Text

Transmittal of Fort Calhoun, Unit 1, Reactor Coolant System Pressure and Temperature Limits Report, Revision 4
	ML081490283
Person / Time
Site:	Fort Calhoun
Issue date:	05/27/2008
From:	Matthews T; Omaha Public Power District
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LIC-08-0071
	Download: ML081490283 (70)
	v • d • e

Oma~a Public PDwer District

'!Q-tt/- 200(,

Fort Calhoun Station Po. Box 550 Fort Calhoun, NE 68023 May 27,2008 LI C-08-0071 U, S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Reference:

Docket No. 50-285

SUBJECT:

Transmittal of Fort Calhoun Station (FCS), Unit No.1, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), Revision 4 Pursuant to FCS Technical Specification 5.9.6c, Revision 4 of the FCS PTLR is attached. Revision 4 was issued on May 18, 2008, to incorporate changes due to the uprate of shutdown cooling initiation conditions.

No regulatory commitments are contained in this submittal.

If you have any questions, please contact Ms. Susan Baughn at (402) 533-7215, Sincerely, T. C. Matthews Manager - Nuclear Licensing TCM/mle

Attachment:

TDB-IX - Technical Data Book - RCS Pressure and Temperature Limits Report - Revision 4

PAGE 1 OF 69 Fort Calhoun Station Unit 1 TDB-IX TECHNICAL DATA BOOK RCS PRESSURE AND TEMPERATURE LIMITS REPORT Change No. EC 35639lEC 43060 Reason for Change Changes made and new reference added due to the uprate of SDC initiation conditions.

Requestor J. Johnson Preparer L. Hautzinger Issue Date 05-18-08 8:30 pm

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 2 OF 69 Table of Contents Section Page INTRODUC-TION ..........................................................................................................................

3

1. NEUTRON FLUENCE VALUES ............................................................................................3

2. REACTOR VESSEL SURVEILLANCE PROGRAM ...............................................................5

3. LTOPSYSTEMLIMITS ......................................................................................................... 5

4. BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURE ....................................6

5. PRESSURE-TEMPERATURE LIMITS USING LINII-TING ART IN THE P-T CURVE 6

CALCULATION ............................................................................................................................

6. MINIMUM TEMPERA1-URE REQUIREMENTS IN THE P-T CURVES .................................7

7. APPLICATION OF SURVEILLANCE DATA TO ART CALCULATIONS ................................8

8. REFERENCES .................................................................................................................... I 0 - CEN.636, Revision 2, "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline IVlaterials. dated July 2000 ................................... 13 List of Figures Figure Number Page Figure 5-1 . FORT CALHOUN STATION UNIT 1 COMPOSITE P/T LINIITS. 40 EFPY.............12

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 3 OF 69 RCS PRESSURE AND TEMPERATURE LlNllTS REPORT INTRODUC'I'ION The purpose of this Technical Data Book (TDB) section is to provide Fort Calhoun Station (FCS) with an administrative document that defines updating the pressure and temperat~lre(P-T) limit curves and low temperature overpressure protection (LTOP) setpoints and delineates Nuclear Regulatory Commission (NRC) review requirements as defined in the Technical Specifications (TSs) Definitions section.

This Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) for FCS Unit No. 1 contains P-T limits corresponding to 40 Effective Full Power Years (EFPY) of operation. In addition, this report references the LTOP methodology and current analysis that contains the system lirr~itsand operatirlg restrictions that protect the P-T limits from being exceeded during limiting LTOP events.. Reference 8.1 allows the relocation of the P-T limit curves and LTOP system limits from the plants TSs and relocates them into a PTLR.

Reference 8.2 is the topical PTLR that forms the basis for this document except as modified by the individual Sections.

This PI-LR will be updated prior to exceeding the adjusted reference temperature (ART (RTNDT))

u,tilizedto develop Figure 5-1. The PTLR, including any revisions or supplements thereto, shall be provided upor1 issuance of P-T limit curves to the NRC Document Control Desk with copies to the Regional Administrator and Senior Resident Inspector. -. .

In addition, anytime it becomes necessary to change the methodology and/or any TSs that were used to develop data generated for this report, a license amendment will also be prepared describing the new methodology and/or TS change and will be submitted for NRC review and approval prior to irr~plementationin this report. -

1. NEUTRON FLUENCE VALUES The most recent reactor vessel beltline neutron fluence has been calculated for the critical locations in Reference 8.3. (Note: The uncertainty associated with the fluence values stated in Reference 8.3 is *I 5.5%.) This reporvreference contains the following:

a) A description of the methodology used to perform the neutron fluence calculation.

b) A description of the computer codes used to calculate the neutron fluence values.

c) A description of how the computer codes for calculating the neutron 'Fluence values were benchmarked.

