ML110730084

1001026.401, Rev 1, Basis for Period of Validity of the Palisades Pressure-Temperature (P-T) Limit Curves, Attachment 6 to Pnp 2011-016

ML110730084
Person / Time
Site:	Palisades
Issue date:	11/12/2010
From:	Griesbach T J Structural Integrity Associates
To:	Office of Nuclear Reactor Regulation
References
PNP 2011-016 1001026.401, Rev 1
Download: ML110730084 (148)

Text

ATTACHMENT 6 Structural Integrity Associates, Inc. Report No. 1001026.401 Basis for Period of Validity of the Palisades Pressure-Temperature (P-T) Limit Curve 147 Pages Follow Report No.: 1001026.401 Revision No.: 1 Project No.: 1001026 November 2010 Basis for Period of Validity of the Palisades Pressure-Temperature (P-T) Limit Curves Prepared for: Entergy Nuclear Corp.Palisades Nuclear Power Plant Covert, Michigan Prepared by: Structural Integrity Associates, Inc.San Jose, California Prepared by: Timothy J. Gfiesbach Date: 11/12/10/I- 1ýý P *W Reviewed by: Danen Heath Date: 11/12/10 Date: 11/12/10 Reviewed by: Vikram Marthandam Timothy J. Griesbach Approved by: Date: 11/12/10 U Structural Integrity Associates, Inc.

REVISION CONTROL SHEET Document Number: 1001026.401 Title: Basis for Period of Validity of the Palisades Pressure-Temperature (P-T)Limit Curves Client: Entergy Nuclear Corp.SI Project Number: 1001026 Quality Program: [ Nuclear E Commercial Section Pages [Revision Date Comments All 1 -33 0 10/29/2010 Initial Issue Appendix A Appendix B Appendix C Pages 12, 17, 1 11/12/2010 Fixed typo and added one capsule data 21, 22, point for plate heat no. C-1279 23, 32, A6 V Structural integrity Associates, Inc.

EXECUTIVE

SUMMARY

The properties for the Palisades reactor vessel beltline materials have been examined to determine the limiting beltline material and validity date for the current pressure-temperature (P-T) limit curves. From the basis document for the P-T curves, it was determined that a limiting 1/4T RTNDT value of 255.3°F, and a limiting 3/4T RTNDT value of 190.71F for circumferential weld 9-112 were used to calculate the existing heatup, cooldown and LTOP setpoint limits. At the time the P-T curves and LTOP setpoints were calculated, the peak fluence in the limiting circumferential weld of 2.192x10 1 9 n/cm 2 (E > 1 MeV) correlated with the period of validity for the P-T curves.However, new data and information is now available and the assumptions used at that time may no longer be valid. Therefore, a reanalysis of the basis for the P-T curves and LTOP setpoints was performed, including up-to-date materials data and fluence projections, to determine the period of validity of the P-T curves to establish the remaining time before reaching the limit. Using the "old" method with a CF value of 228°F for weld heat No. 27204, the current P-T curves are projected to expire in May 2011. However, by using the "new" method with the latest best estimate chemistry results and a CF value = 226.8°F for the limiting circumferential weld, and considering the most recent vessel fluence calculations, the Palisades P-T limit curves are now projected to be valid until March 2012. The March 2012 validity date applies to the pressure-temperature limits for heatup, cooldown, and the corresponding LTOP setpoint limits for the Palisades operating P-T limit curves.Report No. 1001026.401, Rev. 1 i Structural Integrity Associates, Inc.

Table of Contents 1.0 IN T R O D U C T IO N ....................................................................................................

1 2.0 TECHNICAL APPROACH ......................................................................................

2 2.1 P-T C urve M ethodology

............................................................................

2 2.2 LTOP Setpoint Curve ..........................................................................

6 2.2.1 L TOP Setpoint Methodology

...................................................................

6 3.0 RTNDT CALCULATION METHOD .........................................................................

7 3.1 RTNDT Calculations for Current P-T Curves ...........................................

7 3.2 Improved RTNDT Calculations for use with P-T Curves .................................

8 3.2.1 W eld H eat N o. 2 7204 ................................................................................

8 3.2.1.1 Surveillance Data for Plate Heat No. 2 7204 ......................

..............................

9 3.2.2 Weld Heat No. 34B009 ...........................................................................

10 3.2.3 Weld Heat No. W5214 .............................................................................

11 3.2.3.1 Surveillance Data for Weld Heat No. W5214...............................................

11 3.2.4 Vessel Beltline Plate Materials

..............................................................

12 3.2.4.1 Surveillance Data for Plate Heat No. C-1279 ............................

.........

.........

12 4.0 FLUENCE PROJECTIONS AND FLUENCE METHODOLOGY

................

18 5.0 RTNDT PROJECTIONS WITH REVISED VESSEL MATERIAL P R O P E R T IE S ....................................................................................

19 6 .0 D IS C U S S IO N ..................................................................................................................

19

7.0 REFERENCES

................................................................................

23 Appendix A. Data Credibility Assessment for Weld Heat No. 27204 ............................

A-1 Appendix B. Surveillance Capsule Data for Weld Heat No. 27204 ................................

B-1 Appendix C. Surveillance Capsule Data for Plate Heat No. C-1279 ..............................

C-1 Report No. 1001026.401, Rev. 1 ii V Structural Integrily Associates, Inc.

List of Tables Table 1. Initial R TNDT V alues ......................................................................................................

4 Table 2. Chemical Content and Chemistry Factor (CF) Values .............................................

4 Table 3. C alculated M argin Term s ..............................................................................

5 Table 4. End-of-Cycle 9 Vessel Surface Fluence Values ................................................

5 Table 5. Limit of P-T Curves Vessel Surface Fluence Values .........................................

6 Table 6. Limiting Surface Fluence and 1/4T RTNDT Values for Current Palisades P-T Curves and LTO P Setpoints

...............................................................................

13 Table 7. Limiting Surface Fluence and 3/4T RTNDT Values for Current Palisades P-T Curves and LTO P Setpoints

...............................................................................

13 Table 8. RVID2 Vessel Materials Summary Report for Palisades

..................................

14 Table 9. Evaluation of all Surveillance Capsule Results Containing Weld Heat No. 27204 ...........

15 Table 10. Evaluation of Surveillance Capsule Results Containing Weld Heat No. W5214 ...........

16 Table 11. Evaluation of Surveillance Capsule Results Containing Plate Heat No. C-1279 ............

17 Table 12. Calculated Clad-to-Base Metal Interface Fluence in Palisades Vessel ......................

18 Table 13. Revised Limiting Surface Fluence and 1/4T RTNDT Values for Current Palisades P-T C urves and LTO P Setpoints

.........................................................................

21 Table 14. Revised Limiting Surface Fluence and 3/4T RTNDT Values for Current Palisades P-T Curves and LTO P Setpoints

.......................................................................

22(List of Figures Figure 1. Pressure-Temperature Limits for Heatups ...................................................................

27 Figure 2. Pressure-Temperature Limits for Cooldown ......................................................

28 Figure 3. LTO P Setpoint Lim it .............................................................................

29 Figure 4. Best Fit CF for Weld Heat No. 27204 Surveillance Data ...................................

30 Figure 5. Best Fit CF for all W5214 Surveillance Data ................................

31 Figure 6. Best Fit CF for Base Metal Heat No. C-1279 Surveillance Data ..............................

32 Figure 7. Plot of Residual vs. Fast Fluence for A533B-1 HSST-01/HSST-02 CMM with Companion Materials, the Overall 2-Sigma Scatter is 50'F .................................

33 Report No. 1001026.401, Rev. 1 iii Structural Integrity Associates, Inc.

Basis for Period of Validity of the Palisades Pressure-Temperature (P-T) Limit Curves

1.0 INTRODUCTION

The Palisades plant operating heatup and cooldown pressure-temperature (P-T) limit curves were determined in accordance with 1OCFR50, Appendix G [1]. The actual operating P-T limit curves are calculated in accordance with the methods in the ASME Code for protection against non-ductile failure [2]. The procedure for calculating the allowable operating limit curves for heatup and cooldown conditions is described in the plant Technical Specification bases [3, 7, 9] which specifies that the principles of linear elastic fracture mechanics are used with the lower bound of static, dynamic, and crack arrest for defining the critical reference toughness.

Per the ASME Code procedure, the maximum allowable applied stress intensity factor is computed as a function of the nil-ductility reference temperature (RTNDT) and the operating temperature and the vessel wall temperature gradients during heatup or cooldown conditions.

The predicted RTNDT shift in the (limiting) vessel materials was calculated as a function of fluence using the methods in Regulatory Guide 1.99, Rev. 2 [4]. The corresponding low temperature overpressure protection (LTOP)system setpoints were determined to protect the vessel from overpressurization when the system is cold using similar predictions of the vessel material properties.

These operating limit curves are valid for the period of time corresponding to the vessel condition which was assumed when calculating the curves. This vessel condition can be equated to a maximum adjusted reference temperature (RTNDT), or the curves may be correlated to a peak fluence or EFPY. Because fluence and EFPY may vary with time, the actual expiration date of the curves may also vary depending on the fluence and vessel RTNDT projections.

This report examines the basis for the current Palisades' P-T curves to determine the period of validity using up-to-date information for fluence and RTNDT of the limiting vessel beltline material.

A change in the limiting weld chemistry factor (CF) has been determined using the most recent copper and nickel chemistry values for weld heat No. 27204.In this study, the only parameter that was changed was the CF value corresponding to the best estimate chemistry for the limiting weld material.

The corresponding limiting fluence value was determined, and the date for which the 1/4T and 3/4T adjusted reference temperatures would be reached, in accordance with the basis for the current Palisades P-T limit curves and LTOP setpoints.

Report No. 1001026.401, Rev. 1 1 V Structural Integrity Associates, Inc.

2.0 TECHNICAL APPROACH The Palisades P-T limit curves and power-operated relief valve setting (LTOP) limits were last updated in 1995 [3]. These curves and setpoints were calculated in accordance with approved methods as described in the NRC Standard Review Plan, NUREG-0800, Section 5.3.2 [5].The Palisades operating P-T limit curves are shown in Figure 1 (Figure 3.4.3-1 from the Technical Specification) and Figure 2 (Figure 3.4.3-2 from the Technical Specification)

[7]. The Technical Specification basis states that these revised pressure-temperature limits are applicable to a reactor vessel inner wall fluence of up to 2.192x10 1 9 n/cm 2 (E > 1 MeV) [3]. The basis for the P-T curves and the LTOP setpoints are given in a Consumers Power engineering calculation

[6]. The basis for the Palisades P-T limit curves as described in Reference 6 is summarized here.2.1 P-T Curve Methodology ASME B&PV Code Section III, Appendix G [2] presents a procedure for obtaining the allowable loadings for ferritic pressure-retaining materials in Class 1 components.

The procedure used to calculate the Palisades heatup and cooldown limit curves is as follows: The reactor is subjected to both pressure induced membrane hoop stress and thermal stress due to through-wall thermal gradients.

The ASME Appendix G model uses the following relationship for calculating heatup and cooldown curves: KIr >2 Kim + Kit (1)where: Kir = reference stress intensity factor, ksi 4lin Kim = membrane stress intensity factor Kit = thermal stress intensity factor The following relationships are also found in ASME Appendix G reference, Kim = Mm

• cym Fig. G-2214-1 [2]Kit = Mt

• ATw Fig. G-2214-2 [2]where: Mm = membrane stress-to-stress intensity factor Mt = thermal stress-to-stress intensity factor AT, = temperature difference through the vessel wall, 'F cym = membrane hoop stress Report No. 1001026.401, Rev. 1 2 Structural Integrity Associates, Inc.

Now equation (1) can be rewritten as: Kir > 2 Mm

• a'm + Mt

• ATw (2)K1r is the allowable lower bound reference toughness for ferritic pressure vessel steels based on static, dynamic, and crack arrest critical K, values measured as a function of temperature.

Figure G-22 10-1 from Reference 2 shows the K1r reference toughness curve as a function of (T -RTNDT)which can be expressed by the equation: Kir = 26.78 + 1.233 exp[0.0145 (T-RTNDT + 160)] (3)where: T = temperature at which K1r is permitted, 'F RTNDT = nil-ductility reference temperature for the limiting vessel material, 'F In order to calculate the RTNDT and the reference toughness, the shift (ARTNDT) associated with the projected fluence on the vessel must be determined.

Using the procedures in Regulatory Guide 1.99, Rev. 2 to calculate the shift [4], the adjusted reference temperature (ART) or RTNDT is determined for each of the vessel beltline materials using the following expression:

RTNDT = Initial RTNDT + ARTNDT + Margin (4)where: Margin = Margin term to cover for uncertainties in the value of initial RTNDT and the scatter in the shift Margin-=2*

ýl+ A (5)(l = the standard deviation for the Initial RTNDT (0 F). For non-Linde 80 type welds, if a generic Initial RTNDT value is used, (71 = 17'F, if a measured value is used for the Initial RTNDT, c(y = 0°F [23].CT = the standard deviation for ARTNDT ('F). The values for CY are 28°F for welds and 170 for base metal (plates or forgings)

[23].The Initial RTNDT is the unirradiated nil-ductility reference temperature for the material.

For the vessel beltline welds a generic mean value of-56°F was used for Initial RTNDT, and for the limiting plate an Initial RTNDT of 0°F was measured.

The Palisades vessel beltline materials were grouped by material class and the limiting vessel beltline material was determined for each class [6], and Table 1 contains the corresponding Initial RTNDT values for each class of beltline materials, axial welds, circumferential weld and base metal.Report No. 1001026.401, Rev. 1 3 V Structural Integrity Associates, Inc.

Table 1. Initial RTNDT Values 161 Material Class Initial RTNDT Standard Deviation, al Axial Welds -56 0 F 17 0 F Circ. Weld -56 0 F 17 0 F Base Metal 1 0 0 F 0F Limiting plate material The change in reference temperature (ARTNDT) is the mean value of the adjustment caused by irradiation and is calculated as follows: ARTNDT = (CF) fO.2 8-O01log (f)(6)The chemistry factor (CF, 'F) is a function of copper and nickel content if surveillance data is not available.

CF for the vessel beltline materials was determined from the tables in Regulatory Guide 1.99, Rev. 2 using the copper and nickel contents shown in Table 2.Table 2. Chemical Content and Chemistry Factor (CF) Values [61 Material Class Cu Content Ni Content Chemistry Factor, (wt%) (wt%) CF (OF)Axial Welds 0.20 1.02 225 Circ. Weld 0.208 1.00 228 Base Metal 0.24 0.55 165 The fluence at any depth in the vessel wall, f(101 9 n/cm 2 , E > 1 MeV), is determined as follows: f = fstrj (e-0.2 4 x)(7)where f,,srf is the calculated value of fluence at the vessel inner surface and f is the calculated fluence at a depth, x, into the vessel wall (inches).Report No. 1001026.401, Rev. 1 4 Structural Integrity Associates, Inc.

Table 3 shows the corresponding margin term that was calculated from Equation (5) for each class of materials.

Table 3. Calculated Margin Terms 16]Material Class Margin, IF Axial Welds 66*Circ. Weld 66*Base Metal 34 Note: the margin term for welds was rounded up to 66'F in Ref 6 The fluence values at the end of Cycle 9 or the beginning of Cycle 10 were used as a starting point for the fluence extrapolation.

The fluence values at these locations are shown in Table 4.The fluence-to-peak fluence ratio at the limiting axial weld is determined to be 73.9%. These fluences were extrapolated until the peak fluence reached a value of 2.192x 10 1 9 n/cm 2 , as shown in Table 5. This is the stated limiting fluence value for the P-T curves and LTOP setpoints

[3].Table 4. End-of-Cycle 9 Vessel Surface Fluence Values [6]Fluence (EOC-9) Fluence-to-Peak Material Class (n/cm 2) Fluence Ratio Axial Welds 1.19x10 1 9 0.739 Circ. Weld 1.61x10 1 9 1.0 Base Metal 1.61x10 1 9 1.0 Report No. 100 1026.40 1, Rev. I 5 1 Structural Integrity Associates, Inc.

Table 5. Limit of P-T Curves Vessel Surface Fluence Values Maximum Fluence Fluence-to-Peak Material Class (n/cm 2) Fluence Ratio Axial Welds 1.62xl0 1 9 0.739 Circ. Weld 2.192x10 1 9 1.0 Base Metal 2.192xl0 1 9 1.0 2.2 LTOP Setpoint Curve In accordance with NUREG-0800, Standard Review Plan, 5.2.2. Overpressure Protection

[8], each operating PWR is required to have a low temperature overpressure protection (LTOP) system.Overpressure protection for the reactor coolant pressure boundary during low temperature operation of the plant (startup, shutdown) is ensured by the application of pressure relieving systems that function during the low temperature operation.

For Palisades, the pressurizer Power Operated Relief Valves (PORVs) are used for overpressure mitigation against an inadvertent mass addition or energy addition event during low temperature operating conditions.

The LTOP setpoint curve in the Palisades Technical Specifications is shown in Figure 3 [9]. The development of the LTOP setpoint curve includes an additional 10 percent allowed pressure in accordance with ASME Code Case N-514 [10]. As a result, the LTOP setpoint curve is higher than the corresponding pressure-temperature limit curves shown in Figures 1 and 2.2.2.1 L TOP Setpoint Methodology The current LTOP system contains a microprocessor unit that continuously monitors the temperature and pressure of the primary coolant system (PCS) and calculates a pressure setpoint based on the cold leg temperatures

[6]. The LTOP setpoint system was developed in Reference 29 with modifications as described in Reference 6 for the variable setpoint LTOP system. In particular, the modifications incorporated ASME Code Case N-514 for the setpoint pressures.

Other changes included a curve-fit for the primary coolant water bulk modulus of elasticity as a function of temperature, lower pressurizer heatup rates, and lower high pressure safety injection (HPSI) rates Report No. 1001026.401, Rev. 1 6 V Structural Integrity Associates, Inc.

based on the maximum two HPSI train delivery curves developed for the Emergency Operating Procedures.

As a result, the Technical Specification LTOP setpoint curve represents the pressure at which the PORV must be actuated to provide low temperature overpressure protection of the reactor vessel considering the pressure overshoot during the time it takes the PORV to open. The LTOP setpoint curve is determined to be applicable for the same conditions as the current P-T limit curves (i.e., an assumed peak fluence limit in the vessel of 2.192x10'9 n/cm 2) [6]. However, because the LTOP curves are not directly tied to a fluence value, but rather, they are connected to the corresponding P-T limit curves, the LTOP setpoints period of validity will also be reexamined here.3.0 RTNDT CALCULATION METHOD The prediction of embrittlement in the Palisades vessel beltline materials using Regulatory Guide 1.99, Rev. 2 depends on accurate determinations of the vessel material properties and calculated fluence projections in order to monitor the RTNDT. The Palisades reactor vessel beltline materials had been evaluated in previous years based on the initial RTNDT values, best estimate chemistry values for welds and plates, and the corresponding RTNDT shifts due to neutron irradiation of the vessel available at that time. Prior to Generic Letter 92-01 in the early 1990s, each plant was using its data to monitor and predict embrittlement of the vessels. A major revision of the weld best estimate chemistries for the CE-fabricated vessels occurred in the mid-I 990s through the Combustion Engineering Owners Group [11, 12]. The CEOG program compiled and evaluated all the data for the CE vessel weld materials in order to improve the best estimate chemistries and embrittlement predictions.

However, the last update to the Palisades P-T curves and LTOP setpoints was performed in 1994 which preceded much of the industry activities to develop improved best estimate chemistry values for the vessel beltline weld materials.

Consequently, the improvements to the knowledge of best estimate chemistry resulting from this work were not included in the Palisades curves.3.1 RTNDT Calculations for Current P-T Curves Table 6 and Table 7 show the RTNDT calculations at the 1/4T and 3/4T locations for the axial welds, circumferential weld, and base metal plates that were used in developing the current P-T curves and LTOP setpoints for Palisades.

The maximum RTNDT at the 1/4T location is 255°F for the limiting circumferential weld, and 191 'F at the 3/4T location for the limiting circumferential weld. This Report No. 1001026.401, Rev. 1 7 V Structural IntegrityAssociates, Inc.

equates to a peak surface fluence value of 2.192x 10 1 9 n/cm 2 (E > 1 MeV). The limiting material for the P-T curves is the circumferential weld made from heat No. 27204. The early submittals for the best estimate chemistry for this weld heat were Cu = 0.208 wt%, Ni = 1.00 wt%, with a CF value of 228°F, as obtained from Table 1 in Regulatory Guide 1.99, Rev. 2 [4]. Even though the weld is circumferential, an axially-oriented reference flaw was assumed with the properties at the 1/4T and 3/4T depths as shown in these tables. Later ASME Code versions (after 1992) allowed the use of circumferentially-oriented reference flaw in a circumferential weld, but this assumption was not used for the Palisades P-T curves. Consequently, the circumferential weld remains the limiting beltline material for the current P-T curves.3.2 Improved RTNDT Calculations for use with P-T Curves The Palisades reactor vessel consists of the following beltline region materials

[13, 14]: " Intermediate Shell, Axial Welds 2-112 A/B/C, material heat No. W5214," Lower Shell, Axial Welds 3-112 A/B/C, material heat No. W5214 and 34B009,* Intermediate-to-Lower Shell, Circumferential Weld 9-112, material heat No. 27204,* Intermediate Shell, Plate D-3803-1, material heat No. C-1279,* Intermediate Shell, Plate D-3803-2, material heat No. A-0313,* Intermediate Shell, Plate D-3803-3, material heat No. C-1279,* Lower Shell, Plate D-3804-1, material heat No. C-1308A,* Lower Shell, Plate D-3804-2, material heat No. C-1308B,* Lower Shell, Plate D-3804-3, material heat No. B-5294.The data used to evaluate these materials is described below.3.2.1 Weld Wire Heat No. 27204 Weld heat No. 27204 was used in the intermediate-to-lower shell circumferential weld 9-112.Reference 13 describes the basis for the best estimate chemistry for weld heat No. 27204. CE NPSD-1039, Rev. 02 [11] specifies best estimate values of 0.203% Cu and 1.018% Ni for welds fabricated with weld wire heat number 27204. The best estimate values reported June 5, 1992 for the Palisades reactor vessel beltline circumferential weld were 0.208% Cu and 1.00% Ni. In the July 3, 1992 response to Generic Letter 92-01 Rev. 1, the best estimate value for the circumferential weld was reported as 0.21% Cu and 1.00% Ni. In the December 20 and November 17, 1995 submittals, it was acknowledged that additional chemistry information was available and calculated weighted Report No. 1001026.401, Rev. 1 8 V Structural Integrity Associates, Inc.

values of 0.198% Cu and 1.02% Ni. At the time, Consumers Energy chose to continue to report 0.21% Cu and 1.00% Ni as the best estimate value for the circumferential weld. Given that the values reported in CE NPSD-1039, Rev. 02 are comparable to those calculated by Consumers Energy in 1995, the utility concluded that the best estimate chemistry for the Palisades reactor vessel beltline weld fabricated with weld wire heat no. 27204 is 0.203% Cu and 1.018% Ni as reported in CE NPSD-1039, Rev. 02 [11, 13, 24]. The revised CF value for this weld is 226.87F from the tables in Reg. Guide 1.99, Rev. 2. This is consistent with the chemistry and CF value for this weld heat number contained in RVID2 [15], as shown in Table 8.3.2.1.1 Surveillance Data for Weld Heat No. 27204 The previous analysis of surveillance data for weld heat no. 27204 [24] included only two capsule data points from Diablo Canyon Unit 1, Capsule S [25] and Capsule Y [26]. That analysis performed a least-squares fit to the surveillance data results and determined the data to not be credible because the scatter exceeded the allowable scatter for credible surveillance data (i.e., Icy >28°F). Since then, two new data surveillance data points were obtained from the Palisades supplemental capsules SA-60-1 [27] and SA-240-1 [28] and one from Diablo Canyon Unit 1, Capsule V [30]. The surveillance data results for the capsules containing weld heat No. 27204 are shown in Appendix A.- It is noted that the Charpy V-notch test data were fitted with the TANH function to determine the AT 3 0 , or RTNDT shift values. The measured RTNDT shift values from the fitted surveillance capsules were considered for projection of embrittlement in the Palisades vessel weld. The two (2) surveillance capsules from Palisades were found to be credible.

