ML18065A466

Forwards Thermal Annealing Operating Plan Section 1.8, Proposed Annealing Conditions,Requalification Insp & Test Program,Section 2.2,insp Program & Requalification Insp & Test Program,Section 2.3,testing Program

ML18065A466
Person / Time
Site:	Palisades
Issue date:	02/02/1996
From:	Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9602090026
Download: ML18065A466 (21)

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.... _t consumers Power POWERiNii MICHlliAN"S PROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 February 2, 1 996 U S Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT PRELIMINARY THERMAL ANNEALING REPORT, THERMAL ANNEALING OPERATING PLAN, SECTION 1.8, PROPOSED ANNEALING CONDITIONS AND REQUALIFICATION INSPECTION AND TEST PROGRAM, SECTION 2.2, INSPECTION PROGRAM, AND SECTION 2.3, TESTING PROGRAM.

At a meeting on June 6, 1995, we discussed with the staff our plan to anneal the Palisades reactor vessel (RV) during the refueling outage currently scheduled for the middle of 1998. In support of this effort, we plan to submit the final Thermal Annealing Report (TAR) in the third quarter of 1 996 after the results of the Marble Hill reactor vessel annealing demonstration have been evaluated. The TAR will include the information recommended in Draft Regulatory Guide DG-1027, Format and Content of Application For Approval For Thermal Annealing of Reactor Pressure Vessels. To permit NRC review of the TAR to begin before the Marble Hill results are known, we will make a series of submittals of preliminary TAR sections as they are developed. This letter provides the sixth of those submittals.

The Attachment 1 to* this letter contains the Thermal Annealing Operating Plan Section 1 .8, Proposed Annealing Conditions. Attachment 2 contains the Requalification Inspection and Test Program, Section 2.2, Inspection Program.

Attachment 3 contains the Requalification Inspection and Test Program, Section 2.3, Testing Program. These sections are presented in the format recommended by Section C.1 and C.2 of DG-1027.

( 960~090026 960202 PDR P

ADOCK.05000255

, , PDR 'I A CMS' ENERGY COMPANY 1\0f! 1 I

• 2

SUMMARY

OF COMMITMENTS This letter contains no new commitments and no revisions to existing

• commitments.

Richard W Smedley Manager, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades Attachment

ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 THERMAL ANNEALING REPORT SECTION 1 THERMAL ANNEALING OPERATING PLAN SECTION 1.8 PROPOSED ANNEALING CONDITIONS 4 Pages

• 1.8 PROPOSED ANNEALING CONDITIONS 1.8.A Introduction The limiting parameters identified in this section take into consideration the material, temperature, stress, and strain requirements needed to assure an effective and acceptable anneal of the Palisades reactor vessel without further analysis and testing. Administrative limits on key parameters (such as heatup and cooldown rates, and axial, azimuthal, and through-wall gradients) have been identified in Section 1.4 in order to maintain a prudent engineering margin such that the limiting parameters in this section will not be exceeded. These administrative limits have been established based on the thermal and stress analysis results of Section 1 . 7.

A description of the monitoring process to ensure that the limiting parameters are not exceeded is included in Section 2. 1 .

The limiting parameters are classified into four categories: time at temperature, temperature, stress, and displacement. A summary of these limiting parameters is provided in Table 1 .8.A-1. Each category and limiting parameter is described in more detail below.

1.8.B Time at Temperature Limits The limiting parameters associated with time at temperature are to satisfy the material property recovery requirements, and to limit the potential for creep and other forms of elevated temperature metallurgical degradation.

As shown in Table 1.4.B-1 and Figure 1.4.B-1 of Section 1.4, if the RV beltline were to be held at greater than or equal to 800°F for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or more, then adequate annealing recovery is projected for the RV beltline materials based on the annealing equations provided in NUREG/CR-6327. This recovery is sufficient to allow continued safe operation of the Palisades reactor vessel to at least March 2011 . The target annealing time, however, will be one week ( 1 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br />).

