ML20081D517

Alabama Power Co Position for Augmented Reactor Vessel Exam Program

ML20081D517
Person / Time
Site:	Farley
Issue date:	10/06/1983
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20081D502	List: ML20081D499 ML20081D517
References
RTR-REGGD-01.150, RTR-REGGD-1.150 PROC-831006, NUDOCS 8311010224
Download: ML20081D517 (32)

Text

I ALABAMA POWER COMPANY POSITION FOR .

M AUGMENTED REACTOR VESSEL EXAMINATION PROGRAM OCTOBER 6, 1983 SUBMITTED BY:

WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR SERVICES INTEGRATION DIVISION P, O. B0x 78 PITTSBURGH, PA 15230

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ALABAMA POWER COMPANY POSITION FOR AN AUGMENTED REACTOR VESSEL EXAMINATION PROGRAM d

INTRODUCTION The following position has been adopted by Alabama Power Company with respect to implementation of an augmented examination program for the J. M. Farley Unit I and Unit II reactor vessels. The basis for the program is Appendix A to USNRC Regulatory Guide 1.150, Revision 1 where the Electric Power Research Institute Ad Hoc Comittee recomendations are adopted as an acceptable approach to the positions recommended in the base document. The augmented program will be applied to full penetration reactor vessel welds that are subject to volumetric examination as required by the 1974 Edition of Section XI of the ASME Boiler and Pressure Vessel Code including Addenda through Sumer 1975 as noted in the Farley license amendment including any additional amendments and reliefs pertaining to that license. Specifically, the program provides for augmenting examination requirements for all welds as defined in ASME 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 pressurized thermal shock point of view.

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

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

l Meridional and circumferential welds in the bottom head and closure head. l Category B-C -- Vessel-to-flange and closure head-to-flange welds.

t l Category B-D -- Nozzle-to-vessel welds and nozzle inside radius.

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IMPLEMENTATION OF THE AUGMENTED EXAMINATION PROGRAM The scope of the augmented program which will be implemented for examinations  ;

of the J. M. Farley Unit I and Unit II reactor vessels is described below and l summarized in Table I. The Alabama Power Company Technical Position on the Augmented Reactor Vessel Examination Program is found as Attachment A.

i Period 1 and 2 Vessel Examinations During the 40-month and 80-month examination periods, areas of the reactor vessel accessible for examinations include the vessel flange-to-shell weld from the seal surface, outlet nozzle-to-shell welds from the nozzle bores, welds in the closure head, outlet nozzle-to-safe end welds and ligaments around 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 angles which provide normal or near-normal incidence to the plane of the weld. This method of examination establishes a favorable relationship between the interrogating sound beam and any planar reflector which might exist parallel to the weld and provides examination coverage of the volumes of material near the vessel inside and outside diameter surfaces with no limitations due to gating, near field effects, etc. For these examinations, Alabama Power Company will augment Section XI requirements in the areas of instrument performance checks, calibration, examination, and reporting of results per the Technical Position, Attachment A. A 50 percent DAC recording level will be used for examinations of these areas.

Welds in the closure 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. Where access permits, these welds will be manually examined using contact techniques from the outside surface. These techniques provide coverage of the volumes of material near the vessel head inside diameter surface with no limitations due to gating, near field effects, etc.

For examinations of these welds Alabama Power Company will augment Section XI l

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requirements in the areas of instrument performance checks, calibration, examination, and reporting of results per the Technical Position, Attachment A. A 50 percent DAC recording level will be used for examinations of these l

areas. ,

. Examinations of the vessel flange ligaments and outlet nozzle-to-safe end welds 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

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vessel accessible for examination include all those listed for period 1 and 2 examinations and all reactor vessel circumferential and longitudinal shell welds, inlet nozzle-to-shell welds, bottom head meridional and circumferential welds, and inlet nozzle-to-safe end welds.