The Palisades supplemental surveillance data were combined with the Diablo Canyon 1 surveillance data and the results are shown in Table 9. The analysis of the surveillance data was performed using Case 4 of the NRC guidance, "Surveillance Data from Plant and Other Sources" [18]. Adjustments were made to the measured shift values to account for chemistry differences and temperature differences between the capsules and the vessel, as shown in Table 9. The irradiation temperatures for the Diablo Canyon Unit 1 capsules were obtained from Reference

32. Using the least-squares fitting method in Eq. (8), a fitted CF value of 216.13'F was obtained for weld heat No: 27204 for application to the Palisades circumferential weld, 9-112. A discussion of data crediblity for weld heat No. 27204 is given in Appendix A. A plot of all the capsule data for weld heat No. 27204 is shown in Figure 4 along with the 1-sigma (28'F) scatter bound for weld materials.

All surveillance Report No. 1001026.401, Rev. 1 9 Structural Integrity Associates, Inc.

data points fall within 1-sigma (17°F), therefore, the data appear to be credible.

However, the chemistry factor of 226.8°F calculated from the best estimate copper and nickel content of the weld, and using the tables in Reg. Guide 1.99, Rev. 2, is determined to be more conservative for the circumferential weld in the Palisades vessel, and that CF value was retained for the comparison in this analysis.

The surveillance capsule reports with the reported data for weld heat No. 27204 are given in Appendix B.3.2.2 Weld Heat No. 34B009 Weld heat No. 34B009 was used in the Palisades lower shell axial welds 3-112 A/B/C. CE NPSD-1039, Rev. 02 specifies best estimate values of 0.192% Cu and 1.038% Ni for weld fabricated with weld wire heat No. 34B009 [11]. In the submittal dated September 8, 1998, it was determined that a copper value of 0.192% is the best estimate value for the welds made from heat no. 34B009 in the Palisades vessel [13]. However, it was also determined at that time that the best estimate chemistry value for nickel recommended in CE NPSD-1039, Rev. 02 for nickel addition welds could not be endorsed for the Palisades vessel welds. The value of 1.038% Ni was determined by finding the mean of 144 nickel measurements.

It was noted that 45 of those measurements were from the retired Palisades steam generators, and the calculated mean was dominated heavily by just two welds. A new evaluation of the nickel content using a coil-weighted average was performed by the CEOG in"Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content," CE NPSD-1 119, Revision 01, July 1998 [12]. The best estimate value of 1.007% nickel derived using a sample weighted mean was considered a technically superior approach to that used in CE NPSD-1039, Rev. 02. Further studies were performed after the CEOG work was completed in order to establish the best estimate nickel value to be used for this weld in the Palisades vessel. These were described in Reference 16, and after performing these studies it was concluded that the best estimate chemistry for the Palisades reactor vessel beltline welds fabricated with weld wire heat number 34B009 (with nickel addition) is 0.192% Cu and 0.98% Ni.Using Table 1 in Reg. Guide 1.99, Rev. 2, a chemistry factor (CF) of 217.7°F was determined based on the best estimate chemistry for this weld. These are the same values for Cu and Ni content for this weld contained in RVID2 [15], as shown in Table 8.Report No. 1001026.401, Rev. 1 10 I Structural Integrity Associates, Inc.

3.2.3 Weld Heat No. W5214 Weld heat no. W5214 was used in the intermediate shell axial welds 2/112A/B/C.

The weld at the 600 azimuthal location was determined to be the limiting material for PTS, but it is not necessarily the beltline material with the highest RTNDT. The best estimate Cu = 0.213 wt%, and the best estimate Ni = 1.007 wt% for weld heat No. W5214 [12]. The calculated CF value from the Reg.Guide 1.99, Rev. 2 tables is CF = 230.73°F.3.2.3.1 Surveillance Data for Weld Heat No. W5214 To determine whether this weld may also become limiting for P-T curves, a more detailed analysis of the available surveillance capsule data was performed.

A more complete evaluation of the surveillance data for weld heat no. W5214 is given in Reference

17. The results of that study are summarized here. Surveillance capsule data from eleven capsule reports were found to contain this weld heat. These data were obtained from the Palisades supplemental capsules, and additional data from H. B. Robinson 2, Indian Point Unit 2, and Indian Point Unit 3 was evaluated using the credibility requirements in Reg. Guide 1.99, Rev. 2. The NRC guidance for evaluation and use of other plant's surveillance data is contained in Reference

18. The data were adjusted to account for difference in specimen chemistry and plant operating temperatures, and the irradiated Charpy shift results were fitted using the least squares fit relation given in Equation (8) [18].CF = M (8)Oft w 91 1=1 where "n" = the number of surveillance data points,"A," = the measured value of AT 3 0 from the Charpy specimens, and"F," = the fluence for each surveillance capsule data point.The results of that evaluation are given in Table 10 including the least squares fit to the surveillance data results. A plot of the W5214 Charpy shift surveillance data results is given in Figure 5. It is noted that the fitted CF value = 227.74°F for weld heat no. W5214. Although this weld material does not appear to be the limiting weld for the Palisades P-T curves, a more accurate projection of Report No. 1001026.401, Rev. 1 11 Structural Integrity Associates, Inc.

the RTNDT shift was performed using the least-squares fit to the results from the relevant surveillance data.3.2.4 Vessel Beltline Plate Materials The reported copper and nickel chemistry values and calculated chemistry factors for the each of the vessel beltline plates are shown in Table 8 (from RVID2) [15]. It should be noted that the assumptions for base metal properties as shown in Tables 1 and 2 used the highest Cu content (Cu = 0.24%), the highest nickel content (Ni = 0.55%) and the highest initial RTNDT = 0 0 F to represent the beltline plates for calculating the base metal RTNDT in Tables 6 and 7. Despite this conservative assumption, the base metal is not limiting for the P-T curves or LTOP setpoints because the beltline welds have a much higher RTNDT.3.2.4.1 Surveillance Data for Plate Heat No. C-1279 Surveillance data was used to determine a fitted chemistry factor, CF, for plate C-1279. The Palisades surveillance capsule program contains matching base material for plates D-3803-1 and D-3803-3.

Four surveillance capsules with plate heat no. C-1279 were previously removed and tested and the data were used to calculate a fitted chemistry factor to the measured Charpy V-notch test results. The Palisades surveillance capsules which contain this base metal include A-240 [20], W-290 [21], and W-110 [22] and W-100 [31]. There is no new capsule data for the Palisades vessel beltline plate materials.

However, there is new fluence data for the Palisades surveillance capsules containing the surveillance plate heat No. C-1279. The Charpy test results and the evaluated AT 3 0 shift values from these capsule test reports are given in Appendix C. Updated surveillance capsule fluence values were obtained from WCAP-15353, Supplement 1 [19], and the least-squares fit to the data using the revised fluence values is given in Table 11. The least-squares fit to these data is determined to be CF = 147.71 'F, and a plot of the data and fitted results is given in Figure 6.However, it is observed that three data points fall outside the 1-sigma bound (17'F) and one data point falls outside the 2-sigma bound (34°F) for plate materials and, therefore, the data were determined to be non-credible and a fitted CF value should not be used for projections of the vessel RTNDT (ART) for the beltline plates with the same heat number. The projection of RTNDT (ART) for the vessel beltline plates made from plate heat No. C-1279 used a CF value of 157.5°F (from the Reg. Guide 1.99, Rev. 2 tables) and a full margin term of 34°F.Report No. 100 1026.401, Rev. 1 12 V Structural Integrity Associates, Inc.

Table 6. Limiting Surface Fluence and 1/4T RTNDT Values for Current Palisades P-T Curves and LTOP Setpoints Surface I/4T Fluence Fluence 1/4T RPV Material Cu% Ni% CF (xEl9) (xE19) FF RTNDT(U) ARTNDT Margin RTNDT (OF) (n/cm 2) (n/cm 2) (OF) (OF) (OF) (OF)Axial Welds 0.20 1.02 225 1.620 0.973 0.992 -56 223.3 66 233.3 Circ Weld 0.208 1.00 228 2.192' 1.316 1.076 -56 245.3 66 255.32 Base Metal 0.24 0.55 165 2.192 1.316 1.076 0 177.6 34 211.6 1 Peak Surface Fluence 2 Maximum 1/4T Adjusted Reference Temperature Table 7. Limiting Surface Fluence and 3/4T RTNDT Values for Current Palisades P-T Curves and LTOP Setpoints Surface 3/4T Fluence Fluence 3/4T RPV Material Cu% Ni% CF (xEl9) (xE19) FF RTNDT(U) ARTNDT Margin RTNDT (OF) (n/cm2) (n/cm2) (OF) (OF) (OF) (OF)Axial Weld 0.20 1.02 225 1.620 0.351 0.711 -56 160.0 66 170.0 Circ Weld 0.208 1.00 228 2.192 0.475 0.792 -56 180.7 66 190.73 Base Metal 0.24 0.55 165 2.192 0.475 0.792 0 130.7 34 164.7 3 Maximum 3/4T Adjusted Reference Temperature Report No. 100 1026.40 1, Rev. I 13 R Struclural Integrity Associates, Inc.

Table 8. RVID2 Vessel Materials Summary Report for Palisades

[151 NRC -Reactor Vessel Integrity Database PTS Summary Report Printed 9,9S20 10 10 48 55 AS DocketNo 80-255 page 1 PALISADES EOL Date 0 3114/2007_________________________

iO~ts 7ee ~t-rv1,.ý Feaci Doen Ceirrsy Facta Tyoe Heat ID @ EOL @ EOL r, PT-AluivETHOD CEOL @ EOL -acla Metmoo Mwgw kiaeg M~ethod --u % Ni% Pe% S%D-3803-2 1949 2010 -300 lUES !-2 1909 1 190 150 40 TABLE 340 POSmION 11 0240 0O 0010 00 FLATE A-0313 l NO S DATA)C-3804-2 10t" 2010 -a 0 MTEe 5-2 9S7 1 190 8200 TABLE 340 POSITION 1 1 0120 0550 0010 0010 PLATE 6 !.294 INO S DATA)D-3803-1 1944 2.010 -!0 PLANT SPEC IFIC 182 4 1190 15328 ;JRVEILLAN4CE 170 POSInON 2 1 0240 0 510 0009 020 PLATE iZ- 1279 NON-RATI8 S DATA, 0-3803-3 1944 2010 -80 PLANT SPEC FiC '&24 1190 15328 ,RVEILLANCE 1-0 POSmON 2- 0240 8 00 001, 0020 PLATE C- 12779 NON- RATIO ý S DATA)0-3804-1 187 2.010 00 MTES 5.2 2 3 1 190 125 80 TABLE 340 PsmoTN 1 1 01900 480 0016 0 Z-PLATE C-1308A (NO S DATA D-3804-2 18t9 20!1 _30 -NITEE 8-2 8M, 9 1 190 131 00 TABLE 340 POSITION 1 1 0 19 0800 001l8 0020 PLATE I -1 Fixes (NO0 5 DATA)CIRC WELD 9S112 28! t 1080 -1 0 GENERIC 2720 1 199 22681 TABLE 5 t POSMON 11 0203 1018 0013 012 WELD 7 2720_4 _ _N_ S DATA, LOW EP SHELL AXIAL WELD 3- 124.C 2838 1 88-0 -J0 GENERIC 24' 0 i 12' 21770 TABLE M88 POSITION I 1 019 0 98030000 0000 WELD F I009 NO S DATA)INTERMEDIATE SHELLAýQALWELDS2-112 A/C 258 1.I0o -00 GENERI 201%0 1 121 22' 08 TABLE e! ! POSITICN 11 0213 1 0'0 0019 O018 AELL I W5214 ( NO S DATA)LOWER SHELL AXIAL WELDS 3-112A4C 2e-88 1 .60 -T0 2.180 1 12 i 308 TABLE et' S OSITION 1 1 0211 11010 10019 01 WELD = 4 -214 NO S DATAI Plant References and Belttine Material Notes ChemicaI composition.

fluence. and RTndtlu) for the welds are from the September S. 1998 letter from N L. Hlaskell to the USNRC Document Control Desk. subject Palisades Plant Response to Request for Additional Information Regardmg Reactor Pressure Vessel Integrity

-'The September 8 1998 letter states that Consumers Energy continues todiscuss the rate of fluence accumulation to the Palisades reactor vessel vAh the NRC since the sta ff has not accepted the bias measured in the fluence calculations"-

5ased on the safety evaluation report ISERidated Apnil 12, 1995 from Elinor 4densam (NRC)to Kurt Haas the plant is in compliance With the requirements of10 CFR 5061 until the lastquarterof the 1999 C hem ical composition and RTndtlu ý for the plates a re from the May 2E 1994 letter from D W Rogers to the USNRC Document C ontml Desk, subject Palisades Plant -inconsisten ces identified During Venrication of Information Provdede in Response to Generic Letter 92-01. Revision I ULnirradiated upperashelf energy (UUSE forplatesan Wueld 3-112 4C heat numberO4B009)are reported in the May25, 1994 letter from D. W Rogersto theeUSNRC Document Control Desk.subject: Palisades Plant -Inconsstencies Identihfed During Verification of Information P rovided in Response to Generic Letter 92-01, Revison 1 U USE for all other Welds are from the July 3, 1992 letter from S B. Slade to the USfIRC Document Control Desk. subject Palisades Plant-Reactor Vessel Structural Integrity-Response to Generic Letter 92-01. Revision I Chemistry Factor ICF) for the surveillance material (heat number C-1279 YABs calculated from data reported in surveillance reports WCAP-10637 and BCL-585-12 Weld 9-112 Iheat number 272041 has an RTpts value of 281.8 degrees F, hcwe-'er.

this value is not overthe PTS screening criterion for circumferential velds and plates which is 300 degrees F (the screening criteron foraxial Y selos E 270 degrees F;)Report No. 100 1026.40 1, Rev. I 14 R6Structural Integrity Associates, Inc.

Table 9. Evaluation of Surveillance Capsule Results Containing Weld Heat No. 27204 Ratio Chem. &Table Revised Fluence Irrad. Measured Adjusted Temp. Adj. Predicted Adjusted -Capsule %Cu %Ni CF (F) Fluence Factor Temp. ARTndt ARTndt ARTndt ARTndt Predicted (n/cmA2) FF Ti (F) (F) (F) (F) (F) (F)CAP Y (DCPP) 0.198 0.999 222.26 1.05E+19 1.01 542 232.59 237.3 244.1 219.1 25.06 CAP S (DCPP) 0.198 0.999 222.26 2.84E+18 0.66 544 110.79 113.1 121.9 141.8 -19.97 SA-240-1 (PNP) 0.194 1.067 227.8 2.38E+19 1.23 535.7 267.8 266.7 267.2 266.7 0.49 SA-60-1 (PNP) 0.194 1.067 227.8 1.50E+19 1.11 535 253.1 252.0 251.8 240.4 11.43 CAP V (DCPP) 0.198 0.999 222.26 1.37E+19 1.09 541.5 201.07 205.2 211.5 235.0 -23.56 Vessel Best Estimate CF = 226.8 Mean T = 535.2 {___Least Squares Fitted CF = 216.13 Report No. 100 1026.40 1, Rev. I 15 1 Structural Integrity Associates, Inc.

Table 10. Evaluation of all Surveillance Capsule Results Containing Weld Heat No. W5214 [17]Measured Ratio Chem. &Table Revised Fluence Irrad. (Refitted)

Adjusted Temp. Adj. Predicted Adjusted-Capsule %Cu %Ni CF (F) Fluence Factor Temp. ARTndt ARTndt ARTndt ARTndt Predicted (n/cmA2) FF Ti (F) (F) (F) (F) (F) (F)SA-60-1 0.307 1.045 266.5 1.50E+19 1.11 535 259 224.2 224.0 253.3 -29.27 SA-240-1 0.307 1.045 266.5 2.38E+19 1.23 535.7 280.1 242.5 243.0 281.0 -38.00 HB2T 0.34 0.66 217.7 3.87E+19 1.35 547 289.1 306.4 318.2 307.2 10.97 HB2 V 0.34 0.66 217.7 5.30E+18 0.82 547 208.8 221.3 233.1 187.3 45.75 HB2 X 0.34 0.66 217.7 4.49E+19 1.38 547 265.6 281.5 293.3 314.4 -21.14 IP2V 0.20 1.03 226.3 4.92E+18 0.80 524 197.5 201.4 190.2 182.7 7.48 IP2Y 0.20 1.03 226.3 4.55E+18 0.78 529.1 193.9 197.7 191.6 177.8 13.77 IP3T 0.16 1.12 206.2 2.63E+18 0.64 539.4 149.8 167.6 171.8 145.0 26.81 IP3Y 0.16 1.12 206.2 6.92E+18 0.90 539.5 171.1 191.5 195.8 204.2 -8.48 IP3Z 0.16 1.12 206.2 1.04E+19 1.01 538.9 228.3 255.5 259.2 230.2 28.92 IP3X 0.16 1.12 206.2 8.74E+18 0.96 539.7 192.5 215.4 219.9 219.1 0.76 Vessel Best Estimate CF = 230.73 1 Mean T= 535.2 __I I II I I I I I [Least Squares Fitted CF= 227.74 Report No. 100 1026.40 1, Rev. 1 16 V Structural Integrity Associates, Inc.

Table 11. Evaluation of Surveillance Capsule Results Containing Plate Heat No. C-1279 Measured -Material I.D. Capsule Capsule FF Measured Predicted Predicted fluence* ARTndt ARTndt ARTndt (n/cm^2) (OF) (OF) (OF)D3803-1 (Longitudinal)

A-240 [20] 4.09E+19 1.361 205.0 201.0 4.0 D3803-1 (Transverse)

A-240 [20] 4.09E+19 1.361 205.0 201.0 4.0 D3803-1 (Longitudinal)

W-290 [21] 9.38E+18 0.982 155.0 145.1 9.9 D3803-1 (Transverse)

W-290 [21] 9.38E+18 0.982 175.0 145.1 29.9 D3803-1 (Longitudinal)

W-110 [22] 1.64E+19 1.136 180.0 167.9 12.1 D3803-1 (Transverse)

W-100 [31] 2.09E+19 1.201 142.5 177.3 -34.8 D3803-1 (Longitudinal)

W-100 [31] 2.09E+19 1.201 159.1 177.3 -18.2 Fitted CF= 147.71*Revised fluence values from WCAP-15353, Supplement 1-NP [19]Report No. 100 1026.40 1, Rev. I 17 R Structural Integrity Associates, Inc.

4.0 FLUENCE PROJECTIONS AND FLUENCE METHODOLOGY Westinghouse recently performed an updated fluence assessment for the Palisades vessel beltline region [19]. This revised fluence evaluation provided an updated fluence assessment for the vessel beltline region that included cycle specific analyses for known core configuration through operating Cycles 15 through 21, and projections for future operation based on the best available knowledge as a function of EFPY. The calculated and projected neutron fluence values for the limiting 600 weld location and the peak (750) fluence location are given in Table 12 [19]. As shown on this table, the date to reach the limiting peak surface fluence of 2.192x 10'9 n/cm2 (E > I MeV) is May 2011 for the existing P-T curves.Table 12. Calculated Clad-to-Base Metal Interface Fluence in Palisades Vessel [191 Circ. Weld Circ. Weld End of Estimated Cumulative Fluence (n/cmA2, E > 1 MeV) 1/4T 3/4T Fuel Calendar Time Cycle Date (EFPY) Axial Weld Peak @75° RTNoT (-F) RTNDT ({F)@60* tOld/New]

[Old/New]21 10/2010 23.4 1.472E+19 2.157E+19 254.4/252.6 189.7/188.2 11/2010 23.4 1.472E+19 2.157E+19 254.4/252.6 189.7/188.2 12/2010 23.5 1.475E+19 2.163E+19 254.6/252.9 189.8/188.4 1/2011 23.6 1.478E+19 2.168E+19 254.6/252.9 190.0/188.5 2/2011 23.6 1.480E+19 2.174E+19 254.9/253.1 190.1/188.7 3/2011 23.7 1.483E+19 2.179E+19 255.1/253.3 190.3/188.8 4/2011 23.8 1.486E+19 2.185E+19 255.3/253.5 190.5/189.0 5/2011 23.9 1.489E+19 2.192E+19 255.3/253.5 190.7/189.2 6/2011 23.9 1.492E+19 2.196E+19 255.6/253.8 190.8/189.3 7/2011 24.0 1.495E+19 2.202E+19 255.8/254.0 190.9/189.5 8/2011 24.1 1.497E+19 2.207E+19 255.8/254.0 191.1/189.6 9/2011 24.2 1.500E+19 2.213E+19 256.0/254.2 191.2/189.8 10/2011 24.2 1.503E+19 2.218E+19 256.2/254.4 191.4/189.9 11/2011 24.3 1.506E+19 2.224E+19 256.2/254.4 191.6/190.1 12/2011 24.4 1.509E+19 2.230E+19 256.5/254.7 191.7/190.3 1/2012 24.5 1.512E+19 2.235E+19 256.7/254.9 191.9/190.4 2/2012 24.5 1.514E+19 2.240E+19 256.7/254.9 192.0/190.5 312012 24.6 1.517E+19 2.246E+19 256.9/255.1 192.2/190.7 22 4/2012 24.7 1.520E+19 2.252E+19 257.1/255.3 192.3/190.9 Note: bolded values have been interpolated Report No. 1001026.401, Rev. I 18 V Structural Integrity Associates, Inc.

5.0 RTNDT PROJECTIONS WITH REVISED VESSEL MATERIAL PROPERTIES The results from the updated vessel beltline material analyses are shown in Tables'l 3 and 14. The results with and without the use of fitted surveillance data are shown in these tables. The updated CF values obtained from the Reg. Guide 1.99, Rev. 2 are included along with the RTNDT shift results and the corresponding adjusted reference temperature values. The revised fluence projections from WCAP-15353, Supplement 1 [ 19] were incremented to achieve the same maximum 1/4T RTNDT value of 255.3°F and a 3/4T RTNDT value of 190.7°F which is the basis for the current operating P-T curves. Table 12 shows the incremental I.D. surface fluence at the 600 (limiting axial weld) and 750 (peak vessel fluence) locations.