The methodology stated in Reference 8.3 is consistent with the guidance of Draft Regulatory Guide DG-1053 (now Regulatory Guide RG 1.190), as stated by the NRC staff in the safety evaluatio~iscontained in References 8.4 and 8.5.

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 4 OF 69 The values of fast neutron fluence (E > I IWev) used in the ART calculations in Section 4 are located in Table 1-1 and are applicable for 40 EFPYs. (Note: The fluence associated with 40 EFPYs versus 48 EFPYs was used in the ART calculations for Figure 5-1 to prevent a reduction in the operating window between the P-T limit and the reactor coolant pump net positive suction head curves.) The 114 T and the 314 T neutron fluence values were calculated as follows:

a) The cladlbase metal interface fluence values for the plates and circumferential weld use the peak neutron value listed in Table 6.2-1 of Reference 8.3 for 40 EFPY. This is due to these materials would be exposed to the highest fluence.

b) The cladlbase metal interface fluence value used for the limiting axial welds was the value located at the 60" position for 40 EFPY. The axial welds for the 180" position is not limiting due to the fluence at this location is significantly less than at the 60" and 300" locations. The non-limiting 2-410 welds at the O0,120°, and 240" positions are located in geometrically symmetric locations as the 3-410 welds at 60°, 180°,

and 300" positions. In Cycle 14, extreme low radial leakage fuel management was implemented to reduce the reactor vessel fast neutron flux. This management scheme and the incorporation of surveillance data from other nuclear power plants per Reference 8.14 ensures that FCS has the potential to operate to August 9, 2033 without exceeding the 10 CFR 50.61 pressurized thermal shock (PTS) screening criteria as approved by the NRC in Reference 8.5.

c) Equation 3 of Reference 8.22 was then used to calc~~late the 114 T and the 314 T fluence values as shown in Table 1-1.

(Note: The values in parentheses in Table 1-1 refers to weld wired heat numbers.)

Table 1-1, Neutron Fluence Values for 40 EFPY

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 5 OF 69

2. REACTOR VESSEL SURVEILLANCE PROGRAM The reactor vessel surveillance program is described in Section 2, Reference 8.2. The reactor vessel surveillance withdrawal schedule is located in Reference 8.6, Table 4.5-4. This schedule meets the requirements of ASTM-E-185-82 (Reference 8.25). The baseline report describing the pre-irradiation evaluation of the FCS reactor surveillance materials are presented in Reference 8.7. The reports describing the post-irradiation evaluation of the FCS surveillance capsules are contained in References 8.8 - 8.10. Each removed capsule has been evaluated in accordance with the testing requirements of the version of ASTM-E-185 in effect at the time of capsule removal.

3. LTOP SYSTEM I-IMITS The LTOP system setpoints have been developed by making a comparison between the peak transient pressure for each limiting LTOP event and the P-T limit curve of Figure 5-1 to ensure that the P-T limit curve is not exceeded.

These system setpoints and additional lirr~itationsfor LTOP have been established based on NRC-accepted methodology and are described in References 8.15, 8.16 and 8.26. (Note: The I methodology described in Section 3 of Reference 8.2 was not used for the determination of the LTOP system setpoints.)

The LTOP analysis which contains the current system setpoints and operating restrictions to ensure the P-T limit curve is not exceeded during a limiting LTOP event are located in References 8.16 and 8.26. The applicable operating restrictions stated in Reference 8.26 will be maintained in the TSs. Reference 8.21 contains the methodology for incorporating the -

Reference 8.16 setpoints into the LTOP system actuation circuitry. These conservative values will then be used for incorporation into TDB Figures. The LTOP enable temperature is 350°F.

(Reference 8.24)

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 6 OF 69

4. BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURE The calculation of the ART for the reactor vessel beltline region has been performed using the NRC-accepted methodologies as described in Section 4, Reference 8.2. Application of surveillance data was used to refine the chemistry factor and the margin term in Reference 8.14.

(See Section 7) The limiting weld for FCS is the 3-410 weld located at the 60°1300" position using weld wire heat 12008113253. The RTpTsvalue for the limiting weld is projected to be 268°F with a cladlbase metal interface fluence of 2.43 x 10 nlcm2 at the end of license extension (August 9, 2033).

