ML14211A525

WCAP-17628-NP, Revision 1, Palisades, Alternate Pressurized Thermal Shock (PTS) Rule Evaluation.

ML14211A525
Person / Time
Site:	Palisades
Issue date:	07/29/2014
From:	Freed A Westinghouse
To:	Office of Nuclear Reactor Regulation
References
WCAP-17628-NP, Rev. 1
Download: ML14211A525 (47)

Text

Enclosure Alternate Pressurized Thermal Shock (PTS)

Rule Evaluation for Palisades 46 Pages Follow

Westinghouse Non-Proprietary Class 3 WCAP-17628-NP June 2014 Revision 1 Alternate Pressurized Thermal Shock (PTS) Rule Evaluation for Palisades Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17628-NP Revision 1 Alternate Pressurized Thermal Shock (PTS) Rule Evaluation for Palisades Amy E. Freed*

Materials Center of Excellence June 2014 Reviewer: Stephen M. Parker*

Reactor Internals Aging Management Approved: Frank C. Gift*, Manager Materials Center of Excellence

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2014 Westinghouse Electric Company LLC All Rights Reserved

TABLE OF CONTENTS L IST O F TA B L E S ....................................................................................................................................... iii L IST O F F IG U RE S ..................................................................................................................................... iv EX EC U TIV E SU M M A RY ........................................................................................................................... v I IN T RO D UC T ION ......................................................................................................................... 1-1 2 ALTERNATE PRESSURIZED THERMAL SHOCK RULE ................................................. 2-1 3 M ETH OD DISC U SSIO N ............................................................................................................. 3-1 3.1 CALCULATION OF RT .x.x VALUES ..................................................................... 3-1 3.2 SURVEILLANCE CAPSULE DATA STATISTICAL CHECKS .................................... 3-5 3.3 REACTOR VESSEL BELTLINE ISI DATA EVALUATION ......................................... 3-8 4 PLANT-SPECIFIC RV MATERIAL PROPERTIES AND DIMENSIONS ............................ 4-1 5 RV NEUTRON FLUENCE VALUES AND COLD LEG TEMPERATURES ........................ 5-1 6 SURVEILLANCE CAPSULE DATA ........................................................................................... 6-1 7 INSERVICE INSPECTION DATA ........................................................................................... 7-1 8 DETERMINATION OF RTmx-x VALUES FOR ALL BELTLINE AND EXTENDED BELTLINE REGION MATERIALS ............................................................................................ 8-1 8.1 CALCULATION OF RT,,AX-X VALUES ......................................................................... 8-1 8.2 SURVEILLANCE CAPSULE DATA STATISTICAL CHECKS .................................... 8-3 8.3 REACTOR VESSEL BELTLINE ISI DATA EVALUATION ......................................... 8-7 9 C ON C L U SION ............................................................................................................................. 9-1 10 RE FEREN C E S ........................................................................................................................... 10-1 WCAP- 17628-NP June 2014 Revision I

iii LIST OF TABLES Table 3-1 Conservative Estimates for Chemical Element Weight Percentages ............................... 3-4 Table 3-2 PTS Screening C riteria .................................................................................................... 3-5 Table 3-3 Maximum Heat-Average Residual ['F] for Relevant Material Groups by Number of Available Data Points (Significance Level = 1%) ........................................................... 3-6 Table 3-4 TMAX Values for the Slope Deviation Test (Significance Level = 1%) ............................ 3-7 Table 3-5 Threshold Values for the Outlier Deviation Test (Significance Level = 1%) .................. 3-8 Table 3-6 A llowable N um ber of Flaws in Welds ............................................................................. 3-9 Table 3-7 Allowable Number of Flaws in Plates or Forgings .......................................................... 3-9 Table 4-I Details of RTMAX.X Calculation Inputs for Palisades ................................................. 4-I Table 4-2 Palisades RV Dim ensions ................................................................................................ 4-2 Table 5-I Maximum Neutron Fluence on the RV Clad-to-Base Metal Interface for Palisades at 42.1 E F P Y ................................................................................................................................ 5-1 Table 5-2 RV Cold Leg Temperature per Operating Cycle for Palisades ........................................ 5-2 Table 6-1 Surveillance Data for Palisades Base Metal Heat C- 1279 ............................................... 6-2 Table 6-2 Surveillance Data for Palisades Weld Wire Heat W5214 ................................................ 6-3 Table 6-3 Surveillance Data for Palisades Weld Wire Heat 27204 .................................................. 6-4 Table 7-I Reactor Vessel ISI History for Palisades Beltline and Extended Beltline Materials ....... 7-1 Table 7-2 ISI Information for Reactor Vessel Beltline and Extended Beltline Flaws for Palisades 7-2 Table 8-1 RTMAx-Aw Calculation Results for Palisades at 42.1 EFPY ............................................. 8-1 Table 8-2 RTMAAxPL Calculation Results for Palisades at 42.1 EFPY ............................................... 8-2 Table 8-3 RTMAX-CW Calculation Results for Palisades at 42.1 EFPY ............................................. 8-3 Table 8-4 RTMAX-X values for Palisades at 42.1 EFPY .................................................................... 8-3 Table 8-5 Surveillance Capsule M aterials for Palisades .................................................................. 8-4 Table 8-6 Surveillance Data Evaluation for Palisades Base Metal Heat C-1279 ............................. 8-5 Table 8-7 Surveillance Data Evaluation for Palisades Weld Wire Heat W5214 .............................. 8-6 Table 8-8 Surveillance Data Evaluation for Palisades Weld Wire Heat 27204 ................................ 8-7 Table 8-9 Reactor Vessel ISI Evaluation for Potential Beltline and Extended Beltline Flaws for P alisad es ........................................................................................................................... 8-8 Table 8-10 Inspection Length and Area for Palisades ...................................................................... 8-10 Table 8-11 Alternate PTS Rule Allowable Number of Flaws in Plates and Forgings Scaled for P alisad es ......................................................................................................................... 8-10 Table 8-12 Alternate PTS Rule Allowable Number of Flaws in Welds Scaled for Palisades .......... 8-11 WCAP-17628-NP June 2014 Revision I

iv LIST OF FIGURES Figure 4-1 Identification and Location of Beltline Region Materials for the Palisades Reactor Vessel ...4-3 Figure 8-1 Weld and Plate Indication Map for Palisades Beltline and Extended Beltline ...................... 8-12 WCAP- 17628-NP June 2014 Revision I

v EXECUTIVE

SUMMARY

The Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61 a) was approved by the U.S. Nuclear Regulatory Commission (NRC) and included in the Federal Register, with an effective date of February 3, 2010. This Alternate Rule provides a new metric and screening criteria for PTS. This metric, RTMAX-X, and the corresponding screening criteria are far less restrictive than the RTPTS metrics and screening criteria in the original PTS Rule (10 CFR 50.61).

The purpose of this report is to provide Palisades with the basis for implementation of the Alternate PTS Rule up to the end of the 60-year operating license. The evaluation described in this report led to the following conclusions:

I. The Palisades reactor vessel (RV) beltline and extended beltline materials have end-of-license extension (EOLE), 42.1 effective full-power years (EFPY), RTMAX-X values below the Alternate PTS Rule screening criteria;

2. The Palisades surveillance data for the vessel passed all of the surveillance data statistical tests for each material; and

3. The Palisades RV beltline and extended beltline weld flaw density and size distribution are acceptable based on the latest Palisades vessel inservice inspection (ISI) results from an ASME Section XI, Appendix VII1 qualified examination.

WCAP- 17628-NP June 2014 Revision 1

1-1 1 INTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel (RV). A PTS concern arises if one of these transients acts on the beltline region of an RV where a reduced fracture resistance exists because of neutron irradiation. Such an event may initiate and propagate an existing flaw or cause the propagation of a flaw postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.

The purpose of this report is to document evaluations performed for the Palisades RV to meet the requirements of 10 CFR 50.61a, "Alternate fracture toughness requirements for protection against pressurized thermal shock events" (Reference 1). Section 2 discusses the Alternate PTS Rule and its requirements. Section 3 provides the methodology for calculating RTMAx-x and performing the examination and flaw assessment required per the Alternate PTS Rule. Sections 4 through 7 provide inputs necessary to conduct the Alternate PTS Rule evaluations described in Section 3. Specifically, these sections provide the material properties, neutron fluence values, surveillance capsule analysis results, and inservice inspection (ISI) data of the RV beltline and extended beltline materials. The results of the RTMAX-X calculations and flaw assessment are presented in Section 8. The conclusion and references for the PTS evaluation follow in Sections 9 and 10, respectively.

Note that the evaluation contained in this report is focused on the RV beltline and extended beltline regions of the RV. For the purposes of this evaluation, the RV beltline is defined as the region immediately adjacent to the reactor core. This definition is consistent with the definition in the original and Alternate PTS Rules (References 2 and 1), which is as follows:

Reactor Vessel Belline means 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 active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiationdamage to be considered in the selection of the most limiting material with regardto radiationdamage.

The RV extended beltline for Palisades is defined as the region of materials that meet or exceed a neutron fluence exposure of 1.0 x 1017 n/cm 2 (E > 1.0 MeV). These materials were identified within the Palisades extended beltline RV integrity evaluation (Reference 3).

Please note that the Alternate PTS Rule was developed to apply only to materials within the RV beltline.

Requirements of the Rule and other U.S. Nuclear Regulatory Commission (NRC) documents are provided throughout the following sections and often refer to the materials in the beltline region. For this analysis, when referring to beltline region materials, the reader should understand that the materials within the Palisades extended beltline projected to exceed a neutron fluence exposure of 1.0 x 1017 n/cm 2 (E > 1.0 MeV), per Reference 3, are also included.

WCAP-17628-NP June 2014 Revision 1

2-1 2 ALTERNATE PRESSURIZED THERMAL SHOCK RULE The Alternate PTS Rule (Reference 1) primary requirements consist of the following:

• Each licensee shall have projected values of RTMAX-X for each reactor vessel beltline material for the end-of-license (EOL) fluence of the material. The assessment of RTMAXxX values must use the calculation procedures described in Section 3.1 of this report. The assessment must specify the bases for the projected value of RTMAX-X for each reactor vessel beltline material, including the assumptions regarding future plant operation (e.g., core loading patterns, projected capacity factors); the copper (Cu), phosphorus (P), manganese (Mn), and nickel (Ni) contents; the reactor cold leg temperature (Tc); and the neutron flux and fluence values used in the calculation for each beltline material.

• Each licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data satisfy the criteria described in paragraphs (f)(6)(i)(A) and (f)(6)(i)(B) of 10 CFR 50.6 1a.

• Each licensee shall perform an examination and an assessment of flaws in the reactor vessel beltline as described in Section 3.3 of this report. The licensee shall verify that the requirements described in Section 3.3 have been met.

