alabama power co position for augmented reactor vessel ... · implementation of the augmented...

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ALABAMA POWER COMPANY

POSITION FOR .

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AUGMENTED REACTOR VESSEL EXAMINATION

PROGRAM

OCTOBER 6, 1983

SUBMITTED BY:

WESTINGHOUSE ELECTRIC CORPORATIONNUCLEAR SERVICES INTEGRATION DIVISIONP, O. B0x 78PITTSBURGH, PA 15230

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ALABAMA POWER COMPANY POSITION,

FOR AN AUGMENTED REACTOR VESSEL

EXAMINATION PROGRAMd

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INTRODUCTION

The following position has been adopted by Alabama Power Company with respectto implementation of an augmented examination program for the J. M. FarleyUnit I and Unit II reactor vessels. The basis for the program is Appendix Ato USNRC Regulatory Guide 1.150, Revision 1 where the Electric Power ResearchInstitute Ad Hoc Comittee recomendations are adopted as an acceptableapproach to the positions recommended in the base document. The augmented

program will be applied to full penetration reactor vessel welds that aresubject to volumetric examination as required by the 1974 Edition of SectionXI of the ASME Boiler and Pressure Vessel Code including Addenda throughSumer 1975 as noted in the Farley license amendment including any additional

amendments and reliefs pertaining to that license. Specifically, the programprovides for augmenting examination requirements for all welds as defined inASME Section XI categories B-A, B-B, B-C, and B-D with most emphasis placed on

I those welds considered important from a neutron embrittlement and pressurizedthermal shock point of view.

Category B-A -- Longitudinal and circumferential shell welds in the vesselbeltline region.

Category B-B -- Longitudinal and circumferential shell welds other than thosein categories B-A and B-C.

|Meridional and circumferential welds in the bottom head and--

closure head. l

Category B-C -- Vessel-to-flange and closure head-to-flange welds.

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| Category B-D -- Nozzle-to-vessel welds and nozzle inside radius.| '

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IMPLEMENTATION OF THE AUGMENTED EXAMINATION PROGRAM.

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The scope of the augmented program which will be implemented for examinations ;

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of the J. M. Farley Unit I and Unit II reactor vessels is described below and,

| summarized in Table I. The Alabama Power Company Technical Position on the

Augmented Reactor Vessel Examination Program is found as Attachment A.i

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Period 1 and 2 Vessel Examinations

During the 40-month and 80-month examination periods, areas of the reactorvessel accessible for examinations include the vessel flange-to-shell weldfrom the seal surface, outlet nozzle-to-shell welds from the nozzle bores,welds in the closure head, outlet nozzle-to-safe end welds and ligamentsaround threaded stud holes in the vessel flange.

Examinations of the vessel flange-to-shell weld from the seal surface and'

outlet nozzle-to-shell welds from the nozzle bores are conducted with angleswhich provide normal or near-normal incidence to the plane of the weld. Thismethod of examination establishes a favorable relationship between theinterrogating sound beam and any planar reflector which might exist parallelto the weld and provides examination coverage of the volumes of material nearthe vessel inside and outside diameter surfaces with no limitations due togating, near field effects, etc. For these examinations, Alabama PowerCompany will augment Section XI requirements in the areas of instrumentperformance checks, calibration, examination, and reporting of results per theTechnical Position, Attachment A. A 50 percent DAC recording level will beused for examinations of these areas.

Welds in the closure head are located sufficiently far from the reactor coreso as to be relatively unaffected by the concerns of neutron damage orpressurized thermal shock. Where access permits, these welds will be manuallyexamined using contact techniques from the outside surface. These techniques

provide coverage of the volumes of material near the vessel head insidediameter surface with no limitations due to gating, near field effects, etc.For examinations of these welds Alabama Power Company will augment Section XI

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requirements in the areas of instrument performance checks, calibration,,

examination, and reporting of results per the Technical Position, AttachmentA. A 50 percent DAC recording level will be used for examinations of these

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areas. ,

. Examinations of the vessel flange ligaments and outlet nozzle-to-safe endwelds fall outside the scope of the augmented examination program and will be

examined per Section XI.

Period 3 Vessel Examination

During the 120-month, or tenth year, examination period, areas of the reactor~ vessel accessible for examination include all those listed for period 1 and 2

examinations and all reactor vessel circumferential and longitudinal shellwelds, inlet nozzle-to-shell welds, bottom head meridional and circumferentialwelds, and inlet nozzle-to-safe end welds.

Welds in the reactor vessel beltline region are associated with neutronembrittlement and pressurized thermal shock considerations. For examinationsof welds in this region Alabama Power Company will augment Section XI

requirements in the areas of instrument performance checks, calibration,examination, recording and sizing, and reporting of results per the TechnicalPosition, Attachment A. Specifically, examinations of the upper-to-intermediate shell weld, the intermediate-to-lower shell weld, theintermediate shell longitudinal welds, and the lower shell longitudinal weldswill be conducted applying the full scope of the augmented examination programincluding a 20 percent DAC recording level for Section XI angle beamexaminations of volumes within the inner 25 percent of the vessel through-wall

thickness as measured from the vessel inside diameter surface.