Welds in the reactor vessel beltline region are associated with neutron embrittlement and pressurized thermal shock considerations. For examinations of 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 Technical Position, Attachment A. Specifically, examinations of the upper-to-intermediate shell weld, the intermediate-to-lower shell weld, the intermediate shell longitudinal welds, and the lower shell longitudinal welds will be conducted applying the full scope of the augmented examination program including a 20 percent DAC recording level for Section XI angle beam examinations 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 and welds 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 l

l pressurized thermal shock. Examinations of these welds will be augmented in the 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 in the bottom head inaccessible to the remote inspection tool are typically examined manually from the outside surface, where access permits, using contact techniques. These techniques provide coverage of the volumes of material near the head inside diameter surface with no limitations due to gating, near field effects, etc.Section XI examinations of these welds will be 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 structural integrity standpoint because of the high stress concentration present.

Examinations of these welds are performed from the nozzle bores with angles which provide normal or near-normal incidence to the plane of the weld. This method of examination establishes a favorable relationship between the interrogating sound beam and any planar reflector which might exist parallel to the weld and provides examination coverage of the volumes of material near the vessel inside and outside diameter surfaces with no limitations due to gating, near field effects, etc. For these examinations, Alabama Power Company will augment Section XI requirements in the areas of instrument performance checks, calibration, examination, and reporting of results per the Technical Position, Attachment A. A 50 percent DAC recording level will be used for examinations of these areas.

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

SUPPORTING INFORMATION/ DOCUMENTATION A review of mechanical properties, specifically the reference nil-ductility temperature (RTNDT), for the welds and base materials comprising the Farley Unit I and Unit II reactor vessel beltline regions indicates that a pressurized thermal shock concern should not exist in the beltline region of either vessel during their lifetime. These findings are present in Attachment B. " 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 reactor vessels will provide even further assurance of vessel integrity, especially in

areas involved in the pressurized thermal shock issue.

I ATTACHMENTS The following attachments are presented to supplement the Alabama Power Company Position for an Augmented Reactor Vessel Examination Program. j Attachment A - Alabama Power Company Technical Position for an Augmented Reactor Vessel Examination Program Attachment B - Pressurized Thermal Shock Considerations Attachment C - Effects of Low Leakage Fuel Loading Pattern 0337W:42/100683

REFERENCES

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

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

3. "A Quantitative Methodology for Reactor Yessel Thermal Shock Decision Making"; Ackerson, Balkey, et. al.; draft submitted for publication in the International 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 Copper Content", 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 Reporting Examination Automated Performance and of Cat: 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 1

• Refer to Attachment A, " Alabama Power Company Technical Position For An Augnented Reactor Vessel Examination," for specific details.

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

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ATTACHMENT A ALABAMA POWER COMPANY ,

TECHNICAL POSITION FOR AN AUGMENTED REACTOR VESSEL EXAMINATION PROGRAM l

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ATTACHMENT A ALABAMA POWER COMPANY

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TECHNICAL POSITION FOR AN AUGMENTED REACTOR VESSEL EXAMINATION PROGRAM INTRODUCTION USHRC Regulatory Guide 1.150, Revision 1, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examination", has been issued with the intent of increasing reliability for detection and characterization of service induced defects. The guide recommends supplementing current ASME XI requirements in the areas of:

1. Instrument performance

2. System calibration

3. Near surface resolution

4. Beam profile

5. Scanning weld-metal interface

6. Sizing, and

7. Reporting of results.

Alabama Power Company ( APCO) has evaluated Revision 1 of the Regulatory Guide which, in Appendix A, adopts the Electric Power Research Institute (EPRI) Ad Hoc Committee's recommendations as an acceptable approach to the positions stated in the base document. As a result of this review, APC0, in conjunction with Westinghouse, has developed a plan for augmenting ASME XI examinations of the Farley Unit I and II reactor vessels. The technical position for this augmented program, based on Appendix A to Revision 1 of Regulatory Guide 1.150, is described herein.