ASME Code Case N-557 indicates that for metal temperatures exceeding 900°F but less than 940°F and with less than 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> of hold time vessel steels will not undergo significant time dependent behavior such as creep. Similarly for temperatures greater than 850°F but less than 900°F a hold time less than 1000 I

hours will provide the same assurance. These criteria were established based on the desire to allow the time-independent stress intensity (Sm) criteria to be used in lieu of the time-dependent stress intensity (St) criteria. These criteria are not material limitations but are conservative in that they are in conformance with ASME Code allowables.

Thus three time at temperature limiting parameters have been established as indicated in Table 1.8.A-1. One, the RV beltline must be equal to or greater than 800°F for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during the hold or soak period. Second, the TAR 2/2/96 1 .8-1

• reactor vessel cannot exceed 900°F for greater than 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />. And third, the reactor vessel cannot exceed 850°F for greater than 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />.

1.8.C Temperature Limits The limiting parameters associated with temperature are to ensure that potential degradation of specific equipment, components, or structures does not occur.

A maximum temperature has been established to ensure that the material of the in-service reactor vessel would not be subject to significant time dependent material degradation due to annealing temperatures. Using ASME Code Case N-557, this maximum temperature is 940°F. Again as with the time at temperature criteria, the 940°F was based on the desire to allow the time-independent stress intensity (Sm) criteria to be used in lieu of the time-dependent stress intensity (St) criteria.

Secondly, the 94'0°F temperature is sufficiently high to provide for annealing recovery.

At a concrete temperature not exceeding 250°F, it was determined in Section 1.3.C that the biological shield wall concrete will still perform its design basis function. This value was derived by comparing the worst-case high-temperature exposure ~ffects on concrete strength (reduction of concrete compressive strength) to the minimum expected concrete strength gain over time (increase in concrete compressive strength) and showing that there was no significant net effect.

Thus two temperature limiting parameters have been established as indicated in Table 1.8.A-1. The reactor vessel temperature cannot exceed 940°F, and the biological shield wall concrete cannot exceed 250°F.

1.8.D Stress Limits The limiting parameters associated with induced stresses caused by the annealing process are intended to avoid harmful permanent set and the potential for ductile flaw growth.

ASME Code Case N-557 requires that mechanical and thermal loadings on the reactor coolant pressure boundary and its supports be evaluated and provides an acceptable methodology using the stress categories and limits of stress intensities provided in Figure 1 of this code case. In addition the allowable stress intensity values given in Table 1 of this code case are to be used. In the annealing process the dominant stresses are temperature related and are self limiting in nature. Such secondary stresses when combined with primary stresses as shown in Figure 1 of Code Case N-557 have an allowable value of 3Sm. This ASME stress limit (3Sm) is based on an operational range of stress intensities to prevent component damage due to a ratcheting effect resulting from cyclical service loadings. The annealing operation which is analogous to a local heat treatment is essentially a construction activity with a very limited number of occurrences. A 3Sm limit based on a TAR 2/2/96 1.8-2

• maximum calculated stress intensity is therefore conservatively appropriate. The ASME Code Case N-557 does not require the peak stress to be evaluated for the annealing loading.

Thus a limiting parameter for stress has been established as shown in Table 1.8.A-1 . The primary and secondary stress intensities during the annealing operation will be maintained below the 3Sm limit.

1.8.E Displacement Limits The limiting parameters associated with displacements are intended to prevent significant interferences from occurring due to the movement of components caused by thermal expansion.

In Section 1.3.D. 7, it is described that the steam generators are mounted on sliding bases that allow horizontal motion parallel to the hot leg due to thermal expansion or contraction of the PCS loop piping. Hard stops, as discussed in Section 1 . 3. D. 7, are installed to prevent excess horizontal motion. The hard stops are located on supports that are embedded in concrete and which are not readily accessible for temporary modifications during the anneal. Interference with the horizontal motion of the steam generators could damage the supports. The hard stops allow for 1.99 inches of motion for Steam Generator A and for 1.41 inches of motion for Steam Generator B.