The ART values in the beltline region for FCS Unit 1 corresponding to 40 EFPY are listed in Table 4-1. (Note: The limiting ART value for the 114 T and 314 T (Weld 3-410, Weld Wire Heat 12008113523) was incorporated into Figure 5-1 (References 8.19 and 8.23).)

5. PRESSURE-TEMPERATURE LIMITS USING LIMITING ART IN THE P-T CURVE CALCULATION The analytical methods used to develop the beltline RCS P-T limits are based on NRC reviewed methodologies as discussed in Section 5 of Reference 8.2. The NRC approved the use of ASME Code Case N-640 for FCS that allows the use of KlcZo calculate the reference stress intensity factor KIRvalues for the reactor pressure vessel as a function of temperature in Reference 8.17. The limit for the maximum pressure in the vessel is 100 percent of the pressure satisfying Paragraph G-2215 of the 1996 Edition of Appendix G to the ASNlE Code for establishing LTOP lirr~itsetpoints. Additionally, an exemption was granted by the NRC to apply CE NSSS niethods for determining P-T limit curves.

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 7 OF 69

@ The ferritic reactor pressure vessel materials that have accumulated neutron fluences in excess of 1.0 x loi7 n/cm2 (E > I Mev) regardless of whether the materials are located within the region immediately surrounding the active core have been evaluated (Reference 8.18). This evaluation concluded that the limiting material remained the lower shell axial welds, 3-410 AIC.

Figure 5-1 was developed in Reference 8.19 and modified per Reference 8.24. Uncertainty was incorporated into Figure 5-1 as follows (Reference 8.19):

a) Above the LTOP enable temperature (350°F), pressure instrument uncertainty is incorporated into the P-T limit curve and below this temperature it is not. (Note:

Pressure instrument uncertainty is not applied below the LTOP enable temperature due to it being incorporated into the LTOP system setpoint curve). A pressure instrumentation uncertainty of 50 psi is being used, which bounds the wide and narrow range pressurizer pressure instrl-~mentsthat operators would use to determine RCS pressure.

b) The temperature uncertainty used is 14°F which bounds the instruments that operators would use to determine RCS temperature.

6. MINIMUM TEMPERATURE REQUIREMENTS IN THE P-T CURVES The minimum temperature requirements specified in Reference 8.20 are applied to the P-T limit curves using the NRC-reviewed methodologies as described in Section 6 of Reference 8.2.

The minimum temperature values applied to the P-T limit curves for FCS Unit 1 corresponding to 40 EFPY are (Note: These limits were calculated in Reference 8.19 and incorporates instrument uncertainty):

a) Minimum B o l t ~ ~

Temperature:

p 64°F.

b) Minimum Hydrostatic Temperature Test Limits: See Figure 5-1. (Note: The in-service hydrostatic test curve is developed in the same manner as the heatup and cooldown curves with the exception that a safety factor of 1.5 is used in lieu of 2.)

c) Lowest Service Temperature: 164°F.

d) Flange Limit:

1) Normal Operation: 144°F.

2) Hydrostatic and Leak Testing: 114°F.

e) Core Critical Temperature Limit: 515°F per TS 2.10.1(1). (Note: This TS limit is more conservative than the core critical temperature limit required by Reference 8.20. Whenever the P-T limit curve of Figure 5-1 is modified, it must be verified that the new core CI-iticalpeak temperature lirnit is less than 515"F, or else the core critical P-T limit curve must be included on Figure 5-1 and Section 6, iteni

'e' must be updated.)

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 8 OF 69 In the development of P-T lirnits for CE NSSS's, the intent is to utilize the more conservative of either the lowest service temperature or the other mir~imljmterr~peraturerequirements for the reactor vessel when the RCS is pressurized to greater than 20% of the pre-service hydrostatic test pressure (PHTP). The "minimum pressure criteria" specified in Reference 8.20 serves as a regulatory breakpoint in the development of P-T limits and is defined as 20% of PHTP. For CE NSSS plants, the PHTP is defined as 1.25 times the design pressure (Note: Design pressure =

2500 psia). The function of minimum pressure in the development of P-T limits is to provide a transition between the various temperature only based P-T limits, such as miniml-~mbolt LIP and the lowest service temperature of flange limits.