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1.0 INSTRUMENT PERFORMANCE CHECKS 1.1 PRE-EXAM PERFORMANCE CHECKS l Performance checks recommended by this particular sub-paragraph are addressed 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 during the outage. Those specified per 1.2f are conducted prior to examinations, during the calibration sequence.

1.2b Instrument linearity checks will cover the range of instrument settings intended 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 for mechanized scanning.

1.2d Screen Height Linearity Screen height linearity is' determined prior to calibration and before and after each series of reactor vessel examinations during one outage.

1.2e Amplitude Control Linearity Amplitude control linearity is determined prior to calibration and before and 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. These measurements are specified at 20 percent of the distance-amplit'ude-curve as well as at 50 percent of the distance-amplitude-curve and all measurement data are included in the calibration data package.

2.0 CALIBRATION i

System calibration is established on the appropriate basic calibration block (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 SCANNING t

i Procedures used for manual scanning of reactor vessel welds specify investigation of all indications which exceed 20 percent DAC.

Indications which exceed 50 percent DAC are considered recordable. In addition, scanning is performed at two times the calibration sensitivity to provide additional conservatism from a detection standpoint.

2.2 CALIBRATION FOR MECHANIZED SCANNING i

2.2a Distance-amplitude-curves are developed with transducers mounted on the array plate planned for use during the specific reactor vessel examination 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 Code including Addenda through Summer 1975 and verified dynamically at or higher than the specified scanning speed.

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

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2.2d Calibration is verified dynamically thus development of correction

. factors 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. In c.ddition, instrument stability is verified every four hours using an Electronic Block Simulator (EBS).

2.3a Complete ultrasonic system performance is confirmed using an array of cylindrical reflectors called a Mechanical Calibration Transfer Standard (MTS). Responses from reflectors in the array are referenced to the distance-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 design of the array allows at least a two point check of sweep and sensitivity.

Reflectors selected for calibration verification appear at transit times representative of those of the primary reflectors in the basic calibration block where practical.

2.3b Written records are maintained for both the target reflector responses and the distance-amplitude-curves for each transducer / inspection channel combination.

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

2.4 CALIBRATION BLOCKS Basic calibration blocks are designed per Appendix I of the 1974 Edition of Section XI of the ASME Boiler and Pressure Vessel Code including i

Addenda through Sumer 1975. Surfaces of reference reflectors will be protected 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 Code including Addenda through Summer 1975 as noted in the Farley license amendment 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 are conducted with a minimum of 25 percent scan overlap based on the transducer element size.

3.1 INTERNAL SURFACE Examination procedures provide for supplementing code required examinations with techniques specifically intended to interrogate volumes of material near the vessel ID surface when scanning vessel shell beltline region welds and access and geometry permit. Procedures provide for implementation of near surface examination methods as described in either 3.la or 3.lb.

3.la Forty-five degree full node examinations demonstrated capable of 4

identifying the near surface, 2 percent, 90 degree corner reflector, or 3.lb Shallow angle techniques demonstrated capable of identifying the near surface, 2 percent, 90 degree corner reflector or other appropriate reference 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 the vessel ID, as a minimum. ,

l 3.2 SCANNING WELD-METAL INTERFACE Beam angles are selected for examinations of nozzle-to-shell welds from the nozzle bores and the vessel flange-to-shell weld from the flange seal 0316W:42/100683

surface based on their ability to provide complete coverage of the weld l and specified adjacent base material and provide normal or near-normal incidence 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 be applied to augment Section XI recording requirements for examinations of welds in the reactor vessel beltline region. Examinations of all other areas of the reactor vessel will employ the recording criteria 1,pecified in the 1974 Edition of Section XI including Addenda through Summer 1975.

6.1 GEOMETRIC INDICATIONS All indications which exceed the appropriate recording level are listed on the renote inspection tool data printout in terms of amplitude, sweep position, and location in the vessel.

