ML18065A350

Forwards Preliminary Thermal Annealing Rept,Thermal Annealing Operating Plan Sections 1.1, General Considerations & 1.2, Description of Rv, in Format Recommended by Section C.1 of Draft Rg DG-1027

ML18065A350
Person / Time
Site:	Palisades
Issue date:	12/12/1995
From:	Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9512180238
Download: ML18065A350 (75)

Text

consumers Power POWERIN&

/llllCHl&AN"S PRO&RESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 December 12, 1995 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 1.1, GENERAL CONSIDERATIONS, AND SECTION 1.2, DESCRIPTION OF THE REACTOR VESSEL 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 third of those submittals.

Attachment 1 to this letter contains the Thermal Annealing Operating Plan Section 1.1, General Considerations. Attachment 2 to this letter contains Section 1.2, Description of the Reactor Vessel. The attached information is presented in the format recommended by Section C. l of DG-1027.

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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 Attachments

• h.

ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 THERMAL ANNEALING REPORT SECTION 1 THERMAL ANNEALING OPERATING PLAN SECTION 1. 1, GENERAL CONSIDERATIONS

• 1. THERMAL ANNEALING OPERATING PLAN

1. 1 IDENTIFICATION OF GENERAL CONSIDERATIONS

1. 1.A Identification of Reactor Vessel for Annealing This thermal annealing report (TAR) is written for the reactor vessel (RV) of the Palisades Nuclear Plant, located on the eastern shore of Lake Michigan near, the town of Covert, Michigan. Palisades Nuclear Plant, which began commercial operation on December 31, 1971, is currently licensed for operation by Consumers Power Company (CPCo) (Docket Number 50-255) through March 14, 2007.

1.1.B Reasons to Anneal Reactor Vessel The Palisades Nuclear Plant reactor vessel is fabricated from rolled plates (SA-302, Grade 8, Modified) welded both circumferentially and axially. The region of the reactor vessel corresponding to the active height of the core (the RV beltline) is subject to embrittlement by neutron irradiation. The RV beltline region consists of an intermediate and a lower shell course, each with three plates and three connecting axial (longitudinal) welds. The two shell courses are welded together by one continuous (360°) circumferential weld. The reactor vessel is described in further detail in Section 1.2. These plates and welds must continue to meet reactor

• vessel integrity requirements specified by the U.S. Nuclear Regulatory Commission (NRC) throughout the reactor operating life.

The NRC status report on reactor pressure vessel integrity (NUREG-1511, December 1 994), notes that the limiting beltline material for Charpy upper shelf energy (USE) is plate D-3804-1, Heat C-1308 (330 ° to 90 ° in the lower shell).

However, the Charpy USE of this limiting plate is not expected to fall below the 50 ft-lb requirement per 10 CFR Part 50, Appendix G, through the licensed life of the vessel.

The NUREG-1511 status report also noted that the Palisades limiting beltline axial weld (Heat W5214) would exceed the NRC's screening criteria for pressurized thermal shock (PTS), 10 CFR Part 50.61, three years prior to the end-of-license (EOL). To demonstrate that the Palisades reactor vessel would be below the PTS screening criteria at EOL (currently 2007; 2011 if the construction period is recovered), CPCo committed to:

( 1) gather additional materials properties data from its retired steam generators; (2) institute a surveillance program which would include the limiting weld; (3) evaluate annealing of the reactor vessel; and

• TAR 12/11 /95 1 . 1-1

• (4) consider instituting an ultra low leakage fuel strategy .

The retired steam generators have shell welds fabricated with the same heat number of weld wire and flux type used in the fabrication of the RV beltline axial welds. During the Fall of 1 994, CPCo performed material property tests and chemical analyses of newly acquired samples of weld material. from the shells of those retired steam generators. It was expected that the material property test results would prolong the time required to reach the PTS screening criteria by providing more realistic material data. However, these tests and analysis indicated that the degree of embrittlement of the Palisades reactor vessel may be higher than previously calculated. With the steam generator material data included in the PTS evaluation, the Palisades reactor vessel would reach the PTS screening criteria at the end of the 14th refueling outage, scheduled for late 1999.

The NRC staff reviewed the chemistry data and noted that a large variability existed in the reported copper and nickel contents for the limiting RV axial weld.

To assess this concern, the NRC staff performed RV failure frequency calculations using the Palisades plant-specific chemistry and fluence uncertainties. These calculations were similar to those in SECY-82-465, which established the basis for the PTS screening criteria. The results of the Palisades plant specific calculations, which included the observed variability in chemistry data, demonstrated that the margins of safety intended by the PTS screening criteria would in fact be satisfied

• through the 14th refueling outage .

The NRC issued a safety evaluation report (TAC No. M83227) to CPCo in April of 1995. As a result of the safety evaluation, the NRC requested that, in accordance with 10 CFR 50.61, CPCo submit a plant specific analysis three years prior to the date when the limiting material is projected to exceed the PTS screening criteria.

The same transmittal indicated that submission of plans to thermally anneal the reactor vessel would obviate the requirement for a plant specific analysis. Thus, this thermal annealing report pr9vides the required information in accordance with the format and content outlined in Draft Regulatory Guide DG-1 027.

1 . 1 . C Expected Remaining Operating Life After Annealing The objective of the Palisades RV annealing program is to assure sufficient toughness for all materials in the RV beltline region. The Palisades RV anneal is scheduled for the end of the thirteenth operating cycle (middle of 1 998). The nominal annealing conditions will be 850°F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. The anneal will provide adequate recovery of all irradiated vessel materials to allow continued safe operation through the current EOL and beyond. For the purposes of this thermal annealing report, the material properties of the beltline materials have been projected through March 2011. This date represents an extension beyond the current EOL corresponding to the recovery of the construction period .

• TAR 12/11 /95 1.1-2

A detailed characterization of the RV materials is given in Section l. 2 along with projected fluences at each location. A summary of the RV beltline material properties is given in Tables 1.1.C-1 and 1.1.C-2. Table 1.1.C-1 characterizes the vessel "best-estimate" beltline material properties for Palisades prior to annealing

{i.e., at the end of cycle 13). The fluences listed in Table 1.1.C-1 are the peak values at the vessel inside diameter (ID). The irradiation induced shifts in RT NDT were calculated using the procedures outlined in Reg. Guide 1 .99, Rev. 2. This calculation is summarized in Table 1.1 .C-2. Using the transition temperature recovery model in NU REG CR-6327, the projected recovery (as a percentage) for each material was calculated. An irradiation temperature of 533°F was used for the NUREG CR-6327 calculations. These recovery calculations are summarized in Table 1.1.C-3. The projected reembrittlement trends after annealing are also shown in Table 1 .1 .C-3. These projections were made using the lateral shift model as indicated in Draft Regulatory Guide DG-1 027. This model is believed to give conservative results. Projections for the specific surveillance program materials are presented in Section 3.

The Charpy upper shelf energy {USE) responses to irradiation and annealing vvere determined using. similar procedures based on Regulatory Guide 1.99, Revision 2 and NU REG CR-6327. The Charpy USE projections were based on peak fluences at one-quarter thickness ( 1 /4 T) locations in the vessel. These calculations are summarized in Table 1.1.C-4. The Charpy USE after reirradiation was also

• • determined using the lateral shift model. The projected post anneal Charpy USE values for March 2011 are also listed in Table 1.1.C-4. Note that similar projections to these are presented in Section 3 for the specific surveillance program materials which are equivalent or bounding to the actual vessel materials.

The projected values for Charpy USE and RT NDT listed in Tables 1.1 .C-3 and 1.1.C-4 indicate that the pressure vessel will maintain adequate toughness beyond the EOL. These projections are based on the lateral shift embrittlement model, which is believed to give conservative estimates. Further refinement of these projections will be possible when surveillance data is generated. A description of the surveillance strategy is provided in Section 3.

1. 1.D Operating History The operating history of the Palisades reactor vessel from plant startup through the completion of cycle 11 is provided in Table 1.1.D-1. The reactor power generation in terms of total effective full power days (EFPD) of operation is provided for each fuel cycle. Also provided in Table 1.1.D-1 are typical reactor inlet temperatures representative of each fuel cycle. These inlet temperatures are representative of the RV wall temperature in the beltline region. *

• TAR 1 2/11 /95 1.1-3

1.1.E Surveillance Program There have been four Palisades surveillance capsules removed and tested in addition to preirradiation testing conducted in the 1970' s. The following sections provide a general summary of the surveillance program results. Significant parameters such as*transition temperature shift (~RT NDT) and changes in USE are presented. The surveillance program includes additional capsules and materials which will be used to accurately quantify the significant parameters mentioned above for pre-annealing and post-anneal operation.

1.1.E.1 Surveillance Program Configuration The Palisades reactor vessel surveillance program was designed by Combustion Engineering, Inc. to the requirements of ASTM E185-66. Preirradiation Charpy impact and tensile properties of the base metal, weld metal, and heat affected zone (HAZ) metal specimens for the Palisades surveillance program were determined in order to establish baseline data (Perrin and Fromm, 1977). Drop weight properties of base metal specimens were also determined during the preirradiation testing.

The Nil-Ductility Transition (NOT) temperature for the base metal was established from drop weight specimens tested in accordance with ASTM E208-69.

The loca1ion of the surveillance capsule assemblies within the Palisades Nuclear

• Plant reactor vessel are shown schematically in Figure 1.1.E-1. The letter designation preceding the numerical value indicates accelerated (A), wall (W), or*

thermal (T). The numerical value represents the azimuthal location in degrees. For example, Capsule A-240 corresponds to an accelerated capsule irradiated in the 240 degree azimuthal location.

1.1.E.1.1 Specimen Orientation and Compositional Analysis The Palisades RV base metal is fabricated from steel plate to ASTM Specification A-302 Grade B modified, with Class 1 mechanical properties. Surveillance specimens were prepared from one of the actual RV intermediate shell plates (D-3803-1, heat number C-1279-3). All the plate test specimens represent material taken from at least one plate thickness from any water quenched edge. Charpy specimens were machined from the plate in both the longitudinal (major axis of specimen is parallel to the principal rolling direction) and transverse (major axis of the specimen is perpendicular to the principal rolling direction) orientations (See Figure 1.1.E-2). Tensile specimens were machined from the plate with the major axis of the specimen parallel to the principal rolling direction. Charpy and tensile specimens from heat affected zone (HAZ) were oriented with the major axis of the specimens perpendicular to the welding direction. The root of the HAZ Charpy specimen notches was centered on the fusion line between the base metal and weld metal. For the weld metal specimens, the Charpy specimens were oriented

• with the major axis of the specimen perpendicular to the welding direction, while TAR 12/11 /95 1.1-4

the tensile specimens were oriented with the major axis of the specimen parallel to the welding direction.

The original chemical analyses of the surveillance test materials for the three plates and two welds that make up the surveillance program {Groeschel, 1971) are reported in Table 1.1.E-1. The base metal test material was fabricated from Plate No. D-3803-1. HAZ test material was fabricated by welding together intermediate shell plate numbers D-3803-2 and D-3803-3. Weld metal test material was fabricated by welding together intermediate shell plate numbers D-3803-1 and D-3803-2. These welds were fabricated using a submerged arc process, using RACO 3 weld wire heat 3277 and Linde 1092 flux. The process used two arcs and included nickel addition. Weld repairs were performed using E8018 rod.

1.1.E.1.2 Supplemental Surveillance Materials The original surveillance materials will be supplemented with additional weld material corresponding to the Palisades RV beltline welds {Heat No. W5214, 348009 and 27204). \/\Jeld specimens (Heat No. W5214 and 348009) have been fabricated using weld seams from the retired Palisades steam generators. The samples were taken from the higher copper and nickel regions of these welds.

Weld specimens corresponding to the Palisades circumferential weld {Heat No.

27204) are from a RV nozzle dropout. The nozzle dropout weld was manufactured using the same weld wire heat in the same shop as the Palisades circumferentrial weld. However, the Palisades weld used a Linde 124 flux versus th.e Linde 1092 flux used in the nozzle dropout. A comparison between the supplemental materials and the actual RV welds is provided in Table 1.2.C-1. These specimens will be combined with materials from the original surveillance program in supplemental irradiation capsules as part of the Fracture Toughness Recovery and Reembrittlement Assurance Program. This program is described in Section 3.

1.1.E.2 Pre-Irradiated Results 1.1.E.2.1 Drop Weight Tests and Initial RTNDT Drop weight test specimens were fabricated from Plate No. D-3803-1. The NOT temperature was defined as the highest temperature at which a specimen breaks, with a pair of specimens exhibiting "no break" behavior at a temperature 10°F higher. Duplicate drop weight tests were conducted at 10°F intervals from -30°F to 0°F {Perrin and Fromm, 1977). Based on the "break, no break" behavior, the NOT temperature for the base metal was -10 ° F. No drop weight tests were performed on the surveillance weld or HAZ materials. Using results from transverse Charpy samples, measured initial RT NOT for intermediate plate D-3803-1

{Heat No. C-1279-3) was -10°F. Initial RT NOT values were determined for the remaining beltline plate materials using longitudinal Charpy data in accordance with

• MTEB 5-2 .