Thus one limiting parameter on displacement has been established as indicated in Table 1.8.A-1. The horizontal motion of the steam generators is limited to 1.99 inches for Steam Generator A and 1 .41 inches for Steam Generator B.

1.8.F References Eason, E., et al, "Models for Embrittlement Recovery Due to Annealing of Reactor Pressure Vessel Steels", NUREG/CR-6327, February 1995.

ASME Code Case N-557, "In-Place Dry Annealing of a Nuclear Reactor Vessel,Section XI, Division 1 ", Cases of the ASME Boiler and Pressure Vessel Code.

TAR 2/2/96 1.8-3

LIMITING PARAMETER I LOCATION LIMITING PARAMETER I REASON FOR LIMITATION CONDITION CONDITION CATEGORY Time at RV Beltline Soak period of > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Regions of the beltline must exhibit Temperature above 800°F sufficient material property recovery.

Time at Reactor Vessel Cannot exceed 900°F for Precludes a significant time dependent Temperature greater than 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> behavior {creep).

Time at Reactor Vessel Cannot exceed 850°F for Precludes a significant time dependent Temperature greater than 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> behavior {creep).

Temperature Reactor Vessel Cannot exceed 940°F Precludes a significant time dependent behavior {creep).

Temperature Biological Cannot exceed 250°F Precludes potential degradation of the Shield Wall biological shield wall concrete.

Concrete Stress Reactor Maintain primary plus Precludes harmful permanent set and the Vessel, PCS secondary stress intensities potential for ductile flaw growth.

Piping below the 3Sm limit.

Displacement Steam SG "A" displacement limit of The steam generators have permanent hard Generators 1. 99 inches and SG "B"

• stops for horizontal motion which when displacement l!mit of 1 .41 exceeded could result in damage to the inches. supports.

Table 1.8.A-1 Limiting Parameters for the Anneal Without Need for Further Analysis and Testing 1.8-4

• ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 THERMAL ANNEALING REPORT SECTION 2 REOUALIFICATION INSPECTION AND TEST PROGRAM SECTION 2.2 INSPECTION PROGRAM TO REOUALIFY THE REACTOR VESSEL 9 Pages

2.2 INSPECTION PROGRAM TO REOUALIFY THE REACTOR VESSEL This section describes the inspection program that will be implemented to affirm that the annealing operation has not damaged the reactor vessel or related equipment, components or structures.

2.2.A Description of the Inspection Program An inspection program of the critical regions of the reactor vessel, piping and other key components and equipment will be completed prior to and at the end of the annealing cycle.

The inspection program that will be used following the annealing operation will be based on three inspection methodologies: visual, ultrasonics, and dimensional.

Most of the inspections will be concentrated on the reactor vessel, nozzles and safe ends with general inspections within the reactor cavity.

If differing results are obtained from the monitoring, as described in Section 2.1, then consideration will be given to additional post-anneal inspections.

2.2.A.1 Visual Inspection Program The inspection program will include pre and post anneal visual examination of critical regions within the reactor vessel (including the flow skirt, core stabilizing

. lugs, core support lugs, vessel cladding, nozzles, and the surveillance capsule wall holder attachment welds), the biological shield reactor cavity liner plate and the vessel insulation.

All 10-year in-service visual examination requirements for the vessel will be satisfied by the post-anneal examination as well.

The examination requirements, number of examinations, qualification requirements, reporting criteria, and acceptance criteria will be in accordance with the ASME Code Section XI, 1989 Edition. Any non-Section XI examinations will use the ASME Code Section XI, 1989 Edition qualification requirements, reporting criteria, and acceptance criteria as guidelines.

The Section XI and non-Section XI visual examinations will be conducted by one of two different methodologies, VT-1 or VT-3, which are defined in Section XI.

A list of the planned visual inspections and acceptance criteria is provided in Table 2.2.A-1.

