MODIFIED BY B&WAND INDUSTRY TO OBTAIN MORE ACCURATE RESULTS CONCLUSIONS JET TURBULENT iODEL - TOO CONSERVATIVE; AS SHOWN BY CREARE DATA AND FLOW-2D AND COMMIX PREDICTIONS FLOW 2-D -

GOOD AGREEMENT WITH CREARE DATA SUPPORTS COMMIX's RESULTS

MIIXING ANALYSES EMPLOYED !N BW/OG THERMAL SHOCK ANALYSIS (CONTINUED)

COMMIX-1A -

PREDICTS RESULTS FOR BWI/ARC TEST DATA WITH GOOD AGREEMENT

- SUPPORTS RESULTS OBTAINED WITH FLOW-2D CODE FLOW-2D AND COMMIX-lA PROVIDE REASONABLE TEMPERATURE PROFILES FOR USE IN PTS ANALYSES

Figure 4-3. Vessel Azimuthal Fluid Temperature Profile Vs Time 550 600 SEC. 500 400 30 SEC 300

-O 65" BELOW 4 OF NOZZLE 135" -BELOW t OF NOZZLE.

200 100 Shell 50 8.9 13.4 19.0 39.0 62.1 0

Incnes From E of InIet Nozzle 0

X

Figure 4-4.

Comparison of REARE 0ne-rFifth Scale Model Data to Turbulent Jet Analysis DATA FROM TEST 24 7

7" UNDER NOZZLE FOR CREARE TEST DATA 27.5" UNDER NOZZLE FROM B&W ANALYSIS

.5 I8.8" UNDER NOZZLE FOR CREARE TEST DATA, 34.5" UNDER NOZZLE FROM B&W ANALYSIS 4

.3

.2 0~

0 0

10 20 30 40 Distance From Nozzle < Incnes 0

0

Figure 4-14. Azimuthal Downcomer Temperature Comparison at 9.2 Inches Below Cold-Leg Centerline for CREARE Test No. 32 155 150 145

0.

140 135 CREARE DATA 130 FLOW-20 125 I

8.0 10 12 14 16 18 20 22 24 26 Reference distance from not-leg center ine, inches 4-24 Babcock & Wilcox J,-

Figure 4-1.

COMmixlA Computational Mesh for TI-Cold Leg and Vessel Downcomer VENT VRC PUMP DISCHARGE U-BAFFLE COLD LEG PIPING VESSEL DOWNCOMER HIGH PRESSURE INJECTION VENT VALVES.

90

Figure 4-2. Location of TMI-1 Vessel Welds in.6, Z Coordinates for SBLOCA Transient

,,-VENT VALVES IN PLANE 1=2 COLD LEG IN-HOT LEG BLOCKAGE WF-70 207.5" WF-8 WF-25 SA 1526

-300 00 600

-44. 7" O"

69.4" Distance From of Cold Leg 4-13

Figure 4-3. Velocity Profile Along J JCOLD LEG PIPNG VESSEL DOWNCOMER J= 6 0

4-14

Figure 4-5.

ARC Mixing Test Section 0.98 REFERENCE WALL STATION 2 STATION 3 STATION 4 V,D STATION 1 2.8" 3.85" T

,D T

.STATION 5 1.84_

V, T VT 3.78" 0.406*

REFERENCE WALL 6.81' STATION 6 V,T STATION 7 V,T V -

VELOCITY MEASUREMENT T -

TEMPERATURE MEASUREMENT D -

DYE INJECTION POINT 4-16

Figure 4-7. COMMIX-1A Temperature Profile -

ARC Test Run 2 DATA CMM IX ISO STATION 7 TEMPERATURES 180 1.70 Lo STATI ON 7 160 150 140 130 120 =%

110 1

100 80 70 50

40.

0.2 0.4 0.6 0.8 1.0 Distance From Reference Wall (Vertical Leg), in.

4-18

SCREENING CRITERIA 2700 (AXIAL) 3000 (CIRC)

(RTNDT)

INDU STRY T3&WNS OPERATING EPERINC EXPERIENCE PLANT SPECIFIC EVALUATIONS OC0, 'TMI -

1.