• Each licensee shall compare the projected RTMAx-x values for plates, forgings, axial welds, and circumferential welds to the PTS screening criteria in Table 3-2 of this report, for the purpose of evaluating a reactor vessel's susceptibility to fracture due to a PTS event.

• If any of the projected RTMAX-X values are greater than the PTS screening criteria in Table 3-2, then the licensee may propose the compensatory actions or plant-specific analyses as required in paragraphs (d)(3) through (d)(7) of 10 CFR 50.61 a, as applicable, to justify operation beyond the PTS screening criteria in Table 3-2. The licensee shall implement those flux reduction programs that are reasonably practicable to avoid exceeding the PTS screening criteria. If this analysis indicates that no reasonably practicable flux reduction program will prevent the RTMAX-X value for one or more of the reactor vessel beltline materials from exceeding the PTS screening criteria, then the licensee shall perform a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent the potential for an unacceptably high probability of failure of the reactor vessel as a result of postulated PTS events. In the analysis, the licensee may determine the properties of the reactor vessel materials based on available information, research results and plant surveillance data, and may use probabilistic fracture mechanics techniques.

The Alternate PTS Rule subsequent requirements consist of the following:

Whenever there is a significant change in projected values of RTMAX-X, so that the previous value, the current value, or both values, exceed the screening criteria before the expiration of the plant operating license; or upon the licensee's request for a change in the expiration date for operation of the facility; a re-assessment of RTMAX-X values must be conducted. If the surveillance data used to perform the re-assessment of RTMAX-X values meet the requirements discussed in Section 3.2 of this report, the data must be analyzed in accordance with the Alternate PTS Rule and the RTMAx-X values must be recalculated and resubmitted for approval.

WCAP- 17628-NP June 2014 Revision I

2-2 The licensee shall verify that the requirements of paragraphs (e), (e)(l), (e)(2). and (e)(3) of 10 CFR 50.61a have been met. The licensee must submit, within 120 days after completing a volumetric examination of reactor vessel beltline materials as required by ASME Code,Section XI, the adjustments made to the volumetric test data to account for NDE-related uncertainties as described in paragraph (e)(1) of 10 CFR 50.61a and all information required by paragraph (e)(1)(iii) of 10 CFR 50.61a for review and approval. If a licensee is required to implement paragraphs (e)(4), (e)(5), and (e)(6) of 10 CFR 50.61a, the information required in these paragraphs must be submitted within one year after completing a volumetric examination of reactor vessel materials as required by ASME Code,Section XI.

WCAP-17628-NP June 2014 Revision 1

3-1 3 METHOD DISCUSSION 3.1 CALCULATION OF RTMAXx VALUES In accordance with paragraph (f) of 10 CFR 50.61a, each licensee shall calculate RTMAX-x values for each reactor vessel beltline material using pt. Note that the Palisades RV materials determined to be located in the extended beltline (Reference 3) are also included in this evaluation.

The values of RT4AX-AW (Axial Welds), RTMAX-PL (Plates), RTMAX-FO (Forgings), and RTMAX-CW (Circumferential Welds) must be determined using Equations 1 through 4 (Reference 1). NUREG- 1874 (Reference 4) provides additional information on these equations, which are included below. RTMAx-X values are calculated in degrees Fahrenheit (°F) as follows:

RTMAx-AW -MAXL nAWFL F(RTa'-"i(')

MAXAWFL(i) L NLT(u)

+ATad-cni'(j) 30"

(,ta )]11l)

RTMAX=W _ I"-P(' \' NDT(u) +" AI,-

A 30- (05tFL)

Where:

nAWFL is the number of axial weld fusion lines in the beltline region of the vessel, i is a counter that ranges from 1 to nAWFL, Ot FL is the maximum fluence occurring on the vessel inner diameter (ID) along a particular axial weld fusion line, RTDT-'('. is the unirradiated RTNDT of the weld adjacent to the ith axial weld fusion line.

RTa'j--"l) is the unirradiated RTNDT of the plate adjacent to the ith axial weld fusion line, AT~o'"-'~'m is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to 0 tFL of the weld adjacent to the ith axial weld fusion line, and A T3o"-oi-P() is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation due to o5tFL of the plate adjacent to the i" axial weld fusion line.

npL RTMAX-PL ý- ,*MXRTRPLO) + P*0) (PL(1)] (2)

Where:

nPL is the number of plates in the beltline region of the vessel, i is a counter that ranges from I to npL,

• ]tPL(i)

MAV is the maximum fluence occurring over the vessel ID occupied by a particular plate, WCAP- 17628-NP June 2014 Revision I

3-2 RT'L(,) is the unirradiated RTNDT of a particular plate, and AT PL() is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by tPL(i) irradiation to vtM.A of a particular plate.

RTMAX-FO = MAX[RT °08I) +AT FO.,0(oo)] (3)

Where:

nFO is the number of forgings in the beltline region of the vessel, i is a counter that ranges from I to nFo, btg.*j( is the maximum fluence occurring over the vessel ID occupied by a particular forging, RTF*10), is the unirradiated RTNDT of a particular forging, and AT3F°') is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by i*FO(it) _*

irradiation to tlMx of a particular forging.

F(RT`('e-cI) + AT ad1c.(.. (,tFI, L

NDT(u) 30 RTMAX-CW =MMAX MAXCWFL(i) (RT""-P'-') + ATHq-Plul(qttF)(

XNDT(u)(=1 + AT 30m-i) (o) j (4),))

Where:

ncWFL is the number of circumferential weld fusion lines in the beltline region of the vessel, i is a counter that ranges from I to nCWFL, Ot FL is the maximum fluence occurring on the vessel ID along a particular circumferential weld fusion line, RTa(J-'I')is the unirradiated RTNDT of the weld adjacent to the ih circumferential weld fusion line, RT adj-pi) is the unirradiated RTNDT of the plate adjacent to the ith circumferential weld fusion line (if there is no adjacent plate this term is ignored),

RTa"-hfi') is the unirradiated RTNDT of the forging adjacent to the i circumferential weld fusion line (if there is no adjacent forging this term is ignored),

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3-3 AT3o¢-jcl"i) is the shift in the Charpy V-Notch 30-foot-pound (fi-lb) energy produced by irradiation due to tbtFIof the weld adjacent to the ih circumferential weld fusion line, AT*'0a'-"'* is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to qltFL of the plate adjacent to the ih circumferential weld fusion line (if there is no adjacent plate this term is ignored), and AT3odj-f°(i) is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to 0 tFL of the forging adjacent to the ith circumferential weld fusion line (if there is no adjacent forging this term is ignored).

The values of AT30 must be determined using Equations 5, 6, and 7 for each axial weld, plate, forging, and circumferential weld. The AT30 value for each axial weld calculated as specified by Equation I must be calculated for the maximum fluence occurring along a particular axial weld (0tFL) at the clad-to-base metal interface. The AT30 value for each adjacent plate calculated as specified by Equation 1 must also be calculated using the same value of YtFL used for the axial weld. The AT30 value for each plate or forging calculated as specified by Equations 2 and 3 must be calculated for the maximum fluence (YtMAX) occurring at the clad-to-base metal interface over the entire area of each plate or forging. In Equation 4, the fluence (YtFL) value used for calculating the circumferential weld AT30 value is the maximum fluence occurring along the circumferential weld at the clad-to-base metal interface. The AT30 values in Equation4 shall also be calculated for the adjoining plates or forgings using the same maximum circumferential weld fluence. If the conditions for the surveillance capsule data specified in Section 3.2 are not met, licensees must propose AT3o and RTmAX-X values in accordance with paragraph (f)(6)(vi) of 10 CFR 50.61 a (Reference 1).

The equation used to calculate the AT3(oshift is displayed below:

AT3, = MD + CRP (5)

Where:

MD= A(]-0.001718Tc)( l+6.13PMn2 4 7 1) (Pte° 5

(6)

CRP= B(1+3.77Ni 91 ) f(Cue,P)g(Cue,Ni, (Pte) (7)

A= 1.1 40x 10-7 for forgings 1.561 x 10-7 for plates 1.417x 10-7 for welds B= 102.3 for forgings 102.5 for plates in non-Combustion Engineering (CE) manufactured vessels 135.2 for plates in CE manufactured vessels 155.0 for welds WCAP-17628-NP June 2014 Revision I

3-4 Wet= t for y > 4.39x 1010 n/cm 2/sec (pt(4.39x 101°/ () 0 2595 for (p< 4.39 x 1010 n/cm 2/sec f(Cue,P)= 0 for Cu _ 0.072

[Cue - 0.072]0-668 for Cu > 0.072 and P < 0.008

[Cue - 0.072+1.359(P-0.008)] 0 661 for Cu > 0.072 and P > 0.008 Cue 0 for Cu _<0.072 MIN (Cu, Maximum Cue) for Cu > 0.072 MIac. Cue= 0.243 for Linde 80 welds 0.301 for all other materials g(Cu,'Ni, °td= I1+ I tanh IIg1°(pt,)+ '1.390Cu -O'448Ni-18.120.

2 2 1 0.629j TC= Cold leg temperature under normal full power operating conditions (0F) as a time-weighted average

(= Average neutron flux (n/cm 2-/sec) t= Time that the reactor has been in full power operation (sec)

ýpt Neutron Fluence (n/cm 2)

P= Phosphorous content (wt%)

Ni= Nickel content (wt%)

Cu= Copper content (wt%)

Mn= Manganese content (wt%)

The values of Cu, Mn, P, and Ni in Equations 6 and 7 must represent the best estimate values for the material. For a plate or forging, the best estimate value is normally the mean of the measured values for that plate or forging. For a weld, the best estimate value is normally the mean of the measured values for a weld deposit made using the same weld wire heat number as the critical vessel weld. If these values are not available, either the upper limiting values given in the material specifications to which the vessel material was fabricated, or conservative estimates (i.e., mean plus one standard deviation) based on generic data as shown in Table 3-I for P and Mn, must be used.

Table 3-1 Conservative Estimates for Chemical Element Weight Percentages (Reference 1)

Materials P Mn Plates 0.014 1.45 Forgings 0.016 1.11 Welds 0.019 1.63 The values of RTNDT(U) must be evaluated according to the procedures in the ASME Code,Section III, paragraph NB-2331. If any other method is used for this evaluation, the licensee shall submit the proposed method for review and approval by the Director along with the calculation of RTMAX-X values.

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3-5

• If a measured value of RTNDT(U) is not available, a generic mean value of RTNDT(U) for the class of material must be used if there are sufficient test results to establish a mean.

• The following generic mean values of RTNDT(U) must be used unless justification for different values is provided: 0°F for welds made with Linde 80 weld flux; and -56°F for welds made with Linde 0091, 1092, and 124 and ARCOS B-5 weld fluxes.

The value of Tc in Equation 6 of this section must represent the time-weighted average of the reactor cold leg temperature under normal operating full power conditions from the beginning of full power operation through the end of licensed operation. For the surveillance capsule statistical tests, Tc is a time-weighted average from the beginning of full power operation up to the time of capsule withdrawal.

