Text

Documentation for Pressurized Thermal Shock Evaluation Meeting
	ML101540386
Person / Time
Site:	Palisades
Issue date:	06/02/2010
From:	Patricia Anderson; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	WCAP-15353, Rev 0
	Download: ML101540386 (221)
	v • d • e

{{#Wiki_filter:Entergy Nuclear Operations, Inc. -- Entergy Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Paula K. Anderson Licensing Manager June 2,2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Documentation for Pressurized Thermal Shock Evaluation Meeting Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

On February 11, 2010, Entergy Nuclear Operations, Inc. (ENO) and the Nuclear Regulatory Commission (NRC) had a conference call to discuss ENO plans to revise the Palisades Nuclear Plant (PNP) pressurized thermal shock (PTS) analysis. During the conference call, a future meeting at the NRC was proposed to discuss in more detail ENO plans to revise the PTS analysis. Subsequently, a meeting was scheduled between ENO and the NRC on June 22, 2010. The attached documentation is provided in support of the meeting. Attached you will find (1) a PNP reactor vessel fluence evaluation, (2) a preliminary PTS screening criteria assessment, (3) a documented review of the PTS screening criteria assessment, (4) a list of ENO representatives planning to attend the meeting, and (5) a meeting agenda. The attachments contain no proprietary information. This letter contains no new commitments and no revisions to existing commitments. s~ pka/jse

Document Control Desk Page 2 Attachment(s): 1. Background Information

2. Reactor Pressure Vessel Fluence Evaluation

3. Preliminary PTS Screening Criteria Assessment

4. Westinghouse Review of PTS Screening Criteria Assessment

5. List of ENO Representatives Planning to Attend Meeting

6. Planned Meeting Agenda cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ATTACHMENT 1 Background Information 3 pages follow

Background Information On March 22, 2005 (ADAMS accession no. ML050940446), Nuclear Management Company, LLC (former plant operator) submitted a license renewal application (LRA) to the Nuclear Regulatory Commission (NRC) to renew the operating license for the Palisades Nuclear Plant (PNP). The LRA included discussion of a time limiting aging analysis (TLAA) for pressurized thermal shock (PTS) of the PNP reactor pressure vessel (RPV). NRC issued a safety evaluation report (SER) related to the license renewal of the PNP on September 28,2006 (ML062710068). In January 2007, NRC issued NUREG-1871, "Safety Evaluation Report Related to the License Renewal of Palisades Nuclear Plant" (ML070600578). In the SER, it is noted that the RPV is projected to exceed the PTS screening criteria in 2014. The axial welds at the 60° locations, fabricated with weld wires of heat W5214, are the limiting materials in the beltline region of the PNP RPV. The SER states that if an RPV is projected to exceed the PTS screening criteria, 10 CFR 50.61(b)(3) requires flux reduction be implemented reasonably practicable to avoid exceeding the PTS screening criteria. If the flux reduction program does not prevent the RPV from exceeding the PTS screening criteria before the end of the operating license, the applicant, in order to meet PTS requirements, can choose in 10 CFR 50.61 between two options:

1. Submittal of a plant-specific safety analysis, pursuant to 10 CFR 50.61 (b)(4), to determine what, if any, modifications to equipment, systems, and plant operation are necessary to prevent failure of the RPV from a postulated PTS event. This analysis must be submitted at least three years before RT PTS is projected to exceed the PTS screening criteria.

2. Perform a thermal annealing treatment of the RPV, pursuant 10 CFR 50.61 (b )(7), to recover fracture toughness.

However, recognizing that the RPV welds are expected to exceed 10 CFR 50.61 PTS screening criteria during the period of extended operation, Entergy Nuclear Operations (ENO)(current plant owner) has chosen the 10 CFR 54.21 (c)(1 )(iii) option for managing the PTS TLAA. Per Section 4.2.2.2 of the SER, this option involved the following activities: a) An assessment of the current licensing basis TLAA for PTS, b) A discussion of the flux reduction program implemented in accordance with 10 CFR 50.61 (b )(3), and c) An identification of viable options for managing the aging effect in the future ("Pressurized Thermal Shock Analyses for Renewal of Certain Nuclear Power 1 of 3

Plant Operating Licenses," Executive Director Memo to Commissioners, dated May 27, 2004, ML041190564). The license renewal SER states that, prior to exceeding the PTS screening criteria limit, PNP will select the best option to manage PTS in accordance with NRC regulations and make required submittals to obtain NRC review and approval. Subsequent to NRC approval of the LRA, the NRC issued 10 CFR 50.61 a to provide an alternative method for evaluating PTS for plants projected to exceed the screening criteria limit. ENO intends to use this alternative method for evaluating PTS for the PNP RPV at the appropriate time to meet the regulation. The purpose of the scheduled June 22, 2010, meeting with the NRC is to discuss the status of activities recently completed and planned in the near future; under 10 CFR 54.21 (c)(1 )(iii), involving the PTS analysis. Activities recently completed include: .

1. An update of Westinghouse Report, WCAP-15353, "Palisades Reactor Pressure Vessel Neutron Fluence Evaluation," Revision 0, January, 2000, to account for actual capacity factors and core loading patterns since that time.

The update is documented in WCAP-15353-Supplement 1-NP, "Palisades Reactor Pressure Vessel Fluence Evaluation," Revision 0, April 2010 (Attachment 2).

2. An assessment of embrittlement for the limiting axial weld heat number W5214 relative to the PNP RPV.

A new preliminary assessment for the RPV projecting RT PTS and the revised date when weld heat W5214 will exceed the PTS screening criteria of 270°F is contained in Structural Integrity Associates (SIA) Report No. 0901132.401, "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," Revision 0, April 2010 (Attachment 3). This new preliminary assessment uses the updated fluence evaluation performed by Westinghouse and all available surveillance data relevant to the PNP RPV limiting axial weld heat number W5214 as permitted by 10 CFR 50.61 (c)(2). A review of this report was performed by Westinghouse (Attachment 4). The preliminary SIA PTS screening assessment report indicates that, based on the new information, the revised projected date for the PNP RPV to reach the PTS screening criteria limit of 270°F would be approximately April 2017 or later. 2 of 3

10 CFR 50.61 (c)(3) states that any information believed to improve the accuracy of the RT PTS value significantly must be reported to the NRC and any value or RT PTS that has been modified using the procedures of 10 CFR 50.61 (c)(2) is subject to NRC approval. In accordance with 10 CFR 50.61 (c)(3), at a later date, END plans to report new information that improves the accuracy of the RT PTS value to the NRC and to submit the updated fluence evaluation and the new PTS screening assessment report for NRC review and approval. Based upon this new projection when the PNP RPV will reach the PTS screening criteria, ENG has identified the need for the following future activities:

1. Reschedule of the volumetric inspection for the PNP RPV from the fall 2010 refueling outage to the spring 2012 refueling outage.

2. Update and submit the PNP RPV heatup and cooldown curves located fn the PNP Technical Specifications for operation through spring 2017.

3. Following inspection of the PNP RPV in spring 2012, perform an assessment of the PNP RPV in accordance with 10 CFR 50.61 a.

4. Following NRC review and approval of the PTS assessment of the PNP RPV under 10 CFR 50.61 a, update and transmit the PNP RPV heatup and cooldown curves located in the PNP Technical Specifications for operation through the balance of the extended licensed life of the plant.

The ultrasonic inspection of the PNP RPV planned in the spring of 2012 and subsequent evaluation and submittal in accordance with 10 CFR 50.61a will satisfy the PTS regulatory requirements through the remaining period of extended operation. 3 of 3

ATTACHMENT 2 Reactor Pressure Vessel Fluence Evaluation W CAP-15353-N P, "Palisades Reactor Pressure Vessel Fluence Evaluation," Revision 0 - Supplement 1, May 2010 29 pages follow

Westinghouse Non-Proprietary Class 3 WCAP-15353 - -Supplement 1-NP May 2010 Revision 0 Palisades Reactor Pressure Vessel Fluence Evaluation ewe inghou e

Westinghouse Non-Proprietary Class 3 WCAP-15353 - Supplement 1-NP, Revision 0 Palisades Reactor Pressure Vessel Fluence Evaluation

s. L. Anderson*, Fellow Engineer Radiation Engineering & Analysis May 2010 Reviewed:

P. M. Song*, Principal Engineer Radiation Engineering & Analysis Approved: A. R. Dulloo*, Manager Radiation Engineering & Analysis Work Performed Under Shop Order 450 Purchase Order No. 66941 Prepared by Westinghouse for WESTINGHOUSE ELECTRIC COMPANY LLC P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355

                            © 2010 Westinghouse Electric Company LLC All Rights Reserved

Elech'onically approved records are authenticated in the electronic document management system.

Westinghouse Non-Proprietary Class 3 EXECUTIVE

SUMMARY

Calculations of the neutron exposure of the Palisades reactor pressure vessel were previously completed and documented in WCAP-15353, Revision 0, "Palisades Reactor Vessel Neutron Fluence Evaluation," January 2000Y] This evaluation was submitted for review by the NRC Staff and, after consideration of RAI's addressed in Reference 4, the fluence methodology as well as the final results were approved by the Staff. The f1uence' analysis described in WCAP-15353, Revision 0[3] included cycle specific evaluations through Cycle 14 (the then current operating cycle). This supplement to WCAP-15353 provides an updated neutron fluence assessment for the Palisades pressure vessel beltline region that includes cycle specific analysis for additional operating cycles for which the design has been finalized (Cycles 15 through 21) and includes projections for future operation through approximately 44 effective full power years (EFPY). Updated evaluations of surveillance capsule credibility analysis and determination of material chemistry factors are being completed in parallel with this fluence calculation and will be documented elsewhere. Based on the cycle specific analysis through Cycle 21 (approximately 23.4 EFPY) and the projection scenario for future operation provided by Entergy, the maximum neutron exposure of the pressure vessel beltline materials through 44 EFPY is summarized as follows. Neutron (E > 1.0 MeV) Fluence End (n/cm2) of Estimated Cumulative Fuel Calendar Time Cycle Date (EFPY) o Deg. 15 Deg. 30 Deg. 45 Deg_. 21 10/2010 23.4 1.447E+19 2.114E+19 1.590E+19 9.677E+18 22 4/2012 24.7 1.496E+19 2.201E+19 1.652E+19 1.001 E+19 23 10/2P13 26.1 1.545E+19 2.288E+19 1.717E+19 1.038E+19 24 4/5/2015 27.4 1.592E+19 2.372E+19 1.779E+19 1.073E+19 25 10/2016 28.8 1.642E+19 2.461E+19 1.844E+19 1.110E+19 26 4/2018 30.2 1.691 E+19 2.549E+19 1.909E+19 1.147E+19 27 10/2019 31.5 1.741E+19 2.637E+19 1.975E+19 1.184E+19 28 4/2021 32.9 1.790E+19 2.726E+19 2.040E+19 1.221 E+19 29 10/2022 34.3 1.840E+19 2.814E+19 2.105E+19 1.258E+19 30 4/2024 35.7 1.889E+19 2.903E+19 2.170E+19 1.295E+19 31 10/2025 37.1 1.939E+19 2.991E+19 2.236E+19 1.332E+19 32 4/2027 38.4 1.988E+19 3.079E+19 2.301 E+19 1.369E+19 33 10/2028 39.8 2.038E+19 3.168E+19 2.366E+19 1.406E+19 34 4/2030 41.2 2.087E+19 3.256E+19 2.432E+19 1.443E+19 35 10/2031 42.6 2.137E+19 3.344E+19 2.497E+19 1.480E+19 36 4/2033 44.0 2.186E+19 3.433E+19 2.562E+19 1.517E+19 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 Neutron (E > 1.0 MeV) Fluence End (n/cm2) of Estimated Cumulative Fuel Calendar Time Cycle Date (EFPY) 60 Deg. 75 Deg. 90 Deg. 21 10/2010 23.4 1.472E+19 2.157E+19 1.575E+19 22 4/2012 24.7 1.520E+19 2.252E+19 1.647E+19 23 10/2013 26.1 1.571E+19 2.345E+19 1.717E+19 24 4/5/2015 27.4 1.619E+19 2.433E+19 1.784E+19 25 10/2016 28.8 1.670E+19 2.527E+19 1.854E+19 26 4/2018 30.2 1.721E+19 2.621E+19 1.925E+19 27 10/2019 31.5 1.772E+19 2.714E+19 1.995E+19 28 4/2021 32.9 1.823E+19 2.808E+19 2.065E+19 29 10/2022 34.3 1.874E+19 2.902E+19 2.136E+19 30 4/2024 35.7 1.925E+19 2.995E+19 2.206E+19 31 10/2025 37.1 1.976E+19 3.089E+19 2.277E+19 32 4/2027 38.4 2.027E+19 3.182E+19 2.347E+19 33 10/2028 39.8 2.078E+19 3.276E+19 2.417E+19 34 4/2030 41.2 2.129E+19 3.370E+19 2.488E+19 35 10/2031 42.6 2.180E+19 3.463E+19 2.558E+19 36 4/2033 44.0 2.231E+19 3.557E+19 2.628E+19 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 TABLE OF CONTENTS TABLE OF CONTENTS LIST OF TABLES ii LIST OF FIGURES iii

1.0 INTRODUCTION

1-1 2.0 NEUTRON TRANSPORT CALCULATIONS 2-1 2.1 Method of Analysis 2-1 2.2 Calculated Neutron Exposure of Pressure Vessel Beltline Materials 2-5 3.0 NEUTRON DOSIMETRY EVALUATIONS 3-1 3.1 Method of Analysis 3-1 3.2 Dosimetry Evaluations 3-5 4.0 SURVEILLANCE CAPSULE NEUTRON FLUENCE 4-1

5.0 REFERENCES

5-1 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 ii LIST OF TABLES Table Title Page 2.2-1 Summary of Calculated Maximum Pressure Vessel Neutron Flux 2-7 (E> 1.0 MeV) for Cycles 15 through 21 and for Future Projection. 2.2-2 Summary of Calculated Maximum Pressure Vessel Neutron Exposure 2-8 Through the Conclusion of Cycle 21. 2.2-3 Projections of Calculated Maximum Pressure Vessel Neutron Exposure. 2-9 3.2-1 Comparison of Measured and Calculated Threshold Foil Reaction Rates. 3-6 3.2-2 Comparison of Adjusted and Calculated Exposure Parameters. 3-7 4.0-1 Summary of Neutron Fluence (E > 1.0 MeV) Derived from the Application 4-2 of Methodology Meeting the Requirements of Regulatory Guide 1.190. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 iii LIST OF FIGURES Figure Title Page 2.1-1 Palisades r,8 Reactor Geometry 2-3 2.1-2 Palisades r,z Reactor Geometry 2-4 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 1-1 SECTION

1.0 INTRODUCTION

In the assessment of the state of embrittlement of light water reactor (LWR) pressure vessels, an accurate evaluation of the neutron exposure of each of the materials comprising the beltline region of the vessel is required. In Appendix G to 10 CFR 50[1], the beltline region is defined as "the region of the reactor vessel shell material (including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the reactor core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage". Each of the materials comprising the beltline region must be considered in the overall embrittlement assessments for the pressure vessel. Therefore, plant-specific exposure assessments must include evaluations as a function of position over the beltline region. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [2], describes state-of-the-art calculation and measurement procedures that are acceptable to the NRC staff for determining pressure vessel fluence. Also included in Regulatory Guide 1.190 is a discussion of the steps required to qualify and validate the methodology used to determine the neutron exposure of the pressure vessel wall. One important step in the validation process is the comparison of plant-specific neutron calculations with available measurements. In early 2000, WCAP-15353, Revision 0[3] describing the methodology used in the fluence evaluations for the Palisades plant was submitted to the NRC staff for review. Subsequent to that review and a further exchange of information documented in Reference 4, the methodology described in WCAP-15353, Revision 0 was approved for application to the Palisades reactor pressure vessel. Subsequent to that approval additional submittals[7,B] in support of the benchmarking of this

                  , fluence methodology were reviewed and approved by the NRC Staff.

The fluence analysis described in WCAP-15353, Revision 0[3] included cycle specific evaluations through Cycle 14 (the then current operating cycle). This supplement to WCAP-15353 provides an updated neutron fluence assessment for the Palisades pressure vessel beltline region that includes cycle specific analysis for additional operating cycles for which the design has been finalized (Cycles 15 through 21) and includes projections for future operation through approximately 44 effective full power years (EFPY). The results of this evaluation are intended for use as input to vessel materials analyses (to be documented elsewhere) that include updates to surveillance capsule credibility analysis and material chemistry factor determination. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 1-2 Since the PTS screening criterion determination for the Palisades pressure vessel requires the evaluation of all weld heat W5214 surveillance capsule data from Palisades and other PWR's, this report also includes the latest fluence evaluation from capsules containing the W5214 material. This compilation of capsule fluence values is based on the same fluence methodology described in this report. In subsequent sections of this supplement, the methodologies used to perform neutron transport calculations and dosimetry evaluations are described in some detail, the updated results of the plant specific transport calculations are given for the beltline region of the Palisades pressure vessel. Comparisons of calculations and measurements demonstrating that the transport calculations meet the requirements of Regulatory Guide 1.190 that were previously included in Reference 3 are also included in this supplement for completeness. Finally, a listing of updated neutron fluence values based on the use of an approved Regulatory Guide 1.190 compliant fluence methodology for several previously withdrawn surveillance capsules that contain Palisades vessel materials is provided for use in data correlation studies. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-1 SECTION 2.0 NEUTRON TRANSPORT CALCULATIONS As noted in Section 1.0 of this report, the exposure of the Palisades pressure vessel was developed based on a series of fuel cycle-specific neutron transport calculations validated by comparison with plant-specific measurements. Measurement data used in the validation process were obtained from both in-vessel and ex-vessel capsule irradiations. In this section, the neutron transport methodology is discussed in some detail, and the calculated results applicable to the in-vessel surveillance capsules and the pressure vessel beltline materials are presented. A discussion of the Palisades dosimetry evaluations and measurement to calculation comparisons is included in Section 3.0 of this supplement. 2.1 - Method of Analysis In performing the fast neutron exposure evaluations for the Palisades reactor, plant-specific forward transport calculations were carried out using the three-dimensional flux synthesis technique described in Section 1.3.4 of Regulatory Guide 1.190. In particular, the following single channel synthesis approach was employed for all fuel cycles:

                                                             <P(r, z)
                                     <P(r, 8, z) =<P(r, 8) *  <P(r) where     ~(r,e,z) is the synthesized three-dimensional neutron flux distribution,         ~(r,e)  is the transport solution in r,e geometry,  ~(r,z)  is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and       ~(r)  is the one-dimensional solution for a cylindrical reactor model using the same source-per-unit height as that used in the r,e two-dimensional calculation.

For the Palisades analysis, all of the transport calculations were carried out using the DORT two-dimensional discrete ordinates code Version 3.2[5] and the BUGLE-96 cross section library[6]. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a P5 legendre expansion and the angular discretization was modeled with an S16 order of angular quadrature. The geometry used for the Palisades transport analysis is discussed in some detail in Reference 3 and the geometric model established for Cycle 15 and beyond was also used for the current evaluations. A plan views of the r,e model of the reactor geometry at the core midplane is shown in Figure 2.1-1. This model depicts a single quadrant of the reactor. A section view of the r,z model of the Palisades reactor is shown in Figure 2.1-2. The model WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-2 extended radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation one foot below the active fuel to an axial elevation one foot above the active fuel. The one-dimensional radial model used in the synthesis procedure consisted of the same radial mesh intervals included in the r,z model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry. The core power distributions used in the plant-specific transport analysis for the reactor were provided by Entergy.[15] The data used in the source generation included fuel assembly-specific initial enrichments, beginning-of-cycle burnups and end-of-cycle burnups. Appropriate axial distributions were also obtained. For each fuel cycle of operation, the fuel assembly-specific enrichment and burnup data were used to generate the spatially-dependent neutron source throughout the reactor core. This source description included the spatial variation of isotope dependent (U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242) fission spectra, neutron emission rate per fission, and energy release per fission based on the burnup history of individual fuel assemblies. These fuel assembly-specific neutron source strengths derived from the detailed isotopics were then converted from fuel pin cartesian coordinates to the [r,e], [r,z], and [r] spatial mesh arrays used in the DORT discrete ordinates calculations. This same qualified methodology was used along with reactor specific input in the determination of the surveillance capsule fluence values discussed in Section 4.0 of this report. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-3 Figure 2.1-1 Palisades r,a Reactor Geometry WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-4 Figure 2.1-2 Palisades r,z Reactor Geometry WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-5 2.2 - Calculated Neutron Exposure of Pressure Vessel Beltline Materials The plant- and fuel cycle-specific calculated fast neutron (E > 1.0 MeV) flux and fluence experienced by the materials comprising the beltline region of the Palisades pressure vessel is given in Tables 2.2-1 and 2.2-2, respectively, for plant operation through the conclusion of the twenty-first fuel cycle. Cycle 21 represents the last fuel cycle for which final fuel loading patterns have been designed. As presented, the data in Tables 2.2-1 and 2.2-2 represent the maximum neutron exposures at the pressure vessel clad base metal interface at azimuthal angles of 0°, 15°, 30°, 45°, 60°, 75°, and 90° relative to the core major axes. The limiting weld material for the Palisades pressure vessel occurs along the 60° azimuth (Heat W5214, Weld IDs 2-112A/C and 3-112A1C). All of the data provided in Tables 2.2-1 and 2.2-2 were taken at the axial location of the maximum exposure experienced at each azimuth based on the results of the three-dimensional synthesized neutron fluence evaluations. In Table 2.2-3, projections of neutron (E > 1.0 MeV) fluence beyond the end of Cycle 21 are provided. These projections were based on assumed future operating conditions provided by Entergy. In particular the following assumptions were applied to the analysis: 1 - For Cycle 22, the nominal calculated neutron flux based on the average of the prior uprated fuel cycles (18 through 21) was used. This approach is a realistic representation of the neutron flux that would be expected based on existing preliminary designs for Cycle 22. 2 - For Cycles 23 and beyond, the Cycle 21 neutron flux distribution was applied for all fuel cycles. This is a conservative assumption in that, considering Cycles 15 through 21, the Cycle 21 power distribution results in the highest calculated flux at the location of the critical pressure vessel weld (60°). 3 - Projected fuel cycle lengths were provided by Entergy as follows: Design 95% Capacity Cy~le 22 525 EFPD 499 EFPD Cycle 23 525 EFPD 499 EFPD Cycle 24 502 EFPD 477 EFPD Cycles 25+ 530 EFPD 504 EFPD Fuel cycles were assumed to operate with a breaker to breaker capacity factor of 95%. In completing the projections beyond the end of Cycle 21, operation was assumed to a total of 44 EFPY. Given the assumed operating scenario, this would cover a calendar time period extending to 2033. In regard to the f1uence data provided in Tables 2.2-1, 2.2-2, and 2.2-3, it should be noted that WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-6 the critical longitudinal welds (2-112A, 2-112C, 3-112A, and 3-112C) are exposed to the neutron flux characteristic of the 60° azimuthal location. The beltline circumferential weld 9-112 is exposed to the maximum neutron exposure characteristic of the 75° azimuthal location. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-7 Table 2.2-1 Summary of Calculated Maximum Pressure Vessel Neutron Flux (E > 1.0 MeV) For Cycles 15 Through 21 and for Future Projection Cycle Neutron (E > 1.0 MeV) Flux Fuel Time (n/cm2-s) Cycle (EFPY) o Deg. 15 Deg. 30 Deg. 45 Deg. 15 1.1 9.671E+09 1.558E+10 1.277E+10 7.924E+09 16 1.2 1.068E+10 1.604E+10 1.330E+10 7.797E+09 17 1.3 1.080E+10 1.860E+10 1.332E+10 7.613E+09 18 1.3 1.292E+10 2.094E+10 1.352E+10 7.337E+09 19 1.3 1.059E+10 1.924E+10 1.445E+10 7.037E+09 20 1.4 1.123E+10 2.004E+10 1.517E+10 8.143E+09 21 1.4 1.138E+10 2.016E+10 1.501E+10 8.506E+09 22 Proj. 1.153E+10 2.024E+10 1.454E+10 7.756E+09 23+ Proj. 1.138E+10 2.016E+10 1.501E+10 8.506E+09 Cycle Neutron (E > 1.0 MeV) Flux Fuel Time (n/cm 2-s) Cycle (EFPY) 60 Deg. 75 Deg. 90 Deg. 15 1.1 1.105E+10 1.681E+10 1.257E+10 16 1.2 1.135E+10 1.762E+10 1.401 E+1 0 17 1.3 9.781 E+09 1.967E+10 1.539E+10 18 1.3 1.088E+10 2.235E+10 1.664E+10 19 1.3 1.090E+10 2.230E+10 1.743E+10 20 1.4 1.161E+10 2.198E+10 1.650E+10 21 1.4 1.172E+10 2.151E+10 1.618E+10 0 22 Proj. 1.128E+10 2.204E+10 1.669E+10 23+ Proj. 1.172E+10 2.151E+10 1.618E+10 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-8 Table 2.2-2 Summary of Calculated Maximum Pressure Vessel Neutron Exposure Through the Conclusion of Cycle 21 Cycle Cumulative Neutron (E > 1.0 MeV) Fluence Fuel Time Time Cn/cm2) Cycle (EFPY) (EFPY) o Deg. 15 Deg. 30 Deg. 45 Deg. 1-14 14.4 14.4 1.132E+19 1.576E+19 1.192E+19 7.467E+18 15 1.1 15.5 1.165E+19 1.631 E+19 1.237E+19 7.742E+18 16 1.2 16.7 1.206E+19 1.693E+19 1.288E+19 8.041E+18 17 1.3 18.0 1.252E+19 1.773E+19 1.344E+19 8.366E+18 18 1.3 19.3 1.305E+19 1.858E+19 1.400E+19 8.665E+18 19 1.3 20.6 1.347E+19 1.935E+19 1.457E+19 8.944E+18 20 1.4 22.0 1.395E+19 2.023E+19 1.522E+19 9.296E+18 21 1.4 23.4 1.447E+19 2.114E+19 1.590E+19 9.677E+18 Cycle Cumulative Neutron (E > 1.0 MeV) Fluence Fuel Time Time (n/cm2) Cycle (EFPY) (EFPY) 60 Deg. 75 Deg. 90 Deg. 1-14 14.4 14.4 1.158E+19 1.576E+19 1.132E+19 15 1.1 15.5 1.196E+19 1.635E+19 1.175E+19 16 1.2 16.7 1.240E+19 1.702E+19 1.229E+19 17 1.3 18.0 1.282E+19 1.786E+19 1.295E+19 18 1.3 19.3 1.326E+19 1.877E+19 1.363E+19 19 1.3 20.6 1.369E+19 1.966E+19 1.432E+19 20 1.4 22.0 1.419E+19 2.060E+19 1.503E+19 21 1.4 23.4 1.472E+19 2.157E+19 1.575E+19 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 2-9 Table 2.2-3 Projections of Calculated Maximum Pressure Vessel Neutron Exposure End Neutron (E > 1.0 MeV) Fluence of Cycle Cumulative (n/cm2) Fuel Time Time Cycle (EFPY) (EFPY) o Deg. 15 Deg. 30 Deg. 45 Deg. 21 1.4 23.4 1.447E+19 2.114E+19 1.590E+19 9.677E+18 22 1.4 24.7 1.496E+19 2.201E+19 1.652E+19 1.001E+19 23 1.4 26.1 1.545E+19 2.288E+19 1.717E+19 1.038E+19 24 1.3 27.4 1.592E+19 2.372E+19 1.779E+19 1.073E+19 25 1.4 28.8 1.642E+19 2.461E+19 1.844E+19 1.110E+19 26 1.4 30.2 1.691 E+19 2.549E+19 1.909E+19 1.147E+19 27 1.4 31.5 1.741E+19 2.637E+19 1.975E+19 1.184E+19 28 1.4 32.9 1.790E+19 2.726E+19 2.040E+19 1.221E+19 29 1.4 34.3 1.840E+19 2.814E+19 2.105E+19 1.258E+19 30 1.4 35.7 1.889E+19 2.903E+19 2.170E+19 1.295E+19 31 1.4 37.1 1.939E+19 2.991E+19 2.236E+19 1.332E+19 32 1.4 38.4 1.988E+19 3.079E+19 2.301 E+19 1.369E+19 33 1.4 39.8 2.038E+19 3.168E+19 2.366E+19 1.406E+19 34 1.4 41.2 2.087E+19 3.256E+19 2.432E+19 1.443E+19 35 1.4 42.6 2.137E+19 3.344E+19 2.497E+19 1.480E+19 36 1.4 44.0 2.186E+19 3.433E+19 2.562E+19 1.517E+19 End Neutron (E > 1.0 MeV) Fluence of Cycle Cumulative (n/cm2) Fuel Time Time Cycle (EFPY) (EFPY) 60 Deg. 75 Deg. 90 Deg. 21 1.4 23.4 1.472E+19 2.157E+19 1.575E+19 22 1.4 24.7 1.520E+19 2.252E+19 1.647E+19 23 1.~ 26.1 1.571 E+19 2.345E+19 1.717E+19 24 1.3 27.4 1.619E+19 2.433E+19 1.784E+19 25 1.4 28.8 1.670E+19 2.527E+19 1.854E+19 26 1.4 30.2 1.721E+19 2.621E+19 1.925E+19 27 1.4 31.5 1.772E+19 2.714E+19 1.995E+19 28 1.4 32.9 1.823E+19 2.808E+19 2.065E+19 29 1.4 34.3 1.874E+19 2.902E+19 2.136E+19 30 1.4 35.7 1.925E+19 2.995E+19 2.206E+19 31 1.4 37.1 1.976E+19 3.089E+19 2.277E+19 32 1.4 38.4 2.027E+19 3.182E+19 2.347E+19 33 1.4 39.8 2.078E+19 3.276E+19 2.417E+19 34 1.4 41.2 2.129E+19 3.370E+19 2.488E+19 35 1.4 42.6 2.180E+19 3.463E+19 2.558E+19 36 1.4 44.0 2.231E+19 3.557E+19 2.628E+19 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 3-1 SECTION 3.0 NEUTRON DOSIMETRY EVALUATIONS During the first 14 operating fuel cycles at the Palisades plant, five sets of in-vessel surveillance capsule dosimetry and three sets of ex-vessel dosimetry were irradiated, withdrawn, and analyzed. The results of these dosimetry evaluations provide a measurement data base that can be used to demonstrate that the neutron fluence calculations completed for the Palisades reactor meet the uncertainty requirements described in Regulatory Guide 1.190P1 That is, the calculations and measurements should agree within 20% at the 1cr level. These calculation/measurement comparisons were previously completed and documented in Reference 3. However, for completeness, a brief description of the measurement program, dosimetry evaluation procedure, and final results are also included in this supplement to Reference 3. In addition to the Palisades dosimetry evaluations, this general methodology was also used in the determination of capsule exposures from the other PWR's included in Section 4.0 of this report. 3.1 - Method of Analysis Evaluations of neutron sensor sets contained in the in-vessel and ex-vessel dosimetry capsules withdrawn to date from the Palisades reactor were completed using current state-of-the art least-squares methodology that meet the requirements of Regulatory Guide 1.190[81. These least-squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculations resulting in a best estimate neutron energy spectrum with associated uncertainties. Best estimates for key exposure parameters such as ~(E > 1.0 MeV) and iron atom displacement rate (dpa/s) along with their uncertainties are then c easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to reactor dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, R; +/- 6 R, = ~)o-ig +/- 6(T)(¢g +/- 6¢) g relates a set of measured reaction rates, Rj, to a single neutron spectrum, ~g, through the multigroup dosimeter reaction cross section, crig, each with an uncertainty 8. The primary WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 3-2 objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement. For the least-squares evaluation of the Palisades dosimetry, the NRC approved methodology based on the use of the FERRET adjustment code[8] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best estimate values of exposure parameters along with associated uncertainties at the measurement locations. The application of the least-squares methodology requires the following input.