TAR 1 2/11 /95 1.1.,5

• 1. 1.E.2.2 Charpy Impact Tests The preirradiation Charpy impact energy, lateral expansion, and fracture appearance values have been tabulated as a function of temperature. Of particular interest to reactor vessel integrity are the temperatures corresponding to the impact energies at 30 ft-lb and 50 ft-lb, as well as the USE level (ft-lb). Tables 1.1.E-2 through 1 . 1 . E-6 summarize the 30 ft-lb and 50 ft-lb transition temperature values and the USE values for the Palisades RV surveillance capsule material specimens tested to date. The changes in these values were obtained by comparing the unirradiated (Perrin and Fromm, 1977) and irradiated (Perrin et al., 1979; Kunka and Cheney, 1984; Peter et al., 1994; Stallman, 1988) Charpy impact energy values using the hyperbolic tangent curve-fit program CVGRAPH 4.0 (CVGRAPH, 1994). Charpy plots are presented to illustrate the changes in impact properties as

. a function of temperature for the base metal, weld metal and HAZ material.

Figures 1 . 1 . E-3 through 1 . 1 . E-7 present composite curves that correspond to the unirradiated and irradiated surveillance capsule results.

For the unirradiated surveillance testing program, test results were obtained from 1 6 base metal specimens with a longitudinal orientation, 1 5 base metal specimens with a transverse orientation, 18 weld metal specimens and 19 HAZ specimens.

Charpy tests of the additional. weld materials for the supplemental surveillance

• program were also completed. The data from these pre-irradiation tests are summarized in Figures 1.1.E-8, 1.1.E-9 and 1.1.E-10. This data was used as the basis for the design of the Fracture Toughness Recovery and Reembrittlement Assurance Program described in Section 3.

1.1.E.2.3 Standard Reference Material Some Palisades surveillance capsules contain standard reference material (SRM) in addition to actual RV base metal, weld metal and HAZ material. These SRM specimens are cut from "standard" steel plates that are similar in composition and heat treatment to the base material in reactor vessels. The specimens are intended to serve as a reference by comparing the radiation embrittlement of the reference material to the overall trend of SRM data from other surveillance capsules and to detect anomalies in the radiation environment of that surveillance capsule.

Palisades surveillance capsules A-60, W-110, W-80, and *W-260 contain specimens from the 12-inch thick SA-5338-1, Heavy Section Steel Technology Plate No. 1 (HSST-01) fabricated by Lukens Steel.

The HSST-01 specimens were inserted in a limited number of surveillance capsules at commercial reactor plant sites. A baseline Charpy energy curve-fit was generated using the tabulated data from Oak Ridge National Laboratory (Stallman, 1988). This baseline data tabulation was used to develop a hyperbolic tangent

• curve-fit for the HSST-01 material, using the computer program CVGRAPH and is TAR 12/11 /95 1.1-6

shown in the composite curve, Figure 1 . 1 . E-7. Additional details are provided below in the Capsule W-110 discussion.

1.1.E.3 Irradiated Results 1.1.E.3.1 Capsule A-240 Capsule A-240 was removed from the Palisades RV after 2.26 EFPY and analyzed (Perrin et al., 1979). The estimated neutron exposure level of 6.0 x 10 19 n/cm 2 (E

> 1.0 MeV) is presently being evaluated. Test results were obtained from 12 bas*e metal specimens with a longitudinal orientation, 12 base metal specimens with a transverse orientation, 12 weld specimens, and 1 2 *HAZ specimens. No additional chemical analyses were performed on any of the specimens from Capsule A-240.

The values presented in Tables 1.1.E-2 through 1.1.E-5 include the T 30 , LlT30 , T 50 ,

Ll T 50 , USE, and LlUSE results from surveillance Capsule A-240. The CV GRAPH curve-fitted results for the Capsule A-240 Charpy impact energy specimens are also included in the composite curves, Figures 1.1.E-3 through 1.1.E-6.

1.1.E.3.2 Capsule T-330 Capsule T-330 was the first thermal surveillance capsule removed from the

• Palisades reactor vessel. Capsule T-330 was located above the reactor core and was exposed only to the elevated temperature of reactor operation (i.e., no significant neutron flux). This capsule was removed after 5.21 EFPY and analyzed, (Kunka & Cheney, 1984). Test results were obtained from 11 base metal specimens with a longitudinal orientation, 12 base metal specimens with a transverse orientation, 1 2 weld metal specimens, and 1 2 HAZ metal specimens.

Additional chemical analyses were performed on irradiated Charpy specimens from Capsule T-330. The results were as follows:

Charpy Specimen Cr Cu Mn Mo Ni P Si 22J (Plate D-3803-1) 0.11 0.24 1.66 0.45 0.53 0.005 0.20 341 (Weld Ht. 3277) 0.056 0.30 1.20 0.52 1.38 0.014 0.25 460 (Weld/HAZ) 0.050 0.26 1.22 0.47 1 .19 0.015 0.24 46E (Weld/HAZ) 0.050 0.25 1.09 0.45 0.95 0.014 0.19 The values presented in Tables 1.1.E-2 through 1.1.E-5 include the results from thermal surveillance Capsule T-330. In addition, curve-fits for the Capsule T-330 Charpy impact energy specimens are included among the composite curves, Figures 1 . 1 . E-3 through 1 . 1. E-6 .

TAR 1 2/11 /95 1.1-7

1.1.E.3.3 Capsule W-290 Capsule W-290 was also removed from the Palisades reactor vessel after 5.21 EFPY and analyzed, (Kunka & Cheney, 1984). The capsule received an average fast neutron fluence of 1.09 x 1019 n/cm 2 (E > 1.0 MeV). Test results were obtained from 12 base metal specimens with a longitudinal orientation, 12 base metal specimens with a transverse orientation, 12 weld metal specimens, and 12 HAZ specimens. Additional chemical analyses were performed on irradiated Charpy specimens from Capsule W-290. The results were as follows:

Charpy Specimen Cr Cu Mn Mo Ni P Si 25J (Plate D-3803-1} 0.11 0.24 1.61 0.45 0.52 0.004 0.24 37C (Weld Ht. 3277) 0.050 0.25 1.28 0.51 1.60 0.013 0.20 The values presented in Tables 1.1.E-2 through 1.1.E-5 include the T 30 , ~T 30 , T 50 ,

~ T 50 , USE, and ~USE results from surveillance Capsule W-290. Curve-fits for the Capsule W-290 Charpy impact energy specimens are included among the composite curves, Figures 1.1.E-3 through 1.1.E-6.

1.1.E.3.4 Capsule W-110 Capsule W-110 was removed from the Palisades reactor vessel after 9.95 EFPY and analyzed, (Peter, 1994). The capsule received an average fast neutron fluence of 1. 779 x 1019 n/cm 2 (E > 1 .0 MeV). Test results were obtained from 12 base metal specimens with a longitudinal orientation, 1 2 weld metal specimens, and 1 2 HAZ specimens. Additional chemical analyses were performed on some of the specimens tested from Capsule W-110. The results were as follows:

Charpy Specimen Cr Cu Mn Mo Ni P Si 15T (Plate D-3803-1} 0.118 0.215 1.515 0.464 0.523 <0.013 0.134 15E (Plate D-3803-1} 0.112 0.215 1.504 0.464 0.496 <0.013 0.109 150 (Plate D-3803-1} 0.112 0.215 1.510 0.467 0.495 <0.013 0.157 3A4 (Weld Ht. 3277) 0.061 0.239 1.159 0.509 1.617 0.016 0.129 3A3 (Weld Ht. 3277) 0.048 0.231 1.098 0.509 1.502 0.017 0.133 3A7 (Weld Ht. 3277) 0.053 0.233 1.131 0.508 1.494 0.017 0.151 In addition to the base metal, weld and HAZ material, Capsule W-11 0 also contained 12 SRM Charpy specimens (i.e., HSST-01 ). The SRM results were TAR 12/11/95 1.1-8

within two standard deviations of the generic test data results, which indicates no anomalies in the Palisades operating conditions.

The values presented in Tables 1.1.E-2, 1.1.E-4 through 1.1.E-6 include the T 30 ,

11T30 , T 50 , 11T50 , USE, and /1USE results from surveillance Capsule W-110. Curve-fits for the Capsule W-110 Charpy impact energy specimens are also included among the composite curves, Figures 1.1.E-3, 1.1.E-5 through 1.1.E-7.

1.1.E.4 Projected Surveillance Program Annealing Response The projected surveillance material annealing response is detailed in Section 3.

1. 1.F REFERENCES ASTM Designation E185-66, "Surveillance Tests on Structural Materials in Nuclear Reactors," Book of ASTM Standards.

ASTM Designation E208-69, "Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," Book of ASTM Standards.

CPCo Engineering Analyses, 1990, EA-P-PTS-90-001, Covert, Ml

• Wrights, G. N., 1994, "Palisades Cycle 11 Reactor Vessel Inner Radius Fast Neutron Flux Data", Westinghouse Letter SE/REA-169/94 to R. Snuggerud, Pittsburgh, PA.

CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.0, developed by ATI Consulting, December 1994.

Perrin, J. S., and Fromm, E. 0., 1977, "Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties", Battelle Columbus Laboratories, Columbus, OH.

Perrin, J. S., Fromm, E. 0., Farmelo, D. R., Denning, R. S., Jung, R. G., 1979, "Palisades Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule A-240", BCL-585-12, Columbus, OH.

Groeschel, R. C., 1971, "Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Palisades Reactor Vessel Materials", CE Report No. P-NLM-019, Windsor, CT.

Kunka, H. K., and Cheney, C. A., 1984, "Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program", WCAP-10637, Pittsburgh, PA

• TAR 12/11 /95 1.1-9

Peter, P. A., Lippincott, E. P., Wrights, G. N., Madeyski, A., 1994, "Analysis of Capsule W-110 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program", WCAP-14014, Pittsburgh, PA.

RSIC DATA LIBRARY COLLECTION DLC-76, 1987, "SAILOR, Coupled, Self-Shielded, 47-Neutron, 20-Gamma-Ray, P3 , Cross-Section Library for Light Water Reactors".

• Stallman, F. W., 1988, "Analysis of the A302B and A533B Standard Reference Materials in Surveillance Capsules of Commercial Power Reactors", NUREG/CR-4947, Oak Ridge National Laboratory, Oakridge, TN

• TAR 12/11 /95 1.1-10

Beltline Material Heat No. Fluence at RV ID Initial RTNor Cu Ni n/cm 2 (Of) (wt%) (wt%)

(E > 1 MeV)

Shell Plate D-3803-1 C-1279-3 1.93 x 10 19 -10 0.232 0.51 Shell Plate D-3803-2 A-0313-2 1.93 x 10 19 -30 0.24 0.52 Shell Plate D-3803-3 C-1279-1 1.93x1019 0 0.24 0.50 Shell Plate D-3804-1 C-1308-1 1.93 x 1019 0 0.19 0.48 Shell Plate D-3804-2 C-1308-3 1.93 x 1019 '-30 0.19 0.50 Shell Plate D-3804-3 B-5294-2 1.93 x 1019 -25 0.12 0.55 Intermediate Shell W5214 1.45 x 1019 -56 0.212 1.02 Axial Welds 2-112A/C Lower Shell Axial W5214 1.45 x 1019 -56 0.212 1.02 Welds 3-11 2A/C Lower Shell Axial 348009 1.45 x 1019 -56 0.19 0.99 Welds 3-11 2A/C Intermediate/Lower Shell 27204 1.93 x 1019 -56 0.21 1.00 Circumferential Weld 9-11 2 TABLE 1.1.C-1 Prior to Annealing Properties of Palisades RV Beltline Material (Best-Estimate)

TAR 12/11 /95 1 . 1-11

Beltline Material Heat No. Initial Fluence Chemistry ARTNDT Margin RTNDT RTNDT Factor Factor (Of) (Of) (of)

(Of)

Shell Plate D-3803-1 C-1279-3 -10 1.18 155 183 34 207 Shell Plate D-3803-2 A-0313-2 -30 1.18 160 189 34 193 Shell Plate D-3803-3 C-1279-1 0 1.18 158 186 34 220 Shell Plate D-3804-1 C-1308-1 0 1.18 129 152 34 186 Shell Plate D-3804-2 C-1308-3 -30 1.18 131 155 34 159 Shell Plate D-3804-3 8-5294-2 -25 1.18 82 97 34 106 Intermediate Shell W5214 -56 1.10 232 256 66 266 Axial Weld 2-11 2A/C Lower Shell W5214 -56 1.10 232 256 66 266 Axial Weld 3-11 2A/C Lower Shell Axial Welds 348009 -56 1.10 219 242 66 252 3-112A/C ,