W& NRC PROBABLISTIC ANALYSIS OCONEE PLANT SPECIFIC PROBABI LISTIC ANALYSIS EVALUATION COMPLETED SUPPORT THE VALIDITY OF THE SCREENING CRITERIA

B&W OWNERS GROUP THERMAL SHOCK PROGRAM GENERIC APPROACH VESSEL BRITTLE FRACTURE INVESTIGATION BEGAN -

JULY 1981, ANALYSIS CONSISTS OF:

SELECTION OF SB LOCA TRANSIENT, SELECTION OF OVERCOOLING TRANSIENT PERFORM MIXING ANALYSIS ON SB LOCA TRANSIENT VESSEL THERMAL ANALYSIS VESSEL MATERIAL PROPERTIES DETERMINATION VESSEL FLUENCE AT WELD LOCATIONS FRACTURE MECHANICS ANALYSIS o LEFM o WPS

FIGURE 1-4 VESSEL BRITTLE FRACTURE INVESTIGATION SB LOCA ACS TRANSIENT PRESSURE ANALYSIS HISTORY OVERCOrCLING REPRESSURI11 ZATION FRACTURE TRANSIENT RCS FLUID VESSEL MECHANIC VESSEL ANALYSIS PRESS/TEMP MIXING C

WALL THERMAL ANALYSIS LIFE HIjS TOR Y ANALYSIS ANALYSI.S (EFPY).

VESSEL IlATERIAL PROPERTIES 0*

0 VESSEL & WELD FLUENCE DETERMINATION C

B&W OWNERS GROUP THERMAL SHOCK PROGRAI BEGAN JULY 1981 - COMPLETED AUGUST 1982

• GENERIC APPROACH

- VESSEL BRITTLE FRACTURE INVESTIGATION, SB LOCA TRANSIENT SELECTION DATA BASE FOR SELECTION - B&W LICENSING EXPERIENCE

- BAW-1628 (1980) ORIGINAL INDUSTRY T/S WORK

- BAW-1648 (1981)

PROVIDED DATA TO CONSTRUCT REASONABLE AND MECHANISTIC TRANSIENT BREAK SIZE -

LARGE BREAKS RESULT IN COMPLETE DEPRESSURIZING.

SiALL BREAKS RESULT IN HIGHER TEMPERATURES. MlOST PROBABLE FAILURE IS PORVI HOWEVER, PZR CODE SAFETY VALVE,.023 FT.21 is IN THE RANGE THAT RESULTS IN COLDEST DOWNCOMER TEMPERATURE

'WITHOUT COMPLETELY DEPRESSURIZING RCS.

BREAK LOCATION - BREAK ON TOP OF PRESSURIZER OR HOT LEGS. BREAK IN COLD LEG IS LESS SEVERE BECAUSE: 1) HPI FLOWS OUT BREAK, AND 2) INCREASED VENT VALVE FLOW THROUGH COLD LEG BREAK.

INITIATING EVENT -

LOSS OF ALL FW - GIVE MECHANISTIC WAY OF INITIATING TRANSIENT WHICH LIFT PZR CODE SAFETIES AND INTERRUPT N/C.

Babcock&

B&W OWNERS GROUP THERMAL SHOCK APPROACH (CONTINUED)

HPI FLOW - MAXIMUM HPI FLOW FOR OCONEE CLASS PLANTS IS 3 HPI

PUMPS, SG HEAT SINK AVAILABILITY - SINCE NATURAL CIRCULATION WAS NOT MAINTAINED, S.G. HEAT SINK IS NOT IMPORTANT, OP ACTIONJ - TRIP RC PUMPS.

- THROTTLE HPI FLOW TO LIMIT CORE OUTLET SUBCOOLING TO ~1000F (~-93 N INUTES).

Babcock &Wilco

.'~~ti~~Jh.

6 CD~vmr 1

onp0

OVERCOOLI:NG TRANSIENT DATA BASE FOR SELECTIO1 ANALYSIS - B&W LICENSING EXPERIENCE CONSUMERS SENSITIVITY STUDY, DENTON's SHOW CAUSE LETTER (50.54(8))

- ANO-1 SYM. SSLB FOR UPGRADED AFW SYSTEM

- SMUD TRANSIENT (3/20/78); LIGHT BULB INCIDENT

- RESPONSE TO I.E. BULLETIN 79-05C, EFFECTS OF TRIPPING RC PUMPS ON NON-LOCA TRANSIENT SMUD FSAR - MSLB WITH TSV FAILURE IN OPPOSITE LOOP OPERATING EXPERIENCE - OCONEE DPC SUBMITTAL TO NRC 1/15/82 (DPC-RS-1001)

- TMI-1 GPUN SUBMITTAL TO NRC 6/7/82 THIS INFORMATION PROVIDED GREATER THAN 60 INITIATING EVENTS OR COMB.