TAR 2/2/96 2.2-1

• 2.2.A.2 Ultrasonic Inspection Program The post-anneal ir:l-service examination program will include essentially 100%

ultrasonic examination of the accessible reactor vessel welds, beltline base material, reactor vessel nozzle welds, reactor vessel nozzle to safe end welds, reactor vessel safe end to pipe welds, and the required longitudinal pipe welds associated with the safe end welds. All 10-year in-service examination requirements for the vessel will be satisfied by this examination as well.

The examinations will be performed in accordance with the 1 989 Edition of Section XI of the ASME Code as augmented by USNRC Regulatory Guide 1 ..150. The examination requirements, examination techniques, number of examinations, qualification requirements, reporting criteria, and acceptance criteria will be in accordance with the ASME Code Section XI, 1989 Edition. Examination of the beltline base material has been added beyond that required by the referencing Code to ensure that the beltline region of the vessel has not been degraded. A description of the planned ultrasonic examinations is provided in Table 2.2.A-2.

2.2.A.3 Dimensional Inspection Program Distortion of the vessel due to annealing, which would be sufficient to preclude internals installation or which would require rework of the vessel and/or the internals, is considered extremely unlikely. Nevertheless, selected dimensional inspections of the reactor vessel will provide assurance that reassembly can proceed without diffic~lty following the anneal. The key areas of interest are:

• Coplanarity of points on the reactor vessel 0-ring seating surface;

• Head alignment pin positional changes;

• Ovality of the RV flange head seating surface counterbore;

• Ovality of the CSB support ledge counterbore;

• Reactor vessel outlet nozzle to nozzle distance and contour;

• Core support lug elevations;

• Core stabilizing lug azimuthal positions; and

• Internals alignment keyway pattern.

The annealing project measurement tooling will be used to verify that no significant distortion has occurred in critical *areas of the reactor vessel. The areas of concern are primarily related to the fit ups of other components to the vessel and the customized machining which was performed during plant construction. A TAR 2/2/96 2.2-2

description of the planned dimensional inspections and criteria is provided in Table 2.2.A-3.

In order to assess-the effect of the annealing process on the mechanical features of the vessel and its related components, a baseline set of data will be taken prior to annealing. A set of calibration fixtures will also be used to verify the repeatability of the tooling. The post-anneal measurements will be made with the tool in the same. configuration it was during the baseline measurements. This will simplify the comparison of the two sets of data. The results of this comparison will lead to a determination whether the annealing process affected the mechanical features of the vessel.

Should any adjustments be required to the vessel or other components, they will be performed to satisfy the necessary criteria for proper fit up. Based on experience with loop stress relief during initial construction at some plants, jacking of loop components may be required to remove elastic ovalization of the RV flange. This condition is caused by the frictional forces at the steam generator sliding bases restraining the upper RV flange radial displacement. The measurements of the pre-anneal and post-anneal positions of each steam generator sliding base combined with Section 2.1 PCS piping displacement monitoring will be utilized with the results of vessel dimensional measurements to determine jacking needs.

2.2.B Reporting & Acceptance Criteria For typical Section XI examinations the characterization and evaluation of indications will be in accordance with Article IWA-3000, Standards for Examination Evaluation, of the ASME Code Section XI, 1989 Edition. For non-Section XI examinations the characterization and evaluation of indications or conditions will use the requirements of the ASME Code Section XI, 1989 Edition as a guideline.

Specific acceptance criteria are delineated in Tables 2.2.A-1 and 2.2.A-2 which are ASME Code Section XI requirements.

Critical reactor vessel to internals interface data will be evaluated using detailed drawing requirements that define the acceptance criteria for the various dimensional measurements delineated in Table 2.2.A-3.

The certification of the annealing process will contain a summary of the examination results and of the evaluation and disposition of any indications resul~ing from the annealing operation. This will include:

• Pre-anneal and post-anneal in-service visual examination results;

• Post-anneal in-service ultrasonic examination results;

• Pre-anneal and post-anneal dimensional measurement results; and

• Adjustments required to the vessel or other components.