Procedures specify that all indications on the data printout be investigated to determine whether they are valid (i.e., cracks, lack of penetration, inclusions, slag, etc.) or not valid (i.e., geometry, beam redirection, loss of interface gating, etc.). This interpretation is noted 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 the reactor vessel beltline region will be recorded per the augmented criteria specified in 6.2b and 6.2c. Examinations of all ot'her areas of the reactor vessel will employ the recording criteria specified in the 1974 Edition of Section XI including Addenda through Sumer 1975.

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

8 6.2c Valid angle beam indications which are within the inner 25 percent of the vessel through-wall thickness measured from the vessel inner surface will be recorded at 20 percent DAC. If the indication exceeds 50 percent DAC it 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 the reactor vessel beltline region will be recorded per the augmented criteria specified in 6.3a and 6.3b. Examinations of all other areas of the reactor ves::e1 will employ the recording criteria specified in the 1974 Edition of Section XI including Addenda through Summer 1975.

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

6.3b Valid angle beam indications which are within the inner 25 percent of the vessel through-wall thickness measured from the vessel inner surface will be recorded at 20 percent DAC. If the indication exceeds 50 percent DAC it 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,on indication length, beam spread, etc., and are determined on a case-by-case basis.

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

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

7.0 REPORTING OF RESULTS The reactor vessel examination final report will include all records obtained 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 this time. Estimates of this type are subjective in nature and not readily substantiated with quantitative data. The industry is currently assessing procedures and equipment for reflector sizing with the intent of providing a quantitative data base for such estimates and identifying improved methods, where appropriate.

7b. Descriptions of the calibration procedures will be included in the reactor 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 explain its operation will be provided in the reactor vessel examination final report.

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

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9 ATTACHMENT B PRESSURIZED THERMAL SHOCK CONSIDERATIONS

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 which result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary pressure. The PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Su::h an event may produce 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 since this portion of the vessel can be subjected to both significant neutron irradiation and sudden cooldown temperatures coincident with high pressure loadings. The beltline regions of the Farley Unit 1 and 2 vessels are fabricated _by welding two plates together to form the intermediate shell course, two plates for the lower stell course and joining the two courses via an intermediate-to-lower shell circumferential weld. The upper-to-intermediate shell circumferential weld is also considered susceptible to a PTS event. Each plate and weld has different material composition, important in determining the fracture resistance of each location.

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

Reference nil-ductility transition temperature (RTNDT) is a measure of the temperature at which the vessel or weld material undergoes a transition from ductile to brittle behavicr. Its initial value is determined for each location at the time the vessel is fabricated using a destructive specimen testing procedure. During the service life of the reactor vessel, the RTNDT l

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shifts above the initial value of RTNDT because of irradiation. The amount of shift has been characterized in several forms of trend curves, generally as a function of neutron fluence and various residual element compositions of the material, based on toughness measurements of irradiated materials. For a given vessel location, the value of RT NDT will vary in the longitudinal and/or circumferential directions because of the respective variations in vessel fluence in the axial and azimuthal directions and will decrease in value through the vessel wall because of fluence attenuation.

The value of RTNDT at a given time in vessel life is used in fracture mechanics calculations to determine whether an assumed flaw would extend when the vessel is subjected to PTS events. In the fracture calculations for PTS events, flaws are postulated to exist at the inner surface of the limiting location (s) of the vessel beltline regfon because 1) surface flaws are more limiting than buried flaws, and 2) the seitline inner surface experiences the largest amount of neutron irradiation and the highest thermal stress.

Postulated flaws are generally assumed to be oriented in the same direction as the longitudinal or circumferential welds in which they are located.

Specific to the PTS issue for the Farley Units 1 and 2 reactor vessel beltline regions, Table 1 summarizes the inner surface RTNDT values for the locations of interest. These inner surface RTNDT values, which were calculated in accordance 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 PTS on reactor vessel integrity. These values are assessed against RTNDT screening values that are also established in SECY-82-465 to give an indication as to when the risk of vessel damage from PTS may reach an unacceptable level. The appropriate RTNDT screenine a O n s are also presented in Table 1 for reference.