1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2. The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set.

3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the Palisades application, the calculated neutron spectrum at each measurement location was obtained from the results of plant-specific neutron transport calculations based on the methodology described in section 2.0 of this report. The calculated spectrum at each sensor set location was input to the adjustment procedure in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements. The sensor reaction rates were derived from the measured specific activities of each sensor set and the operating history of the respective fuel cycles. The dosimetry reaction cross sections were obtained from the SNLRML dosimetry cross-section library. [9] e In addition to the magnitude of the calculated neutron spectra, the measured sensor set reaction rates, and the dosimeter set reaction cross sections, the least-squares procedure requires uncertainty estimates for each of these input parameters. The following provides a summary of the uncertainties associated with the least-squares evaluation of the Palisades dosimetry. Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, the irradiation history corrections, and the corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM national consensus standards for reaction rate determinations for each sensor type. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 3-3 After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least-squares evaluation: Reaction Uncertainty 6O Cu 63 (n,a)C0 5% Ti 46 (n,p)Sc46 5% Fe54 (n,p)Mn 54 5% Ni 58 (n,p)C0 58 5% U238(n,f)Cs 137 10% Nb93(n,n')Nb93m 5% Np23\n,f)Cs 137 10% C0 59 (n,y)C06O 5% These uncertainties are given at the 1cr level. Dosimetry Cross-Section Uncertainties As noted above, the reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library. This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multigroup structure for use in least-squares adjustment applications. These cross sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the f1uence and energy characterization of 14 MeV neutron sources. Detailed discussions of the contents of the SNLRML library along with the evaluation process for each of the sensors is provided in Reference 9. For sensors included in the Palisades dosimetry sets, the following uncertainties in the fission spectrum-averaged cross sections are provided in the SNLRML documentation package: Reaction Uncertainty Cu 63 (n,a)C0 6O 4.08-4.16% Ti 46 (n,p)Sc46 4.50-4.87% Fe 54 (n,p)Mn 54 3.05-3.11% Ni 58 (n,p)C0 58 4.49-4.56% U238 (n,f)FP 0.54-0.64% Nb93(n,n')Nb93m 6.96-7.23% Np237(n,f)FP 10.32-10.97% C0 59 (n,y)C0 6O 0.79-3.59% WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 3-4 These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations. Calculated Neutron Spectrum Uncertainties While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks, and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship: where Rn specifies an overall fractional normalization uncertainty, and the fractional uncertainties Rg, and Rg specify additional random groupwise uncertainties that are correlated with a correlation matrix given by: where H = (g_g')2 2r2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (8 specifies the strength of the latter term). The value of 0 is 1.0 when g = g' and 0.0 otherwise. The set of parameters defining the input covariance matrix for the Palisades calculated spectra was as follows: Flux Normalization Uncertainty (Rn) 15% c Flux Group Uncertainties (R g, Rg,) (E > 0.0055 MeV) 15% (0.68 eV < E < 0.0055 MeV) 29% (E < 0.68 eV) 52% Short-Range Correlation (8) (E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 3-5 Flux Group Correlation Range (y) (E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 These uncertainty assignments are consistent with an industry consensus uncertainty of 15-20% (10') for the fast neutron portion of the spectrum and provide for a reasonable increase in the uncertainty for neutrons in the intermediate and thermal energy ranges. 3.2 - Dosimetry Evaluations In this section, comparisons of the measurement results from the Palisades surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels. In the first instance, calculations of individual sensor reaction rates are compared directly with the measured reaction rates derived from the counting data obtained from the radiochemical laboratories. In the second case, the calculated values of neutron exposure expressed in terms of (E > 1.0 MeV), (E > 0.1 MeV), and iron atom displacements (dpa) are compared with the results of the least squares adjustment procedure described in Section 3.1. It is shown that these two levels of comparison yield consistent and similar results which demonstrate that the transport calculations for Palisades reactor produce neutron exposure results that meet the requirements of Regulatory Guide 1.190.[2] In Table 3.2-1, measurement/calculation (M/C) ratios for each fast neutron sensor reaction from surveillance capsule and reactor cavity irradiations are listed. This comparison provides a direct comparison, on an absolute basis, of calculation and measurement prior to the application of the least squares adjustment procedure. In Table 3.2-2, comparisons of measured and adjusted neutron exposures are given in terms of adjusted/calculated ratios for the five surveillance capsule dosimetry sets withdrawn to date as well as for the three cycles of reactor cavity midplane dosime\ry sets irradiated during Cycles 8, 9, and 10/11. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 3-6 Table 3.2-1 Comparison of Measured and Calculated Threshold Foil Reaction Rates M/C Ratio Capsule 63Cu(n,a.) 46Ti(n,p) 54Fe(n,p) 58Ni(n,p) 238U(n,f) 237Np(n,f) A240 1.09 1.21 1.02 0.95 W290 1.15 1.11 0.99 1.00 0.98 W290-9 1.12 1.16 0.96 0.98 0.96 0.92 W110 1.17 1.17 1.02 1.01 SA60-1 1.13 1.19 1.05 1.07 1.15 840 Cavity Cycle 9 1.11 1.10 1.08 1.03 1.13 1.21 Cycle 10/11 1.15 1.11 1.10 1.08 1.32 1.11 740 Cavity Cycle 8 1.09 1.14 1.08 1.07 1.06 1.40 Cycle 9 1.03 1.07 1.01 1.01 0.93 1.13 Cycle 10/11 1.08 1.05 1.02 1.03 1.07 1.08 640 Cavity Cycle 8 1.09 1.15 1.08 1.06 1.04 1.32 Cycle 9 1.05 1.08 1.01 1.03 1.09 1.24 Cycle 10/11 1.07 1.10 1.05 1.03 1.10 1.12 540 Cavity Cycle 10/11 1.09 1.05 1.00 1.06 1.04 39 0 Cavity Cycle 8 1.08 1.21 1.14 1.11 1.06 1.32 Cycle 9 1.06 1.06 0.99 1.00 0.87 0.98 Cycle 10/11 1.03 1.12 1.05 1.05 1.06 1.06 240 Cavity Cycle 10/11 1.03 1.08 1.03 1.04 1.19 0.96 Average 1.09 1.12 1.04 1.03 1.07 1.14 % std dev 3.9 4.7 4.4 3.8 10.0 12.8 Reaction Average M/C % Standard Deviation 63Cu(n,a.) 1.09 3.9 46Ti(n,p) 1.12 4.7 54Fe(n,p) 1.04 4.4 58Ni(n,p) 1.03 3.8 238U(n,f) 1.07 10.0 237Np(n,f) 1.14 12.8 Linear Average 1.08 7.9 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 3-7 Table 3.2-2 Comparison of Adjusted and Calculated Exposure Parameters Adjusted/Calculated (AlC) Ratio Capsule 4>(E> 1.0 MeV) 4>(E > 0.1 MeV) dpa A240 0.983 0.972 0.988 W290 0.988 0.981 0.997 W290-9 0.955 0.937 0.966 W110 1.011 1.001 1.020 SA60-1 1.078 1.067 1.077 84° Cavity Cycle 9 1.091 1.083 1.084 Cycle 10/11 1.142 1.133 1.134 74° Cavity Cycle 8 1.108 1.120 1.116 Cycle 9 0.999 0.993 0.996 Cycle 10/11 1.044 1.058 1.055 64° Cavity Cycle 8 1.086 1.096 1.092 Cycle 9 1.055 1.033 1.038 Cycle 10/11 1.065 1.078 1.075 54° Cavity Cycle 10/11 1.026 1.039 1.036 39° Cavity Cycle 8 1.116 1.139 1.135 Cycle 9 0.949 0.956 0.957 Cycle 10/11 1.058 1.060 1.060 24° Cavity Cycle 10/11 1.062 1.050 1.053 Average 1.05 1.04 1.05 % std dev 5.3 5.8 5.1 WCAP-15353 - Supplement 1-NP, Revision 0 May 2010

Westinghouse Non-Proprietary Class 3 4-1 SECTION 4.0 SURVEILLANCE CAPSULE NEUTRON FLUENCE In support of embritllement evaluations for the Palisades reactor pressure vessel, a compilation of calculated neutron fluence (E > 1.0 MeV) values for a series of materials surveillance capsules that contain test samples that apply to the Palisades plant is provided in this section. The compilation, encompassing a total of 18 surveillance capsules irradiated at the Palisades, Indian Point Unit 2, H. B. Robinson Unit 2, and Indian Point Unit 3 reactors is provided in Table 4-1. For each surveillance capsule listed in Table 4-1, the reported fluence value was calculated using an NRC approved methodology that meets the requirements of Regulatory Guide 1.190[2]. Therefore, this tabulation represents a consistent set of fluence values for use in data correlations. Details of the analysis methodology as applied to each of the four host reactors are given in References 3, 12, 13, and 14. In providing the data listed in Table 4-1, no new fluence calculations were performed. The data were obtained either from Palisades specific documents[10, 11] or from public domain documents[3, 12, 13, 14] that have been submitted to the NRC and are available on the ADAMS document system. It should be noted that, relative to the Palisades data listed in Table 4-1, References 3, 10, and 11 did not explicitly report fluence (E > 1.0 MeV) values for the individual capsules. Rather, the irradiation environment was reported in terms of irradiation time and calculated neutron flux (E > 1.0 MeV) averaged over the irradiation period. The fluence values listed in Table 4-1 were computed as the product of the irradiation time and the average neutron flux reported in these documents. Relative to the data in Table 4-1 and the listed references, it should also be noted that, in addition to the Reg.

, Guide 1.190 derived fluence values for Indian Point Unit 2, Table 3 of Reference 12 also lists fluence values for H. B. Robinson Unit 2 and Indian Point Unit 3 that were extracted from older references. These older values have been updated and superseded by the fluence values documented in References 13 and 14, respectively. All of these updated fluence values reflect the application of a fluence methodology that meets the requirements of Reg. Guide 1.190.
WCAP-15353 - Supplement 1-NP, Revision 0 May 2010
Westinghouse Non-Proprietary Class 3 Palisades A240 Palisades W290 Palisades W100-1 Palisades SA60-1 Palisades SA240-1 Palisades W100-2 Indian Point 2 Indian Point 2 Y* Indian Point 2 Z Indian Point 2 V* H. B. Robinson S H. B. Robinson V* H. B. Robinson T* H. B. Robinson X* Indian Point 3 T* Indian Point 3 Y* Indian Point 3 Z* Indian Point 3 X* Notes: 1- Relative to 2- In addition 4-2 Table 4-1 Summary of Neutron Fluence (E > 1.0 MeV) Derived from the Application of Methodology Meeting the Requirements of Regulatory Guide 1.190 Surveillance Fluence Reactor Capsule (E> 1.0 Mev) Reference Designation [n/cm2] 4.0ge+19 WCAP-15353, RO (Ref. 3) 9.38e+18 WCAP-15353, RO (Ref. 3) 1.64e+19 WCAP-15353, RO (Ref. 3) 1.50e+19 WCAP-15353, RO (Ref. 3) 2.38e+19 CPAL-01-009 (Ref. 10) 2.0ge+19 CPAL-04-8 (Ref. 11) T 2.53e+18 WCAP-15629, R1 (Table 3) (Ref. 12) 4.55e+18 WCAP-15629, R1 (Table 3) (Ref: 12) 1.02e+19 WCAP-15629, R1 (Table 3) (Ref. 12) 4.92e+18 WCAP-15629, R1 (Table 3) (Ref. 12) 4.7ge+18 WCAP-15805, RO (Table 5-10) (Ref. 13) 5.30e+18 WCAP-15805, RO (Table 5-10) (Ref. 13) 3.87e+19 WCAP-15805, RO (Table 5-10) (Ref. 13) 4.4ge+19 WCAP-15805, RO (Table 5-10) (Ref. 13) 2.63e+18 WCAP-16251-NP, RO (Table 5-10) (Ref. 14) 6.92e+18 WCAP-16251-NP, RO (Table 5-10) (Ref. 14) 1.04e+19 WCAP-16251-NP, RO (Table 5-10) (Ref. 14) 8.74e+18 WCAP-16251-NP, RO (Table 5-10) (Ref. 14) the Palisades data, References 1, 10, and 11 did not explicitly report fluence values for ,the listed capsules. Rather, the irradiation environment was reported in terms of irradiation time and neutron flux averaged over the irradiation period. The fluence values listed in Table 4-1 were computed as the product of the irradiation time and the average neutron flux (E > 1.0 MeV) reported in those documents. to the Reg. Guide 1.190 derived fluence values for Indian Point Unit 2, Table 3 of Reference 12 also lists fluence values for H. B. Robinson and Indian Point Unit 3 that were taken from older references. These values have been updated and superseded by the fluence values documented in References 13 and 14 that are based on a methodology that meets the requirements of Reg. Guide 1.190.
Indicates Capsules in other plants that contain W5214 weld material.
WCAP-15353 - Supplement 1-NP, Revision 0 May 2010
Westinghouse Non-Proprietary Class 3 5-1 SECTION
5.0 REFERENCES
1. Code of Federal Regulations Title 10 Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix G, "Fracture Toughness Requirements" and Appendix H, "Reactor Vessel Materials Surveillance Requirements," January 1992.

2. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

3. WCAP-15353, Revision 0, "Palisades Reactor Pressure Vessel Neutron Fluence Evaluation," G. K. Roberts et aI., January 2000.

4. LTR-REA-00-630, "Transmittal of Responses to Requests for Additional Information on WCAP-15353 in Support of the Palisades Pressure Vessel Fluence Evaluation,"
G. K. Roberts, July 13, 2000.
5. CCC-650, "DOORS 3.2, One-, Two-, and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998. Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.

6. DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996. Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.

7. WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure c
Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004 ..
8. WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," S. L. Anderson, May 2006.

9. DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium," July 1994.
Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory. 1O. Westinghouse Project Letter CPAL-01-009, "Fluence Analysis of Palisades Surveillance Capsule SA-240-1 ," W. R. Rice, April 30, 2001. WCAP-15353 - Supplement 1-NP, Revision 0 May 2010
Westinghouse Non-Proprietary Class 3 5-2
11. Westinghouse Project Letter CPAL-04-8, "Fluence Analysis for Reactor Vessel Surveillance Capsule W100," S. P. Swigart, February 11 2004.

12. WCAP-15629, Revision 1, "Indian Point Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," T. J. Laubham, December 2001.

13. WCAP-15805, Revision 0, "Analysis of Capsule X from the Carolina Power & Light Company H. B. Robinson Unit 2 Reactor Vessel Radiation Surveillance Program,"
T. J. Laubham, et aI., March 2002.
14. WCAP-16251-NP, Revision 0, "Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, et ai, July 2004.