Also shown are the projected RTNDT values in the limiting circumferential weld at the 1/4T and 3/4T depths using both the "old" (CF 228°F) and "new" (CF = 226.8°F) methods. The corresponding fluence to reach the matching 1/4T or 3/4T RTNDT limits can be determined in Table 12. As noted in Tables 12 and 14, this correspondence occurs in the circumferential weld 9-112 at a peak I.D. surface fluence of 2.246xl01 9 n/cm 2 (E > 1 MeV) when the maximum vessel 3/4T RTNDT value reaches the limit of 190.7°F. From Table 12 it can also be seen that the peak fluence of 2.246x10 1 9 n/cm 2 (E > I MeV) is projected to be reached in March 2012 using an interpolation of the fluence projections for Cycle 22 from Reference

19. By comparison, the maximum 1/4T RTNDT limit of 255.3°F in the circumferential weld would be reached when the peak I.D. surface fluence reaches 2.252x 1019 n/cm 2 in April 2012, as confirmed by the limiting vessel 1/4T RTNDT values in Table 13. The March 2012 date for the weld heat No. 27204 to reach the 3/4T RTNDT limit is determined to be the more limiting case.6.0 DISCUSSION The properties for the Palisades reactor vessel beltline materials have been examined to determine the limiting material condition and validity date for the current P-T limit curves. All available surveillance data and updated material chemistries were considered, and it was determined that the limiting beltline material for the P-T curves remains the circumferential weld made from weld heat No. 27204. From the basis documents for the P-T curves, it was determined that the limiting 1/4T RTNDT values of 255.3°F and 3/4T RTNDT value of 190.71F were used to calculate the current heatup, cooldown and LTOP setpoint limits. At the time the curves were calculated, the peak Report No. 1001026.40 1, Rev. 1 19 I Structural Integrity Associates, Inc.

fluence for the limiting circumferential weld (9-112) to be correlated with the current P-T curves was 2.192x10 1 9 n/cm 2 (E > 1 MeV). The assumptions used at that time may no longer be valid, so a reanalysis of the basis for the P-T curves and LTOP setpoints was performed, including up-to-date materials data and fluence projections.

The basis is equated to the maximum 1/4T and 3/4T RTNDT values for the limiting circumferential weld used in developing the P-T curves and LTOP setpoints.

Using the "old" method with a CF value of 2287F for weld heat No. 27204, the current P-T curves and LTOP setpoints are projected to expire in May 2011. However, by using the"new" method with the latest best estimate chemistry results and a CF value = 226.8'F for limiting weld 9-112, and considering the most recent vessel fluence calculations and the actual margin term of 65.5°F, the Palisades P-T limit curves and LTOP setpoints are projected to be valid until March 2012 or a peak I.D. surface fluence of 2.246x10 1 9 n/cm 2 (E > 1 MeV).Use of a "new" value for CF = 226.8°F versus the "old" value of CF = 228'F for the limiting circumferential weld is acceptable because it is based the latest industry best estimate copper and nickel values for weld heat No. 27204 as submitted by Palisades

[24] and reviewed and approved by NRC as shown in RVID2 (see Table 8) [15]. The original methodology.and other design input parameters have been maintained for this analysis to determine the new fluence value (and corresponding applicability date) when the circumferential weld would experience a 1/4T RTNDT value of 255.3'F and a 3/4T RTNDT value of 190.71F. Based on these analyses, the revised validity date of March 2012 applies to the pressure-temperature limits for heatup, cooldown, and the corresponding LTOP setpoint limits for the Palisades operating P-T limit curves.Report No. 1001026.40 1, Rev. 1 20 V Structural Integrity Associates, Inc.

Table 13. Revised Limiting Surface Fluence and 1/4T RTNDT Values for Current Palisades P-T Curves and LTOP Setpoints Surface 1/4T Fluence 1 Fluence 1/4T RPV Material Heat No. Cu% Ni% CF (xE19) (xE19) FF RTNDT(U) ARTNDT Margin RTNDT (OF) (n/cm 2) (n/cm 2) (OF) (OF) (OF) (OF)Axial Weld 3-112A/C 34B009 0.192 0.98 217.7 1.520 0.913 0.974 -56 212.1 65.5 221.6 226.8 2.252 1.352 1.084 -56 245.8 65.5 255.32 Circ Weld 9-112 27204 0.203 1.018 216.133 2.252 1.352 1.084 -56 234.3 44 222.3 230.73 1.520 0.913 0.974 -56 224.8 65.5 234.3 Axial Weld 2-112A/C W5214 0.213 1.007 227.743 1.520 0.913 0.974 -56 221.9 65.5 231.4 157.5 2.252 1.352 1.084 -5 170.7 34 199.7 Plate D-3803-1 C-1279 0.24 0.50 147.7' 2.252 1.352 1.084 -5 160.1 34 189.1 Plate D-3803-2 A-0313 0.24 0.52 160.4 2.252 1.352 1.084 -30 173.9 34 177.9 157.5 2.252 1.352 1.084 -5 170.7 34 199.7 Plate D-3803-3 C-1279 0.24 0.50 147.73 2.252 1.352 1.084 -5 160.1 34 189.1 Plate D-3804-1 C-1308A 0.19 0.48 128.8 2.252 1.352 1.084 0 139.6 34 173.6 Plate D-3804-2 C-1308B 0.19 0.50 131.0 2.252 1.352 1.084 -30 142.0 34 146.0 Plate D-3804-3 B-5294 0.12 0.55 82.0 2.252 1.352 1.084 -25 88.9 34 97.9 1 Revised Peak Surface Fluence for April 2012 2 Maximum 1/4T Adjusted Reference Temperature for April 2012 3 Surveillance data used to calculate CF Report No. 100 1026.40 1, Rev. I 21 R2 Struclural Integrity Associates, Inc.

Table 14. Revised Limiting Surface Fluence and 3/4T RTNDT Values for Current Palisades P-T Curves and LTOP Setpoints Surface 3/4T Fluence' Fluence 3/4T RPV Material Heat No. Cu% Ni% CF (xE19) (xEl9) FF RTNDT(U) ARTNDT Margin RTNDT (OF) (n/cm 2) (n/cm 2) (OF) (OF) (OF) (OF)Axial Weld 3-112A/C 34B009 0.192 0.98 217.7 1.517 0.328 0.694 -56 151.0 65.5 160.5 226.8 2.246 0.486 0.799 -56 181.2 65.5 190.72 Cire Weld 9-112 27204 0.203 1.018 216.13' 2.246 0.486 0.799 -56 172.7 44 160.7 230.73 1.517 0.328 0.694 -56 160.1 65.5 169.6 Axial Weld 2-112A!C W5214 0.213 1.007 227.74' 1.517 0.328 0.694 -56 158.0 65.5 167.5 157.5 2.246 0.486 0.799 -5 125.8 34 154.8 Plate D-3803-1 C-1279 0.24 0.50 2.246 0.486 0.799 -5 118.0 34 147.0 Plate D-3803-2 A-0313 0.24 0.52 160.4 2.246 0.486 0.799 -30 128.2 34 132.2 157.5 2.246 0.486 0.799 -5 125.8 34 154.8 Plate D-3803-3 C-1279 0.24 0.50 147.73 2.246 0.486 0.799 -5 118.0 34 147.0 Plate D-3804-1 C- 1308A 0.19 0.48 128.8 2.246 0.486 0.799 0 102.9 34 136.9 Plate D-3804-2 C-1308B 0.19 0.50 131.0 2.246 0.486 0.799 -30 104.7 34 108.7 Plate D-3804-3 B-5294 0.12 0.55 82.0 2.246 0.486 0.799 -25 65.5 34 74.5 Revised Peak Surface Fluence for March 2012 2 Maximum 3/4T Adjusted Reference Temperature for March 2012 3 Surveillance data used to calculate CF Report No. 100 1026.40 1, Rev. I 22 V Structural Integrity Associates, Inc.

7.0 REFERENCES

1. 1 OCFR50, Appendix G, "Fracture Toughness Requirements," May 31, 1983 as amended November 6, 1986.2. ASME Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Non-Ductile Failure," 1992 Edition.3. Palisades Plant Technical Specifications, Section 3.0 Limiting Conditions for Operation, 3.1 Primary Coolant System, Amendment No. 163, effective March 2, 1995. (SI File No. 1001026.201)

4. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, May, 1988.5. U.S. Nuclear Regulatory Commission Standard Review Plan, NUREG-0800, Section 5.3.2, "Pressure-Temperature Limits," Rev. 1, July 1981. (SI File No. 1001026.202)

6. Consumers Power Company Engineering Analysis EA-A-PAL-92-095-01, Rev. 0,"Pressure Temperature Curves and LTOP Limit Curve for Maximum Reactor Vessel Fluence of 2.192x 10 1 9 Neutron/cm 2 ,"' September 1994. (SI File No. 1001026.203)

7. Palisades Plant Technical Specifications, LCO 3.4.3, PCS Pressure and Temperature (P/T) Limits, Amendment No. 189. (SI File No. 100 1026.204)8. U.S. Nuclear Regulatory Commission Standard Review Plan, NUREG-0800, Section 5.2.2, "Overpressure Protection," Rev. 2, November 1988, (SI File No. 1001026.205)

9. Palisades Plant Technical Specifications, LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) System, Revised 10/29/2009. (SI File No. 1001026.206)

10. ASME Code Case N-514, Low Temperature Overpressure Protection,Section XI, Division 1, American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Code Cases -Nuclear Components, 1992 Edition, Supplement No. 4. (SI File No. 1001026.207)

11. "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Rev. 02, Combustion Engineering Owners Group, June 1997.(SI File No. 1000915.202)

Report No. 1001026.401, Rev. 1 23 Structural Integrity Associates, Inc.

12. "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)

13. Letter from Darl S. Hood (USNRC) to Nathan Haskell (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).

14. "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)

15. U.S. Nuclear Regulatory Commission, "Reactor Vessel Integrity Database Version 2.0.1," September 7, 2000.16. Smedley (Consumers Energy) to NRC, "Docket 50-255-License DPR-20 -Palisades Plant Response to NRC Generic Letter 92-01, Revision 1, Supplement 1: Reactor Vessel Structural Integrity," November 17, 1995. (SI File No. 1000915.212)

17. "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," SI Report No. 0901132.401, Rev. 0, April 2010.18. "Generic Letter 92-01 and RPV Integrity Assessment," NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. (SI File No. 0901132.213).

19. "Palisades Reactor Pressure Vessel Neutron Fluence Evaluation," WCAP-15353-Supplement 1-NP, May, 2010 (SI File No. 1000915.211).

20. Perrin, J. S., et al., "Palisades Nuclear Plant Reactor Pressure Vessel Surveillance Program Capsule A-240," BCL-585-12, March 13, 1979. (SI File No. 1000915.216)

21. Kunka, M. K., Cheney, C. A., "Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program," WCAP-10637, September 1984. (SI File No. 1000915.217)

22. Peter, P. A., et al., "Analysis of Capsule W-I 10 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program," WCAP-14014. (SI File No. 1000915.218)

Report No. 1001026.401, Rev. 1 24 Structural Integrity Associates, Inc.

23. 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).

24. 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)

25. Westinghouse Report, WCAP- 11567, "Analysis of Capsule S from the Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," December 1987. (SI File No. 1000915.209)

26. Westinghouse Report, WCAP-13750, "Analysis of Capsule Y from the Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," July 1993. (SI File No. 1000915.210)

27. 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).

28. 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).

29. Consumers Power Company Engineering Analysis EA-FC-809-13, Rev. 1, "Pressure Response Effects of VLTOP with Replacement PORVs," September 1989. (SI File No. 1001026.209)

30. Westinghouse Report WCAP-15958, Revision 0, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," January 2003. (SI File No. 1000915.219).

31. BWXT Services Report No. 1295-OD 1-03-08:00, "Analysis of Capsule W-100 from the Nuclear Management Company Palisades Reactor Vessel Material Surveillance Program," February 2004. (SI File No. 1000915.220).

32. Westinghouse Letter Report No. CPAL-10-34 Dated October 15, 2010,"Westinghouse Review of Structural Integrity Associates Reports on Pressurized Thermal Shock for Palisades." (SI File No. 1000915.22 1).Report No. 1001026.401, Rev. 1 25 Structural IntegrityAssociates, Inc.

33. ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition, including all addenda through Winter 1965, American Society of Mechanical Engineers.

34. ASTM E 185-66, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors." 35. 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, April 15, 2010. (SI File No. 0901132.224)

36. ORNL Report, "Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials," Oak Ridge National Laboratory, NUREG/CR-6413, ORNL/TM-13133, April 1996. (SI File No. 1000915.215)

37. "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)

Report No. 1001026.401, Rev. 1 26 Structural Integrity Associates, Inc.

U)0.j 1.0 0 0 1~0.1.0'.4 1.Si Si S£22509 1750: 20 F/Hi-1759-&-43 F114- 68 F/Hr 1258-758 -__ __ __ __ORAfl X2 I/Vtm Inlet Ava. Hrf-/ .JIM IRM TI 170F 20*F/Hr.250 1 T 170F 40*F/Hr.350 ) T )250F 60F/Hr.dh T 350*F t28T/Hr-.*When shutdown, colsn Isolation valves 14O-3015 and -03016 are.,en. PCs heataw rate 5116ll be maintained

ý 40F/Hie.4.rJI I 11 111 11Ri l1 11111 11111111111 1ii 50 100 150 200 250 300 350 480 459 RV Inlet Temperature F Figure 3.4.3-1 (Page 1 of 1)Pressure -Temperature Umits for Heatups'Figure 1. Pressure-Temperature Limits for Heatups [71 Report No. 1001026.401, Rev. 1 27 V Structural Integrity Associates, Inc.

C1.L L L CL 2250 2000' i--9 F/Hr--- 20 F/k-1750 -A- 40 F/Hr-e- 65 F/Hr 1500-- 80 F/-Hr--108 F/W.1250 i 7RI. ___ __ _ ,___ __751A~.JkJ -RV Inlet Avo. L 1,.Temperature C1 %9m;T I 170-F 40*F/Hr.260 ý T > 170T 40F/Hr..350 > T ) 250T 60F/Hr.T I 350T l1rnF/Hr.CV-I I 1 50 I 100 150 200 2II I 300 II 5I IIIIIIII 250 300 350 400 450 RV Inlet Temperature.

F PAldffanal resetiidons when head is an readmwveseL

1. Mmbaitin avWnie wze exi bunper3auie:

135'F >T > IUff for > 3houis 2- Fdkngompletion of item 4. mxzintainaverage h~uiy cocdoam (00)2 Oimit of 20'Fihour based an Figure 3.4.3-2 (Page 1 of 1)Pressure -Temperature Limits for Cooldown Figure 2. Pressure-Temperature Limits for Cooldown [71 Report No. 1001026.401, Rev. 1 28! Structural Integrity Associates, Inc.K I.CL L, a-2250-2000.1750'15001 less 759.59 I I lI I IIi I I I II I 111 I I I I I 1 190 159 209 250 339 I to I 41ll 450 RV Hnlot Temperoetre.

F Figure 3-4.2-1 (Page I of 1lLTOP Selpoint Limit Figure 3. LTOP Setpoint Limit [91 Report No. 100 1026.40 1, Rev. I 29 V Structural Integrity Associates, Inc.

350.00 300.00 250.00 200.00 u.0+ 14igma,+1-Sigma-__ _ _ _ __ _ _ _ _ ---------

--Best Fit CF =216.137F~ ~ ~~~~~~~~~~~~~~~~~~.

...... .................................

.. ..............

--- ----.. .... ---150.00 100.00 50.00 0.00 O.OOE+00 1.OOE+19 2.OOE+19 Fluence (n/cm 2)3.OOE+19 4.OOE+19 Figure 4. Best Fit for all Weld Heat No. 27204 Surveillance Data Report No. 1001026.401, Rev. 1 30 V Structural Integrity Associates, Inc.

40000 350 00 300.00 250.00 20000 0 150.00 100.00 50.00 0.00 O 00EO00 2.erEn)19 4.OOE+19 Fkence (n/cm 2)Figure 5. Best Fit CF for all W5214 Surveillance Data 117]31 V Structural Integrity Associates, Inc.Report No. 1001026.401, Rev. 1 300,00 250.00 I --+ ~Z-Sigpna 200.00 A A-------------

.....................

150.00 100.00 50.00-2-Sigma Best Fit CF= 147.71'F---- _ _ _ -_ _ _ _ _ ------_ _ I -0.00 O.OOE+00 1.OOE+19 2.00E+19 3.00E+19 4OOE+19 5.00E+19 Fhuence (n/cm 2)Figure 6. Best Fit CF for Base Metal Heat No. C-1279 Surveillance Data Report No. 1001026.401, Rev. 1 32 V Structural Integrity Associates, Inc.

100 80----Two Sigma 50-!o A5338-1 HSST01/02 CMM o Pho Materials x Weld MaterMs t Forging Materials K U)U)40-20-0--20-x a 0-x 0 rO 0 A~l~1 01 0 0 X KaADJ x x 5.,, I n M a-100 0 0 On 0 x o0gq X~ 0 I K 0 04o0 84 1k 4 V6 R-40-4-K a-80.(9 x x-804 a 0-1n Oi" 1.OOE+17 1AIOE4IG .00E+19 Fluence, E > 1 IeV [W=21~I.OOE+20 Figure 7. Plot of Residual vs. Fast Fluence for A533B-1 HSST-O1/HSST-02 CMM with Companion Materials, the Overall 2-Sigma Scatter is 50°F [36].Report No. 1001026.401, Rev. 1 33 Structural Integrity Associates, Inc.

APPENDIX A DATA CREDIBILITY ASSESSMENT FOR WELD HEAT NO. 27204 Report No. 100"1026.401, Rev. 1 A-I V Structural Integrity Associates, Inc.K DATA CREDIBILITY ASSESSMENT FOR WELD HEAT NO. 27204 The purpose of this evaluation is to apply the credibility requirements in Reg. Guide 1.99, Rev. 2[4] to the Palisades, and Diablo Canyon 1 surveillance capsule data and to determine if the surveillance capsule data is credible and can be used to improve the RTNDT predictions for the vessel circumferential weld heat No. 27204.Reg. Guide 1.99, Rev. 2 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. Reg. Guide 1.99, Rev. 2 also provides two methods for calculating the adjusted reference temperature of the reactor vessel beltline materials.

The first method is described in paragraph (c)(1), "Surveillance Data Not Available".

The second method is described in paragraph (c)(2), Surveillance Data. Available".

The procedures in paragraph (c)(2)can only be applied when two or more credible surveillance data sets-become available.

NRC provided additional guidance for evaluation and use of surveillance data in Reference 18.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.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 defined 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:

• Intermediate Shell, Axial Welds 2-112 A/B/C, material heat No. W5214,* Lower Shell, Axial Welds 3-112 A/B/C, material heat No. W5214 and 34B009,* Intermediate to Lower Shell, Circumferential Weld 9-112, material heat No. 27204,* Intermediate Shell, Plate D-3803-1, material heat No. C- 1279,* Intermediate Shell, Plate D-3803-2, material heat No. A-0313, Report No. 1001026.401, Rev. 1 A- 2 Structural Integrity Associates, Inc.Reporto.100026.41,Rev

• Intermediate Shell, Plate D-3803-3, material heat No. C-1279," Lower Shell, Plate D-3804-1, material heat No. C-1308A," Lower Shell, Plate D-3804-2, material heat No. C-1308B,* 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[33]. The Palisades reactor vessel surveillance program was originally developed with the intent to comply, where possible, with the guidance of ASTM E185-66, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors" [34]. At the time that the Palisades surveillance capsules were built, 10 CFR50 Appendices G and H did not exist.Palisades supplemental capsules SA-240-1 and SA-60-1 were reinserted into the Palisades vessel at the end of Cycle 11 and removed for testing at the end of Cycle 13. The capsules contain reconstituted Charpy specimens made from weld heat No. 27204 obtained from Fort Calhoun, another C-E designed plant with the same weld heat. The weld material was carefully chosen, including the post weld heat treatment condition, in order to match the Palisades vessel beltline weld. Because weld heat No. 27204 in the capsules matches the limiting circumferential weld, the beltline material with the highest adjusted reference temperature and the limiting material for P-T curves, 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. 27204 can be fitted to determine the 30 ft-lb temperature (T 3 0) and upper shelf energy (USE) unambiguously.

The data and Charpy energy curve fits for weld heat No. 27204 are shown in Appendix B. It was determined that the Charpy curve-fits have produced accurate 30 ft-lb temperatures and USE values. Hence, Criterion 2 is met for all the surveillance capsules evaluated here which contain weld metal heat No. 27204.Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT 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 El185.Report No. 1001026.401, Rev. 1 A- 3 Structural Integrity Associates, Inc.

The functional form of the least squares method as described in paragraph (c)(2) of Reg. guide 1.99, Rev. 2 will be utilized.

A best-fit line is generated for this data to determine if the scatter of the ARTNDT values about this line is less than 28°F for weld metal heat No. 27204.The Palisades limiting weld metal will be evaluated for credibility using the NRC recommended guidelines

[ 18]. Of the recommended methods, Case 4 most closely represents the situation for the Palisades surveillance weld metal where data is available from the plant of interest and from other plants.Case 4a Credibility Assessment

-Palisades 2 7204 Data Only The data most representative for the Palisades limiting vessel weld are the supplemental surveillance capsules containing weld heat No. 27204 since the irradiation environment of the surveillance capsules and the reactor vessel are the same. The data requires the least adjustment since the radiation conditions are the same as the vessel. The Palisades weld heat No. 27204 capsule data are shown in Table A-1 and in Figure 4, along with the fitted solution (i.e., mean shift prediction) result, and the comparison of the measured -predicted scatter from the fitted CF of 221.8°F. A plot of the measured AT 3 0 vs. fluence results for the Palisades supplemental capsule weld (27204) is shown in Figure 4 along with the +/- Icy bounds for credible data scatter.The data clearly fall within the 1-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.Table A-i, Evaluation of Palisades S~urveillance Data Results for Weld Heat No. 27204 Material Capsule Heat Capsule fluence FF 6RTndt FF*ARTndt FF-2 27204 SA-240-1 (PNP) 27204 2.38E+19 1.234 267.8 330 1.52 27204 SA-60-1 (PNP) 27204 1.50E+19 1.112 253.1 282 1.24 CF (SumSqrs)

Measured 1RTndt Predicted

&RTndt Scatter ARTndt 27204 5A-240-1 (PNP) 27204 221.8 267.8 273.6 -5.8 27204 SA-60-1 (PNP) 27204 221.8 253.1 246.7 6.4 IFFr*RTndt FFA2 Sum 611.9 2.8 Fitted CF 221.8Structural Integrity Associates, Inc.Report No. 1001026.401, Rev. 1 A- 4 Case 4b Credibility Assessment

-All 2 7204 Surveillance Capsule Data Following the guidance in Case 4 [18], the data from all sources should also be considered.

For weld heat No. 27204 there are a total of five surveillance capsules, two from Palisades and three from Diablo Canyon Unit 1. Since data are from multiple sources, the data must be adjusted first for chemical composition differences and then for irradiation temperature differences before/determining the least-squares fit.The five capsule results and the fitted CF value, as shown in Table A-2, is determined to be 216.13'F for this case. The results for.(measured

-predicted) scatter for all the 27204 surveillance data results are also shown in Table A-2. The results for all the surveillance capsule data are plotted in Figure 4 along with the +/- I a scatter bands. The scatter in the measured -predicted values does not exceed 28°F (1-sigma).