Lower shell plate B6919-1 and intermediate shel} plate 87212-1 are the respective limiting locations for Farley Unit 1 and Unit 2 in terms of calculated RT NDT values for PTS. The corresponding limiting RTNDT values, along with the RTNDT values for the other vessel locations, are consistent with those calculated for the Westinghouse Owners Group (WOG) and reportcJ in letter, WOG-82-290, December 31, 1982 to the W0G representatives. The data 0075W:42/100583

indicates that a PTS concern should not exist at any location in the Farley l Units 1 and 2 reactor vessel beltline regions during plant lifetime. The

i. 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 WOG J

j material properties were used in calculating the RTNDT values for Farley l 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 l Westinghouse defense for the WOG values at a " working" meeting held on  ;

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October 25, 1982 at the Westinghouse Bethesda office between NRC, l 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 NRC 1

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 officially j 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.

l 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 l plant emergency procedures which address the PTS concern. These procedures j and supplementary operator training programs are directed toward increasing operator awareness of the potential for PTS, recognizing a potential or imminent 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 l reactor vessel integrity. Guidance is also provided on any subsequent RCS cooldown restrictions required to safely achieve cold shutdown conditions.

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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." '

)

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

the RT NDT values provided in Table 1 indicates that a PTS concern should not

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

i lifetime, 2) Alabama Power Company is implementing specific plant emergency I procedures and training programs to minimize the potential for a severe PTS 1

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. event, and 3) implementation of the Alabama Power Company Position for an Augmented Reactor Vessel Examination Program will provide even further assurance of vessel integrity; especially in the areas involved in the PTS issue.

References

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

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TABLE lA INNER SURFACE RTNDT VALUES FOR FARLEY UNIT 1 REACTOR VESSEL BELTLINE REGION RT F NDT LOCATION PRESENT LIFE END-OF-LIFE NRC SCREENING (3 EFPY) (32 EFPY)* VALUE Intermediate shell 113 172 270 plate B6903-2 Intermediate shell 116 169 270 plate B6903-3 Lower shell 132 194 270 plate B6919-1 Lower shell 122 185 270 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 NDT VALUES FOR FARLEY

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UNIT 2 REACTOR VESSEL BELTLINE REGION RT F NDT LOCATION PRESENT LIFE END-OF-LIFE NRC SCREENING (3 EFPY) (32 EFPY)* VALUE Intermediate shell 116 198 270 plate B7203-1 Intermediate shell 116 237 270 plate B7212-1 Lower shell 113 187 270 plate B7210-1 Lower shell 100 180 270 plate B7210-2 Vertical Welds -15 -4 270 Circumferential Welds 45 103 300

• Values based on original out-in-in-fuel management loading patterns. Low leakage loading patterns which were initiated with cycle 3 will reduce the End-of-Life RTNDT values below those given in the table.

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e ATTACHMENT C EFFECTS OF LOW LEAKAGE FUEL PATTERN I

ATTACHMENT C l

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 fuel economy, the major effect of this fuel loading pattern is reduced neutron embrittlement damage to the reactor vessel shell material. The data shown in Figure 1 rnd Figure 2 illustrate the' reduced fast and thermal flux predicted for the peripheral fuel assemblies of Farley Unit II at the beginning of cycle 2 versus beginning of cycle 3, when the low leakage fuel design was first used. An absolute value for the flux reduction at the vessel wall is not available; however, an estimate of this trend can be obtained from the core flux and power distribution plots given in Figures 1 and 2 and from the letter concerning the excore power range detector calibration (Attachment C-1). This letter summarizes reductions in the excore power range detector currents experienced at Farley Unit 1 during the changeover to the low leakage fuel pattern initiated beginning with cycle 5. Detector currents were reduced by about 26.6% due to the fuel patte'n change. It is expected that reductions in the neutron flux would be proportionate to the reductions in detector current, and that the effects of neutron damage at the vessel wall would be reduced similarly, again by a proportionate amount.