TAR 2/2/96 2.2-3

Description Exam Exam Exam Locations and Extent Acceptance Category Item Type Criteria Reactor Vessel:

Vessel Interior Interior Attachments Within B-N-1 B-N-2 B13.10 B13.50 VT-3 VT-1 The areas in the spaces above and below the reactor core position.

All surveillance IWB-3520.2 IWB-3520.1 Belt Line Region capsule wall holder welds.

Interior Attachments Beyond B-N-2 B13.60 VT-3 All core stabilizing lug IWB-3520.2 Belt Line Region welds, core support lug welds and flow skirt welds.

Interior Attachments Beyond Note 1 Note 1 VT-1 All core stabilizing lug IWB-3520.1 Belt Line Region welds, core support lug welds and flow -

skirt welds.

TABLE 2.2.A-1 Pre-Anneal and Post-Anneal Visual Examination Program (Page 1 of 2)

TAR 2/2/96 2.2-4

Description Exam Exam Exam Locations and Extent Acceptance Category Item Type Criteria Insulation:

Vessel Insulation Exterior Note 1 Note 1 VT-3 Reactor cavity floor IWB-3520.2 insulation covering directly below the vessel, the vessel shell insulation up to and including the lower vessel nozzle insulation.

See Note 2.

Biological Shield:

Reactor Cavity Note 1 Note 1 VT-3 The biological shield IWB-3520.2 liner plate surface up to the vessel transition area.

See Note 2.

Notes 1. Not an ASME Code Section XI examination. Inspection added as a supplemental examination to satisfy criteria in 1 OCFR50.66. Indication/condition characterization and acceptance criteria of ASME Code Section XI used as a guideline.

2. Only limited areas will be examined due to ALARA and access concerns at the locations identified.

Inaccessible and unexamined areas will be identified in the examination report.

TABLE 2.2.A-1 Pre-Anneal and Post-Anneal Visual Examination Program {Page 2 of 2)

TAR 2/2/96 2.2-5

Description Exam Exam Exam Locations and Extent Acceptance Category Item Type Criteria Reactor Vessel:

Shell Welds:

Circumferential Longitudinal Nozzle Welds:

Nozzle to Vessel B-A B-A B-D 81 .11 81 .12 83.90 UT UT UT 3 locations, full volume 9 locations, full volume 6 locations, full volume IWB-3510 IWB-3510 IWB-3512 Nozzle Inner Radius B-D 83.100 UT 6 locations, near IWB-3512 surface Flange Welds:

Shell to Flange B-A 81 .30 UT 1 location, full volume IWB-3510 Bottom Head Welds:

Circumferential B-A 81 .21 UT 1 location, full volume IWB-3510 Meridional B-A 81 .22 UT 6 locations, full volume IWB-3510 Belt Line Base Metal:

132 inches of Belt Line Height Note 1 Note 1 UT Essentially 100%, near surface including cladding IWB-3510 TABLE 2.2.A-2 Post-Anneal Ultrasonic Examination Program (Page 1 of 2)

TAR 2/2/96 2.2-6

Description Exam Exam Exam Locations and Acceptance Category Item Type Extent Criteria Primary Coolant Piping Pipe Welds Adjacent to R/V Nozzles:

Longitudinal Pipe Welds 8-J 89.12 UT 8 locations, full IW8-3514 volume. (Note 2)

Transition Piece to Nozzle 8-J 89.11 UT 6 locations, full IW8-3514 volume (Note 2)

Elbow to Transition Piece 8-J 89.11 UT 4 locations, full IW8-3514 volume (Note 2)

Transition Piece to Pipe 8-J 89.11 UT 2 locations, full IW8-3514 volume (Note 2)

Notes 1. Not an ASME Code Section XI examination. Inspection added as a supplemental examination to satisfy criteria in 1 OCFR50.66. Indication/condition characterization and acceptance criteria o.f ASME Code Section XI used as a guideline.