OF I.E./MULTI-FAILURES.

INFORMATION PROVIDED DATA TO CONSTRUCT REASONABLE AND MECHANISTIC TRANSIENT.

BREAK SIZE - FAILURE OF ALL TBV's -28% STEAM FLOW.

- MSLB FOR FPC, AP&L, SMUD.

INITIATING EVENT -

ICS SINGLE FAILURE - MECHAiNISTIC AND REALISTIC WAY OF INITIATING TRANSIENT SINGLE MSLB UPSTREAM OF MSIV's Babcock &Wilcox

OVERCOOLING TRANSIENT (CONTINUED)

AFW SYSTEM -

INITIATES, FILLS AT vAX FLOW UNTIL LEVEL SETPOINT IS REACHED: CONTROLS AT SETPOINT.

OP ACTIO4

- TRIP RC PUMPS NO THROTTLE OF HPI FOR MSLB TRANSIENT

- THROTTLE HPI AND SHUTOFF FOR TMI-1 TBV TRANSIENT 0 THROTTLE OF HPI FOR TBV 0-I TRANSIENT Babcock Wilcor a Ucetmoott cmt-c.

OPERATING PROCEDURES AND TRAINING.

• PLAN

T. PROCEDURE

S BASED ON GENERIC GUIDELINES ARE IN THE PROCESS OF-BEING IMPLEMENTED ON A SCHEDULE CONSISTENT WITH REQUIREMENTS OF SECY 82-111

• CERTAIN ACTIONS HAVE ALREADY BEEN INCORPORATED INTO EXISTING PROCEDURES.

• B&W OWNERS GROUP HAS BEEN WORKING CLOSELY WITH THE NRC STAFF IN THE REVIEW AND APPROVAL OF THE GENERIC GUIDELINES

• GENERIC TRAINING MODULE ON PTS DEVELOPED AND IMPLEMENTED ON UTILITY SPECIFIC SCHEDULES

• NRC PTS AUDIT REVEALED NO SIGNIFICANT PROCEDURAL OR TRAINING DEFICIENCIES

PRESSURIZED THERMAL SHOCK TRAINING OUTLINE REACTOR VESSEL THERMAL SHOCK DESCRIPTION FACTORS AFFECTING REACTOR VESSEL THERMAL SHOCK EFFECTS OF THESE FACTORS OPERATOR ACTIONS SYMPTOM RECOGNITION AND SYMPTOM ORIENTED PROCEDURES

PLANT MODIFICATIONS COMPLETED/UNDERWAY WHICH IMPROVE PLANT RESPONSE TO PTS o

IMPROVED ICS/NNI POWER SUPPLY RELIABILITY (ADDRESSES 1978 R-S AND 1980 CR-3 EVENTS) o UPGRADED Aux. FEEDWATER SYST.EMS o

ANTICIPATORY TRIPs, PORV RELIABILITY IMPROVED 0

IMPROVED SYSTEMS To REDUCE OVERFEED. AND EXCESS HEAT REMOVAL TRANSIENTS

PgW/pwNFR' TO VESE vATEPI.~LS mRG PHASE SHORT TERM PROGRAM TAS KA-BEST ESTMIATE DESIGN CURVE.S TASK B REFINEMENT OF EUTRON FLUENCE TASK C -

CHARACTERIZATION OF CHEMICAL COMPOSITION TASK D INTEGRATED SURVEI-LLANCE PROGRAM TASK E - COMPLIANCE WITH SECTION V, PARAGRAPH E.,

10CFR50 APPENDIX G PHASE II-IRRADIATION PROGRAM -

PREPARATION OF SPECIMENS PHASE II IRRADIATION OF TEST SPECIMENS PHASE IV -EVALUATION OF IRRADIATED MATERIALS DATA PHASE FRACTURE MECHANICS ANALYSIS PHASE VI PLANT SPECIFIC ANALYSIS PHASE VII IN-PLACE ANNEALING PROGRAM PHASE VIII -