According to Reg. Guide 1.99, Rev. 2 paragraph (c)(2), 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 mustbe used for vessel RTPTS predictions, if the CF value is lower, it may be used. In this case the fitted CF value is lower.The chemistry factor from the tables in paragraph (c)(1) is 226.8°F, and the adjusted chemistry factor using the Palisades surveillance capsule data is 216.13'F.

It is noted that per NRC guidance that itis possible to use a lower value of chemistry factor based upon all sources of surveillance capsule data with a reduced margin term. if the data is credible in all other ways.Therefore, the weld data meets this criterion, and the Palisades surveillance program weld metal chemistry factor is determined to be 216.13'F.

Per the criteria in Reg. Guide 1.99, Rev. 2, the data are credible and the lower CF value may be used with a reduced (1-sigma) margin term of 44°F when calculating RTNDT for this weld heat in the Palisades vessel.Report No. 1001026.401, Rev. 1 A- 5 Structural Integrity Associates, Inc.

Table A-2. Evaluation of all Surveillance Capsule Results Containing Weld Heat No. 27204 Ratio Chem. &Table Revised Fluence Irrad. Measured Adjusted Temp. Adj. Predicted Adjusted -Capsule %Cu %Ni CF (F) Fhie nce Factor Temp. ARTndt ARTndt ARTndt ARTndt Predicted (n/cmA2) FF Ti (F) (F) (F) (F) (F) (F)CAP.Y (DCPP) 0.198 0.999 222.26 1.05E+19 1.01 542 232.59 237.3 244.1 219.1 25.06 CAP S (DCPP) 0.198 0.999 222.26 2.84E+18 0.66 544 110.79 113.1 121.9 141.8 -19.97 SA-240-1 (PNP) 0.194 1.067 227.8 2.38E+19 1.23 535.7 267.8 266.7 267.2 266.7 0.49 SA-60-1 (PNP) 0.194 1.067 227.8 1.50E+19 1.11 535 253.1 252.0 251.8 240.4 11.43 CAP V (DCPP) 0.198 0.999 222.26 1.37E+19 1.09 541.5 201.07 205.2 211.5 235.0 -23.56 Vessel Best Estimate CF= 226.8 Mean T= 535.2 ]I_ j _ _ [Least Squares Fitted CF = 216.13 Note: all five (measured

-predicted) data points are within the 1 standard deviation of 287F for credible data for welds.Report No. 100 1026.40 1, Rev. 1 A- 6 Structural Integrity Associates, Inc.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base 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 A-3 provides a history of the time-weighted temperature for the Palisades supplemental surveillance capsules and reactor vessel wall.Table A-3. 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 (°F)(OF)1 371.7 523 2 440.1 529 A-240 526 3 342.5 534 4 321.0 536 5 386.7 536 W-290 531 6 326.7 536 7 362.5 536 8 366.1 537 9 292.5 534 10 349.7 534 W-110 533 11 421.9 533 12 399.3 534 13 419.6 536 SA-60-1 535.0 14 449.3 537 SA-240-1 535.7 15 401.3 537 16 444.3 537 1 17 493.1 537 1 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)(b)Cycle length (EFPD) values obtained from Reference 35 Cycle average vessel temperatures obtained from Reference 37 (c) Cycles 1-12 EFPDs are reduced by 2% to account for power reduction factor per the guidance in [35]Report No. 10oo 1026.401, Rev. I A-7 q~r Structural Integrity 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 25°F. Therefore, this criterion is satisfied for the Palisades capsules.The Diablo Canyon 1 capsule irradiation temperatures are shown in Table A-2 [32, 24].These temperatures are also within 25°F of the Palisades average vessel irradiation temperature of 535.2°F.Criterion 5: 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 A 1195-1) from SA-60-1 and SA-240-1 are documented in BAW-2341 Rev 2 [27] and BAW-2398 [28], respectively.

Charpy energy versus temperature for the unirradiated correlation monitoring material (HSST Plate 02, Heat Al 195-1) is taken from NUREG/CR-6413, ORNL/TM-13133

[36]. Tables A-4 and A-5 provide the updated calculation of (measured

-predicted) scatter versus fast fluence in the correlation monitor material (HSST 02) data. Figure 7 (from Reference

36) shows that the measured scatter band for the correlation monitor materials is 50 0 F.Table A-4. Correlation Monitor Material HSST Plate 02 Calculation of Fitted CF Capsule Fluence Fluence Factor ARTNDT(c)

FF

• ARTNDT FF 2 (X 1019) (a) (FF) (b) (*F)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 = Y (FF x RTNDT) / Y (FF 2)= 300.305/2.7588

= 108.853 Slope of best fit line is 108.853 Notes: (a) Calculated fluence (x 1019 n/cm 2 , E>1.0 MeV)(b) FF = fluence factor = f(O.28-0.*iogf)(c) Irradiated values of 30 ft-lb Transition Temperature From BAW-2341 Rev 2 and BAW-2398 [27, 28]Report No. 100 1026.40 1, Rev. I A-8 V Structural Integrity Associates, Inc.

Table A-5. Correlation Monitor Material HSST Plate 02 Calculation of Measured -Predicted Scatter Capsule Fluence Fluence Factor ARTNDT(c)

Predicted (Measured

-(x 1019) (a) (FF) (b) ARTNDT Predicted)

ARTNDT 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 ARTNDT = (slope best fi)*(Fluence Factor)Slope of best fit line is 108.853 Notes: (a) Calculated fluence (x 1019 n/cm 2 , E>1.0 MeV)(b) FF = fluence factor = f(O.28-0.1*logf)(c) Irradiated values of 30 ft-lb Transition Temperature From BAW-2341 Rev 2 and BAW-2398 [27, 28]Table A-5 shows that the scatter in these data is less than 507F, which is the allowable scatter in NUREG/CR-6413, ORNL/TM-13133

[36]. Thus, criterion 5 is satisfied for the correlation monitor materials.

Report No. 1001026.401, Rev. 1 A-9 Structural Integrity Associates, Inc.

APPENDIX B SURVEILLANCE CAPSULE DATA FOR WELD HEAT NO. 27204 Report No. 100 1026.40 1, Rev. 1 B-i V Structural Integrity Associates, Inc.

BAW-2341.

Revision 2 May 2001 Test Results of Capsule SA-60-1 Consumers Energy Palisades Nuclear Plant Reactor Vessel Material Surveillance Program-..

by M. J. DeVan, FTI Document'No.

77.-2:341-02 (See Section 7 for document signatures.)

,/ Prepared for Consumers Energy Prepared by Framatome ANtP, Inc., 3315 Old Forest'Road

p. P* Box 10935 Lynchburg, Virginia 24506-0935 fPRAMATOME ANP Executive Summary This report describes the results of the test specimensfrom the first supplemental capsule (Capsule SA-60-1) of the Consumers Energy Palisades Nuclear Plant as partof their reactor vessel surveillance program. The objective0of the program is to monitor 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-60-4 was removed from the Palisades-reactor vessel at the end-of-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 !Snmm Charpy V-notch inserts for three weld metals fabricated with weld Wire heats W5214, 34B009, and 27204 and-standard Chaipy 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.Following the initial Charpy V-notch impact testing, the laboratoryperformed a calibration of the temperature indicator used in the Palisades Capsule SA-60-1 testing. The results of the'laboratory calibration indicated the instrument was out-of to1erance.

Based on the results of this calibration test, the laboratory revised the Chaipy impactltest temperatures accordingly.

Revision 1 corrects the test temperatures for the Supplemental Capsule SA-60-1 reconstituted weld metal'Charpy V-notchimpact specimens and the HSST Plate 02 Charpy V-notch impact specimens.

Revision2 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 ft-bs for all cases, and for each materials the upper-shelf.energy was fixed atthe average of all test energies exhibiting 1001%o shear, consistent with ASTM Standard E 185-82.ii / RAMATOME ANP of full size Type A Charpy V-notch specimens in accordance with ASTM Standard E:23-91.The recOnstituted Charpy specimen dimensions for each specimen are shown in Table 4-2. Upon completion of the machining of the reconstituted Charpy specimens, twelve (12) specimens were selected from each weld~metal for Charpy impact'testing.

4.5. Charpy V-Notch Impact Test Results LI The Charpy V-notch impact testing was performed in accordance with the applicable requirements of ASjTStandardB 23-91. Impact~energy, lateral expansion, and percent shear fracturewere, measured at numerous test temperatures and recorded for each specimen.

The impact energywas measured using a certified Satec Sl-1K Impact.tester (traceable to NIST Standard) with 240 ft-lb of available energy. The lateral'expahsion was measured using a certified dial indicator.

The specimen percent shear was estimated by video examination and comparison with the visual>standards presented in ASTM Standard E 23-9 1. IT addition, all Charpy V-n6tch impact testing was performed using instrumentation to record a load-versus-time trace and-energy-versus-time trace for each, impact event. The load-versus-time traces were analyzed to determine time, load, I and impact energy: for general yielding, ma:imum load, fast fracture, and crack-arrest properties during the test. Thedynamic yield stress iscalculated from the three-point bend, formula:/I o-y,= 33.33 *(general fielding load:)The dynamic flow'stress is calculated from the average of the' yield and'maximum loads, also using the three-point.bend formula: ii C.oo= 3 3.3 3* (generaly ielding load + maximum loadf ) 91 The results of the Charpy V-notch impact testing are shown'in Tables 4-3 through 4-10 and Figures 4-2 through .4-5, and the individual load-versus-timetraces for the instrumented Charpy V-notch impact tests are presented in Appendix B. The curves were generated using a hyperbolic' tangent curve-fitting program to produce the best-fit curve through the data, The hyperbolic tangent,(TANH) function (test response, i.e., absorbed energy, lateral expansion, and percent I shear fracture, "R," as a function of test temperature, 'T") used to evaluate the surveillance data is as follows: ]4-3 t RAMATOME ANP j ri For the absorbed (impact) energy curves, the lower-shelf energy was fixed at 2.2 ft-lbs for all materials, :and the upper-shelf energy was fixed atthe average of all test energiesexhibiting 100 percent shear for each material, consistent with~the ASTM Standard E 185-82. The lateral expansion curves were generated with the lower-shelf mils lateral expansion fixed at I mil,and the upper-shelf mils lateral expansion not constrained (i.e., not fixed). The percent shea fracture cutves for each material were generated with the lower-shelves and upper-shelves fixed at 0 and 100 respectively.

2 The Charpy V-notchdata was entered, and the coefficients A, B, To, and C are determined by the program minimizing the sum of the errors squared (least-squares fit) of the data points about the fitted curve. Using these coefficients and the above TANH function, a fsmooth curve is.generated through the data for interpretation of the material transition region behavior.

The coefficients determined for irradiated materials in Capsule SA-60t1, are shown in Table 4-11.The transition temperature shifts and upper-shelf"energy decreases for theCapsule SA-60-1 materials with respect to the unirradiatedmaterial properties are summarized in Table 4-12.Photographs of the Charpy V-notch specimen fracture surfaces are presented in Figures 4-6.through 4-9., 4-4 YFRAMATOME ANP Table 4-5. Charpy Impact Results for Palisades Capsule SA-60-1 Irradiated Weld Metal 27204 Test Impact Lateral Shear Specimen Temperature, Energy, Expansion, Fracture, ID OF ft-lbs mil %_PB68 74 12.5 7 0 PB56 129 16.5 13 40 PB81 154 17 10 30 PB78 204 25 19 45 PB93 229 28 27 70 PB91 254 39.5 35 85 PB28 279 44.5 39 95 PB96 329 52.51* 48 100 PB94 329 52* 50 100 PB15' 404 57* 53 100 PB42 454 55* 49 100 PB95 479 48.5* 43 100* Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82."5.1 I I I I I Ii I j I I I]7]I i FRAMATOME ANP 4-9 Table 4-11. Hyperbolic Tangent Curve-Fit Coefficients for the Palisades Capsule SA-60-1 Surveillance Materials Material Hyperbolic Tangent Curve Fit. Coefficients Description Absorbed Energy Lateral Expansion_

Percent Shear Fracture Weld Metal A: 28.4 A: 25.o0 A: 50.0 W5214 B: 26.2 B: 24.0 B.: 50.0 C: 158.1 C: 160.0 C: 80.5 TO,: 188.8 TO: 239.6 TO: 214.9 Weld Metal A: 28.7 A: 25.3 A: 50.0 34BO09 B: 26.5 B: 24.3 B: 50i0,-C: 123..8 C:, 97.61 C 89.6 TO: 161.8 TO: 196.4 TO: 179,6 Weld Metal A:: 27.6 A: 25.9. A: 50.0 27204 B: 25.4 B: 24.9 'B: 50.0 C: 111.4 C: 101.8 C:; 92.1 TO: 201.4 TO: 214.4 TO: i87.1 C2orrelation A: 44.3 A: 41.3' A: 50.0 Monitor Plate, B: 42.1 B: 40.:3. B: 50.0 HSST Plate 02 C: 95.1 C: 104.9 C: 85.2 (Heat No. A1195.-) TO: 193.0 TO" 208.6 :TO: 183.7 I I I I 71 1]2]I]I 71 71 11 I L]ii)I fFRAMA-TOME AN P 4-15 r~ F~ V~7 V~ ~** LZ2 ~2 U r r--.~ ~---~ [j~~ --_Table 412. Summary of Charpy Impact Test Results for the Palisades Capsule SA,60-1 Surveillance Materials 30 ftNb Transition Temperature, 50 ft-lb Transition Temperature, 35 mil Lateral Expansion Material OF -OF Transition Temperature, 0 F Upper-Shelf Energy, ft-lb Descr iptionj__________J_____I Uiadad Iraied A _______[_____

____Unirradiated.

Irradiated AT Uriirradiated Irradiated AT Unirradiated Irradiated AIrradiatd Decrease Weld Metal -6 0.2 a) 198.8 2ý59.0 -17.4(a) 375.6 393.0 -29.60 310. 1 339.7, 102.7(*) 54.5 48.2 W5214 WeId Metal -82.0(4) 167.8 249.8. -45,00) 298.6 343.6 -51.61") ,237.5 289.1 1413.9* ,55.25 58.65 34B009 Weld Metal -4 1 2 (b) 211.9 253.1 -6.1(b: 355.6 361.7 Not' 249.4 --- 108.4c' 53.0 55.4 27204 ava.ilable.

HSST Piate02 45.7ic) 159.4 1137. 7 8.31cJ 206.0 127.7 Not 187.9 -.. 120.3(cý 86.3 34.0.Heat.:NoAl195-1 available.

2 I-.0\(a) Data reported in ABA Technology Report AEA-TSD-0774., (b) Data reported'in C Report No..TR-MCC-189) 6 (c) Data reported in NUREGICR-6413.

1°0 3: rn T.

Figure 4-4. Charpy Impact Data-for Irradiated Weld Metal 27204 LL cc 100, 75 50 25 0-1 00 0 100 200 300!Temperature, F 400!500 ,600 1,0080 60 x 40' 20 fl 0-10...........

---

)00 0 100 2o00 30o TemperatUre, F 400 500 600 2 120 1 o0 CD'U.80 60 TUMLE: +23.4 F T'16: +35.6 CvUSE: 53+/-Ib......................

.: -.. ....,.. ... .. .....Matefil:l Weld Metal Heat Nurriber:

27204 140 1 20 I 0-100 0 100 200 300* Temperature, F 400 500 660 IARAMATOME ANP 4-20 BAW-2398 May 2001 Test Results of Capsule SA-240-1 Consumers Energy Palisades Nuclear Plant-- Reactor Vessel Material Surveillance Program --by M. J. DeVan FTI 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. 0. Box 10935 Lynchburg, Virginia 24506-0935 fA 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 monitor 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 ARAMATOME ANP the center position of the temperature verification mockup insert ranged from 347WF to 511 0 F, which is less than the Palisades reactor vessel cold-leg temperature and meets the temperature requirement of ASTM Standard E 1253-88.Twelve (12) stud-welded inserts were then selected from each of the weld metals W5214, 34B009, and 27204 for machining of full size Type A Charpy V-notch specimens in accordance with ASTM Standard E 23-91. The reconstituted Charpy specimen dimensions for each specimen are shown in Table 4-2.4.5. Charpy V-Notch Impact Test Results The Charpy V-notch impact testing was performed in accordance with the applicable requirements of ASTM Standard E 23-91. Prior to testing, the specimens were temperature-controlled in liquid immersion baths, capable of covering the temperature range -100WF to +550'F. Specimens remain immersed in the liquid medium at the test temperature

+/-2°F for at least 10 minutes before testing to assure achievement of thermal equilibrium.

A certified Omega Model 462 device was used to measure the temperature.

Impact energy, lateral expansion, and percent shear fracture were measured at numerous test temperatures and recorded for each specimen.

The impact energy was measured using a certified Satec S 1-1K Impact tester (traceable to NIST Standarda) with a striker Velocity of 16.90 ft/sec and 240 ft-lb of available energy. The lateral expansion was measured using a certified dial indicator.

The specimen percent shear was estimated by video examination and comparison with the visual standards presented in ASTM Standard E 23-91. In addition, all Charpy V-notch impact testing was performed using instrumentation to record a load-versus-time trace and energy-versus-time trace for each impact event. The load-versus-time traces were analyzed to determine time, load, and impact energy for general yielding, maximum load, fast fracture, and crack arrest properties during the test. The dynamic yield stress is calculated from the three-point bend formula: Cry = 33.33 * (general yielding load)The dynamic flow stress is calculated from the average of the yield and maximum loads, also using the three-point bend formula: a Each year, two sets of Charpy specimens are purchased from NIST and tested on the Charpy test machine. The results are then sent to NIST for evaluation.

A letter is then issued by NIST certifying the calibration of the Charpy test machine. The accuracy of the Charpy tester is +/-1 ft-lb or 5% of the dial reading whichever is greater.4-3 ARAMATOME AMP N' The results of the Charpy V-notch impact testing are shown in Tables 4-3 through 4-10 and Figures 4-2 through 4-5, and the individual load-versus-time traces for the instrumented Charpy V-notch impact tests are presented in Appendix B. The curves were generated using a hyperbolic tangent curve-fitting program to produce the best-fit curve through the data. The hyperbolic tangent (TANH) function (test response, i.e., absorbed energy, lateral expansion, and percent shear fracture, "R," as a function of test temperature, "T") used to evaluate the surveillance data is as follows: R =A + B *tanh[(TT°j For the absorbed (impact) energy curves, the lower-shelf energy was fixed at 2.2 ft-lbs'for all materials, and the upper-shelf energy was fixed at the average of all test energies exhibiting 100 percent shear for each material, consistent with the ASTM Standard E 185-82. The lateral expansion curves were generated with the lower-shelf mils lateral expansion fixed at 1 mil and the upper-shelf mils lateral expansion not constrained (i.e., not fixed). The percent shear fracture curves for each material were generated with the lower-shelves and upper-shelves fixed at 0 and 100 respectively.

The Charpy V-notch data was entered, and the coefficients A, B, To, and C are determined by the program minimizing the sum of the errors squared (least-squares fit) of the data points about the fitted curve. Using these coefficients and the above TANH function, a smooth curve is generated through the data for interpretation of the material transition region behavior.

The coefficients determined for irradiated materials in Capsule SA-240-1 are shown in Table 4-11.The transition temperature shifts and upper-shelf energy decreases for the Capsule SA-240-1 materials with respect to the unirradiated material properties are summarized in Table 4-12.Photographs of the Charpy V-notch specimen fracture surfaces are presented in Figures 4-6 through 4-9.4-4 fFRAMATOME ANP Table 4-5. Charpy Impact Results for Palisades Capsule SA-240-1 Irradiated Weld Metal 27204 Test Impact Lateral Shear Specimen Temperature, Energy, Expansion, Fracture, ID OF ft-lbs mil %PB45 70 5.5 3 0 PB62 125 16.5 12 10 PB71 175 16 18 30 PB54 200 26.5 29 55 PB07 200 33.5 27 60 PB73 225 29 24 65 PB52 250 34.5 26 55 PB35 300 36 32 65 P1306 350 44.5 43 95 PB58 400 49.5* 42 100 PB57 450 59* 52 100 PB61 500 53* 47 100 Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.[t17 4-9 ARAMAIOME ANP 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: 21.8 B: 50.0 C: 111.6 C: 83.5 C: 72.5 TO: 208.1 TO: 231.7 TO: 223.2 Weld Metal A: 29.8 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. Al 195-1) TO: 211.8 TO: 222.2 TO: 206.5 fFRAMATOME ANP 4-15

_j Table 4-12. Summary of Charpy Impact Test Results for the Palisades Capsule SA-240-1 Surveillance Materials 30 ft-lb Transition Temperature, , 50 ft-lb Transition Temperature, 35 mil Lateral Expansion

-Material OF OF Transition Temperature, 'F Upper-Shelf Energy, ft-lb Description Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated Decrease Weld Metal -60.2(a) 219.9 280.1 -17.4(a) 372.7 390.1 -29.6(") 284.3 313.9 102.7(a) 52.5 50.2 W5214 Weld Metal -82.0(a) 177.4 259.4 -45.0(2) 280.8 325.8 -51.6(a) 238.6 290.2 113.9s' 57.4 56.5 34B009 Weld Metal -41.2() 226.6 267.8 -6. 1I" 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.

0fl'(a) Data reported in AEA Technology Report AEA-TSD-0774.

1 9 1 (b) Data reported in CE Report No. TR-MCC-189.

1 1 8 1 (c) Data reported in NUREG/CR-6413.l I;A 0 3: m z I I ( I I I I .1 f I Figure 4-4. Palisades Capsule SA-240-1 Charpy Impact Data for Irradiated Weld Metal 27204*U al.U U 100 75 50 25 0 0 100 200 300 400 500 600 Temperature, F I w 100 80 60 40 20 0--- -- -- -- --- --.. ... ...0 100 200 1 300 400 500 600 Temperature, F 120 100 T 3%&E -+293.7 F T50: +399.7F T 3 o: +226.6F CvUSE: 53.8 ft-lb CL E 80 0 40 Material:

Weld Metal Heat NNu~mbber:

27204 20 0 0 100 200 300 400 500 600 Temperature, F 4-20 AFRAMATOME ANP WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION p WCAP-11567 I ANALYSIS OF CAPSULE-S FROM THE: PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON UNIT' 1REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S.E. Yanichko ,S. L. Anderson 3. C. Schmertz L. Albertin DECEMBER.

1987 ,Approved by: T. rA. eyerl Manager Structural Materials and Reliability Technology Work Performed -Under Shop*Order PFUJ-1I06 Prepared by.Westinghouse Electric Corporation for the Pacif ic Gas and Electric Company .Although information-contained in this report.is nonproprietary, no distribution shall be made outside Westinghouse or its.licensees without the customer's .approval..WESTINGHOUSE ELECTRIC CORPORATION Generation Technology Systems Divisiron P.O. Box 2728.Pittsburgh, Pennsylvania 15230-2728 890203003a 890129 PDR ADOCK 05000275 p -PDR SECTION I

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule S, the first cap'sule to be removed from the Pacific Gas and Electric Company Diablo Canyon Unit 1 reactor pressure -vessel, led to the following conclusions:

o The capsule received an average-fast neutron fluence, (E > 1.0 MeV)of 2.98 x 1018 n/cm 2.o Irradiation of specimens made from the reactor Vessel intermediate, shell plate B4106-3 to 2.98 x. 118 n/cm 2 resulted in 30 and 50 ft-lb transition temperature shifts of -2F and'4 0 F respectively, for specimens oriented parallel to the major working direction (longitudinal orientation).

o Specimens made from weld metal irradiated to 298 x 1018 n/cm2 S" resulted in 30 and 50 ft-lb transition temperature increases of 110*F and 148*F respectively.