Intermediate/Lower Shell 27204 -56 1.18 229 270 66 280 Circumferential Weld 9-112 Note: RT NOT = Initial RTNoT + ~RT NOT + Margin TABLE 1.1.C-2 Prior to Annealing Properties of Palisades RV Beltline Material (Assumed RV Properties for Analyses)

TAR 12/11 /95 1 . 1-12

Material Pre-Anneal RT NDT @ Anneal Recovery RTNoT After Reirradiated RTNoT After fluence at RV ID (Of) (%) Anneal fluence at RV ID Reirradiation (n/cm 2 , E > 1MeV) (Of) (n/cm 2 , E > 1MeV) (Of)

Plate D-3803-1 1.93 x 1019 207 89 44 0.78 x 10 19 169 Plate D-3803-2 1.93 x 1019 193 88 27 0.78 x 1019 154 Plate D-3803-3 1.93 x 1019 220 88 56 0.78 x 1019 182 Plate D-3804-1 1.93 x 1019 186 92 46 0.78 x 1019 155 Plate D-3804-2 1.93 x 1019 159 92 16 0.78 x 1019 126 Plate D-3804-3 1.93 x 1019 106 96 13 0.78 x 10 19 85 Weld W5214 1.45 x 1019 266 90 36 0.62 x 1019 212 Weld 348009 1.45 x 1019 252 92 29 0.62 x 1019 201 Weld 27204 1.93 x 1019 280 90 37 0.78 x 1019 224 TABLE 1 . 1 . C-3 Projected RTNDT Annealing Response of Palisades RV Materials TAR 1 2/11 /95 1 . 1-1 3

Material Pre-Anneal CvUSE CvUSE 1 Predicted CvUSE1* Reirradiated CvUSE1.,

Fluence at 1 /4 T (ft-lb) (ft-lb) Recovery (ft-lb) Fluence at 1 /4 T (ft-lb) n/cm 2 (%) n/cm 2 (E > 1 MeVI (E > 1 MeV)

Plate D-3803-1 1.16 x 1019 102 68 100 102 0.47 x 10 19 76 Plate D-3803-2 1.16x1019 87 57 100 87 0.47 x 1019 63 Plate D-3803-3 1.16 x 1019 91 60 100 91 0.47 x 1019 66 Plate D-3804-1 1.16x10 19 72 51 lOb 72 0.47 x 1019 55 Plate D-3804-2 1.16x1019 76 54 100 76 0.47 x 10 19 58 Plate D-3804-3 1.16 x 10 19 73 57 100 73 0.47 x 1019 60 Weld W5214 0.87 x 1019 118 78 95 116 0.38 x 1019 84 Weld 348009 0.87 x 1019 111 76 100 111 0.38 x 1019 82

. Weld 27204 1 .16 x 10 19 98 63 100 98 0.47 x 1019 69 CvUSE Unirradiated Charpy upper shelf energy CvUSE1 Irradiated Charpy upper shelf energy (prior to anneal) estimated from Regulatory Guide 1.99, Revision 2 based upon best estimate reactor vessel material copper chemistry values in Table 1.1.C-1 Predicted Recovery Percent Charpy upper shelf energy recovery estimated using NUREG/CR-6327 CvUSE 1* = Predicted Irradiated-Annealed Charpy upper shelf energy CvUSE1., Predicted Charpy upper shelf energy after reirradiation estimated using the lateral shift in Draft Regulatory Guide DG-1027 TABLE 1.1.C-4 Projected CvUSE Annealing Response of Palisades RV Materials TAR 12/11/95 1.1-14

Cycle Length Inlet Temperature Cycle (EFPD) (OF) 1 379.4 523 2 449.1 525 3 349.5 535 4 327.6 537 5 394.6 536 6 333.4 536

• 7 8

9 369.9 373.6 298.5 536 537 533 10 356.9 533 11 430.4 533 TABLE 1.1.D-1 Operating History of the Palisades RV Cycles 1 through 11

• TAR 12/11 /95 1.1-15

Element 03803-1 1 03803-2 03803-3 03803-3/03803-2 2 Weld 03803-2/03803-1 3 .4 Weld Plate Plate Plate @2in. @2in.

Weld Wire Heat 3277 Weld Wire Heat 3277 Root Face Root Face Si 0.23 0.32 0.24 0.24 0.25 0.25 0.22 s 0.019 0.021 0.020 0.009 0.010 0.010 0.010 p 0.011 0.012 0.010 0.011 0.012 0.011 0.011 Mn 1.55 1.43 1.56 1.08 1.03 1.01 1.02 c 0.22 0.23 0.21 0.098 0.080 0.088 0.086 Cr 0.13 0.42 0.13 0.05 0.04 0.05 0.03 Ni 0.53 0.55 0.53 0.43 1.28 0.63 1.27 Mo 0.58 0.58 0.59 0.54 0.53 0.55 0.52 Al 0.037 0.022 0.037 Nil Nil Nil Nil v 0.003 0.003 0.003 Nil Nil Nil Nil Cu 0.25 0.25 0.25 0.25 0.20 0.26 0.22 Notes:

1 Plate 03803-1 was used to fabricate the base metal specimens 2 Weld 03803-3/03803-2 was used to fabricate HAZ metal specimens 3 Weld 03803-2/03803-1 was used to fabricate weld metal specimens 4 Several additional chemistry measurements have been performed on the surveillance weld.

TABLE 1.1.E-1 Chemical Composition of the Original Surveillance Materials for Palisades Reactor Vessel TAR 12/11 /95 1.1-16

Capsule Fluence, x 30 ft-lb l1T30 50 ft-lb 11Tso Upper Shelf Decrease in 10 19 n/cm 2 Transition (oF) Transition (oF) Energy USE (E > 1 MeVJ Temperature Temperaturn (ft-lb) (ft-lb)

(oF) (oF)

Un irradiated 0.0 0 0 25 0 155 0 A-240 6.0* 204 204 243 218 93 62 T-330 0.0 5 5 34 9 172 0 W-290 1.09 176 176 198 173 112 43 W-110 1.779 179 179 203 178 103 52

• This value is presently being evaluated.

TABLE 1.1.E-2 Transition Temperature and USE Results for Palisades RV Surveillance Program Base Metal (LT) Plate No. 03803-1, Heat No. C-1279-3 from CVGRAPH TAR 12/11 /95 1.1-17

Capsule Fluence, x 30 ft-lb AT30 50 ft-lb AT so Upper Shelf Decrease in 10 19 n/cm2 Transition (oF) Transition (OF) Energy USE (E > 1 MeV) Temperature Temperature (ft-lb) (ft-lb)

(oF) (oF)

Unirradiated 0.0 18 0 49 0 102 0 A-240 6.0* 212 194 265 216 68 34 T-330 0.0 19 1 56 7 108 0 W-290 1.09 176 158 202 153 84 18

• This value is presently being evaluated.

TABLE 1.1.E-3 Transition Temperature and USE Results for Palisades RV Surveillance Program Base Metal (TL) Plate No. 03803-1, Heat No. C-1279-3 from CVGRAPH TAR 12/11/95 1.1-18

Capsule Fluence, x 30 ft-lb l1T30 50 ft-lb l1Tso Upper Shelf Decrease in 10 19 n/cm 2 Transition (oF) Transition (oF) Energy USE (E > 1 MeV) Temperature Temperature (ft-lb) (ft-lb)

(oF) (oF)

Un irradiated 0.0 -87 0 -47 0 118 0 A-240 6.0* 255 342 390 437 51 67 T-330 0.0 -63 24 -19 28 132 0 W-290 1.09 199 286 254 301 63 55 W-110 1.779 218 305 306 353 58 60

• This value is presently being evaluated.

TABLE 1.1.E-4 Transition Temperature and USE Results for Palisades RV Surveillance Program Weld Metal Heat 3277 from CVGRAPH TAR 12/11 /95 1.1-19

Capsule fluence, x 30 ft-lb i1T3a 50 ft-lb fl.Tso Upper Shelf Decrease in 10 19 n/cm 2 Transition (Of) Transition (Of) Energy USE (E > 1 MeV) Temperature Temperature (ft-lb) (ft-lb)

(Of) (Of)

Unirradiated 0.0 -89 0 -49 0 116 0 A-240 6.0* 184 273 276 325 59 57 T-330 0.0 -40 49 -9 40 119 0 W-290 1.09 105 194 188 237 79 37 W-110 1.779 137 226 190 239 81 35

• This value is presently being evaluated.

TABLE 1.1.E-5 Transition Temperature and USE Results for Palisades RV Surveillance Program HAZ Material from CVGRAPH TAR 12/11 /95 1 .1-20

Capsule Fluence, x 30 ft-lb l1T30 50 ft-lb - l1Tso Upper Shelf Decrease in 10 19 n/cm 2 Transition (oF) Transition (oF) Energy USE (E > 1 MeV) Temperature Temperature (ft-lb) (ft-lb)

(oF) (oF)

Unirradiated 0.0 14 0 43 0 136 0 W-110 1.779 168 154 200 157 99 37 TABLE 1.1.E-6 Transition Temperature and USE Results for Palisades RV Surveillance Program SRM Material from CVGRAPH TAR 12/11/95 1 . 1-21

lsrP f-- - - - --: i Outlet Nozzle

/

/ ' '~_/

( 1 Thermal t Accelerated T-150 A-240

____ , . . \ Inlet Thermal

\Nozzle Capsules

\ Assembly

\

.... J I

I

,,....I Accelerated

-.-.--Wall Caosule Wall W-260 Assembly W-100 Wall Wall---r-_.. Capsule-~...

W-80 Assembly I

r""

j

\

' \

~""-

I I I

'\ /

' 'j (

/

l. _______,

Enlarged Plan View rP Elevation View Figure 1.1.E-1 Location of Palisades Surveillance Capsule Assemblies TAR 12/11 /95 1 .1-22

Rolling or Working Direction, L Longitudinal [ L-T or L-C] ~

Orientation (Strong Direction)

L----,,,' -'I L..--+--+--tlt"1-, r - - - '.._)" - - I

"',1 __ ... ----I f-1 I

-*1 . I I

---~~__:1--~~~-il~ I I

' I I Tiansverse

( I I Direction, I I I I 1) Tore

• Transverse [ T-L or C-L]

Orientation (Weak Direction)

Short Transverse or Radial Direction Figure 1.1.E-2 Charpy Specimen Machining Orientation

• TAR 12/11 /95 1 .1-23

• Palisades Nuclear Plant - Base (Long.)

CVGRAPH 4.0 Hyperbolic Tangent Curve Printed at 17:57:23 on 10-02-1995 Results Curve F1uence I.SE d-I.SE USE d-USE To 30 d-T o 30 To 50 d-T o 50 1 0 2.19 a 154.B a -.36 a 25.29 0 2 6E+19 2.09 -.09 92.5 --£2.3 204.46 204.82 243.08 217.79 3 0 2.19 0 171.B 17 4.71 5.07 33.75 B.45 4 1.09E+19 2.2 0 112.39 -42.4 176.11 176.47 197.76 172.46 5 1.779E+19 2.1 -.09 102.69 -52.1 17914 179.5 203.59 178.29 300 rn 250

,.0

~

I I

+.)

• ~

~ ~

~ -** . _.......- .-**----- D bJ) ' /--:: *-1]

A

~ 150 [

OJ  ;@

~ I

~

• f' v:;;

~ ~,0! - A A

/ *17 100 li'__...Q.. ___ --*-----

z u

DU

.~

1 ~el D ,...,..... .....----i

/'

~ ~ A

. *.~ ~

u

-300 -200 -100 0 100 200 300 400 500 600 Temperature lll Degrees F Curve U!gend 10 20---------* 3\7---*- 4 5 Data Se4s) Plotted Curve Plant CajmI.le Material Ori Heati 1 PAL Unirr PLATE SA302BM LT C-1279-3 2 PAL A-240 PLATE SA302BM LT C-1279-3 3 PAL T-330 PLATE SA302BM LT C-1279-3 4 PAL W-290 PLATE SA302BM LT C-1279-3 5 PAL W-110 PLATE SA302BM LT C-1279-3 Figure 1.1.E-3 Surveillance Capsule Results for Palisades Base Metal (longitudinal) from CVGRAPH TAR 12/11 /95 1. 1-24

• Palisades Nuclear Plant - Base (Transverse)

CVGRAPH 4.0 Hyperbolic Tangent Curve Printed at 1ail4:0l on llHJ2-1995 Results Curve Fluence 1SE d-lSE USE d-USE To 30 d-T ii 30 T ii 50 d-T ii 50 1 0 2.2 0 101.6 0 la46 0 49.28 0 2 6E+19 2.2 0 68.39 -332 212.15 193.69 264.92 215.63 3 0 2.2 0 107.8 619 la54 1l8 5618 6.89 4 to9Etl9 2.2 0 83.8 -17.8 176.03 157b7 202.36 153.07 300 50 D ~

• ~ .a.---; 1--*--*--*-j i;._,__________ _,.,,_ _____ ,_,_,. .. '

00

~ ~ [ .<fl

) ~*

~~ ~.~-9---' ~--*----- -------

I// & ...