Vertical and circumferential welds outside the vessel beltline region andwelds in the bottom head are located sufficiently far from the reactor core so

|as to be relatively unaffected by the concerns of neutron damage or

| pressurized thermal shock. Examinations of these welds will be augmented inthe areas of instrument performance checks, calibration, examination, and

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reporting of results per the Technical Position, Attachment A. A 50 percent.

DAC recording level will be used during examinations of these welds. Welds inthe bottom head inaccessible to the remote inspection tool are typicallyexamined manually from the outside surface, where access permits, usingcontact techniques. These techniques provide coverage of the volumes ofmaterial near the head inside diameter surface with no limitations due togating, near field effects, etc. Section XI examinations of these welds willbe augmented in the areas of instrument performance checks, calibration,examination, and reporting of results per the Technical Position, Attachment A.A 50 percent recording level will be used during examinations of these welds.

The reactor vessel inlet nozzle-to-shell welds are important from a structuralintegrity standpoint because of the high stress concentration present.Examinations of these welds are performed from the nozzle bores with angleswhich provide normal or near-normal incidence to the plane of the weld. Thismethod of examination establishes a favorable relationship between theinterrogating sound beam and any planar reflector which might exist parallelto the weld and provides examination coverage of the volumes of material nearthe vessel inside and outside diameter surfaces with no limitations due togating, near field effects, etc. For these examinations, Alabama PowerCompany will augment Section XI requirements in the areas of instrument

performance checks, calibration, examination, and reporting of results per theTechnical Position, Attachment A. A 50 percent DAC recording level will beused for examinations of these areas.

Examinations of inlet nozzle-to-safe end welds fall outside the scope of theaugmented examination program and will be examined per Section XI.

SUPPORTING INFORMATION/ DOCUMENTATION

A review of mechanical properties, specifically the reference nil-ductilitytemperature (RTNDT), for the welds and base materials comprising the FarleyUnit I and Unit II reactor vessel beltline regions indicates that apressurized thermal shock concern should not exist in the beltline region ofeither vessel during their lifetime. These findings are present in AttachmentB. " Pressurized Thermal Shock Considerations".

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Further, Alabama Power Company has adopted a low-leakage fuel loading design.Information relative to this measure is presented in Attachment C, " Effects of*

Low Leakage Fuel Loading Pattern".

Lnplementation of the Alabama Power Company Position for an Augmented Reactor

Vessel Examination Program for the J. M. Farley Unit I and Unit II reactorvessels will provide even further assurance of vessel integrity, especially in

; areas involved in the pressurized thermal shock issue.

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ATTACHMENTS

The following attachments are presented to supplement the Alabama PowerCompany Position for an Augmented Reactor Vessel Examination Program. j

Attachment A - Alabama Power Company Technical Position for anAugmented Reactor Vessel Examination Program

Attachment B - Pressurized Thermal Shock Considerations

Attachment C - Effects of Low Leakage Fuel Loading Pattern

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REFERENCES

1. Section XI of the ASME Boiler and Pressure Vessel Code with Addendathrough Summer 1975.

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2. USNRC Regulatory Guide 1.150, " Ultrasonic Testing of Reactor Vessel WeldsDuring Preservice and Inservice Examinations", Revision 1 with Appendix A;February 1983. .

3. "A Quantitative Methodology for Reactor Yessel Thermal Shock DecisionMaking"; Ackerson, Balkey, et. al.; draft submitted for publication in theInternational Journal of Nuclear Engineering and Design; March 1983.

4. United States Nuclear Regulatory Commission Policy Issue, SECY-82-465;

1982.

5. USNRC Regulatory Guide 1.99, " Trend Curves: Predicted Adjustments of

Reference Temperature RTNDT as a Function of Fluence and CopperContent", Revision 1.

6. NUREG 0737, Supplement 1, " Requirements for Emergency Response

Capability"; January 1983.

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TABLE I

SUMMARY OF ALABAMA POWER COMPANY POSITION FOR AN AUGMENTED

REACTOR VESSEL EXAMINATION PROGRAM

AREAS WHERE AUGMENTED PROGRAM WILL APPLY *

Manual or Instrument Recording ReportingExamination Automated Performance and ofCat: gory Description Examination Checks Calibration Examination Sizing Results

B-A Upper-to-Intermediate Shell Circ. Weld A X X X X X

B-A Intermediate-to-Lower Shell Circ. Weld A X X X X X

B-A Intermediate Shell Long. Welds A X X X X X

B-A Lower Shell Long. Welds A X X X X X

B-B Lower Shell-to-Bottom Head Circ. Weld A/M** X X X X

B-B Lower Head Ring-to-Meridional Circ. Weld A/M** X X X X

B-B Lower Head Heridional-to-Bottom Plate Circ. Weld A/M** X X X X

B-B Bottom Head Meridional Long. Welds A/M** X X X X,B-B Closure Head Meridional-to-Dollar Plate Circ. Weld M X X X X.