Irrditio t 2.8 118 nc 2 o Irradiation to 238 x 10 n/cm2 resulted in a 11 ft-lb decrease in the upper shelf energy of the weld metal specimens and no decrease in the upper shelf of the shell plate 84106-3 specimens.

Both materials exhibit a more than adequate upper shelf level for continued safe plant operation.

o Comparison of the 30 ft-lb transition temperature increases for the Diablo Canyon Unit 1 surveillance materialwith predicted increases' using the methods of NRCRegu)atory Guide 1.99 proposed Revision 2 shows that the plate material and weld metal transition temperature increase are less than predicted.

o Capsule S contained specimens from the same heat of weld wire (Heat 27204) as the limiting reactor vessel weld seam. The surveillance program is therefore representative of the limiting reactor vessel material.25f-121087:10

• 1-1 C SECTION 5 TESTING OF SPECIMENS FROM CAPSULE S 5.1 Overview The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and:Development Laboratory with consultation by Westinghouse Power Systems Division personnel.

Testing was performed in accordance with IOCFR50, Appendices d and H,[21 ASTM Specification E185-82, and Westinghouse Procedure RMF-8402, Revision 0 as modifiedkby RMF Procedures 8102 and 8103.Upon receipt of the capsule at the laboratory,;

the specimens and spacer blocks were carefully removed, inspected foridentification number, and checked against the master list in WCAP-8465.[1J No discrepancies were found.Examination of the two low-melting point 3040C (579°F) and 3106C (590VF)eutectic alloys indicated no melting of either type of thermal monitor, Based on this-examination, the maximum temperature to which the test specimens were exposed-was less than 304C (579°F).The Chatpy impact tests: were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy instrumented with an Effects Technology Model 500 instrumentation system. With this system, and'energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve, the load of general'yielding (PGy), the time to general yielding (tGy), the maximum load Pm) and the time to maximum load (tM) can be determined.

The load at which fast fracture was initiated is identified as the fast fracture load (p- )and the load at which fast fracture terminated is identified as the arrest load (P K The energy at maximum load (EM) was, determined by comparing the energy-time record and the load-time record. The energy .at maximum load is roughly eqUiValent to the energy required to initiate a crack in the specimen,.

2Ms-12t2108710 5-1 Therefore, the propagation energy for the crack (E )is the difference, between the total energy to fracture (ED) and the energy at maximum load.The yield stress (oy) is talculated from the three-point bend formula.The flow stress is. calculated from the average ofthe yield and maximum loads, also using the three-point bend formula.Percent -shear ýwas determined from postfracture photographs using.the ratio'of-areas methods in compliance withASTM Specification A370 The lateral expansion was measured using a dial gage rig similar to that:shown in the same specification.

Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115),per ASTM Specifications E883 and E21-79, and RMF Procedure'8102.

All pull rod s, grips, and pins, were made of Inconel 718 hardened to Rc 45. The upper pull rod was connected througha universal joint to improve axiality of loading. The tests were conducted at a constant.crosshead speed of 0.05 inches per minute throughout the test.Deflection measurements Were made with a linear variable displacement transducer (LVDT) extensometer.

The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class 8'-2 per ASTME83-67.

Elevatedtest temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air.Because of the difficulty in remotely attaching a thervocouple directly to the specimen, the following procedure was used to monitor specimen temporature.Q Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. IF the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550*F (288*C) The upper grip

-52 Was used to control the furnace temperature.

During the actual testing the grip temperatures were used to obtain desired specimen temperatures.

Experiments indicated that this method is accurate to + 2"F The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, Ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were.determined from post-fracture photographs.

The. fracture area used to calculate the fracture stress (true stress at fracture)*and percent reduction in area was computed using the final diameter measurement.

5.2 Charpy V-Notch Impact-Test Results The results of Charpy V-notch impact tests performed on the various materials contained in Capsule S irradiated at 2.98 x 1018n/cm are presented in tables 5-1 through 5-5 and figures 5-l through 5-4. Initial and irradiated transition temperature and upper shelf energy levels were determined using a hyperbolic tangent (TANH) curve fitting model as used by 01dfield [3] to fit the data. Tables containing the results of the curve fitting were suppli ed to Westinghouse by the Pacific Gas.and Electric Company for use in determining the impact property changes. The transition temperature increases and upper shelf energy 'decreases for the Capsule S material are summarized in table 5-7.Irradiation of vessel intermediate shell plate B4106-3 material (longitudinal orientation) specimens to 2.98 x 1018 n/cm 2 (figure 5-1) resulted in a 30 and 50 ft-lb transition temperature shift of -2 and 4!F, respectively, and an average upper shelf energy increase of 4 ft-lb. The small increase in upper shelf energy is not unusual and is considered to be the result of data scatter.Weld metal irradiated to 2.98 x 1018 n/cm 2 (figure 5-2) resulted in 30 and 50 ft-lb' transition temperature increases of 110 'and 148*F respectively and an average upper shelf energy decrease of ft-lb.

10 5-3 Weld HAZ metal irradiated to 2.98 x 1018 n/cm (figure 5-3) resulted in So and 50 ft-lb transitions temperature increases of 77 and 56*Frespectively and an average upper shelf energy decrease of 22 ft-lb.HSST plate 02 correlation monitor material irradiated to 2.98 x ;018 n/cm2 (figure 5-4) resulted in a 30 and 50 ft-lb transition temperature increase of 66 and 60BF, respectively, and an average upper shelf energy decrease of 1 ft-lb,, which fall within the data base for this material.The fracture appearance of each irradiated Charpy specimen from the various materials is shown in figure 5-5 through 5-8and show an increasing ductile or tougher appearance

'With increasing test temperature.

Table 5-8 shows a comparison of the 30 ft-lb transition temperature increases for the various Diablo Canyon Unit 1 surveillance materials:

with predicted increases using the methodsof proposed NRC Regulatory Guide 1.99 Revision 2.t4] This comparison shows that the transition temperature increase resulting from irradiation to 2.98. x 1018 n/cm 2 is less than predicted by the Guide, for plate B4106-3, weld metal and correlation monitor material.Four weld metal Charpy V-notch impact specimens from Capsule S were reconstituted by Westinghouse to obtain additional impact toughness data to better define -the transition region and the'upper shelf of-the weld metal. A separate report `[5] describes the reconstitution procedure and discusses the analysis of the test. data.. Two of the four reconstituted specimens were notched too. deep and therefore were considered inappropriate for obtaining reliable test data. Table5-,9 shows the results of the impact tests performed at 125 and 400LF on the other two reconstituted specimens.

The toughness results for these two tests when compared to the original irradiated weld metal test resultsishown in figure 5-2 appear to be questionable since they do not fit the irradited energy and lateral expansion transition curves-. A review of the reconstitution and testing techniques used in this prog-Pam was conducted which did not identify any obvious abnormalities that could have produced the lower than expected toughness values,. The reconstituted Charpy data points have not been included in the development of the Charpy transition curves. Only two of the four reconstituted specimens were successful in 2SM9-121087:10 514 TABLE 5-7 EFFECT OF 550"F IRRADIATION AT 2.986 10i8 n/cm2 (E > I.0 MeV)ONANOTCH TOUGHNESS PROPERTIES OF DIABLO CANYON UNIT I REACTOR VESSEL MATERIALS Average, 50 ftIb Temperature

!(.,F)Unirradiated Irradiated AT Averagie 35,mil Lateral Expantlon Temperature

(*F.)UnIrradiated Irradiated AT Average 30 ft-lb Temperature

(°F)Unirradlated Irradiated AT Average Upper Shelf Energy (ft-lb)'Unirradiated Irradiated A(ft-lb)Material I Plate 84106-3 (Longltudinal)ý Weld Metal HAZ Metal Correlation Monitor MtV1 41 45 4 125 148*29 29 0 5-23--111-46 96 142-67 3 -2 122 43 10 98-91 77 147 112 66 124 126 87 125 123.4-11-22-55, 56 -107-64 43 -168 124 65 46 78 146 68 59-*1 U'I-i-, 2589s:-121OS7:l0 0 A 0 S U a 6 N a I.S 4I.1 110 too go so'70, 50 40 30 Is 10 --.00-100 too

• o00 S00 Teazap.w.tuu, (p')U a 4'-U I.I N I" A I.a A U 50 30 SO 0-300-too10 300 300 C UAIwsfdkisfe A b'ditatd Figure 5-2.Irradiated Charpy.V-Notch Impact Properties for Diablo Canyon Unit 1 Reactor Vessel Weld Metal 5-4.

WESTINGHOUSE CLASS 3 WCAP-13750

((ANALYSIS OF CAPSULE Y FROM THE PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON UNITl1 REACTOR VESSEL RADIATION SURVEILANCE PROGRAM E. Terek S. L Anderson A. Madeyski July 1993 Work Performed Under Shop Order LUKP-106 Prepared by Westinghouse Electric Corporation for the Pacific Gas and Electric Company Approved by.; 1 T 6 T. A. Meyer, Managef Structural Reliability and Plant life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355

© -1993 Westinghouse Electric Corporation 1.0

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule Y, the second capsule to be removed from the Pacific Gas and Electric Company Diablo Canyon'Unit 1 reactor pressure vessel, led to the following conclusions:

o The capsule received an average fast neutron fluen#e (E> 1.0 MeV) of 1.02,x i0T' n/cm 2 after 5.86 EFPY of plant operation.

irradiation of the reactoruvessel inter mediate shell plate B4106,3 CQarpy "speimens, oriented wih the. longitudinal axis of the specimen parallel to the major rollingdirection (longitudinal orientation), to 1.02 X i019 n/cm 2 (E 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 47 0 F and a 50 ft-lb transition temperature increase of 53,*F. 'iis results in a 30 ft-lb transition temperature of 52 0 F and a 50 ft-lb transition temperature of 94*F for. the longitudinally oriented specimens.

o Irradiation of the weld metal Charpy specimens to 1.02 x 10" n/cm 2 (E'> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 2347 and a 50 ft-lb transition temperature increase of This results in a 30 ft-lb transition tempeature of 167 0 F and a 50 ft-lb transition temperature of 2537F.o Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 1.02 x 1019 n/CM 2 (E > 1.0 MeV) resulted.

in a 30 ft-lb transition temperature increase of and a 50 ft-lb transition temperature increase of 757F. Thlis results in *a30 ft-lb transition temperature of -84OF and a 50 ft-lb transition temperature of6 -36F.o Irradiation of the Correlation Monitor Material Plate HSST 02 Charpy specimens to 1.02 x 10" ncm 2 (E> 1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 1127.This result$ in a 30 ft-lb transition temperature of 158 0 F..and, a 50 ft-lb transition temperature of 190T.1-1 o The surveillance Capsule Y test results indicate that the Correlation Monitor Plate HSST 02 material 30 ft-lb transition temperature shift is 10TF greater than the RegUlatory Guide 1.99, Revision 2 prediction.

This increase is bounded by the 2 sigma allowance for shift prediction of 34TF. The average upper shelfenergy decrease of the Correlation Monitor Material is less than the Regulatory Guide 1.99, Revision 2 prediction.

o The surveillance capsule materials exhibit a more than adequate upper shelfoenergy level for continued safr plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life (32 EFPY) of- the vessel as required by 1OCFR50, Appendix G.The calculated end.-of-life (32EFPY) maximum neutron fluencer(E

> 1.0 MCV) for the Diablo Canyon Unit 1 reactor vessel is as follows-, Vessel inner radius' = 154 x 1019 n/cm2 Vessel 1/4 thickness

= 8.10 x'018 Vessel 3/4 thickness

= 1.62 x 10 18 n/cn 2 Clad/base metal interface o Based on the criteria given in Regulatory Guide 1.99i Revision 2, the Diablo Canyon Unit 1 Surveillance Proigram is judged to be credible.1-3 SECTION 5.0 TESTING OF SPECIMENS FROM CAPSULE Y 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology -Center hot cell with consultationby -Westinghouse Power Systems personnel.

Testing was performed in accordance with 1OCFR5O, Appendices G and HI3, ASTM Specification E185-82m, and Westinghouse Remote Metallographic Facility (RMF) Procedure 8402, Revision 2 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master' list in WCAP-8465 1 1 1.No discrepancies were found., Examination of the two low-melting point 579F (304 0 C) and 590"F (3100C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579SF (304"C).The Charpy impact tests were performed per ASTM Specification E23-92P) and RMF Procedure 8103, Revision 1 on a. Tinius-OlsenModel 74, 358i machine. The tup-(striker) of the (Clarpy machine is instrumented with a GRC 8301 instrumentation system,ifeeding into an IBM XT Computer.

With this system, load-time" and energy-time signals can be recorded in addition to the.standard measurement of Charpy energy (ED). From the load-time curve (Appendix A), the -load of general yielding the time to general yielding (tay), the maximum load (PM), and. thetime to maximum load (ti) can be determined.

Under some test conditions,.

a sharp drop in load indicative of fast fracture was observed.The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA). The energy at maximum load (EM)was determined by comparing the energy-time record and the load-time record. The energy at maximum-load "is roughly equivalent to the energy required to initiate a crack in the specimen.Therefore, the propagation energy for the crack is thedifference between:the total energy to fracture (ED) and the energy at maximum load (Em).5-1 The yield'stress(o) was calculated from the three-point bend formula having the; following expression:P * { L/[B (W-a )2 c] (1)where L --distance between the specimen supports in the impact testing machine; B the width ofthe spdcimen measured parallel to the notch; W.= height of the specimen, measured perpendicularly to the notch; a = notch depth. The constant C is dependlent on the notch flank angle (0), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending).In three-point bending a Charpy specimen in which ) = 450 and p = 0.010", Equation 1 is valid with C = 1.21. Therefore (for L =-4W), L=P { /L-[B(W- a)2 , .21 [3 P.W ]/[B(W -a)2-] (2)For the Caarpy specimens, B = 0394 in., W = 0.394 in., and a = 0.079 in. Equation 2 then reduces to: o= 33.3 *Pay (3)where oyis in units of psi and Pcy is in units of lbs. The flow stress was calculaied from the average of the yield and maximum loads, also using the three-point bend formula.Percent shear was determined from post-fracture photographs using the ratio-of-area methods in compliance with ASTM Specification A370.92t9.

The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests'were performed on a 20,000-pound Instron Model 1115, split-console testmachine, per ASTM Specification E8-9 1 10(1 and E21-79 (1988)['11, and RMF Procedure:8102, Revision.

1. All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45. Theupper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of '0.05 ,inches per-minute throughout the test.5-2 Deflection measurements were made with a linear variable displacement transducer (LVDT)extensometer.

The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 :inch. The extensometer is rated as ClassB-2 per ASTM Elevated test temperatures were obtained with a three-zone electric resistance split4tube furnace with a.9-inch, hot zone. All tests- were conducted in air.Because of thedifficultyin remotely attachinga thermocouple directly to the specimen, the following procedure was, used to monitor specimen!temperature.

Chrome l-alumel thermocouples were inserted in shallow holes in the center and each end ofthe:gage section of a dummy specimen and in each grip.In the test configuration, with'a slight load on the specimen, a plot of specimen temperature versus upper and lower. grip and controller temperatures was developed' over the range. of room temperature to/ 5507F (288°C). The upper grip was used to control the" furnace temperature.

During the actual testing the grip temperatures were used to obtain desired specimen temperatures.

Experiments indicated that this method is accurate to +2 0 F.(ii The yield load, ultimate load, fracture load, total elongation, and uniform elongation were detrmined directly from the load-extension curve. The yield strength, ultimate strength,.

and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from Post-fracture photographs.

The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was the finaldiameter measurement.

5.2 Charmv V-Notch ImDactTest Results, The results of the Charpy V-notch impact tests performed on the various materials.

contained in.Capsule Y, which was irradiated to 1.02 x 10W9 n/cmr 2 (E > 1.0 MeV), are presented in Tables 5-1 through 5$8 and are compared with inirradiated results" 1 2 as shown in Figures 5-1 through 5-12. 'The transition temperature increases and upper rself energy decreases for the Capsule.Y materials are summarized in Table 5-9.'5-3 Irradiation of the reactor vessel intermediate shell plate B4106-3 Charpy specimens oriented with, the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal (orientation) to 1.02 x 101.0 n/cm 2 (E > 1.0 MeV) at 550T (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 47*F and a 50 ft-lb ,transition temperature increasef 53-F. This resulted in a 30'ft-lb transition temperature of 52*F0and a, 0 ft-lb transition temperature of 94.F (longitudinal orientation).

The average, upper shelf energy (USE) of:the intermediate shell plate B41063 Charpy specimens (longitudinal orientation) resulted in a energy decrease of 3 ft-lb after irradiation to 1.02 x 1019 n/cm 2 (E > 1.0 MeV) at 5507F, This results in an average USE of 119 ft-lb (Figure 5-1).Irradiation of the surveillance weld metal Charpy specieiis to 1.02 x 1019 n/cm 2 (E> :1.0 MeV) at 550 ,(Figure 5-4) resulted in a 30 ft-lb transition temperature shift of 234*F and-a 50 ft-lb transition temperature incease of 276°F. TIs results in a 30 ft-lb transition temperature of 167TF and a 50 ft-lb tranition temperature of 253-F.The average upper, shelf energy (USE) of the surveillance weld metal resulted inan energydecrease ofd 32 ft-lb after irradiation to 1.02 x 109 n/cm 2 (E> 1.0 MeV) at,550-?.

This resulted in an average (USE of 66 ft-lb (Figure 5-4).Irradiation of the reactor vesse weld HAZ metal Charpy specimens to 1.02 x I1' n/cm 2 (E> 1..0 MeV):at 550F (Figure 5-7) resulted in a 30 ft-lb transition temperature increase of 847F and a 50 ft-lb transition temperature increase of 750F. This resulted in a 30 ft7lb transition temperature of-84,F and a 50 ft-lb transition temperature f -360F.The average upper shelf energy (TUSE),of'the weld HAZ metal resulted in an energy de of 37 ft-lb after irradiation to 1.02 x 1019 n/cm 2 (E >1.0 MeV) at 5500F. This resulted in an average USE of 110 ft-lb (Figure 5-7).Irradiation of the HSST 02 correlation monitor material Charpy specimens to 1.02 x 101" n/cm 2 (E.>1.0 MeV) at 5500I (Figure 5-10) resulted in 30 and 50 ft-lb transition temperature increases of 1120F.This results in a 30 ft-lb transitionr temperature of 1587? and a 50 ft-b transition temperature of 1900F.5-4 The average upper shelf energy of the HSST 02 correlation monitor material experienced an energy decrease of 2 ft-lb after irradiation to 1.02 x 1019 n/cm 2 (E > 1.0 MeV) at 550TF. This resulted in an average USE of 122 ft-lb (Figure 5-10).Plots of the Capsule Y CMarpy test results are presented in Figures 5-13 through 5.24.Thie fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-25 through 5-28, and :show an increasingly ductile or tougher appearance with increasing test temperature.

A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the various Diablo Canyon Unit 1 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 214',is presented in Table 5,10 and led to the following conclusions:

o This comparison indicates that the transition temperature increases and the USE decreases for intermediate shell Plate B4106-3 resulting from irradiation to 1.02 x 10"9 n/cm 2 (E > 1.0 MeV) are less than the Regulatory Guide predictions.

o This comparison indicates that the surveillance weld metal 30 ft-lb shift in transition temperature is 10*F greater than the Regulatory Guide 1.99, Revision 2 prediction.

However, this increase is bounded by the 2 sigma allowance for shift prediction of 56 0 F. The average upper shelf energy decrease of the surveillance weld metal is less than the Regulatory Guide 1.99, Revision 2.prediction.

S bThis comparison indicates that the HSST 02 correlation monitor material'30 ft-lb shift in transition temperature is iO 0 F greater than the Regulatory Guide 1.99, Revision 2 prediction.

However, this increase is bounded by the 2 sigma allowance for shift prediction of 34 0 F. The average upper shelf energy decrease of the HSST 02 correlation monitor material is less than the Regulatory Guide 1.99, Revision 2 prediction.

The load-time records for the individual instrumented Charpy specimen tests are presented in Appendix A.5Z5

-- TABLE 5-9 Effect of 550*F Irradiation to 1.02 X 1W)9 n/cm 2 (E > 1.0 MeV) on the Notch ToughnessProperties of the Diablo Canyon Unit1 Reactor Vessel Surveillance Matcriaglsb Average.30 (ft-lb) ° Average-35 mil Lateral ( Average 50 ft-lb { Average'Energy Absorption

°Transition Temperature

(*F) Expansion Temperature (OF) Transition Temperature(0 F) at Full Shear (ft-lb)Material F)a-Fl'ha f-b Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated' Irradiated A Plate B4106-3 5 52 47 29 75 46 41 94 53 122,1 119 -3 (longitudinal)

(110) (-12)-Weld Metal -67 167 234 -46 195 241 -23 253 276 98 66 -32 ,(60) (-38)HAZ Metal -168 -84 84 -107 -36 71 -111 -36 75 147 110 -37_(109) (-38)Correlation 46 158 112 59 178 119 78 i90 112 124 122 -2 Monitor Mt'l ......._L(112) (-12)(a) "Average" is defined as the value read from the curve fit through the data .oints rof the Charpy tests (see Figures 5-1 through 5-4)(b) The data were fit by PG&E using the EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2-0")(c), Values in parenthesis were calculated per the definition of Upper Shelf Energy given in ASTM E185-82)(d) Unirradiated values presented here are from the Capsule "S" AnalysisM 5-14

/TABLE 5-10 Comparison of the Diablo Canyon-Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shiftsand Upper Shelf Energy.Decreases withRegulatoryGuide 1.99 Revision 2 Predictions 30 ft-lb Transition Upper Shelf Energy Temperature, Shift Decrease Fluence:Material Capsule (X .10 1 9-r/cm 2) Predicted (a) Measured Predicted (a) Measured"),_(°.) ( (9) (M)Plate B4106,3 S 0.305, 35 -2 14 0 (Longitudinal)

Y L102 52 47 '19 3 (10)Surveillance .Weld S 0.305 150 110 26 11 Metal Y 1.02 224 234 34 33(39)Heat Affected Zone S 0.305 -- 77 -15 Metal Y 1.02 -- 84 -- 26(26)Correlation Monitor S 0.305 68 66 18 2 Plate HSST 02 Y 1.02 102 112 23' 2(10)(a) Based on Regulatory Guide 1.99, Revision 2 methodology, using Mean w. % values of Cu and Ni.(b) Values in parenthesis were calculated per the definition of Upper:Shelf Energy given in ASTM E185-82M 5-15 (f C v R r 9 1F L B a-300 -200 -100 0 100 200 300 400 Soo 600 Temperature in Degrees F D 0 Unirradiated Test Data CapsuleS Test Data CapsuleY' Test Data5-16 Carpy V-Notch Impact Energy vs.- Temperature for Diablo Canyon Unit 1 Surveillance Weld Metal.11 5-32 Westinghouse Non-Proprietary Class 3 WCAP-15958 Revision 0 January 2003 Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program* Westinghouse WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15958, Revision 0 Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit I Reactor Vessel Radiation Surveillance Program A. R. Rawluszki J. Conermann R. J. Hagler January 2003 Approved:

J, V -J.A. Gresham, Mat~ger Engineering

& Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

©2003 Westinghouse Electric Company LLC All Rights Reserved ix EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule V from Diablo Canyon Unit 1. Capsule V was removed at 14.27 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed.