EVALUATION OF ATYPICAL WELD METAL PHASE IX -

DESIGN BASIS FLAW SIZE (ENHANCED ISI)

PHASE X -DOSIMETRY

Pr I ~

-r~lo PlI;SrS il-tv PP ASSSIII-IV DEVELOP IIELD IETAL RESOLVE SitOtT FrA\\CTURE TOUGMIESS TERM ISSUE DATA BASE

.PHIASES, IX_ A X RECUCE COurYATIS 'S 4IN DeSIGa CASES DEVELOP ANALYSIS PROCEDURES rrE VII M41 I TOl /EVALUATE DMttSTRATE fttACTOR VESSEL R.Y. A J\\ALING INTEGRITY BY ATU1LYSIS PHASE VI SUCCESSFUL.

REACTOR VESSEL Pg!iTAli CONTIUED A~ttE~l.LICEff5AILITY PAIMEAL ItcsuILi EMPHASIS ON ANALYSIS APPLICATIONS AND PPOCtiES PASED ON STATE-OF-THE-ART DEVELOPMENTS IH INJDUSTRY-WIDE R&D PcXtPAPS.

BRP O\\'ERS GROUP PTS PROGRAM PER ENCLOSURE A (CHAPTER 9) OF SECY-82-465:

ANALYSIS/ACTIONS REQUIRED WHEN RTNDT SCREENING CRITERIA ARE EXCEEDED OR WILL EXCEED WITHIN 3 YRS A. VESSEL MATERIAL PROPERTIES o IMPROVE BASIS FOR INITIAL RTNDT o REFINEMENT OF CHEMISTRY INFO FOR CRITERIA MATERIALS B.

DETERMINISTIC FRACTURE MECHANICS EVALUATION C. FLUX REDUCTION PROGRAMS D.

ISI/NDE PROGRAM F.

INSITU ANNEALING ABOVE ARE ADDRESSED BY ONGOING B&W OG INTEGRATED RV MATERIAL SURVEILLANCE PROGRAM'(IRVMSP)

VESSEL MATERIAL PROPERTIES A.

INITIAL RTNDT DOCUMENTED IN BAW-1006AI MARCH 1976 B.

CHEMISTRY INFO FOR CRITICAL MATERIALS PHASE I OF IRVMSP DOCUMENTED IN.

BAW-1511P, OCTOBER 1980, BAVI-1500 C. VESSEL FLUENCE PHASE I OF IRVMSP DOCUMENTED IN BAW-1511P, OCTOBER 1980, ALSO IN BAW-1485.

PHASE X -

DOSIMETRY -

METHODS TO FURTHER REDUCE UNCERTAINTY D.

DEVELOPMENT OF J-INTEGRAL TEST METHOD E.

DEVELOPMENT OF RTNDT SHIFT CORRELATION

DETERMINI ST CFRACTURE M CH ANICS o

OCONEE-I REPORT o

TMI-1 REPORT ITEMS DISCUSSED INCLUDE:

(SECY-82-465) o VESSEL WALL THICKNESS a CLAD THICKNESS:

VESSEL INNER RADIUS o

LOCATION & ORIENTATION OF THE ASSUMED INITIAL CRACK o

HEAT TRANSFER COEFFICIENT USED MATERIAL PROPERTIES, K, E

vs. TEMPERATURE ASSUMED CRACK SHAPE AT INITIATION 8 TIME OF INITIATION o

CRACK SHAPE AT ARREST o

TREATMENT OF CLADDING INDUCED STRESSES o

UPPER SHELF TOUGHNESS o

BASES FOR THE DETERMINATION OF LIMITING RTNDT (AT THE INNER VESSEL RADIUS)

NOTE:

ADDITIONAL PLANT SPECIFIC REPORTS ON HOLD PENDING RESULTS OF STAFF REVIEW OF TWO REPORTS SUBMITTED

Peak Reactor Vessel Fluence In Oconee 1 3.0 Initial Predictions 2.5 -

(FSAR)

Predicted Based On Owners Group Integrated 2.0 RV Surveillance Program 1.5 30%

20%

1.0/

Low Leakage Fuel 0.5 os Cycle!