B-B Closure Head Meridional Long. Welds M X X X X

B-C Vessel Flange-to-Upper Shell Circ. Weld A X X X X

B-C Closure Head Flange-to-Meridional Circ. Weld M X X X X

B-D Inlet Nozzle-to-Shell Welds A X X X X

B-D Outlet Nozzle-to-Shell Welds A X X X X

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* Refer to Attachment A, " Alabama Power Company Technical Position For AnAugnented Reactor Vessel Examination," for specific details.

** Automatic and/or manual, as access permits.,

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ATTACHMENT A

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ALABAMA POWER COMPANY ,

TECHNICAL POSITION

FOR AN

AUGMENTED REACTOR VESSEL

EXAMINATION PROGRAM"

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ATTACHMENT A.

ALABAMA POWER COMPANY~

TECHNICAL POSITION

FOR AN AUGMENTED REACTOR VESSEL EXAMINATION PROGRAM

INTRODUCTION

USHRC Regulatory Guide 1.150, Revision 1, " Ultrasonic Testing of ReactorVessel Welds During Preservice and Inservice Examination", has been issuedwith the intent of increasing reliability for detection and characterizationof service induced defects. The guide recommends supplementing currentASME XI requirements in the areas of:

1. Instrument performance2. System calibration3. Near surface resolution4. Beam profile5. Scanning weld-metal interface6. Sizing, and7. Reporting of results.

Alabama Power Company ( APCO) has evaluated Revision 1 of the Regulatory Guidewhich, in Appendix A, adopts the Electric Power Research Institute (EPRI) AdHoc Committee's recommendations as an acceptable approach to the positionsstated in the base document. As a result of this review, APC0, in conjunctionwith Westinghouse, has developed a plan for augmenting ASME XI examinations of

the Farley Unit I and II reactor vessels. The technical position for thisaugmented program, based on Appendix A to Revision 1 of Regulatory Guide1.150, is described herein.

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1.0 INSTRUMENT PERFORMANCE CHECKS

1.1 PRE-EXAM PERFORMANCE CHECKS |

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Performance checks recommended by this particular sub-paragraph areaddressed via implementation of those checks recommended in 1.2.

1.2 FIELD PERFORMANCE CHECKS

1.2a Frequency of Checks

Performance checks recontiended in 1.2c,1.2d, and 1.2e are conducted

before and after examining all vessel welds that need be examined duringthe outage. Those specified per 1.2f are conducted prior toexaminations, during the calibration sequence.

1.2b Instrument linearity checks will cover the range of instrument settingsintended for use during the examinations.

1.2c RF Waveform

Photographic records of the RF pulse waveforms are obtained before and

after each reactor vessel examination for each transducer used formechanized scanning.

1.2d Screen Height Linearity

Screen height linearity is' determined prior to calibration and before andafter each series of reactor vessel examinations during one outage.

1.2e Amplitude Control Linearity

Amplitude control linearity is determined prior to calibration and beforeand after each series of reactor vessel examinations during the outage.

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1.2f Angle Beam Profile Characterization.

Vertical beam profile measurements are conducted for each transducer used

during the examinations as part of the calibration sequence. Thesemeasurements are specified at 20 percent of the distance-amplit'ude-curveas well as at 50 percent of the distance-amplitude-curve and all<

measurement data are included in the calibration data package.

2.0 CALIBRATION,

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System calibration is established on the appropriate basic calibrationblock (s) in accordance with Appendix I of the 1974 Edition of Section XI

of the ASME Boiler and Pressure Vessel Code including Addenda through,

: Summer 1975.t

2.1 CALIBRATION FOR MANUAL SCANNINGt

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Procedures used for manual scanning of reactor vessel welds specifyinvestigation of all indications which exceed 20 percent DAC.Indications which exceed 50 percent DAC are considered recordable. Inaddition, scanning is performed at two times the calibration sensitivityto provide additional conservatism from a detection standpoint.

2.2 CALIBRATION FOR MECHANIZED SCANNING

2.2a Distance-amplitude-curves are developed with transducers mounted on thei

array plate planned for use during the specific reactor vesselexamination when possible.

2.2b Distance-amplitude-curves are developed statically per Appendix I of the

1974 Edition of Section XI of the ASME Boiler and Pressure Vessel Codeincluding Addenda through Summer 1975 and verified dynamically at orhigher than the specified scanning speed.

2.2c The scanning motion of the reactor vessel inspection tool is such thattransducers scan forward during one scan increment and backward on thenext successive increment. Experience indicates the difference in scanmotion has no significant effect on detection.