A fluence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI database.

Capsule V received a fluence of 1.37 x 1019 n/cm 2 after irradiation to 14.27 EFPY. This is equivalent to a vessel fluence at the end of the current license (32 EFPY). The peak clad/base metal interface vessel fluence after 14.27 EFPY of plant operation was 6.07 x 1018 n/cm 2.This evaluation lead to the following conclusions:

Specimen results are behaving in accordance with predictions.

The surveillance program, however, does not meet the regulatory criteria for credibility.

Regulatory Guide 1.99 requires that all five criteria for credibility be met. For the Diablo Canyon Unit I surveillance program, four out of five of the criteria for~credibility were met. A brief summary of the 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.

1-1 1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule V, the fifth capsule removed (third capsule tested) from the Diablo Canyon Unit 1 reactor pressure vessel, led to the following conclusions:

The Charpy V-notch data'presented in WCAP-8465" 3', WCAP-1 1567411 andWCAP-13750 1 5 3 were based on Charpy curves using a hyperbolic tangent curve-fitting routine. The results presented in this report are based on a re-plot of all capsule data using CVGRAPH, Version 4.1. which is a symmetric hyperbolic tangent curve-fitting program. Appendix B presents a comparison of the Charpy V-Notch test results for each capsule based on previous fit vs. symmetric hyperbolic tangent fit. Appendix C presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data., o Capsule V received an average fast neutron fluence (E> 1.0 MeV) of 1.37 x 1019 n/cm 2 after 14.27 effective full power years (EFPY) of plant operation.

Irradiation of the reactor vessel intermediate shell plate B4106-3 (heat number C2793-1) Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 39.46°F and an irradiated 50 ft-lb transition temperature of 77.510F. This results in a 30 ft-lb transition temperatu'e increase of 34.32'F and a 50 ft-lb transition tempeiature increase of 38.19'F for the longituldinal oriented specimens.

See Table 5-9.Irradiation of the weld metal (heat number 27204) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 135.45°F and an irradiated 50 ft-lb transition temperature of 21 9.26F. This results in a 30 ft-lb transition temperature increase of 201.07*F and a 50 ft-lb transition temperature increase of 243.43F. See Table 5-9.Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -52.65'F and an irradiated 50 ft-lb transition temperatureof-l.98'F.

This results in a 30 ft-lb transition temperature increase of 110.91F and a 50 ft-lb transition temperature increase of 109.77'F.

See Table 5-9.Irradiation of the Correlation Monitor Material Plate HSST02 Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 163.05*F and an irradiated 50 ft-lb transition temperature of 197.42'F.

This results in a 30 ft-lb transition temperature increase of 116.61 IF and a 50 ft-lb transition temperature'increase of 119.12'F.

See Table 5-9.The average upper shelf energy of the intermediate shell plate B4106-3 (longitudinal orientation) resulted in no energy decrease after irradiation.

This results in an irradiated average upper shelf energy of 118 ft-lb for the longitudinal oriented specimens.

See Table 5-9.Summary of Results 1-2 The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 25 ft-lb after irradiation.

This results in an irradiated average upper shelf energy of 66 ft-lb for the weld metal specimens.

See Table 5-9.The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation.

This results in an irradiated average upper shelf energy of 116 ft-lb for the weld HAZ metal. See Table 5-9.The average upper shelf energy of the Correlation Monitor Material Plate HSST02 Charpy specimens resulted in an average energy decrease of 6 ft-lb after irradiation.

This results in an irradiated average upper shelf energy of .117 ft-lb for the weld correlation monitor metal. See Table 5-9.A comparison, as presented in Table 5-10, of the Diablo Canyon Unit I reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 211] predictions led to the following conclusions:

The measured 30 ft-lb shift in transition temperature for all the surveillance materials of Capsule V contained in the Diablo Canyon Unit I surveillance program are in good agreement or less than the Regulatory Guide 1.99, Revision 2, predictions.

, The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules V contained in the Diablo Canyon Unit I surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions.

The credibility evaluation of the Diablo Canyon Unit I surveillance program presented in Appendix D of this report indicates that the surveillance results are not credible.

This is based on not satisfying the third criterion for credibility.

o 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 of the vessel (32 EFPY) as required by IOCFR50, Appendix G 12)The calculated and best estimate end-of-license (32 EFPY) neutron fluence (E> 1.0 MeV) at the core midplane for the Diablo Canyon Unit I reactor vessel using the Regulatory Guide 1 .99, Revision 2 attenuation formula (i.e., Equation #3 in the guide) are as follows: Calculated:

Vessel inner radius* = 1.26 x 1019 n/cm 2 Vessel 1/4 thickness

= 7.51 x 1018 n/cm 2 Vessel 3/4 thickness

= 2.67 x 1018 n/cm 2 Summary of Results 4-4 Table 4-3 Chemical Composition (wt%) of the Diablo Canyon Unit I Reactor Vessel Surveillance Materials (Unirradiated)(c)

HSST 02 Element Intermediate Shell Plate Weld Metal 'b)B4106-3 Ladle Check N 0.010 0.009 --C 0.200 0.140 0.22 0.22 Si 0.250 0.450 0.22 0.25 Mo 0.460 0.480 0.53 0.52 Cu 0.077 0.210 -0.14 Ni 0.460 0.980 0.62 0.68 Mn 1.330 1.360 1.45 1.48 Cr 0.035 0.060 --V 0.001 0.001 -Co 0.001 a) 0.001(a) -Sn 0.007 0.010 -Zn 0.001 a) 0.056 -Ti 0.0011a1 0.010 --Zr 0.0011a, 0.030 --As 0.009 0.016 --Sb 0.001 0.003 --S) 0.012 0.025 0.019 0.018 P 0.011 0.016 0.011 0.012 Al 0.036 0.018 --B 0.(x)3la, 0.03'a) -Notes: (a) Not detected, the number represents the minimum of detection.(b) Surveillance weld was made of the same weld wire Heat 27204 and Linde 1092 Flux as the beltline region reactor vessel intermediate and lower shell longitudinal weld seams. Linde 1092 flux lot 3714 was used to fabricate the surveillance weld whereas flux lot 3724 and 3774 was used to fabricate the intermediate and lower shell longitudinal weld seams respectively.(c) This table was taken from WCAP-13750 1 5 1.Description of Program 4-5 The best estimate copper and nickel weight percent remains as presented in the Diablo Canyon Unit I FSAR. The values used for the intermediate shell plate B4106-3 (Heat Number C2793-1) in all calculations documented in this report are as follows: Cu wt. % 0.086, and Ni wt. % =0.476 The values used for the surveillance weld (Heat Number 27204)* in all calculations documented in this report are as follows: Cu wt. % = 0.198, and Ni wt. % = 0.999* The overall best estimate Cu and Ni for heat 27204 is 0.203 Cu and 1.018 Ni. These values are documented in the Diablo Canyon Unit I FSAR.Description of Program 5-3 calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs.

The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

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 V, which received a fluence of 1.37 x 10 ' n/cm'(E> 1.0 MeV) in 14.27 EFPY of operation, are presented in Tables 5-1 through 5-8 and are compared with unirradiated results 1 3 1 as shown in Figures 5-1 through 5-12.The transition temperature increases and upper shelf energy decreases for the Capsule V materials are summarized in Table 5-9 and led to the following results: Irradiation of the reactor vessel Intermediate Shell Plate B4106-3 (heat number C2793-1) Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation) resulted in an irradiated 30 ft-lb transition temperature of 39.46tF and an irradiated 50 ft-lb transition temperature of 77.51'F. This results in a 30 ft-lb transition temperature increase of 34.321F and a 50 ft-lb transition temperature increase of 38.19'F for the longitudinal oriented specimens.

See Table 5-9.Irradiation of the weld metal (heat number 27204) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 135.45'F and an irradiated 50 ft-lb transition temperature of 219.26 0 F. This results in a 30 ft-lb transition temperature increase of 201.07'F and a 50 ft-lb transition temperature increase of 243.43°F.

See Table 5-9.Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-52.65°F and an irradiated 50 ft-lb transition temperature of -1.98 0 F.This results in a 30 ft-lb transition temperature increase of 11 0.9°F and a 50 ft-lb transition temperature increase of 109.77°F.

See Table 5-9.Irradiation of the Correlation Monitor Material Plate HSST02 Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 163.05'F and an irradiated 50 ft-lb transition temperature of 197.42°F.This results in a 30 ft-lb transition temperature increase of 116.61°F and a 50 ft-lb transition temperature increase of 119.12'F.

See Table 5-9.The average upper shelf energy of the Intermediate Shell Plate B4106-3 (longitudinal orientation) resulted in no energy decrease after irradiation.

This results in an irradiated average upper shelf energy of 118 ft-lb 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 25 ft-lb after irradiation.

This results in an irradiated average upper shelf energy of 66 ft-lb for the weld metal specimens.

See Table 5-9.Testing of Specimens from Capsule V 5-4 The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation.

This results in an irradiated average upper shelf energy of 116 ft-lb for the weld HAZ metal. See Table 5-9.The average upper shelf energy of the weld correlation monitor metal Charpy specimens resulted in an average energy decrease of 6 ft-lb after irradiation.

This results in an irradiated average upper shelf energy of 117 ft-lb for the weld correlation monitor metal. See Table 5-9.A comparison, as presented in Table 5-10, of the Diablo Canyon Unit I reactor vessel beltline material test results with the Regulatory Guide 1.99, Revision 211) predictions led to the following conclusions:

-The measured 30 ft-lb shift in transition temperature for all the surveillance materials of Capsule V contained in the Diablo Canyon Unit I surveillance program are in good agreement or less than the Regulatory Guide 1.99, Revision 2, predictions.

-The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules V contained in the Diablo Canyon Unit 1 surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions.

The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule V materials is shown in Figures 5-13 through 5-16 and shows an increasingly ductile or tougher appearance with increasing test temperature.

All beltline materials exhibit a more than adequate upper shelfenergy 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 (32 EFPY) as required by IOCFR50, Appendix G t2]The load-time records for individual instrumented Charpy specimen tests are shown in Appendix A.The Charpy V-notch data presented in WCAP-8465' 1', WCAP-1 1 5 6 7 t41 and WCAP-13750E 5 1 were based on hyperbolic tangent curve fitting. The results presented in this report are based on a re-plot of all capsule data using CVGRAPH. Version 4.11141, which is a symmetric hyperbolic tangent curve-fitting program. Appendix B presents a comparison of the Charpy V-Notch test results for each capsule based on previous fit vs. symmetric hyperbolic tangent fit. Appendix C presents the CVGRAPH, Version 4. 1, Charpy V-notch plots and the program input data.Testing of Specimens from Capsule V 5-7 Table 5-2 Charpy V-notch Data for the Diablo Canyon Unit I Surveillance Weld Metal Irradiated to a Fluence of 1.37 x 1019 n/cm 2 (E> 1.0 MeV)Sample Temperature Impact Energy Lateral Expansion Shear Number OF °C ft-lbs Joules mils mm %WII .25 -4 11 15 1 0.03 5 W13 100 38 23 31 13 0.33 15 W12 150 66 36 49 25 0.64 25 W9 200 93 37 50 24 0,61 30 W10 225 107 52 71 37 0.94 80 WI5 300 149 71 96 51 1.30 100 W14 325 163 60 81 50 1.27 100 W16 350 177 66 89 48 1.22 100 Testing of Specimens from Capsule V 5-15 Table 5-10 Comparison of the Diablo Canyon Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(d' Predicted Measured Predicted Measured (x 1019 n/cm2,, (OF) (a) (OF) Wb m(a) (, .)(C)E > 1.0 MeV)Inter. Shell Plate S 0.284 36.2 -1.78 14 0 B4106-3 Y 1.05 56.0 48.66 19 6.8 (Longitudinal)

V 1.37 60.0 34.32 20 0 Weld Metal S 0.284 145.8 110.79 25.5 11.0 (heat # 27204) Y 1.05 225.4 232.59 34.5 34.1 V 1.37 241.6 201.07 36.5 27.5 I-AZ Metal S 0.284 --72.31 --8.1 Y 1.05 3,.79.77 19.9 V 1.37 -- 110.9 14.7 Correlation Monitor S 0.284 73.01 65.62 2.4 Material Y 1.05 112.9 115.79 8.9 V 1.37 121.0 116.61 4.9 Notes: (a) Based on 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 4.1 (See Appendix C)(c) Values are based on the definition of upper shelf energy given in ASTM El 85-82.(d) The fluence values presented here are the calculated fluence values, not the best estimate.

For best estimate values see Section 6 of this report.Testing of Specimens from Capsule V 5-20 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Diablo Canyon Unit I Reactor Vessel Weld Metal Testing of Specimens from Capsule V 5-21 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 4J Hyperbolic Tangent Curve Printed at 12239 on 09-19-2m92 Results Curve Fluence USE d-USE T o LE35 d-T o LI35 1 2 3 4 0, 2,84E+18 1.0~5E+19=27+19 6817 7391 6124 N.47 0-1425-26.93-33.7-46.52 9528 19425 220S6 0 14 1B 240.7?26719 U)F4-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F SCurve Legend ID 20---- ----- 30 4 Data Set(s) Plotted Material Curve Plant Caisule Ofl. Ileal I Curve Plant (ansule Material ori Heatl 4 2 3 4 DCI DCI DIO UNIRR S Y V WELD LINDE 1092 WEL LIDNDE 10M hEMD LINDE WELD LINDE 102 Z714 FLUX LOT 3714 Z7204 FLUX LOT 3714 Z2204 FLUX WDT 3714 2=4 FLUX LWT 3714 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Diablo Canyon Unit I Reactor Vessel Weld Metal Testing of Specimens from Capsule V 5-22 SURVEILLANCE PROGRAM WELD METAL CVCRAPII 4.1 Hyperbolic Tangent Curve Printed at 12-44:5 on 08-19-26)2 Results Curve Fluence T o 50/ Shear d-T o 5&t. Shear 1 2 3 4 0 2B4E+18 1.05E+19 1.37E+19-1593 110.74 168.75 2D].5 0 12.67 184.68 217.5;-q 4-)-3w0 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 20 ----------

30" 4 I ný-Data Set(s) Plotted Material Curve Plant Cansule Orn Heat#-i ,I!2 3 4 DC1 DO1 DC1 DC1 UNIRR S Y V WELD WELD WELD WELD LINDE 1092 LINDE 1092 LINDE 109 LINDE 1092 Z7204 FLUX Wr 3714 27204 FLUX LOT 3714 2:Z04 FLUX LOT 3714 27204 FLUX LOT 3714 Figure 5-6 Charpy V-Notch Percent Shear-vs Temperature for Diablo Canyon Unit I Reactor Vessel Weld Metal Testing of Specimens from Capsule V APPENDIX C SURVEILLANCE CAPSULE DATA FOR PLATE HEAT NO. C-1279 Report No. 100 1026.40 1, Rev. I C-1 R Structural Integrity Associates, Inc.

BCL-585-12 ,FINAL REPORT on, PALISADES NUCLEAR PLANT REACTOR PRESSURE VESSEL SURVEILLANCE PROGRAM: CAPSULE A-240 to CONSUMERS POWER COMPANY March 13, 1979 by J. S. Perrin, E. 0. Froimm, D. R. Farmelo, R. S. Denning, and R. G. Jung BATTELLE Columbus Laboratories 505 King Avenue Columbus, Ohio 43201 FINAL REPORT on PALISADES NUCLEAR PLANT REACTOR PRESSURE VESSEL SURVEILLANCE PROGRAM;-'CAPSULE A-240 to CONSUIERS POWER COMPANY from BATTELLE Columbus Laboratories-March 13, 1979

SUMMARY

Capsule A-240 was removed from the Palisades Nuclear Power Plant after 2.26 equivalent full power years of reactor operation, The, capsule was: sent to the Battelle Columbus Hot Laboratory for examination and evaluation.

The irradiation temperature did not exceed 536 F as indicated by the examination of the 12 thermal monitors.

The neutron fluence at the location of the specimens was, determined to be 4.4 x 10-9 nvt (E>P MeV), using neutron dosimeters from within the capsule. At the vessel wall the maximum exposure.

was detei-mined to be 3.2 x 109 n/cm2 at 32 full power years.The radiation-induced changes in the mechanical properties of pressure vessel material specimens were determined.

Charpy impact specimens were used to determine changes in the impact behavior, including the shifts in the transition temperature region and the drops in the upper shelf'energy level. Evaluation of the. tensile property specimens included the yield and.ultimate strengths as well as .elongation and reduction in area.

34 Charpy Impact Properties This section Contains results and discussion pertaining to the Charpy impact testing. Appendix A contains further results anddidscussioh relating to the instrumented procedures used during the impact testing.The impact properties determined asia function of temperature are listed in Tables 9 through 12. In addition to the impact energy values, the tables also list .the measured values, of lateral expansion And the'estimated.

fracture appearance for each specimen.

The lateral expansion is -.a measure ,of -the deformation produced by the striking edge of the impact machine hammer when it impacts the specimen; it is the change in specimen.

thickness directly adjacent to the notch location.

The fracture appearance is a visual estimate of the:amount of shear or ductile type of fracture appearing on the specimen fracture surface.The-impact data are graphically shown in ,FiguKes l:through 14.These figures show the change in impact pr6pertiesý as a function of tempgra-ture, including both the impact energy and the lateral expansion.

Figures 1.5 through 18 show the fracture surfaces of the.Charpy specimens.

Table 13 summarizes the Palisades 30 and 50 :ft-lb transition tempera-ture, the 35 mils lateral expansion temperature, and the upper shelf energy for the present program and for the earlier unirradiated program. As indicated.

previously in the neutron dosimetry section, the Charpy specimens, received a fairly uniform exposure.,.

The neutron exposures based on the iron dosimeters r anged from 4.3 to 4.6 x 19 n/cm2 (>l1MeV)..

Pa'ticular exposure values can be assigned to, each of the-.f our Charpy materials Since specimens of a particU-lar material were all located in a given Charpy compartment.

Using a conserva-tive approach, the four Charpy materials from Capsule A-240 received exposures estimated as follows: Base. longitudinal, 4.5 x 109 n/cm2 (>1 MeV)Base transverse, 4.4-x 109 -n/hml ( 1 MeV)Weld,4.6 x10 19 nlcm 2 (>i MeV)19 2 Weld, 4.6 x 10 n/cm (>1 mqV)HAZ, 4.3 x 1019 n/cm 2 (>1 14eV)The impact properties ofthe Palisades balse metal, weld metal, and HAZ metal are all significantly affected by irradiation, as can ýbe seen in the figures of impact energy and lateral expansion versus temperature (Figures 11 through 14).

\ 37.100 0 0 Unirradiated Baseline M Irradiated Capsule A7240'E I, ,6 C-60 Q 0.x-LU 0 040'6.A -20'140 0 .120./-r100-6.0: 80- : E 60: ...4 0 -040' l,OO-200 -100 0 loo 200 300 400 500 Temperature, F FIGURE 11. CHARPY IMPACT PROPERTIES FOR BASE METAL, PLATE NO. D3803-1, LONGITUDINAL oRIENTATION 38:0 CLýx wjýcu-200 -100 0 100 200 300 400 500 Temperature, F FIGURE 12.CHARPY IMPACT PROPERTIES FOR BASE METAL, PLATE NO. D3803-1, TRANSVERSE ORIENTATION

&L^45 TABLE 13.

SUMMARY

OF CHARPY IMPACT PROPERTIES FOR PALISADES Fluence. Upper 35-Miu-E>lMeV, .30 ft-lb 50 ft-lb Shelf Lateral 19n 2 TranSition Trans2tion Energy, Expansion Material 'Program x 10 n/cm Temp, F Temp, F ft-lb Temp, F Base L (a Ref 10 0 0 +20 165 +5 Base Present 4.5 +205. +25 95 +220 ,Base T(b) Refi10 0 25 +55- 105 +40 Base T (b). Present 4.4 +230 +270 68 +235 Weld Ref 10 85 -50 120 -85 Weld, Present 4ýh. +265 +305 54 +285 HAZ Ref 10 0 65 125 -55 HAZ Present 4.3 +200 +240 60 +205 (a) Base metal, longitudinal orientation.(b) Base metal, transverse orientation.

For the four materials in Capsule A-240-,the

50. ft-lb and 30 ft-lb transition temperatures range from 240 F to 305 F-and 200 F to 265 FT, respectively.

The 35,-mil lateral expansion temperature ranges from 205 F to 2.85 F, and the'upper shelf energy levels range from 54 to6-95 ft-lb. Theupper shelf energy levels were taken as- being the highest point- of the curve drawn through the points.

Act ckg-_46 Table 14 is a comparison of the 50 ft!lb and 30 ft-lb transition temperature shifts and the 35-miu lateral expansion temperature shift due to irradiation for the present program. The 50 ft-lb transition temperature shift is defined as the increase in the irradiated 50 ft-lb temperature with respect to the unirradiated 50 ft -lb temperature.

The 30 ft-lb transition templerature shift and the 35-mil lateral expansion temperature Shift are similarly defined. As can be seen, the greatest shift occurs. for the weld material in all three cases.TABLE 14. 50 FT-LB', 30 FT-LB, ANT) 35-MIL LATERAL EXPANSION TEMPERATURE SHIFTS DUE1 QO IRRADIATION FOR "PALISADES CAPSULE A-240 35-Mll Fluence, .30 'ft-lb 50 ft-lb Lateral E>lIMeV, Transition TranSition Expansion Temperature Temperature, Temperature Mat-erial X 109 n/cm Shift, F Shift, F Shift ' "F Base L(a) 4.5 +205 230 215 Base T(b) 4i4 +205 2i5 195 Weld 4.6 +350 355 37.0 HAZ 4.3 +290 '305 260 (a) Base longitudinal orientation.(b) Base transverse orientation.

in comparing the 50 ft-lb shift to the 35-mil lateral 'expansion

ýshift for each of the four materials, note thatthe weld metal 35-mil lateral expansion shift is greater than the weld metal. 50. ft-lb shift, but*the reverse is true for the other three materials.

'In considering the Charpy results., it should be realized that this surveillance capsule is an accelerated one, and has a veryhigh lead factor of 19.4. This means the flux the specimens in the-capsule receive is much higher than any point in the vessel wall. Further Palisades, capsules to be examined include ones with significantly lower lead factors, more closely approximately the vessel wall. The actual location of the various surveilance.capsules inside the pressure vessel are given, in Figure C-1 of Appendix C.

AO fv,,,SL 47 The reference temperature, RTD was determined previously for the unirradiated base transverse materiaJ to be 0 F(23). The procedure for the determination of the RTNT is defined by the ASME Boiler and Pressure Vessel Code(2 4). Apppndix.H, "Reactor Vessel Material Surveillance Program Requirementsý", to 10CFR5O specifies how an adjusted reference temperature for (25 irradiated specimens can be determined.