15. Palisades Calculation EA-DOR-09-01, "Reactor Pressure Vessel Fluence Calculations for Cycle 20 and an Estimate of Cycle 21," Thomas W. Allen, June 10, 2009.
WCAP-15353 - Supplement 1-NP, Revision 0 May 2010
ATTACHMENT 3 Preliminary PTS Screening Criteria Assessment Structural Integrity Associates (SIA) Report No. 0901132.401, "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," Revision 0, April 2010 178 pages follow
Report No. 0901132.401 Revision 0 Project No. 0901132 April 2010 Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis Prepared for: Entergy Nuclear Corp. Palisades Nuclear Power Plant Prepared by: Structural Integrity Associates, Inc. San Jose, California Prepared by: ~L~ Timothy . Gnesbach Date: 412012010
, \I \!t1A~
Prepared by: Date: 4/20/2010 Vikram Marthandam Reviewed by: tu'~--"" Date: 4/20/2010 Clark Oberembt Approved by: 4£-*jjArkL Date: 4/20/2010 Timothy J. Griesbach
REVISION CONTROL SHEET Document Number: 0901132.401
Title:
Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis Client: Entergy Nuclear Corp. SI Project Number: 0901132 Quality Program: IZI Nuclear D Commercial Section Pages Revision Date Comments 1.0 1 0 4/20/2010 Initial Issue 2.0 1-2 3.0 2-6 4.0 6 - 12 5.0 13 - 23 6.0 23 - 29 7.0 29 8.0 30 - 33 Appendix A A1-A2 AppendixB B1 - B7 Appendix C C1 - C22 AppendixD Dl - D19 Appendix E El - E27 AppendixF F1 - F24 Appendix G G1 - G23 AppendixH H1-H6 Appendix I 11 - I4
EXECUTIVE
SUMMARY
This evaluation was perfonned as part of a review of the Palisades Pressurized Thermal Shock (PTS) re-evaluation. A previous analysis perfonned for the Palisades vessel in 2000 determined that the PTS screening criteria limit of 270°F for weld heat No. W5214 would not be reached until January 2014. That evaluation was based on the fluence projections and weld material chemistry for weld heat No. W5214 available at that time; no credit was given for surveillance data to improve the RT PTS projection. In the fall of 2009 it became apparent to Entergy that new infonnation was available that could affect the RTNDT of the limiting Palisades vessel beltline material. The new data included revised fluence calculations and a total of eleven irradiated surveillance capsules that contain Charpy V -notch data for weld heat No. W5214. This report examines the updated fluence calculations perfonned by Westinghouse and all the available surveillance data relevant to the Palisades reactor pressure vessel weld heat No. W5214. Using the revised fluences and chemistry factors based on the refitted surveillance data for this weld heat, this re-evaluation shows that the projected date to reach the PTS screening criteria limit using the surveillance weld data would be approximately April 2017 or later. Report No. 0901132.401, Rev. 0 STrifU!f}IYfU 1"'t;If!Uf'i'hf ASSOICJares, Inc.
Table of Contents Section
1.0 INTRODUCTION
............................................................................................................... 1 2.0 APPLICABILITy ................................................................................................................ 1 3.0 METHODOLOGY .............................................................................................................. 2 4.0 DATA EVALUATION RESULTS ..................................................................................... 6 5.0 DATA CREDIBILITY ASSESSMENT ANDFLUENCEEVALUATION ................. 13 6.0 DISCUSSION ......................................................................................... 23 7.0
SUMMARY
AND CONCLUSIONS .............................................................. 29
8.0 REFERENCES
................................................................................................. :................ 30 APPENDIX A: CEOG DETERMINATION OF BEST-ESTIMATE CHEMISTRY FOR WELD HEAT NUMBER W5214 ........................................................................ A-l APPENDIX B: EXCERPT FROM GENERIC LETTER 92-01 AND RPV INTEGRITY ASSESSMENT NRC/INDUSTRY WORKSHOP ON RPV INTEGRITY ISSUES .............. B-l APPENDIX C: PALISADES SUPPLEMENTAL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 ..................................................... C-l APPENDIX D: INDIAN POINT 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 .............................. D-l APPENDIX E: INDIAN POINT 3 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 ................................ E-l APPENDIX F: H. B. ROBINSON 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 .............................. F-l APPENDIX G: CVGRAPH TANH CURVE-FITS FOR W5214 SURVEILLANCE WELD DATA ......................................................................................................... G-l APPENDIX H: CALCULATION OF TIME-WEIGHTED AVERAGE TEMPERATURES FOR SURVEILLANCE CAPSULES CONTAINING WELD HEAT NO. W5214 ............... H-l APPENDIX I: LISTING OF DESIGN INPUTS FOR WELD HEAT NO. W5214 SURVEILLANCE DATA RE-EVALUATION ....................................................... .1-1 Report No. 0901132.401, Rev. 0 11
List of Tables Table 1: Results of all W5214 Surveillance Data with Reported Fluence and Vendor Shift Results ................................................................................................ 10 Table 2: Summary of Revised Capsule Fluences and Time-Weighted Average Temperatures for Surveillance Capsules Containing Weld Heat No. W5214 .............................. 11 Table 3: Summary of Revised (Refitted) Surveillance Capsule Results for Weld Heat No. W5214 ............................................................................................. 12 Table 4: Test Specimens Contained in Palisades Capsules SA-60-1 and SA-240-1 ............... 15 Table 5: Evaluation of Palisades Surveillance Data Results for Weld Heat No. W5214 .......... 17 Table 6: Evaluation of all Surveillance Capsule Results Containing Weld Heat No. W5214 .............................................................................................. 18 Table 7: Scatter in Fit to all Surveillance Capsule Results Containing Weld Heat No. W5214 .............................................................................................. 19 Table 8: History of Time-Weighted Operating Temperature for Palisades .......................... 20 Table 9: Correlation Monitor Material HSST Plate 02 Calculation of Fitted CF ................... 21 Table 10: Correlation Monitor Material HSST Plate 02 Calculation of Measured - Predicted Scatter ............................................................................................... 22 Table 11: Calculc~ted and Projected Fluence Values at 60° Weld Location ......................... 23 Table 12: Limiting Fluence Determination for Current Licensing Basis (Case 1) ................. 25 Table 13: Limiting Fluence Determination for Revised Best-Estimate CF Value (Case 2) ...... 25 Table 14: Limiting Fluence Determination for Case 4a ................................................ 26 Table 15: Limiting Fluence Determination for Case 4b ............................................... .26 Table 16: Interpolation ofPTS Limit Date Based on Current Licensing Basis and Revised Fluence ............................................................................................ 27 Report No. 0901132.401, Rev. 0 III 'l:tY"u'tfYfF Integrity Assodates, Inc.
Table 17: Interpolation ofPTS Limit Date Based on Limiting Fluence for Case 4b ............. .28 Table 18: Projected Maximum Fluence and Estimated PTS Limit Dates for Palisades Weld W5214 ............................................................................................. 29 List of Figures Figure Figure 1. Best Fit to Data for all W5214 Surveillance Data with Reported Fluence and Vendor Shift Values .............................................................................. 34 Figure 2. Palisades Supplemental Surveillance Data (W5214) with Revised Fluence and Refitted Shift (Case 4a) ............................................................ : ............ 35 Figure 3. Best Fit for all W5214 Surveillance Data with Revised Fluence and Refitted Shift (Case 4b) ........................................................................................... 36 Figure 4. Plot of Residual vs. Fast Fluence for A533B-l HSST-01lHSST-02 CMM with Companion Materials, the Overa1l2-Sigma Scatter is 50°F [24] ......................... 37 Figure 5: Projected Peak Fluence at 60° Weld Location (from [18]) and RTPTs Limit Dates ... 38 Report No. 0901132.401, Rev. 0 IV Sti'ifuml.ral int&mriflJ Associates, Inc.
EVALUATION OF SURVEILLANCE DATA FOR WELD HEAT NO. W5214 FOR APPLICATION TO PALISADES PTS ANALYSIS
1.0 INTRODUCTION
The Palisades Nuclear Plant submitted to NRC a Pressurized Thermal Shock (PTS) evaluation in 2000 that projected the value for RTpTs, or maximum Adjusted Reference Temperature (ART) of the limiting vessel weld or plate, based on the calculated fluences and material properties available at that time [1]. The limiting vessel belt1ine material was determined to be weld heat No. W5214, and the projected ART value was based on the method in the PTS Rule given in 10CFR50.61, Paragraph (c)(l) [2] using the best estimate chemistry for this weld, the corresponding chemistry factor, CF, and the fluence values from the vessel fluence evaluation in WCAP-15353 [3]. These inputs to the PTS Rule equations were used to calculate RTPTs , and the Palisades vessel was projected to reach the screening criterion limit of 270°F for the limiting weld in January 2014. Since the time that the previous PTS evaluation was performed for the Palisades vessel, ten years have passed and more data and information are available now to update the projected RTNDT value for the limiting Palisades vessel beltline material. In particular, this evaluation considers all the available surveillance data for weld heat No. W5214 that can be used to refme the projected RTNDT in accordance with 10CFR50.61, Paragraph (c)(2) [2]. This evaluation is being updated now because there is new information that changes the projected values ofRTpTs for the Palisades vessel. The new information was generated by performing a survey of all relevant surveillance data for weld heat No. W5214. Eleven irradiated surveillance capsule reports were found containing this weld, and the Charpy data contained in these capsules was compiled and refitted consistently using the CVGRAPH hyperbolic tangent curve-fitting methodology [4]. Also, the eleven capsule fluence values have been updated (over time) by Westinghouse using their NRC approved fluence methodology for implementing the Regulatory Guide 1.190 benchmarking procedure [5]. AU data was evaluated in accordance with the PTS Rule approach to determine the shift values, fitted CF from the surveillance data, scatter from the mean predicted shift, and credibility of the data. It was determined that this new evaluation provides the most technically complete and sound assessment for the Palisades weld heat No. W5214 and allows for more accurate life projections of the limiting material in the Palisades vessel. The latest W5214 life projection provides improvement (i.e., more time) from the previous prediction to reach the PTS screening criterion limit. 2.0 APPLICABILITY This evaluation is applicable to the Palisades Reactor vessel PTS analysis relative to intermediate shell axial welds 2-112A1B/C fabricated from weld heat No. W5214 [1]. The results of this evaluation are used to revise the RTPTs projection to determine the maximum fluence, or Report No. 0901132.401, Rev. 0 1 '"OnE ,",w""""." Associates, Inc.
equivalent date, to reach the PTS screening criteron of 270°F for the limiting axial weld at the 60° azimuthal location. Evaluation of other beltline region materials w~s not performed at this time because they are not projected or expected to exceed the PTS screening criterion over the next several years. A complete PTS analysis for all of the vessel beltline materials will be performed at a later time per IOCFR50.61. 3.0 METHODOLOGY 3.1 Charpy TANH Curve-Fitting Method All Charpy data has been re-evaluated to assure that the Charpy curve fits and the 30 ft-lb shift values from the surveillance capsule reports are performed in a consistent manner. The general shape of Charpy test data (energy versus temperature, or lateral expansion versus temperature) is that of an "S", generally with definable lower and upper shelves and a connecting region between the shelves called the transition region. The hyperbolic tangent (TANH) function has been used for some time as a simple statistical curve-fit tool to describe this "S"-shaped response [25]. Other functional relationships could have been used to produce a similar shape (e.g., an error function), but the benefit of the TANH function is that the curve fit parameters defining the "S" shape have physical meaning relative to what is generally evaluated from the test results. As a result, the hyperbolic TANH curve fitting ofCharpy V-notch (CV) impact energy data has been a standard practice within the industry. The TANH model used for modeling Charpy V-notch curves is given by Equation (1) [25]: Cv = A + B tanh [(T - To) / C] (1)
where, Cv = Charpy V-notch impact energy T test temperature A = the'mean energy level between the upper and lower shelves B = the + or - deviation from the mean energy level To = a parameter that represents the mid-energy transition temperature C = the + or - deviation of the intercepts of the tangent to the transition of To and the upper and lower shelves The lower shelf (A - B) was fixed at 2.2 ft-lb, and the upper shelf (A + B) was fixed III accordance with standard practice for applying hyperbolic tangent fits to Charpy V -notch data
[24]. Report No. 0901132.401, Rev. 0 2 J'ifFjrlf:TIJr:u lnTIlnrtnl Associates, Inc.
3.2 Fluence Analysis Method The surveillance capsule fluence values were re-evaluated by Westinghouse using the DORT neutron transport calculation method which has been benchmarked to meet the criteria in NRC Regulatory Guide 1.190 [5], shows close agreement between calculations and neutron dosimetry measurements, and has been approved for use by NRC. The capsule fluence values were provided as design inputs by Westinghouse [18]. 3.3 10CFR50.61 (PTS Rule) Embrittlement Prediction Methods The PTS Rule in 10CFR50.61 [2] provides two methods for determining the reference temperature. The first method considers only the copper and nickel chemistry, fluence, and initial RTNDT of the weld, plate, or forging material in the reactor vessel beltline. For those beltline materials, Equation (2) is used to determine the adjusted RTNDT for comparison to the PTS screening criteria limits. RTNDT = Initial RTNDT + ~RTNDT + Margin (2) where ~RTNDT is the mean value of the transition temperature due to irradiation, and must be calculated using the Equation (3):
~RTNDT = (CF)f(o.28-0.101og t) (3) where CF (OF) is the chemistry factor, which is a function of the copper and nickel content. CF is determined by using Table 1 (from 10CFR50.61) for welds and by using Table 2 (from 10CFR50.61) for base metals (plates and forgings). "Wt % copper" and "Wt % nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging. For a weld, the best estimate values will normally be the mean of the measured values for a weld deposit made using the same weld wire heat number as the critical vessel weld. For weld heat number W5214, the best-estimate chemistry, as determined by the industry best-estimate results from the CEOG report [20], is Cu = 0.213 wt%, Ni = 1.007 wt%.
The best-estimate chemistry values for the C-E fabricated welds are shown in Appendix A. The corresponding chemistry factor for this weld heat is CF = 230.73°F. This value for best-estimate nickel content varies slightly from the value used previously in the PTS submittal (Ni = 1.01 %) [1]. It is noted that the basis for the best estimate Cu and Ni values came from the CEOG report [20] which is considered to be the industry standard for the C-E fabricated welds, but Palisades chose to round up the nickel content from 1.007% to 1.01 % in the 1998 RAI response to Generic Letter 92-01 [19]. For the current analysis we have also used the CEOG determined actual nickel best estimate chemistry which gives a CF value of230.73°F for comparison to a CF value of 231.08 of for the rounded up nickel content. Report No. 0901132.401, Rev. 0 3 :srl'l'!ffUII'JU InttH"JiI'ihl Associates, Inc.
The Initial RTNDT for weld heat No. WS214 is detennined from the generic value of -S6°F for C-E fabricated Linde 1092 flux type welds [31], and the margin tenn is detennined from Equation 4: Margin = 2 ~(J"2 + (J"2 (4) I ~ where aI is the standard deviation for the initial RTNDT. If the generic mean Initial RTNDT value for a Linde 1092 weld is used, then aI = 17°F [2]. The (I-sigma) standard deviation for i1RTNDT' at, , is 28°F for welds, so the margin tenn for this case is 6S.S0F. Using this approach to detennine the RTPTS at the screening criteria limit for axial welds (i.e., 270°F) yields a maximum allowable fluence at the 60° azimuthal weld location of: RTpTS = 270°F = Initial RTNDT + i1RTNDT + Margin = -S6 + L1RTNDT + 6S.S~F (S) i1RTNDT = 260.soF = (CF) f(0.28-0.10 logf) = (230.73) f(0.28-0.10 logf) (6) (7) The second method for detennining the chemistry factor and the RTNDT states that, "To verify that R T NDT for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific infonnation that could affect the level of embrittlement. This infonnation includes but is not limited to the reactor vessel operating temperature and any related surveillance program results." Surveillance program results means any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR Part SO, appendix H. This is the case for Palisades; eleven previously tested surveillance capsules are now available that contain the limiting vessel weld heat No. WS214. The axial weld in the Palisades vessel made from weld heat No. WS214 C was determined to be the limiting vessel beltline material [1]. Results from the plant-specific surveillance program must be integrated into the RTNDT estimate if the plant-specific surveillance data has been deemed credible as judged by the following criteria [2]: (A) The materials in the surveillance capsules must be those which are the controlling materials with regard to radiation embrittlement, (B) Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions must be small enough to pennit the detennination of the 30-foot-pound temperature unambiguously, (C) Where there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values must be less than 28°F for welds and 17°F for base metal. Even if the range Report No. 0901132.401, Rev. 0 4 Stflllctllfal,.. t1tll'i'f" Associates, Inc.
in the capsule fluences is large (two or more orders of magnitude), the scatter may not exceed twice those values (i.e., 56°F), (D) The irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the claddinglbase metal interface within 25°F, and (E) The surveillance data for the correlation monitor material in the capsule, if present, must fall within the scatter band of the data base for the material. Surveillance data deemed credible according to these criteria must be used to determine a material-specific value of CF for use in Equation (2). A material-specific value of CF is determined from Equation (8) [19].
'£ [AI x f:fJ.~ - IMI1Iot CF", .;. .1=..;;1_ _ _ _ _ __ (8) fk ~ l1 61 * - 0.20 fjlJ i"'1 where "n" is the number of surveillance data points, "At is the measured value of LlT 30 from the Charpy specimens, and "~" is the fluence for each surveillance capsule data point.
If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld (i.e. differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld), the measured values of LlT 30 must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld [2]. For cases in which the results from a credible plant-specific surveillance program are used, the value of (J ~ to be used in the margin term of Equation (4) is 14°F for welds; this is called a reduced margin tehn. The value of (J ~ need not exceed one-half of LlRT NDT [2]. The use of results from the plant-specific surveillance program method may result in a RTNDT that is higher or lower than that determined from the first (chemistry table) method. If the resulting RTNDT from using the surveillance data gives a higher value it must be used. If the resulting RTNDT from using the surveillance data gives a lower value it may be used. NRC provided additional guidance for evaluation and use of surveillance data in Reference 19. The guidance provides examples of Case 4 that may be used for evaluating the Palisades related surveillance data, as shown in Appendix B. Two cases are considered, Case 4a considers surveillance data from the plant of interest, and Case 4b for use of surveillance capsule data from both the plant of interest and also from its sister plants containing the same weld heat. Report No. 0901132.401, Rev. 0 5 StfllJctllfallnttlffYihfAssociates, Inc.
Also, if there is clear evidence that the copper or nickel content of the surveillance weld differs from that of the vessel weld, i.e., differs from the average of the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of i1RTNDT should be adjusted by multiplying them by the ratio of the chemistry factor of the vessel weld to that of the surveillance weld using the equation [19]: (9) According to the NRC guidance [19], further adjustment to the L1RTNDT data from other sources is needed if there is a difference between the capsule temperature from the other plant and the plant of interest. A temperature correction of 1°F;oF is made to the i1RTNDT values to account for this difference; a positive temperature adjustment is made to capsules exposed to (time-weighted average) temperatures below the mean vessel temperature, and a negative temperature adjustment is made to capsules exposed to (time-weighted average) temperatures above the mean vessel temperature. The mean vessel temperature for the Palisades vessel, using a time-weighted average for the plant operating cycles, is determined to be 535.2°F, as shown in Table 8. Guidance from Reference 19, Case 4, "Surveillance Data from Plant and Other Sources," and 10CFR50.61, method 2 for inclusion of plant-specific surveillance data, has been applied to evaluation of the W5214 surveillance data as described in Section 5.0. 4.0 DATA EVALUATION RESULTS In 1998, Consumers Energy provided a response to a Request for Information from NRC regarding pressure vessel integrity for the Palisades vessel [26]. That response evaluated seven surveillance capsules containing weld heat No. W5214 which were available at that time (two capsules from H. 13. Robinson 2, two capsules from Indian Point 2, and three capsules from Indian Point 3) and determined those data were not credible and, therefore, the data was not used to improve the projected RTpTs for the Palisades vessel. Since then, four more capsules containing this weld heat can be included in the analysis for use of surveillance data related to the Palisades limiting weld material. An evaluation of these data starts with the original capsule reports. The new data survey was performed to gather all the unirradiated and irradiated capsule test results for the Palisades limiting weld material. The data from all related surveillance capsules containing weld heat No. W5214 were compiled and the results were reviewed for applicability to the Palisades vessel weld. New data were discovered in the process of compiling these capsule reports. For example, there are two capsules from the Palisades supplemental Report No. 0901132.401, Rev. 0 6 Associates, Inc.
surveillance program that were previously unreported (References 14 - 17). In addition, three capsules from the H. B. Robinson 2 surveillance program (References 12), two capsules from the Indian Point 2 surveillance program (References 6 - 8), and four capsules from the Indian Point 3 surveillance program (References 9, 10, 11 and 13) were compiled and the Charpy V-notch test results were reviewed. These reports include surveillance capsule fluences and comparisons between the unirradiated and irradiated Charpy V -notch curves to determine the L1RTNDT (or L1 T 30) shifts. The reported fluence values from these capsule reports, the average of measured surveillance weld copper and nickel chemistries, and the (measured and reported) L1T30 shift results for these eleven capsules are shown in Table 1. The capsule reports are included in Appendix C (Palisades), Appendix D (Indian Point Unit 2), Appendix E (Indian Point Unit 3) and Appendix F (H. B. Robinson Unit 2). These are considered to be the reported data (or original data), with the exception of the Palisades capsule reports that were considered to be supplemental capsule test results. It is useful to fIrst combine these data without any adjustments for chemistry or irradiation temperature to determine the mean trend in irradiation damage behavior. The mean trend, or average chemistry factor, can be calculated directly from a least squares fIt to the data using Equation (8). The least squares fIt method was used and a best fIt chemistry factor (CF) of 217.67°F was determined from these data, as shown in Table 1. The results are plotted in Figure 1 and are shown here for information only. The scatter in the measured - predicted results show that the scatter exceeds the 28°F (I-sigma) margin for two out of eleven points, but these two points are within the 56°F (2-sigma) margin. The average copper content for these surveillance materials is Cu = 0.243 wt%, and the average nickel content for these surveillance materials is Ni = 0.965 wt%. The predicted (average) chemistry factor for the surveillance specimens based on chemistry (from the PTS Rule Table 1) is CF = 234.37°F. The mean fIt to the data shows that the CF value and the fItted trend for these data is well below that predicted by the PTS Rule method 1 (i.e., surveillance data not available). 4.1 Original ~nd Re-evaluated Surveillance Capsule Fluence Westinghouse recalculated the capsule fluences from Palisades, Indian Point 2, Indian Point 3, and H. B. Robinson 2 using a consistent methodology to establish a common basis for the fluence values. This was an essential step so that all the surveillance capsule data could be evaluated properly for credibility and applicability to the Palisades vessel limiting weld material. The revised fluence values for capsules containing weld heat No. W52I4 are shown in Table 2 [18]. It is noted that there were changes in the fluences from the originally calculated fluence values (shown in Table 1), and the new calculated fluence results (shown in Table 2) that were used to re-evaluate all the relevant surveillance data. The same Westinghouse fluence methodology was used to calculate fluence in the wall of the Palisades vessel for prediction of the vessel embrittlement. Report No. 0901132.401, Rev. 0 7 StfltlctllraI 1ntt:lf1i'ihlAssociates, Inc.
4.2 Surveillance Capsule Temperatures Surveillance capsule temperatures are necessary for the temperature corrections of the surveillance data when applying these data to the plant of interest. Time-weighted average temperatures were determined for the Palisades, Indian Point 2, Indian Point 3, and H. B. Robinson 2 capsules containing weld heat No. W5214. The data and method for determining the time-weighted average temperatures is given in Appendix H. The time-weighted average temperatures for the Indian Point Units 2 & 3 capsules were verified in Reference 33. 4.3 Original and Re-evaluated Charpy V-notch Surveillance Data The surveillance capsule test results for weld heat No. W5214 from the Palisades supplemental capsules SA-60-1 and SA-240-1 are provided in Appendix C. The supplemental capsules with this weld heat were irradiated for a number of cycles, and removed and tested; capsule SA-60-1 was removed at the end of cycle 13, and capsule SA-240-1 was removed at the end of cycle 14. The specimens containing weld metal inserts were reconstituted to full size Charpy V -notch specimens. The capsule materials were tested by Framatome in 2001 [14, 15]. The unirradiated Charpy energy values for the weld metals are documented in a letter from John R. Kneeland to Matthew J. DeVan dated February 2, 1999 [16]. The baseline (unirradiated) curve and weld metal chemistry data for these specimens are also provided in Appendix C. The surveillance capsule test results for weld heat No. W5214 from the Indian Point Unit 2 reactor surveillance capsule program are provided in Appendix D. The unirradiated data is contained in a Westinghouse report [8]. Southwest Research Institute tested two irradiated capsules, capsule Y [7] and capsule V [6]. The Indian Point 2 surveillance weld metal chemistry is also contained in these reports. There are four irradiated surveillance capsules containing weld heat No. W5214 and one baseline test report for the Indian Point Unit 3 plant, as shown in Appendix E. The baseline capsule report from Westinghouse contains the unirradiated data and one chemistry measurement for the surveillance weld [9]. The results for the four irradiated capsules are also given in one Westinghouse (WCAP-16251-NP) report [13]. This WCAP report contains the surveillance weld Charpy V-notch test results and measured chemistry data for the Indian Point Unit 3 plant. The surveillance capsule test results for weld heat No. W5214 from the H. B. Robinson Unit 2 reactor surveillance capsule program are provided in Appendix F. The unirradiated data is contained in a Westinghouse report [27]. There are three irradiated capsules from H. B. Robinson Unit 2 which contain weld heat No. W5214, capsule T, capsule V, and capsule X. The Charpy V-notch test results from these three capsules are contained in one Westinghouse (WCAP-15805) report [12]. This WCAP report also documents the measured surveillance weld chemistry for H. B. Robinson Unit 2. Report No. 0901132.401, Rev. 0 8 StrJ!JctllfSI1ntl,!nri:hlAssociates, Inc.
These original Charpy V-notch energy data were refitted using the CVGRAPH 5.0 hyperbolic tangent curve-fitting method [4]. The data were carefully fitted to obtain the best TANH fits. The CVGRAPH curve-fit results are shown in Appendix G from Reference 32. The results of the refitted and reanalyzed weld heat No. W5214 data for 30 ft-lb shift (~T30) are shown in Table
3. These refitted Charpy data results have been verified for use in the new credibility evaluation
[32]. The results presented here are considered to be "new data" because of the updated fluences and refitted ~T3o values and because several additional capsules containing weld heat No. W5214 were uncovered in this survey that had not been previously evaluated together with the other data. The results from these new data were evaluated for applicability to the prediction of the RTPTS value for weld heat No. W5214 per 10CFR50.61 [2] and the NRC guidance shown in Appendix B [19]. An evaluation of the credibility for the use of these data for the Palisades limiting weld is given in Section 5.0. Report No. 0901132.401, Rev. 0 9 Strl'lctllfa/1f!te;'f!!Pi:h;Associates, Inc.
Table 1. Results for all W5214 Surveillance Data with Reported Fluence and Vendor Reported Shift Results Reported Reported Predicted Measured - Capsule %Cu(a) %Ni(a) CF (F) Fluence (b) FF LlRTndt LlRTndt Predicted (n/cmI\2) (F) (F) LlRTndt (F) SA-60-1 0.307 1.045 266.5 1.61E+19 1.13 259 246.3 12.7 SA-240-1 0.307 1.045 266.5 2.60E+19 1.26 280.1 273.4 6.7 HB2 T 0.34 0.66 217.7 3.87E+19 1.35 288.15 293.6 -5.5 HB2 V 0.34 0.66 217.7 5.30E+18 0.82 209.32 179.1 30.3 HB2 X 0.34 0.66 217.7 4.49E+19 1.38 265.93 300.5 -34.6 IP2 V 0.20 1.03 226.3 5.59E+18 0.84 204 182.3 21.7 IP2 Y 0.20 1.03 226.3 5.89E+18 0.85 195 185.4 9.6 IP3 T 0.16 1.12 206.2 2.63E+18 0.64 151.6 138.6 13.0 IP3 Y 0.16 1.12 206.2 6.92E+18 0.90 172 195.2 -23.2 IP3 Z 0.16 1.12 206.2 1.04E+19 1.01 229.2 220.1 9.1 IP3 X 0.16 1.12 206.2 8.74E+18 0.96 193.2 209.4 -16.2 Average = 0.243 0.965 Table CF = 234.37°F Best fit CF = 217.67°F (a) Measured capsule weld materials Cu and Ni values obtained from [6, 12, 13, 14, 15,20,26] (b) Reported capsule fluence values from [6, 7, 12, 13,28] Report No. 0901132.401,Rev.0 10 .')tructurallnteoritll Associates, Inc.
Table 2. Summary of Revised Capsule Fluences and Time-Weighted Average Temperatures for Surveillance Capsules Containing Weld Heat No. W5214 Surveillance Time-Weighted Fluence (a) Reactor Capsule Average Temperature (E > 1 MeV) Designation (OF) [n/cm2] Palisades SA-60-1 535.0 [from Table 8] 1.50E19 Palisades SA-240-1 535.7 [from Table 8] 2.38E19 H. B. Robinson 2 T 547 [12] 3.87E19 H. B. Robinson 2 V 547 [12] 5.30E18 H. B. Robinson 2 X 547 [12] 4.49E19 Indian Point 2 V 524 [12] 4.92E18 Indian Point 2 Y 529.1 [from App. H) 4.55E18 Indian Point 3 T 539.4 [from App. H) 2.63E18 Indian Point 3 Y 539.5 [from App. H) 6.92E18 Indian Point 3 Z 538.9 [from App. H) 1.04E19 Indian Point 3 X 539.7 [from App. H) 8.74E18 (a) Revised capsule fluence values from Reference 18. Report No. 0901 132.401,Rev. 0 II ~~tfJ!Jt:tlmlJ UIIu!nnm Associates, Inc.
Table 3. Summary of Revised (Refitted) Surveillance Capsule Results for Weld Heat No. W5214 n Unirradiated Irradiated Revised Upper Shelf Capsule (Refitted) (Refitted) (Refitted) Energy T30 (F)(a) T30 (F)(a) LlT30(F) (ft-Ibs)(a) SA-60-1 -60.1 198.9 259 54.5 SA-240-1 -60.1 220 280.1 52.5 HB2 T -85.8 203.3 289.1 60.5 HB2 V -85.8 123 208.8 70.5 HB2 X -85.8 179.8 265.6 79.8 IP2 V -65.4 132.1 197.5 76 IP2 Y -65.4 128.5 193.9 66.5 IP3 T -63.8 86 149.8 90.5 IP3 Y -63.8 107.3 171.1 69 IP3 Z -63.8 164.5 228.3 76 IP3 X -63.8 128.7 192.5 75 (a) Charpy TANH curve-fit parameters, T30 values and plots are shown in Appendix G [32] Report No. 0901132.401,Rev. 0 12 .I::tmr.tllY:IIllntlmrifll Associates, Inc.
5.0 DATA CREDIBILITY ASSESSMENT AND FLUENCE EVALUATION The purpose of this evaluation is to apply the credibility requirements in 10CFR50.61 to the Palisades, H.B Robinson Unit 2, Indian Point Unit 2, and Indian Point Unit 3 surveillance capsule data and to determine if the surveillance capsule data is credible and can be used to improve the RTNDT predictions for the limiting vessel weld heat No. W5214. 10CFR50.61 describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of low-alloy steels currently used for light-water-cooled reactor vessels. 10CFR50.61 provides two methods for calculating the adjusted reference temperature of the reactor vessel beltline materials. The first method is described in paragraph (c)(1). The second method is described in paragraphs (c)(2) and (c)(3). The procedures in paragraphs (c)(2) and (c)(3) can only be applied when two or more credible surveillance data sets become available. These tests of surveillance data credibility are also stated in Section 3.3. NRC provided additional guidance for evaluation and use of surveillance data in Attachment 3 of Reference 19. The evaluation presented herein is organized like Case 4 from this guidance document, the case for plants with surveillance data for their plant and from other sources. 5.1 Credibility Evaluation: Criterion 1: The materials in the surveillance capsules must be those which are the controlling materials with regard to radiation embrittlement. The beltline region of the reactor vessel is defmed in Appendix G to 10 CFR 50, "Fracture Toughness Requirements" as follows:
"the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of The reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material and regard to radiation damage."
The Palisades reactor vessel consists of the following beltline region materials [1, 31]: e Intermediate Shell, Axial Welds 2-112 AlBIC, material heat No. W5214, e Lower Shell, Axial Welds 3-112 AlBIC, material heat No. W5214 and 34B009, e Intermediate to Lower Shell, Circumferential Weld 9-112, material heat No. 27204, e Intermediate Shell, Plate D-3803-1, material heat No. C-1279, e Intermediate Shell, Plate D-3803-2, material heat No. A-0313, e Intermediate Shell, Plate D-3803-3, material heat No. C-I279, Report No. 0901132.401, Rev. 0 13 l::#rJI11'fillfi4i IntG.'nri:til Associates, Inc.
19 Lower Shell, Plate D-3804-1, material heat No. C-1308A, 19 Lower Shell, Plate D-3804-2, material heat No. C-1308B, 19 Lower Shell, Plate D-3804-3, material heat No. B-5294. The Palisades reactor vessel was designed and fabricated in accordance with the ASME Boiler and Pressure Vessel Code, Section III, 1965 Edition, including all addenda through Winter 1965 [21]. The Palisades reactor vessel surveillance program was originally developed with the intent to comply, where possible, with the guidance of ASTM EI85-66, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors" [22]. At the time that the Palisades surveillance capsules were built, 10 CFR50 Appendices G and H did not exist. 5.1.1 Description of Original Palisades Surveillance Capsule Program ASTM E 185-66 [22] describes the requirements for test specimens. ASTM E 185-66 requires the base metal specimen be from " ... one heat with the highest initial ductile-brittle transition temperature", also known as the nil-ductility transition temperature (NDTT). Drop weight tests of Palisade's beltline samples identified five of the plates in contention for the highest initial NDTT at -30°F. The base material from shell plate D-3803-1 was selected over the other base metal specimens for the capsule base metal because it had the highest initial RTNDT temperature [31]. ASTM E 185-66 requires a sample to represent one vessel weld if a weld occurs in the irradiated region. The original Palisades surveillance weld specimens were fabricated with the same procedure used to fabricate the reactor vessel axial welds, and were fabricated with a similar filler wire and fluxes as the reactor vessel beltline welds. However, the original Palisades surveillance capsules did not contain limiting axial weld heat No. W5214. 5.1.2 Description of Supplemental Surveillance Capsules SA-60-1 and SA-240-1 At the end of Cycle 11, the Palisades surveillance capsule program was augmented to contain two supplemental ,surveillance capsules, designated as SA-60-1 and SA-240-1, installed in the capsule holders located on the core support barreL The new surveillance capsules, SA-60-1 and SA-240-1, included welds fabricated with weld wires of identical heats to those of the Palisades reactor vessel beltline welds. Surveillance capsule SA-60-1 and SA-240-1 contained test specimens from the following material heat No.'s: W5214, 34B009, 27204, and standard reference material HSST -02. All of these materials are the same heats as the materials used to fabricate portions of the reactor vessel that surround the active core and adjacent regions of the reactor vesseL Table 4 provides a tabulation of the specimens included in the Palisades supplemental surveillance capsules SA-60-1 and SA-240-1. Report No. 0901132.401, Rev. 0 14 Associates, Inc.
Table 4. Test Specimens Contained in Palisades Capsules SA-60-1 and SA-240-1 Material Description Tension Standard Charpy V- 18 mm Charpy Notch Impact V-Notch Inserts Weld Metal W5214 --- --- 42 (39)* Weld Metal 348009 --- --- 36 (39)* Weld Metal 27204 3 12 36 Correlation Monitor --- 12 --- Material, HSST Plate 02 (Heat No. A1195-1)
number of specimens in SA-60-1 capsule [15]
Capsules SA-60-1 and SA-240-1 were removed from the Palisades reactor vessel at the end of cycles 13 and 14, respectively. Twelve Charpy V-notch specimens made from weld heat No. WS214 were tested in capsule SA-60-1 [IS], and twelve Charpy specimens from heat No. WS214 were tested in capsule SA-240-1 [14]. Twelve Charpy V-notch specimens made from the HSST-02 correlation monitor material were tested from capsule SA-240-1 [14]. Because weld heat No. WS214 in the supplemental capsules matches the limiting axial welds, Criterion 1 is met for the Palisades reactor vessel. Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously. Criterion 2 is satisfied if the Charpy energy data for the surveillance capsules containing weld heat No. WS214 can be fitted to determine the 30 ft-lb temperature (T30) and upper shelf energy (USE) unambigudilsly. An accurate determination of the 30 ft-Ib shift (~T30) values is the reason these data were re-evaluated. The TANH curve fit method provides an accurate and reproducible determination of these values and can be used to establish the T 30 and USE values for a given Charpy data set [24]. Unirradiated and irradiated Charpy energy versus temperature data for the weld metal were fitted and plotted using the CVGRAPH hyperbolic tangent curve fitting program [4]. The Charpy energy fitted results for the eleven surveillance capsules, including the calculated 30 ft-lb temperatures and upper shelf energy values, are shown in the Appendix G and summarized in Table 3. Based on engineering judgment by looking at the fitting parameters and the plots, the scatter in the data is small enough, and the correlation coefficients are high enough, to permit the determination of the 30 ft-lb temperature and upper shelf energy of the surveillance weld materials unambiguously. Hence, Criterion 2 is met for all the surveillance capsules evaluated here which contain weld metal heat No. WS214. Report No. 0901132.401, Rev. 0 IS Stfltfctllrall!lIj.mrlhi Associates, Inc.
Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ~RTNDT values about a best-fit line drawn as described in Position 2 (surveillance data available) normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fails this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition in ASTM E185. The functional form of the least squares method as described in paragraph (c)(2) of 10CFR 50.61 will be utilized. A best-fit line is generated for this data to determine if the scatter of the ~RTNDT values about this line is less than 28°F for weld metal heat No. W5214. The Palisades limiting weld metal will be evaluated for credibility. This weld is made from weld heat No. W5214. This weld metal is also contained in the Indian Point Unit 2, Indian Point Unit 3, and H. B. Robinson Unit 2 surveillance programs. Since the welds in question utilized data from other surveillance programs, the recommended NRC methods for determining creditability will be followed. Of the recommended methods, Case 4 most closely represents the situation listed above for the Palisades surveillance weld metal. Case 4a Credibility Assessment - Palisades W5214 Data Only The data most representative for the Palisades limiting vessel weld are the supplemental surveillance capsules containing weld heat No. W5214 since the irradiation environment of the surveillance capsules and the reactor vessel are the same. The data requires the least adjustment. An adjustment can be made for the difference between the chemistry of the capsule specimens (CF = 266.5°F) and the best estimate chemistry of the vessel (CF = 230.73°F) using the ratio procedure. The updated fluence and ratio adjusted shift values were used to calculate a new least-squares fitted chemistry factor. The Palisades capsule data are shown in Table 5 along with the fitted solution (i.e., mean shift prediction) result, and the comparison of the measured-predicted scatter from the fitted CF of 198.8°F. A plot of the measured ~T3o vs. fluence results for the Palisades supplemental capsule weld (W5214) is shown in Figure 2 along with the +/- lcr bounds for credible data scatter. The data clearly fall within the I-sigma scatter band for credible surveillance data and the margin term can be reduced when using credible data. Based on criterion 3, the Palisades surveillance data is credible since the scatter is less than 28°F for both of these surveillance capsules. Report No. 0901132.401, Rev. 0 16
Table 5. Evaluation of Palisades Surveillance Data Results for Weld Heat No. W5214 Weld Heat No. W5214 Surveillance Data Palisades Revised ARTlldt!f) Predicted Measured - Capsule eu Ni TableCf fluence Refitted Adjusted to aJUndt Predicted Number (Wt"h) \'vvt%) (F) (n/em A 2) FF MlTndtlFJ Vessel CF (f) IF) aRTndt(F) SA-6{H 0.307 1.045 2665 1.501:+19 1.11 259.0 224.2 221.13 3.11 SA-240-1 (,-307 1.045 266.5 2.3SE+19 1,23 280.1 242.5 245.30 -2.30 Fitted CF = 198.8°F Case 4b Credibility Assessment - All W5214 Surveillance Capsule Data Following the guidance in Case 4 [19], the data from all sources should also be considered. For weld heat No. W5214 there are a total of eleven surveillance capsules from Palisades, Indian Point Unit 2, Indian Point Unit 3, and H.B. Robinson Unit 2. Since data are from multiple sources, the data must be adjusted first for chemical composition differences and then fbr irradiation temperature differences before determining the least-squares fit. For a credibility determination, the measured and refitted T30 shift data for all the relevant plant data was normalized to the mean chemistry factor of the vessel (230.73°F) using the ratio procedure and then to the mean operating temperature (535.2°F) for the Palisades vessel (see Table H-7). The fitted CF value, shown in Table 6, is determined to be 227.74°F for this case. The results for (measured - predicted) scatter for all the W5214 surveillance data results are shown in Table 7. The results for all the surveillance capsule data are plotted in Figure 3 along with the +/- 2cr scatter bands. The scatter in the measured - predicted values exceeds 28°F (1-sigma) for a few points. Four of the measured - predicted LlRTNDT values are outside the I-sigma hand of 28°F, but all data points are within the 56°F (2-sigma) scatter band for welds. According to 10CFR50.61 paragraph (c)(2)(iv), the use of results from the plant-specific surveillance program may result in an RTNDT that is higher or lower than that determined from the chemistry of the weld and a chemistry factor using the tables. If the CF value is higher, it must be used for ~essel RTPTS predictions, if the CF value is lower, it may be used. The chemistry factor from paragraph (c)(1) is 230.73°F, and the adjusted chemistry factor using the Palisades surveillance capsule data is 227.74°F. It is noted that per NRC guidance that it is possible to use a lower value of chemistry factor based upon all sources of surveillance capsule data with a full margin term (i.e., 56°F) if the data is credible in all other ways but the scatter. In summary, the (measured - predicted) scatter for all the W5214 weld data is within the acceptable range of 56°F for a wide range of fluence. For this case, the surveillance capsule fluence ranges between 2.63xlO 18 n/cm2 to 4.49xlO 19 n/cm2 . Therefore, the weld data meets this criterion, and the Palisades surveillance program weld metal chemistry factor to be used for determining RT pTs and RTNDT is 227.74°F in combination with a full (2-sigma) margin term. Report No. 0901132.401, Rev. 0 17 Stnl~fiirr:d intc,.nri:hf Associates, Inc.
Table 6. Evaluation of all Surveillance Capsule Results Containing Weld Heat No. W5214 Measured Ratio Chern. & Table Revised Fiuence IHad. (Refitted) Adjusted Temp. Adj. Capsule %Cll 1 %Ni 1 Cf (F) F!llence (n/cmA2) Factor fF Temp. Ti (F) ARTndt (F) ARTndt {F} ARTndt I (F) HA2 I ll.HOx fF 5.A-60-1 0,307 UY.+5 266,5 1,50E+l9 1.11 535.0 259 224.2 224.0 1~237 249.186 SA-240-1 0,307 1.045 266.5 2.38E+l9 1.23 535,7 280.1 242.5 243.0 1.522 299.830 HB2T 0:>;'
~~""+ 0)56 217.7 3.87E+19 1.35 547 28~tl 306A 318.2 1.820 429.2J53 HB2 V 1),,34 0.66 217,7 530E+18 0.82 547 208.8 2.2.1.3 233.1 0,677 191.749 HB2 ::< [1.34 OJj6* 217.7 4A9E+19 1,38 547 265,6 281.5 293.3 1.906 404343 !P2 V 0.20 :t03 226.3 it92E+18 0.80 524,0 1975 201A 190.2 0.643 15:L544 IP2 Y 0.20 1,03 226,3 4.55E+18 0.78 529.1 193.9 197~7 191J5 0.610 149.601 !P3T 0.16 1.12 206.2 2.63E+18 0.64 539.4 149.8 167.6 171.13 0.405 109 ADO IP3Y 0.16 1 "" ~..lLL 206.2 6,92E+18 0.90 539.5 171.1 191.5 195.8 0.804 175.543 IP3Z 0.16 1.12 2062 1,04E+19 1.01 538.9 228.3 255.5 259.2 1.022 262,003 IP3 )( 0.16 1.12 206.2 8~74E+18 0.96 539.7 192.5 215.4 219,9 0.926 211.596 SUM 11.573 263.5.058 Vessel Best Est~mate CF = 230.73* F c
Mean Vessel T =535.2"F I Least Squares Fitted CF = 227.74"'F Measured capsule v'J'e!d materials Cu and values obtained frorn [61 12, 13, 20,26J Fluencevalues obtalned from Reference 18 (c JTime-'Neighted average temperatures obtained frorn References 23 and 33 and Appendix H Refitted Charp'f V-notch shift data obtained from Reference 32 and .ADDendix G Report No. 0901132.401, Rev. 0 18 StriliCillnli Il'Ir§:mN.f!fAssociates, Inc.
Table 7. Scatter in Fit to all Surveillance Capsule Results Containing Weld Heat No. W5214 Irrad. Revised Fluence Adjusted Predicted Adjusted - Capsule Temp. Fluence Factor ilRTndt ilRTndt Predicted Ti (F) (n/cmI\2) FF (F) (F) (F) SA-60-1 535 1.50E+19 1.11 224.0 253.31 -29.27 SA-240-1 535.7 2.38E+19 1.23 243.0 281.00 -38.00 HB2 T 547 3.87E+19 1.35 318.2 307.23 10.97 HB2 V 547 5.30E+18 0.82 233.1 187.34 45.75 HB2 X 547 4.49E+19 1.38 293.3 314.44 -21.14 IP2 V 524 4.92E+18 0.80 190.2 182.69 7.48 IP2 Y 529.1 4.55E+18 0.78 191.6 177.83 13.77 IP3 T 539.4 2.63E+18 0.64 171.8 145.01 26.81 IP3 Y 539.5 6.92E+18 0.90 195.8 204.23 -8.48 IP3 Z 538.9 1.04E+19 1.01 259.2 230.24 28.92 IP3 X 539.7 8.74E+18 0.96 219.9 219.14 0.76 Note: four of the eleven (measured - predicted) data points exceed the 1 standard deviation of28°P for credible data for welds. All eleven (measured - predicted) data points fall within 2 standard
, deviations of 56°P for welds.
Report No. 0901 132.401,Rev. 0 19
Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the claddinglbase metal interface within +/- 25°F. The Palisades supplemental surveillance capsules SA-60-1 and SA-240-1 were located in the reactor vessel between the core barrel and the vessel wall opposite the center of the core. These supplemental surveillance capsules were installed in the capsule holders located on the core support barrel. Table 8 provides a history of the time-weighted temperature for the Palisades supplemental surveillance capsules and reactor vessel wall. Table 8. History of Time-Weighted Operating Temperature for Palisades Operating Cycle Cycle Average Surveillance Time Weighted Cycle length(a) Vessel Capsule Capsule Avg. T Number (EFPD) Temp.(b) Removed (OF) (OF) 1 371.7 523 2 440.1 529 3 342.5 534 4 321.0 536 5 386.7 536 6 326.7 536 7 362.5 536 8 366.1 537 9 292.5 534 10 349.7 534 11 421.9 533 12 399.3 534 13 419.6 536 SA-60-1 535.0 c 14 449.3 537 SA-240-1 535.7 15 401.3 537 16 444.3 537 17 493.1 537 18 472 537 19 459.2 537 Time Weighted 20 499.8 537 Vessel Avg. T 21 519.2 537 (OF) 22 498.8 537 535.2 (a) Cycle length (EFPD) values obtained from Reference 23 (b) Cycle average vessel temperatures obtained from Reference 28 Report No. 0901132.401, Rev. 0 20 Strjr/elllffllln/i:~nf'i:hf Associates, Inc.
The location of the specimens with respect to the reactor vessel beltline assured that the reactor vessel wall and the specimens have experienced equivalent operating conditions such that the temperatures did not differ by more than 2S0F. Therefore, this criterion is satisfied for the Palisades capsules. The Indian Point Unit 2 and Indian Point Unit 3 average surveillance capsule temperatures have been also reviewed and updated. The H. B. Robinson Unit 2 average capsule temperature was confirmed by the utility. The time-weighted average temperature values for these capsules are listed in Table 2, and the method for calculating these temperatures is given in Appendix H. Criterion S: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band for that materiaL The Palisades supplemental surveillance capsules, SA-60-1 and SA-240-1, both contain standard reference material HSST02 plate. Plots of the Charpy energy versus temperature for the irradiated condition of correlation monitoring material (HSST Plate 02, Heat AI19S-1) from SA-60-1 and SA-240-1 are documented in BAW-2341 Rev 2 [IS] and BAW-2398 [14], respectively. Charpy energy versus temperature for the unirradiated correlation monitoring material (HSST Plate 02, Heat A119S-1) is taken from NUREG/CR-64 13 , ORNLlTM-13133 [24]. Tables 9 and 10 provide the updated calculation of (measured - predicted) scatter versus fast fluence in the correlation monitor material (HSST 02) data. Figure 4 (from Reference 24) shows that the measured scatter band for the correlation monitor materials is SO°F. Table 9. Correlation Monitor Material HSST Plate 02 Calculation of Fitted CF (e) Capsule Fluence Fluence Factor ilRTNDT FF
ilRTNDT FF2 (x 1019 ) {al (FF) (b) (OF)
SA-60-1 1.5 1.112 113.7 126.4344 1.2365 SA-240-1 2.38 1.234 140.9 173.871 1.5223 Sum 300.305 2.7588 CF Surveillance weld = L: (FF x RT NDT) / L: (FF2)= 300.305/2.7588 = 108.853 Slope of best fit line is 108.853 Notes: (a) Calculated fluence (x 1019 n/cm 2, E>l.O MeV) (b) FF = fluence factor = f(O.28-0.1*logf) (c) Irradiated values of 30 ft-Ib Transition Temperature From BAW-2341 Rev 2 and BAW-2398 [15, 14] Report No. 0901132.401, Rev. 0 21 SlfjrlCllfffJlintegrity Associates, Inc.
Table 10. Correlation Monitor Material HSST Plate 02 Calculation of Measured - Predicted Scatter (e) Capsule Fluence Fluence Factor LiRTNDT Predicted (Measured - (x 1019 ) (al (FF) (b) LiRTNDT Predicted) LlRTNOT SA-60-1 1.5 1.112 113.7 121.044 -7.344 SA-240-1 2.38 1.234 140.9 134.324 6.575 Where predicted L'l.RTNDT = (slope bestfit)*(Fluence Factor) Slope of best fit line is 108.853 Notes: 19 2 (a) Calculated fluence (x 10 n/cm , E>1.0 MeV) (b) FF = fluence factor = f(o.28-0.1*logfl (c) Irradiated values of 30 ft-Ib Transition Temperature From BAW-2341 Rev 2 and BAW-2398 [15, 14] Table 10 shows that the scatter in these data is less than 50°F, which is the allowable scatter in NUREG/CR-6413, ORNLlTM-13133 [24]. Thus, criterion 5 is satisfied for the correlation monitor materials. 5.2 Palisades Vessel Fluence Evaluation Fluence in the Palisades vessel beltline has been tracked to manage the PTS issue. The fluence projections are important to be able to predict the future levels of embrittlement in the vessel beltline materials. Calculations of the neutron exposure of the Palisades reactor pressure vessel were previously completed and documented in WCAP-I5353, Revision 0 [3]. That evaluation, along with the benchmarking method, was submitted for review by the NRC Staff and the methodology and the final results were approved as part of the PTS evaluation in 2000 [1]. The previous evaluation determined that the peak fluence at the clad-to-base-metal interface at the 60° limiting axial weld was 1. 158xl0 19 n1cm2 (E > 1 MeV) at the end of Cycle 14 (i.e., October 1999) [34]. Since then, ten more years of plant operation has occurred and, as a result, the vessel has accumulated additional fluence. Recently, Westinghouse provided an updated fluence assessment for the Palisades vessel beltline region that includes cycle specific analysis for additional operating cycles for which the design has been finalized and operations are known (Cycles 15 through 21) and projections for future operation based on the best available knowledge as a function ofEFPY and estimated calendar dates [18]. The calculated and projected neutron fluence values for the limiting 60° weld location are given in Table 11. Note: the cycle specific projections for the designs of Cycles 21 and beyond were provided by Entergy and include an assumed load factor of95% for future plant operation [18]. Report No. 0901132.401, Rev. 0 22 l:jm:u:tEIfBllntegrity Associates, Inc.
Table 11. Calculated and Projected Fluence Values at 60° Weld Location [18] End of Estimated Cumulative Neutron Fuel Calendar Time Fluence @ 60° 2 Cycle Date (EFPV) n/cm (E > 1 MeV) 14 October 1999 14.4 1.158E+19 15 March 2001 15.5 1.196E+19 16 March 2003 16.7 1.240E+19 17 September 2004 18.0 1.282E+19 18 April 2006 19.3 1.326E+19 19 September 2007 20.6 1.369E+19 20 March 2009 22.0 1.419E+19 21 October 2010 23.4 1.472E+19 22 April 2012 24.7 1.520E+19 23 October 2013 26.1 1.571E+19 24 April 2015 27.4 1.619E+19 25 October 2016 28.8 1.670E+19 26 April 2018 30.2 1.721E+19 6.0 DISCUSSION The results for the surveillance capsules containing weld heat No. W5214 have been re-evaluated for applicability to the Palisades vessel. The refitted Charpy data results have been incorporated along with updates to the average capsule irradiation temperatures, corrections to account for chemistry differences, and results of the revised fluence calculations for the capsules and for the Palisades vessel have been included. Four cases are considered for prediction of the date to reach the PTS screening criteria limit. The first case is Position 1 of the PTS Rule using the current licensing basis method and considering the revised fluence calculations and projections, as shown in Table 11. The chemistry factor for the weld heat No. W5214 was based on best-estimate Cu = .213%, Ni = 2 1.01 %, and a CF = 231.08°F. The maximum fluence limit was calculated to be 1.584xl019 nlcm according to the embrittlement prediction method when surveillance data is not available. That prediction ofRTNDT shift is shown in Table 12, and the corresponding date to reach the PTS screening criteria limit of 270°F for axial welds is March 2014 using the fluence interpolation given in Table 16. The current licensing basis date to reach the PTS screening criteria date is Report No. 0901132.401, Rev. 0 23 StrJ'mtlm'lllntlt.'firi:fll Associates, Inc.
January 2014 as given in Reference 1. It has been determined that the Palisades plant is still operating within that licensing basis. The second case follows Position 1 of the PTS Rule but considers that the actual best-estimate chemistry for weld heat No. W5214 has slightly lower nickel as determined by the CEOG report in 1998 [20]. This evaluation was performed after the initial PTS submittal in 1995 and, using this information, the revised Ni = 1.007% which gives a new value ofCP = 230.73°P. It is permitted under the PTS Rule in 10CPR50.61 to use the best-estimate values for Cu and Ni, however there is only a slight difference in the maximum fluence to reach the PTS screening criteria limit (i.e., 1.595xl0 19 nlcm2) as shown in Table 13. The projected date to reach that limit using the interpolated fluence projections is July 2014 as shown in Table 16. Although this case does not show much additional margin from the January 2014 date, it is provided here to show that there is still slightly more time to be gained within the Position 1 approach of the PTS Rule method before the vessel reaches the 270 0 P screening criteria limit. The third case considered plant-specific surveillance data from the Palisades supplemental capsules containing weld heat No. W5214. This case is labeled as Case 4a per the guidance document for use of surveillance data [19]. The Case 4a credibility assessment calculated a fitted chemistry factor of 198.8°P from the two Palisades capsule data points., The data were deemed to be credible based on meeting all the credibility criteria including scatter within the I-sigma (i.e., 28°P) scatter bounds. The limiting fluence for the vessel for this case is shown in Table 14. Using these results and a reduced margin term to account for credible data, the projected date to reach the PTS screening criteria limit would be beyond 2034 for the limiting vessel weld heat No. W5214, as shown in Table 18. Note: It is likely that some other beltline material would become limiting if this case was used for weld heat No. W5214. However, this case demonstrates that surveillance data can provide significant improvement in determining the effects of embrittlement on the limiting vessel beltline weld material. The fourth case, pe:l:1nitted under the PTS Rule, is to use all sources of surveillance data that match the limiting weld heat No. W5214. This is designated as Case 4b, and the credibility assessment determined the fitted CP = 227.74°P. The data meet credibility criteria 1,2,4, & 5, and the scatter of the (measured - predicted) data was within 2-sigma (i.e., 56°P) such that it can be considered to be credible data for the chemistry factor, however the margin term, (J11, cannot be reduced in half. Use of Case 4b for the Palisades vessel is acceptable because the (measured - predicted) scatter in the weld data is within the acceptable range of 56°P for a wide range of fluence. The limiting fluence for Case 4b is 1.685xl0 19 nlcm2 (E > 1 MeV) as shown in Table 15. Table 17 interpolates the vessel fluence and shows the projected date to reach the screening criteria limit is April 2017, a difference of three years compared to the first case using the current licensing basis and Position 1 approach. A summary of the four cases considered in this analysis is given in Table 18. Report No. 0901132.401, Rev. 0 24 XIif,rl!!UlrJJilintegrity Associates, Inc.
Table 12. Limiting Fluence Determination for Current Licensing Basis (Case 1) 2 FLUENCE= i.58-4E+1@ n/cm f= 1.584 f FACTOR= fA(0.28-0.1 *@LOG(f))
= 1.1270 CHEM FACTOR = 231.08 of ARTNDT= 260.5 OF RTNDTO= -56.0 OF MARGIN= 65.5 OF TOTAL RTNDT = 270.0 of Table 13.
Limiting Fluence Determination for Revised Best-Estimate CF Value (Case 2) 2 FLUENCE= 1 .5~5E+19 n/cm f= 1.595 f FACTOR= fA(0.28-0.1*@LOG(f))
= 1.1289 CHEM FACTOR = 230.73 of ARTNDT= 260.5 of RTNDTO= -56.0 of MARGIN= 65.5 OF TOTAL RTNDT = 270.0 of Report No. 0901132.401, Rev. 0 25 tr $j05'Mr:KII!1{f'&lQinteorlfy Associates, Inc.
Table 14. Limiting Fluence Determination for Case 4a FLUENCE= 5438[=+19 n/cm 2 f= 5.438 f FACTOR= f"(0.28-0.1*@LOG(f))
= 1.4185 CHEM FACTOR = 198.8 of ~RTNDT = 282.0 of RTNDTO= -56.0 of MARGIN= 44.0* OF TOTAL RTN DT = 270.0 of
reduced margin term based on credible surveillance data Table 15.
Limiting Fluence Determination for Case 4b FLUENCE= 1.6'8-5"E+19 n/crn 2 f= 1.685 fFACTOR= f"(0.28-0.1 *@LOG(f)) 1.1437 CHEM FACTOR = 227.74 of
~RTNDT = 260.5 of RTNDTO= -56.0 OF MARGIN= 65.5 of TOTAL RTNDT = 270.0 of Report No. 0901132.401, Rev. 0 26
Table 16. Interpolation ofPTS Limit Date Based on Current Licensing Basis and Revised Fluence Neutron Date Fluence @ 60 0 2 n/cm (E > 1 MeV) November 2013 1.571E+19 December 2013 1.574E+19 January 2014 1.577E+19 February 2014 1.579E+19 Martn 20~4 1.582E+H~
i April 2014 1.585E+19 May 2014 1.588E+19 June 2014 1.591E+19 July 2014 1.594EH9'"*
August 2014 1.596E+19 September 2014 1.599E+19 October 2014 1.602E+19 November 2014 1.605E+19 December 2014 1.608E+19 January 2015 1.611E+19 February 2015 1.613E+19 March 2015 1.616E+19 April 2015 1.619E+19
Maximum fluence limit = 1.584xlO 19 nlcm2 for current licensing basis material case, CF = 231.08°F
**Maximum fluence limit = 1.595xl0 19 nlcm2 for revised best-estimate weld, CF = 230.73°F Report No. 0901132.401 , Rev. 0 27
Table 17. Interpolation of PTS Limit Date Based on Limiting Fluence for Case 4b Neutron Date Fluence @ 60° 2 n/cm (E > 1 MeV) November 2016 1.670E+19 December 2016 1.673E+19 January 2017 1.676E+19 February 2017 1.679E+19 March 2017 1.682E+19 April 2,017 L685~+Hi* i May 2017 1.688E+19 June 2017 1.691E+19 July 2017 1.694E+19 August 2017 1.697E+19 September 2017 1.700E+19 October 2017 1.703E+19 November 2017 1.706E+19 December 2017 1.709E+19 January 2018 1.712E+19 February 2018 1.715E+19 March 2018 1.718E+19 April 2018 1.721E+19
Maximum fluence limit = 1.685xl0 19 nlcm2 for Case 4b using revised fluence and W5214 surveillance data CP = 227.74°P and full (2-sigma) margin term Report No. 0901132.401, Rev. 0 28
Table 18. Projected Maximum Fluence and Estimated PTS Limit Dates for Palisades Weld W5214 Case No. CF IRTNDT Fluence FF DRTNDT Margin RTPTS Est. PTS Date (OF) (OF) 19 2 (OF) (OF) (OF) (10 n/cm ) (1) Current LB w/revised 231.08 -56 1.584 1.1270 260.5 65.5 270 March 2014 fluence (2) Current LB w/revised 230.73 -56 1.595 1.1289 260.5 65.5 270 July 2014 fluence and revised CF value 4a 198.80 -56 5.438 1.4185 282 44 270 >2034* 4b 227.74 -56 1.685 1.1437 260.5 65.5 270 April 2017
Other beltllne matenals will hkelybecome more limiting and will affect this date Case 1- Current licensing basis CF value for W5214 weld and revised fluence calculation Case 2 - CEOG best estimate chemistry and CF value for W5214 weld and revised fluence calculation Case 4a - Use of credible Palisades W5214 surveillance data and revised fluence with reduced margin term Case 4b - Use of all W5214 su[veillance data and revised fluence with full margin term 7.0
SUMMARY
AND CONCLUSIONS The results for all available surveillance capsules containing weld heat No. W5214 have been evaluated for applicability to the Palisades limiting vessel weld. Updates to the surveillance capsule fluences and the projected fluence in the Palisades vessel were also reviewed and included in these analyses. The methods of 10CFR50.61 were applied including options for considering the effects of surveillance data on the projected RTNDT values. Using Position 1 of the PTS Rule (without the use of surveillance data) shows a projected date to reach the PTS screening criteria limit as late as July 2014. However, use of the weld heat No. W214 surveillance data can improve the projections of embrittlement and significantly changes the date to reach the screening criteria limit. Since weld heat no. W5214 is currently identified as the limiting material, the projections using Case 4a with the credible Palisades supplemental surveillance data show that the PTS screening criteria limit of 270°F would not be reached until after 2034; however, other vessel beltline materials would become limiting and that would change that date. For Case 4b, the surveillance data for weld heat No. W5214 were shown to be credible for determination of the CF value, but the scatter in the data would not permit a reduction in the margin term. However, use of the *fitted chemistry factor for Case 4b with the revised fluence projections and the full margin term provides a better determination of the vessel embrittlement prediction for the limiting vessel weld. Using all the available weld heat No . W5214 surveillance data, a CF value of 227.74°F was determined for Case 4b and a projected date to reach the screening criteria limit of approximately April 2017 was estimated using the updated fluence projections from Westinghouse. Report No. 0901132.401 , Rev. 0 29
8.0 REFERENCES
1. Letter from Dar! S. Hood (USNRC) to Nathan Haskall (Palisades), "Palisades Plant-Reactor Vessel Neutron Fluence Evaluation and Revised Schedule for Reaching Pressurized Thermal Shock Screening Criteria (TAC No. MA8250)," November 14,2000. (SI File No.
0901025.206).
2. Code of Federal Regulations, Title 10, Part 50, Section 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U. S. Nuclear Regulatory Commission. (SI File No. 0901025.201).