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2.2d Calibration is verified dynamically thus development of correctionfactors is not appropriate..

2.3 CALIBRATION CONFIRMATION.

System calibration is confirmed, as a minimum, before and after each

series of vessel examinations with a particular array plate. Inc.ddition, instrument stability is verified every four hours using anElectronic Block Simulator (EBS).

2.3a Complete ultrasonic system performance is confirmed using an array ofcylindrical reflectors called a Mechanical Calibration Transfer Standard(MTS). Responses from reflectors in the array are referenced to thedistance-amplitude-curves generated with the basic calibration block (s)per Appendix I of the 1974 Edition of Section XI of the ASME Boiler and

Pressure Vessel Code including Addenda through Summer 1975. The designof the array allows at least a two point check of sweep and sensitivity.Reflectors selected for calibration verification appear at transit timesrepresentative of those of the primary reflectors in the basiccalibration block where practical.

2.3b Written records are maintained for both the target reflector responsesand the distance-amplitude-curves for each transducer / inspection channelcombination.

2.3c The entire ultrasonic system is protected from temperature, vibration,and shock via a trailer mounted control center with controlledenvironment.

; 2.4 CALIBRATION BLOCKS

Basic calibration blocks are designed per Appendix I of the 1974 Editionof Section XI of the ASME Boiler and Pressure Vessel Code including

i Addenda through Sumer 1975. Surfaces of reference reflectors will beprotected from the environment by a suitable plugging method.

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3.0 EXAMINATION.

The scope and extent of the ultrasonic examinations are per IWA-2000,

1974 Edition of Section XI of the ASME Boiler and Pressure Vessel Codeincluding Addenda through Summer 1975 as noted in the Farley licenseamendment including any additional amendments and reliefs pertaining to

that license.

The ultrasonic system used for these examinations is capable of recording

of multiple indications appearing simultaneously and all examinations areconducted with a minimum of 25 percent scan overlap based on the

transducer element size.

3.1 INTERNAL SURFACE

Examination procedures provide for supplementing code requiredexaminations with techniques specifically intended to interrogate volumesof material near the vessel ID surface when scanning vessel shellbeltline region welds and access and geometry permit. Procedures providefor implementation of near surface examination methods as described ineither 3.la or 3.lb.

3.la Forty-five degree full node examinations demonstrated capable ofidentifying the near surface, 2 percent, 90 degree corner reflector, or4

3.lb Shallow angle techniques demonstrated capable of identifying the nearsurface, 2 percent, 90 degree corner reflector or other appropriatereference defect.

3.lc The near surface examination technique selected, either 3.la or 3.lb,will be conducted to provide coverage of one inch of material near thevessel ID, as a minimum.

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3.2 SCANNING WELD-METAL INTERFACE

Beam angles are selected for examinations of nozzle-to-shell welds fromthe nozzle bores and the vessel flange-to-shell weld from the flange seal

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surface based on their ability to provide complete coverage of the weld |

and specified adjacent base material and provide normal or near-normalincidence to the weld / base metal interface. Ability to sdhere to the+ 15 degree tolerance suggested is dependent upon component geometry.

4.0 BEAM PROFILE

The APC0 position on this item is stated in 1.2f.

5.0 SCANNING WELD-METAL INTERFACE

The APC0 position on this item is stated in 3.2.

6.0 RECORDING AND SIZING

The recording and sizing criteria recomended in 6.2 and 6.3 will beapplied to augment Section XI recording requirements for examinations ofwelds in the reactor vessel beltline region. Examinations of all otherareas of the reactor vessel will employ the recording criteria 1,pecifiedin the 1974 Edition of Section XI including Addenda through Summer 1975.

6.1 GEOMETRIC INDICATIONS

All indications which exceed the appropriate recording level are listedon the renote inspection tool data printout in terms of amplitude, sweepposition, and location in the vessel.

Procedures specify that all indications on the data printout beinvestigated to determine whether they are valid (i.e., cracks, lack ofpenetration, inclusions, slag, etc.) or not valid (i.e., geometry, beamredirection, loss of interface gating, etc.). This interpretation isnoted on the data printout.

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6.2 INDICATIONS WITH CHANGING METAL PATH

6.2a Valid angle beam indications detected via Section XI techniques in thereactor vessel beltline region will be recorded per the augmentedcriteria specified in 6.2b and 6.2c. Examinations of all ot'her areas ofthe reactor vessel will employ the recording criteria specified in the1974 Edition of Section XI including Addenda through Sumer 1975.

6.2b Va?id angle beam indications at metal paths representing 25 percent andgreater of the through-wall thickness of the vessel wall measured fromthe inner surface will be recorded at 50 percent DAC.

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6.2c Valid angle beam indications which are within the inner 25 percent of thevessel through-wall thickness measured from the vessel inner surface willbe recorded at 20 percent DAC. If the indication exceeds 50 percent DACit will be recorded to 50 percent DAC limits as well.