If any of the calculated RTMAX.X values for Palisades are greater than the PTS screening criteria, defined in Table 3-2, further evaluation or action, consistent with paragraphs (d)(3) through (d)(7) of the Rule, is required.

Table 3-2 PTS Screening Criteria RTMAX-X limits [7F] for different vessel wall thicknesses' Product form and RTMAX-X (TWALL) values 9.5 in. < TWALL 10.5 in. < TWALL T<10.5 in. _11.5 in.

Axial Weld-RTMAXAW 269 230 222 Plate-RTMAXPL 356 305 293 Forging without underclad cracks-RTMAXFo2 356 305 293 Axial Weld and Plate-RTMAXAW + RTMAX-PL 538 476 445 Circumferential Weld-RTMgx-cw 3 312 277 269 Forging with underclad cracks-RTNIAXFo 4 246 241 239 Notes:

1. Wall thickness is the beltline wall thickness including the clad thickness.

2. Forgings without underclad cracks apply to forgings for which no underclad cracks have been detected and that were fabricated in accordance with Regulatory Guide 1.43.

3. RTp.s limits contribute 1 x 10-8 per reactor year to the reactor vessel TWCF.

4. Forgings with underclad cracks apply to forgings that have detected underclad cracking or were not fabricated in accordance with Regulatory Guide 1.43.

3.2 SURVEILLANCE CAPSULE DATA STATISTICAL CHECKS As a condition of the Alternate PTS Rule, the licensee must consider plant-specific information that could affect the use of this equation for the determination of a material's AT 30 value. In order to make this determination, the Alternate PTS Rule provides requirements for evaluation of surveillance capsule data.

These requirements are specified in paragraphs (f)(6)(i), (f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of Reference 1.

The requirements consist of a Mean Deviation Test, a Slope Deviation Test, and an Outlier Deviation Test.

Specifically, the Rule states that the licensee shall verify that an appropriate RTMAX-X value has been calculated for each reactor vessel beltline material by considering plant-specific information that could WCAP-17628-NP June 2014 Revision 1

3-6 affect the use of the model (i.e., Equations 5, 6, and 7) for the determination of a material's AT 30 value.

The licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data satisfy the following criteria:

• The surveillance material must be a heat-specific match for one or more of the materials for which RTMAX-X is being calculated. The 30-foot-pound transition temperature must be determined as specified by the requirements of 10 CFR part 50, Appendix H.

• If three or more surveillance data points measured at three or more different neutron fluences exist for a specific material, the licensee shall determine if the surveillance data show a significantly different trend than the embrittlement model predicts. If fewer than three surveillance data points exist for a specific material, then the embrittlement model must be used without performing the consistency check.

The licensee shall estimate the mean deviation from the embrittlement model for the specific data set (i.e.,

a group of surveillance data points representative of a given material). The mean deviation from the embrittlement model for a given data set must be calculated using Equations 8 and 9. The mean deviation for the data set must be compared to the maximum heat-average residual given in Table 3-3 or derived using Equation 10. The maximum heat-average residual is based on the material group into which the surveillance material falls and the number of surveillance data points. For surveillance data sets with greater than 8 data points, the maximum credible heat-average residual must be calculated using Equation

10. The value of o used in Equation 10 must be obtained from Table 3-3.

Residual (r) = MeasuredATo - PredictedATI3o (8)

Mean deviationfor a data set of n datapoints =(I / n)x r, (9)

Maximum credible heat-amerage residual = 2.33o/n° 5 (10)

Where:

n = number of surveillance data points (sample size) in the specific data set a= standard deviation of the residuals about the model for a relevant material group given in Table 3-3.

Table 3-3 Maximum Heat-Average Residual 1[F1 for Relevant Material Groups by Number of Available Data Points (Significance Level = 1%)

Material group o [0F]4 Number of5available6data points 7 8 3 14 5 6 7 8 Welds, for Cu > 0.072 26.4 35.5 30.8 27.5 25.1 23.2 21.7 Plates, for Cu > 0.072 21.2 28.5 24.7 22.1 20.2 18.7 17.5 Forgings, for Cu > 0.072 19.6 26.4 22.8 20.4 18.6 17.3 16.1 Weld, Plate or Forging, for Cu < 0.072 18.6 25.0 21.7 19.4 17.7 16.4 15.3 WCAP- 17628-NP June 2014 Revision I

3-7 The licensee shall estimate the slope of the embrittlement model residuals (estimated using Equation 8) plotted as a function of the base 10 logarithm of neutron fluence for the specific data set. The licensee shall estimate the T-statistic for this slope (TsuRv) using Equation I I and compare this value to the maximum permissible T-statistic (TMAX) in Table 3-4. For surveillance data sets with greater than 15 data points, the TMAX value must be calculated using Student's T distribution with a significance level (a) of I percent for a one-tailed test.

m (11)

(se(m))

Where:

ni - The slope of a plot of all of the r values (estimated using Equation 8) versus the base 10 logarithm of the neutron fluence for each r value.

The slope shall be estimated using the method of least squares.

(se(in)) = The least-squares estimate of the standard-error associated with the estimated slope value m.

Table 3-4 TNAx Values for the Slope Deviation Test (Significance Level = 1%)

Number of available data points (n) TMAX 3 31.82 4 6.96 5 4.54 6 3.75 7 3.36 8 3.14 9 3.00 10 2.90 11 2.82 12 2.76 13 2.72 14 2.68 15 2.65 The licensee shall estimate the two largest positive deviations (i.e., outliers) from the embrittlement model for the specific data set using Equations 8 and 12. The licensee shall compare the largest normalized residual (r *) to the appropriate allowable value from the third column in Table 3-5 and the second largest normalized residual to the appropriate allowable value from the second column in Table 3-5.

r (12)

Where r is defined using Equation 8 and a is given in Table 3-3.

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3-8 Table 3-5 Threshold Values for the Outlier Deviation Test (Significance Level = 1%)

Number of available data points Second largest allowable normalized Largest allowable normalized (n) residual value (r*) residual value (r*)

3 1.55 2.71 4 1.73 2.81 5 1.84 2.88 6 1.93 2.93 7 2.00 2.98 8 2.05 3.02 9 2.11 3.06 10 2.16 3.09 11 2.19 3.12 12 2.23 3.14 13 2.26 3.17 14 2.29 3.19 15 2.32 3.21 The AT 30 value must be determined using Equations 5, 6, and 7 if all three of the following criteria are satisfied:

• The mean deviation from the embrittlement model for the data set is equal to or less than the value in Table 3-3 or the value derived using Equation 10 of this section;

" The T-statistic for the slope (TsuRv) estimated using Equation I I is equal to or less than the maximum permissible T-statistic (TMAx) in Table 3-4; and

• The largest normalized residual value is equal to or less than the appropriate allowable value from the third column in Table 3-5 and the second largest normalized residual value is equal to or less than the appropriate allowable value from the second column in Table 3-5.

If any of these criteria are not satisfied, the licensee shall review the data base for that heat in detail.,

including all parameters used in Equations 5, 6, and 7 of this section and the data used to determine the baseline Charpy V-notch curve for the material in an unirradiated condition. The licensee shall propose AT 30 and RTMAX-X values, considering their plant-specific surveillance data, to be used for evaluation relative to the acceptance criteria of this rule.

3.3 REACTOR VESSEL BELTLINE ISI DATA EVALUATION The licensee must have performed an examination of the RV beltline welds using procedures, equipment, and personnel that have been qualified under the ASME Code Section XI, Appendix VIII, Supplement 4 and Supplement 6, as specified in 10 CFR 50.55a(b)(2)(xv). The licensee shall verify that the flaw density and size distributions within the volume described in ASME Code,Section XI (Reference 5),

Figures IWB-2500-1 and IWB-2500-2 and limited to a depth from the clad-to-base metal interface of 1-inch or 10 percent of the vessel thickness, whichever is greater, do not exceed the limits in Tables 3-6 and 3-7 based on the test results from the volumetric examination. The verification of the flaw density and size distributions shall be performed line-by-line for Tables 3-6 and 3-7.

WCAP-17628-NP June 2014 Revision I

3-9 Table 3-6 Allowable Number of Flaws in Welds Through-Wall Extent (TWE) of Flaw (in.) Maximum number of flaws per 1000 inches of weld length in the inspection volume that are greater than or equal to TWEMN TWEMAx TWEM[N and less than TWEMAx 0 0.075 No Limit 0.075 0.475 166.70 0.125 0.475 90.80 0.175 0.475 22.82 0.225 0.475 8.66 0.275 0.475 4.01 0.325 0.475 3.01 0.375 0.475 1.49 0.425 0.475 1.00 0.475 Infinite 0.00 The licensee shall determine the allowable number of weld flaws in the reactor vessel beltline by multiplying the values in Table 3-6 by the total length of the reactor vessel beltline welds that were volumetrically inspected and dividing by 1000 inches of weld length.

Table 3-7 Allowable Number of Flaws in Plates or Forgings Through-Wall Extent (TWE) of Flaw (in.) Maximum number of flaws per 1000 square inches of inside surface area in the inspection volume that are greater than or TWEMN TWEMAx equal to TWEMrN and less than TWEMAx 0 0.075 No Limit 0.075 0.375 8.05 0.125 0.375 3.15 0.175 0.375 0.85 0.225 0.375 0.29 0.275 0.375 0.08 0.325 0.375 0.01 0.375 Infinite 0.00 The licensee shall determine the allowable number of plate or forging flaws in their reactor vessel beltline by multiplying the values in Table 3-7 by the total surface area of the reactor vessel beltline plates or forgings that were volumetrically inspected and dividing by 1000 square inches.

For each flaw detected within the greater of 1-inch or 10 percent of the vessel thickness inspection volume, measured from the clad-to-base metal interface, with a through-wall extent equal to or greater than 0.075 inches, the licensee shall document the dimensions of the flaw, including through-wall extent and length, whether the flaw is axial or circumferential in orientation and its location within the reactor vessel, including its azimuthal and axial positions and its depth embedded from the clad-to-base metal interface.

WCAP- 17628-NP June 2014 Revision 1

3-10 The licensee shall verify that axially oriented flaws located at the clad-to-base metal interface do not open to the vessel inside surface using surface or visual examination techniques capable of detecting and characterizing service induced cracking of the reactor vessel cladding. The licensee shall verify that all flaws between the clad-to-base metal interface and three-eighths of the reactor vessel thickness from the interior surface are within the allowable values in ASME Code,Section XI, Table IWB-35 10-1.

The licensee shall perform analyses to demonstrate that the reactor vessel will have a through-wall cracking frequency (TWCF) of less than I x 10.6 per reactor year if the ASME Code, Section X1 volumetric examination indicates any of the following:

• The flaw density and size in the inspection volume exceed the limits in Tables 3-6 and 3-7;

" There are axial flaws that penetrate through the clad into the low alloy steel reactor vessel shell, at a depth equal to or greater than 0.075 inches in through-wall extent from the clad-to-base metal interface; or

• Any flaws between the clad-to-base metal interface and three-eighths of the vessel thickness exceed the size allowable in ASME Code,Section XI, Table IWB-3510-1.