OU

~

~--411 v """

_o..

~JI i F>-_4,,o*

D u

-300 -200 -100 0 100 200 300 400 500 600 Temperature lll Degrees F Curve Legend 1 [J 20-------- 3 ~-----*-**- 4 Data Se~s) Plotted Curve Plant Capsule Material OrL Heat~

1 PAL Unirr PLATE SA302BM TL C-1279-3 2 PAL A-240 PLATE SA302BM TL C-1279-3 3 PAL T-330 . PLATE SA302BM TL C-1279-3 4 PAL W-290 PLATE SA302BM TL C-1279-3 Figure 1. 1.E-4 Surveillance Capsule Results for Palisades Base Metal (Transverse) from CVGRAPH TAR 12/11 /95 1.1-25

• Palisades Nuclear Plant - Weld (Heat 3277}

CVGRAPH 4.0 Hyperbolic Tangent Curve Printed at IB:11ill on 10-02-1995 Results Curve Fluence ISE d-ISE USE d-USE To 30 d-T o 30 To 50 d-T o 50 1 0 2.1 0 117.69 0 -86.46 0 -47.05 0 2 6Et19 22 .09 50.79 -86.89 255.45 341.91 389.78 436.84 3 0 22 j 132.39 14.69 -63.05 23.4 -IB.93 28.11 4 1.09E+19 2.19 .09 63 -54.69 198.85 28.5.31 254.25 301.31 5 1.7/9E+19 22 .09 57.79 -59.89 21B 304.46 30619 35325 300 a>lJ 200 *-

~

150

-u-----~ -*---**--*- -*-*--... ......_ ......,....,.,,.*,*._,,.*,,, __

O,.~ ~ n

..,--*u [

100

-* ~ n,*:

./,/

A

/,

A A,

/

~~'7

,.. ... .... 0

~-/

~

00 2:uu *-

~

~ -** .

Ql)

~ 150 Cl,)

r7- u u Q

~ ()

q..- .!.!.-*---*--

DI 100 v z - J!f

./

> 0 vu

/

u /

,/

50 /

/

/ <D

@.7 0

./ -*-*-*--*.J

~ .....

-300 -200 -100 0 100 200 300 400 500 600 Temperature lil Degrees F Curve Legend l Cl 20----------

Data Se~s) Plotted Curve Plant Ca~sule Material Ori. Heatfi 1 PAL Unirr SRM ~1 LT 2 PAL W-110 SRM H5m'Ol LT Figure 1.1.E-7 Surveillance Capsule Results for Palisades Standard Reference Material from CVGRAPH TAR 12/11 /95 1. 1-28

Palisades Nuclear Plant - Weld/34B009 (Unir)

C\'GRAPH 4.0 Hyperbolic Tangent Cun'e Printed at 15:12:53 on 12--01-1995 Page 1 Coefficients of Cum I A= 58 B = 55.9 c =91.09 TO =-31.89 F.quation is: CVN = A + B * [ tanh((T - TO}/C) ]

Upper Shelf Energy: 113.9 Fixed Temp. at 30 ft-lbs: -82 Temp. at 50 ft-lbs: -45 Lower Shelf Energy: 2.09 Fixed Material: 1fELD Heat Number. 340009 Orientation:

Capsule: Unirr Total Fluence 0.0 30 0 0

0 0

0 D p.......- ~

0 v

g;I~

50 ya

__/a 0 I

. -300 -200 -100 0 100 200 300 400 500 600 Temperature lil Degrees F Data Se<<s) Plotted Plant: PAL Cap: Unirr Material: WELD Ori: Heat #: 34!3009

• Figure 1.1.E-8 TAR 1 2/11 /95 Unirradiated Charpy Test Results for Supplemental Surveillance Material Weld Heat 348009

1. 1-29

Palisades Nuclear Plant l\Teld/W5214 (Unirr)

C\'GRAPH 4.0 Hyperbolic Tangenl Curve Printed at 15:08:08 on 12--01-1995 Page 1 Coefficien Is of Cum 1 A =52.4 B = 50.29 c = 99.19 TO = -12.59 F.qualion Th: CVN = A + B ' l tanh((T - TO)/C) I Upper Shelf Energy: 102.7 Fixed Temp. at 30 ft-lbs: -502 Temp. at 50 ft-lbs: -17.4- Lower Shelf Energy: 2.1 Fixed Material: 1\"ELD Heat Number: ~-5214- Orientation: TL Capsule: Unirr Total Fluence: 0.0 3uu rn 250

,..0 7

~ 200

• ~

z H

~

Q.D Q) q 150 100 0 0 0

10  ?

u ~

5u

. u

~

v I

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Se~s) Plotted Plant PAL Cap~ Unirr Material: lfELD Ori~ TL Heat ffe. lf521~.

• Figure 1.1.E-9 TAR 12/11 /95 Unirradiated Charpy Test Results for Supplemental Surveillance Material Weld Heat W5214 1 .1-30

• Palisades Nuclear Plant - Weld/27204 (Unirr)

C\'GRAPH 4.0 Hyperbolic Tangent Curi*e Printed at 15:l5:2B on 12--01-199'5 Page l Coefficien.ts of Curve 1 A = 53.29 B =5L09 c =81.9 TO = -.89 F.quation is: CVN =A+ B * [ tanh((T - TO)/C) I Upper Shelf Energy: 104.39 Fixed Temp. at 30 ft-lbs: -412 Temp. at 50 ft-lbs: -6.1 Lower Shelf Energy: 22 Fixed Materia~ ll"ELD Heat Number. 27204 (Circum. Weld) Orientation:

Capsule: Unirr Total Fluence:

300 250 200

• 150 8 -

r*

[

100

[)

50 u

-300 ,. -200

.J

-100 v 0 100 I

200 300 400 500 600 Temperature in Degrees F Data Se~s) Plotted Plant: PAL Cap~ Unirr Material: ll"ELD Ori~ Heat ffe. 2ia\4 (Circum. Weld)

• Figure 1.1.E-10 TAR 12/11 /95 Unirradiated Charpy Test Results for Supplemental Surveillance Material Weld Heat 27204

1. 1-31

• ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 THERMAL ANNEALING REPORT SECTION 1

• THERMAL ANNEALING OPERATING PLAN SECTION 1.2, DESCRIPTION OF THE REACTOR VESSEL

1.2 DESCRIPTION

OF REACTOR VESSEL

• 1.2.A Detailed Description of Reactor Vessel The reactor vessel (RV) assembly is a vertically mounted cylindrical vessel with an integral hemispherical lower head and a removable hemispherical upper head. The reactor vessel is approximately 483 inches high with a 172 inch inside diameter, with all welded manganese molybdenum steel plate and forging construction, see Figure 1.2.A-1. The internal surfaces, which are in contact with the reactor coolant are clad with 1 /4 inch nominal (3/16 inch minimum) stainless steel having an as-deposited and buffed surface. The cladding consists of a 1 /8 inch layer of 309 stainless for bonding with the ferritic base metal and a 1 /8 inch layer of 308 stainless steel in contact with the reactor coolant. Unless additional machining or welding was required in the clad area, the clad surfaces were ground or buffed to provide a suitable surface for pre-service inspection. The closure head flange and the RV shell flange provide the structural rigidity necessary for bolting the head to the shell.

The hemispherical bottom and closure (upper) heads were made from formed low alloy steel plates and forgings. Each head was made from a small rounded dome welded to a torus, which was made of four smaller welded pieces. The bottom head was attached to the lower shell course with a low alloy steel weld while the closure head was welded to a flange. The lower head contains no penetrations.

• Two concentric gasket grooves are machined in the closure head flange for 0-ring gaskets. These gaskets are polished and silver plated Ni-Cr-Fe alloy. The 0-rings are installed on the head flange, which is secured by hydraulically tensioned studs to the vessel flange. Two connections for a monitoring device are provided in the vessel flange to detect any leakage past the 0-rings. The closure head is attached to the vessel shell by 54, 7 inch diameter studs, threaded into the vessel .flange and extending through the closure head flange. The closure head is held in place by means of nuts which are screwed onto the studs and seated on spherical washers.

The six RV nozzles, two outlet and four inlet, are located with their horizontal center lines on a common horizontal plane. The nozzles are tapered internally to reduce coolant pressure losses. An internal boss around the outlet nozzles provides a mating surface for the core support barrel (CSB) outlet nozzles. Each nozzle. has a short section of forged carbon steel safe end (stainless steel clad) welded to it, allowing a carbon-to-carbon steel field weld to be made between the reactor coolant piping and the nozzles during plant construction.

The reactor vessel is supported by three pads welded to the underside of the coolant nozzles. Two pads are located under inlet nozzles and one pad under an outlet nozzle. These pads are bolted to sole plates which can slide on a set of plates mounted to steel I-beam supports which are .anchored in the concrete biological shield that surrounds the reactor vessel. This support arrangement

• permits radial thermal growth of the reactor vessel while maintaining it centered and restrained from any movement resulting from seismic forces.

12/11/95 1.2-1

A ledge is machined on the inside surface of the RV flange that is used to suspend

• the CSB, carrying the entire weight of the reactor internals .

Six core stabilizing lugs (snubbers) built into the lower portion of the RV shell limit the amplitude of flow induced vibrations in the CSB. The core stabilizing lugs are Ni-Cr-Fe alloy. At the vessel wall, the stainless steel clad is interrupted at the six core stabilizing lug locations. Ni-Cr-Fe weld wire (EN-82) was used to form a local pad to which each Ni-Cr-Fe alloy core stabilizing lug was welded. Around the periphery of the pad, a Ni-Cr-Fe stick electrode (alloy 182) was used to tie the pad to the stainless clad.

Nine core support lugs are built into the bottom head of the reactor vessel to limit the downward drop of the core if the CSB should fall. In the area of the core

.*support lugs, the bottom head is clad with a 360 degree band using Ni-Cr-Fe wire (EN-62). The Ni-Cr-Fe alloy core support lugs are welded to this band using Ni-Cr-Fe stick electrodes (alloy 182).

The Ni-Cr-Fe flow skirt assembly is located in the bottom head of the reactor vessel and is welded to the core support lugs. The flow skirt is a perforated cylinder designed to provide more uniform flow distribution to the core.

Six vessel wall mounted surveillance holder assemblies are positioned close to the RV wall. The holder assemblies are rectangular tubes made of Ni-Cr-Fe alloy. A

• series of eighteen Ni-Cr-Fe alloy supports (nine per side) is welded to each tube and to the stainless cladding of the RV using Ni-Cr-Fe electrodes (alloy 182).

1.2.B Palisades Reactor Vessel Annealing Zone The RV beltline region is the steel volume adjacent to the active fuel length. For the Palisades reactor vessel, the primary thermal annealing zone will be the base plates, axial welds and circumferential weld in the beltline region as previously described in Section 1.1 .C of this report. These materials are identified in Table 1 .2.B-1. The beltline temperature gradients will be controlled to assure recovery of the relevant embrittled materials. The entire volume of material between the RV ID and OD must be annealed. The projected properties of these materials after annealing and reirradiation are summarized in Tables 1. 1 .C-3 and 4. These calculations indicate that annealing at the nominal conditions of 850 ° F for 1 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br /> should produce a level of recovery sufficient to keep all of the beltline materials below the relevant screening criteria beyond the end of life. The thermal annealing apparatus will be constructed to span an axial height (at .a minimum) from the top of the core to the bottom of the core. Within this region, the specified annealing temperature will be maintained throughout the annealing time period. A detailed description of the actual operating limits on annealing temperature and time is included in Section 1 .4.

• Above and below the height of the core, there is base metal and weld metal that has also been embrittled to a lesser extent than the beltline materials. The plates in 12/11 /95 1.2-2

the core beltline extend more than 14.5 inches above the active core and

• approximately 44 inches below the active core. Within these near beltline transition regions, the plate and axial weld materials are exactly the same as the beltline materials. The extent of embrittlement is determined by the axial fluence profile. At the top and bottom edges of the beltline region, the fast neutron flux is less than 50% of the peak beltline axial value. This implies that the peak end of life fluence in the near beltline plate materials will be less than 1.36x10 19 n/cm 2 (E > 1 MeV) and the peak fluences in the near beltline axial welds will be less than 1.04x10 19 n/cm 2 (E > 1 MeV). These peak end of life fluences in the near beltline region are less than the peak fluences for the equivalent materials in the beltline prior to the anneal at the end of the 13th cycle, see Table 1.1.C-3. Due to this axial reduction in flux, these materials are not projected to exceed any RV integrity screening criteria during the period of licensed operation. As a result of the physical arrangement of the annealing apparatus, these materials will also experience a certain amount of 'recovery. However, there is no. need to take credit for this recovery. In the event that additional analysis of recovery in this region should be deemed desirable at a future date, the time and temperature records for the near core transition region will be maintained.