This temperature can be used in. revising the plant pressure-temperature operating curves in: those cases where the -fluence of the irradiated specimens is in the range to be exper-i-.enced by the pressure vessel. The adjusted reference temperature defined by Appendix H is determined by adding to the reference temperature the; amount of the temperature shift in the :Charpy curves between the unirradiated material and the irradiated'material, measured at the '50 ft'lb level, or 'that measured at the 35-mil lateral expansion level, whichever temperature shift is greater.Tensile Properties The tensile properties determined for 'the tensile specimen contained in the Palisades capsule A-240 are listed in Table 15. The table. lists test temperaturefluence, 0.2 percent offset yield strength, ultimate tensile strength, uniform elongation, total elongation,, and reduction in area for the present program as well as for the unirradiated b'aselrine program. Post-test photographs of the tensile specimens are shown in Figure 19. These photographs show the necked down region of the gage length and the fracture.

A typical tensile test curve is shown in Figure 20' the particular test shown is'for base'metal specimen 1D4 tested at 72 F.Tensile tests were run at.room temperature (69 to 73 F), 535 and'565 F.The higher temperature tests exhibited a decrease in 0.2 percent !offset yield strength and a decrease in ultimate tensile strength for each material with respect to the room temperature tests. In general, ductility values (as determined by total elongation and reduction in area) decreased at higher temperatures as compared to room temperature for each material.Palisades tensile specimens were located in the vicinity, of iron...19 2 dosimeters which received fluences rangingvfrom 4.3. to 4.6 x 10 n/cm (E>l MeV). When the tensile data for Capsule A-240 is compared to the unirradiated baseline data (Table 5)'it can be seen that as fluence increases, the yield strength and tensile strength increase while ductility decreases.

WCAP-10637 WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSULES, T-330 AND W-290 FROM THE, CONSUMERS POWER COMPANY PALISADES REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM M. K,. Kunka C.A. Cheney September i984.Work performed under Shop Order Nos. ENVJ-106 and ENVJ-450* -7 ,..K 1 APPROVED:

1/7 Ti A. Meyer, Manager Structlural Materials ,and Reliabil ty Technology Prepared by Westinghouse for the.Consumers Power Company Although information contained -in this report is nonproprietary,ý no distribution shall be made outside Westinghouse or Its licensees.

w'ithout the customer's approval;ýWESTINGHOUSE ELECTRIC CORPORATION Nuicleear Energy Systems P.:O. Box. 355 Pittsburgh, Pennsylvania 15230 Wa 4 -P M 6,!; ýSECTION 1

SUMMARY

OF RESULTS The analysis of the ma1terial contained in Capsule, T-330, the 'first thermal surveillance capsule removed from the Consumers Power Compahy'ýs Palisades.reactor presýsure:

vessel, led to the following conclusions:

!o The weld and heat-affected zone metal has experienced a 60-70F. shift in the ductile to brittl~e transition

temperatures due to exposure to-elevated temperature.

The analysis of the material contained in Capsule W-290, the second jrradi.ated surveillance capsule to be removed from the Consumers, Power Company Palisades reactor pressure vessel, led to the following conc.lusions:

o The capsule received an average fast neutron fluenc ,(E>I..OMev) of 1.09 x 1019 n/cm 2.o Irradiation of the reactor vessel intermediate shell course plate D-3803-1, to -1.09:x 1019 n/cm, resulted in 30 and 50 ft-lb transition temperature increases of 155 and 160'F, respectively, for specimens oriented perpendicular to the principal rolling direction (trahsverse orientation),, and 175 0 TF and 180 0 F, respecti.vely, for specimens oriented paralleI to the principal rolling direct.ion (longitudiha'l orientation).

o Weld metal irradiated to 1..09 x I0 1 9 n/cm 2 resulted in 30 and 50 ft-lb transition temperature increase of 290.and 300'F, respectively.

o The average upper shelf energy of all the surveillance materials remained above 50 ft-lbsý, thereby providing adequate toughness for continued safe plant operation.

8092B:lb-092684 1-1 o Comparison of the 30 ft-lb transition temperature increases for the Palisades surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1, shows that the weld metal transition temperature increase was greater than predicted., It is suspected that the relatively high nicke~l content of the weld metal contributed to the greater than predicted.

transition temperature increase experienced by the weld metal.8092B~lb-092684 2 these test results it appears that a mixup in monitors occurred during the initial loading of the capsules and therefore a reliable estimate of the capsule temperature cannot be determined from the thermal monitors.5-3. CHEMICAL.ANALYSIS Chemical analyses were performed on fractured Charpy V-notch. spPecimens in order to confi'rm the chemical cbmposition of the !surveillance plate aid weld materials.

The chemical analysis results are summari~zed in Table 5-1. The most notable feature of these analyses is the. greatv ariance measured in the.nickel content, specifically from .95 to 1.60 wt. %. From the high nickel content, it is evident that. a Nickel-200 addition was made% to the, survei.l1ance weldment, and from the nickel. Variances ob.served it can be concluded that the rate of Nickel-200 addition was varied during welding.5-4. CHARPY V-NOTCH IMPACT TEST RESULTS Capsul-e T-,330: The results of the Charpy V-notc.h impact tests performed on the various materials cdntained in Capsule T-330, the thermal capsule, are presented in Tables 5-2 through 5-9 and Figures 5-1 through 5-4.. From the ,Charpy V-notch, plots based, on best engineering judgement it appear's that the weld and hea.t-affected zone metals have experienced a 60 to 70'F shift in the ductile to brittle transition temperatures due to exposure.

to elevated temperature;, but no decrease in upper shel.f .energy.The. fracture appearance of each Charpy specimen from the various materials is shown in Figures 5-5-through 51-8, and show an increasing ductile or-tougher appearance with increasing test temperature..

A typical instrumented Charpy curve, representing the curves of both, Capsule T-330 and-Capsule W-290, is presented in Figure 5-9.8092B:lb-102984 5-4 Capsule W-2-90: The results of the Charpy V-notch impact tests performed on the various material.s contained in Capsule W-290, irradiated at 1.,09 x 101 n/cm 2 , are presented in Tables 5-10 through 5-17 and Figures 5-10 through 5-13. A summary of the transition temperature increases and upper shelf energy decreases for th CGapsule W-29Q mateia.l is shown in Table 5-18.Irradiation Pf t-he vessel intermediate shell course plate D-3803-1 (transverse 19 2.orientation) to 1.09 .X 10 n/cm (Figure 5-10) resulted in 30 and 50 ft-lb transition temperature increases of 155 and 1606F, respectively, and an upper shelf energy decrease of 18 ft-lb. Irradiation of the vessel inte'mediate she.l plate material (longitudinal orientation) to 1.:09x 1019 nhcm 2'(Figure '5-11) resulted in 30 and 50. 'ft-l.b transition.

temperature of 175 and 180 0 F, respectively, and an Upper shelf energy ýdecrease of 43 f-t-lb.Weld metal irradiated to 1.,09 x 10!9 n/cm 2(Figure 5 -12), re.sulted in 30 and 50 ft-lb transition tempera'ture increases of-290:and 300 0 F., respectively, and an upper Shelf energy decrease of 54 ft-lb..Weld HAZ metal irradiated to 1.09 x 10I9 n/cm 2 (Figure 5-13) resulted ih 30 and 50 ft-lb transition temperature increases of 235 and 245 0 F, respectively, and an upper shelf energy decrease of 44 ft-lb.The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-14 through 5-17 and show an increasing ductile or tougher appearance, with increasing test temperature.

Figure. 5-18 shows a comparison of the 30 frt-lb transi~tion temperature increases for the Various Palisades surveillance materials with predicted increases using the. methods of NRC Regulatory Guide 1.99, Revision .1.[3 8092B:lb-102984 5-5 The regulatory curves used for comparison were developed from the average copper and phosphorus contents (averages of the analyses presented in Tables 4-1 and 5-1) of plate D-3803-1 and the weld metal. This comparison shows that the plate transition temperature increases resulting from irradiation to 1.09 x 1019 n/cm 2 are less than predicted by the Guide for plate D-3803-1.

The weld metal transition temperature increase resulting from 1.09 x 10 1 9 n/cm2 is greater than predicted by the Guide. This can be explained by the high nickel content of the weld metal. It is widely recognized today that nickel has a profound effect upon the irradiation damage of reactor vessel materials, whereas the current revision of Regulatory Guide 1.99 does not incorporate this important variable.5-5. TENSION TEST RESULTS Capsule T-330: The results of the thermil capsule tension tests performed on plate D-3803-1 (longitudinal orientation) and weld metal are shown in Table 5-19 and Figures 5-19 and 5-20, respectively.

These results show that the thermal environment produced little change in the 0.2 percent yield strength of the plate and weld material.

Fractured tension specimens for each of the materials are shown in Figures 5-22 through 5-24. A typical stress-strain curve for the tension specimens, representing the curves of both Capsule T-330 and Capsule W-290, is shown in Figure 5-25.Capsule W-290: The results of the irradiated capsule tension tests performed on plate D-3803-1 (longitudinal orientation) and weld metal irradiated to 1.09 x 1019 n/cm 2 are show in Table 5-20 and Figures 5-26 and 5-27, respectively.

These results show that irradiation produced an increase in the 0.2 percent yield strength of approximately 20 ksi for plate D-3803-1 and of approximately 30 ksi for the weld metal. Fractured tension specimens for each of the materials are shown in Figures 5-29 through 5-31.8092B:lb-102984 5-6 fco%0 co TABLE 5-18 EFFECT OF IRRADIATION AT 1.09 x 1019 (E > 1 MeY)ON THE NOTCH TOUGHNESS PROPERTIES.

OF THE PALISADES SURVEILLANCE-VESSEL MATERIALS Average 30 ft-Ib'Temp (F)Material Unirradiated Irradiated Average 35 mil Average Lateral Expansion Temp (*F) 50"ft-lb:Temp

(-F)AT Unirradiated Irradiated AT Unirradiated Irradiated Average.Energy Absorption at Full Shear (ft-lb)Unirradiated Irradiated AT.0b Plate D73803-1 (Transverse)

Plate D-3803-1 (Longitudinal) 125 180-155 25 175' -5 195, 170 .55 190 185 20 215 160 102 ,200 180 -155 84.112 A (ft-lb)18 43 0 175)Weld-Metal

-85 HAZMetal -90 205 290 -75 240 315 -50;250 300 -118 180 245 116 64 54 145! 235 -55 160 215 -.65 72 44 0oo ,a%VC PC aa 02 50 0 100 U U 0.M-I I"'Sm's0 0 to0*150-0---200 Figure 5-10.-100 0 100 200 300 400?ZMPIEAOoU R IF)Irradiated Capsule'Charpy V-NotCh Impact Properties for Palisades Intermediate Shell Plate D-3803-1 (Transverse Orientation) 80928:lb-091984 5-36 goo 44 0'ego ,50 1.0 150 U 0-2I Figure 5-11 a.1 Fý1.1650)0 pit,-100 0 NO0 to0 300 400 YIIPIIAUII ( F)Irradiated Capsule Charpy V-Notch Impact Properties for Palisades Intermediate Shell Plate D-3803-1 (Longitudinal Orientation) 8092B:lb-092684 5-37 I WC AP- 140 1'4 WESTINGHOUSE CLASS 3 (Non-Proprletary)

A'NALYSIS OF CAPSULE W- 110 FROM THE.CONSUMERS POWER COMPANY P.,ALISADES REACTOR VESSEL.RA-kDIATION SURVEILLA-NCE PROGRA.M CONTROLLED COpy ERC SECTION. 1.0

SUMMARY

OF RESULTS The analysis of the reactor vessel. materials contained in surveillance Capsule W-I 10, the second vessel wall capsule assembly to be removed from the Consumers Power Company Palisades'reactor pressure vessel, led to the following conclusions:

o The capsue received an averageI fast neutron fluence of 1 .779'k 10'9 n/cm 2 (E > 1.0 MeV)after 9295 EFPY of plant operation.

o Irradiation of the reactor vessel intermediate,shell plate D-3803-.1 Charpy specimens, oriented with the longitudinal -axis of the specimen parallel tO the.major rolling direction (lorngitudinal orientation), to 1.779 x 10'9 t/cm- (E> 10 MeV) resulted in a 30 ft-lb transition temperature increase of 180'F and a 50 ft-lb transition temperature increase of M90OF. This results in an irradiated 30 ft-lb transition temperature of 180F :.and an irradiated 50 ftMib transition temperature of 210F ,for longitudinally Oriented specimens.

0 of the surveillance weld metal Charpy specimens to 13779x x0'9 ri/cm 2 (E > .-MeV) resulted in a 30ft4b transition temperature increase of 314 0 F and a 50 ft-lb transition temperature increase of 355 0 F. This results in an irradiated,30 ft-b transition-temperature of 229 0 F and an irradiated 50: ft-lb transition temperature of. 305 OF for the weld metal.Irradiation of the reactor vessel weld Heat-Affected-Zone H )metalCharpy.

specimens to 1.779 x 10'" n/cm 2' (E,> 1.0 MeV) resulted in a 30 ft-lb transition temperaure increase of 240°F and a 50 ft4b.transition temperature increase of:275 0 F° This.restultsin an irradiated 30.ft-lb transition temperature of 150,F and. an irradiated

.5.0 ft-lb transition temperature of 210 0 F for the weld HAZ metal.0 Irradiation of the ieactor.vessel Correlation Monitor Standard Reference MateriAl (SRM)metal Charpy specimens to 1.779.x .109 n/cmuz (E >. 1.0 MeV) resulted in a 30 ft-lb. transition.

temperature increase of 148OF and a 50 ft-lb transition temperaturelincrease of 158 0 F, This results in an irradiated 30 ft-lb transition temperature of 163°F and an-irradiated 50 ft-lb transition.

temperature of 203IF for the weld HAZ metal.14I o Irradiation of intermediate shell plate D-3803-1 (longitudinal orientation) to 1..779 x 10'9 n/cm 2 (E > 1.0 MeV) resulted in an irradiated average upper shelf energy decrease of 52 ft-lbs, resulting in an irradiated upper shelf energy of 103 ft-lbs.o The average upper shelf energy of the weld metal decreased 56 ft-lb after irradiation to 1,779 x i019 n/cm 2 (E > 1.0 MeV). This results in an irradiated upper shelf energy of 62 ft-lb for the weld metal specimens.

0 The average upper shelf energy of the weld HAZ metal decreased 35 ft-lb after irradiation to 1.779 x 10'9 n/cm 2 (E > 1.0 MeV). This results in an irradiated upper shelf energy of 81 ft-lb for the weld HAZ metal.0 Irradiation of SRM metal to 1.779 x 10'9 n/cm 2 (E > 1.0 MeV) resulted in an irradiated average upper shelf energy decrease of 34 ft-lbs, resulting in an irradiated upper shelf energy of 99 ft-lbs.o The surveillance Capsule W- 110 test results indicate that the 30 ft-lb transition temperature shift of the surveillance materials is in good agreement with Regulatory Guide 1.99, Revision 2 predictions and that the upper shelf energy decrease of the surveillance materials, except for the weld metal, is less than the Regulatory Guide L99, Revision 2 predictions (Table 5-8).o Per Reference 6, the Surveillance Capsule Removal Schedule will not be generated as part of this analysis.1-2 The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-93t"'Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.

Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip., In the test configuration, with a slight load-on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range of room temperature to 550°F (2881C). The upper grip was used to control the furnace temperature.

During the actual testing the grip temperatures were used to obtain desired specimen temperatures.

Experiments indicated that this method is accurate to S2o1i 4 4lJ.The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs.

The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

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 W- 110, which was irradiated to 1.779 x 1019 n/cm 2 (E > 1.0 MeV), are presented'in Tables 5-1 through 5-6 and are compared with unirradiated resultst 2) as shown in Figures 5-2 through 5-5.The transition temperature increases and upper shelf energy decreases for the Capsule W-1 10 materials are summarized in Table 5-7.Irradiation of the reactor vessel intermediate shell plate D-3803-1 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 1.779 x i0 V n/cm- (E > 1.0 MeV) (Figure 5-1) resulted in a 30 ft-lb transition tempeiature increase of 1801F and in a 50 ft-lb transition temperature increase of 190 0 F. This results in an irradiated 30 ft-lb transition temperature of 180IF and an irradiated 50 ft-lb transition temperature of 210°F (longitudinal orientation).

5-3 The average Upper Shelf Energy (USE) of the intermediate shell plate D-3803-1 Charpy specimens (longitudinal orientation ) resulted in an energy decrease of 52 ft-lb after irradiation to 1.779 x 10'9 n/cm 2 (E > 1.0 MeV), This results in an irradiated average USE of 103 ft-lb (Figure 5-2)Irradiation of the surveillance weld metal Charpy specimens to 1.779 x 10'9 n/cm 2 (E > 1[0 MeV)(Figure 5-3) resulted in a 3141F increase in 30 ft-lb transition temperature and a 50 ft-lb transition temperature increase of 355°F. This results in an irradiated 30 ft-lb transition temperature of 229GF and an irradiated 50 ft-lb transition temperature of 305 0 F.The average USE of the reactor vessel core region weld metal resulted in an energy decrease of 56 ft-lb after irradiation to 1,779 x 10'9 n/cm 2-(E > 1,0 MeV), This results in an irradiated average USE of 62 ft-lb (Figure 5-3).Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 1.779 x 10'9 n/cm2 (E > 10 MeV) (Figure 5-4) resulted in a 30 ft-lb transition temperature increase of 240"F and a 50 ft-lb transition temperature increase of 275 0 F. This results in an irradiated 30 ft-lb transition temperature of 150TF and an irradiated 50 ft-lb transition temperature of 210 0 F.Ihe average USE of the reactor vessel weld HAZ metal experienced an energy decrease of 35 ft-lb after irradiation to 1.779 x l0'9 n/cm'- (E > 1.0 MeV). This results in an irradiated average USE of 81 ft-lb (Figure 5-4).Irradiation of the reactor vessel Correlation Monitor Standard Reference Material (SRM) specimens to 1.779 x 10'9 n/cm 2 a (E > 1.0 MeV) (Figure 5-5) resulted in a 30 ft-lb transition temperature increase of 148°F and a 50 ftlb transition temperature increase of 158TF. This results in an irradiated 30 ft-lb transition temperature of 163°F and an irradiated 50 ft-lb transition temperature of 203 0 F.The average USE of the reactor vessel Correlation Monitor Standard Reference Material (SRM)experienced an energy decrease of 34 ft-lb after irradiation to 1.779 x 10'9 nfcm 2 (E > 1.0 MeV). This results in an irradiated average USE of 99 ft-lb (Figure 5-5).The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-6 through 5-9' and show an increasingly ductile or tougher appearance with increasing test temperature.

5-4 A comparison Of the 30 ft-lb transition temperature increases, and upper shelf energy. decreases for the various Palisades surveillance materials, with. predicted values using the methods of NRC Rcgulatory Guide 1 .99. Revision 21"1 is presented in Table 5-8. This comparison indicates that the 30 -ft-lb'transition temperature shift Of the surveillance materials lis in good agreement'with, the Regulatory Guide 1.99, Revision 2 predictions and the USE decrease of the surveillance materials, -except the weld metal, is less than the Regulatory Guide 1.99, Revision 2 predictions.

The load-.time records for the individual instrumented Charpy specimen tests are shown in Appendix A..53 Tension Test Results The results of the tension tests performed

on the -various materials contained in:Capsule W;. 110 irradiated to 1.779 x .019 n/cm 2 (E > _0 MWeV) are presented in Table 5-9 and are compared with unirradiated results" 2) as shown .in Figures 10 through 5-12.The results of the tension'tests performed on the intermediate shell plate D-3803-1 '(longitudinal orientation) indicated that irradiation to 1.779 x 1019 n/cm 2 ,(E > 1.0 MeV) caused an .18 to 23 ksi increase ini the 0.2 percent-offset yield strength and a 12 to 15 ksi increase in the ultimate tensile strength when,compared to unirradiated dataM 2 1 (Figure 5- i0).The results of the tension tests performed on the surveillance weld-metal indicated that.irradiafionto 1.779,x 10'9 n/cm 2-(E > 1.0 MeV caused a 17 to.35 ksi increase in the 0o2 percent offset yield strength:

and a 17 to 27 ksi increase in the ultimate tenile strength when compared to unirradiated data0 2 1 (Figure 5-11).The results of the tensionitests-performed on the heat-affected

zone (HAZ).metal indicated that iradiation to 1.779 x 10'.!n/cim 2 (EM> [.0 MeV) caused a 12 to 20 ksi increase in the 0.2 percent.

offset yield strength and a 15 to !17 ksi increase in the ultimate tensile strength when compared to unirradiated data'2 (Figure 5-12).The fractured tension specimens for the surveillance materials are shown in Figures 5-13 through 5-15.The engineering stress-strain curves for the tensile tests are shown.in Figures 5-16 through 5-21.5-5 TABLE 5-7 Effect of Irradiation to 1.779 x 10"' n/cm 2 (E > 1.0 MeV)on the Notch Toughness Properties of the Palisades Reactor Vessel Surveillance Materials Average 30 ft-lb (a) Average 35 roil (a) Average 50 ft-lb01 Average Energy Absorption Material Transition Temperature

(*F) Lateral Expansion Temperature

(*F) Transition Temperature

("1i) at Full Shear (ft-lb)Unirradiated Irradiated AT Unirradlated Irradiated AT Unirradialed Irradiated AT Unirradiated Irradiated AE Plate D-3803-1 0 180 180 5 200 195 20 210 190 155 103 -52 Weld Metal -85 229 314 -75 280 355 -50 305 355 118 62 -56 HAZ Metal -90 150 240 -55 187 242 -65 210 275 116 81 -35 SRM O0MY 15 163-148 25 192 167 45 203 158 133 99-34 U I I J ___________________

.8- -4. --8 I .1- S (a) "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5- 1 through 5-4).

TABLE 5-8 SComparison Of the Palisades Surveillance

Material

.30, ft-lb TransitionTemperatbre Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99. Revision 2 Predictions 30 ft-lb Transition Temperatu'e Shift Upper, Shelf Energy Decrease~Jm Fluence Iti Ca' 0nc Predicted (a), Measured:

Predicted (a) Measured Material, Capsule E>10Oe 9 Itk (%)"E> .0MeV) Oe Plate D-3803-t A-240 6.0' 223 205 48 35 (Transverse)

W-290 1.09 159 155 33 18 W-110 1.779 180 -- 37 Plate D-3803- 1 Aý-240 6.0* 223 205 48 42.(Longitudinal)

W-290 1.09 159 175 33 28 W-1 10 1.779 180 180. 37 34 Weld Metal A-240 6.r0 423 350 56. 55 W-290 1.09 302 290 40 46',W-llOr 1.779 341 314 ý44 47 HAZ Metal A-240 6.0 -- 290" 52!W-290 1.09 -- 235 '38.W-1.10 1.779 -- .240 -- 30.SRM W-10 -l.779 -158 148 31 26&(a) Based on Regulatory Guide 41.99, Revision 2 methodology..

• Seekeference

42.