3. Westinghouse Report, "Palisades Reactor Pressure Vessel Neutron Fluence Evaluation,"
WCAP-15353, Rev. 0, January, 2000. (SI File No. 0901025.203).
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5. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, March 2001.

6. SWRI Report, "Reactor Vessel Material Surveillance Program for Indian Point Unit No.2 Analysis of Capsule V," SWRI Project No. 17-2108, October 1988. (SI File No.
0901132.201).
7. SWRI Report, "Reactor Vessel Material Surveillance Program for Indian Point Unit No.2 Analysis of Capsule Y," SWRI Project No. 02-5212, November 1980. (SI File No.
0901132.202).
8. Westinghouse Report, "Consolidated Edison Co. Indian Point Unit No.2 Reactor Vessel Radiation Surveillance Program," WCAP-7323, May 1969. (SI File No. 0901132.203).

9. Westinghouse Report, "Consolidated Edison Co. of New York Indian Point Unit No.3 Reactor Vessel Radiation Surveillance Program," WCAP-8475, January 1975. (SI File No.
0901132.204).
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0901132.205). Report No. 0901132.401, Rev. 0 30 SffJf}ctllfaJ Integrity Associates, Inc.
11. Westinghouse Report, "Analysis of Capsule Y from the Power Authority of the State of New York Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," WCAP-I0300, March 1983. (SI File No. 0901132.206).

12. Westinghouse Report, "Analysis of Capsule X from the Carolina Power & Light Company H. B. Robinson Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-15805, March 2002. (SI File No. 0901132.207).

13. Westinghouse Report, "Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," WCAP-16251-NP, Revision 0, July 2004. (SI File No. 0901132.208).

14. Framatome ANP Report, "Test Results of Capsule SA-240-1 Consumers Energy Palisades Nuclear Plant - Reactor Vessel Material Surveillance Program," BAW-2398, May 2001. (SI File No. 0901132.209).

15. Framatome ANP Report, "Test Results of Capsule SA-60-1 Consumers Energy Palisades Nuclear Plant - Reactor Vessel Material Surveillance Program," BAW -2341, Revision 2, May 2001. (SI File No. 0901132.210).

16. Letter from J. Kneeland to M. 1. DeVan dated February 2, 1999, Enclosure. (SI File No.
0901132.211).
17. Westinghouse Report, "Neutron Fluence Analysis for Palisades Surveillance Capsule SA-240-1," CPAL-OI-009, April 30, 2001. (SI File No. 0901132.212).

18. Email Transmittal from Stanwood L. Anderson (Westinghouse) to Timothy Griesbach,
Subject:
Revised Fluence Values for Design Inputs to PTS Evaluation, April 15, 2010. (SI File No. 0901132.223)
19. "Generic Letter 92-01 and RPV Integrity Assessment," NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. (SI File No. 0901132.213).

20. "Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content," Combustion Engineering Owners Group, CEOG Task 1054, CE NPSD-1119, Rev. 01, July 1998. (SI File No. 0901025.204)

21. ASME Boiler and Pressure Vessel Code, Section III, 1965 Edition, including all addenda through Winter 1965, American Society of Mechanical Engineers.
Report No. 0901132.401, Rev. 0 31 :sm!U'mU'81 II.. ,,,,,,,,*.,,, Associates, Inc.
22. ASTM E185-66, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors."

23. Design Input Record from Thomas Allen (Entergy) to Timothy Griesbach (SIA) for basis/reference for adjusting the Palisades Cycle 1 through 12's cycle length expressed as Effective Full Power Day (EFPD) & unadjusted cycle lengths and operating dates, Attachments LAR of2-21-2000, and EA-DOR-09-01, Rev. 0, April 15, 2010. (SI File No.
0901132.224)
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25. EricksonKirk, M. A., EricksonKirk, M. T., Rosinski, S., Spanner, J., "A Comparison of the tanh and Exponential Fitting Methods for Charpy V-Notch Energy Data," Journal of Pressure Vessel Technology, Volume 131, June 2009. (SI File No. 0901132.225)

26. Haskell (Consumers Energy) to NRC, "Docket 50-255 -License DPR-20 -Palisades Plant Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity (TAC No. MA0560)," September 8, 1998. (SI File No. 0901132.217)

27. Westinghouse Report, "Carolina Power & Light Co. H. B. Robinson Unit No.2 Reactor Vessel Radiation Surveillance Program," WCAP-7373, January 1970. (SI File No. 0901132.218)

28. "Evaluation of Palisades Nuclear Plant Reactor Pressure Vessel Through the Period of Extended Operation," Constellation Nuclear Services Report, CNS-04-02-01, Rev. 1, June 2004. (SI File No. 0901132.219)

29. "Estimation ofEDYS for IP2 Reactor Vessel Head by 2R17 and 2R18," Entergy Nuclear
(' Calculation Number FCX-00538, 7120/05. (SI File No. 0901132.220)
30. Entergy Nuclear Calculation Number IP3-CALC-RV-03720, Rev. 2, Page 7 of 8. (SI File No. 0901132.221)

31. RVID2, NRC Reactor Vessel Integrity Database, U. S. Nuclear Regulatory Commission, Version 2.0.1,2000.

32. "Determination of30 ft-lb Shift (i1T30) Values for the Palisades Reactor Vessel Heat No.
W5214," Structural Integrity Associates Calculation No. 0901132.301, Rev. 0,4/15110. (SI File No. 0901132.301) Report No. 0901132.401, Rev. 0 32 SlririclilrtiJ lnt~'nythl Associates, Inc.
33. "Verification of the Time Weighted Average Temperatures for Indian Point Units 2 and 3 Vessel Weld Surveillance Capsules," Structural Integrity Associates Calculation No.
0901132.302, Rev. 0,4/15/10. (SI File No. 0901132.302)
34. Letter from Nathan Haskall (Palisades) to U. S. Nuclear Regulatory Commission, "Reply to Request for Additional Information Regarding Reactor Pressure Vessel Neutron Fluence Evaluation (TAC No. MA8250)," July 6,2000. (SI File No. 0901132.226).
Report No. 0901132.401, Rev. 0 33 :ilfjrltmlral'nt~ln¥it!l Associates, Inc.
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_____.. _ .,.-_ _ ,__ ."...._ _ .__._________._ ______ .J Figure 5. Projected Peak Fluence at 60° Weld Location (from [18]) and Revised RTPTs Limit Dates Report No. 0901132.401, Rev, 0 38 e Slt1.l.fJlUralln.IHfjfl/J' ASS~Cjates, Inc_
APPENDIX A CEOG DETERMINATION OF BEST-ESTIMATE CHEMISTRY FOR WELD HEAT NUMBER W5214 (from CE NPSD-1119, Rev. 1 [20]) Report No. 0901132.401, Rev. 0 A-I 5lftJ1CllJI'B1 m:r"I',:tlJ A,SSOi'Jlafl1S Inc.
Table 5 (CI)f:[OOI!td) Estl~te C;ll'tlTll~!1' and Nickel 90130 0.133 00136 0269 0'.010 90144 0.044 OJrl5 90146 0.082 90209 0,126 9565 0.052 A8746 0.13 BOLA BODA U, Ni200 BE
- Se~ Report CE NPSD*I0:39. Rev. 02 for method of lk:terminarkH},
It"" In
/'1; J:. <1 J'~
Report No. 0901132.401, Rev. 0 A-2
APPENDIXB EXCERPT FROM GENERIC LETTER 92-01 AND RPV INTEGRITY ASSESSMENT NRC/INDUSTRY WORKSHOP ON RPV INTEGRITY ISSUES Report No. 0901132.401, Rev. 0 B-1
RECOMMENDATIONS IN 1 P~r10 CFR 50.61 (e)(2):
" for each vessel beltlin, meteria! i.$ Si bounding reflictor vessel. licensees !ineU eorl$id~r ;pIa"t~
S~01flC tnformatlon could affl!j;(;t the lewt of ernbrU11emmt information includes but iii not iimlted to the reector vessel op.rllting tempet'iilture and anyrelst.d ,urvell'.nt;e pl'OgrllmfJ :result*.# Per Footnote 5: s, Surveillance data thetdemon..,trate; the embriWement for limitinG material, inoludlng but Umited to deta fromtfltt reat;lor; or from a1.ilfVe~!lance programs other plants 0,( without $tlrv:eit"ll;n~e program integrated ~ 10 CFR SO , Appandb: PerRG1.99 Rev, 2: POlitioo and 10 erR SO.61 (e)(2) (iiXB):
'the:r$' is, tlesr evidence th.tthfit copper or nickel: content of the $U('I.fe.11IMC.e weld differ. from thstol the vane! weld, I.e" differs from the av&rage for the weid wiFe heat number sS$OOl;ded with 1ha ve.lsill wekJ and the sUl'\I'$iUanc:e we~dj the measul'fld v.lues of l1RTNDT shoufdba .JdjU$tf)d by multiplyIng them by the ratio of the chf)mI,fty mctor tor ;tn. vessel weld to that for the survellJam;e Report No. 0901132.401, Rev. 0 B-2 Slru'ctlJl'BJ In:'~nt'j"1f f\;SSOl;larf1S, Inc.
Irradiation temperature aMi f!uence are flrst order envtronmental variables in as.eslliing lnadiation damage contributora

StucU..fJ hav**hown, fortemper.aturee near 550 "'F. a 1 deci"ij;aae rn irradiation temperature wilt ",.un In appro.xlm.l.telya 1 "F inceaae In ARTNDT Report No. 0901132.401, Rev. 0 B-3 Smtlt:tllf8I lnT* mn ;rtf n.;>.;>v'v,,,,,vv, Inc.
OF
~Best",ftt lill.,i'i thrQugh 8UNeiltane. d.tl (plot of 4RTNDT 'IS.
fluence, orig,in ApFoprJate chemica' composition for multiple Nelllance ci!I,ps..desfrom
IIln9:le .our~ mea>> w.uefor *.11 CapSl.dH sour:ee)
Appropriate normafb:ing parameters for Stl NeiflanCfJ data 8I!!UllleS;SifiICii credibility (I.e., m.a;n of .unrelnanee dam.) and determining' CF (i.e.! best .Um __ of v~'.* Report No. 0901132.401, Rev. 0 B-4 .~/l!'i'8!'tJli't:I;l mr.~nnrI!Assocjates, Inc.
Vesse' being analyzf)d P'ant X S_teatlma.te cbemistty for h.at (WeJd metal) 0.'61% NI - Table C'V..._I ChliM. Z Cr~i,bUity assessment -m Using Plant .jtXit data only NQ ~m'~f.tUt. adjustment needed Determ ilne SurveUla;nCe elF for Plant X data only (214.8"f) Report No. 0901132.401, Rev. 0 B-5
Table CF SlJfV, ch.n., -182.S"F Oe-termlne Survellliilnce OF iildjustment ":"dld S!U:iNeiIIIJliCe elF ;r;; 259.0"f Finiill* RTNIYf{U) ;r;; ..7.0I>F; "'" 49,S; f - 0.8745 RTNOT "'" ,;1.0" + (259.0 '" O.614S) ::: 269.i<!oF Report No. 0901132.401, Rev. 0 B-6 {SS(JClarBS, Inc.
adjusted to mean ~mlica' Cu ~ 0,31% H.! := Data adj~ted to mean, temperatlJ,re of surveillance cap8uf.1I Temp. 11: !550!>P' Data are not credible 3inc., .catler is greater than OJ), (28GF) for
**val'al $~nt.illanc, .p.elm.n.
Report No. 0901132.401, Rev. 0 B-7 S1Iflm"'I'II'!!I1 Inttroritll Ass(J!cia,tes. Inc.
APPENDIXC PALISADES SUPPLEMENTAL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 Report No. 0901132.401, Rev. 0 C-l "Hn"""iF"" in.tpm!Jffli Associates, Inc.
BAW-2398 May 2001 Test Results of Capsule SA*240-1 Consumers Energy Palisades Nuclear Plant -- Reactor Vessel Material Surveillance Program DD by M. J. DeVan. Ffr Document No. 77-2398-00 (See Section 7 for document signatures.) Prepared for Consumers Energy Prepared by Framatome ANP, Inc. 3315 Old Forest Road P. O. Box 10935 Lynchburg, Virginia 24506-0935 fRAMATOME AHP
Executive Summary This report describes the results of the tests performed on the specimens contained in the second supplemental reactor vessel surveillance capsule (Capsule SA-240-1) from the Consumers Energy Palisades Nuclear Plant. The objective of the program is to mom tor the effects of neutron irradiation on the mechanical properties of the reactor vessel materials by testing and evaluation of Charpy impact specimens. Supplemental Capsule SA-240-1 was removed from the Palisades reactor vessel at the end-of-cycle 14 (EOC-14) for testing and evaluation. The test specimens included modified 18mm Charpy V-notch inserts for three weld metals fabricated with weld wire heats W5214, 34B009, and 27204 and standard Charpy V-notch specimens fabricated from the correlation monitor plate material, HSST Plate 02. The weld metal Charpy inserts were reconstituted to full size Charpy V-notch specimens. The reconstituted weld metals along with HSST Plate 02 material were Charpy*impact tested. The results of these tests are presented in this document. ii FRAMATOME AHP
Table 3*2. Chemical Composition of Palisades Capsule SA*240*1 Surveillance Materials Chemical Composition, wt% Weld Metal Weld Metal Weld Metal Correlation Monitor Plate Element W5214(a) 34B009(a) 27204(b) Heat No. A1195-1(C) C 0.094 0.110 0.142 0.23 Mn 1.161 1.269 1.281 1.39 P 0.009 0.012 0.009 0.013 S 0.012 0.016 0.008 0.013 Si 0.252 0.181 0.217 0.21 Ni 1.04S(b) 1.121{b) 1.067 0.64 Cr 0.040 0.040 0.071 --- Mo 0.510 0.543 0.525 0.50 Cu 0.307(b) 0.18S{b) 0.194 0.17 (a) ABA Technology analysis.[9] (b) Analysis provided by Consumers Energy.£lO] (c) ORNL analysis. Ill ] 3-4 FRAMATOME ANP
Table 4-3. Charpy Impact Results for Palisades Capsule SA*240*1 Irradiated Weld Metal W5214 Test Impact Lateral Shear Specimen Temperature, Energy, . Expansion,_ Fracture, ill of ft-Ibs mil % 2ALl 70 14 9 0 2AB3 125 15.5 6 20 AW5 175 24.5 15 10 2An 200 13 10 40 AU4 200 26.5 15 35 2AL3 225 25 11 50 API 250 40 26 65 AU5 300 54.5 47 95 2AE5 350 49 42 95 2AK5 400 50.5* 35 100 AP5 450 52.5* 45 100 AS2 500 54.5* 43 100
Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.£17]
4-7 FR.AMATOME .ANP
Table 4~6. Charpy Impact Results for Palisades Capsule SA*240*1 Irradiated Correlation Monitor Plate Material (HSST Plate 02) Heat No. All9S-! Test Impact Lateral Shear Specimen Temperature, Energy, Expansion, Fracture, ID OF ft-lbs mil % 02D2-10 70 6.5 4 0 02D2-13 125 15.5 10 20 02D2-23 175 27.5 24 30 02D2-17 200 26 19 35 02D2-2 200 44.5 29 55 OCD2-22 225 44.5 30 55 02D2-19 240 54 40 70 02D2-8 250 70 50 80 02D2-5 300 83* 66 100 02D2-15 350 82.5* 72 100 02D2-24 400 89.5* 67 100 02D2-11 500 82.5* 65 100
Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.[171 4-10 ""
FRAMATOME A .... P
Table 4*11. Hyperbolic Tangent Curve Fit Coefficients for the Palisades Capsule SA*240*1 Surveillance Materials Material Hyperbolic Tangent Curve Fit Coefficients Description Absorbed Energy Lateral Expansion Percent Shear Fracture Weld Metal A: 27.4 A: 22.8 A: 50.0 W5214 B: 25.2 B: 21.S B: 50.0 C: 111.6 C: 83.5 C: 72.5 TO: 20S.1 TO: 231.7 TO: 223.2 Weld Metal A: 29.S A: 22.9 A: 50.0 34B009 B: 27.6 B: 21.9 B: 50.0 C: 111.7 C: 88.0 C: 109.8 TO: 176.6 TO: 184.3 TO: 192.6 Weld Metal A: 28.0 A: 25.6 A: 50.0 27204 B: 25.8 B: 24.6 B: 50.0 C: 145.7 C: 169.2 C: 118.4 TO: 215.3 TO: 225.9 TO: 210.1 Correlation A: 43.3 A: 35.8 A: 50.0 Monitor Plate, B: 41.1 B: 34.8 B: 50.0 HSST Plate 02 C: 75.3 C: 83.1 C: 75.9 (Heat No. A1l95-1) TO: 211.8 TO: 222.2 TO: 206.5 4-15 FRAMATOME ANP
Table 4-12. Summary of Charpy Impact Test Results for the Palisades Capsule SA-240-1 Surveillance Materials 30 ft-Ib Transition Temperature, 50 ft-Ib Transition Temperature, 35 mil Lateral Expansion Material OF OF Transition Temperature, OF Upper-Shelf Energy, ft-Jb Description Unirradiated Irradiated ll.T Un irradiated Irradiated AT Un irradiated Irradiated AT Unirradiated Irradiated Decrease WeJdMetal -60.i') 219.9 280.1 _17.4(a) 372.7 390.1 -29.6(0) 284.3 313.9 102.ia) 52.5 50.2 W5214 Weld Metal -82.0(') 177.4 259.4 -45.0(') 280.8 325.8 -51.6(3) 238.6 290.2 113.9(0) 57.4 56.5 34BOO9 J Weld Metal -41.2(0) 226.6 267.8 -6. I (b) 399.7 405.8 Not 293.7 --- 108.4(b) 53.8 54.6 27204- available. HSST Plate 02 45.7(c) 186.6 140.9 78.3(C) 224.2 145.9 Not 220.3 -- 120.3(C) 84.4 35.9 Heat No A1195-1 available. t-I-' 0\ (a) Data reported in AEA Technology Report AEA-TSD-0774.[9] (b) Data reported in CE Report No. TR_MCC_189.[181 (c) Data reported in NUREGlCR-6413.[11J
~
0
~
o m
!Z "V
Figure 4~2. Palisades Capsule SA-240-1 Charpy Impact Data for Irradiated Weld Metal W5214 100 (ft.
...ai 75
.a t) (!l
'- 50
!.I.
...III (IJ 25
.c (/) 0 0 100 200 300 400 500 600 Temperature, F ~ 100 E c:- 0 80 "iii 60 f: w (!l
c. 40
'iii 20 ~ ...I 0
0 100 200 300 400 500 600 Temperature, F 120 TasMlE : ~
T50: +372.7 F 100 T30: +219.9 F CvUSE: ~ 80 1l ~ til (IJ 60 l:: W Col
CI.
.5 4D 20 Material: Weld Metal It Heal Number. W5214 0 0 100 200 300 400 500 600 Temperature, F 4*18 FRAMATOME ANP
Figure 4-5. Palisades Capsule SA*240-1 Charpy Impact Data for Irradiated Correlation Monitor Plate Material (HSST Plate 02), Heat No. A1195-1 100
?fl.
l!! 75
.au I! 50 .m Il.
25
.c t.fI 0
0 100 200 300 400 500 600 Temperature, F
!l 100 E
c" eo
.2 c 60
m a.
I< 40 W ! 20 S
!l 0 0 100 200 300 400 500 600 Temperature, F 120 r;==========~----------------------------------,
T35MLE : .f-22Q.3F T50: +224,2 F 100 T3(): +186.6 F CvUSE: 84.4 ft-Ib .8 80 ~
~CD 60
~ '0
!i
.5 40 20 Material: HSST Plate 02 Orientation; Longitudinal HI' Nllmhar: A1195*1 o 100 200 300 400 500 600 Temperature, F 4-21 FRAMATOME ANP
6.0 Summary of Results The investigation of the post-irradiation test results of the materials contained in the second supplemental surveillance capsule, Capsule SA-240-1 removed from the Consumers Power Company Palisades reactor vessel, led to the following conclusions:
1. Observation of the Capsule SA-240-1 thermal monitors indicated that the irradiated test specimens were exposed to a maximum irradiation temperature less than 558°P.