1.2.C Reactor Vessel Data & Programs for Recovery and Reembrittlement The Palisades reactor vessel was fabricated by Combustion Engineering (CE) in Chattanooga, Tennessee under Contract No. 2966A. Data on the material

• properties and vessel condition have been generated through a combination of original qualification programs, in-service inspections and on-going surveillance programs. These data provide the basis for subsequent analysis used to evaluate the requirements for reactor vessel annealing.

1.2.C.1 Reactor Vessel Materials The RV shells and bottom head were fabricated from rolled plates of SA-3028 modified material. The primary coolant nozzles and the vessel flange were fabricated from forgings of A508-Class 2 material. The leakage monitors were fabricated using SB-166 and SB-167 materials. Welding of the low alloy steel components was performed using automatic submerged arc and manual metal arc processes.

The beltline region of the Palisades reactor vessel incorporates portions of the intermediate and lower shell courses along with the associated axial welds and the intermediate to lower circumferential weld. The beltline materials are identified in Table 1.2.B-1 and their locations shown in Figure 1.2.A-1.

The method presented in Section Ill of the ASME Boiler and Pressure Vessel Coc:le, Appendix G, "Protection Against Nonductile Failure", uses fracture mechanics concepts and is based on a reference nil-ductility temperature (RTNOT). RT NOT is

• defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 60°F less than the 50 ft-lb (and 35-mil 12/11/95 1.2-3

-~-----

lateral expansion) temperature as determined from transverse Charpy specimens.

• Although Code qualification data are available for each material in the RV beltline, the RT NDT requirements were established after the reactor vessel was fabricated.

Therefore, welds were tested to meet Charpy requirements at a single temperature and plate Charpy specimens' orientations were longitudinal rather than transverse as per NB 2300. Therefore, RT NDT values for plates were determined using MTEB 5-2 procedures.

Neutron irradiation can produce an increase in hardness and tensile properties and a decrease in ductility of low alloy ferritic materials, such as the Palisades RV base metal material. In addition to the Code qualification tests on the beltline materials, surveillance program base metal and weld metal materials were also tested to ascertain important mechanical properties. The Palisades surveillance base plate, weld metal and HAZ materials were tested in the form of Charpy and tensile specimens. The results of the Charpy tests performed on the surveillance materials are presented in Section 1.1.E. A comparison of the 30 ft-lb transition temperature increases and upper shelf energy (USE) decreases for the Palisades surveillance materials shows good agreement with the Regulatory Guide 1 .99, Revision 2 predictions (Peter et al., 1994). The information in Table 1. 1.C-4 includes initial USE values for the beltline materials. Initial RT NDT values used to determine adjusted RT NDT values are described below.

1.2.C.1.1 Reactor Vessel Beltline Plates

• The Palisades RV plates were purchased to ASME specification SA-302 Grade B-modified. The "modified" refers to the addition of approximately 1 /2% nickel, which makes the plate comparable to SA-5338 Class 1 . There are six plates in the lower and intermediate shell courses, which encompass the entire beltline region of the reactor vessel. These plates are identified in Table 1.2.B-1. The surveillance program specimens were taken from Plate D-3803-1 (Heat C1 279-3), used in the fabrication of the intermediate shell course. Plate fabrication records and test certificates for the remaining beltline base plates were reviewed to compile the necessary chemistry information. The best estimate chemistry, provided in Table 1.2.B-1, illustrates that all of the plates are comparable in terms of composition.

Plate D-3803-1 (Heat C1279-3) is an appropriate material for assessing the embrittlement of all RV plate materials.

The initial RT NDT value for Intermediate Shell Plate No. D-3803-1 surveillance base plate material was determined to be -10° F based on unirradiated transverse Charpy tests and longitudinally oriented drop weight test specimens. The remaining base plates in the RV beltline region do not have measured transverse orientation Charpy data and as a result, initial RT NDT values are based upon the MTEB 5-2 approach.

Initial RT NDT values for the beltline base plate materials are summarized in Table 1.2.B-1.

• There is no record of weld repairs to any of the beltline plates, see Section 1.2.C.2.3. Shallow repairs, if performed, would consist of a manual arc weld made 12/11 /95 1.2-4

using E8018 stick electrodes which is 3/8 inch or less deep on the inside or outside

• surface of the plate. For a discussion of the chemistry factor of a manual arc electrode weld deposit, see Section 1.2.C.1 .2.

1.2.C.1.2 Reactor Vessel Beltline Welds The Palisades RV beltline welds, consisting of seam numbers 9-11 2, 2-11 2 A, B &

C and' 3-112 A, B & C were fabricated using automatic submerged arc (S/A) welding. The intermediate and lower shell course plates were welded axially using three seams in each shell course. The axial seam weld numbers are 2-112 A, B &

C and 3-112 A, B & C. They were deposited using a tandem arc process (i.e., two coils of wire fed through the welding heads) with Linde 1092 flux. A Ni 200 wire was also fed into the weld puddle to attain approximately 1 % nickel in the weld deposit. The intermediate and lower shells were joined by circumferential weld seam number 9-112. The circumferential weld was deposited using a single arc process with a Linde 124 flux. There was no Ni 200 addition because the wire was alloyed with approximately 1 % nickel.

The chemical composition and initial toughness properties of the beltiine axial and circumferential welds are given in Table 1.2.B-1. Welds similar to the axial and circumferential welds in the Palisades beltline region have been analyzed numerous times over the past decade. As a result, there is a relatively large body of available data for the class of welds in the Palisades reactor vessel. In addition to compiling

• data from other reactor vessels fabricated with the same heats of weld wire, retired steam generators containing the same weld heat were sampled in detail for chemical composition (Fenech, 1994). All available chemistry information was reviewed and compiled to develop the best estimate chemistry values for the welds, see Table 1.2.B-1. These properties provide the basis for predicting the embrittlement condition of the RV beltline welds prior to annealing.

Three heats of weld wire (27204, W5214 and 348009) were used in the fabrication of the beltline welds. The surveillance weld was fabricated using a fourth heat of weld wire (3277). To provide more representative surveillance data, three supplemental welds have been included in the Fracture Toughness Recovery and Reembrittlement Assurance Program, see Section 3. These supplemental materials were carefully chosen to match the materials and procedures used in the fabrication of the RV beltline welds. A comparison of the RV beltline weld materials and the surveillance materials is provided in Table 1.2.C-1.

All weld materials in the beltline were assumed to have a generic initial RT NOT value of -56 °F. Initial RT NOT values for the beltline welds are summarized in Table 1. 2. B-1.

Fit-up, root welding, back welding and some of the repair welds involved manual arc (M/A) welding using E8018 electrodes. Fit-up included items such as the

• attachment of backing bars prior to circumferential seam welding. The attachments were typically made to the base metal. Root welding consisted of 1 2/11 /95 1.2-5

"tacking" the machine prep of two adjoining plates prior to axial seam welding.

• The majority of the weld prep, including the M/A weld, was removed by back-gouging the weld root. Back welding included procedures such as surface welding to smooth the weld and plate surface after the backing bar was removed. This was done only if removal of the backing bar and subsequent grinding left a recess in the weld surface. Repair welding involved the filling of a cavity in the weld or plate which had been back-gouged (i.e., to remove a defect discovered during the welding process). Shallow cavities were repaired by M/A welding, whereas large repairs were often accomplished by S/A welding.

The intermediate to lower shell circumferential (9-112) and the lower shell axial (3-112 A, B & C} weld records indicate that E8018 electrodes were used in addition to the S/A weld wire and flux, as indicated in Table 1.2.B-1. The weld wires for the S/A welds were copper coated. The E8018 electrodes consisted of a low alloy steel wire which was coated with flux. The wires (E8018 stick electrodes} for the M/A welds were not copper coated and therefore, the copper content of the deposited M/A welds was typically under 0.10% Cu. The wire was alloyed with Ni to a nominal concentration of 1 .0%. Therefore, the nominal chemistry factor determined using the procedures outlined in 10 CFR 50.61 is 135 °F. This is significantly less than the chemistry factors for the submerged arc weldments listed in Table 1.1.C-2. Given the relatively low chemistry factor of an E8018 electrode, any beltline area containing the M/A weld deposit will be less embrittled (i.e.,

tougher} after neutron irradiation than the adjacent S/A weld deposit. Furthermore,

• the volume of M/A weld deposit will be very small compared to the volume of S/A weld deposit. Therefore, the E8018 electrode material used for the Palisades beltline will not be considered in RV integrity analyses.

1.2.C.1.3 Reactor Vessel Cladding For the Palisades RV, the primary cladding surfaces (i.e., heads and shell surfaces}

were clad using shielded metal arc, single and three wire submerged arc, single wire gas-tungsten arc welding (GTAW} and gas-metal arc welding (GMAW}

processes. The electrodes used for cladding (as well as attachment welds and surface buildup} were 308, 309 and Ni-Cr-Fe (i.e., alloys 62, 82 and 182).

In the early 1980's, pressurized thermal shock (PTS} evaluations typically used a stress free temperature of 550 ° F. This value was conservative in terms of tensile force that would be applied to the vessel base metal during a cooldown. A more recent analysis to better understand the effect was performed (Ganta, Ayres and Hijeck, 1991 & Kanninen and Chell, 1993) to determine the room temperature residual stress distribution in the cladding and base metal and an estimate of the stress-free temperature for use in RV integrity evaluations. The analysis considered best estimate temperature dependent thermal, elastic, plastic and creep material properties, including an assumption that the room temperature yield stress of the cladding is 45 ksi. The analysis followed the clad application, the subsequent heat

• treatment as defined by the manufacturing records, the hydro test and the first cycle of normal operation. After all these loadings, the RV temperature was raised 12/11 /95 1.2-6

uniformly until the cladding was approximately stress free. This occurred at about

• 400° F.

In the mid 1980's Westinghouse conducted a series of residual stress measurements of clad specimens from nozzle cutouts of operating nuclear power plant vessels. The measurements were taken using the classic hole-drilling method utilizing a range of temperatures from room temperature up to 300 ° F. The results at 300 ° F showed very low stresses from the cladding and when a curve-fitting routine was applied, the conclusion was that the clad stress free temperature is approximately 400° F, which agrees with the value above.

A stress free temperature value of 400 ° F, as well as the yield stress assumption (45 ksi at room temperature) were applied to the thermal stress analysis of Section 1 . 7 in order to evaluate their effect on the anneal. The effect on the anneal was small since the anneal occurs in a unpressurized vessel. and the cooldown rate on the base metal is low. Since the largest stresses* in a thermal annealing process occur at the end of heatup when the vessel it still undergoing thermal expansion, a lower reference temperature would result in larger deformations and stresses. For the 'bounding' evaluation, the thermal annealing stress analysis conservatively used a reference temperature of 70 ° F.

1.2.C.2 Reactor Vessel Assembly Sequence

• The following is a general description of the sequence of fabrication operations followed by CE.

1.2.C.2.1 Shell Courses The shell plates were hot formed, quenched and tempered. Each plate was machined, fit-up, joined by axial weld seams and stress relieved. Each shell was then clad with stainless steel on internal surfaces. The RV flange was clad, stress relieved, and joined to the uppermost shell course by a circumferential weld seam, and stress relieved.

An upper assembly, including the intermediate and upper shell courses and the RV flange, was fabricated. The shell courses were fit-up and joined by a circumferential weld seam and then stress relieved (interstage). The circumferential welds were back clad. The nozzle openings were cut by burning and then machined. The nozzles were fit-up to the shell, welded and the welds back clad.

A lower assembly, including the lower shell course and the bottom head, was fabricated. The bottom head torus and dome plates were hot formed and then quenched and tempered. The torus plate sections were machined (weld prep),

joined by axial seams and stress relieved. The dome was machined and then joined to the torus by a circumferential weld seam and stress relieved. The bottom head was clad, joined to the lower shell and then stress relieved. The circumferential welds were back clad.

12/11 /95 1.2-7

1.2.C.2.2 Final Assembly

• The upper assembly and lower assembly were joined by the closing circumferential seam weld. The circumferential weld was back clad and the completed RV assembly was given a final stress relief (furnace).

1.2.C.2.3 Pre-Service, Non-Destructive Examination (NOE)

Plate used in the manufacture of the Palisades RV was received at CE and processed in the following manner. Upon receipt, the plate dimensions and supplier identification were checked and recorded using the Certified Material Test Report. The plate was then steel stamped with the unique code number identification assigned by CE. The plate then underwent an ultrasonic examination for compliance with ASME Code and contract requirements. Subsequent to ultrasonic examination, the plate was placed in the furnace and heated for "hot forming" to the required radius of the component. After completion of the forming operations, the plate was again placed in the furnace and heated to the required temperature for the quench operation. Subsequent to review of the quench documentation, the plate was again placed in the furnace for the tempering operation. Subsequent to the tempering operation, a test plate was removed from the parent plate and sent to the metallurgical laboratory for analysis. After review of the mechanical test data, the plate was released for fabrication.