(0 C)-150 -100 -50 0 50 100 150 200 250 100 so 60~40 20 0 100~80 c" 60-40 S20 200 180 160 140.120"100" 80' 60 40 20 0 2.5 2,0 1,5 1.0 0.5 0 E 240 200 160 120 80 40 0-200 -100 0 100 200 TEMPERATURE (OF)300 400 500 0 MAMM)SNAMED M A W IF L779 x 1019(/Vj 2 (E) 1U HeV)Charpy V-Notch impact Properties for Palisades Reactor Vessel Intermediate Shell Plate D-3803-1 (longitudinal orientation)

Figure 5-2 5.16 pT cr SWXT Services, Inc.ANALYSIS OF CAPSULE W-100 FROM THE NUCLEAR MANAGEMENT COMPANY PALISADES-' VESSEL MATERIAL SURVEILLANCE PROGRAM FEBRUARY 2004 1295-001-03-08:00 FEBRUARY 2004 12500-3-80 FEBRUARY 2004111 BWXT Services, Inc.Analysis of Palisades Capsule W-100 4 4.0 DESCRIPTION OF THE PALISADES REACTOR VESSEL SURVEILLANCE PROGRAM Prior to initial plant start-up, ten surveillance capsules were inserted into the Palisades reactor vessel near the reactor vessel wall as shown in Figure 4-1. The capsules contain specimens made from intermediate shell plate D-3803-1, heat-affected-zone (HAZ) metal fabricated by welding intermediate shell plates D-3803-2 and D-3803-3 with submerged arc process using Linde 1092 flux, and weld metal fabricated by welding intermediate shell plates D-3803-1 and D-3803-2 with submerged arc process using Linde 1092 flux and a MIL-B4 electrode and a 1/16-inch diameter Nickel-200 wire feed. Capsule W-100 was removed after 16.93 effective full power years (EFPY) of plant operation.

This capsule contained Charpy impact and tensile specimens made of intermediate shell plate D-3803-1, submerged arc weld metal, and HAZ metal as describe above. All test, specimens were machined from material taken at least one plate thickness from any water quenched edge.The surveillance plate material was cut directly from the intermediate, shell course plate after being subjected to 1.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> of interstage and 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of final heat treatment at 1150 t 25 0 F. Charpy impact specimens from surveillance plate D-3803-1 were machined in the longitudinal orientation (longitudinal axis of the specimen longitudinal to the major working direction).

The weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular to the weld direction.

The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction.

Tensile specimens from surveillance plate D-3803-1 were machined with the major axis both in the tangential and longitudinal orientations.

Tensile specimens from the weld metal were oriented with the long dimension of the specimen perpendicular to the weld direction.

The chemical compositions of the surveillance materials are presented in Table 4-1. The chemical analysis reported in Table 4-1 was obtained from unirradiated material used in the surveillance program[3].Capsule W-100 contained dosimeter wires of uranium, sulfur, iron, nickel, titanium, and copper.Cadmium covers were used for materials that have competing thermal activities (i.e., uranium, nickel, and copper). Dosimeters are used to determine flux spectrum and flux attenuation through the thickness of the Charpy specimens.

-- The temperature monitor assemblies consist of four separate coil-shaped monitors, each of different composition and thus having different melting points. They are identified by varying capsule lengths, with melting temperatures increasing with increasing capsule length. The alloys compositions and-- melting points are listed as follows.Composition Melting Point 92.5% Pb, 5.0% Sn, 2.5% Ag 536 0 F 90.0% Pb, 5.0% Sn, 5.0% Ag 558TF 97.5% Pb, 2.5% Ag 580OF 97.5% Pb, 0.75% Sn, 1.75% Ag 590OF The arrangement of the various mechanical specimens, dosimeters, and thermal monitors contained in* Capsule W-100 is shown in Figure 4-2.

13WXT Services, Inc.Analysis of Palisades Capsule W-100 5 TABLE 4-1 Chemical Composition (wt%) of the Palisades Reactor Vessel Surveillance Materials[

3 J Weld Weld Weld Weld Element D-3803-1 D-3803-2 D-3803-3 D-3803-3/

D-3803-31 D-3803.21 D.3803.2/D-3803-2 D-3803-2 D-3803-2 D-3803-1 Root Face Root Face Si 0.23 0.32 0.24 0.24 0.25 0.25 0.22 S 0.019 0.021 0.020 0.009 0.010 0.010 0.010 P 0.011 0.12 0.010 -0.011 0.012 0.011 0.011 Mn 1.55 1.43 1.56 1.08 1.03 1.01, 1.02 C 0.22 0.23 0.21 0.098 0.080 0.088 0.086 Cr 0.13 0.42 0.13 0.05 0.04 0.05 0.03 Ni 0.53 0.55 0.53 0.43 1.28 0.63 1.27 Mo 0.58 0.58 0.59 0.54 0.53 0.55 0.52 AI(T) 0.037 0.022 0.037 Nil Nil Nil Nil V 0.003 0.003 0.003 Nil Nil Nil Nil Cu 0.25 0.25 0.25 0.25 0.20 0.26 0.22 BWXT Services, Inc.Analysis of Palisades Capsule W-100 12 Table 6-1.Charpy Impact Data for the Palisades W-100 Capsule Plate D-3803-1 Specimen Number Temperature, *F Impact Energy, Lateral Shear Fracture, %ft-lb Expansion, mils Transverse 213 70 9.5 2 0 255 110 14.0 7 5 25E 150 27.5 19 20 25B 200 44.0 32 40 25D 225 52.5 39 50 211 240 50.0 36 , 50 257 250 71.5 54 70 256 260 69.0 47 90 214 270 71.0 54 95 25A 285 77.0 56 100 25C 300 76.5 63 100 212 325 67.5 56 100.... Longitudinal 152 70 5.5 1 0 151 110 14.0 7 5 153 130 29.5 18 25 157 175 44.0 27 40 15A 200 45.0 34 45 154 225 74.4 51 70 1SY 250 86.5 50 85 156 260 73.0 50 80 15C 270 102.0, 57 95 15B 280 100.5 58 100 15U 300 104.5 72 100 155 325 101.0 68 100 C r (~ I I I I I I BWXT Services, Inc. -Analysis of Palisades Capsule W-100 14 Table 6-3.Comnarison of Palisades Surveillance Material (Capsule W-100) 30 ft-lb Transition Temperature Shifts (Position 2.1) and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions Material Unirradiated Capsule Measured Predicted Unirradiated Capsule Predicted USE 30 ft-lb W-100 30 ft-lb 30 ft-lb Temp. USE W-100 (ft-lb)Temp. 30 ft-lb Temp. Shift Shift (ft-lb) USE (OF) Temp. (CF) (OF) (ft-lb)D-3803-1 Tra8vers 18.3 160.8 142.5 189.1 101.6 73.0 61.7 TransverseI D-3803- -0.5 158.6 159.1 189.1 154.8 102.0 93.8 Longitudinal

....Weld Metal- -86.6 218.8 305.4

• 117.7 51.8 63.7 HAZ Metal -89.6 101.4 191.0 ** 115.5 59.7 ***Could not be calculated because RG 1.99 Rev 2 Table 2 does not provide chemistry factors for base metals with Ni content greater than 1.2 wt%.** The RG 1.99 Rev 2 CF tables do not cover HAZ metal Table 6-4.Tensile Properties of the Palisades Capsule W-100 Materials Material Specimen Test 0.2% Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Temp Yield Strength Load Stress Strength Elongation Elongation In Area Strength (Ksi) (Kip) (Ksi) (Ksi) (%) (%) (%)HAZ 4JK 70 92.0 106.3 3.60 175.9 73.4 6.0 N/A ; 58.3 4JJ 200 88.6 101.3 3.47 171.4 70.7 4.9 N/A 58.8 4ME 550 83.1 98.9 3.84 146.8 78.2 4.3 N/A 46.7 D-3803- l31 70 89.8 108.2 3.51 190.6 71.4 10.8 23.8 62.5 1 IEY , 250 83.0 101.7 3.35 169.9 68.3 9.9 21.6 59.8 (Long.) 1E7 550 76.7 98.3 3.53 150.1 71.9 8.8 19.2 52.1 Weld 3DT 70 101.7 115.0 4.52 192.1 92.0 11.7 23.5 52.1 Metal 3DP 300 96.4 109.7 4.59 167.1 93.5 9.4 16.6 44.0 3DM 550 92.9 109.8 4.78 178.6 97.3 7.9 13.6 45.5 NA: Specimen failed outside the gage length.

BVIXT Services, Inc.Analysis of Palisades Capsule W-100 15 Palisades Nuclear Plant -Base (Transverse)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 02/03/2004 10:22 AM Data Set(s) Plotted Curve 1 2 3 4 Plant PALISADES PALISADES PALISADES PALISADES Capsule UNIRR W-290 W-100 A-240 Material SA302BM SA302BM SA302BM SA302BM Ori.TL TL TL TL Heat #C-1279-3 C-1279-3 C- 1279-3 C-1279-3 120 100 0 0 U.wU z>0 80 60 40 20 0 i:::-200-100 0 100 200 300 400 500 o Set1 Temperature in Deg F a Set2 0 Set3 Results Set 4 Curve 2 3 4 Fluence LSE USE 0 2.2 101.6 d-USE T @30 d-T @30 T @50 d-T @50.0 18.3.0 49.1 9.26E18 2.2 83.8 -17.8 2.09E19 2.2, 73.0 -28.6 176.3 160. 8 212. 6 158.0 142.5 194.3 202.5 206.3 265.1.0 153. 4 157. 2 216. 0 4.01E19 2.2 68.4-33.2 Figure 6-1. Charpy Impact Energy vs. Temperature for Palisades Surveillance Plate D-3803-1 (Transverse Orientation) col BWXT Services, Inc.Analysis of Palisades Capsule W-100 16 Palisades Nuclear Plant -Base (Transverse)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 02/03/2004 10:25 AM Data Set(s) Plotted Curve 1 2 3 4 Plant PALISADES PALISADES PALISADES PALISADES Capsule UNIRR W-290 W-100 A-240 Material SA302BM SA302BM SA302BM SA302BM Orn.TL TL TL TL Heat #.C-1279-3 C-1279-3 C-1279-3 C-1279-3 100 80 0 i;C 60 40 20 0-200-100 0 100 -200 300 400 500 o Set 1 Temperature in Deg F Set2 C Set3 Results a Set4 Curve 2 3 4 Fluence LSE USE d-USE T @35.0 23.5 d-T @35.0 0 9. 26E18 2. 09E19 4. 01E19.0 78.6.0 69.3.0 57.3-9.3-21.3 181.7 158.2 205.9 182.4 219.2 195.7.0 66.3 -12.3 Figure 6-2. Lateral Expansion vs. Temperature for Palisades Surveillance Plate D-3803-1 (Transverse Orientation)

BWXT Services, Inc.Analysis of Palisades Capsule W-100 17 Palisades Nuclear Plant -Base (Transverse)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 02/03/2004 10:23 AM Data Set(s) Plotted Curve 1 2 3 4 Plant PALISADES PALISADES PALISADES PALISADES Capsule UNIRR W-290 W-100 A-240 Material SA302BM SA302BM SA302BM SA302BM Ori.TL TL TL TL Heat #C-1279-3 C-1279-3 C-1279-3 C-1279-3 120 100 80 40)~60 40-K 20 0 --200-100 0 100 200 300 400 500 Temperature in Deg F c Set2 0 Set3 o Set I a Set 4 Results Curve-1 2 3 4 Fluence LSE USE 0 .0 100.0 d-USE T @50.0 85.5 9. 26E18 2. 09EI 9 4. 01E19.0 100.0.0 100.0.0 100.0.0 193.6 d-T @50.0 108. 1 133.2 158.0.0.0 218.7 243.5 Figure 6-3. Percent Shear vs. Temperature for Palisades Surveillance Plate D-3803-1 (Transverse Orientation)

BWXT Services, Inc.Analysis of Palisades Capsule W-100 18 Palisades Nuclear Plant -Base (Long.)CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 02/03/2004 09:49 AM Data Set(s) Plotted Curve 1 2 3 4 5 200 175 150 o 125 ,,oo L 7 z 75 Plant PALISADES PALISADES PALISADES PALISADES PALISADES Capsule UNIRR W-290 W-110 W-100 A-240 Material SA302BM SA302BM SA302BM SA302BM SA302BM On.LT LT LT LT LT Heat #C-1279-3 C-1279-3 C-1279-3 C-1279-3 C-1279-3 50-25 00-200-100 0 100 200 300 400 500 Temperature in Deg F 3 Set2 0 Set3 &- Set4 v Set5 o Set1 Results Curve 2 3 4 5 Fluence 0 9. 26E18 1. 66E19 2. 09E19 4. 01E19 Figure 6-4.LSE 2.2 2.2 2. 1 2.2 2.1 USE 154.8 112.3 102.7 102.0 92.3 d-USE.0-42.5-52.1-52.8-62.5 T @30-.5 176.3 179.0 158.6 204.6 d-T @30.0 176. 8 179.5 159. 1 205. 1 T @50 25.3 198.0 203.5 190. 4 243.0 d-T @50.0 172.7 178.2 165. 1 217.7 Charpy Impact Energy vs. Temperature for Palisades Surveillance Plate D-3803-1 (Longitudinal Orientation)

CO,!-

BWXT Services, Inc.Analysis of Palisades Capsule W-100 1(19 Palisades Nuclear Plant -Base (Long.)CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 02/03/2004 09:51 AM Data Set(s) Plotted Curve 1 2 3 4 5 Plant PALISADES PALISADES PALISADES PALISADES PALISADES Capsule UNIRR W-290 W-110 W-100 A-240 Material SA302BM SA302BM SA302BM SA302BM SA302BM Ori.LT LT LT LT LT Heat #C-1279-3 C-1279-3 C-1279-3 C-1279-3 C-1279-3 100 80 E C9 0 60 40 2:1 0--200 0 Set I-100 0 100 200 300 400 Temperature in Deg F 1 Set2 0 Set3 A Set4 500 Set 5 Results Curve 1 Fluence LSE USE d-USE T @35.0 10.0 d-T @35.0 0 2. 9.26E18.0 87.8.0 81.0.0 75.3 3 4 5 1. 66E19 2. 09E19 4. 01E19-6.8-12.5-24.0-7.8 186.9 176.9 190.0 180.0 195.7 185.7 214.4 204.4.0 63.8.0 80.0 Figure 6-5.Lateral Expansion vs. Temperature for Palisades Surveillance Plate D-3803-1 (Longitudinal Orientation)

'--05 BWXT Services, Inc.Analysis of Palisades Capsule W-100 20 Palisades Nuclear Plant -Base (Long.)CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 02/03/2004 09:50 AM Data Set(s) Plotted Curve 1 2 3 4 5 Plant PALISADES PALISADES PALISADES PALISADES PALISADES Capsule UNIRR W-290 W-110 W-100 A-240 Material SA302BM SA302BM SA302BM SA302BM SA302BM Ori.LT LT LT LT LT Heat #C-1279-3 C-1279-3 C-1279-3 C-1279-3 C-1279-3 6..0 IL 120 100 80 60 40 20 0-200 0 Set1-100 0 100 200 300 400 Temperature in Deg F a Set2 0 Set3 a Set4 v Set5 Results 500 Curve-1 2 3 4 5 Fluence LSE USE d-USE.0 T @50 75.0 0 9. 26E18 1. 66E19 2. 09E19 4. 01E19.0 100.0.0 100.0.0 100.0.0 100.0.0 100.0.0 203.5.0 220.5.01 195.7.0 231.0 d-T @50.0 128.5 145.5 120.7 156.0 Figure 6-6. Percent Shear vs. Temperature for Palisades Surveillance Plate D-3803-1 (Longitudinal Orientation)

W-100 PLATE (LONGITUDINAL)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 01/08/2004 10:29 AM Page 1 Coefficients of Curve 1 A = 52.1 B = 49.9 C = 73.2 TO = 193.41 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=102.0(Fixed)

Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=158.6 Deg F Temp@50 ft-lbs=1 90.4 Deg F Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

LT Capsule: W-100 Fluence: 2.09E19 n/cma^2 120 100.0>0 80 60 40 20 0 00 0 0 0 0 100 200 300 400 Temperature In Deg F Charpy V-Notch Data Temperature 70.00 110. 00 150.00 175. 00 200. 00 225. 00 250. 00 260. 00 270.00 Input CVN 5.50 14.00 29.50 44.00 45. 00 74.50 86.50 73.00 102.00 Computed CVN 5.51 11. 47 25.55 39. 81 56.58 72. 39 84. 47 88. 08 91.04 Differential

-.01 2.53 3.95 4.19-11.58 2. 11 2. 03-15.08 10.96 W-100 PLATE (LONGITUDINAL)

Page 2 Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

LT Capsule: W-100 Fluence: 2.09E19 n/cMA^2 Charpy V-Notch Data Temperature 280. 00 300.00 325. 00 Input CVN 100. 50 104. 50 10.1.00 Computed CVN 93. 44 96. 86 99.33 Differential

7. 06 7. 64 1. 67 Correlation Coefficient

= .978 W-100 PLATE (LONGITUDINAL)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 01/08/2004 10:30 AM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 66.65 TO = 195.62 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 195.7 Plant: PAUSADES Material:

D-3803-1 Heat: C-1279 Orientation:

LT Capsule: W-100 Fluence: 2.09E19 n/cmA2 120 100 80 (L 40 20 0 0 100 200 300 400 Temperature in Deg F Charpy V-Notch Data Temperature 70.00 110.00 150. 00 175.00 200. 00 225.00 250.00 260.00 270.00 Input Percent Shear.00 5.00 25.,00 40.00 45. 00 70.00 85. 00 80. 00 95. 00 Computed Percent Shear 2. 25 7.11 20.28 35.01 53.28 70.71 83. 64 87. 34 90. 31 Differential

-2. 25-2.11 4.72 4.99-8.28-.71 1.36'-7.34 4. 69 W-100 PLATE (LONGITUDINAL)

Page 2 Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

LT Capsule: W-100 Fluence: 2.09E19 n/cmA2 Charpy V-Notch Data Temperature 280.00 300. 00 325. 00 Input Percent Shear 100. 00 100.00 100.00 Computed Percent Shear 92. 63 95. 82 97.98 Differential

7. 37 4. 18 2.02 Correlation Coefficient

=.992 W-100 PLATE (LONGITUDINAL)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 01/08/2004 10:35 AM Page 1 Coefficients of Curve 1 A = 31.9 B = 31.9 C = 73.39 TO = 188.47 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.--63.8(Fixed)

Lower Shelf L.E.=.0(Fixed)

Temp.@L.E.

35 mils=195.7 Deg F Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

LT Capsule: W-100 Fluence: 2.09E19 n/cmA2 80 70 60_ 50.2 1140 30 20 10 0 0 Temperature 70.00 110.00 150. 00 175. 00 200. 00 225.00 250. 00 260. 00 270.00 100 200 300 Temperature in Deg F 400 Charpy V-Notch Data Input L.E.1.00 7.00 18.00 27.00 34. 00 51.00 50. 00 50.00 57. 00 Computed L.E.Differential

2. 43 6. 73 16. 56 26. 11 36. 87 46.59 53. 75 55. 85 57.56-1.43.27 1.44.89-2.87 4.41-3.75-5. 85-. 56 W-100 PLATE (LONGITUDINAL)

Page 2 Plant: PALISADES Material:

D-3803-1 Orientation:

LT Capsule: W-100 Fluence: Heat: C-1279 2.09E19 n/cm^2 Charpy V-Notch Data Temperature 280.00 300. 00 325. 00 Input LE.58.00 72. 00 68. 00 Computed LE.58.94 60. 89 62.29 Differential

-.94 11. 11 5.71 Correlation Coefficient

= .981 W-100 PLATE (TRANSVERSE)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 01108/2004 10:37 AM Page I Coefficients of Curve 1 A = 37.6 B = 35.4 C = 77.92 TO = 177.79 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=73.0(Fixed)

Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=160.8 Deg F Temp@50 ft-lbs=206.3 Deg F Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

TL Capsule: W-100 Fluence: 2.09E19, nr'cmA2 80 70 60 0 50 40 nL z 30 20 10 0 0 100 200 300 Temperature in Deg F Charpy V-Notch Data 400 K K Temperature

70. 00 110.00 150.00 200. 00 225.00 240. 00 250. 00 260. 00 270. 00 Input CVN 9.50 14.50 27.50 44. 00 52.50 50.00 71.50 69. 00 71.00 Computed CVN 6. 39 12. 77 25.48 47.42 56. 76 61. 07 63.41 65.34 66.93 Differential

3. 11 1 .73 2.02-3. 42-4. 26-11. 07 8. 09 3. 66 4. 07 L W-100 PLATE (TRANSVERSE)

Page 2 Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

TL Capsule: W-100 Fluence: 2.09E19 n/cm^2 Charpy V-Notch Data Temperature 285. 00 300.00 325. 00 Input CVN 77.00 76.50 67.50 Computed CVN Differential

68. 75 70. 05 71.42 8.25 6.45-3.92 Correlation Coefficient

= .970 W-100 PLATE (TRANSVERSE)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 01/08/2004 10:38 AM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 58.39 TO = 218.6 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)I(C+DT))]

Temperature at 50% Shear = 218.7 Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

TL Capsule: W-100 Fluence: 2.09E19 n/erA2 120 100 80 U-~60 L I..40 20 0 0 00 0 *0(0 100 200 Temperature in Deg F Charpy V-Notch Data 300 400 Temperature 70.-00 110.00 150. 00 200. 00 225. 00 240. 00 250. 00 260. 00 270. 00 Input Percent Shear.00 5.00 20. 00 40. 00 50. 00 50. 00 70. 00 90. 00 95. 00 Computed Percent Shear.61 2. 37 8.71 34. 59 55. 46 67. 55 74. 56 80. 50 85. 33 Differential

-.61 2.63 11. 29 5.41-5.46-17.55-4.56 9. 50 9. 67

-W-100 PLATE (TRANSVERSE)

Page 2 Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

TL Capsule: W-100 Fluence: 2.09E19 nrcmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 285.00 100.00 90.67 9.33 300. 00 100.00 94. 20 5. 80 325.00 100.00 97. 45 2.55 Correlation Coefficient

= .975 80 70 60= 50.9 2 30 20 10 W-100 PLATE (TRANSVERSE)

CVGRAPH 5.0.1 Hyperbolic Tangent Curve Printed on 01/08/2004 10:39 AM Page 1 Coefficients of Curve 1 A = 28.65 B = 28.65 C = 77.19 TO = 188.4 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=57.3(Fixed)

Lower Shelf L.E.=.O(Fixed)

Temp.@L.E.

35 mils=205.9 Deg F Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

TL Capsule: W-100 Fluence: 2.09E19 n/ceaA2 2 0 0 ...0-0 0 0 I 100 Temperature

70. 00 110.00 150. 00 200. 00 225.00 240. 00 250. 00 260. 00 270. 00 Input 200 Temperature in Deg F Charpy V-Notch Data LE. Computed L.E.O0 2.55 O0 6.64 00 15.47 O0 32.92 00 41.30 O0 45.38 00 47.64 00 49.55 00 51.13 300 400 Differential 2.7.19.32.39.36.54.47.54.-.55.36 3. 53-. 92-2.30-9.38 6.36-2.55 2. 87 W-100 PLATE (TRANSVERSE)

Page 2 Plant: PALISADES Material:

D-3803-1 Heat: C-1279 Orientation:

TL Capsule: W-100 Fluence: 2.09E19 n/cm^2 Charpy V-Notch Data Temperature 285. 00 300. 00 325. 00 Input L.E.56.00 63. 00 56. 00 Computed L.E.52. 96 54. 29 55. 68 Differential

3. 04 8.71.32 Correlation Coefficient

"-.973