6.3 INDICATIONS WITHOUT CHANGING METAL PATH

Valid angle beam indications detected via Section XI techniques in thereactor vessel beltline region will be recorded per the augmentedcriteria specified in 6.3a and 6.3b. Examinations of all other areas ofthe reactor ves::e1 will employ the recording criteria specified in the1974 Edition of Section XI including Addenda through Summer 1975.

6.3a Valid angle beam indications at metal paths representing 25 percent andgreater of the through-wall thickness of the vessel wall measured fromthe inner surface will be recorded at 50 percent DAC.

6.3b Valid angle beam indications which are within the inner 25 percent of thevessel through-wall thickness measured from the vessel inner surface willbe recorded at 20 percent DAC. If the indication exceeds 50 percent DACit will be recorded to 50 percent DAC limited as well.

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6.4 ADDITIONAL RECORDING CRITERIA

6.4a Approprite scan intervals for recording of indications depend,onindication length, beam spread, etc., and are determined on acase-by-case basis.

6.4b Recorded information includes the sweep position and transducer positionfor the peak amplitude point and 100 percent, 50 percent, and 20 percentDAC points where appropriate.

6.4c Information presented on the data printout includes the indicationamplitude, sweep position, and location in the vessel. The system iscapable of recording multiple indications which may appear simultaneously.

7.0 REPORTING OF RESULTS

The reactor vessel examination final report will include all recordsobtained per implementation of items 1, 2, 3, and 6 as described herein.

7a. An estimate of the error band for sizing flaws cannot be provided at thistime. Estimates of this type are subjective in nature and not readilysubstantiated with quantitative data. The industry is currentlyassessing procedures and equipment for reflector sizing with the intentof providing a quantitative data base for such estimates and identifyingimproved methods, where appropriate.

7b. Descriptions of the calibration procedures will be included in thereactor vessel examination final report.

7c. Estimates of examination limitations due to geometry, access, gating,etc., will be provided in the reactor vessel examination final report.

7d. Descriptions and sketches of the remote examination system which explainits operation will be provided in the reactor vessel examination finalreport.

7e. Alternative volumetric techniques, if applied, will be described in thefinal report.

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ATTACHMENT B

PRESSURIZED THERMAL SHOCK CONSIDERATIONS

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ATTACHMENT B

Pressurized Thermal Shock Considerations.

Reactor vessel pressurized thermal shock (PTS) events have been shown from

pressurized water reactor (PWR) operating experience to be transients whichresult in a rapid and severe cooldown in the primary system coincident with ahigh or increasing primary pressure. The PTS concern arises if one of thesetransients acts on the beltline region of a reactor vessel where a reducedfracture resistance exists because of neutron irradiation. Su::h an event mayproduce propagation of flaws postulated to exist near the inner wall surface,thereby potentially affecting the integrity of the vessel.

The PTS concern is associated with the reactor vessel beltline region sincethis portion of the vessel can be subjected to both significant neutronirradiation and sudden cooldown temperatures coincident with high pressureloadings. The beltline regions of the Farley Unit 1 and 2 vessels arefabricated _by welding two plates together to form the intermediate shellcourse, two plates for the lower stell course and joining the two courses viaan intermediate-to-lower shell circumferential weld. The upper-to-intermediate shell circumferential weld is also considered susceptible to aPTS event. Each plate and weld has different material composition, importantin determining the fracture resistance of each location.

The overall effects of fast neutron irradiation on the mechanical propertiesof low alloy ferritic pressure vessel steels are well documented in theliterature. N In general, low alloy ferritic materials show an increase inhardness and tensile properties and a decrease in ductility and toughness,which is importent for fracture considerations, under certain conditions ofirradiation.

Reference nil-ductility transition temperature (RTNDT) is a measure of thetemperature at which the vessel or weld material undergoes a transition fromductile to brittle behavicr. Its initial value is determined for eachlocation at the time the vessel is fabricated using a destructive specimen

testing procedure. During the service life of the reactor vessel, the RTNDT|

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shifts above the initial value of RT because of irradiation. The amountNDT

of shift has been characterized in several forms of trend curves, generally asa function of neutron fluence and various residual element compositions of thematerial, based on toughness measurements of irradiated materials. For agiven vessel location, the value of RT will vary in the longitudinal

NDTand/or circumferential directions because of the respective variations invessel fluence in the axial and azimuthal directions and will decrease invalue through the vessel wall because of fluence attenuation.

The value of RTNDT at a given time in vessel life is used in fracturemechanics calculations to determine whether an assumed flaw would extend whenthe vessel is subjected to PTS events. In the fracture calculations for PTSevents, flaws are postulated to exist at the inner surface of the limitinglocation (s) of the vessel beltline regfon because 1) surface flaws are morelimiting than buried flaws, and 2) the seitline inner surface experiences thelargest amount of neutron irradiation and the highest thermal stress.Postulated flaws are generally assumed to be oriented in the same direction asthe longitudinal or circumferential welds in which they are located.