• In-process NDE was performed on each RV weld using magnetic particle and dye penetrant testing. Each completed weld was radiographed and repairs to welds were made if necessary. All weld indications greater than 3/8 inch deep were radiographed after the repair was completed. The recordable radiographic indications are listed in Table 1.2.C-2.

If the ultrasonic examination (UT) of the Palisades RV plates had identified any indications greater than 3/8 inch, the plate would have been repaired and radiographed. No radiograph and repair records were found with the fabrication records for the RV shell plates indicating thqt no such repairs were made. It is possible that plate surfaces were welded (i.e., to repair shallow UT indications or to smooth over where the straps were removed), but records were not generated in such cases. No base metal repairs requiring radiography were recorded with the exception of two base metal repairs discovered during vessel inlet nozzle weld inspections. The base metal repairs were to the upper shell course plates adjacent to welds 5-114C and 5-1140. These repairs are listed in Table 1 .2.C-3.

1.2.C.2.4 Post-Weld Heat Treatment Each weld received a post-weld heat treatment (PWHT) to reduce residual stresses caused by the welding process. The closing circumferential seam weld and subassemblies (i.e., closure head, upper assembly and lower assembly) received the final PWHT in a furnace in addition to the interstage PWHTs given to each subassembly.

12/11 /95 1.2-8

The cumulative total PWHT (at temperatures between 1100° F and 1175° F) time

• for the beltline materials was 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, 5 minutes for the lower shell and 1 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 1 5 minutes for the intermediate shell.

1.2.C.3 Reactor Vessel Neutron Fluences Fast neutron fluence evaluations through the completion of Cycle 11 were performed using two-dimensional discrete ordinates techniques in r,(} geometry.

The methodology and results are documented in Section 6.0, "Radiation Analysis and Neutron Dosimetry" of WCAP-14014 (Peter et al., 1994).

All of the transport calculations used the SAILOR cross-section library. The SAILOR library is a 47 group ENDF/B-IV data set produced specifically for light water reactor applications. In the Palisades analysis, anisotropic scattering was treated with a P3 expansion of the scattering cross-sections and the angular discretization was modeled with an S8 order of angular quadrature.

The core power distributions utilized in the transport calculations were generated with the normalized nodal cycle energy database. The aforementioned core power distributions were supplied in terms of average pin power within the outer assemblies, axial power shapes and the beginning of cycle and end of cycle assembly burnups. From the fuel burnup for each assembly, an average fission spectrum for that assembly was calculated using ENDF/B-V fission spectra for each

• fissionable isotope. Using the fission spectrum for each assembly, the assembly power generation obtained from the cycle burnup and the power shape within the assembly from relative pin powers, a 47 group neutron source was calculated for each mesh point in the transport model.

Results of the fluence evaluations for the Palisades RV are provided in Table 1.2.C-

4. The data provided in this table represent the axial maximum fluence at several azimuthal locations along the RV clad/base metal interface. Data are provided for each operating cycle to date. Along with the vessel fluence values, the reactor operating history in terms of effective full power days per fuel cycle is also provided.

Projected fluence values for individual heats of material can be obtained from the data provided in Table 1.2.C-4 by correlating the location of the individual heats with the maximum exposures given in the table. In that regard, it is noted that each of the shell plates as well as the intermediate to lower shell circumferential weld experience the maximum fluence at the 16° azimuthal location. At the end of Cycle 11, the peak fluence at this location was 1. 76x 10 19 n/cm 2 (E > 1 MeV).

Based on these measurements, the projected peak fluence at the end of Cycle 1 3 is 1.93x1019 n/cm 2

• The lower and intermediate shell axial welds are limited to specific azimuthal angles of 0°. or 30°and the peak fluences at these locations will be lower than the peak fluences in the plates and circumferential weld. For the axial welds, the projected peak fluence at the end of Cycle 13 is 1.45x1019 n/cm 2

• 12/11 /95 1.2-9

1.2.C.4 Palisades Reactor Vessel Beltline lnservice Inspection (151) Results

• lnservice examinations involving the RV beltline region were performed in 1 983 during the 10-year examination and in 1995 at the 20-year examination. Both inspections included 100% of the length of the intermediate shell course longitudinal welds, the circumferential weld seam connecting the intermediate and lower shell courses and the lower shell course longitudinal welds. The volume of metal subject to inspection in both examination efforts included the welds and base metal on each side of the welds for a distance of one-half the wall thickness. From an ultrasonic testing standpoint, both examinations included four-directional shear wave angle beam interrogation of the code-specified volumes and straight beam inspection of the volumes. Additionally, supplemental "near-surface" examinations using low-angle longitudinal waves were conducted for detection of defects directly underneath the cladding. Both the 10 and 20-year examination efforts were conducted in compliance with the requirements of the ASME Code,Section XI, and the recommendations of USNRC Regulatory Guide 1.150.

The two examinations were conducted almost 12 years apart and as expected, technoiogy upgrades were implemented in the 1995 examination. These included complete processing of the distance amplitude corrected ultrasonic signal for each .

tran'sducer and archive storage of all examination data on optical disk for "off line" analysis.

• Both examinations yielded a number of recorded indications. All indications in the Palisades RV beltline region can be characterized as one of two basic types, plate segregates and embedded fabrication flaws. The plate segregates are seen with the 0° transducer or "straight beam" in the portion of the examination volume extending about five inches on either side of the intermediate and lower shell course longitudinal seams and the RV circumferential weld (which joins the two sections). These indications, confined to the middle one-third of the plate thickness, are small laminar inclusions formed during material processing and are seen, to some degree, in all reactor vessels fabricated with SA-302 Grade B or SA-533 Grade B plate material. A number of measurements of these straight beam indications were conducted in the 1983 examinations. All were found to be well within the acceptance criteria for laminations, which for nine inch thick plate, is 28 square inches per indication. A second criteria for evaluation of the straight beam indications is whether or not they interfere with transmission of the angle beams ..

According to this criteria, none of the straight beam indications was large enough to interfere with the angle beam examinations.

The fairest and most accurate way to describe the distribution and size of the plate segregates is simply to assume that they exist intermittently in all six shell segments comprising the beltline region. In the 1983 examination, 14.segregate indications had peak amplitude responses high enough to require measurement with the 0° transducer. Measured as a laminar indication with the 0° transducer, the average laminar area was 0.44 square inches with the largest measurement being 0.98 square inches which is 3.5% of the allowable laminar area. Straight 12/11 /95 1. 2-10

beam amplitude based measurements of this type can be exaggerated and the

• likely size for the segregates is about 0.2 inches by 0.2 inches .

Angle beam indications in the Palisades RV beltline were recorded in 1 983 and 1995. A summary of the indications is presented in Table 1.2.C-5. From the table, a total of 15 relevant indications were identified in 1983. The majority of the indications were detected and measured with the 60° longitudinal wave transducer, a sensitive detection technique used for inspection of the material volumes directly beneath the cladding. A total of 12 of the 15 indications were seen again in the 1995 examination at or below the recording threshold.

Of particular interest is an indication area located near the top of intermediate shell long seam 2-1128, reference Figure 1 . 2.A-1, at a distance of about 140 inches from the top of the vessel flange and a vessel azimuth of between 25 and 30 degrees. Found in 1983 during the 10th year examination with 60° Longitudinal (L-wave) techniques, the indications were interpreted to be two small clusters of reheat cracks situated on either side of the long seam. The indications were individually assigned a cons~rvative flaw size and were determined to be within the allowable limits of Section XI, l\tVB-3500. This interpretation was supported by additional scanning of reheat cracked test specimens at the Waltz Mill facility after the 1983 examination. Details of the inspections and investigations can be found in the 1983 RV examination report (Westinghouse, 1983).

• In the 1995 examination, the area of interest was scanned with shallow angle 70° L-wave techniques. A relatively small grouping of indications was noted at a distance of 140" at 27 degrees vessel azimuth. Other indications were seen at a distance of 140 inches, 33° vessel azimuth and 135", 39° vessel azimuth. The indication depths (0.3 inches), echodynamic signature and rate of occurrence particularly at the 140" elevation are consistent with the previous interpretation of reheat cracking. There is no evidence this relatively small band of indications has expanded in number or size.

1.2.D Reactor Vessel Dimensions The nominal Palisades RV dimensions utilized for the analysis of the thermal annealing are listed in Table 1.2.D-1, with additional information shown in Figure 1.2.D-1.

In 1973, motion of the RV internals was hypothesized due to gradually increasing ex-core signal noise. In October 1973, inspections of the reactor vessel and RV internals confirmed the suspected motion of the internals. As of that time, about 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation had occurred. The motion of the RV internals was attributed to loss of "clamping" at the RV ledge area. The effect of this motion was to wear away approximately 1/4" of the RV ledge upon which the CSB flange rests. Thus, at the ledge, most or all of the cladding is worn away and the distance from the vessel mating surface to the CSB seating surface is closer to 8 1/2 inches, than the 8 1/4 listed in Table 1.2.D-1. The wearing condition was 12/11 /95 1 .2-11

stabilized by the introduction of a new internals hold-down device in 1974. Several

• inspections since 1974 confirmed the effectiveness of the repair. The wear areas are of no consequence to the RV anneal because they are located remotely from the beltline area, and because the minor amount of material loss (mostly cladding) will not materially change the structural or thermal response of the reactor vessel during the annealing.

1.2.E Attachments to the Vessel and Expected Effects of Annealing There are a number of attachments to the Palisades reactor vessel, both ext~rnal and internal, which may be affected by the thermal annealing operations.

• 1.2.E.1 External Attachments The external attachments to the Palisades reactor vessel which will be subjected to the effects of the thermal annealing include:

• Nozzle extensions

• 0-ring leakage monitor tubes

• Reactor vessel support pads

• Seal ledge These external attachments and the expected effects of thermal annealing on them are discussed in the following paragraphs. Inspections and tests for the external attachments after completion of the thermal annealing are addressed in Sections 2.2 and 2.3.

1.2.E.1.1 Nozzle Extensions Nozzle extensions are welded onto each of the six Palisades RV primary nozzles.

The nozzles extensions are A-508-64, Class 1 carbon. steel forgings which allow the carbon-to-carbon field weld to the primary coolant system piping. The nozzle extensions like the RV nozzles are clad internally with weld deposited austenitic stainless steel.

The main effects of the thermal annealing on the nozzle are the loadings due to the thermal displacements of the reactor vessel and piping. The reactor vessel will expand outward as it is heated up during the annealing operations thereby causing compressive loading of the nozzles extensions. Additionally, the variation in the heating from the core region up to the vessel flange will impose a bending moment loading on the nozzle extensions. The nozzle extensions will be loaded in tension with decreasing bending moments during annealing cooldown. Since only the reactor vessel is being heated during the annealing operations as opposed to the entire primary coolant system during reactor operation, the compressive and tensile 12/11 /95 1 .2-12

thermal loadings on the nozzles extensions associated with annealing are

• enveloped by the design loads even though the maximum annealing temperature is higher. However, the maximum bending moment loadings at the RV nozzle extensions exceed the design thermal loadings. The loadings at the nozzle extensions due to thermal annealing are developed from the thermal and structural analyses in Section 1 . 7 and are justified by either comparison with the design loadings or by stress analysis in accordance with Section Ill of the ASME Boiler and Pressure Vessel Code.

The primary coolant system piping is discussed in Sections 1.3.A.1 and 1.3.B.1.

1.2.E.1.2 0-ring Leakage Monitor Tubes The 3/4 inch schedule 80 0-ring leakage monitor tubes at the vessel flange are connected to a piping system which pipes any leakage to the drain tank. The monitor tubes are ASME SB-166 Ni-Cr-Fe Alloy 600 and are welded into the RV flange penetrations by Ni-Cr-Fe alloy partial penetration welds at the flange mating surface. The monitor tubes exit the penetrations on the exterior of the RV flange

• j 6° 40' to either side of the 0° axis.

The effects of the thermal annealing on the monitor tubes are loads due to the thermal displacements of the reactor vessel and interaction with the attached piping system. Resolution of the loads is not expected to be an issue since the heating of the vessel during the annealing operations will be concentrated below the primary nozzles well below the RV flange. Therefore, the thermal loads on the monitor tubes and the attached piping due to the thermal displacement of the reactor vessel will be enveloped by the thermal, pressure and seismic loads to which the monitor tubes might be subjected during operation.

The attached leakoff piping system is discussed in Sections 1.3.A.9 and 1.3.B.9.

1.2.E.1.3 Reactor Vessel Support Pads The reactor vessel is supported by the RV support pads on the underside of three (3) of the RV nozzles 120 degrees apart. The support pads, which are bolted to a sole plates and rest on the plates of the RV supports, can slide to accommodate the thermal growth at the support pads as the reactor vessel heats up and cools down. The support pads are SA-508, Class 2 low alloy steel forgings which are welded to low alloy steel weld build-up preparations on the nozzles. The bearing surface on the bottom of each support pad is 18 inches x 48 inches.