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.879 1.017 1.007 1.180 35.783 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 N 4.10 5.87 5.85 5.81

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FIGURE 2 ' '

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Sus TORTISE VERSION- 4.12 DATE- D4/13/83 USER-NEARL 0.00 TIME = 0.00 PAGE= 162 MAINFR AME-MFB JOBNAME-AVAA004 AVER AGE ASSEMBLY FAST FLUX, THERMAL FLUX, AND FAST / THERMAL FLUX RATIO 34.819 1 5.312 6.55 37.245 33.471 TORT 15E VERSION- 4.12 DATE- 04/13/03 8U= 0.0 2 5.476 5.0F9 USER-NE AXL MAINFRAME-MFS J08MA 6.80 6.59 AVERAGE ASSEMBLY POWER, PEAK ASSEMBLY POWER AND AVER AGE ASSEMBLY S 38.714 35.819 34.914 3 5.797 4.824 5.251 6.68 F.43 6.65 .926 1 .966 4

39.621 40.830 39.498 36.829 5.525 6.021 5.268 5.530 T.17 6.78 T.50 6.66 1.176 .886 2 1.326 .928 37.113 41.398 42.18F 37.901 24.994 0 23928

, 5 5.591 5.502 4.237 5.345 3.804 l 6.64 T.52 6.76 7. 09 6.57 1.153 1.003 942 3 1.258 1.152 1.037 l '

i 6 37.734 5.667 41.5'56 41.0F0 32.959 13.584 6.190 5.478 5.150 2.071 6.66 6.71 F.50 6.40 6.56 1.134 1.208 1.171 .997 4 1.232 1.306 1.316 096 1.053 7

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ATTACHMENT C-1 SATO-NO-173

.:- NTD e.n 284-5604 ~

cre June 20,1983 see:- Low Leakage Loading Effects at Farley Nuclear Plant Cycle 5 Startup n W. L. Orr, Manager PHOB-301 NFD Major Programs cc: A. J. Impink MNC-469 U. L. Brown MMOB-310 E. M. Spier MMOB-303 C. R. Tuley MNC-421 Nuclear Operations (1)

The effects of Low Leakage Loading Pattern (L3 P) loadings on the excore detectors has to be considered for equivalent impact on the Vantage-5 (V-5) fuel product line. Of primary importance is the ability of excore instrumentation to sense changes in the core power distribution with the V-5prodgetcore. A complete evaluation requires plant data associated with a L P loading and 3D design calculations of a V-5 loading. This mqmo documents the plant data from J.H. Farley Unit 1 which installed a LJP core for cycle 5.

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

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

2. were reduced by s26.6% between cycles ThedetectorcurrentgPloading.

4 and 5 due to the L

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

1

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

axial power shape is indicated by the slope of the line, which remained effectively constant.

..r-

SATO-NO-173 In conclusion, if sufficient detector signal is available and the design predictions show coupling between the outer assemblies and the inner assemblies, then the excore detectors should perform quite well on a V-5 core. 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 O hy W L. R. Grobmyer NTD Nuclear Operations

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SATO-NO-173 TABLE 1

SUMMARY

OF ALA EXCORE CALIBRATION DATA ,

PARAMETER CYCLE 1 CYCLE 4 _ CYCLE 5 N-41 Top Current 266.9 254.1 194.3 Bot Current 277.0 262.3 194.3 Total Current 543.9 51G.4 388.6 Gain (%/v) 11.459 10.786 11.009 N-42

. Top Current 269.1 l 255.6 189.6 Bot Current 273.1 259.0 183.9 Total Current 542.2 514.6 373.5 Gain (%/v) 10.905 10.002 10.152 i

N-43 Top Current 265.1 250.4 180.7 Bot Current 293.5 277.8 Total Current 193.5 558.6 528.2 374.2 Gain (%/v) 11.113 10.550 10.349 N-44 Top Current 261.4 244.9 184.7 Bot Current 274.4 255.7 190.1 Total Current 535.8 500.6 374.8 Gain (%/v) 10.537 10.139 9.949 Avg Total Current 545.1 515.0 377.8 i

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