If required, these analyses must address the effects on TWCF of the known sizes and locations of all flaws detected by the ASME Code,Section XI, Appendix VIII, Supplement 4 and Supplement 6 ultrasonic examination out to three-eighths of the vessel thickness from the inner surface, and may also take into account other reactor vessel-specific information, including fracture toughness information. The licensee shall also prepare and submit a neutron fluence map, projected to the date of license expiration, for the reactor vessel beltline clad-to-base metal interface and indexed in a manner that allows the determination of the neutron fluence at the location of the detected flaws.

WCAP-1 7628-NP June 2014 Revision I

4-1 4 PLANT-SPECIFIC RV MATERIAL PROPERTIES AND DIMENSIONS Before performing the Alternate PTS evaluation, a review of the latest plant-specific beltline and extended beltline region material properties for the Palisades RV was performed. See Section I for the definition of the beltline and extended beltline regions.

Table 4-I summarizes the best estimate copper, manganese, phosphorus, and nickel contents and RTNDT(U) values of the beltline and extended beltline materials for the Palisades RV. RTNDT(U) values for the Palisades RV plate materials were determined in accordance with the fracture toughness requirements in NUREG-0800, Revision 2, Branch Technical Position MTEB 5-3 (Reference 6); and the requirements of Subparagraph NB-2331 of Section III of the ASME B&PV Code (Reference 7). RTNDT(U) values for the weld materials are generic values for Linde 1092 and 124 weld fluxes per Reference 1.

The Palisades RV was fabricated by CE. Table 4-2 provides the dimensions of the Palisades RV necessary for the Alternate PTS Rule evaluation. This table includes wall thicknesses for the RV shell courses, RV inner diameter, and dimensions necessary to determine the location of the extended beltline region. Per Reference 8, the Palisades extended beltline region extends approximately 245 cm (96.5 inches) above and below the core midplane. The location of the extended beltline was then determined using these dimensions. Figure 4-1 shows the location of the beltline and extended beltline materials.

Table 4-1 Details of RTNAx.x Calculation Inputs for Palisades Region and Material Material Material CuM* Ni°l) P Mn RTNDT(u)

No. Component Identification Type Heat No. [wt%] [wt%] [wt%] [wt%]

Description [OF](') Method I Upper Shell Plate D-3802-1 SA-302BM C-1279-2 0.21 0.48 0.016-(2) 1.45(2) 10 Plant Specific Plant 2 Upper Shell Plate D-3802-2 SA-302BM C-1308-2 0.19 0.52 0.019(2) 1.29(2) 19 Pecfi Specific 3 Upper Shell Plate D-3802-3 SA-302BM C-1281-1 0.25 0.57 0.018(2) 1.32(2) 10 Plant Specific 4 Intermediate Shell D-3803-1 SA-302BM C-1279-3 0.24 0.50 0.011(2) 1.55(22 -5 Plant Plate Specifil D-3803-2 SA-302BM A-0313-2 0.24 0.52 0.01212) 1.43(2) -30 Plant 5 Intermediate Shell Plate Specific 6 Intermediate Shell D-3803-3 SA-302BM C-1279-1 0.24 0.50 0.010(2) 1.56(2) -5 Plant Plate I Specific Plant 7 Lower Shell Plate D-3804-1 SA-302BM C-1308-1 0.19 0.48 0.017(2) 1.22(2) 0 Plcit Specific 8 Lower Shell Plate D-3804-2 SA-302BM C-1308-3 0.19 0.50 0.013(2) 1.27(2) -30 Plant Specific

-25 Plant Pecfi 9 Lower Shell Plate D-3804-3 SA-302BM B-5294-2 0.12 0.55 0.010(2) 1.27(2)

Specific 0 Uper Shell 1-1 12A Linde 1092 W5214 0.213 1.007 0.0 19("* 1.63(') -56 Generic Longitudinal Weld WCAP-17628-NP June 2014 Revision I

4-2 Table 4-1 Details of RTNtAxx Calculation Inputs for Palisades Region and Material Material Material CuM'} Ni") P Mn RTNDT(u)

No. Component Identification Type Heat No. [wt%] [wt%] [wt%] [wt%]

Description [°F]" Method II Upper Shell 1-112B Linde 1092 W5214 0.213 1.007 0.019(3" 1.63( -56 Generic Longitudinal Weld 12 Upper Shell I-112C Linde 1092 W5214 0.213 1.007 0.019(') 1.63(3 -56 Generic Longitudinal Weld 13 Intermediate Shell 2-112A Linde 1092 W5214 0.213 1.007 0.019(3) 1.63(3 -56 Generic Longitudinal Weld 14 Intermediate Shell 2-112B Linde 1092 W5214 0.213 1.007 0.019,31 1.63(31 -56 Generic Longitudinal Weld 15 Intermediate Shell ILongitudinal Weld 2-112C Linde 1092 W5214 0.213 1.007 0.0 19(') 1.63 3 -56 Generic 16 Lower LwrSel Shell 3-112A Linde 1092 34B009 0.192 0.98 0.01913) 1.63(3) -56 Generic Longitudinal Weld W5214 0.213 1.007 0.01913) 1.63(3 ) -56 Generic Lower Shell 34B009 0.192 0.98 0.019013 1.63 3) -56 Generic 17 Longitudinal Weld 3-112B Linde 1092 W5214 0.213 1.007 0.019("* 1.63() -56 Generic 18 Lower LwrSel Shell 3-1 12C Linde 1092 34B009 0.192 0.98 0.0 19") 1.63(3) -56 Generic Longitudinal Weld W5214 0.213 1.007 0.019"3 1.63") -56 Generic Upper to 19 Intermediate Shell 8-112 Linde 1092 34B009 0.192 0.98 0.0191") 1.63(3 ) -56 Generic Circ. Weld Intermediate to 20 Lower Shell Circ. 9-112 Linde 124 27204 0.203 1.018 0.019(3) 1.63 3) -56 Generic Weld Notes:

1. Material chemistry and initial RTNTobtained from WCAP-17341-NP (Reference 9) and WCAP-17403-NP (Reference 3).

2. Phosphorus and manganese content is from plant-specific certified material test reports (Reference 10).

3. Plant-specific data is not available for weld materials. Weld material phosphorus and manganese content are conservative estimates provided in Table 3-1.

Table 4-2 Palisades RV Dimensions Parameter Dimension (in.) Reference No.

Inside Radius of RV (to clad/base metal interface) 86.35 9 Cladding Thickness 0.25 11 RV Thickness of Upper Shell Region (excluding cladding) 10.86 II RV Thickness of Intermediate and Lower Shell Region (excluding cladding) 8.74V) 11 Active Fuel Height 132 Plant-Specific Drawing Upper to Intermediate Shell Circ. Weld (Distance downward from Flange) 133.72 11 Intermediate to Lower Shell Circ. Weld (Distance downward from Flange) 231.03 II Bottom of Active Fuel (Distance downward from Flange) 280.4 Plant-Specific Drawing Note:

1. The RV beitline thickness used in the ISI inspection was reported as 8.74 inches. This value is slightly less than the value (8.79 inches) used in previous reactor vessel integrity analyses. Note that the slight difference in thickness does not change the applicable PTS screening criteria (See Table 3-2).

WCAP-17628-NP June 2014 Revision 1

4-3 117.

31 0.86"(2)

Figure 4-1 Identification and Location of Beltline Region Materials for the Palisades Reactor Vessel 3" Notes:

1. 117.94" corresponds to the upper extent of the extended beltline region. See Table 4-2 for references used to determine this location.

2. 310.86" corresponds to the lower extent of the extended beltline region. See Table 4-2 for references used to determine this location.

3. Map is not drawn to scale. Numbers in parentheses correspond to "No." column in Table 4-1.

Dimensions are measured downward from the RV flange surface.

WCAP-17628-NP June 2014 Revision I

5-1 5 RV NEUTRON FLUENCE VALUES AND COLD LEG TEMPERATURES The projected maximum neutron fluence (E > 1.0 MeV) values at the clad-to-base metal interface of the Palisades RV for 42.1 effective full-power years (EFPY) are shown in Table 5-1 for the beltline and extended beltline materials. Palisades is projected to have a total operating time of approximately 42.1 EFPY at end-of-life extension (EOLE).

In addition to neutron fluence data, the Palisades reactor cold leg temperature under normal operating full-power conditions from the beginning of full-power operation through the last operating cycle is presented in Table 5-2. The effective full-power operation times for each cycle are also included in this table. These temperatures and cycle times will be used to determine the time-weighted average of the reactor cold leg temperature, Tc, used in Equation 6 of Reference I. Data for this table was obtained from the latest PTS evaluation for Palisades (Reference 13).

Table 5-1 Maximum Neutron Fluence on the RV Clad-to-Base Metal Interface for Palisades at 42.1 EFPY Maximum Fluence Region and Component Description [1019 Neutron/cm 2, E > 1.0 MeV]"'

Upper Shell Plates 0.1529 Intermediate Shell Plates 3.42912, Lower Shell Plates 3.42912' Upper Shell Longitudinal Welds 0.09707 Intermediate Shell Longitudinal Welds 2.161 Lower Shell Longitudinal Welds 2.161 Upper to Intermediate Shell Circ. Weld 0.1529 Intermediate to Lower Shell Circ. Weld 3.429121 Notes:

1. Unless otherwise noted, the maximum fluence was obtained from WCAP-15353-Supplement 2-NP (Reference 8).

2. Maximum fluence obtained from Structural Integrity Associates (SIA) Report No. 1000915.401, Revision I (Reference 13).

WCAP-17628-NP June 2014 Revision I

5-2 Table 5-2 RV Cold Leg Temperature per Operating Cycle for Palisades Cycle Time Cycle Time Cumulative Cycle Time Cycle Tc (F) (EFPD) (EFPY) (EFPY) 1 523 371.7 1.018 1.0 2 529 440.1 1.205 2.2 3 534 342.5 0.938 3.2 4 536 321.0 0.879 4.0 5 536 386.7 1.059 5.1 6 536 326.7 0.894 6.0 7 536 362.5 0.992 7.0 8 537 366.1 1.002 8.0 9 534 292.5 0.801 8.8 10 534 349.7 0.957 9.7 11 533 421.9 1.155 10.9 12 534 399.3 1.093 12.0 13 536 419.6 1.149 13.1 14 537 449.3 1.230 14.4 15 537 401.3 1.099 15.5 16 537 444.3 1.216 16.7 17 537 493.1 1.350 18.0 18 537 472 1.292 19.3 19 537 459.2 1.257 20.6 20 537 499.8 1.368 22.0 21 537 519.2 1.422 23.4 22 537 498.8 1.366 24.7 WCAP- 17628-NP June 2014 Revision 1

6-1 6 SURVEILLANCE CAPSULE DATA There are three surveillance materials that are representative of the Palisades RV. The base metal surveillance material is a heat-specific match for upper shell plate D-3802-1 and intermediate shell plates D-3803-1 and D-3803-3 (Heat C-1279). The weld wire surveillance materials are heat-specific matches for the upper, intermediate, and lower shell longitudinal welds (Heat W5214) and the intermediate to lower shell circumferential weld (Heat 27204). The 30-foot-pound transition temperatures for each capsule were determined using measured Charpy V-notch data plotted using the requirements of 10 CFR part 50, Appendix H.