The nozzle support pads will be subjected to temperatures which are less than the RV design temperature of 650°F during the thermal annealing operations.

Furthermore, the loading of the pad bearing surfaces during annealing will be less than the design loading since the RV head, reactor internals, fuel and water volume will not be supported by the reactor vessel during annealing. Therefore, there are no detrimental effects of thermal annealing on the support pads.

12/11 /95 1.2-13

The RV sliding supports are included in Sections 1.3.A.4 and 1.3.B.4 .

• 1.2.E.1.4 Seal Ledge The seal ledge is a ring of SA-212 carbon steel which is weld attached circumferentially around the top of the RV flange OD. The seal ring facilitates sealing of the reactor cavity for refueling pool flood-up. A stainless steel sheet metal drip pan is attached to the refueling cavity liner extension and is positioned in the annulus all the way around the circumference under the seal ledge for the purpose of collecting minor leakage past the cavity seal.

There are no expected effects of the annealing on these items since they are so far removed from the thermal annealing heating zone that the temperatures should remain within design. Furthermore, the drip pan is not attached to the seal ledge and there is sufficient clearance at the top of the reactor cavity to permit thermal growth so that no interference between the seal ledge and the refueling cavity extension or drip pan will develop at elevated temperatures.

The drip pan and its drain lines are further described and discussed in Sections 1.3.A. 11 and 1.3.B.11.

1.2.E.2 Internal Attachments

• The internal attachments to the Palisades reactor vessel which will be subjected to the effects of thermal annealing include:

• Core support lugs

• Core stabilizing lugs

• Flow skirt

• Surveillance holder assemblies These internal attachments and the their anticipated effects of the thermal annealing are discussed in the subsequent paragraphs. Inspections and tests of the internal attachments after completion of the thermal annealing are addressed in Sections 2.2 and 2.3.

1.2.E.2.1 Core Support Lugs and Core Stabilizing Lugs Nine (9) core support lugs of SB-168 Ni-Cr-Fe Alloy 600 plate are equally spaced around the interior wall of the bottom head just below the bottom head to shell juncture. The core support lugs provide support for the flow skirt and for the core support barrel if a core support barrel drop event should occur .

Six (6) core stabilizing lugs of SB-166 Ni-Cr-Fe Alloy 600 bar are weld attached to 12/11 /95 1. 2-14

the interior wall at the bottom section of the lower shell course. The lugs are

• spaced around the reactor vessel for the purpose of providing lateral support for the bottom of the core support barrel.

The 900°F maximum annealing temperature is well below the heat treatment temperature of the Ni-Cr-Fe Alloy 600 material. Therefore, no degradation of the mechanical properties due to the heating is anticipated. There will be a small degree of grain boundary sensitization when holding the Ni-Cr-Fe Alloy 600 material at 900°F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. However, sensitized Ni-Cr-Fe Alloy 600 material is not a concern in the normal PWR environment.

The core stabilizing lugs are not included in the thermal model in Section 1 . 7 since the coefficients of thermal expansion for the Ni-Cr-Fe Alloy 600 lugs and the carbon-molybdenum steel reactor vessel shell material are very similar throughout the temperature range and the average annealing heatup and cooldown rates are low. The thermal stresses in the core stabilizing lugs are small. The thermal stresses in the core support lugs at the RV wall attachment due to annealing are evaluated in the RV stress analysis in Section 1 . 7, since they are attached to the flow skirt.

1.2.E.2.2 Flow Skirt The flow skirt is a welded assembly of SB-168 Ni-Cr-Fe Alloy 600 plate in the

• bottom head region of the Palisades reactor vessel for the purpose of directing the primary coolant flow into the reactor core. The flow skirt is supported in the bottom head by the nine core support lugs by attachment welds which were performed at the site.

The main effect of the thermal annealing operations on the flow skirt are expected to be stresses at the attachment points resulting from the thermal displacements.

The thermal stresses in the flow skirt are evaluated in the reactor vessel stress analysis in Section 1 . 7.

The SB-1 68 Ni-Cr-Fe Alloy 600 material will be slightly sensitized at temperature during annealing. However, the sensitized material is not a concern in the normal PWR environment.

1.2.E.2.3 Surveillance Holder Assemblies There are six (6) surveillance holder tubes mounted vertically on the inside wall of the reactor vessel below the primary nozzles for the purpose of positioning surveillance capsules on the interior wall of the core region shell. The surveillance holders are attached to the wall by 1 08 SB-166 Ni-Cr-Fe Alloy 600 bracket supports which are weld attached to the RV wall.

• The effect of the thermal annealing temperatures on the surveillance holders ,is thermal stress in the brackets at the RV wall interface. Since the coefficients of 12/11/95 1.2-15

thermal expansion for the carbon-molybdenum steel vessel shell material and the

• Ni-Cr-Fe Alloy 600 support material are very similar over the entire annealing temperature range, the thermal stress at the RV wall interface during annealing will be acceptably small. Even though the support brackets are welded to the austenitic stainless steel cladding, the base metal dictates the thermal expansion through the bond with the thin cladding. The thermal stresses in the support brackets are quantified in the Section 1 . 7.

The Ni-Cr-Fe Alloy 600 bracket material may be slightly sensitized at temperature during annealing. However, the sensitized material is not a concern in the normal PWR environment.

1.2.E.3 Other Equipment, Components, Structures and Instrumentation Equipment, components, structures and instrumentation in and around the reactor cavity other than RV attachments are described along with the expected effects of the thermal annealing in Section 1.3. In addition to the previously identified sections addressing the nozzle extensions, the RV support pads, the leakage monitor tube leakoff piping, and the seal ledge, Section 1 .3 also includes discussions of the effects of thermal annealing on RV insulation, primary coolant.

system piping, RV sliding supports, cavity seal drip pan and drain lines, primary coolant piping insulation, the biological shield and the biological shield insulation.

Moreover, Section 1.3.E addresses the effects of the heat of annealing on instrumentation such as neutron detectors and the primary coolant flow and temperature measurement instrumentation.

1.2.F REFERENCES Combustion Engineering, "Palisades General Reactor Vessel Arrangement",

Drawing Number SE-2005507-322-001, Revision 2. Windsor, CT Fenech, R. A. 1994, "Docket 50-255 License DPR-20, Palisades Plant - Response to the January 9, 1994 NRC Request for Additional Information - 10 CFR 50.61 Screening Criterion", Washington, D. C.

Peter, P. A., Lippincott, E.P., \JYrights, G.N., Madeyski, A., 1994, "Analysis of Capsule W-110 form the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program", WCAP-14014, Pittsburgh, PA.

Porter, N. J. (CE), 1984a, "(Palisades Plant) Reactor Vessel Weld Documentation,"

P-CE-7747, letter to J. B. Toskey (CPCo), Windsor, CT Porter, N. J. (CE), 1984b, "Palisades Vessel Weld Documentation," P-CE-7752, letter to J. B. Toskey (CPCo), Covert, Ml

• RSIC Data Library Collection DLC-76, 1987, "SAILOR, Coupled, Self-Shielded, 47-Neutron, 20-Gamma-Ray, P3 , Cross-Section Library for Light Water Reactors".

12/11 /95 1.2-16

Section Ill of the ASME Boiler and Pressure Vessel Code, Appendix G "Protection

• Against Nonductile Failure" .

Westinghouse, 1983, "Palisades Reactor Vessel Examination Report, Volumes 1-4".

Wrights, G. N., 1994, "Palisades Cycle 11 Reactor Vessel Inner Radius Fast Neutron Flux Data", Westinghouse Letter SE/REA-169/94 to R. Snuggerud, Pittsburgh, PA .

• 12/11 /95 1.2-17

Material Unirradiated Properties Chemical Content (wt%)

. Location Identity Heat USE RTNDT RTNDT s p Cu Ni Method Intermediate D-3803-1 C1279-3 102 -10 Measured 0.020 0.009 0.232 0.51 Shell D-3803-2 A0313-2 87 . -30 MTEB 0.023 0.010 0.24 0.52 D-3803-3 C1279-1 91 0 MTEB 0.020 0.011 0.24 0.50 Lower Shell D-3804-1 C1308-1 72 0 MTEB 0.022 0.016 0.19 0.48 D-3804-2 C1308-3 76 -30 MTEB 0.020 0.015 0.19 0.50 D-3804-3 85294-2 73 -25 MTEB 0.020 0.010 0.12 0.55 Intermediate to 9-112 27204 98 -56 Generic 0.011 0.013 0.21 1.00 Lower Shell (Linde 124 Circumferential Lot #3687)

Weld E8018 #LODG N/A N/A --- N/A N/A N/A N/A Intermediate 2-112 A, W5214 118 -5*6 Generic 0.016 0.017 0.212 1.02 Shell Axial B&C (Linde 1092 Welds Lot# 3617 Ni 200 #N7753A)

Lower Shell 3-112 A, W5214 118 -56 Generic 0.016 0.017 0.21 1.02 Axial Welds B&C 348009 111 -56 Generic N/A N/A 0.19 0.99 (Linde 1092 Lot #3692 &3617 Ni 200 #N7753A)

E8018 #CBBF N/A N/A ---- N/A N/A N/A N/A TABLE 1.2.B-1 .Palisades Reactor Vessel Beltline Materials 12/11 /95 1.2-18

Weld Application Ni Cu Chemistry Heat No. Content Content Factor (wt%) (wt%)

Reactor Vessel 1.02 0.212 232 W5214 Supplemental 1.045 0.307 267 Surveillance Reactor Vessel 0.99 0.19 219 348009 Supplemental 1.121 0.185 227 Surveillance Reactor Vessel 1.00 0.21 229

• 27204 3277 Supplemental Surveillance Original Surveillance 1.067 1.388 0.194 0.246 228 270 1 Notes: 1 Weld 3277 Chemistry Factor based on Surveillance Data from Capsules W-290 and W-110. The High Nickel Content Prevents Interpolation of the Chemistry Factor from 1 OCFR 50.61, Table 1 Table 1.2.C-1 Comparison of Best Estimate Reactor Vessel Weld Chemistry and Surveillance Weld Chemistry for Palisades

• 12/11 /95 1 .2-19

• Weld Seam Description Type Indication 1-11 2A, 1-11 28, Upper Shell Axial Seam Surface Defects, 1-112C Porosity, Slag Inclusions 2-112A, 2-112C Intermediate Shell Surface Defects Axial Seam 3-112A, 3-1128, Lower Shell Axial Seam Slag Inclusions, 3-112C Surface Defects, Porosity 7-112 Upper Shell to Flange Surface Defects,

~

Porosity, Slag Inclusions, Overlap 8-112 Upper to Intermediate Slag Inclusions, Overlap, Shell Porosity

• 10-112 1-113A, 1-11 3B, 1-113D Bottom Head to Lower Shell Bottom Head Torus Slag Inclusion, Porosity, Surface Defects Porosity 4-113 Bottom Head Dome Porosity 1-114D Inlet Nozzle Extensions Porosity, Slag Inclusion 1-114E Outlet Nozzle Extensions Processing Marks 5-114A, 5-1148, Inlet Nozzle to Shell Slag Inclusions, Overlap, 5-114D Surface Defects, Processing Marks, Porosity 5-114F Nozzle to Shell Slag Inclusion TABLE 1.2.C-2 Palisades Reactor Vessel Recordable Radiographic Indications

• 1 2/11 /95 1.2-20

I Weld Seam I Location I

7-112 Flange to Upper Shell Various (90 ° - 360 °)

8-112 Upper to Intermediate Shell 270° 9-112 Intermediate to Lower Shell 60 O I 315 O 10-112 Lower Shell to Bottom Head 80° 3-112B&C Lower Shell Axial Various 1-113A&B Bottom Head Torus Axial Various 1-114C Inlet Nozzle to Nozzle Extension Various

• 5-114A 5-114B 5-114C 5-114C Inlet Nozzle to Shell Inlet Nozzle to Shell Inlet Nozzle to Shell Inlet Nozzle to Shell 120 25° O I 260° 225 Base metal O

5-1140 Inlet Nozzle to Shell Base metal Table 1.2.C-3 Palisades Reactor Vessel Recorded Weld Repairs

• 12/11 /95 1.2-21

0° AZIMUTHAL ANGLE

• Cycle 1

2 Cycle Length (EFPD) 379.4' 449.1 Neutron Flux (n/cm 2 -sec) 4.59x10 10 4.59x10 10 Cycle Fluence (n/cm 2 )

1.50x10 18

1. 78x10 18 Total Fluence (n/cm 2 )