2. Includes O.D. equivalency examinations in lieu of O.D. surface examinations.

TABLE 2.2.A-2 Post-Anneal Ultrasonic Examination Program (Page 2 of 2)

TAR 2/2/96 2.2-7

Me~surement Reason for Measurement Number of Measurements Basis for Number of Measurements 1 Reactor Vessel Verify changes are Approximately every 30 Minimum reasonable Flange Mating insignificant with respect degrees around* the number to determine Surface Flatness to sealing capability of the periphery at inboard and flange sealing surface surfaces. outboard seal locations. contour.

2 Head Guide Pin Positional Changes (cross vessel measurement)

Verify no significant changes to guide pin positions. Pins are required for reactor internals and head Two close to the vessel flange, two higher.

Checks the span and tilt of the guide pins for changes.

• reinstallation and are located over the outlet nozzles where pipe loadings act.

3 Closure Head Seating Verify clearance required Four places, 90 degrees Four will show increase Surface Counterbore for reassembly. Nominal apart. .or decrease, if any, with Ovality .050 inch radial gap with sufficient detail to closure head. evaluate ovality.

4 Internals Support Ledge Counterbore (may be dropped if 3 shows little or no change)

Verify adequate radial clearance with UGS hold down device, and CSB support flange.

Four places, 90 degrees apart.

Four will show increase or decrease, if any, with sufficient detail to evaluate clearance changes.

TABLE 2.2.A-3 Dimensional Inspection Program {Page 1 of 2)

TAR 2/2/96 2.2-8

Measurement Reason for Measurement Number of Measurements Basis for Number of Measurements 5 Vessel Inner Verify proper clearances Four per nozzle around For determining vessel Diameter at Outlet with mating CSB nozzles. periphery of the nozzle. diameter changes and tilt Nozzles across the outlet nozzles.

6 7

Core Support Lug Elevations Relative to Vessel Flange Core Stabilizing Lug Verify absence of axial residual changes. Verifies clearances with bottom of CSB.

Verify post-anneal pattern Three out of nine core support lugs.

One pattern of six core Determine ample clearance with the CSB bottom; three is sufficient.

All six must be correct for Pattern for CSB-installation. stabilizing lugs. CSB installation. Relation Nominal gap is 0.040 inch to upper elevation will total at each lug. determine offset, if any.

8 Alignment Key Slot Verify CSB keys will One pattern of four slots. Key ways serve as Pattern engage slots. reference point for other measurements.

9 Steam Generator Utilize with the results of One location for each One measurement each, Sliding Base Position vessel measurements to steam generator. parallel to the hot leg determine jacking needs. combined with Section 2.1 PCS piping displacement monitoring will provide sufficient information.

TABLE 2.2.A-3 Dimensional lnspe.ction Program (Page 2 of 2)

TAR 2/2/96 2.2-9

• ATTACHMENT 3 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 THERMAL ANNEALING REPORT SECTION 2 REQUALIFICATION INSPECTION AND TEST PROGRAM SECTION 2.3 TESTING PROGRAM TO REQUALIFY THE REACTOR VESSEL 3 Pages

2.3 TESTING PROGRAM TO REQUALIFY THE REACTOR VESSEL 2.3.A Description of the Test Program The strategy used to determine the need to conduct special testing to verify that annealing has not degraded the reactor vessel or systems, structures and components (SSC) effected by the annealing operation was based on an integrated approach. Sections 1.2 and 1.3 performed a comprehensive evaluation of the reactor vessel and effected SSC's to determine potential areas of concern. The analyses conducted in Section 1. 7 was utilized to confirm which of the potential concerns could be eliminated based on results of the analysis. The development of Sections 2.1 and 2.2 identified the required monitoring and inspections needed to insure that the bounds of the analyses and dimensional stability are maintained.