There have been four surveillance capsule analyses conducted for the Palisades base metal heat C-1279.

As a result, there are seven surveillance data points measured in the longitudinal and transverse directions. There have been eleven surveillance capsule analyses conducted for weld wire heat W5214 within the domestic PWR fleet, including Palisades. Lastly, there have been five surveillance capsule analyses conducted for weld wire heat 27204 within the domestic PWR fleet, including Palisades.

Tables 6-1 through 6-3 contain surveillance data of the Palisades beltline and extended beltline materials required to perform the surveillance data evaluation. Tables 6-2 and 6-3 contain sister plant material data from H. B. Robinson Unit 2 (HB2), Indian Point Units 2 and 3 (IP2 and IP3), and Diablo Canyon Unit I (DCI). A majority of the data in these tables was obtained from the latest PTS evaluation (Reference 13).

References for additional data are noted in the tables.

WCAP- 17628-NP June 2014 Revision I

6-2 Table 6-1 Surveillance Data for Palisades Base Metal Heat C-1279 eTime-Averaged Measured AT3o Transition Chemical Composition Fluence Withdraw Coolant Temperature (OF)(3) 2 Plant Capsule (xl 01' n/cm , EFPY Cycle Temperature 3

Cu Ni P Mn E > I.OMeV) (oF)( ) Longitudinal Transverse

[wvt%]o) [%vt%](l [%vt%](,) [wt%]m(Fj ogtdnl Tases Palisades A-240 0.25 0.53 0.011 1.55 4.09 2.2 2 526 205.0 205.0 Palisades W-290 0.25 0.53 0.011 1.55 0.938 5.1 5 531 155.0 175.0 Palisades W-1 10 0.25 0.53 0.011 1.55 1.64 9.7 10 533 180.0 (2)

Palisades W-100 0.25 0.53 0.011 1.55 2.09 16.7 16(l) 534 159.1 142.5 Notes:

1. Data obtained from Palisades W-100 capsule analysis (Reference 14).

2. No reported value for Transverse AT 30 (Reference 13).

3. Values are reported to the precision level calculated in Reference 13.

WCAP- 17628-NP June 2014 Revision I

6-3 Table 6-2 Surveillance Data for Palisades Weld Wire Heat W5214 PlantCapsulChmica Chemical Compoitio Composition Fun/ce2 Fluence Withdraw oln eprtre Time-Averaged Tasto Measuredeprtr AT3 0 r Plant Capsule Cu Ni P Mn (xl0' nlcm 2 , E EFPY C t u Te (Op)(7 7

[wt%] > 1.0MeV) (F)( ) F

[wt%] [wt%] [wvt%]

Palisades SA-60-1t6) 0.307 1.045 0.009") 1.161") 1.50 2.2(6) 13(6) 535.0 259 6

Palisades SA-240-11 ) 0.307 1.045 0.009") 1.161(') 2.38 3.5(6) 14(6) 535.7 280.1 HB2 T 0.34 0.66 0.021(2) 0.98(2) 3.87 7.27(2) 8(2) 547 289.1 HB2 V 0.34 0.66 0.021(2) 0.98(2) 0.530 3.18(2) 3(2) 547 208.8 HB2 X 0.34 0.66 0.021(2) 0.98(2) 4.49 20.39(2) 20)2) 547 265.6 IP2 V 0.20 1.03 0.019(s) 1.63(') 0.492 8.622(') 8(s) 524 197.5 IP2 Y 0.20 1.03 0.019(3) 1.63(') 0.455 2.337(') 2B(5) 529.1 193.9 5

IP3 T 0.16 1.12 0.019414 1.18(4) 0.263 1.342(') 1A( ) 539.4 149.8 IP3 Y 0.16 1.12 0.019(4) 1.18 (4) 0.692 3.3(') 3(5) 539.5 171.1 IP3 Z 0.16 1.12 0.019(4) 1.18(4) 1.04 5.56605) 5(5) 538.9 228.3 IP3 X 0.16 1.12 0.019(4) 1.18(4) 0.874 15.601(5) 12B(5 ) 539.7 192.5 Notes:

1. Phosphorous and manganese content obtained from Palisades Capsule SA-240-1 analysis (Reference 15).

2. Surveillance material data obtained from H. B. Robinson Capsule X analysis (Reference 16).

3. Data not available. Conservative estimate per Reference 1.

4. Phosphorous and manganese content obtained from Indian Point Unit 3 Capsule X analysis (Reference 17).

5. Surveillance capsule information obtained from SIA Report No. 0901132.401 (Reference 18).

6. Palisades Capsules SA-60-1 and SA-240-1 were installed at the end of cycle 11 in the RV. Capsule-accrued EFPY values were calculated using EFPD data provided in Table 5-2.

7. Values are reported to the precision level calculated in Reference 13.

WCAP- 17628-NP June 2014 Revision I

6-4 Table 6-3 Surveillance Data for Palisades Weld Wire Heat 27204 Chemical Composition Fluence Withdraw Time-Averaged Measured AT30 Plant Capsule Cu Ni P Mn (xl019 n/cm 2, E EFPY Coolant Temperature Transition

[wt%] [wt%] [%] [%] > 1.MeV) Cycle Temperature (°F)(4) 4 f)( )

DC1 Y 0.198 0.999 0.016(l) 1.360") 1.05 5.87... 5"1) 542 232.59 DC1 S 0.198 0.999 0.016(l) 1.360(') 0.284 1.25(" 1(1) 544 110.79 Palisades SA-240-1 3" 0.194 1.067 0.009(21 1.281(2) 2.38 3.5(3 14 535.7 267.8 3

Palisades SA-60-1( ) 0.194 1.067 0.009(2) 1.281(2) 1.50 2.20) 13(3) 535.0 253.1 DC1 V 0.198 0.999 0.0160') 1.360(') 1.37 14.27(* 110) 541.5 201.07 Notes:

1. Surveillance material data obtained from Diablo Canyon Unit 1 Capsule V analysis (Reference 19)

2. Phosphorous and manganese content obtained from Palisades Capsule SA-240-1 analysis (Reference 15).

3. Palisades Capsules SA-60-1 and SA-240-1 were installed at the end of cycle I1 in the RV. Capsule-accrued EFPY values were calculated using EFPD data provided in Table 5-2.

4. Values are reported to the precision level used in Reference 13.

WCAP-17628-NP June 2014 Revision I

7-1 7 INSERVICE INSPECTION DATA Three 10-year ISIs have been performed for the Palisades RV welds. The most recent ISI (Interval 3, Period 3) was conducted in accordance with the ASME Code, Section Xl, Appendix VIII, 2001 Edition with the 2003 Addenda (Reference 5) as modified by 10 CFR 50.55a(b)(2)(xiv., xv, and xvi).

Examinations of the Category B-A welds were governed by the ASME Code Section XI, 1989 Edition, no addenda. An ASME Section XI., Performance Demonstration Initiative (PDI) qualified supplement 4 examination was completed using a dual 45 degree ultrasonic longitudinal wave transducer operated at a center frequency of 3.6 MHz. The contact examination from the reactor vessel inside diameter (ID) uses water as a couplant. This PDI technique is effective in detection of planar flaws oriented axially or circumferentially in the reactor vessel.

All of the recorded indications were assessed in accordance with the criteria in the ASME Code Section XI, 1989 Edition, no addenda, Article IWB-3000, Paragraph IWB-3500; and were found to be within the allowable limits specified with no further evaluation required. Tables 7-1 and 7-2 contain data on the welds and the characteristics of any indications within the beltline and extended beltline regions of the RV obtained from the latest ISI report (Reference 11).

A supplemental eddy current examination was performed at the location of ultrasonic identified flaws reported within 1.0" of the reactor vessel ID surface to validate that these flaws do not open to the vessel inside surface. The eddy current probe for this inspection was the Plus-Point coil operated at 250 KHz in the Driver/Pick-up mode. This technique is normally used in reactor vessel nozzle examinations to detect and characterize surface flaws in the clad material. This examination was conducted to detect flaws oriented axially or circumferentially in the reactor vessel. No surface indications were revealed by the eddy current examinations in the locations where ultrasonic flaws were reported.

Table 7-1 Reactor Vessel IS1 History for Palisades Beltline and Extended Beltline Materials Number of Number of Weld151 Region and Component Description DateeLast Recordable Reportable No. Inspected Indications FlawsI1 )

1-1 12A Upper Shell Longitudinal Weld at 900 2014 6 None 1-1 12B. Upper Shell Longitudinal Weld at 2100 2014 4 None I-I 12C Upper Shell Longitudinal Weld at 3300 2014 3 None 2-112A Intermediate Shell Longitudinal Weld at 2700 2014 2 None 2-112B Intermediate Shell Longitudinal Weld at 300 2014 18 None 2-112C Intermediate Shell Longitudinal Weld at 1500 2014 No Indications None 3-112A Lower Shell Longitudinal Weld at 900 2014 No Indications None 3-112B Lower Shell Longitudinal Weld at 210' 2014 1 None 3-1 12C Lower Shell Longitudinal Weld at 3300 2014 1 None 8-112 Upper to Intermediate Shell Circ. Weld 2014 1 None 9-112 Intermediate to Lower Shell Circ. Weld 2014 6 None Note:

1. Flaws that are reportable are those that exceed the ASME Section XI Table IWB-35 10-1 acceptance standards.

WCAP- 17628-NP June 2014 Revision I

7-2 Table 7-2 ISI Information for Reactor Vessel Beltline and Extended Beltline Flaws for Palisades Weld Weld Table UT Beam t (in.) L b Weld ISI Type Width Indication No. No. Direction)ti (in.) S (in.) 2a (in.) a (in.) IWB-3510-1 (A or C)") (in.) Disposition 1 CW 10.86 0.60 0.512 0.125 0.063 Allowable 2 CW 10.86 0.60 1.372 0.125 0.063 Allowable I-I12A A 1.44 3 CW 10.86 0.60 0.835 0.125 0.0625 Allowable 4 CCW 10.86 1.50 10.11 0.417 0.2085 Allowable 5 CCW 10.86 2.00 10.86 (2) 0.125 Allowable 6 DN 10.86 1.60 0.103 0.176 0.088 Allowable 1 CCW 10.86 0.6 0.08 0.08 0.040 Allowable 1-112B A 1.44 2 CW 10.86 0.6 0.50 0.10 0.050 Allowable 3 CW 10.86 21.75 2.50 0.41 0.203 Allowable 4 CCW 10.86 1.75 3.04 0.16 0.081 Allowable I UP 10.86 3.0 9.70 0.125 0.063 Allowable I-112C A 1.44 2 CCW 10.86 1.6 0.36 0.118 0.059 Allowable 3 CW 10.86 2.25 2.74 0.081 0.041 Allowable 2-112A A 1.44 1 CW 8.74 1.1 0.34 0.14 0.07 Allowable 2 CW 8.74 0.6 0.34 0.13 0.06 Allowable I CCW 8.74 0.75 3.23 0.26 0.13 Allowable 2 CCW 8.74 0.6 0.30 0.06 0.03 Allowable 3 CCW 8.74 0.6 0.24 0.06 0.03 Allowable 4 CCW 8.74 0.6 0.36 0.06 0.03 Allowable 5 CCW 8.74 0.6 0.34 0.06 0.03 Allowable 6 CCW 8.74 0.6 0.24 0.06 0.03 Allowable 7 CCW 8.74 0.6 0.22 0.06 0.03 Allowable 8 CCW 8.74 0.6 0.42 0.06 0.03 Allowable 2-112B A 1.44 9 CCW 8.74 0.6 0.44 0.06 0.03 Allowable 10 CCW 8.74 0.6 0.44 0.06 0.03 Allowable 11 CCW 8.74 0.6 0.42 0.06 0.03 Allowable 12 CCW 8.74 0.6 0.28 0.06 0.03 Allowable 13 CCW 8.74 0.6 0.32 0.06 0.03 Allowable 14 CCW 8.74 0.6 0.26 0.06 0.03 Allowable 15 CCW 8.74 0.6 0.32 0.06 0.03 Allowable 16 CCW 8.74 0.6 0.34 0.06 0.03 Allowable 17 CCW 8.74 0.6 0.30 0.06 0.03 Allowable 18 CCW 8.74 0.6 0.22 0.06 0.03 Allowable 3-112B A 1.44 1 CCW 8.74 1.25 3.69 0.37 0.185 Allowable 3-112C A 1.44 1 CCW 8.74 1.6 0.40 0.26 0.129 Allowable 8-112 C 1.31 1 CCW 8.74 0.6 0.16 0.20 0.10 Allowable WCAP-17628-NP June 2014 Revision I

7-3 Table 7-2 IS[ Information for Reactor Vessel Beltline and Extended Beltline Flaws for Palisades Weld Weld Table Type Width Indication UT Beam L Snb.)e No.ISI Weld No. Direction t (in.) S (in.) 2a (in.) a (in.) IWB-3510-1 (A or C)(') (in.)

Disposition 1 UP 8.74 5.1 0.037 0.125 0.063 Allowable 2 UP 8.74 12.6 0.213 0.125 0.063 Allowable 9-112 C 1.31 3 UP 8.74 14.6 0.350 0.125 0.0625 Allowable 4 UP 8.74 3.1 0.056 0.125 0.0625 Allowable 5 UP 8.74 1.6 0.095 0.125 0.0625 Allowable 6 UP 8.74 5.6 0.174 0.125 0.0625 Allowable Notes:

1. A = Axial, C = Circumferential, CW = Clockwise, CCW = Counterclockwise, UT = Ultrasonic

2. Indication was reported as an outside surface indication in WesDyne ISI Report (Reference 11).

WCAP-17628-NP June 2014 Revision 1

8-1 8 DETERMINATION OF RTMAX-X VALUES FOR ALL BELTLINE AND EXTENDED BELTLINE REGION MATERIALS 8.1 CALCULATION OF RTMAXX VALUES Using the Alternate PTS Rule methodology described in Section 3.1, RTMAx-x values were generated for the beltline and extended beltline region materials of the Palisades RV for fluence values at the EOLE (42.1 EFPY). These values were calculated using RV beltline and extended beltline material copper, nickel, phosphorus, and manganese content, unirradiated RTNDT, projected EOLE neutron fluence values, and time-weighted averaged reactor cold-leg temperature provided in Sections 4 and 5.

The Palisades RV wall thickness transitions between the intermediate and upper shell courses. The upper shell course has a total wall thickness (including the cladding) of 11.11 inches; whereas, the lower and intermediate shell courses have a total wall thickness of 8.99 inches. Since the RTMAXx limits vary depending on vessel wall thickness per 10 CFR 50.61 a (Reference 1), the Palisades RTMAX.X values were determined for the two vessel wall thicknesses within the beltline and extended beltline regions of the RV and compared to the applicable PTS screening criteria.

Tables 8-1 through 8-3 summarize the results of Equations 1 through 7 to calculate RTMAX-X for the Palisades axial welds, plates, and circumferential welds. The calculated RTMAX.X values and applicable PTS screening criteria are provided in Table 8-4.

Table 8-1 RTMAX-AW Calculation Results for Palisades at 42.1 EFPY Fluence Total Weld Group RV Location Material (X1019 n/cm2, AT3 0 RTNDTlu RTNDT RTMAX-AwV)

Heat No. E > 1.0MeV) (OF) (OF) (OF) (OF)

Upper Shel LgUpper Shell W5214 0.09707 154.4 -56 98.4 Longitudinal Longitudinal Weld Weld 1-1 12A Upper Shell Plate C-1308-2 0.09707 95.6 19 114.6 135.2 Upper Shell Plate C-1281-1 0.09707 125.2 10 135.2 Upper Shell Upper Shell W5214 0.09707 154.4 -56 98.4 Longitudinal Longitudinal Weld Upper Shell Plate C-1279-2 0.09707 99.8 10 109.8 114.6 WeldniI 12B Upper Shell Plate C-1308-2 0.09707 95.6 19 114.6 Upper Shell Upper Shell W5214 0.09707 154.4 -56 98.4 Longitudinall ongitudinal Weld35.2 Weld ]-i 12C Upper Shell Plate C-1279-2 0.09707 99.8 10 109.8 Upper Shell Plate C-1281-1 0.09707 125.2 10 135.2 Intermediate SelLntuiaWed Intermediate Shell W5214 2.161 293.0 -56 237.0 Shell Long;itudinal W eld23 .

Longitudinal Intermediate Shell Plate A-0313-2 2.161 191.4 -30 161.4 237.0 Weld 2-112A Intermediate Shell Plate C-1279-1 2.161 187.4 -5 182.4 Intermediate Intermediate Shell W5214 2.161 293.0 -56 237.0 Shell Longitudinal Weld Longitudinal Intermediate Shell Plate C-1279-3 2.161 189.0 -5 184.0 237.0 Weld 2-112B Intermediate Shell Plate C-1279-1 2.161 187.4 -5 182.4 WCAP- 17628-NP June 2014 Revision I

8-2 Table 8-1 RTMAX-AW Calculation Results for Palisades at 42.1 EFPY Fluence Total Weld Group RV Location HeateNo (Xl' 9 n/cm 2 AT,, RTNDTu RTNDT RTMAXoA(F)

E > 1.0MeV) (OF) (OF) (OF) (OF)

Intermediate Intermediate Shell SelLniunaWed W5214 2.161 293.0 -56 237.0 Shell Longitudinal Weld 237.0 Longitudinal Intermediate Shell Plate C-1279-3 2.161 189.0 -5 184.0 Weld 2-112C Intermediate Shell Plate A-0313-2 2.161 191.4 -30 161.4 Lower Shell 34B009 2.161 268.7 -56 212.7 Lower Shell Longitudinal Weld W5214 2.161 293.0 -56 237.0 Longitudinal Lower Shell Plate C-1308-1 2.161 165.7 0 165.7 237.0 Weld 3-112A Lower Shell Plate C-1308-3 2.161 164.0 -30 134.0 Lower Shell 3413009 2.161 268.7 -56 212.7 Lower Shell Longitudinal Weld W5214 2.161 293.0 -56 237.0 Longitudinal Lower Shell Plate C-1308-3 2.161 164.0 -30 134.0 237.0 Weld 3-1 12B Lower Shell Plate B-5294-2 2.161 125.3 -25 100.3 Lower Shell 34B009 2.161 268.7 -56 212.7 Lower Shell Longitudinal Weld W5214 2.161 293.0 -56 237.0 Longitudinal Lower Shell Plate C-1308-1 2.161 165.7 0 165.7 237.0 Lower Shell Plate B-5294-2 2.161 125.3 -25 100.3 Table 8-2 RTMAX.PL Calculation Results for Palisades at 42.1 EFPY 0

Material Fluence AT3o RTNT(-*) Total )

RV LocationHeat No.

HeatNo.

(x10' 9 n/cm, E > 1.0MeV)

(.F)

(OF) ((F0F) RTN(-r (OF_)(OF) RT_______F Upper Shell Plate D-3802-1 C-1279-2 0.1529 112.0 10 122.0 Upper Shell Plate D-3802-2 C-1308-2 0.1529 108.2 19 127.2 149.8 Upper Shell Plate D-3802-3 C-1281-1 0.1529 139.8 10 149.8 Intermediate Shell Plate D-3803-1 C-1279-3 3.429 204.4 -5 199.4 Intermediate Shell Plate D-3803-2 A-0313-2 3.429 206.5 -30 176.5 Intermediate Shell Plate D-3803-3 C-1279-1 3.429 202.6 -5 197.6 199.4 Lower Shell Plate D-3804-1 C-1308-1 3.429 180.6 0 180.6 Lower Shell Plate D-3804-2 C-1308-3 3.429 178.7 -30 148.7 Lower Shell Plate D-3804-3 B-5294-2 3.429 139.5 -25 114.5 A WCAP-17628-NP June 2014 Revision 1

8-3 Table 8-3 RTNAx-cw Calculation Results for Palisades at 42.1 EFPY FluenceToa Material Fune-~ AT,, RTNDT Total RT(Ax-cwF)

TA-Wi Weld Group RV Location HeatHetN.

No. E>

(x10'91.0MeV) n/cm, .O (O)

E )(.OF)

(°F) (OF)

(OF) (u) RT )RTNDT (F Upper to Intermediate 34B009 0.1529 163.2 -56 107.2 Shell Circ. Weld Upper to Upper Shell Plate C-1279-2 0.1529 112.0 10 122.0 Intermediate Upper Shell Plate C-1308-2 0.1529 108.2 19 127.2 Shell Circ. 149.8 Weld Upper Shell Plate C-1281-1 0.1529 139.8 10 149.8 8-112 Intermediate Shell Plate C-1279-3 0.1529 123.0 -5 118.0 Intermediate Shell Plate A-0313-2 0.1529 125.3 -30 95.3 Intermediate Shell Plate C-1279-1 0.1529 122.1 -5 117.1 Intermediate Inermediate to Lower Wer 27204 3.429 303.6 -56 247.6 ShellCirc._Weld ___ ____

Intermediate to Intermediate Shell Plate C- 1279-3 3.429 204.4 -5 199.4 Lower Shell Intermediate Shell Plate A-0313-2 3.429 206.5 -30 176.5 Circ. Weld Intermediate Shell Plate C-1279-1 3.429 202.6 -5 197.6 9-112 Lower Shell Plate C-1308-1 3.429 180.6 0 180.6 Lower Shell Plate C-1308-3 3.429 178.7 -30 148.7 Lower Shell Plate B-5294-2 3.429 139.5 -25 114.5 Table 8-4 RTNAx.x values for Palisades at 42.1 EFPY Lower and Intermediate Shell RegionUpper Shell Region T L9ion (10.5 in. < TWALL -- 11.5 in.)".