2. Thirty-six pre-machined irradiated 18mm Charpy inserts were successfully reconstituted and machined to Type A Charpy impact specimens. The reconstituted Charpy specimens were subsequently impact tested.

3. The 30 ft-lb and 50 ft-Ib transition temperatures for the weld metal W5214 increased 280.1 °P and 390.1 °P, respectively. In addition, the CvUSE for this material decreased 48.9%.

4. The 30 fr-lb and 50 ft-Ib transition temperatures for the weld metal 34B009 increased 259.4°P and 325.8°P, respectively. In addition, the CyUSE for this material decreased 49.6%.

5. The 30 ft-lb and 50 ft-Ib transition temperatures for the weld metal 27204 increased 267 .8°P and 399.7°P, respectively. In addition, the CvUSE for this material decreased 50.4%.

6. The correlation monitor plate demonstrated similar behavior with an increase in the 30 ft-Ib and 50 ft-Ib transition temperatures of 140.9°P and 145.9°P, respectively.
The percent decrease in the CvUSE for this material is 29.8%. 6-1 ERAMATOME ANP
J BAW-2341, Revision 2 o May 2001 o o Test Results of Capsule SA-GO-l Consumers Energy TJ L. Palisades Nuclear Plant [I ** Reactor Vessel Material Surveillance Program ** o by M. J.DeVan [] D o o FTI Document No. 77-2341-02 (See Section 7 for document signatures.) o o u Prepared for Consumers Energy [J Prepared by Framatome ANP, Inc. 3315 Old Forest Road U P. O. Box 10935 Lynchburg, Virginia 24506-0935 J II FRAMATOME ANP
f\ U o Executive Summary u This report describes the results of the test specimens from the first supplemental capsule o (Capsule SA-60-1) of the Consumers Energy Palisades Nuclear Plant as part of their reactor vessel surveillance program. The objective of the program is to monitor the effects of neutron irradiation on the mechanical properties of the reactor vessel materials by testing and evaluation o of Charpy impact specimens. Supplemental Capsule SA-60-1 was removed from the Palisades reactor vessel at the end-of-D cycle 13 (EOC-13) for testing and evaluation. The capsule contents were removed from Capsule SA-60-1 for testing and examination. The test specimeps included modified 18mm Charpy V-o notch inserts for three weld metals fabricated with weld wire heats W5214, 34B009, and 27204 and standard Charpy V -notch specimens fabricated from the correlation monitor plate material, D HSST Plate 02. The weld metal Charpy inserts were reconstituted to full size Charpy V-notch specimens. The reconstituted weld metals along with HSST Plate 02 material were Charpy o impact tested. Following the initial Charpy V-notch impact testing, the laboratory perfonned-a calibration of o the temperature indicator used in the Palisades Capsule SA-60-1 testing. The results of the laboratory calibration indicated the instrument was out-of-tolerance. Based on the results of this calibration test, the laboratory revised the Charpy impact test temperatures accordingly. Revision tJ 1 corrects the test temperatures for the Supplemental Capsule SA-60-1 reconstituted weld metal Charpy V-notch impact specimens and the HSST Plate 02 Charpy V-notch impact specimens. Revision 2 provides an update to the hyperbolic tangent.curve fits of the Charpy impact curves by restraining the upper-shelf energy. For these curve fits, the lower-shelf energy was fixed at 2 2.2 ft-Ibs for all cases, and for each materials the upper-shelf energy was fixed at the average of all test energies exhibiting 100% shear, consistent with ASTM Standard E 185-82. o o [J D ii FRAMATOME ANP
r] Table 3-2. Chemical Composition of Palisades Capsule SA-60-1 Surveillance Materials LJ Chemical Composition, wt% Weld Metal Weld Metal Weld Metal Correlation Monitor Plate [J Element W5214(a) 34B009(a) 27204(b) Heat No. A1195-1 (el C 0.094 0.110 0.142 0.23 [J Mn 1.161 1.269 1.281 1.39 P 0.013 o 0.009 0.012 0.009 S 0.012 0.016 0.008 0.013 Si 0.252 0.181 0.217 0.21 o Ni Cr
- 1.04S(b) 0.040 1.121 (b) 0.040 1.067 0.071 0.64
[1 Mo 0.510 0.543 0.525 0.50 Cu 0.30ib) 0.185(b) 0.194 0.17 o (a) ABA Technology analysis. 8 J (b) Analysis provided by Consumers Energy.9 o (c) ORNL analysis. 10 [] IIu u o u q L. o 3-4 FRAMATOME ANP
J Table 4*3. Charpy Impact Results for Palisades Capsule SA 60 1 Irradiated Weld Metal W5214 G 0 J Specimen Test Temperature, Impact Energy, Lateral Expansion, Shear Fracture, J ID OF ft-lbs mil %
]
AA2 74 10 4 0 AWl AW2 129 154 24 23.5 15 15 5 15 J 2AF5 AL3 204 229 30 33.5 16 23 50 65 J AA4 2AL6 254 279 28 43.5 19 35 60 80 J 2AB2 2AH6 279 329 48.5 47.5 38 90 90
]
35 100 AR94 2AHl 404 454 51.5* 55* 43 47 100 J AV4 57* 100 479 46 J
Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.15 J
] ] "l r-I ~
4-7 FRAMATOME At-.lP
D [J Table 4-6. Charpy Impact Results for Palisades Capsule SA*60*1 Irradiated Correlation Monitor Plate Material (HSST Plate 02) Heat No. A1195*1 0 Test Impact Lateral Shear Specimen Temperature, Energy, Expansion, Fracture, 0 ID of ft-lbs mil % 02D2-9 74 8 3 5 [J 02D2-3 104 20.5 16 10 02D2-1 129 24.5 18 20 0 02D2-7 154 26.5 23 40 02D2-18 179 35.5 28 45 J OCD2-21 204 48.5 40 70 02D2-14 229 53.5 43 65 0 02D2-16 229 51.5 43 70 02D2-4 254 73.5 65 80 0 02D2-6 279 85* 70 100 02D2-12 329 -87.5* 74 100 ~J 02D2-20 404 86.5* 77 100 0 Value used to detennine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82. 15 D 0 0 [J 0 0 [] lJ 4-10 FRAMATOME ANP
]
Table 4*11. Hyperbolic Tangent Curve Fit Coefficients for the Palisades Capsule SA*60-1 Surveillance Materials J Material Hyperbolic Tangent Curve Fit Coefficients J Description Absorbed Energy Lateral Expansion Percent Shear Fracture Weld Metal A: 28.4 A: 25.0 A: 50.0 J W5214 B: 26.2 B: 24.0 B: 50.0 '. C: TO: 158.1 188.8 c: 160.0 TO: 239.6 C: 80.5 TO: 214.9 J Weld Metal 34B009 A: B: 28.7 26.5 A: B: 25.3 24.3 A: B: 50.0 50.0
]
C: 123.8 c: 97.6 C: 89.6 Weld Metal TO: A: 161.8 27.6 TO: A: 196.4 25.9 TO: A: 179.6 50.0 2J 27204 B: 25.4 B: 24.9 B: C: 111.4 C: 101.8 C: 50.0 92.1 ] TO: 201.4 TO: 214.4 TO: 187.1 Correlation Monitor Plate, A: B: 44.3 42.1 A: B: 41.3 40.3 A: B: 50.0 50.0
]
HSST Plate 02 C: 95.1 C; 104.9 C: 85.2 (Heat No. A1195-1) TO: 193.0 TO: 208.6 TO: 183.7 J J J J J J J J J 4-15 FRAMATOMEANP ]
~ r ------I 1--', r-'-J C::J f-"~ L.:J ~ CJ c:.:.J CJ L.J r --I [----J c:.J c:J c=J r -1 L-i
-.~.--,
Table 4-12. Summary of Charpy Impact Test Results for the Palisades Capsule SA-60-1 SurveiHam;e Materials 30 ft-lb Transition Temperature, 50 ft-Ib Transition Temperature, 35 mil Lateral Expansion Material OF OF Transition Temperature, OF Upper-Shelf Energy, ft-lb Description Unirradiated Irradiated L\T Unirradiated Irradiated L\T Unirradiated Irradiated ll.T Unirradiated Irradiated Decrease Weld Metal -60.2(0) 198.8 259.0 -17.4(') 375.6 393.0 -29.6(0) 310.1 339.7 102.7(0) 54.5 48.2 W5214 Weld Metal _82.0(B) 167.8 249.8 -45.0(0) 298.6 343.6 _51.6(0) 237.5 289.1 113.9(0) 55.25 58.65 34B009 2 Weld Metal _41.2(b) 211.9 253.1 -6. 1(b) 355.6 361.7 Not 249.4 --- 108.4(b) 53.0 55.4-27204- available. HSST Plate 02 45.7'0) 159.4 113.7 78.3(0) 206.0 127.7 Not 187.9 -- 120.3(0) 86.3 34.0 Heat No Al 195-1 available. I ..j:>. ~ (a) Data reported in ABA Technology Report ABA-TSD-0774. 8 16 (b) Data reported in CE Report No. TR-MCC-189. 1o (c) Data reported in NUREG/CR-6413. z "tI
o o Figure 4-2. Charpy Impact Data for Irradiated Weld Metal W5214 o ~
~ 75 100 o = ~
u..~ I.- m II) 50 25 o w 0
*100 0 100 200 300 Temperature, F 400 500 600 ~ 100 E
[J r:
.2 c
II) 80 60 IU Q. 40 J
)(
w 1! 20
.II:!III .,' 0 o ...J *100 0 100 200 300 400 500 600 Temperature, F 2
J 120 TSSMLE: +310.1 F o 100 T50: T30:
+375.6 F +198.8 F CvUSE; 54.5 ft*lb o II) .Q ;if 80 o .- ~
III 60 w c u . ...................... .- o m Q.
.§ 40 o 20 Material; Weld Metal Heat Number: W5214 L1 0 -100 0 100 200 300 400 500 600 Temperature, F o
4:-18 FRAMATOME A .... P
]
Figure 4*5. Charpy Impact Data for Irradiated Correlation Monitor Plate Material (HSST Plate 02), Heat No. A1195-1
] </i.
100 J
.e u.
CJ III 75 50 J 10 III
.s::
r.II 25 0 0
*100 0 100 200 300 400 500 600 Temperature, F J .!!! 100 'E c0 80 J 'u; 60 c
w III 1:1. x 40 J
'iii 20 ~
III
...I 0 -100 0 100 200 300 400 500 600 ]
Temperature, F 2 J 1~~==========~----------------------------1 100 TS5MLE: Tso: Tso:
+/-l§Z.li +206.0 F +159.4 F J
CvUSE: 86.3 ft*!b
]
III
..c 80 t ~ ~
s: 60 J w t) 111 c.. ]
.5 40 ]
20 Material: HSST Plate 02 Orientation: Longitudinal oL-~~~~~~~~~==~~~~==~ Heat Number: A1i95-1 J
-100 o 100 200 300 400 500 600 Temperature, F J ]
4-21 FRAMATOMEANP ]
Palisades Nuclear Plant - Weld/W5214 (Unirr) CVGRAPH 4.0 Hyperbolic Tangent Curve Printed at j5:08:08 on 12-01-1995 Page 1 Coefficients of Curve 1 A = 52.4 B = 50.29 C =99.19 TO = -12.69 Equation is: CVN :: A + B
I tanh((T - TO)/C) 1 Upper Shelf Energy: 102.7 Fixed Temp. at 30 ft-lbs: -60.2 Temp. at 50 ft-lbs: -17.4 Lower Shelf Energy: 2.1 Fixed Material: 'YELD Heat Number. W5214 Orientation: TL Capsule: Unirr Total Fluence: 0.0 300 250 200 150 0 0 0 100 ~ u E
~
50
~
yr u
-300 0200 -100 0 100 200 300 400 500 600 Temperature In Degrees F Data Set(s) Plotted Plant: PAL Cap.: Unirr Material: WELD Ori~ TL Heat t. W5214 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential -110 11.8 14.5 -2.7 -110 H.8 14.5 -2.7 -80 33.9 22.69 112 -80 20.6 22:69 -2.09 -80 29.5 22.69 6.B -40 47.9 38.89 9 -40 43.5 38.89 4.6 -40 29.5 38.89 -9.39 -40 41.29 38.89 2.4 **** Data continued on next page ....
Palisades Nuclear Plant - Weld/W5214 (Unirr) Page 2 Material: WELD Heat Number: W5214 Orientation: TL Capsule: Unirr Total Fluence: 0.0 Charpy V-Notch Data (Continued) Temperature Input CVN Energy Computed CVN Energy Differential o 64.19 58.8 5.39 o 39.09 5B.8 -19.7 20 62.7 68.4 -5.7 20 60.5 68.4 -7.9 30 78.19 72.B 5.39 40 612 76.86 -15.66 60 S7 83.82 3.17 60 75.19 83.82 -8.62 60 111.4 83.82 27.57 110 110.59 94.88 15.71 110 9S.8 94;00 3.91 210 110.59 lO1.58 9.01 210 95.9 lO1.58 -5.68 300 97.4 lO2.51 -5.11 300 94.4 102.51 -B.ll SUM of RFSIDUAIB = 10.77
APPENDIXD INDIAN POINT 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 Report No. 0901132.401, Rev. 0 D-l Xtr/itWf't:ii Infrmriftv Associates, Inc.
, ~\ . . ,~l\~Pr::'~TP~~Tlf.:"1!""1! r;~e,Tr?~ 'T'ON ~.,..::r,lI
).\
., FOR 11 . . . . . . 1:1 ....... ~.J ...... ' 14 04 ....... ". ~.,,. a b......- * ..-.J :.. ...." ",~.v J .... **.** ' Is DATE: _ _ _ _ _ _ _ SE_P_2_9_1_97_1_1_WNES
c. I CONSOLIDATED EDISON CO.
INDIAN POINT UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM By S. E. Yanichko May 1969 IFP 106 APproved:~~~----~~~~
£O~___ ..
E. Landerman Westinghouse Electric Corporation Nuclear Energy Systems Division Box 355 Pittsburgh, Pa. 15230
SECTION 4 POST-IRRADIATION TESTING Specimen capsules will be removed from the reactor only during normal refueling periods. The recommended scbedule for removal of capsules is as follows: Capsule Capsule Type Identification Exposure Time I T (Replacement of 1st Region) II S (Replacement of 2nd Region) I Z (Replacement of 4th Region) II V 10 years I U I I W Extra capsules for complementary I X testing or additional exposure II Y Each specimen capsule upon removal after radiation exposure will be trans-ferred to a post-irradiation test facility for disassembly of the capsule and testing of all specimens. 4.1 CHARPY V-NOTCH IMPACT TESTS The testing of the eight Charpy impact specimens from each of the IPP vessel plates, the weld and HAZ metal, and the correlation monitor material in the capsules can be -done singulary, ~ possible the performanGe of Charpy imp.act tests at five different temperatures, with three extra specimens to provide an optimum curve for each plate. Tbe initial Charpy specimen from the first capsule removed shpu1d be tested at room temperature. The impact energy value for this te$t temperature should be compared with pre-irradiation test data. The testing temperatures for the remaining specimen should then be appropriately raised or lowered. The test temperatures of 9
7 specimens from capsules exposed to longer irradiation periods should be determined by the test results for the previous capsule. 4.2 TENSILE TESTS The two tensile specimens per plate or weld from each of the capsules should be tested at room temperature and the approximate operating temperature of the reactor (550 0 P). 4.3 WEDGE OPENING LOADING TESTS The WOL specimens from each individual capsule should be tested at a temper-ature based on the transition temperature shift obtained from the associated Charpy impact specimens. fA. mean temperatur.e of -200 "'F plus the transition temperature shift should be the initial test temperature. 4.4 POST-IRRADIATION TEST EQUIPMENT
1. Milling machine or special cut~off wheel for ope~ capsules and dosimeter blocks.

2. Hot-cell tensile testing machine with:

a. pin-type adapter for pulling tensile tests

b. clevis and extensometer for pulling WOL specimens.