The development of this section has taken credit for these analyses, monitoring and inspections and determined that a very limited number of potential .concerns need to be verified by special te~ting. These tests are basically reactor vessel fit up tests which will provide assurance that vessel internals reassembly can be achieved. Additionally this section discusses several standard operational tests which will be conducted to provide additional assurance that primary coolant system integrity and SSC's functionality is verified; however, these tests are not considered unique to the annealing operations and therefore are not specifically required by 10CFR50.66.

The fit-up tests described below along with the evaluations performed in sections 1.2, 1.3, 1. 7 and the monitoring/inspection efforts described in sections 2.1 and 2.2 will affirm that the annealing operation has not degraded the reactor vessel, piping and appurtenances, and other key components and equipment.

A discussion of the program to demonstrate annealing effectiveness on the vessel is located in Section 3.1, Fracture Toughness Recovery Program.

2.3.A. 1 Equipment Fit Up Tests A pre-anneal and post-anneal test of the surveillance capsule wall holders will be performed. A gauge conservatively representing the envelope of the surveillance capsule will be inserted into each holder planned for future use and any abnormal obstructions or resistance will be recorded. The same test will be performed after annealing to det.ermine if any distortion or other damage occurred. The test can be performed from the operating floor elevation in conjunction with other vessel measurements.

In view of the dimensional verification measurements discussed in Section 2.2.A.3 undetected distortion of the reactor vessel due to annealing, which would be sufficient to preclude internals installation or which would require rework of the vessel and/or the reactor internals, is considered unlikely. Nevertheless, successful TAR 2/2/96 2.3-1

installation of the core support barrel into the reactor vessel will provide assurance that reassembly can proceed in accordance with the normal refueling procedures.

2.3.B Other Testing Performance of the following tests are not required to recertify the vessel for reuse per 1OCRF50.66. The tests are discussed for information to show how required standard testing will be utilized to provide additional assurance that annealing did not have any detrimental effects on plant structures, systems or components.

2.3.B.1 System Tests Primary coolant system (PCS) pressure testing at 250 pounds will be performed during heat-up from cold shut down to hot shut down in accordance with the plant general operating procedure. Execution of the heat-up checklist will include a thorough leak check of accessible equipment. This includes PCS equipment such as the reactor vessel head, piping loops and the biological shield cooling system.

A second PCS system leak test which includes a cpmplete walk down of the PCS will be performed at operating temperature and pressure in accordance with the ASME Code Section XI, 1989 Edition.

• The biological shield cooling system performance and cooling coil leak rate will be monitored during annealing and plant heatup to identify any changes resulting from annealing.

The plant general operating procedure also provides for the on-line monitoring of closure head flange inner and outer seal leakage. This system has alarms that are initiated by sensing pressure between the seals or detecting leakage past the outer seal.

During normal plant operation neutron noise monitoring will be utilized to develop a comparison between the noise characteristics after the anneal operation and the previous operating noise data base. This comparison will provide additional assurance that the annealing operation did not affect reactor vessel internal equipment fit ups or core support barrel vibration amplitude and frequency.

2.3.B.2 Instrumentation Tests The post-anneal functioning of instrumentation either left in place or removed during the annealing such as RTD's in PCS piping and excore start up, wide range and power range neutron detectors located in the biological shield wall will be verified operable via the normal surveillance program established for the equipment.

TAR 2/2/96 2.3-2

- -=~ ====-=-

• ~

2.3.B.3 Structures Tests The refueling cavity liner leak rate will be monitored during post annealing refueling operations to identify any changes from existing known leakage due to the loads associated with the combined storage of the reactor internals and shielding or distortions due to the annealing thermal loads.

2.3.C Reporting The results of the fit up tests will be included in the recertification report.

Any anomalies during plant heat-up and normal operations will be reported to the NRC in accordance with the operating license requirements.

2.3.D Qualification Requirements Reactor re-assembly fit up tests will be performed in accordance with approved refueling procedures. Any tests involving the manipulation of plant equipment during plant heat-up will be performed with approved procedures by CPCo plant operators.

TAR 2/2/96 2.3-3