(TwALL:-- 9.5 in.)")*

Palisades 10 CFR 50.61a 10 CFR 50.61 a Screening Criteria Screening Criteria Axial Weld--RTMAXAW (OF) 237.0 269 135.2 222 Plate-RTMAXPL (OF) 199.4 356 149.8 293 Axial Weld and Plate-RTMAXAW + 436.4 538 285.0 445 RTMAX-PL (OF) I I I I Circumferential Weld-RTMAX.CW (OF) 247.6 312 149.8 269 Note:

1. TWALL is the RV wall thickness including the cladding.

The RTMAX-X values calculated for Palisades are less than PTS screening criteria and therefore meet this requirement of the Alternate PTS Rule.

8.2 SURVEILLANCE CAPSULE DATA STATISTICAL CHECKS As discussed in Section 3.2, the Alternate PTS Rule (Reference 1) requires that surveillance data that could affect the calculation of AT 30 be evaluated. This requirement is only applicable for materials for which three (3) or more points of surveillance data exist at three (3) or more unique fluence values.

WCAP- 17628-NP June 2014 Revision I

8-4 The Palisades and sister plant surveillance materials, along with their calculated values of AT 30, can be seen in Table 8-5. This table includes a list of the tested and analyzed capsules for each material.

Table 8-5 Surveillance Capsule Materials for Palisades No Region and Material Fluence

11) Component Identification Plant Capsule Direction (X10' 9 n/cm 2, Calculated (Heat No.) E > 1.0MeV) AT 30 (OF)" 2 '

Description Palisades A-240 Longitudinal 4.09 233.9 Palisades W-290 Longitudinal 0.938 163.5 1 Upper and D-3803-1 D-3803- 1 Palisades W-1 10 Longitudinal 1.64 180.4 4 Intermediate D-3803-3 Palisades W-100 Longitudinal 2.09 189.0 6 Shell Plates Palisades A-240 Transverse 4.09 233.9 Palisades W-290 Transverse 0.938 163.5 Palisades W-100 Transverse 2.09 189.0 Palisades SA-60-1 N/A 1.50 322.7 10 Palisades SA-240-1 N/A 2.38 340.8 11 HB2 T N/A 3.87 258.7 12 HB2 V N/A 0.530 200.3 13 Surveillance 1-I 12A, 2-1 12A, B, B, &

&C C HB2 X N/A 4.49 263.6 14 Program 2IP2 3-112A, B, & C V N/A 0.492 215.0 15 Weld Metal IP2 Y N/A 0.455 188.5 16 (W5214) IP3 T N/A 0.263 107.2 17 IP3 Y N/A 0.692 173.0 18 IP3 Z N/A 1.04 195.4 IP3 X N/A 0.874 197.7 DCI Y N/A 1.05 213.7 Surveillance 9-112 DCI S N/A 0.284 133.9 20 Program Palisades SA-240-1 N/A 2.38 247.2 Weld Metal (27204) Palisades SA-60-1 N/A 1.50 229.0 DCI V N/A 1.37 229.2 Note:

1. Numbers listed correspond to the item numbers from Table 4-1

2. Values calculated using Alternate PTS Rule correlation as described in Sections 3.1 and 3.2 All of the materials listed have at least three data points at three or more different neutron fluences; therefore, it needed to be determined if the surveillance data showed a significantly different trend than the embrittlement model predicts per the Alternate PTS Rule (Reference 1). Using the methodology described in Section 3.2, a Mean Deviation Test, a Slope Deviation Test, and an Outlier Deviation Test were conducted for each surveillance material. The inputs for the surveillance data evaluations, including the measured values of AT 30 , are provided in Tables 6-1 through 6-3 for the three surveillance materials.

The results of the evaluations are shown in Tables 8-6, 8-7, and 8-8.

June 2014 WCAP- 17628-NPP WCAP-1!7628-N June 2014 Revision I

8-5 Table 8-6 Surveillance Data Evaluation for Palisades Base Metal Heat C-1279 Plant Capsule Pln asle Direction Drcin Fluence Log of Residual 'Y" (x - xa,,g) r* (r/sigma)

Palisades A-240 Longitudinal 19.61 -28.9 0.104 -1.37 Palisades W-290 Longitudinal 18.97 -8.5 0.100 -0.40 Palisades W-1 10 Longitudinal 19.21 -0.4 0.005 -0.02 Palisades W-100 Longitudinal 19.32 -29.9 0.001 -1.41 Palisades A-240 Transverse 19.61 -28.9 0.104 -1.37 Palisades W-290 Transverse 18.97 11.5 0.100 0.54 Palisades W-100 Transverse 19.32 -46.5 0.001 -2.19 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard Deviation 21.2 Slope (m) -52.84 Largest r* 0.54 (sigma)

Mean Deviation -18.8 Standard Error of Fit 16.13 Largest allowable r* 2.98 Maximum Mean 18.7 Standard Error of 24.99 Pass/Fail? Pass Residual Slope Pass/Fail? Pass T-Statistic -2.11 Second largest r* -0.02 Critical T-Statistic 3.36 Second largest 2.00 allowable r*

Pass/Fail? Pass Pass/Fail? Pass June 2014 WCAP-l 7628-NP WCAP-17628-NP June 2014 Revision I

8-6 Table 8-7 Surveillance Data Evaluation for Palisades Weld Wire Heat W5214 2

Plant Capsule Direction Fluence Residual r' (x - Xav.g) r* (r/sigma)

Palisades SA-60-1 N/A 19.18 -63.7 0.0283 -2.41 Palisades SA-240-1 N/A 19.38 -60.7 0.1360 -2.30 HB2 T N/A 19.59 30.4 0.3363 1.15 HB2 V N/A 18.72 8.5 0.0804 0.32 HB2 X N/A 19.65 2.0 0.4153 0.08 IP2 V N/A 18.69 -17.5 0.0997 -0.66 IP2 Y N/A 18.66 5.4 0.1223 0.21 1P3 T N/A 18.42 42.6 0.3455 1.61 IP3 Y N/A 18.84 -1.9 0.0281 -0.07 IP3 Z N/A 19.02 32.9 0.0001 1.24 IP3 X N/A 18.94 -5.2 0.0044 -0.20 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard Siama) Deviation 26.4 Slope (m) -23.22 Largest r*

(sigma) 1.61 Mean Deviation -2.5 Standard Error of Fit 35.09 Largest allowable r* 3.12 Maximum Mean 18.5 Standard Error of 27.77 Pass/Fail? Pass Residual Slope Pass/Fail? Pass T-Statistic -0.84 Second largest r* 1.24 Second largest 21 Critical T-Statistic 2.82 2.19 1_ allowable r*

Pass/Fail? Pass Pass/Fail? Pass June 2014 WCAP- 17628-NP WCAP-1i7628-NP June 2014 Revision I

8-7 Table 8-8 Surveillance Data Evaluation for Palisades Weld Wire Heat 27204 Plant Plan Capule Capsule irecion Direction Fluence Log of Residual 'r' (X- Xavg) 2 r* (r/sigma)

DC1 Y N/A 19.02 18.9 0.0001 0.72 DC1 S N/A 18.45 -23.1 0.3358 -0.87 Palisades SA-240-1 N/A 19.38 20.6 0.1182 0.78 Palisades SA-60-1 N/A 19.18 24.1 0.0205 0.91 DCi V N/A 19.14 -28.2 0.0108 -1.07 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard Deviation 26.4 Slope (m) 42.78 Largest r* 0.91 (sigma)

Mean Deviation 2.5 Standard Error of Fit 24.30 Largest allowable r* 2.88 Maximum Mean 27.5 Standard Error of 27534.87 Pass/Fail? Pass Residual Slope Pass/Fail? Pass T-Statistic 1.23 Second largest r* 0.78 Critical T-Statistic 4.54 Second largest 1.84 1 allowable r*

Pass/Fail? Pass Pass/Fail? Pass As shown in Tables 8-6 through 8-8, the surveillance results for the plate and weld surveillance materials satisfy the criteria in the Alternate PTS Rule for all three tests. Therefore, the use of Equations (5) to (7) in the Alternate PTS Rule (Reference I) for calculation of AT30 is acceptable for Palisades.

8.3 REACTOR VESSEL BELTLINE ISI DATA EVALUATION Per the requirements of Section 3.3, the results of the latest ISI of the Palisades RV were analyzed in detail to ensure that the recorded indications met the acceptance criteria. The RV 1SI data specified in Section 7 indicates that the Category B-A examinations of the Palisades RV beltline and extended beltline region welds have been performed to ASME Section XI, Appendix VIII requirements. At least one inspection has been performed on each weld per Code requirements. Inspection coverage of the welds within the beltline region has been greater than 90%. All indications found in the beltline and extended beltline were reported to be either subsurface or located on the outside surface. In accordance with the Alternate PTS rule, examinations were performed during the latest Palisades RV ISI in locations where flaws were detected within the first 1.0" from the vessel inside diameter to verify that they are not open to the inside surface. Automated eddy current and ultrasonic examination techniques capable of detecting and characterizing service induced cracking of the reactor vessel cladding were used for the inspection of these flaws. No inside surface flaws were found in the beltline or extended beltline welds.

After reviewing the data from the latest ISI (Reference 11), 42 indications with the potential to be located within the beltline and extended beltline regions of the Palisades RV were recorded. Thirty of these indications were within the 1.0 x 1017 n/cm2 neutron fluence region, and therefore within the RV beltline and extended beltline. Twenty-nine of these indications also fall within the inner 3/8th of the RV thickness and are allowable per Table IWB-3510-1 of Section XI of the ASME Code (Reference 5).

WCAP-17628-NP June 2014 Revision I

8-8 Twenty-eight of the indications fall within the inner 1 inch or 10 percent thickness of the RV, whichever is greater; therefore, further evaluation is required to confirm that they satisfy the flaw requirements provided in Section 3.3. These indications are shaded in gray in Table 8-9 and shown in Figure 8-1.

These flaws are all located either adjacent to or within the weld fusion lines. However, for the purposes of this evaluation, all twenty-eight indications are considered to be in the plate material and are evaluated against the more conservative plate flaw limits.