Potential Ultra Low Leakage Fuel Cycle

'onversion To Low Leakage Fuel Cycle 0

J 0

4 8

12 16 20 24 28 32 Effective Full Power Years,(EFPY)

.Lq

' UY U u1KY K)U u

U Lm UI Per NRC Procedure AS OF 12/31/81-GUTHRIE + 2(T 32EFPY GUTHRIE + 2 O cwa R-S LWA' cw TMI-I LW I OCONEEcw OCONEE cw LW cw X

ANO-I LW I CR-3 OCONEE 3AVIS-cw ESSE MID-cw LAND-1 MID-LAND-2 FORGING 0

50 100 150 200 250 300.

50

.I END

INSERVICE gSPECTION OF VESSEL BELTL@E REGIONS 6 WITH ARIS (AUTOMATIC REACTOR INSPECTION SYSTEM), 0-I, O-II, o-III, AND ANO-I HAVE COMPLETED THEIR 10-YEAR INSPECTION OF THE REACTOR VESSELS.

• MEETINGS WITH THE NRC STAFF ESTABLISHED THAT THIS PLANNED INSPECTION OF 0-I WOULD SATISFY THE INTENT OF RG 1.150 BEFORE THE TESTING WAS

BEGUN,

• THE NEAR SURFACE OF ALL BELTLINE WELDS WERE THOROUGHLY INSPECTED.

• WITH ARIS, IT WAS DEMONSTRATED WITH A HIGH DEGREE OF CONFIDENCE THAT THE BELTLINE REGION OF THE VESSEL IS FREE OF SIGNIFICANT DEFECTS.

• THE INSPECTION RESULTS FOR 0-1, II, AND III ARE:

NO INDICATION HAD A THRU-WALL DIMENSION GREATER THAN.15 INCHES.

OF 133 INDICATIONS FOUND IN 0-1 VESSEL, THE VAST MAJORITY (114)

WERE CHARACTERIZED AS LAMINAR INDICATIONS, 3 AS PLANER SUBSURFACE INDICATIONS AND 16 AS SLAG INCLUSIONS. II HAD 4 INDICATIONS. III HAD, 1 INDICATION.

ANO-1 HAD 40 INDICATIONS (PRELIMINARY RESULTS - NOTHING SIGNIFICANT).

• ACCEPTANCE CRITERIA, FROM SECTION XI OF ASME CODE, IS THAT ALL INDICATIONS BE LESS THAN 1/40 T OR SUPPLEMENTARY ANALYSIS BE PERFORMED TO DEMONSTRATE THAT SUBSEQUENT OPERATION FOR THAT COMPONENT IS JUSTIFIED.

Babcock &Wilcor, f-

OTHER PROGRAMS SUPPORTED BY OWNERS GROUP o TAP PROGRAM TO FOLLOW AL REACTOR TRIP TRANS ENT o HSST MATERIALS COMMITTEE HAS ONGOING PROGRAM TO FOLLOW THIS WORK o ORNL

-SUPPORTING ONGOING EVALUATION OF 0-1 BY ORN EPR PROVIDED ORIGINAL INPUT FOR CREARE MIXING 1/5 SCALE FACILITY ON B08 PLANTS PROVIDED INFORMATION FOR CREARE 1/2 SCALE MODEL FACILITY EVALUATION OF TMI-1 PTS BY EPRI

B&W OWNERS GROUP PTS PROGRAM PER ENCLOSURE A (CHAPTER.9)

OF SECY-82-,465:

ANALYSIS/ACTIONS REQUIRED WHEN RTNDT SCREENING CRITERIA ARE EXCEEDED OR WILL EXCEED WITHIN 3 YRS A. VESSEL MATERIAL PROPERTIES o IMPROVE BASIS FOR INITIAL RTNDT o REFINEMENT OF CHEMISTRY INFO FOR CRITERIA MATERIALS B. DETERMINISTIC FRACTURE MECHANICS EVALUATION C. FLUX REDUCTION PROGRAMIS P. ISI/NDE PROGRAM F.

INSITU ANNEALING ABOVE ARE ADDRESSED BY ONGOING B&W OG INTEGRATED RV MATERIAL SURVEILLANCE PROGRAM.(IRVMSP)

VESSEL MATERIAL PROPERTIES A.