3. Hot-cell Charpy impact testing machin~.

4. NaI scintillation detector and pulse he~ght analyzer for gamma counting of the speci*fic activities of the dosimeter$.
10
TABLE 7 PRE-IRRADIATION TENSILE PROPERTIES FOR THE INDIAN POINT UNIT NO. 2 PRESSURE VESSEL PLATE MATERIAL AND WELD METAL Test 0.2% Yield Tensile Total Reduction Plate Temp. , Strength, Strength, Elongation In Area, No. OF psi psi % % B2002-1 Room 68,500 89,000 25.1 67.8 B2002-1 Room 65,850 87,800 25.3 67.4 B2002-1 200 61,550 79,900 24.1 68.6 B2002-1 200 67,950 89,400 23.8 67,6 B2002-1 400 57,900 79,900 23.1 64.7 B2002-l 400 59,800 82,200 22.2 67.8 B2002-1 600 56,750 80,550 21. 9 64.3 B2002-l 600 57,750 85,700 22.9 64.2 B2002-2 Room 62,350 83,800 27.1 70.0 B2002-2 Room 66,750 90,500 28.2 69.6 B2002-2 200 63,650 84,450 24.8 70.5 B2002-2 200 63,200 83,800 25.5 67.3 B2002-2 400 53,800 77,900 23.1 68.5 B2002-2 400 52,650 73,150 22.4 67.6 B2002-2 600 53,500 78,800 22.7 64.4 B2002-2 600 54,700 81,450 24.7 64.4 B2002-3 Room 65,650 87,300 27.6 67.3 B2002-3 Room 65,000 87,350 24.8 66.7 B2002-3 200 67,800 88,900 23.4 68.6 B2002-3 200 67,700 89,150 22.1 64.9 B2002-3. 400 57,950 79,550 22.3 68.7 B2002-3 400 55,350 77,100 23.2 64.9 B2002-3 600 57,750 83,850 24.9 68.2 B2002-3 600 58,350 86,500 24.9 64.7 Weld Room 64,500 80,700 28.5 73.9 Weld Room 65,000 81,000 26.9 71.5 Weld 200 63,450 76,100 28.4 72.9 Weld 200 61,050 75,200 25.2 73,0 Weld 400 57,550 75,000 . 22.9 68.1 Weld 400 58,300 75,800 22,6 69.6 We14 600 56,650 79,800 24.4 62.0 Weld 600 56,650 79,200 24.0 66.9 21
1297-8 140 120 100 V) co ....l ~ 80 u.. 0:: u.J 3S 60 40 20 Q ~--------~----------~--------~----------~--------~
-200 -100 o 100 200 300 TEMPERATURE (OF)
Figure 10" Pre-Irradiation Charpy V-Notch Impact Energy for the Indian Point Unit #2 Reactor Pressure Vessel Weld Metal 37
SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Cutebr8 Road San Antonio, Texas 78284 by E. B. Norris FINAL REPORT SwRI Project No. 02-5212 Prepared for Consolid'lted Edison Company of New York, Inc. 4 Irving Place New York, New York 10003 November 1980 U*. S. lindholm, Director Department of Materials Sciences
!ABL! V!I CHARrY V-NOTCH L~ACT DAtA L"IDIA..'t POINT UNIT NO. 2 PRESS';,..'R.E nSSE!. WELD MEtAL Lat@ral ~
Spee. Temp. £n'!rgy Shear Expansion Condition ~ .J.:!L (ft-lbf) ..J!L (m:l.llJ ) Baseline (a) -150 12.5 10 10
-150 10.5 1.5 11 I! -100 -100 35.0 9.0 25 20 29 9 -100 18.0 .30 19 -80 1.3.0 20 12 -80 32.5 20 Zi -80 21).0 20 23 -40 34.0 .30 30 -40 35.S 35 31 -40 48.0 35 40 10 78.5 60 64 10 74 .0 60 60 10 8LO 70 68 ~o 102 .. 5 SO i8 60 10 ** 0 85 82 60 100.0 85 80 110 112.5 99 88 110 108.5 90 87 110 108.5 98 88 160 11.5.5 100 90 , 160 113.0 100 92 I
I 160 1::0.0 lOG 93
,I I: I 210 121.0 100 n f t 210 210 123.5 117.5 100 100 91 92 Capsule Y I
W-li 74 ..,.,, .,- :lil 14 W-19 110 23.:J ,; 19 I! ~*i-20 160 t.O.O 2.5 34
'J-21 190 :'i.J 50 43 ~-23 210 55.J 60 53 ;~-24 260 7:.5 100 51
~-18 (b) 300 61.0 :"')0 45 r :J-22 350 6i.:J 100 52 (a) ~ot reported.
(b) Specimen n~mber stamped on !~ac: sice.
TABLE VIII CHARPY V-NOTCH IMPACT DATA COR.R!UTION MOHnOR MAn:IU.AL (SU1'PLIED BY U. s. STEEL) Lateral Spec. Teml' * £':I!rgy Shear Expansion Condition ~ J.:!L (£t-105) ...ill.... (mils) !aseline (a) -80 4 2 6 fI -80 4 2 6 I a I
-60 -60 6 3
3 6
6
-40 12 10 14 -40 10 5 10 -40 6 5 7 -20 14 l.5 14 -20 13 15 14 0 22 30 22 0 18 25 18 20 29 35 28 20 23 35 23 ':'0 36 45 ~3 I 40 26 45 26 I
I , 60 60 36 33 50 45 40 35 I 80 67 100 60 I 80 50 70 48 t Capsule '{ R-60 I r 100 100 40 68 62 5.0 98 85 nil 60 58 4 i R-57 74 26.0 5 2: R-62 90 30.5 10 26 a-58 110 23.0 15 25
, R-59 135 36.0 :0 32 \ R.-63 160 51.5 40 !i3 I a-64 210 60.0 90 33 r c
\.-61 260 68.5 100 58 (a) ~ot reported.
Coad: -0 0 !.:nirt'.lciiated
5.89,. ,x 10 18 (ll: ~ ~e\')
.x 120 ""
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£. /0 Q~ 1-1 ./ ./ ,",01 ~ 1: *. , h l ".- /' v y 4!1 o -200 -100 100 200 300 ':'00 Temperacure, deg F 110 C"de:
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, G -. ~o ~ !.I ~ !&l 1/ / --) ""-F... Y u / Code:
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-200 -100 lelO :00 300 :'00 Temperature, deg .
60 !:: ~ 11&
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REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR lNDIAN POINT UNIT NO.2 ANALYSIS OF CAPSULE V FINAL REPORT SwRI Project No. 17*2108 Prepared for Consolidated Edison Company of New York, Inc. 4 Irving Place New York, New York 10003 October 1988 Written by Approved by .) B. T. Cross Director Department of NDE Science and Research Southwest Research Institute
l SUlIlMARY OF RESULTS AND CONCLUSIONS o The analysis of the fourth material smveillance capsule removed from the Indian Point Unit No. 2 reactor pressure vessel led to the following conclusions:
-i' (1)
Based on a. calculated neutron spectral distribution, Capsule V received a fast fiuence of 5.3.x 1018 n/em 2 (E > 1 MeV) at its radial center line,
I (2) The surveillance specimens of the core beltline plate materials experienced shifts in RTNDT (50 ft-Ib. values) over the range of *79°F (Plate B2002*2) to 239°F (Weld) as a result of fast neutron exposure up to the 1987 refueling outage.
(3) The core beltline weld exhibited the largest shift in RTNDT and is projected to control the heatup and cooldown limitations throughout the 'design lifetime of the pressure vessel. i ' (4) . From the previous capsule, Z, the estimated maximum neutron fiuence of 3.33 x lOIS'" neutrons/cm 2 (E > 1 MeV) was received by the vessel wall in 5.17 effective full power years (EFPY) through Cycle 5, which is equal to a fiuence rate of 6.44 x 1017 '" per EFPY. At the end of Cycle 8 (8.6 EFPY) the neutron fiuence was 4.45 .x 1018 n/cm 2 giving 3.26 t.". x lOF n/cm2 per ErFY for Cycles 6 through 8. This calculated value for the decrease in fiuence per EFPY agrees well with the experimental value for the decrease in fiuence rate; i.e., 50.6% vs. 48.9%. The use of a low leakage core loading pattern beginning with ( "', Cycle 6 did significantly reduce the fIuence rate on the pressure vessel wall. U
$Revised from Capsule Z report using the la.test plant specific lea.d factors. \i u
1-1
TABLEIV-5 CHARPYlMPACT DATA WITH PHOTOS OF FRACTURE FACES SPECIMEN TEMP ENERGY LATERAL FRACTURE PHOTOGRAPH NO. of FT-LSS EXPANSION APPEARANCE X w- 9 74"'F 24.0 .019 0
. . . -10 +130 26.5 .023 ~O ;~- 11 +180 40.5 .OJ5 !to ~~ .. 12 +220 5J.0 .048 65 ;-1 J +260 62.5 . 054 95 * ,.;- 14 +300 76.0 .Oo!. 95 \.;-16 +325 7:2.5 .065 95 \y-15 +350 76.0 .067 100 I
I
~ . .' \ - \
I IV-15
TABLEIV-5 CHARPYIMPACT DATA WITH PHOTOS OF FRACTURE FACES I .
~ I ::a..:e ..!ur.e 2, :SSo "'.AT~RrAL ~ (WELD) --~------~--------
SPECIMEN TEMP ENERGY LATERAL FRACnJRE PHOTOGRAPH NO. of FT-LBS EXPANSION APPEARANCE '~.-~'
..LX w- 9 74°F 24.0 .019 0 r . . . . -LO +130 26.5 .023 20 \,1-11 +180 40.5 .035 40 ~~~- t2 +220 53.0 .048 65 W-13 +260 62.5 .05<' 95 ,.;-14 +300 76.0
Db.!. 95
\~-16 +325 72.5 .065 95 \v-15 +350 76.0 .067 100 l ..:; , I I ..,
I'; 8Y
WELD METAL 160 rr-:-- o
-..,----,--------r BASELINE 1 I I ... CAPSULE Y + CAPSULE V I I ";' 120 r ..;
I I
..0 I - I
1. I
.:::.. 80 ~
i II 40 j..I I I I ot~~.~~~~~~~~~_ ,t", r--' -200 I , t __: ['
----1----- ----r"]
100 j (I)
...0 BASELINE CAPSULE Y I ~I + CAPSULE V l "
z E
......., 75 t I
0 I
~O ~ * * * *J Vi i ~
Z A
<{
a.. I I I X I W I I I
....l ,
(1
, I ~ 25 i U w 35 mil r--~
I-
....l I! !J:..J 0 -I I
I I *
! . -200 -1 00 0 100 200 300 400 l __ :
TEMPERATURE (deg F) ( j Figure IV-3. Radiation Response of Indian Point Unit No.2 Weld Metal U IV-20
TableIV-9
SUMMARY
OF RTNDT SHIFTS.AND UPPER SHELF ENERGY REDUCTION (Cy) 0 FOR MATERIALS IN CAPSULE V ,'I, A. So.mmar.r ofFlueDCl!! and:M:euured ~TNl)'l' Values fur Test Speci.menB in Capsule V I _J FIuance Type of Neutron Measured .6.RTNDT ell') Material cm2 50 Ft-Lbs 30 Ft-Lbs 35 mils'" t**i Weld 5.59E18 239 204 230 77 Ft-Lbs 46 Ft-Lbs Plate :62002*2 4.57E18 85 80 97 HAZ 5.59E18 loo 162 184 Correlation Monitor 4.57E18 NA"' 104 108 B. Decrease in U~ Shelf Energy (c;.) Initial Shelf Capsule V"' .... Cv Material Ft-Ib Ft-lb Ft-Ib % Decrease B2002-2 117 111 6 5 Weld Metal 118 75 43 36 HAZ 100 98 2 (nil) 2 Correlation Monitor 118 70 48 41
. ..J "'3/5 mil + 20°F included in table. ""*The upper shelf energy for this capsule was below 77 :ft lbs . . I ..... Average of 3 Charpy measurements at (::I 100% ductile failure. \ ;
IV-28
TABLE IV-l1(Cont'd)
SUMMARY
OF CHEMISTRY VALUES FOR INDIAN POINT UNIT NO. 2. MATERIALS rt U ,. , Material .Source of Data CuW% NiW% i Plate B2002.-3 WCAP7328 (.14)"' (,57)- Capsule-Z: Cv Specimen 3-33 .30 .64 Capsule-Z: Cv Specimen 3-38 .27 .59 Capsule-Z: Tensile Specimen 3-5 .23 .58 Capsule-Y: Cv Specimen 3-41 .21
.. Capsule-Y: Cv Specimen 3-45 .22 Capsule-Y: Tensile Specimen 8-6 (.11)"
Capsule-Y: Tensile Specimen 3-7 (.10)- Capsule-T: Cv Specimen 3-2 .27 Capsule-T: Cv Specimen 3-3 .28 Capsule-T: Tensile Specimen 3-1 (.09)- Average .25 .SO
.~
HAZ Capsule-V: Cv Specimen H-l6 .08 1.2 Capsule-V: Cv Specimen H-12 .06 1.2 Capsule-Y: Cv Specimen B-2l .15 Capsule-Y: Cv Specimen H-23 .20 Average .12 1.2 Weld Capsule-V: CvSpecimen W-13 .23 1.02 Capsule-V: Cv Specimen W-12 .20 1.06 Capsule.V: Tensile Specimen W-g .20 (.69)" Capsule-V: Cv Tehsile Specimen W-4 (.12)'" 1.00 Capsule-Y: CvSpecimen W-17 .19 Capsule-Y: CvSpecimen W-l9 .22 Capsule-Y: Tensile Specimen W-5 .18 Capsule-Y: Tensile Specimen W-6 .20 Average .20 1.03 QorrelatiQn Moni:tQr Capsule.V: Cv Specimen R-56 .20 .18 ..; Capsule-V: Cv Specimen R-52 .18 .27 Capsule-Z: Cv Specimen R-83 .85 .28 Capsule-Z: Cv Specimen R-36 .81 .27
,; Capsule-Z: Cv Specimen R-40 .21 .21
, .. Capsule-Y: Cv Specimen R*60 .17 Capsule-Y: Cv Specimen &*62 .19 I i ":
, Capsule-T: Cv Specimen R*2 .25
~J Average .23 .24 IIIV&UOO in parenthooes discarded because of excessive deviation or were WCAP values. Surveillance specimen WCAP values not used since chemical analyses were aVailable. i 1
~~
IV*t.7
.t.:;
APPENDIXE INDIAN POINT 3 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 Report No. 0901132.401, Rev. 0 £-1 "I:f"ff,W''' ifllfpfIlYl't!l Associates, Inc.
Westinghouse Class 3 CONSOLIDATED EDISON CO. OF NEW YORK INmAN POINT UNIT NO.3 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko J. A. Davidson January 1975 APPROVED: Work Performed Under I NT 106 WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems Division P. O. Box 355 Pittsburgh, Pennsylvania 15230
SECTION 3 PREIRRADIATIO TESTING 3-1. CHARPY V-NOTCH TESTS Charpv V-Notch impact tests were performed on the vessel plates at various temperatul-es from -100 to 210°F to obtain full Charpv V-Notch transition curves (refer to tables 3-1 and 3-2, and to figures 3-1 thru 3-5). Charpv V-Notch impact tests were performed on weld metal and :-IAZ metal at various temperatures ranging from -150 to 160°F. The results are reported >-: table 3-3 and in figures 3-6 and 3-7, respectively. The Charpy V.Notch impact data for the correlation monitor material are shown in table 3-4 and figure 3-8. 3-2. . ENSILE TESTS Tensile tE>:5 were performed on the shell plates and on weld material at room temperature, 300 and f/)O°F, respectively. The results are shown in table 3-5 and in figures 3-9 through 3-14. 3-3. DROPWEIGHT NDTT TESTS Dropweight NDTT tests (ASTM E208) were performed on each plate by the fabricator. The NDTT obtbled on each plate follows: Plate Temperature 82802-1 -50°F 82802-2 -50°F 82802-3 -40°F 82803-3 -10°F 3-1
TABLE 3-3 PREIRRADIATION CHARPY V*NOTCH IMPACT DATA FOR THE INDIAN POINT UNIT NO.3 REACTOR PRESSURE VESSEL WELD METAL AND WELD HEAT AFFECTED ZONE MATERiAL Weld Metal Weld Heat*Affected Zone Test Lateral Test Lateral Temp Energy Shear Expansion Temp. Energy Shear Expansion of FT*LBS % Mils of FT-LBS % Mils -150 5.0 5 2 -150 4.0 0 2 -150 2.0 5 2 -150 6.0 5 1 -150 4.5 9 4 -150 14.0 9 11 -100 29.0 20 22 -125 16.0 9 14 -100 18.0 18 16 -125 7.0 5 8 -100 25.5 23 23 -125 34.0 -14 28 -50 35.0 40 34 -100 48.5 20 35 -50 33.0 47 30 -100 59.0 25 40 -50 32.5 40 30 -100 30.0 18 20 -35 78.0 64 66 -50 30.0 29 22 -35 69.5 67 56 -50 62.5 40 44 -35 54.5 40 47 -50 60.0 36 44 -20 87.0 77 69 10 111.0 85 79 -20 82.0 77 63 10 51.5 47 48 -20 89.0 81 74 10 83.0 62 62 10 100.0 81 78 60 142.0 100 90 10 105.0 82 81 60 127.0 100 82 10 113.5 100 85 60 121.5 100 91 60 115.0 100 89 160 111.0 100 81 60 119.0 100 84 160 125.0 100 85 60 121.5 100 90 160 143.0 100 88 160 124.0 100 88 160 125.0 100 89 160 112.0 100 90 3-9
8328-10 Figule 3-6. Preirradiation Charpy V-Notch Impact Energy for the Indian Point No.3 Reactor Pressure Vessel We Id Metal 3-10
TABLE A-1 CHEMICAL COMPOSITION OF THE INDIAN POINT UNIT NO.3 REACTOR VESSEL MATERIALS Chemical Composition (wt-%) Lower Shell Intermediate Shell Course Plate Course Plate As-Deposited Element B2802-' 82802-2 B2802-3 B2803-3 Weld Metal C 0.22 0.19 0.20 0.22 0.08 Mn 1.41 1.33 1.32 1.30 1.18 P 0.010 0.015 0.011 0.012 0.019 S 0.023 0.019 0.025 0;024 0,016 Si 0.28 0.21 0.26 0.28 0.17 Ni 0.50 0.53 0.49 0.52 1.02 Cr 0.08 0.09 0.08 0.08 0.04 Mo 0.46 0.48 0.50 0.45 0.53 Cu 0.18 0.20 0.19 0.24 0.15 A1 0.036 0.027 0.042 0.03 <0.01 V <0.01 <0.01 <0.01 <0.01 <0.01 Sn 0.014 0.017 0.014 <0.01 0.007 Cb <0.01 <0.01 <0.01 <0.01 <0.01 Zr <0.01 <0.01 <0.01 <0.01 <0.01 Ti <0.01 <0.01 <0.01 <0.01 <0.01 All other elements (except Fe) were <0.01 %. A-2
Westinghouse Non-Proprietary Class 3 WCAP-16251-NP July 2004 Revision 0 alysis Capsule X from Entergy's I i it 3 actor Vessel iati Surveillance rogram
WESTINGHOUSE NON-PROPRIETARY CLASS :3 WCAP 16251 NP, Revision 0 p p Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program T.J. Laubham J. Conermann S.L. Anderson July 2004 Weslinghouse Electric Company LLC Bm:rgy Syf;lcms P.O. Box 355 Pittsburgh, PA 15230-0355
©2004 Westinghouse Electric Company LLC All Rights Reserved
IX EXECUTIVE SUM:MARY The purpose of this report is to docwnent the results of the testing of surveillance Capsule X from Indian Point Unit 3. Capsule X was removed at 15.5 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A f]uence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDFIB-VI data-base. Capsule X received a f]uence of 0.874 x 10 19 nlcnl after irradiation to 15.5 EFPY. The peak clad/base metal interface vessel f]uence after 15.5 EFPY of plant operation was 5.86 x 10 18 n/cm 2 . This evaluation lead to the following conclusions: 1) The measured 30 ft-Jb shift il1 transition temperature values of the lower shell plate B2803-3 contained in capsule X (longitudinal & transverse) are greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift values are less than the two sigma allowance by Regulatory Guide 1.99., Revision 2. 2) The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule X is less than the Regulatory Guide 1.99, Revision 2, prediction. 3) The measured 30 ft-lb shift in transition temperature value of the intermediate shell plate B2802-2 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revision 2, prediction. However, the shift value is less than the two sigma allowance by Regulatory Guide J .99, Revision 2. 4) The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Indian Point Unit 3 surveillance program are in good agreement with the Regulatory Guide 1.99, Revision 2 predictions. 5) An beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the life of the vessel (27.J EFPy) as required by 10CFR50, Appendix G [2]. 6) The Indian Point Unit 3 surveillance data from the lower shell plate B2803-3 was found to be credible. This evaluation can be found inAppendix D. Lastly, a brief summary ofthe Charpy V-notch testing can be found in Section 1. All Charpy V-notch data was plotted using a symmetric hyperbolic tangent curve fitting program.
1- J 1
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule X, the fourth capsule removed and tested from the Indian Point Unit 3 reactor pressure vessel, led to the following conclusions: I) The Charpy V-notch data presented in WCAP-8475[31, WCAP-9491 [4], WCAP-] 0300[5], and WCAP- J 1815[6] were based on hand-fit Charpy curves using engineering judgment. However, the results presented in this report are based on a re-plot of all applicable capsule data using CVGRAPH, Version 5.0.2, which is a byperbolic tangent curve-fitting program. Appendix C presents the CVGRAPH, Version 5.02, Charpy V-notch plots and the program input data. Capsule X received an average fast neutron fluence (E> 1.0 MeV) of 0.874 x 10 19 n/cm" after 15.5 effective full power years (EFPY) of plant operation. Irradiation of the reactor vessel lower shell plate B2803-3 Charpy specimens, oriented with the longitudinal axis oftbe specimen parallel to tbe major working direction (longitudinal orientation), resulted in an irradiated 30 ft-Ib transition temperature of 191.6°F and an irradiated 50*ft-Ib transition temperature of 223. 8°F. This results in a 30 ft-lb transition temperature increase of 159.6°F and a 50 ft-Ib transition temperature increase of 161.7°F for the longitudinal oriented specimens. See Table 5-9. Irradiation of the reactor vessel lower shell plate B2803-3 Charpy specimens, oriented with the longitudinal axis ofthe specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-Ib transition temperature of216.5°F and an ilTadiated 50 ft-Ib transition temperature of 327.4°F. This results in a 30 ft-Ib transition temperature increase of 158.2°F and a 50 ft-lb transition temperature increase of2l7.9°F for the longitudinal oriented specimens. See Table 5-9. Irradiation of the weld metal (heat number W52J4) Charpy specimens resulted in an irradiated 30 ft-Ib transition temperature of 128.5°F and an irradiated 50 ft-Ib transition temperature of 196.8°F. This results in a 30 ft-Ib transition temperature increase of 193.2°F and a 50 ft-Ib transition temperature increase of 242.8°F. See Table 5-9. Irradiation of the reactor vessel intermediate shell plate B2802-2 Charpy specimens, oriented with the longitudinal axis ofthe specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 98.1 OF and an irradiated 50 ft-lb transition temperature of 145.0°F. This results in a 30 ft-Ib transition temperature increase of 152.6°F and a 50 ft-lb transition temperature increase of 166.5°F for the longitudinal oriented specimens. See Table 5-9. The average upper shelf energy of the lower shell plate B2803-3 (longitudinal orientation) resulted in an average energy decrease of 24 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 81 ft-Ib for the longitudinal oriented specimens. See Table 5-9. Summary of Results
1-2 The average upper shelf energy of the lower shell plate B2803-3 (transverse orientation) resulted in an average energy decrease of 16 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 52 ft-Ib for the longitudinal oriented specimens. See Table 5-9. The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 46 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 74 ft-Ib for the weld metal specimens. See Table 5-9. The average upper shelf energy of the intermediate shell plate B2802-2 (longitudinal orientation) resulted in an average energy decrease of 20 ft-Jb after irradiation. This results in an irradiated average upper shelf energy of 105 ft-Ib for the longitudinal oriented specimens. See Table 5-9. A comparison, as presented in Table 5-10, of the Indian Point Unit 3 reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2fJ ] predictions led to the following conclusions: The measured 30 ft-lb s}rift in transition temperature values of the lower shell plate B2803-3 contained in capsule X (longitudinal & transverse) are greater than the Regulatory Guide 1.99, Revision 2, predictions. However, each shift value is less than the two sigma allowance by Regulatory Guide 1.99, Revision 2. The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule X is less than the Regulatory Guide 1.99, Revision 2, prediction. The measured 30 ft-lb shift in transition temperature values of the intermediate shell plate B2802-2 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revision 2, prediction. However, the shift value is Jess than the two sigma allowance by Regulatory Guide 1.99, Revision 2. The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Indian Point Unit 3 surveillance program are in good agreement with the Regulatory Guide 1.99, Revision 2 predictions. All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the life ofthevessel (27.1 EFPy) as required by I OCFR50, Appendix G [2]. SUlmnary of Results
2-1 2 INTRODUCTION This repOli presents the results of the examination of Capsule X, the fourth capsule removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the hldian Point Unit 3 reactor pressure vessel materials under actual operating conditions. The surveillance program for the Indian Point Unit 3 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties ofthe reactor vessel materials are presented in WCAP-8475, "Consolidated Edison Co. of New York Indian Point Unit No.3 Reactor Vessel Radiation Surveillance Program,,[31. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-62, "RecOlmnended Practice for Surveillance Tests on Structural Materials for Nuclear Reactors." Capsule X was removed from the reactor after 15.5 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed. This report summarizes the testing of and the post-irradiation data obtained from surveillance capsule X removed from the Indian Point Unit 3 reactor vessel and discusses the analysis of the data. Introduction
4-3 Table 4-1 Chemical Composition (wt%) of the Indian Point Unit 3 Reactor Vessel Sunreillance Materials (Unirradiated) (a) Intermediate Shell Plate Lower Shell Plate Element B2802-1 B2802-2 B2802-3 B2803-3 Weld Metal (b) C 0.22 0.19 0.20 0.22 0.08 Mn 1.41 1.33 1.32 1.30 1.18 P 0.010 0.015 0.011 0.012 0.019 S 0.023 0.019 0.025 0.024 0.016 Si 0.28 0.21 0.26 0.28 0.17 Ni 0.50 0.53 0.49 0.52 1.02 (1.21 )(C) Cr 0.08 0.09 0.08 0.08 0.04 Mo 0.46 0.48 0.50 0.45 0.53 Cu 0.18 0.20 0.19 0.24 0.15 (0.166)(0) Al 0.036 0.027 0.042 0.03 <0.01 V <0.01 <0.01 <0.01 <0.01 <0.01 Sn 0.014 0.017 0.014 <0.01 0.007 Cb <0.01 <0.01 <0.01 <0.01 <0.01 Zr <0.01 <0.01 <0.01 <0.01 <0.01 Ti <0.01 <0.01 <0.01 <0.01 <0.01 Notes: (a) Data obtained from WCAP-11815 and duplicated herein for completeness. (b) Weld wire Heat Number W5214, Flux Type Linde 1092, and Flux Lot Number 3692. Surveillance weldment has the same heat and flux as the nozzle shell longitudinal weld seams J-042A, B & c. (c) Results of chemical analysis performed on in-adiated Charpy V-notch Specimen \11'-15 from Capsule Y. e Description of Program
5-4 The average upper shelf energy of the intermediate shell plate B2802-2 (longitudinal orientation) resulted in an average enerb'Y decrease of20 ft-Ib after irradiation. This results in an irradiated average upper shelf energy of I 05 ft-lb for the longitudinal oriented .specimens. See Table 5-9. A comparison, as presented in Table 5-10, of the Indian Point Unit 3 reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision ill predictions led to the folJowing conclusions: The measured 30 ft-Ib shift in transition temperature values ofthc lower shell plate B2803-3 contained in capsule X (longitudinal & transverse) are greater than the Regulatory Guide 1.99, Revision 2, predictions. However, each shift value is less than the two sigma allowance by Regulatory Guide 1.99, Revision 2. The measured 30 ft-lb shifl in transition t.emperature value of the weld metal contained in capsule X is less than the Regulatory Guide 1.99, Revision 2, predictions The measured 30 ft-lb shift in transition temperature values of the intermediate shell plate B2802-2 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revisi.on 2, prediction. However, the shift value is less than the two sigma allowance by Regulatory Guide 1.99, Revision 2. The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Indian Point Unit 3 surveillance program are in good agreement with the Regulatory Guide 1.99, Revision 2 predictions. All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-Jb throughout the extended life of the vessel (27.1 EFPy) as required by I OCFR50, Appendix G [21. The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule X materials is shown in Figures 5-13 through 5-16 and shows an increasingly ductile or tougher appearance with increasing test temperature. The load-time recbrds for individual instrumented Charpy specimen tests are shown in Appendix B. The CharpyV-notch data presented in WCAP-8475[31, WCAP-949 I [41, WCAP-I0300[51, and WCAP-I] 815[6] were based on hand-fit Charpy curves using engineering judgment. However, the results presented in this report are based on a re-plot of all applicable capsule data using CVGRAPH, Version 5.0.2, which is a hyperbolic tangent curve-fitting program. This report also shows the composite plots that show the results from the previous capSUle. Appendix C presents the CVGRAPH, Version 5.02, Charpy V-notch plots and the program input data. Testing of Specimens fi'om Capsule X
5-8 Table 5-3 Charpy V-notch Data for the Indian Point Unit 3 Surveillance Weld Metal Irradiated to a Fluence of 0.874 x 10 19 nlcm2 (E> 1.0 MeV) Sample Temperature Impact Energy Lateral Expansion Shear Number OF °C ft-Ibs Joules mils mm % W42 75 24 9 12 5 0.13 20 W4] ]25 52 49 66 36 0.9] 50 W43 125 52 24 33 19 0.48 40 W48 150 66 -)~ 47 26 0.66 45 W47 200 93 37 50 30 0.76 70 W44 250 121 67 91 52 1.32 95 W45 300 149 72 98 56 1.42 98 W46 350 ]77 75 ]02 57 1.45 100 Testing of Specimens fi'om Capsule X
5-12 Table 5-7 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Surveillance Weld Metal Irradiated to a Fluence of 0.874 x 1019 n/cm2 (E>1.0 MeV) Charpy Normalized Energies Yield Time to Time to Fast Energy ~ft-Ib/in2) Test Load Yicld Max. Max. !I<'ract. Arrest Yicld Flow Eo Sample Temp. Charpy Max. .P1"Op. rC;Y tGY Load t", Load .PI'" Load Stress Stress No. (OF) (ft-Ib) Eo/A E~iA Ep/A (lb) (msec) .PM (lb) (msec) (I b) P" (Ib) cry (ksi) (ksi) W42 75 9 73 36 36 3426 0.14 3696 0.16 3687 0 J 14 119 W41 .125 49 395 226 169 341 ] 0.15 4363 0.52 4288 617 114 129 W43 125 24 193 68 126 3341 0.14 4109 0.22 4058 1313 111 124 W48 150 35 282 184 98 3416 0.14 4449 0.42 4417 1141 114 131 W47 200 37 298 150 148 3371 0.14 4260 0.37 4222 1713 J 12 127 W44 250 67 540 227 313 3486 0.14 4432 0.50 4251 2819 116 132 I W45 300 72 580 218 362 3329 0.14 4303 0.50 3029 2501 111 ]27 I I W46 350 75 604 221 383 3285 0.14 4309 0.51 n/a n/a 109 126 I
-~ , -- - - - - ------ -~- -----~
Testing of Specimens fi'oll1 Capsule X
5-15 Table 5-10 Comparison of the Indian Point Unit 3 Surveillance Material 30 ft-Ib Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(d) Predicted Measured Predicted Measured (x 1019 n/cm 2, (OF) (oj (oF) (bJ ('Yo) (aJ (%)(C) E> 1.0 MeV) Lmver Shell Plate T 0.263 101.9 139.4 24 12 B2803-3 Z 1.04 161.6 167.8 33.5 22 (Longitudinal) X 0.874 153.9 159.6 32 23 Lower Shell Plate T 0.263 101.9 ]05.9 24 16 B2803-3 Y 0.692 143.5 148.9 30 25 Z 1.04 161.6 157.9 33.5 18 (Transverse) X 0.874 153.9 158.2 32 24 Surveillance T 0.263 131.3 151.6 22 30 Program Y 0.692 185.0 172.0 27 43 Weld Metal Z 1.04 208.3 229.2 31 37 X 0.874 198.4 193.2 29 38 Intermediate Shell Plate B2802-2 X 0.874 146.2 152.6 30 16 (Longitudinal) Notes: (a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel ofthe surveillance material. (b) Calculated using measured Charpy data plotted using CVGRAPH, Version 5.0.2 (See Appendix C) (c) Values are based on the definition of upper shelf energy given in ASTM EI8S-82. (d) The tluence values presented here are the calculated values, not the best estimate values. Testing of Specimens from Capsule X
5-23 SURVEILLANCE
._- ~. -- WELD MATERIAL----- .~- .. -.. -
CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/021200402:31 PM Data Sct(s) Plott.ed Curve Plant Capsule Material Ori. Heat # 1 Indian Point 3 UNIRR SAW NA W5214 2 Indian Point 3 T SAW NA W5214 3 Indian Point 3 Y SAW NA W5214 4 Indian Point 3 Z SAW NA W5214 5 lndian Point 3 X SAW NA W5214
]'i" 200 .. ---+-_....- '0 af e> 150 _. ---+....- -.......- -....- - i r - - - - j -
QI C W o Z o (; 100 r--' 50--' 0, - .-
*300 *200 *100 o 100 200 300 400 Temperature In Deg F o Set 1 o $et2 ¢ Set 3 '" Set 4 Set 5 Results Curve Fluence h<:;E USE doUSE T@30 d*T@30 T@50 doT @50 1 2. 2 120. 0 .0 *64.7 .0 *46.0 .0 2 2.2 84.0 - 36.0 86. 9 151. 6 130.6 176.6 3 2. 2 69.0 *51. 0 107.3 172.0 J64.2 210.2-4 2.2 76. 0 -44.0 164.5 229.2 218. 7 264,7 5 2, 2 74.0 - 46.0 128.5 193.2 196. 8 242. 8
,~----------------------------------------------------------------~----------- Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Testing of Specimens from Capsule X
5-24 SURVElILLANCE
... - .- WELD MATERIAL .... _.- ...
CVGRAPH 5,0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:11 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat # 1 Indian Point 3 UNIRR SAW NA W5214 2 Indian Point 3 T SAW NA W5214 3 Indian Point 3 Y SAW NA W5214 4 Indian Point 3 Z SAW NA W5214 5 Indian Point 3 X SAW NA W5214 200 '-1--'- -.. ---- ....,,--_ _ - ..... .. r*---**** I 150 r - - - - - _..._ - + - _ .. _. - - - - - - ~----.... .._----\
---_.. - - + - - - ...._---- .. _+----_ ...- - _ . -
600 Temperature in Oeg F o Set 1 <> Set 3 b Set 4 Set 5
-_... ~ ..... , -'_..
ReSlllts Curve F1ucnc.e LSE USE I d*USE T@35 doT @35 i I .0 90. 8 .0 -59.3 .0 2 .0 BO. 3 .. 10.5 113.3 172.6 3 .0 65. 6 *25.2 145.0 204. 3 4 .0 73.5 .. 17. 3 187.8 247.1 5 .0 62.1 - 28.7 184.6 243.9 Figure 5-8 Testing of Specimens from Capsule X
5-25 SURVEILLANCE WELD MATERIAL CVGRAPH 5.0.2 Hyperbolic Tangen! Curve Printed on 04/0212004 03:57 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat # 1 Indian Point 3 UNrRR SAW NA W5214 2 Indian Point 3 T SAW NA W5214 3 Indian Point 3 Y SAW NA W5214 4 Indian Point 3 Z SAW NA W5214 5 Indian Point 3 X SAW NA W5214
-+--_. '--~-I 0 100 200 300 400 500 600 Temperature In Ceg F o Set 1 o Set 2 <> Set 3 to. Set 4 Set 5 Results eliI'Ve Fluence LSE USE d-USE T@SO doT @50 I .0 100.0 .0 -47.8 .0 2 .0 100.0 .0 124.0 171. 8 3 .0 100.0 .0 132.6 180.4 4 .0 100.0 .0 147.5 195.3 5 .0 100.0 .0 144.5 192.3 I
.~------------------------------------------------------------~ Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Testing of Specimens from Capsule X
c-o APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Appendix C
UNIRRADIA TED (WELD) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 02:26 PM Page 1 Coefficients of Curve 1 A =61.1 B =58..9 C =47.03 TO::;: -37.11 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy;:: 120.0(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=-64.7 Deg F Temp@50 ft-lbs=-46.0 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: UNIRR Fluence: nlcmll2 300 - - -,.--,----, 250
~ 200 -...... -~..-+--- +J o ~
E> 150
~
w z {; 100 '-- 50 - - - t - -....-
-1---*....,-+-+--+-_.......-- -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential
-150.00 5. 00 3. 16 l. S4 -150.00 2.00 3. 16 - 1 . 16 -150.00 4.50 3. 16 1. 34 -100.00 29. 00 9. SO 19. 2 a -100.00 IS. 00 9. 8a 8. 20
100.00 25.50 9. 8a 15. 70
- 50. 00 35. 00 45. 35 -10.35 -50.00 33. 00 45. 35 .. 12. 35 -50.00 32. 50 45. 35 .. 12. 85 C-45
UNIRRADIATED (WELD) Page 2 Plant: Indian Point 3 Materia1: SAW Heat: W5214 Orientation: NA Capsule: UNlRR Fluence: n/cm"2 Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
-35.00 78. 00 63. 74 14. 26 . 35. 00 69. 50 63. 74 5. 76 -35.00 54. 50 63. 74 - 9. 24 -20.00 87. 00 81. 63 5. 37 *20.00 82. 00 8 I . 63 . 37 *20.00 89. 00 81. 63 7. 37