In order to scale the number of allowable weld flaws per the Alternate PTS Rule, the total inspected plate area and weld length within the beltline and extended beltline regions of the RV were calculated as shown in Table 8-10. The acceptance criteria for plate and weld flaws based on Palisades RV dimensions are shown in Tables 8-11 and 8-12, respectively. After identifying the size, location, and orientation of the indications, it was determined that they satisfy the Alternate PTS Rule requirements provided in Section 3.3 because the number of actual flaws is less than the number allowable for all flaw size increments.

Note that RV TWCF calculations are not needed due to the following factors:

• The flaw density and size in the inspection volume are less than the limits in Tables 8-11 and 8-12;

" There are no axial flaws that penetrate through the clad into the low alloy steel reactor vessel shell, at a depth equal to or greater than 0.075 inches in through-wall extent from the clad-to-base metal interface; and

• All flaws between the clad-to-base metal interface and three-eighths of the vessel thickness meet the size allowable in ASME Code,Section XI, Table IWB-3510-1.

Table 8-9 Reactor Vessel IS! Evaluation for Potential Beltline and Extended Beltline Flaws for Palisades Weld Idicatin TWE~~ ~2 Locaion~ Within InrFa W eld Indication TW E "' Location(2 Beltline Beithin or Inner (Inner

/ 0t Flaw Flaw Limits E auto (1/10)t Evaluation ISI No. No. (in.) (Plate/Weld) Extended (3/8)t? or 1"? Orientation Required?

Beltline?

1 0.125 Plate No Yes Yes Axial No 2 0.125 Plate No Yes No Axial No 3 0.125 Plate No Yes Yes Axial No 1-112A 4 0.417 Plate No No No Axial No 5 0.125 Plate No No No Axial No 6 0.176 Plate Yes Yes Yes Circ. Yes 1 0.08 Plate No Yes Yes Axial No 2 0.10 Plate No Yes Yes Axial No 1-112B 0.41 Plate No Yes No Axial No 4 0.16 Plate No Yes No Axial No 1 0.125 Plate No No No Circ. No 1-112C 2 0.118 Plate No Yes Yes Axial No 0.081 Plate No Yes No Axial No 2 1 0.14 Plate Yes Yes Yes Axial Yes 2-112A 0.13 Plate Yes Yes Yes Axial Yes WCAP- 17628-NP June 2014 Revision 1

8-9 Table 8-9 Reactor Vessel ISI Evaluation for Potential Beltline and Extended Beltline Flaws for Palisades Weld Indication TWE"'L Location Beltlineetleor(

or Inner Inner Flaw Flaw Limits 1/l10)t Evaluation ISI No. No. (in.) (Plate/Weld) Extended (3/8)t? or I"? Orientation Required?

Beltline?

1 0.26 Plate Yes Yes No Axial No 2-112B 3-112B 1 0.37 Plate Yes No No Axial No 3-112C 1 0.26 Plate Yes Yes Yes Axial Yes 8-112 1 0.20 Plate Yes Yes Yes Axial Yes 1 0.125 Plate Yes Yes Yes Circ. Yes 2 0.125 Plate Yes Yes Yes Circ. Yes 3 0.125 Plate Yes Yes Yes Circ. Yes 9-112 4 0.125 Plate Yes Yes Yes Cir . Yes 5 0.125 Plate Yes Yes Yes Circ. Yes 6 0.125 Plate Yes Yes Yes Circ. Yes Note:

1. TWE is the same as the dimension "2a" for subsurface indications or "a" for surface indications from Table 7-2.

2. For the purposes of this evaluation, all indications were considered to be in the plate material and are be evaluated against the more conservative plate flaw limits.

WCAP- 17628-NP June 2014 Revision I

8-10 Table 8-10 Inspection Length and Area for Palisades Inside Diameter of RV (to clad surface) 172.2 in.

Cladding Thickness 0.25 in.

Inside Diameter of Weld Inspection Volume 172.7 in.

RV Thickness of Upper Shell Region 10.86 in.

RV Thickness of Intermediate and Lower Shell Region 8.74 in.

Number of Upper Shell Axial Welds 3 Number of Intermediate Shell Axial Welds 3 Number of Lower Shell Axial Welds 3 Number of Circumferential Welds 2 Height of Upper Shell Region (in extended beltline) 15.78 in.

Height of Intermediate Shell Region 97.31 in.

Height of Lower Shell Region (in extended beltline) 79.83 in.

Width of Upper Shell Axial Welds 1.44 in.

Width of Intermediate Shell Axial Welds 1.44 in.

Width of Lower Shell Axial Welds 1.44 in.

Width of Circumferential Welds 1.31 in.

Total Plate Area (in beltline and extended beltline)°) 14500 in. 2 Total Weld Length (in beltline and extended beltline)(1) 1600 in.

Note:

I. The total calculated plate area and weld length were rounded down, which is conservative for the calculation of the 10 CFR 50.61a weld and plate flaw limits (See Tables 8-11 and 8-12).

Table 8-11 Alternate PTS Rule Allowable Number of Flaws in Plates and Forgings Scaled for Palisades Scaled maximum number of flaws per 14500 square-inches Through-Wall Extent, TWE (in.) of inside surface area in the inspection volume that are Number of Flaws greater than or equal to TWEMTN and less than TWENAX. Number/ofrFla This flaw density does not include underclad cracks in TWEN, TWEMAx forgings 0 0.075 No Limit 17(17/0) 0.075 0.375 117 11 (4/7) 0.125 0.375 46 11 (4/7) 0.175 0.375 13 3(2/1) 0.225 0.375 5 1(1/0) 0.275 0.375 2 0 (0/0) 0.325 0.375 1 0 (0/0) 0.375 Infinite 0 0 (0/0)

WCAP- 17628-NP June 2014 Revision 1

8-11 Table 8-12 Alternate PTS Rule Allowable Number of Flaws in Welds Scaled for Palisades Through-Wall Extent, TWE (in.) Scaled maximum number of flaws per 1600 inches of weld Number of Flaw length in the inspection volume that are greater than or equal (Axial/Circ.)

to TWEMjN and less than TWEMAX.

TWEMIN TWEMAX 0 0.075 No Limit 0 (0/0) 0.075 0.475 267 0 (0/0) 0.125 0.475 146 0(0/0) 0.175 0.475 37 0 (0/0) 0.225 0.475 14 0 (0/0) 0.275 0.475 7 0 (0/0) 0.325 0.475 5 0 (0/0) 0.375 0.475 3 0 (0/0) 0.425 0.475 2 0 (0/0) 0.475 Infinite 0 0 (0/0)

WCAP-17628-NP June 2014 Revision I

8-12 13600 Figure 8-1 Weld and Plate Indication Map for Palisades Beltline and Extended Beltline(3)

Notes:

1. 117.94" corresponds to the upper extent of the extended beltline region. See Table 4-2 for references used to determine this location.

2. 310.86" corresponds to the lower extent of the extended beltline region. See Table 4-2 for references used to determine this location.

3. Map is not drawn to scale and indication locations are approximate. Numbers in parentheses correspond to "No." column in Table 4-1. Dimensions are measured downward from the RV flange surface.

WCAP- 17628-NP June 2014 Revision 1

9-1 9 CONCLUSION After conducting this evaluation, it was concluded that the Palisades reactor pressure vessel is acceptable per the Alternate PTS Rule acceptance criteria. As shown in Section 8.1, all of the beltline and extended beltline region materials in the Palisades RV have EOLE (42.1 EFPY) RTMAX.X values below the screening criteria values. After conducting surveillance data statistical tests, it was determined that the surveillance data satisfied the Alternate PTS Rule requirements. Lastly, a review of the latest RV ISI report for Palisades showed that the flaw density and size distribution is acceptable per the Alternate PTS Rule requirements.

WCAP-17628-NP June 2014 Revision I

10-1 10 REFERENCES I. Code of Federal Regulations, 10 CFR Part 50.61 a, "Alternate fracture toughness requirements for protection against pressurized thermal shock events," U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No. L1dated January 4, 2010, and No. 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010.

2. Code of Federal Regulations, 10 CFR Part 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events," U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 60, No. 243, dated December 19, 1995 and last updated on January 4, 2010.

3. Westinghouse Report, WCAP-17403-NP, Revision 1, "Palisades Nuclear Power Plant Extended Beltline Reactor Vessel Integrity Evaluation," January 2013.

4. U.S. Nuclear Regulatory Commission NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March 2010.

5. ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition with no addenda up to and including the 2007 Edition with the 2008 Addenda, American Society of Mechanical Engineers, New York.

6. Branch Technical Position 5-3, "Fracture Toughness Requirements," Revision 2, contained in Chapter 5 of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LRW Edition," March 2007.

7. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NB, Section NB-2300, "Fracture Toughness Requirements for Material." American Society of Mechanical Engineers, New York.

8. Westinghouse Report, WCAP-15353-Supplement 2-NP, Revision 0, "Palisades Reactor Pressure Vessel Fluence Evaluation," July 2011.

9. Westinghouse Report., WCAP-17341-NP. Revision 0, "Palisades Nuclear Power Plant Heatup and Cooldown Limit Curves for Normal Operation and Upper-Shelf Energy Evaluation," February 2011.

10. Combustion Engineering Report P-PENG-ER-006, Revision 0, "The Reactor Vessel Group Records Evaluation Program Phase 11 Final Report for the Palisades Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," Combustion Engineering, Inc., October 1995.

II. WesDyne ISI Report, "Entergy Palisades Unit 1 Nuclear Power Plant 10 Year Reactor Vessel Inservice Inspection," February 2014

12. SE-REA-96-122, "Transmittal of the Updated Palisades Reactor Vessel Fluence Submittal Response to NRC Request for Additional Information," June 21, 1996.

WCAP-17628-NP June 2014 Revision 1

10-2

13. Structural Integrity Associates, Inc. Report No. 1000915.401, Revision 1, "Revised Pressurized Thermal Shock Evaluation for the Palisades Reactor Pressure Vessel," November 12, 2010.

14. BWXT Services., Inc. Report, "Analysis of Capsule W-100 from the Nuclear Management Company Palisades Reactor Vessel Material Surveillance Program," February 2004.

15. Framatome ANP Report BAW-2398, "Test Results of Capsule SA-240-1 Consumers Energy Palisades Nuclear Plant," May 2001.

16. Westinghouse Report, WCAP-15805, Revision 0, "Analysis of Capsule X from the Carolina Power & Light Company H.B. Robinson Unit 2 Reactor Vessel Radiation Surveillance Program,"

March 2002.

17. Westinghouse Report, WCAP-16251-NP, Revision 0, "Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," July 2004.

18. Structural Integrity Associates, Inc. Report No. 0901132.401, Revision 0, "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," April 20, 2010.

19. Westinghouse Report, WCAP-15958, Revision 0, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit I Reactor Vessel Radiation Surveillance Program,"

January 2003.

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