Specific to the PTS issue for the Farley Units 1 and 2 reactor vessel beltlineregions, Table 1 summarizes the inner surface RTNDT values for the locationsof interest. These inner surface RTNDT values, which were calculated inaccordance with the prescribed method in SECY-82-465 (dated November 23,1982)

- NRC Policy Issue on PTS, are the current way of assessing the impact of PTSon reactor vessel integrity. These values are assessed against RT

NDT

screening values that are also established in SECY-82-465 to give anindication as to when the risk of vessel damage from PTS may reach an

unacceptable level. The appropriate RTNDT screenine a O n s are alsopresented in Table 1 for reference.

Lower shell plate B6919-1 and intermediate shel} plate 87212-1 are therespective limiting locations for Farley Unit 1 and Unit 2 in terms of

values,calculated RT values for PTS. The corresponding limiting RTNDTNDT

along with the RTNDT values for the other vessel locations, are consistentwith those calculated for the Westinghouse Owners Group (WOG) and reportcJ inletter, WOG-82-290, December 31, 1982 to the W0G representatives. The data

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indicates that a PTS concern should not exist at any location in the Farley-

| Units 1 and 2 reactor vessel beltline regions during plant lifetime. Thei. values are conservative in they do not reflect any fluence adjustments as a

i result of low leakage cores already in place or planned. Also, thi WOGJj material properties were used in calculating the RT values for Farley

NDT

| Unit 1, and they are different from those currently provided by the NRC in i'

} SECY-82-465. However, the NR: verbally and i;nofficially accepted the

| Westinghouse defense for the WOG values at a " working" meeting held on ;

}: October 25, 1982 at the Westinghouse Bethesda office between NRC,

| Westinghouse, and WOG representatives. The NRC staff stated that they would

! not change their values prior to the issuance of SECY-82-465 since the NRC1

; will request plant-specific data as part of the NRC Rule on PTS. The NRC will

correct their RTNDT values once the plant-specific data is officiallyj requested and received. To this end, the WOG values should be considered to

be the correct values for use in NRC related activites at this time.

|I To provide further assurance that a severe PTS event is not likely to occur,

Alabama Power Company is implementing per NUREG-0737, Supplement 1, specific;

| plant emergency procedures which address the PTS concern. These procedures

j and supplementary operator training programs are directed toward increasingoperator awareness of the potential for PTS, recognizing a potential orimminent PTS event, and providing instructions in the event of an unexpectedly

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i severe RCS cooldown following a reactor trip or safety injection actuation.Operators are provided with instruction to attempt to prevent further cooldown

j and to minimize pressure in the RCS in order to respond to a challenge to

| reactor vessel integrity. Guidance is also provided on any subsequent RCS

) cooldown restrictions required to safely achieve cold shutdown conditions.

i WOG input to this program includes Functional Restoration Guidelines FR-P.1,

! " Response to Imminent Pressurized Thermal Shock Condition" and FR-P.2,

i " Response to Anticipated Pressurized Thermal Shock Condition.")

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Summarizing the PTS issue as it relates to Farley Units 1 and 2, 1) review of

| the RT values provided in Table 1 indicates that a PTS concern should notNDT

! exist in any location in the beltline region of either vessel during theiri

i lifetime, 2) Alabama Power Company is implementing specific plant emergencyI procedures and training programs to minimize the potential for a severe PTS1

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event, and 3) implementation of the Alabama Power Company Position for an.

Augmented Reactor Vessel Examination Program will provide even furtherassurance of vessel integrity; especially in the areas involved in the PTS

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issue.

References

[1]UnitedStatesNuclearRegulatoryCommissionPolicyIssue,SECY-82-465,1982,

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TABLE lA-

INNER SURFACE RT VALUES FOR FARLEYNDT.

UNIT 1 REACTOR VESSEL BELTLINE REGION

RT FNDT

LOCATION PRESENT LIFE END-OF-LIFE NRC SCREENING

(3 EFPY) (32 EFPY)* VALUE

Intermediate shell 113 172 270

plate B6903-2


plate B6903-3

Lower shell 132 194 270

plate B6919-1


plate B6919-7

Vertical Walds 78 146 270

Circumferential Welds 103 192 '300

* Values based on original out-in-in-fuel managenent loading patterns. Low

leakage patterns which were initiated beginning with cycle 5 will reduce the

End-of-Life RTNDT values below those given in the table.|

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TABLE 1B-

INNER SURFACE RT VALUES FOR FARLEYNDT~

UNIT 2 REACTOR VESSEL BELTLINE REGION

RT FNDT

LOCATION PRESENT LIFE END-OF-LIFE NRC SCREENING

(3 EFPY) (32 EFPY)* VALUE


plate B7203-1.


plate B7212-1


plate B7210-1


plate B7210-2

Vertical Welds -15 -4 270

Circumferential Welds 45 103 300

* Values based on original out-in-in-fuel management loading patterns. Lowleakage loading patterns which were initiated with cycle 3 will reduce the

End-of-Life RTNDT values below those given in the table.