10. 00 100.00 106.00 - 6. 00 10.00 105.00 106.00 - 1. 00 10.00 113.50 106.00 7. 50 60.00 1 15. 00 I I 8. 14 - 3. 14 60.00 1 19. 00 118.14 . 86 60.00 121.50 118.14 3. 36 160.00 124.00 119,97 4. 03 160.00 125.00 119.97 5.03 160.00 112.00 119, 97 - 7. 97 Correlation Coefficient ;;0 .981 C-46
CAPSULE T (WELD) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/200402:26 PM Page 1 Coefficients of Curve 2 A:;; 43.1 B = 40.9 C:;; 87.09 TO = 115.75 D = O.OOE+OO Equation is A + B * [Tanh<<T-To)/(C+DT))] Upper Shelf Energy",84.0(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30ft-lbs::=86.9 Deg F Temp@50 ft-lbs=130.6 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: T Fluence: nlcm"2 300 ..,-------r-- ---,-----,----_..-_... - - - - , 250*-*-----+------~- --~----+------ - ...--- ------_ . -
~ 200 +---+----+ --1---+----+---.--+-.---+---:+----1 '0 o
u.. e; 150 - - -........_.. -.-..-----1---+---- ........... .---j----+--
~
w z .............. --1----1----- J._.........-j------+---- ..- ...... . ......_ - - () 100
~~--~----- ---------- ----------- -----------
p' .......... 50 -j----+----+ ...... - .., --l----+rl..,r:;./,.!-IOL--I-- ,1i
. . .-6' o ---- ----- ----- . ---- -----.;:..:::-::.--1---..;---1---1---1-- -300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential .00 ! 3. 00 7.56 5.44 70.00 17. 50 23. 39 .. 5. 89 I 10. 00 48.00 40.40 7. 60 150.00 55. 50 58. 40 .. 2. 90 150.00 53. 00 58. 40 - 5. 40 165. 00 66.00 64.04 1. 96 210.00 78. 00 75. 58 2.42 300. 00 90. 50 82. 83 7. 67 Correlation Coefficient = .979
CAPSULE Y (WELD) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 02:26 PM Page I Coefficients of Curve 3 A ::: 35.6 B = 33.4 C = 90.16 TO = 122.54 D = O.OOE+OO Equation is A + B * [Tanh<<T-To)/(C+DT>>} Upper Shelf Energy:=69.0(Fixed) Lower Shelf Energy:::2.2(Fixed) Temp@30 ft-lbs==107.3 Deg F Temp@50 ft-lbs=164.2 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Y Fluence: n/cm A 2 300 ,-.-.. - . . , - - - - - . - r - - - -..., - - - , - . -......- - - , - - - - - ..- . . . , - - - - . . , - - - . - - , - - -.. - 250 I---- ***..* I - - - -..t - - - t - *..- - - - I - - ..-+----- . _ - - + -.. -.--+-_.....-
~ 200 .. '0 o - - f - - . - .---1----.- -+-----. -.--+--- ' - - - _ . " -r---::- . - - - - 1 u..
E> 150 .. _--+-- --+--- ... --+---- . - . - - - - f - - - - " ... -.----'1---.-.-1----1
~
w
z
($ 100 +----- - - - 1 . - . - - - - - + - - - ...- - - ! - - -..... - - -....... _ - - j - - - - - - f - - - . -... - o ..."........................"....-..-......................_....+. .....
"--1---
r - - - - I - - - j .... - - -....... + - - - ; - - , ..- j - - -........ - -
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Oeg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 25.00 20.00 9. 08 10. 92
72. 00 19. 50 18. 62 . 88 125.00 31. 00 36. 51 . 5. 51 125.00 29. 50 36. 51 - 7. 01 150.00 49.00 45. 47 3. 53 200. 00 67. 50 58. 84 8. 66 300.00 69. 50 67. 72 1. 78 400.00 68. 50 68. 86 - . 36 Correlation Coefficient == .960
CAPSULE Z (WELD) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/021200402:26 PM Page 1 Coefficients of Curve 4 A;:;; 39.1 B:: 36.9 C = 97.52 TO;;; 188.96 D;;; O.OOE+OO Equation is A + B * [Tanh<<T*To)/(C+DT))] Upper Shelf Energy=76.0(Fixed) Lower Shelf Encrgy=2.2(Fixed) Temp@30ft-lbs=164.5 Deg F Tcmp@50ft-lbs=218.7 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Z Fluence: nfcm"2 250 .!----I-----~e__- - . - ; - - - - - -..........- - f - - - -... -+------j-- - - _ - - j - _ .... _ 111
=F 200 . - - - - +.....- .. ..--t----.- +----+-- ......- - - j - - - -.. .. -------4 '0 If 21150 - - - - - I ....- - - - ... _-+--- ..
t - - - + _ ------1----. ..' ..
(I) C W Z
~ 100 ./--_. -.... .-... -.-.--!--.-....... --+---. ...--- ---f--...... ..- /:'6' ' ._.~._.~ . * . -.... ' . ' . ' . ' . ' . ' 0 ' * *' * . * * * * * * * * *
50 1--.-.--+---- . I "'~: -+----.... .- .-
o .J=:.~c~';- .'.'.. ........ , .......... -.::.L:.~ j - -.. I . ..--1-_-1--_.. '" I I
~300 u200 -100 0 100 200 300 400 500 GOO Temperature in Ceg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 100. 00 10.00 12 . 45 *2. 45 150.00 2 I .00 25. 10 - 4. 10 150.00 44.00 25. 10 18. 90 175. 00 26.00 33. 85 . 7. 85 200.00 33. 00 43.26 .10.26 225.00 57.00 52. 15 4. 85 300.00 75. 00 69. 14 5. 86 400.00 77. 00 75. 04 1.96 Correlation Coefficient = .929 C*57
CAPSULE X (WELD) CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 02:36 PM Page 1 Coefficients of Curve 5 A =38.1 B:; 35.9 C::: 118.98 TO == 155.76 D == O.OOE+OO Equatjon is A + B * [Tanh<<T-To)/(C+DT))l Upper Shelf Energy==74.0(Fixed) Lower Shelf Energy:=2.2(fixed) Temp@30 ft-Jbs=128.5 Deg F Temp@50 fHbs=196.8 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: X Fluence: n/cm"2 300,--..........,- ... -~-..- - - -.... " - - ' - - ..... "~------'-i _ ......- - , - - - - - , 250
l - - - - f * . _--+-..._ - -.-... --1----- - - - - .. - - - -
II)
~ 200 1-- - - + - - ... - - - ..- + - - -+---_.... . . . - - - .... -~+-.- '0o u.
21 150 - ' " '" - -... -+--- +---_. "'+--_. "'--". - - -..... _ - (!) c w z
~ 100 f - - - - .. " " " ' - _ ......- - -.... _ - + .._ - - - l - - - ----1--.- --+-- - -......-
50 I - - ...- --1--_ .........- + - - - - - o .--
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Oeg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential 7 5. 00 9.00 16. 89 - 7. 89 125.00 49. 00 29. 02 19. 98 125. 00 24.00 29.02 .. 5.02 150.00 35.00 36. 36 .. 1. 36 200.00 37.00 50. 87 -13.87 250. 00 67. 00 61. 78 5.22 300.00 72.00 68. 16 3. 84 350.00 75. 00 71. 36 3.64 Correlation Coefficient =.906 C-60
APPENDIXF H. B. ROBINSON 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM RESULTS FOR WELD NO. W5214 Report No. 0901132.401, Rev. 0 F-l ,~t.iflil'~tllr'f!illntegrity Associates, Inc.
"*;-)prl.f.'ti.,ry l .
~ ('/ I'- ('/ I'-
~
'1 u
3 i'
CAROLINA POWER AND LIGHT CO. H. B. ROBINSON UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM By S. E. Yanichko January 1970 Work Performed Under CPI.-106
/:' ' /':) U "-..,.-'
Approved: E. Landerman Westinghouse Electric Corporation Nuclear Energy Systems Box 355 Pittsburgh, Pennsylvania 15230
TABLE 4 PRE-IRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE CPL H. B. ROBINSON UNIT NO. 2 REACTOR PRESSURE VESSEL WELD METAL AND WELD HEAT AFFECTED ZONE METAL Weld ~leL~ Heat Affected Zone Test Lateral Test Lateral Specimen Temp Energy Shear Expansion Speciinen Temp Energy Shear Expansion
. (ft-1b) (0 F) ~ - -No.-- (OF) (%) ___(mils)_ No. Qt-lb) __iJ!111sl... .125 -15019.0 29 16 1128 -150 11.0 10 17 W26 -150 10.0 23 9 1129 -150 J1. 0 18 29 W27 -150 30.0 29 25 IlJO -150 3/,.0 18 30 W34 -150 3.0 9 2 W35 -150 34.5 26 28 1125 -100 39.5 _29 30 W36 -150 2.0 9 2 1126 -100 42.0 37 1127 -100 41 .0 34 30 \0128 -100 38.0 30 34 W29 -100 29.0 23 19 II31 - 50 1,2.5 35 W30 -100 25.0 20 22 1.32 - 50 (,0.0 45 1133 - 50 37.5 32 W31 - 50 21.0 25 20 'W32 - 50 54.5 36 49 H34 - 20 75.0 42 61 W33 - 50 36.5 30 31 1135 - 20 86.0 6l 62 H36 - 20 45.0 34 37 WJ7 10 73.5 64 62 W38 10 68.0 61 58 H37 10 83.0 81 67 W39 10 65.5 59 57 H38 10 119.0 90 81 H39 10 94.0 77 70 W40 60 97.0 90 80 W41 60 99.0 91 80 60 116.0 100 88 \<142 60 116.0 94 88 GO 111. 5 100 87 H42 60 110.0 95 87 W43 110 97.0 95 74 W44 110 104.0 100 85 110 117.0 100 93 W45 100 107.5 98 89 110 140.0 100 83 110 119. 0 100 86 W46 210 112.0 100 90 W47 210 111.0 100 91 H46 210 130.0 100 89 W48 210 115.0 100 83 H47 210 l3 t"O 100 86 H48 210 123.0 100 84 18
1179-5 130 120 100 V) a:l 80 -1 I l-Ll... () 60 0::: UJ Z UJ 0 40 0 0 0 20 0 0 0
-200 -100 o 100 200 300 TEMPERATURE (OF)
Figure 10. Pre-Irradiation Charpy V-Notch Impact Energy for the CPL H. B. Robinson Unit ff2 Reactor Pressure Vessel Weld Metal 32
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15805 Analysis of Capsule X from the Carolina Power & Light Company HoB. Robinson Unit 2 Reactor Vessel Radiation Surveillance Program T. J. Laubham E. P. Lippincott J. Conermann MARCH 2002 Prepared by the Westinghouse Electric Company for the Carolina Power & Light Company Approved: C. H. Boyd, Manager Equipment & Materials Technology Westinghouse Electric Company LLC Nuclear Services Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
© 2002 Westinghouse Electric Company LLC All Rights Reserved Analysis ofH.B. Robinson Unit 2 Capsule X
VllI EXECUTIVE
SUMMARY
The purpose of this report is to docmnent the results of the testing of surveillance capsule X from H.B. Robinson Unit 2. Capsule X was removed at 20.39 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens was performed, along with a fluence evaluation based methodology and nuclear data including recently released neutron transport and dosimetry cross-section libraries derived from the ENDFIB* VI database. The calculated peak clad base/metal vessel fluence after 20,39 EFPY of plant operation was 2,76 x 1019 n/cm2 and the surveillance Capsule X calculated fluence was 4.49 x 10 19 n/cm 2, A brief summary of the Charpy V-notch testing results can be found in Section 1 and the updated capsule removal schedule can be found in Section 7, A supplement to this report is a credibility evaluation, which can be found in Appendix D, that shows the H.B, Robinson Unit 2 surveillance weld data, while including all surveillance data for weld heat W5214, is credible. Of the three surveillance plates, only intermediate shell plate W10201-5 was found to be credibJe. AnalysisofH.B. Robinson Unit 2 CapsuJe X
1-1 1
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance capsule X the fourth capsule to be removed from the H.B, Robinson Unit 2 reactor pressure vessel, led to the following conclusions: (General Note: Temperatures are reported to two significant digits only to match CVGraph output.) e The capsule received an average fast neutron calculated fluence (E> 1.0 MeV) of 4.49 x 10 19 nlcm 2 after 20.39 effective full power years (EFPy) of plant operation. Irradiation of the reactor vessel intermediate shell plate WI0201-4 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation), to 4.49 x 10 19 nlcm 2 (E> 1.0MeV) resulted in a 30 ft-Ib transition temperature increase of 104. 73°P and a 50 ft-Ib transition temperature increase of98.68°P. This results in an irradiated 30 ft-Ib transition temperature of 86.55°P and an irradiated 50 ft-lb transition temperature of 116.04°P for the longitudinally oriented specimens Irradiation of the weld metal Charpy specimens to 4.49 x 10 19 n/cm2 (E> l.OMeV) resulted in a 30 ft-lb transition temperature increase of 265.93°P and a 50 ft-lb transition temperature increase of 251. 74°P. This results in an irradiated 30 ft-Ib transition temperature of 179.64°P and an irradiated 50 ft-lb transition temperature of 211.38°F. Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 4.49 x 1019 nlcm 2 (E > 1.0 MeV) resulted in a 30 ft-Ib transition temperature increase of 210. I3 0 P and a 50 ft*lb transition temperature increase of216.59°P. This results in an irradiated 30 ft-lb transition temperature of l00.47°F and an irradiated 50 ft-Ib transition temperature of 150.54°P. Irradiation of the correlation monitor material Charpy specimens to 4.49 x 1019 nlcm 2 (E > 1.0 MeV) resulted in a 30 ft-Ib transition temperature increase of 125.21 OF which resulted in an irradiated 30 ft-lb transition temperature of 188.1SoP. The tested specimens did not reach the 50 ft-Ib transition temperature. The average upper shelf energy ofllie intermediate shell plate WI0201-4 (longitudinal orientation) resulted in an average energy decrease of 1 flAb after irradiation to 4.49 x 10 19 n/cm2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 94 ft-lb for the longitudinally oriented specimens. The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 33 ft-Ib after irradiation to 4.49 x 10 19 nlcm2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 80 ft-lb for the weld metal specimens. The average upper shelf energy of the weld HAl metal Charpy specimens resulted in an average energy decrease of 24 ft-lb after irradiation to 4.49 x 10 19 nlcm2 (E> 1,0 MeV). Hence, this results in an irradiated average upper shelf energy of 105 ft-lb for the weld HAZ metal. Analysis~of H.B. Robinson Unit 2 Capsule X
4-4 Table 4-1 Chemical Composition (wt %) and Heat Treatment of Material for the H.B. Robinson Unit 2 Reactor Vessel Surveillance Material(a) Chemical Composition Element Plate W 1020 1-4 Plate WI0201-S Plate WI0201-6 Weld Metal Correlation Monitor Material C 0.19 0.20 0.19 0.16 0.24 Mn 1.35 1.29 1.32 0.98 1.34 P 0007 0.010 0.010 0.021 0.011 S 0.019 0.021 0.0]5 0.014 0.023 Si 0.23 0.22 0.19 0.34 0.23 Mo 0.48 0.46 0.49 0.46 0.51 Cu 0.12 0.10 0.09 0.34 0.20 V --- -- - --- 0.001 -- - Ni --- --- --- 0.66 0.18 Cr --- --- . -- 0.024 0.11 Co --- --- --- --- --- Heat Treatment Plate WI0201-4, 1550°F to 1600°F, 4 hours, Water Quench Plate WI0201-5,'& 1200°F to J250 oP, 4 hours, Air Cooled Plate WI0201-6 1125°F to 1 I75°F, 15 1/2 hours, Furnace cooled to 60Q o F Weld Metal 1125"F to J J75"F, 30 hours, Furnace cooled to 600"F t Correlation Monitor I 650°F, 4 hours, Water Quenched I200°F - 6 hours, Air Cooled Notes: a) The data given in this column (originally) is from WCAP-7373 & WCAP-J0304. AnalysisofH.B. Robinson Unit 2 Capsule X
5-3 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in capsule X, which received a fluence of 4.49 x 1019 nlcm2 (E > 1.0 MeV) in 20.39 EFPY of operation, are presented in Tables 5-1 through 5-8, and are compared with unirradiated results as shown in Figures 5*1 through 5-12. The transition temperature increases and upper shelf energy decreases for the capsule X materials are summarized in Table 5-9. These results led to the following conclusions: Irradiation of the reactor vessel intermediate shell plate WI 020 1-4 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation), to 4.49 x 1019 nJcm 2 (E> l.OMeV) resulted in a 30 ft-Ib transition temperature increase of 104.73°F and a 50 ft-Ib transition temperature increase of 98.68°F. This results in an irradiated 30 ft-Ib transition temperature of 86.55°F and an irradiated 50 ft-Jb transition temperature of 116.04°F for the longitudinally oriented specimens Irradiation of the weld metal Charpy specimens to 4.49 x 10 19 nJcm2 (E> l.OMeV) resulted in a 30 ft-Ib transition temperature increase of265.93°F and a 50 ft-Ib transition temperature increase of 251.74°F. This results in an irradiated 30 ft*lb transition temperature of 179.64°F and an irradiated 50 ft-Ib transition temperature of211.38°F. Irradiation of the weld Heat-Mected-Zone (HAZ) metal Charpy specimens to 4.49 x 10 19 n/cm 2 (E > 1.0 MeV) resulted in a 30 ft-Ib transition temperature increase of210.13°F and a 50 ft-lb transition temperature increase of 216.59°F. This results in an irradiated 30 ft-Ib transition temperature of . 100.47°F and an irradiated 50 ft-Ib transition temperature of 150.54°F. Irradiation of the correlation monitor material Charpy specimens to 4.49 x 10 19 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 125.21 OF which resulted in an irradiated 30 ft-Ib transition temperature of 188 .1soF. The tested specimens did not reach the 50 ft-Ib transition temperature. The average upper shelf energy of the intermediate shell plate WI 020 1-4 (longitudinal orientation) resulted in an average energy decrease of 1 ft-Ib after irradiation to 4.49 x 10 19 n/cm2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of94 ft-Ib for the longitudinally oriented specimens. The average upper shelf energy of tlie weld metaJ Charpy specimens resulted in an average energy decrease of 33 ft-lb after irradiation to 4.49 x 10 19 n/cm2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 80 ft-Ib for the weld metal specimens. The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 24 ft-lb after irradiation to 4.49 x 1019 n/cm2 (E > 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 105 ft-Ib for the weld HAZ metal. The average upper shelf energy of the correlation monitor material Charpy specimens resulted in no energy decrease after irradiation to 4.49 x 10 J9 n/cm2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 42 ft-Ib for the correlation monitor material. Analysis ofH.B. Robinson Unit 2 Capsule X
5-4 A comparison of the H.B. Robinson Unit 2 reactor vessel beltline material test results with the Regulatory Guide 1.99, Revision i!J, predictions led to the following conclusions: The measured 30 ft-Ib shift in transition temperature values of the intermediate shell plate Wi 0201-4 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2. The measured 30 ft*lb shift in transition temperature values of the weld metal contained in capsule X (longitudinal) is less than the Regulatory Guide 1.99, Revision 2, predictions. The measured percent decrease in upper shelf energy of the Capsule X surveillance material is less than the Regulatory Gtride 1.99, Revision 2, predictions. The fracture appearance of each irradiated Charpy specimen from the various surveillance capsule X materials is shown in Figw-es 5-13 through 5-16 and show an increasingly ductile or tougher appearance with increasing test temperature. The load-time records for individual instrumented Charpy specimen tests are shown in Appendix A The Charpy V-notch data presented in this report is based on a re-plot of all capsule data using CVGRAPH, Version 4.1, which is a hyperbolic tangent curve-fitting program. Hence, Appendix C contains a comparison of the Charpy V-notch shift results for each surveillance material (hand-fitting versus hyperbolic tangent curve-fitting). Additionally, Appendix B presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data. Analysis of H.B. Robinson Unit 2 Capsule X
5*7 Table 5-2 Charpy V-notch Data for the H.B. Robinson Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 4.49 x 1019 nfcm2 (E> 1.0 MeV) Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm % W3 0 -18 4 5 0 0.00 0 W2 100 38 14 19 4 0.10 15 W6 175 79 28 38 16 0.41 35 W4 200 93 38 52 22 0.56 40 W8 250 121 74 100 49 1.24 100 W7 350 177 78 106 51 1.30 100 W5 375 191 85 115 56 1.42 100 WI 425 218 82 III 54 1.37 100 Analysis 9fH.B. Robinson Unit 2 Capsule X
5*15 Table 5-10 Comparison of the H.B. Robinson Unit 2 Surveillance Material ':;0 ft-Ib Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured 19 (x 10 nlcm ) 1 (OF) (II) (OF) (b) (%)(0) (%t) Inter. Shell Plate S 0.479 45.39 32.51 18 10 WI0201-4 X 4.49 78.86 104.73 30 I (Longitudinal) Surveinance V 0.530 179.17 209.32 39 38 Program T 3.87 293.68 288.15 52 46 Weld Metal X 4.49 300.64 265.93 54 29 Heat Affected V 0.530 *. 59.21 -. - 26 Zone Material T 3.87 -. (d) --- 24 X 4.49 *. 210.13 -- - 19 Correlation S 0.479 -. 72.79 . .- 3 Monitor Material V 0.530 *. 69.39 -- . 5 T 3.87 *. 156.83 . .. 5 X 4.49 -- 125.21 --- 0 Notes: (a) Based orr: Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material. (b) Calculated using measured Charpy data plotted using CVGRAPH, Version 4. J (See Appendix B) (c) Values are based on the definition of upper shelf energy given in ASTM E 185-82. (d) Only 2 specimens were tested from capsule T to confirm the upper shelf energy, thus, there is insufficient data to detennine the measured 30 ft-Ib shift. Analysis ofH.B. Robinson Unit 2 Capsule X
5*20 SURVELLIANCE PROGRAM WELD MATERIAL CVGRAPH 4.1 Hyperbolic Tangent Curve Prinled at 00:54:57 on ]()-24-2001 Results Curve FJuence LSE d-LSE USE d-USE T I) 30 d-T I) 30 'I' 6 50 d-T 0 50 1 0 2.19 0 113 0 -8629 0 -40.35 0 2 0 2.19 () 70 -~3 123.02, 209.32 21t59 254.94 3 0 22 0 6l -52 201.86 2B8.l5 205.94 246.3 0 2.19 0 80 -3,1 179.64 265.93 211.38 251.74 300 250 200 150 0 l<=l h-100 u 5u 0/ 7 ....y"!
&.c..... ~-I o£,," __.:;;,
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L>. B ~ o <;?..... 9--.....~ <0/ I ~
~ ---_._-
0
,-_.,..-- ~~/00 . /
0 I I
-300 -200 -100 0 100 200 300 400 500 600 Temperature In Degrees F Curve tegenl!
1D 20**_****** 3~-- 4'" Data Set(s} Plotted Curve Planl Capsule Malerial Ori. Heali 1 HB2 UNrRll WELD N/A W5214 2 Hffi V YiELD N/A Wn214 3 HB2 T WELD N/A l'f5214 4 HB2 X WELD N/A 1'15214 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Surveillance Weld Metal Analysis ofH.B. Robinson Unit 2 Capsule X
7- J 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The fonowing surveillance capsule removal schedule meets the intent of ASTM E18S-82 and is recommended for future capsules to be removed from the H.B. Robinson Unit 2 reactor vessel. This recommended removal schedule is applicable to 29 EFPY of operation. TABLE 7-1 H.B. Robinson Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluence Capsule Location Lead Factor(') (EFPy)(b) (n/cm', E> 1.0 MeVia) S 280 0 1.90 1.28 4.79 x 10 18 (c) V 290 0 0.91 3.18 5.30 x 10 18 (c) T 270* 2.80 7,27 3.87 x 10 19 (c) X 50° 1.63 20.39 4.49 x 10 19 (c) U(l) 30* 1.41 (2.02) 29.8 6.00 x 1019 Cd) Y 150* 0.92 (1.04) Standby (e) W 40 0 0.59 (0.61) Standby (e) ZW 230* 0.59 (0,61) Standby (e) Notes: (8.) Updated in Capsule X dosimetry analysis. Lead Factor in Parentheses are for Future Cycles. (b) Effective FuIl Power Years (EFPY) from plant startup. (c) Plant specific evaluation. (d) Capsule U will reach a fluence of approximately 6.00 x lO l9 (50 EFPY Peak Fluence) at approximately 29.8 EFPY. Thus, it should pe pulled at the closest outage to 29.8 EFPY. (e) If further material data is desired, then it is reconunended that these capsules be moved to a higher lead factor location and then removed once their accumulated neutron fluence equals the license renewal (50 EFPy) fluence on the vessel inner surface. (f) Moved to Capsule "S" Location (280°) at Cycle 8, eg) Capsule Z was inadvertently removed from the H.B. Robinson 2 Reactor VesseL At this time jt is unconfumed that Capsule Z was re-installed into the vessel or placed in the spent fuel pool. Analysi~ ofH.B. Robinson Unit 2 Capsule X
4-4 Table 4-1 Chemical Composition (wi %) and Heat Treatment of Material for the H.B. Robinson Unit 2 Reactor Vessel Surveillance Material(a) Chemical Composition Element Plate WI0201-4 Plate WI0201.S Plate WI0201-6 Weld Metal Correlation Monitor Material C 0.19 0.20 0.19 0.16 0.24 Mn 1.35 1.29 1.32 0.98 1.34 P 0007 0.010 0.010 0.02J 0.01 I S 0.019 0.021 0.015 0.014 0.023 Si 0.23 0.22 0.19 0.34 0.23 Mo 0.48 0.46 0.49 0.46 0.51 Cu 0.12 0.10 0.09 0.34 0.20 V - -- --- -.. 0.001 --- Ni - .. * .. * .. 0.66 0.J8 Cr . -- * -. *.- 0.024 0.11 Co -. - *.- * .. . -. -- . Heat Treatment Plate WI0201-4, 1550°F to 1600°F, 4 hours, Water Quench Plate WI0201.S, & 1200°F to 1250°F, 4 bours, Air Cooled Plate WI0201*6 112S"F to 1175oF, 1S 112 bours, Furnace cooled to 600"F Weld Metal 1125°F to 1175°F, 30 hours, Furnace cooled to 600°F Correlation Monitor 1650°F, 4 hours, Water Quenched 1200°F - 6 hours, Air Cooled Notes: a) The data given in this column (originally) is from WCAP*7373 & WCAP*I0304. Analysis ofH.B. Robinson Unit 2 Capsule X
5-7 Table 5-2 Charpy V-notch Data for the H.B. Robinson Unit 2 Surveillance Weld Metal Irradiated to It Fluente of 4.49 x 1019 n1cm 2 (E> 1.0 Me V) Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm % W3 0 -18 4 5 0 0.00 0 W2 100 38 14 19 4 0.10 15 W6 175 79 28 38 16 0.41 35 W4 200 93 38 52 22 0.56 40 W8 250 121 74 100 49 1.24 100 W7 350 177 78 106 51 1.30 100 W5 375 191 85 115 56 1.42 "l00 WI 425 218 82 III 54 1.37 100 Analysis ofH.B. Robinson Unit 2 Capsule X
5- J5 Table 5-10 Comparison of the H.B. Robinson Unit 2. Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (x 1019 nlcm 2 ) (OF) (a) (OF) (b) (%) (a) (%)<<) Inter. Shell Plate S 0.479 45.39 32.51 18 10 WI0201-4 X 4.49 78.86 104.73 30 1 (Longitudinal) Surveillance V 0.530 179.17 209.32 39 38 Program T 3.87 293.68 288.15 52 46 Weld Metal X 4.49 300.64 265.93 54 29 Heat Affected V 0.530 -- 59.21 --- 26 Zone Material T 3.87 -- Cd) - -- 24 X 4.49 -. 210.13 --. 19 Correlation S 0.479 -- 72.79 --. 3 Monitor Material V 0.530 -- 69.39 -.. 5 T 3.87 .- 156.83 --. 5 X 4.49 .- 125.21 --- 0 Notes: (a) Based ontReguiatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material. (b) CalCUlated using measured Charpy data plotted using CVGRAPH, Version 4.1 (See Appendix B) (c) Values are based on the definition of upper shelf energy gjven in ASTM EI 85-82. Cd) Only 2 specimens were tested from capsule T to confirm the upper shelf energy, thus, there is insufficient data to determine the measured 30 ft-Ib shill. Analysis ofH.B. Robinson Unit 2 Capsule X
5*20 SURVELLIANCE PROGRAM WELD MATERIAL CVGllAPH 4.1 Hyperbolic Tangent Curve Printed at 09:54:57 on 10-24-2001 ResaJts Curve Fluence lSE d-lSE USE d-USE 1'1>30 d-T 0 30 T Ii 50 d-T 4) 50 1 0 2.19 0 113 0 -8629 0 -40.35 0 2 0 2.19 0 70 -43 123.02 209.32 214.59 ~.94 3 0 2.2 0 61 -52 201.86 288.l5 205.94 246.3 4 0 2J9 0 BO -33 179.64 265.93 211.36 251.74 30a 25)0 200 150 0 Q
!b-1oo IU ~
A A A A/ ._--'-- ------ ----- 0/ 1~:,--*-:a. Jd~ 50 .9- ~ o 9.,- 9-'/~
._--_._- .-..... .-- ..... -- ~:: ,/00 ~.
0 Figure 5-4 Cbarpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Surveillance Weld Metal Analysis ofH.B. Robinson Unit 2 Capsule X
5-21 SURVEILLANCE PROGRAM WELD MATERIAL CVGRAPH 4.1 Hyperbolic Tangent Curve Printed al 09:59:52 on 10-24-2001 Results Curve Fluence USE d-USE T 0 LE35 d-T p LE35 I 0 91.98 0 -60.64 0 2 0 65.7 -2627 1902£ 25ll.9 3 0 41.57 -50.4 204.2 264.65 0 542 -37.77 21924 279.ll9 200 150 100 0 =
~ 0 50 ? 0 Of;7L~" .-o .....t:S,- ------ .------
B l1 CJ I-e"'.....
....0,. ,,:A Y F?~ ,j ~ ,--~,.,.,-,g..,.
U
-300 -200 -100 0 100 200 300 400 500 600 Temperature In Degrees F Curve Legend 10 20-------- 3~-- 4" Data Set(s) Plotted Curve Plant Ca~ule Malerial On. Heali 1 HB2 UNIRR lYEI1J N/ A W5214 2 Hll2 V WELD N/A lf5Z14 3 HB2 T WEIJl NIl, lf5214 HB2 X lfElJl NIp. 1f5214 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Surveillance Weld Metal Analysis ofH.B. Robinson Unil 2 Capsule X
H B Robinson Weld Data PREP-4 Hyperbolic Tangent Curve Printed on 10/19/2009 10:38:40 PM Page 1 Coefficients of Curve 1 A = 59.8 B = 57.6 C = 118.8 TO = -18.8 Equation is A + B * [Tanh ( (T-Tol/C)] Upper Shelf Energy=117.4 Lower Shelf Energy=2.2 Temp.@30 ft-lbs=-86.8 Deg F Temp.@50 ft-lbs=-39.2 Deg F Plant: H B Robinson 2 Material: SAW Heat: W5214 (S) Orientation: TL Capsule: Unirr. Fluence: O. 250 200 til
.c "'i' 0
u. 150 0
e' C1) 0 c (~ 11;;1 w 100 U
z
(.) 50 0/ V 8 0' o
-300 --t -200 -100 0 -J---l-t --I o 100 200 300 400 500 600 Temperature in Deg F Charpy !¥~~otch ,W Data Temperature Input CVN Computed CVN Differential -150.00 30.00 13.60 16.40 -150.00 2.00 13.60 -11. 60 -150.00 3.00 13.60 -10.60 -150.00 10.00 13.60 -3.60 -150.00 19.00 13.60 5.40 -150.00 34.50 13.60 20.90 -100.00 29.00 25.60 3.40
H B Robinson Weld Data Page 2 Plant: H B Robinson 2 Material: SAW Heat: W5214 (3) Orientation: TL Capsule: Unirr. Fluence: O. Charpy V-Notch Data Temperature Input CVN Computed CVN Differential
-100.00 38.00 25.60 12.40 -100.00 25.00 25.60 -.60 -50.00 54.50 45.01 9.49 -50.00 21.00 45.01 -24.01 -50.00 36.50 45.01 -8.51 10.00 73.50 73.50 .00 10.00 65.50 73.50 -8.00 10.00 68.00 73.50 -5.50 60.00 97.00 93.24 3.76 60.00 116.00 93.24 22.76 60.00 99.00 93.24 5.76 100.00 107.50 103.67 3.83 110.00 104.00 105.58 -1. 58 110.00 97.00 105.58 -8.58 210.00 112.00 115.00 -3.00 210.00 111.00 115.00 -4.00 210.00 115.00 115.00 .00 Correlation Coefficient .000
H B Robinson Weld Data PREP-4 Hyperbolic Tangent Curve Printed on 10/19/2009 10:40:29 PM Page 1 Coefficients of Curve 2 A = 31.6 B = 29.4 C = 63.57 TO = 225.94 Equation is A + B * [Tanh((T-To)/C)] Upper Shelf Energy=61.0 Lower Shelf Energy:2.2 Ternp.@30 ft-lbs=222.5 Deg F Ternp.@50 ft-lbs=272.6 Deg F Plant: H B Robinson 2 Material: SAW Heat: W5214 (S) Orientation: TL Capsule: T Fluence: 4.42E+19 250,-----,-----,----,-----,-----,-----,----,-----,-----, 2004------~----r---~----~------+-----r----~----+---~ 1l
.a!.
o o 150+-----+-----+-----+-----+-----+-----+-----+-----+----4 u.. 0') L.. CIl C 100,------r-----r----_r----~----+_----r_--~----_r----~ W Z U o
~~ -~--------- ---------- -----------
50+-----+-----r----4----~-----+--~~~-----r----~-----1 1/
/ ,-6' o - - -- - - -- - - --.-, - -- - - - --~----:_- ----~-~-~'l, 1 1 ---J-+-4-+1~'I-l-I-+1-!-,j--f-I--l- -300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 175.00 14.00 12.06 1. 94 200.00 23.50 20.23 3.27 200.00 17.00 20.23 -3.23 225.00 64.00 31.17 32.83 250.00 38.50 42.22 -3.72 275.00 51.50 50.65 .85 300.00 60.50 55.79 4.71 Correlation Coefficient .000
H B Robinson Weld Data PREP-4 Hyperbolic Tangent Curve Printed on 10/19/2009 10:41:23 PM Page 1 Coefficients of Curv,e 3 A = 34.35 B = 32.15 C = 142.03 TO = 141.21 Equation is A + B * [Tanh((T-To)/C)] Upper Shelf Energy=66.5 Lower Shelf Energy=2.2 Temp.@30 ft-lbs=121.9 Deg F Temp.@50 ft-lbs=216.7 Deg F Plant: H B Robinson 2 Material: SAW Heat: W5214 (S) Orientation: TL Capsule: V Fluence: 6.01E+18 250~----~----~----~----~----~----~----~----~--~ 200
.-(I) .Q 0
0 150 - u..
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o ..........................................
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 30.00 23.50 13.31 10.19 75.00 23.50 20.36 3.14 110.00 30.00 27.40 2.60 160.00 14.00 38.58 -24.58 180.00 44.00 42.92 1. 08 210.00 58.50 48.81 9.69 300.00 72.50 60.29 12.21
H B Robinson Weld Data Page 2 Plant: H B Robinson 2 Material: SAW Heat: W5214 (S) Orientation: TL Capsule: V Fluence: 6.01E+18 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 400.00 68.50 64.86 3.64 Correlation Coefficient .000
APPENDIXG CVGRAPH TANH CURVE-FITS FOR W5214 SURVEILLANCE WELD DATA (from Reference 32) Report No. 0901132.401, Rev. 0 G-l StrlJ'ctulti/ 1!!1ffrmritll Associates, fnco
Table G-l. Fitted Results for CVGRAPH Hyperbolic Tangent Curve-Fits [32] Plant Capsule A B C TO T30 Palisades Unirradiated 53.14 50.94 100.7 -10.76 -60.1 Palisades SA-60-1 28.35 26.15 158.11 188.85 198.9 Palisades SA-240-1 27.35 25.15 111.62 208.13 220 H. B. Robinson 2 Unirradiated 56.05 53.85 107.57 -29.1 -85.8 H. B. Robinson 2 T 31.35 29.15 9.09 203.64 203.3 H. B. Robinson 2 V 36.35 34.15 150.19 151.23 123 H. B. Robinson 2 X 40.97 38.78 59.96 197.18 179.8 Indian Point 2 Unirradiated 59.32 57.12 86.23 -16.54 -65.4 Indian Point 2 V 39.1 36.9 123.56 163.14 132.1 Indian Point 2 Y 34.35 32.15 91.6 140.88 128.5 Indian Point 3 Unirradiated 60.38 58.19 45.25 -37.64 -63.8 Indian Point 3 T 46.35 44.15 98.27 124.16 86 Indian Point 3 Y 35.6 33.4 90.16 122.54 107.3 Indian Point 3 Z 39.1 36.9 97.52 188.96 164.5 Indian Point 3 X 38.6 36.4 121.83 157.96 128.7 Report No. 0901132.401, Rev. 0 G-2 '!>i.fI'U'!"fW!"!lH Eii1!fDtI'!!'!fif Associates, Inc.
Palisades Unirradiated Capsule Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/0912010 06:36 PM Page 1 Coefficients of Curve 1 A =53.14 B =50.94 C =100.7 TO =-10.76 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=104.1 Lower Shelf Energy=2.2(Fixed) Temp@30 ft-Ibs=-60.1 Deg F Temp@50 ft-Ibs=-16.9 Deg F Plant: PALISADES Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: Unirradiat nlcml\2 300 250
~ '0o 200 I.L ~ 150 (I>
W Z 0 0 0
~ 100 >- 0 t~
50 / (
~1310 yf'~ ~
o
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential
-110.00 11 . 80 14. 66 - 2. 86 -110.00 11. 80 14. 66 - 2. 86
-80.00 33. 90 22. 76 1 1. 14 -80.00 20. 60 22. 76 - 2. 16 -80.00 29. 50 22. 76 6. 74 - 40.00 47. 90 38. 75 9. 15 - 40.00 43. 50 38. 75 4. 75 - 40.00 29. 50 38. 75 - 9. 25 - 40.00 41. 29 38. 75 2. 54
Palisades Unirradiated Capsule Report Page 2 Plant: PALISADES Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: Unirradiat n/cmJ\2 Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
.00 64. 19 58. 56 5. 63 .00 39. 09 58. 56 -19.47 20.00 62. 70 68. 24 - 5. 54
20. 00 60. 50 68. 24 - 7. 74