1.50x1018 3.28x10 18 3 349.5 4.59x10 10 1.39x10 18 4.67x10 18 4 327.6 4.59x10 10 1.30x10 18 5.97x10 18 5 394.6 4.59x10 10 1.56x10 18 7.53x10 18 6 333.4 4.87x10 10 1.40x10 18 8.94x10 18 7 369.9 4.87x10 10 1.56x10 18 1.05x10 19 8 373.6 2.16x10 10 6.97x10 17 1.12x10 19 9 298.5 2.08x10 10 5.36x10 17 1.17x10 19 10 356.9 1.51x10 10 4.66x10 17 1.22x10 19 11 430.4 1.31x1010 5.09x10 17 1.27x10 19 16° AZIMUTHAL ANGLE

• Cycle 1

2 Cycle Length (EFPD) 379.4 449.1 Neutron Flux (n/cm 2 -sec) 6.03x10 10 6.03x10 10 Cycle Fluence (n/cm 2 )

1.98x10 18 2.34x10 18 Total Fluence (n/cm 2 )

1.98x10 18 4.31x10 18 3 349.5 6.03x10 10 1.82x1018 6.13x10 18 4 327.6 6.03x10 10 1.71x10 18 7.84x10 18 5 394.6 6.03x10 10 2.05x10 18 9.89x10 18 6 333.4 6.25x10 10 1.80x1018 1.17x10 19 7 369.9 6.25x10 10 2.00x10 18 1.37x1019 8 373.6 4.89x10 10 1.58x10 18 1.53x1019 9 298.5 3.06x10 10 7.89x10 17 1.61x10 19 10 356.9 2.40x10 10 7.40x10 17 1.68x10 19 11 430.4 2.07x10 10 7.70x10 17 1. 76x10 19 TABLE 1.2.C-4 Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface (Page 1 of 2)

• 12/11 /95 1.2-22

• Cycle 1

Cycle Length (EFPD) 379.4 30° AZIMUTHAL ANGLE Neutron Flux (n/cm 2 -sec) 4.70x10 10 Cycle Fluence (n/cm 2 )

1.54x10 18 Total Fluence (n/cm 2 )

1.54x10 18 2 449.1 4. 70x10 10 1.82x1018 3.36x10 18 3 349.5 4.70x10 10 1.42x1018 4. 78x10 18 4 327.6 4.70x10 10 . 1.33x10 18 6.11x10 18 5 394.6 4.70x10 10 1.60x10 18 7.71x10 18 6 333.4 4.79x10 10 1.38x10 18 9.09x10 18 7 369.9 4. 79x10 10 1.53x10 18 1.06x10 19 8 373.6 2.34x10 10 7.55x10 17 1.14x10 19 9 298.5 2.00x10 10 5.16x10 17 1.19x10 19 10 356.9 1.94x1010 5.98x10 17 1.25x10 19 11 430.4 1.59x1010 5.91x10 17 1.31x10 19 45° AZIMUTHAL ANGLE Cycle Length Neutron Flux Cycle Fluence Total Fluence Cycle (EFPD) (n/cm 2 -sec) (n/cm 2 ) (n/cm 2 )

1 379.4 2.98x10 10 9. 78x10 17 9.78x10 17 2 449.1 2.98x10 10 1.16x10 18 2.13x10 18 3 349.5 2.98x10 10 9.00x10 17 3.04x10 18 4 327.6 2.98x10 10 8.44x10 17 3.88x10 18 5 394.6 2.98x10 10 1.02x1018 4.90x10 18 6 333.4 3.03x10 10 8. 73x10 17 5. 77x10 18 7 369.9 3.03x10 10 9.68x10 17 6.74x10 18 8 373.6 1. 77x10 10 5.71x10 17 7.31x10 18 9 298.5 1.15x10 10 2.97x10 17 7.61x10 18 10 356.9 1.32x1010 4.07x10 17 8.02x10 18 11 430.4 1.04x1010 3.87x10 17 8.41x10 18 TABLE 1.2.C-4 Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel

• 12/11 /95 Clad/Base Metal Interface (Page 2 of 2) 1.2-23

Weld# Weld Type 1983 Results 1995 Results

• Indication Type Depth From m Surface Description Vessel Azimuth Distance From Flange Status (Inches) (Degrees) (Inches) 2-112A lnterm. Shell 60°L Indication Verified/70 ° L/Recorded Embedded 0.6 267 184.3 Note 1 Long Seam 2-1128 Intermediate 60°S Indication Not Verified Embedded 1.4 27.5 172 Note 1 Shell 31 %DAC Long Seam 60°L,Note 2 Verified/70 ° L/Recorded Embedded 0.3 27 140 Note 1 & 2 2-112C lnterm. Shell 60°L Indication Verified/70 ° L/NRI Embedded 0.5 153 219.66 Note 1 Long Seam 60°L Indication Verified/70 ° L/NRI Embedded 0.6 153.7 146 Note 1 60°L Indication Not Verified Embedded 0.5 147 128 Note 1 9-112 lnterm/Low NRI NRI NRI NRI NRI NRI NRI Circ.

3-112A Lower Shell 60°L Indication Verified/70 ° L/NRI Embedded 1.25 91 308 Note 1 Long Seam 60°L Indication Verified/70 ° L/NRI Embedded 0.78 89.7 306 Note 1 60°L Indication Verified/70 ° L/NRI Embedded 0.85 90.7 294 Note 1 60°L Indication Verified/70 ° L/Recorded Embedded 0.4 87 269 Note 1 60 ° L Indication Not Verified Embedded 0.5 88 220 Note 1 & 3 3-1128 Lower Shell NRI NRI NRI NRI NRI NRI NRI Long Seam 3-112C Lower Shell 60 ° S Indication Verified/45 °S/NRI Embedded 2.16 330 237 Note 1 Long Seam 60 ° L Indication Verified/7 0 ° L/Recorded Embedded 0.46 330 250 Note 1 60 ° L Indication Verified/7 0 ° L/Recorded Embedded 0.66 331 246 Note 1 60°L Indication Verified/70 ° L/Recorded Embedded 0.6 _J 330 239 Note 1 Notes: 1) Indication acceptable per Section XI, Paragraph IWB 3500

2) Small pocket of re-heat cracks

3) Vessel elevation listed is uppermost location of indication scanning boundry Table 1.2.C-5 Angle Beam Indication Results For Palisades Reactor Vessel Beltline Area 12/11 /95 1.2-24

Item Nominal Dimension I I (Inches) I RV Inside Diameter (Beltline) 172 RV Inside Diameter (at Outlet Nozzle) 165-1/2 RV Inside Diameter (at CSB Ledqe) 166 RV Wall Thickness (Beltline Minimum, Excluding 8-1/2 Cladding RV Claddinq Thickness (Nominai) 1I-.-

I II RV Claddinq Thickness (Minimum) 3/16 Outlet Nozzle Inside Diameter (at Safe End) 42 Outlet Nozzle Inside Diameter (at RV l.D.) 48

• Inlet Nozzle Inside Diameter (at Safe End)

Inlet Nozzle Inside Diameter (at RV l.D.)

Too of RV Flange to Lower RV Tanqent Line Too of RV Flanqe to Center Line of Nozzles 30 35-1 /8 305-7/16 75-1 /2 Too of RV Flanae to RV/Closure Head Matina Surface 1-1/2 RV /Closure Head Mating Surface to CSB Mating 8-1 /4 Surface (As Built)

Center Line of RV to Core Stabilizina Lua 77-5/32 Center Line of RV to Core Support Luq 73-11 /16 Top of Support Lug to CSB Mating Surface (As Built) 315-1 /2 TABLE 1.2.D-1 Nominal Reactor Vessel Dimensions

• 1 2/11 /95 1.2-25

LOOP 1 LOOP 1B LOOP 2A LCOP 2 LOOP 2B LOOP 1A OUTL.ET#1 1NLET111 lNU::-1112 OUTLEl#Z INLET#.3 INLF.Tn4 0* 60' 120* 1s0* 240' 300' I I I I 270' I 360' I I I

-7494.lillil

.*j

<( CJ

-~ N

- I -I 1o----------a-112--~1o----

---

~-.------1--------------

~-i:q i !_c:,;i J 1

~ I i;:* 132. r2ni N

..- ACl IVE FUEL ..-

~ ~:~T ~

' ID-3:~::;_'~'-~~1o-----__....______....,_____1t_J--3_F._.~_3_-'_'__... o.... -...o...~""'--lo_-_3_B0_3_-.3-I. ........,

ID-330t,-1( N I

I()

~----*- .. - -~--- ........ *-----+----------1~-----

Figure 1.2.A-1 Palisades Reactor Vessel Materials 12/11 /95 1.2-26

• Al OUTLET NOZZLE 180" INLET NOZZLE 90" 270*

• A

~60" OUTLET NOZZLE a J VESSEL PLAN VIEW Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 1 of 14) 12/11/95 1 .2-27

t0.7~MIH EXCUJOING CUii>.

11 .2 £ST. MAX 126.500 APPROX 112CS.OOO

• 218.000 11182.200

• 172.boo I

I

---

~-------

---

~

- ; - - - - -.......+ - - - - - - + - 6 X 135.000  ;~ I 1i--~---11s.aoo-----1 r--------------~--~305.4.JS-----------------1 i-------------------329.625-------------------1 r------------:tr-----------~-~398,125-----------------------1 SECTION A-A Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 2 of 14) 12/11 /95 1.2-28

t 14.4-37 J_ J_

1.843 4.180 APPROX INLET N*OZZLE ELEVATION CROSS SECTION Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 3 of 14) 12/11 /95 1.2-29

TO VESSEL C/L TO VESSEL C/L R86.000

+

119.000 123.562 APPROX 18.000

- ,~';,;_1:

I- - - - - - 1.531---,

INLET NOZZLE ELEVATION CROSS SECTION Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 4 of 14) 12/11/95 1.2-30

i------ 050.625 ---------.~

I~~t----- !642.000 - - - . - i I R 6.375 4.000 1.688 APPROX 15.093 126.500

~

APPROX

+

TO VESSEL C/L W.P._/~~--- 1648.000 -----1*-I'- W.P. I 82.750 OUTLET NOZZLE ELEVATION CROSS SECTION +

TO VESSEL C/L Figure 1.2.D-1 Palisades General Reac~or Vessel Arrangement (Page 5 of 14) 12/11/95 1.2-31

!""""*~~- 41.531 _ ___,--i 6.250 4.750 I

18.000

_i l

119.000 t

TO VESSEL C/L

---

~68.062-------i OUTLET NOZZLE ELEVATION CROSS SECTION Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 6 of 14) 12/11 /95 1 . 2-32

E 1.500 TO VESSEL C/L 108.500 54X 7-8N-2B R1.000 2.250 -

i---- 17.625 *-

- - 20.500 - -

- - - - 21.625 - - - - - i

~--*-- 29.000 - - - - i DETAIL W Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 7 of 14) 12/11 /95 1.2-33

Ni-Cr-Fe WELD

-**- - 84.062 --Y-- TO VESSEL i-----~172_000 + C/L D[TAIL X Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 8 of 14) 12/11 /95 1.2-34

2-500 4r*x5.500 I

• 135.000---------------i 8X 10.053 DETAIL Y F DETAIL Z Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 9 of 14) 12/11/95 1.2-35

• * - 3X 48.000 r------+-- 3X 24.000 i---+---+---+- 3X 2.5.00

-j 3X 12.500 180" I

r 240" II II

• x go* -----1.._,,._;.. - - - -- - - - -- -

I

-+- - - - - - - - - - - - -

/

I I

I I

I 60" Ir

--t-r I

I I

0-SECTION 8-G

• Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 10 of 14) 12/11 /95 1 .2-36

• 17ff 180" II II 2so*

I

---+-------

I go* ...t::::=:EI-------

1 I

I I

I I

1ff 0*

SECTION C-C

• Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 11 of 14) 12/11/95 1.2-37

• TO VESSEL C/L H~.

t 73.687 91146.375 i=-:-28.125~

t 121172.000 I

I

.250 NOM CLAD .

. 1BB MIN CLAD.

~ 305.4-38----<-

• TO TOP OF VESSEL FLANGE V

/\_ _ 324.406----1"""

C/L WELD SECTION .,...._--

D-D 4.953 KF:fWAY #3 3.953 KEYWAY #1.2&:4 VIEW E-E

• Figure 1.2.D-1 12/11 /95 Palisades General Reactor Vessel Arrangement (Page 12 of 14) 1.2-38

• TO VESSEL C/L t77r6 --j .9.000t L_:tl 7.000 VIEW F-F Ni-Cr-Fe TIE TO STAINLESS STEEL CLAD STAINLESS CLAD VIEW G-G

• Figure 1.2.D-1 12/11 /95 Palisade~ General Reactor Vessel Arrangement (Page 13 of 14) 1.2-39

J *"

SECTION H-H

• Ni-Cr-Fe ALLOY f.188 THRU SECTION J-J 2 PLACES

- Figure 1.2.D-1 Palisades General Reactor Vessel Arrangement (Page 14 of 14) 12/11 /95 1.2-40