0075W:42/100583- _ - .-. -- - ,

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ATTACHMENT C

EFFECTS OF LOW LEAKAGE FUEL PATTERN

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ATTACHMENT C.

| EFFECTS OF LOW LEAKAGE FUEL LOADING PATTERN

Alabama Power Company has adopted a low Leakage Loading Pattern for J. M.

Farley Unit 1 and Unit 2. In addition to the benefits of increased fueleconomy, the major effect of this fuel loading pattern is reduced neutronembrittlement damage to the reactor vessel shell material. The data shown inFigure 1 rnd Figure 2 illustrate the' reduced fast and thermal flux predictedfor the peripheral fuel assemblies of Farley Unit II at the beginning of cycle2 versus beginning of cycle 3, when the low leakage fuel design was firstused. An absolute value for the flux reduction at the vessel wall is notavailable; however, an estimate of this trend can be obtained from the coreflux and power distribution plots given in Figures 1 and 2 and from the letterconcerning the excore power range detector calibration (Attachment C-1). Thisletter summarizes reductions in the excore power range detector currentsexperienced at Farley Unit 1 during the changeover to the low leakage fuelpattern initiated beginning with cycle 5. Detector currents were reduced byabout 26.6% due to the fuel patte'n change. It is expected that reductions inthe neutron flux would be proportionate to the reductions in detector current,and that the effects of neutron damage at the vessel wall would be reducedsimilarly, again by a proportionate amount.

0080W.42/100683

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

* , .,FIGURE 1

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-htE C,. CY2 9 sot.,'

AMt CYCLE 2 80L TORTISE H PLETI0es 8U=0 SAFLFACT=.89 E-Y DEPLETIost 8U= 0.00 TIME = 0.00 PAGE= 59TORTISE VER$10er- 4.10 DATE- 11/06/82 USER-etASAP MAlteFRAME-esF8 JosasAME-AKZGAGZ

AVERAGE ASSDOLY FA5T FLUI, TMRMAL FLUX, A8eD FAST / THERMAL FLUX AATIO

~

32.6211 5.928 --

5.50

35.558 35.288 APS CYCLE 2 80L TORTISE DEPLETIces 8U=0 SAFLFACT=.89 I-T bEPLETI0tt Gua 0.0 '

.2 5.503 5.434 TORTISE VER510st- 4.10 DATE- 11/06/82 USER-stASAP MAlmFRAME-MF8 J08ena'

AVERAGE A55DWLY POWER, PEAK ASSEseLY POWER Ase0 AVERAGE A55DOLf BURetAP DISTRIBUT10el*

*34.397 33.727 35.954. 3 5.316 5.476 5.592

6.4T 4.16 6.43.854

31.944 34.301 34.163 34.305 1 .8746 5.200 5.184 5.864 5.953 13592

4.14 6.61 4.1T 4.431.0F9 1.052

32.382 35.310 38.150 34.808 34.221 2 1.1T3 1.121-5 5.263 5.325 5.921 5.983 5.371 1111' 1313'

6.15 6.63 4.44 4.15 6.3T

34.313 35.345 35.853 36.808 23.543 3 1.079 1.002 1.2270 5.633 5.754 5.858 6.280 4.016 13139 16182 10100

6.45 6.14 4.12 5.86 5.86 .879 1.017 1.007 1.18035.783 37.462 30.771 21.799 4 .935 1.095 1.099 1.283

.i 7 5.862 6.386 5.258 3.755 17208 14WT 15847 N4.10 5.87 5.85 5.81

.891 1.047 1.175 1.007 1.06325.494 21.219 5 .954 1.147 1.279 1.071 1.231

8 4.393 3.651 17648 14709 8873 18386 9199! 5.00 5.81

1.113 .971 .989 1.232 .7864 1.216 1.056 1.064 1.378 1.178

10100 18031 17859 0 0'.

i

.990 1.252 1.030 .734 i,

7 1.062 1.334 1.301 1.145! 17872 0 0 0

.859 7148 1.133 1.117

! 0 0

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_ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.