30. 00 78. 19 72. 70 5.49 40.00 61. 20 76. 85 -15.65

60. 00 87. 00 84. 02 2. 98

60. 00 75. 19 84. 02 - 8. 83

60. 00 I l l . 40 84. 02 27. 38 110.00 110.59 95. 60 14. 99 110.00 98. 80 95. 60 3. 20 210.00 110.59 102.83 7. 76 210.00 95. 90 102.83 - 6. 93 300.00 97. 40 103.87 - 6. 47 300.00 94. 40 103.87 - 9. 47 Correlation Coefficient = .947
Palisades SA -60= 1 Capsule Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/25/2010 10:39 AM Page 1 Coefficients of Curve 1 A =28.35 B =26.15 C =158.11 TO =188.85 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=54.5(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=198.9 Deg F Temp@50 ft-lbs=375.7 Deg F Plant: PALISADES Material: SAW Heat: W5214 Orientation: NA Capsule: SA Fluence: 1.5E19 nlcm l\2 300 250 11
'0o 200 u..
21150 (I) s: w z (; 100 f'lO 50
~ ~ ---cJ ~
o
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Ceg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential 74.00 10. 00 12. 11 - 2. 11 129.00 24.00 18. 90 5. 10 154.00 23. 50 22. 68 . 82 204.00 30. 00 30. 85 - . 85 229.00 33. 50 34. 85 - 1. 35 254.00 28. 00 38. 55 -10.55 279.00 43. 50 41. 83 1. 67 279.00 48. 50 41. 83 6. 67 329.00 47. 50 46. 91 . 59
Palisades SA 1 Capsule Report Page 2 Plant: PALISADES Material: SAW Heat: W5214 Orientation: NA Capsule: SA Fluence: 1.5E19 nlcm A 2 Charpy V-Notch Data Temperature InputCVN Computed CVN Differential 404.00 51. 50 51. 27 . 23 454.00 55. 00 52.73 2. 27 479.00 57. 00 53. 20 3. 80 Correlation Coefficient = .957
Palisades SA-240-! Capsule CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0312512010 10:34 AM Page 1 Coefficients of Curve 1 A =27.35 B =25.15 C =111.62 TO =208.13 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=52.5(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=220.0 Deg F Temp@50 ft-lbs=372.9 Deg F Plant: PALISADES Material: SAW Heat: W5214 Orientation: NA Capsule: SA-240 Fluence: 2.38E19 n/cmA 2 300 250
,~ 200 '0o LL ~ 150 CI)
W Z (; 100 () ,.. 50 -~
~
o 0
-~ ~ -300 -200 -100 o 100 200 300 400 500 600 Temperature in Ceg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
70. 00 14. 00 6. 10 7. 90 125.00 15. 50 1 1. 45 4. 05 175.00 24. 50 20. 10 4.40 200.00 13. 00 25. 52 -12.52 200.00 26. 50 25. 52 .98 225.00 25. 00 3 l. 12 - 6. 12 250.00 40.00 36. 36 3. 64 300.00 54. 50 44. 37 10. 1 3 350.00 49.00 48. 83 . 17
Palisades SA-240-1 Capsule Report Page 2 Plant: PALISADES Material: SAW Heat: W5214 Orientation: NA Capsule: SA-240 Fluence: 2.38E19 nlcml\2 Charpy V-Notch Data Temperature InputCVN Computed CVN Differential 400.00 50. 50 50. 93 - . 43 450.00 52. 50 5 l. 85 . 65 500.00 54. 50 52. 23 2. 27 Correlation Coefficient = .935
H. B. Robinson 2 Unirradiated Capsule CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/25/2010 11:43 AM Page 1 Coefficients of Curve 1 A =56.05 B =53.85 C =107.57 TO =-29.1 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=109.9(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-Ibs=-85.8 Deg F Temp@50 ft-Ibs=-41.2 Deg F Plant: H B Robinson 2 Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: 300 250
,! 200 '0o u. ~ 150 cu s::::
w z 0 (Ib-lfil
~ 100 0 V
1 0/ 50 8 (
~
o
-300 W -200 0
o 200 400
-100 100 300 500 600 Temperature in Deg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
-150.00 19. 00 12. 49 6. 51 -150.00 10. 00 12. 49 - 2. 49 -150.00 30. 00 12. 49 17. 5 1 -150.00 34. 50 12. 49 22. 01 -150.00 3. 00 12. 49 - 9. 49 -150.00 2.00 12. 49 -10.49 -100.00 25. 00 24. 94 . 06 -100.00 38. 00 24.94 13. 06 -100.00 29. 00 24.94 4.06
H. Robinson 2 Unirradiated Capsule Report Page 2 Plant: H B Robinson 2 Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
-50.00 36. 50 45. 72 - 9. 22 -50.00 21. 00 45. 72 -24.72 -50.00 54. 50 45. 72 8. 78
10. 00 65. 50 74. 80 - 9. 30

10. 00 73.50 74. 80 - 1 . 30

10. 00 68. 00 74. 80 - 6. 80 60: 00 97. 00 92. 64 4. 36

60. 00 99. 00 92. 64 6. 36

60. 00 116.00 92. 64 23. 36 100.00 107.50 100.94 6. 56 110.00 97. 00 102.36 - 5. 36 110.00 104.00 102.36 1. 64 210.00 115.00 108.65 6. 35 210.00 Ill. 00 108.65 2. 35 210.00 112.00 108.65 3. 35 Correlation Coefficient = .962
B. Robinson 2 Capsule T Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/1412010 09:40 PM Page 1 Coefficients of Curve 1 A =31.35 B =29.15 C =9.09 TO =203.64 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT>>] Upper Shelf Energy=60. 5(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=203.3 Deg F Temp@50 ft-lbs=210.6 Deg F Plant: H B ROBINSON 2 Material: SAW Heat: W52l4 Orientation: NA Capsule: T Fluence: 3.87E19 nlcm"2 300 250 s:
/I) 200 15 o
u..
~ 150 CI>
W Z (; 100 0 ( n 50 II ° OJ:~ o ,
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Oeg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential 175.00 14. 00 2. 31 1 1 . 69 200.00 17. 00 20. 27 - 3. 27 200.00 23. 50 20. 27 3. 23 225.00 64.00 59. 97 4. 03 250.00 38. 50 60. 50 -22.00 275.00 51. 50 60. 50 - 9. 00 300.00 60. 50 60. 50 .00 Correlation Coefficient = .908
H. B. Robinson 2 Capsule V Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0312512010 11:47 AM Page 1 Coefficients of Curve 1
= = = =
A 36.35 B 34.15 C 150.19 TO 151.23 D O.OOE+OO = Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=70.5(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=123.0 Deg F Temp@50 ft-lbs=214.9 Deg F Plant: H B ROBINSON 2 Material: SAW Heat: W5214 Orientation: NA Capsule: V Fluence: 5.30E18 nlcml\2 300 250 III
.r:; 200 15 o
I.L
~ 150 I.. <I) s:
W
z
() 100
~
0 ____ r-- 50
~
o
-300 -200 -100 -o~ 100 0
200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
30. 00 23. 50 13. 54 9. 96

75. 00 23. 50 20. 37 3. 1 3 110.00 30. 00 27. 20 2. 80 160.00 14. 00 38. 34 -24.34 180.00 44.00 42. 81 l. 19 210.00 58. 50 49. 07 9. 43 300.00 72. 50 62. 22 10. 28 400.00 68. 50 68. 10 .40 Correlation Coefficient = .865
B. Robinson 2 Capsule X CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/25/2010 11:45 AM Page 1 Coefficients of Curve 1 A =40.97 B =38.78 C =59.96 TO = 197.18 D = O.OOE+OO Equation is A + B * [Tanh<<T-To)/(C+DT))] Upper Shelf Energy=79.8(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30ft-lbs=179.8 Deg F Temp@50ft-lbs=211.4DegF Plant: H B ROBINSON 2 Material: SAW Heat: W-5214 Orientation: NA Capsule: X Fluence: 4.49E19 nlcm"2 300 250
~ 200 "'i' '0 o
u..
~ 150 (I)
W Z C; 100 0 r () 50 o
,/ -300 -200 -100 o 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential .00 4.00 2. 31 1. 69 100.00 14. 00 5. 12 8. 88 175.00 28. 00 27. 25 . 75 200.00 38. 00 42.80 - 4. 80 250.00 74. 00 68. 39 5. 61 350.00 78. 00 79. 28 - 1. 28 375.00 85. 00 79. 54 5. 46 425.00 82. 00 79. 71 2. 29 Correlation Coefficient = .992
Indian Point 2 Unirradiated Capsule CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/2512010 10:55 AM Page 1 Coefficients of Curve 1 A =59.32 B =57.12 C = 86.23 TO = -16.54 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT>>] Upper Shelf Energy=116.4(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-Ibs=-65.4 Deg F Temp@50 ft-Ibs=-30.7 Deg F Plant: INDIAN POINT 2 Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: Unirradiat n/cm"2 300 250
~ '0o 200 u.. ~ 150 CI)
I: W ,... Q il Z a,/ tr () 100
/
V 50 o
~
00
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Ceg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
-150.00 12. 50 7. 15 5. 35 -150.00 10. 50 7. 15 3. 35 -100.00 35. *00 16. 61 18. 39 -100.00 18. 00 16. 61 l. 39 -100.00 9. 00 16. 61 - 7. 61
- 80.00 13. 00 23. 52 -10.52 - 80.00 26. 00 23. 52 2. 48 - 80.00 32. 50 23. 52 8. 98 -40.00 35. 50 44. 15 - 8. 65
Indian Point 2 Unirradiated C~psule -"-"'-UI'UJL Page 2 Plant: INDIAN POINT 2 Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: Unirradiat nlcm A 2 Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
-40.00 48. 00 44. 15 3. 85 -40.00 34. 00 44. 15 -10.15
10. 00 74. 00 76. 37 - 2. 37

10. 00 81. 00 76. 37 4. 63

10. 00 78. 50 76. 37 2. 13

60. 00 102.50 99. 89 2. 61

60. 00 102.00 99. 89 2. 1 1

60. 00 100.00 99. 89 . 11 110.00 112.50 110.68 l. 82 110.00 108.50 110.68 - 2. 18 110.00 108.50 110.68 - 2. 18 160.00 120.00 114.57 5. 43 160.00 115.50 114.57 . 93 160.00 113.00 114.57 - 1 . 57 210.00 123.50 115.85 7. 65 210.00 12l. 00 115.85 5. 15 210.00 117.50 115.85 l. 65 Correlation Coefficient = .990
Indian Point 2 Capsule V Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/25/2010 02:35 PM Page 1 Coefficients of Curve 1
= = = =
A 39.1 B 36.9 C 123.56 TO 163.14 D O.OOE+OO = Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=76.0(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=132.1 Deg F Temp@50 ft-lbs=200.8 Deg F Plant: INDIAN POINT 2 Material: SAW Heat: W5214 Orientation: NA Capsule: V Fluence: 4.92E18 nlcm"2 300 250
~ -; 200 '0o LL ~ 150 I...
Q.) W Z (; 100 50 ~ () 1"\ (")
~
V o
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Ceg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
74. 00 24. 00 16. 30 7. 70 130.00 26. 50 29. 44 - 2. 94 180.00 40. 50 44. 11 - 3. 61 220.00 53. 00 54. 98 - 1. 98 260.00 62. 50 63. 27 -.77 300.00 76. 00 68. 74 7. 26 325.00 72. 50 70. 99 1. 51 350.00 76. 00 72. 58 3. 42 Correlation Coefficient = .978
Indian Point 2 Capsule Y Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0312512010 10:59 AM Page 1 Coefficients of Curve 1 A =34.35 B =32.15 C =91.6 TO =140.88 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=66.5(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=128.5 Deg F Temp@50 ft-lbs=189.6 Deg F Plant: INDIAN POINT 2 Material: SAW Heat: W5214 Orientation: NA Capsule: Y Fluence: 4.55E18 nlcmA 2 300 250
~ 200 '0o u.. ~ 150 CI>
t: W Z (; 100 0 ,...
-(
50 V-
~ /
o
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Ceg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
74. 00 17. 50 14. 32 3. 18 110.00 23. 00 23. 90 -.90 160.00 40. 00 40. 96 -.96 190.00 47. 00 50. 11 - 3. 1 1 210.00 55. 00 54. 86 . 14 260.00 71. 50 62. 06 9.44 300.00 61. 00 64. 57 - 3. 57 350.00 67. 00 65. 84 1. 16 Correlation Coefficient = .978
Indian Point 3 Unirradiated Capsule CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/25/2010 11:16 AM Page 1 Coefficients of Curve 1 A =60.38 B =58.19 C =45.25 TO =-37.64 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))r Upper Shelf Energy=118.6(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=-63.8 Deg F Temp@50 ft-lbs=-45.8 Deg F Plant: INDIAN POINT 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: Unirradiat n/cmA 2 300 250 11
""i 200 '0o I.L e; 150 CI) s:
w e 0 z (; 100 r- 0 50 f o
~ .Q. . - -
v'
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Ceg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
-150.00 5. 00 3. 01 1. 99 -150.00 2.00 3. 01 - 1. 01 -150.00 4. 50 3. 01 1. 49 -100.00 29. 00 9. 15 19. 85 -100.00 18. 00 9. 15 8. 85 -100.00 25. 50 9. 15 16. 35
-50.00 35. 00 44. 88 - 9. 88 -50.00 33. 00 44. 88 -11.88 - 50.00 32. 50 44. 88 - 12. 38
Indian Point 3 Unirradiated Capsule Page 2 Plant: INDIAN POINT 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Unirra Fluence: Unirradiat nlcm A 2 Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
-35.00 78. 00 63. 78 14. 22 -35.00 69. 50 63. 78 5. 72 -35.00 54. 50 63. 78 - 9. 28 - 20.00 87. 00 81. 98 5. 02 - 20.00 82. 00 81. 98 .02 - 20.00 89. 00 81. 98 7. 02
10. 00 100.00 105.94 - 5. 94

10. 00 105.00 105.94 -.94

10. 00 113.50 105.94 7. 56

60. 00 115.00 117.04 - 2. 04 60.00 119.00 117.04 1. 96

60. 00 121. 50 117.04 4.46 160.00 124.00 118.55 5.45 160.00 125.00 118.55 6.45 160.00 112.00 118.55 - 6. 55 Correlation Coefficient = .981
Indian Point 3 Capsule T Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0312512010 11:19 AM Page 1 Coefficients of Curve 1 A = 46.35 B = 44.15 C = 98.27 TO = 124.16 D = O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT>>] Upper Shelf Energy=90.5(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-Ibs=86.0 Deg F Temp@50 ft-Ibs=132.4 Deg F Plant: INDIAN POINT 3 Material: SAW Heat: W5214 Orientation: NA Capsule: T Fluence: 2.63E18 nlcm"2 300 250
~ 200 ""i '5 o
1.1..
~ 150 CI)
W
z
(; 100 V-50
'7 /
o
~ Vo -300 -200 -100 o 100 200 300 400 500 600 Temperature in Ceg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential .00 13. 00 8. 73 4. 27
70. 00 17. 50 24. 21 - 6. 71 110.00 48. 00 40.03 7. 97 150.00 55. 50 57. 70 - 2. 20 150.00 53. 00 57. 70 - 4. 70 165.00 66. 00 63. 71 2. 29 210.00 78. 00 77. 39 . 61 300.00 90. 50 88. 10 2.40 Correlation Coefficient = .984
Indian Point 3 Capsule Y Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/2512010 11:18 AM Page 1 Coefficients of Curve 1 A =35.6 B =33.4 C =90.16 TO =122.54 D =O.OOE+OO Equation is A + B * [Tanh<<T-To)/(C+DT))] Upper Shelf Energy=69.0(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=107.3 Deg F Temp@50 ft-lbs=164.2 Deg F Plant: INDIAN POINT 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Y Fluence: 6.92E18 nlcmA 2 300 250 II)
.r; 200 '0 o
IJ.. e; 150 (U t: W Z () 100 ( 50 /"
~
o
~ -300 -200 -100 o 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential
25. 00 20.00 9. 08 10. 92

72. 00 19. 50 18. 62 . 88 125.00 31. 00 36. 51 - 5. 51 125.00 29. 50 36. 51 - 7. 01 150.00 49.00 45. 47 3. 53 200.00 67. 50 58. 84 8. 66 300.00 69. 50 67. 72 1. 78 400.00 68. 50 68. 86 - . 36 Correlation Coefficient = .960
Indian Point 3 Capsule Z Report CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0312512010 11:21 AM Page 1 Coefficients of Curve 1 A =39.1 B =36.9 C =97.52 TO =188.96 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))] Upper Shelf Energy=76.0(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-Ibs=164.5 Deg F Temp@50 ft-Ibs=218.7 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Z Fluence: l.04E19 n/cml\2 300 250
~ 200 15o u.. ~ 150 Q)
I: W Z
~ 100 ~
( 50 V
~/ o ( ~
o
-300 -200 -100 o 100 200 300 400 500 600 Temperature in Oeg F Charpy V-Notch Data Temperature InputCVN Computed CVN Differential 100.00 10. 00 12. 45 - 2. 45 150.00 21. 00 25. 10 - 4. 10 150.00 44.00 25. 10 18. 90 175.00 26. 00 33. 85 - 7. 85 200.00 33. 00 43. 26 -10.26 225.00 57. 00 52. 15 4. 85 300.00 75. 00 69. 14 5. 86 400.00 77. 00 75. 04 1. 96 Correlation Coefficient = .929
Indian Point 3 Capsule X CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/25/2010 02:34 PM Page 1 Coefficients of Curve 1 A =38.6 B =36.4 C =121.83 TO =157.96 D =O.OOE+OO Equation is A + B * [Tanh<<T-To)/(C+DT>>] Upper Shelf Energy=75.0(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs=128.7 Deg F Temp@50 ft-lbs=197.5 Deg F Plant: INDIAN POINT 3 Material: SAW Heat: W5214 Orientation: NA Capsule: X Fluence: 8.74E18 nlcm l\2 300 250 I/)
.s; 200 15o u..
2' 150 CI) I: W Z () 100 I"l
~
50

/

~-RC Palis ades Fluence '.VC.-\P-ij353~ Rev 0 3~3.+ Eyaluation x Email from S. Anderson Re;;ised Flumee Yalues (\VeHmghouse) to Tim Griesba::h (51A): IS for Design Inputs to x FrS Enuuation u.s '15/~010 fluence e;;aluation and D. S.Hood Letterto X Haskall (Palisades) s::hedule for reaching x PTS Screening Criteria Report No. 0901132.401, Rev. 0 1-3 Strjrlf!tll1'~llnte!l'ity Associates, Inc. Table 1-1. Listing of Design Inputs and References Used (cont.) Report TA,'\tI T;m.WeigliteIetiH>dGl<>gy DOCUU1E'ntsfCodes Average Temperature-Package )Jeutron Fluence ~~:t1a1ysis t01~ Palisades CAPL-O i .{H39 Sur>:eiUance Capsule x S_4..-240-1 CEOGCu.Ni CE NPSD-l119 2:1 x Chemistries .-\S1:\-1 E-l!3-5-66 22 Sl..lr". "eillance Testing x X ?d ..-\. EricksonKirk H.al. c~\:')J Data Fitting JOU1nal of?ressure ":esse! 25 X )vlethodology TechtlOlogv~ v. L'1, 2':};}9 Sofn.,\"at--e pl"ogm,m to fit: CVGR..A.?H 5 .; X CVNData 10CFR50.6~ 2 ¥IS Rille fluence Calcu1ation Reg Guide: .190 x ,\-ftho dolo ZY R..P'o;.:" Inte gtity Generic Lette1* 92-01 19 Assessment 'Workshop x ~4.S::YiE Boile1" and AS:vrE Section HI 21 Pt.es5ur.e Y.essel Code x 'h-raruation Data for XCREG 'CR*:6! ~ 3. A3G'2B and A533B OR..."'L*T:...:Z-~3133 21 Con:elation l\fonitor x X ~'iaterials NRC Reactor \. esse1 ")JRC R"',/IDl 31 Integ,11t"..:- Database x x [P2 Yes-sel Head FCX-OG538 29 Temperature x IP3 Ve5~el Head I?3-CALC-R\¥*03720 3'<< Temperature x Evaluations of Pausade:. RP"/ Through C)';S*D-*L'J2*Di. Rev. 2£ retied efE:nended x Operation Cycle i -11EFPDfor L_..:L:t of 2-21-:!GOO Tim.e \1.2"eighted 23 A ~perage Temp eratuce x E..:'\.-DOR-09**0! Rey, {} Calculation D.::terrnination of h.T D 09:C<: 132.301, Re~:. 32 ';/alues: for the Heat No. x '1.~5214 "/1illcatlQn of the Ti."11.e ',':li.7 eighted _~~""erage D9GU31.3D2. R.ev. 33 Temperatures forIP2 x and IP3 Capsules Report No. 0901132.401, Rev. 0 1-4 ~i!l'lNi'ttll'!IJIllnte!Jrity Associates, Inc. ATTACHMENT 4 Westinghouse Review of PTS Screening Criteria Assessment Westinghouse Letter, "Westinghouse Review of Reactor Vessel Weld Heat W5214 Surveillance Data for Palisades," Stephen T. Byrne (W) to Keith Smith (PNP), LTR-RIDA-1 0-1 07-NP, Rev. 0, May 13, 2010. 3 pages follow Westinghouse Non-Proprietary Class 3 Westinghouse Westinghouse Electric Company Nuclear Services Engineering Services 20 International Drive Windsor, Connecticut 06095 USA To: Keith Smith, Palisades Plant Date: May 13,2010 cc: Nathan A. Palm From: Stephen T. Byrne Ext: 860-731-6703 Our ref: LTR-RIDA-10-107-NP, Rev. 0 Fax: 860-731-6709
Subject:
Review of Reactor Vessel Weld Heat W5214 Surveillance Data for Palisades
References:
1. U.S. Nuclear Regulatory Commission, 10 CFR 50.61, "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," 60 FR 65468, December 19, 1995, as amended at 61 FR 39300, July 29, 1996; 72 FR 49500, August 28,2007; 73 FR 5722, January 31, 2008.

2. Structural Integrity Associates Report, 0901132.401, Rev. 0, "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," April 20, 2010.
Executive Summary The purpose of this letter is to document a review of the SIA report [2] that updated the pressurized thermal shock (PTS) screening criteria assessment for the Palisades reactor pressure vessel. A chemistry factor (CF) value of 227.4°F was determined in [2] based on post-irradiation surveillance test results specific to the Palisades vessel weld 2-112 NC, that contains weld wire heat number W5214 with Ni-200 wire addition. The CF evaluation incorporates data from four different surveillance programs, and applies a conservative methodology to normalize the data for differences in chemistry and irradiation temperature. The calculated 227.4°F CF value is more precise than a CF derived only from Cu and Ni content and more conservative than the results from the plant-specific surveillance capsules. RT PTS projections for Palisades weld 2-112 NC were made using the 227.4°F CF. The 270°F PTS screening criterion [1] for this axial weld is projected to be reached at a neutron fluence of 1.685E19 n/cm2 (E> 1 MeV), which will be attained during the first quarter of 20 17. The assessment performed in [2] is reasonably and conservatively based. Therefore, it is conservative to project that the PTS screening criterion will not be reached in the Palisades reactor vessel until the first quarter of the year 2017. This projection is based on the assumption that the neutron fluence
Page 2 of3 LTR-RIDA-1O-107-NP, Rev. 0 May l3, 2010 accumulation rate will continue as expected such that the 60° weld location fluence will reach 1.685E19 n/ cm2 (E > 1 MeV) in the March 2017 - April 2017 time-frame. Review and Discussion The integrity of the Palisades reactor pressure vessel is determined in part through an evaluation of the limiting RTPTS for the beltline materials. The RT pTs is determined in accordance with 10 CFR 50.61 [1] and compared to screening criteria specific to the orientation of an assumed flaw. The Palisades reactor pressure vessel beltline consists of the lower and intermediate shell courses that are fabricated from six low-alloy steel plates and the adjoining welds. The focus of the evaluation is the limiting weld fabricated using wire heat number W5214 with Ni-200 wire addition. The chemistry factor determined for this weld from Table 1 of [1] using best estimate copper (Cu) and nickel (Ni) content is stated in [2] as 231.08°F. In [2], an update is provided to an earlier evaluation of data from reactor vessel surveillance capsules irradiated in Palisades and from three other reactor vessels. The data are specific to submerged arc weldments produced using wire heat number W5214 with Ni-200 wire addition. The data specific to Palisades are from supplemental capsules SA-60-1 and SA-240-1. The weld material was obtained from the Palisades retired steam generator and irradiated in the Palisades vessel. The three other sources of weld data applicable to wire heat number W5214 plus Ni-200 wire are H.B. Robinson Unit 2, Indian Point Unit 2, and Indian Point Unit 3. These other sources represent three unique surveillance program welds. The transition temperature shift measurements were determined using a consistent curve-fitting methodology. The neutron fluence calculated for each evaluated surveillance capsule is obtained from recent sources. The surveillance data credibility analysis followed the criteria in [1], (c)(2)(i). Each of the five criteria were addressed subject to the extent of the information available from each of the four sources. The evaluation of surveillance data follows the approach described in [1], (c)(2) and (c)(3) with the addition of other considerations. Those considerations include differences in best estimate weld chemistry and irradiation temperature. Determination of chemistry factor (CF) given differences in copper and nickel content between the surveillance program weld and the Palisades reactor vessel weld is addressed in accordance with [1], (c)(2)(ii). Accounting for the effect of differences in the irradiation temperature between the individual plant-specific surveillance capsules and the Palisades reactor vessel is performed using informal guidance provided by the NRC Staff and included in the referenced report. (For each 1°F that the capsule temperature exceeds the mean vessel temperature, the transition temperature shift is decreased by 1OF. Likewise, a positive adjustment was applied to the shift for capsule temperatures lower than that of the vessel.) The two measured transition temperature shifts from Palisades (supplemental capsules SA-60-1 and SA-240-1) and the combined set of twelve measured shifts from the four sources, including adjustments for differences in copper and nickel content and for differences in irradiation temperature, were then analyzed using Equation 5 from [1]. The resultant CF
Page 3 on LTR-RIDA-lO-l07-NP, Rev. 0 May 13, 2010 value was then assessed for predictability relative to each of the twelve adjusted transition temperature shifts. The evaluation of the two measured transition temperature shifts from Palisades resulted in a CP of 198.8°F. When assessed for predictability, the SA-60-1 and SA-240-1 measured shifts are within approximately 3°P of the prediction (within the one-sigma scatter of 28°P). An additional evaluation of the correlation monitor material from those Palisades supplemental capsules demonstrated the correlation monitor shift measurements to also be consistent with predictions. Therefore, the plant-specific measurements are fully consistent with predictions. The evaluation of the combined twelve measured tranSItion temperature shifts for four separate weldments produced using wire heat number W5214 with Ni-200 wire addition resulted in a CP of 227.4°P. When assessed for predictability, the twelve measured shifts are within approximately 46°P of the prediction (within the two-sigma scatter of 56°P). The 227.4°P CP value was used to project the RT PTS for the Palisades reactor vessel weld 2-112 AlC. The 2700 P PTS screening criterion [1] is projected to be reached at a neutron fluence of 1.685E19 n/cm2 (E> 1 MeV). That fluence will be attained during the first quarter of 2017. The RT PTS projection is based on the assumption that the neutron fluence accumulation rate will continue as expected such that the 60° weld location fluence will reach 1.685E19 n/cm2 (E > 1 MeV) in the March 2017 - April 2017 time-frame. In conclusion, this review finds that the assessment performed in [2], including the CP evaluation and the RTPTS projection, is reasonably and conservatively based. If you have any questions or desire further information, please contact the undersigned. ELECTRONICALL,Y APPROVED] Verified by: ELECTRONICALLY APPROVED] Stephen T. Byrne, Author Paul R. Sotherland, Reviewer Reactor Internals Design & Analysis II Reactor Internals Design & Analysis II Approved by: ELECTRONICALLY APPROVED] John P. Kielb, Acting Manager Reactor Internals Design & Analysis II
© 2010 Westinghouse Electric Company LLC All Rights Reserved I Electronically approved records are authenticated in the electronic document management system.
ATTACHMENT 5 List of ENO Representatives Planning to Attend Meeting Brian Kemp, Design Engineering Manager, PNP Paula Anderson, Licensing Manager, PNP Jeff Erickson, Licensing, PNP Keith Smith, Project Manager, PNP Bill Server, President, ATI Consulting Tim Griesbach, Senior Associate, Structural Integrity Associates Steve Byrne, Fellow Engineer, Westinghouse
ATTACHMENT 6 Planned Meeting Agenda Introduction of Meeting Attendees Purpose of Meeting
Historical Status of PTS at Palisades Current Activities to Manage PTS at Palisades Integration of New Data Update of WCAP-15353 PTS Assessment for Limiting Weld Metal o Future Plans

Proposed Schedule

Questions e Identification of Action Items}}

v t e Letter Sequence Request
TAC:ME4021, (Open)
Initiation Request, Request Acceptance... Administration Meeting, Meeting Questions Answers Results Approval... Other:
MONTHYEAR2010-06-22 [Table View]

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