INITIAL RTNDT DOCUMENTED IN BAW-10046A, MARCH 1976 B, CHEMISTRY INFO FOR CRITICAL MATERIALS PHASE I OF IRVMSP DOCUMENTED IN BAW-1511P, OCTOBER 1980.

BA1-l500 C

VESSEL FLUENCE PHASE I OF IRVMSP DOCUMENTED IN BAW-1511P, OCTOBER 1980, ALSO IN BAW-1485, PHASE X -

DOSIMETRY -

METHODS TO FURTHER REDUCE UNCERTAINTY D. DEVELOPMENT OF J-INTEGRAL TEST METHOD E. DEVELOPMENT OF RTNIDT SHIFT CORRELATION

DETERMINISTIC FRACTURE MECHANICS o

OCONEE-I REPORT o

TMI-1 REPORT ITEMS DISCUSSED INCLUDE:

(SECY-82-65) o VESSEL WALL THICKNESS CLAD THICKNESS:

VESSEL INNER RADIUS o

LOCATION & ORIENTATION OF.THE-ASSUMED INITIAL CRACK o

HEAT TRANSFER COEFFICIENT USED & MATERIAL PROPERTIES, K, Ed, vs. TEMPERATURE o

ASSUMED CRACK SHAPE AT INITIATION 9 TIME OF IN ITIATION o

CRACK SHAPE-AT ARREST o

TREATMENT OF CLADDING INDUCED STRESSES o

UPPER SHELF TOUGHNESS o

BASES FOR THE DETERMINATION OF LIMITING RTNDT (AT THE INNER VESSEL RADIUS)

NOTE:

ADDITIONAL PLANT SPECIFIC REPORTS ON HOLD-PENDING RESULTS OF STAFF REVIEW OF TWO REPORTS SUBMITTED

OTHER PROGRAMS SUPPORTED BY ONERS GROUP o TAP

-PROGRAM TO FOLLOW ALL REACTOR TRIP TRANSIENT o HSST MATERIALS COMMITTEE HAS ONGOING PROGRAM TO FOLLOW THIS WORK o'. ORNL SUPPORTING ONGOING EVALUATION OF 0-I BY ORNL o EPRI PROVIDED ORIGINAL INPUT FOR CREARE MIXING 1/5 SCALE FACILITYON B&W PLANTS PROVIDED INFORMATION FOR CREARE 1/2 SCALE MODEL FACLITY EVALUATION 0 OF TMI-1 PTS BY EPRI

FUTURE ACTIONS SUPPORTING PTS POSITION ON BW PLANTS 0

CONTINUE ASSESSMENT OF EVENTS TO DEMONSTRATE THAT TRANSIENT SELECTION WAS CONSERVATIVE O

COMPLETE PLANT SPECIFIC EVALUATION FOR PLANTS PREDICTED TO REACH SCREENING CRITERIA o

IMPLEMENT PLANT.MODS AS NEEDED,TO RESPOND TO ANY NEW ACTUAL PTS EVENTS 0

CONTINUE -OPERATOR TRAINING AND PROCEDURE UPGRADE 0

CONTINUE RV MATERIALS PROGRAM o

EVALUATE FURTHER FLUENCE REDUCTION MEASURES O

CONDUCT 10-YEAR INSERVICE INSPECTIONS OF REACTOR VESSEL WELDS WITH ENHANCED ISM 1ETHODOLOGY o

CONTINUE TRANSIENT ASSESSMENT PROGRAM o

CONTINUE INVOLVEMENT AND SUPPORT OF INDUSTRY PROGRAMS

CONICLUS IONS o

B&W OWNERS GROUP HAS A QUALITY PROGRAM IN PLACE WHICH IS RESPONSIVE TO THE ITEMS IN CHAPTER 9 OF STAFF'S REPORT.

o SCREENING CRITERIA IS CONSERVATIVELY VALID FOR B&W PLAN,!TS, AMPLE MARGIN ISAVAILABLE BETWEEN SCREENING CPITERIA AND PLANT SPECIFIC RT VALUES.

O FURTHER DETAILED DISCUSSIONS WITH NRC STAFF AND BWOG ARE ESSENTIAL TO COMPLETE THE DEMONSTRATION THAT PTS RISK ON B&W PLANTS IS ACCEPTABLY LOW,