*. FIGURE 2 ' '

!-

YARt.EYZ CY3 G DL -

TORTISE VERSION- 4.12 DATE- D4/13/83 USER-NEARL MAINFR AME-MFB JOBNAME-AVAA004

. Sus 0.00 TIME = 0.00 PAGE= 162

AVER AGE ASSEMBLY FAST FLUX, THERMAL FLUX, AND FAST / THERMAL FLUX RATIO

34.8191 5.312

6.55

37.245 33.471 TORT 15E VERSION- 4.12 DATE- 04/13/03 USER-NE AXL MAINFRAME-MFS J08MA8U= 0.0

2 5.476 5.0F9AVERAGE ASSEMBLY POWER, PEAK ASSEMBLY POWER AND AVER AGE ASSEMBLY SURNUP DISTRISUTION

6.80 6.59

38.714 35.819 34.9143 5.797 4.824 5.251

6.68 F.43 6.65 .9261 .966

39.621 40.830 39.498 36.8294 5.525 6.021 5.268 5.530 1.176 .886T.17 6.78 T.50 6.66 2 1.326 .928' 0 2392837.113 41.398 42.18F 37.901 24.9945 5.591 5.502 4.237 5.345 3.804,

1.153 1.003 942| 6.64 T.52 6.76 7. 09 6.57 3 1.258 1.152 1.037l 37.734 41.5'56 41.0F0 32.959 13.584 'i 6 5.667 6.190 5.478 5.150 2.071 1.134 1.208 1.171 .9976.66 6.71 F.50 6.40 6.56 4 1.232 1.306 1.316 1.053

38.694 37.430 29.773 15.627 096 7753 0 207947 5.878 5.605 4.644 2.387

6.58 6.68 6.41 6.55 .993 1.224 1.220 1.165 .6955 1.039 1.356 1. 29 2 1.364 9687 39 22376 0 11709 0 195198 3.752 2.674

6.34 6.60 1.019 1.223 1.218 1.096 .3616 1.096 1.332 1.354 1.397 .69521176 9934 0 0 253261.175 1.196 .988 .4357 1.262 1.337 1.344 .9177096 0 0 20426,

.797 .518*8 1.127 942;

O 11933,

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ATTACHMENT C-1SATO-NO-173

<

:.:- NTD

e.n 284-5604cre June 20,1983 ~

see:- Low Leakage Loading Effects atFarley Nuclear Plant Cycle 5 Startup

n W. L. Orr, Manager PHOB-301NFD Major Programs

cc: A. J. Impink MNC-469U. L. Brown MMOB-310E. M. Spier MMOB-303C. R. Tuley MNC-421Nuclear Operations (1)

.

3The effects of Low Leakage Loading Pattern (L P) loadings on the excoredetectors has to be considered for equivalent impact on the Vantage-5(V-5) fuel product line. Of primary importance is the ability of excoreinstrumentation to sense changes in the core power distribution with theV-5prodgetcore. A complete evaluation requires plant data associatedwith a L P loading and 3D design calculations of a V-5 loading. Thismqmo documents the plant data from J.H. Farley Unit 1 which installed aLJP core for cycle 5.

Table 1 sumarizes the excore power range calibration data from FarleyUnit 1, cycles 1, 4 and 5. In the table, the top, bottom and totaldetector currents (u-amps) are shown. Also, the average total currentsand the detector gain, sensitivity as %at per calibrated avolts, areprovided. The following observations are made:

1. The detector currents were reduced by s5.5% between cycles1 and 4 due to detector depletion, loading pattern changes, etc.

were reduced by s26.6% between cyclesThedetectorcurrentgPloading.

2.4 and 5 due to the L

3. The detector currents are suf"ficient (>100u-amps) forl standard calibration techniques.1

4. As indicated by the gains, the sensitivity of the excore detectorsto axial power shape was virtually unaffected. This is furtherdemonstrated by the plot of Incore Axial Offset (AOIN) vs ExcoreAxial Offset (A0EX) shown on Figure 1. The detector sensitivity to

axial power shape is indicated by the slope of the line, which| -

remained effectively constant.'

..r-

- . - _ _ . _ .

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SATO-NO-173'-

,

.

In conclusion, if sufficient detector signal is available and the designpredictions show coupling between the outer assemblies and the innerassemblies, then the excore detectors should perform quite well on a V-5core. This will be further evaluated when more data becomes available.

This completes milestone 2 of DGRF-40847, V-5 Core Monitoring.

If there are any questions, please call.

f

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L. R. GrobmyerNTD Nuclear Operations

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SATO-NO-173

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TABLE 1.

SUMMARY OF ALA EXCORE CALIBRATION DATA,

PARAMETER CYCLE 1 CYCLE 4 _ CYCLE 5

N-41Top Current 266.9 254.1 194.3Bot Current 277.0 262.3 194.3Total Current 543.9 51G.4 388.6Gain (%/v) 11.459 10.786 11.009

N-42Top Current 269.1 255.6 189.6

.

| Bot Current 273.1 259.0 183.9Total Current 542.2 514.6 373.5Gain (%/v) 10.905 10.002 10.152,

i N-43Top Current 265.1 250.4 180.7Bot Current 293.5 277.8 193.5Total Current 558.6 528.2 374.2Gain (%/v) 11.113 10.550 10.349

N-44Top Current 261.4 244.9 184.7Bot Current 274.4 255.7 190.1Total Current 535.8 500.6 374.8Gain (%/v) 10.537 10.139 9.949

Avg Total Current 545.1 515.0 377.8

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