For FCS Unit 1, the minimum pressure is calculated as follows:

Minimum Pressure = (1.25 x design pressure) x 0.20

= 1.25 x 2500 psia x 0.20

= 625 psia Therefore, when the pressure correction factors (Reference 8.19) are applied to 625 psia, the minimum pressure(s) are as follows:

Actual RCS Temperature < 210°F = 564 psi Actual RCS Temperature > 210°F = 558 psi The pressure of 564 psi is the most significant value due to the RCS can not exceed this pressure until RCS temperature is greater than the lowest service temperature value stated in Section 6 item 'c' above. The lowest service te~iiperatureis the limiting minimum temperature value and is incorporated into Figure 5-1. The heatup and cooldown limit curve is more conservative than the minimum pressure value in the temperature range specified, but the in-service hydrostatic test curve is limited by the regulatory requirement (Reference 8.20).

7. APPLICA'TION OF SURVEILLANCE DATA TO ART CALCULATIONS Post-irradiation surveillance capsule test results for FCS Unit 1 are given in References 8.8 - 8.10. Additional reports containing surveilla~icecapsule data from other nuclear power plants are located in References 8.11 - 8.13. These additional surveillance reports, along with others that are contained in Reference 8.14 (Attachment I ) , were deemed credible and approved for use in the FCS surveillance program as stated by the NRC staff in Reference 8.5.

Additionally, Reference 8.5 requires the following:

a) Future core loadings are limited to the core neutron leakage to v a l ~ ~ e similar s to those for Cycles 15 and 16 which will satisfy the re uirement of end of license 9

(August 9, 2033) fluence accumulation of 2.43 x 10 9 neutrons/cm2 to the limiting welds.

b) Caution is exercised to preclude misloading any of the peripheral assemblies which would invalidate the loading requirements.

c) New data from the Mihama Unit 1, Diablo Canyon Unit 1 and Palisades plants is assessed by the FCS staff as it becomes available, since the data from these plants were used in.the FCS PTS analysis.

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 9 OF 69 The use of surveillance data from ,these "Sister" reactor vessels (as stated in Section 7 item 'c' above) is required to ensure that FCS does not exceed PTS screening criteria during its extended lifetime (August 9, 2033).

A review of the surveillance programs of Mihama Unit 1 (12008127204), Diablo Canyon Unit 1 (27204); Palisades Supplemental Capsules (27204), and the FCS W-275s Capsule (27204 and 12008113253) concluded further data should be available for use in the FCS reactor vessel s~~rveillanceprogram as follows: (Note: The values in parentheses correspond to weld wire heat numbers.)

a) Mihama Unit 1 (Weld Wire Heat 12008127204)

The data from Capsules 1-3 were used in Reference 8.14. The removal schedule for the remaining Mihama Unit 1 capsules as of 2000 was:

1) Capsule 4 was scheduled for removal in 2001; results are expected in 2002.

2) Capsule 5 is scheduled for removal in 2010; results are expected in 201 1.

3) Capsule 6 is currently considered in standby with no scheduled removal date.

Attempts to obtain additional information from KANSAI Electric Company by OPPD, MHI, and AREVA NP have not yielded any response or additional data.

b) Palisades (Weld Wire Heat 27204127204)

The removal schedule for the Palisades capsules are: -

1) Capsule SA-60-1 was pulled and evaluation data are found in internal report -

ATI-99-006-002 (814199). The capsule report should be submitted to the NRC in 2003 or 2004. The data was used in Reference 8.14.

2) Capsule SA-240-1 was pulled and was evaluated by Framatome. A summary of the data was provided to OPPD by Palisades Staff and evaluated by Westinghouse for continued validity.

c) Diablo Canyon Unit 1 (Weld Wire Heat 27204)

The removal schedule for the Diablo Canyon Unit 1 capsules and the status of the results that are reported to the NRC are:

1) Caps~.~leDCI-S data are contained in Reference 8.11 and was used in Reference 8.14.

2) Capsule DCI-Y data are contained in Reference 8.12 and was used in Reference 8.14.

3) Capsule DCI-V was removed in 2002 and submitted to the NRC (ML031400347). This is the last of the three original capsules containing 27204 weld material.

4) Capsule DCI-C (supplemental) and DC1-D (supplemental) were removed, but were stored in the spent fuel pool. Due to planned changes to 10CFR50.61, there are presently no plans for re-insertion or evaluation. (Note: DCI-D was fabricated using the FCS 1-410B (27204) nozzle dropout.)

R4

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 10 OF 69

8. REFERENCES 8.1 NRC GL 96-03, "Relocation of Pressure-Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," January 31, 1996.

8.2 CE NPSD-683-A, Rev 06, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," April 2001.

8.3 WCAP-15443, Revision 0, "Fast Neutron Fluence Evaluations for the Fort Calhoun Unit 1 Reactor Pressure Vessel," July 2000 [Contained in Letter LIC-00-0064 from OPPD (W.

G. Gates) to NRC (Document Control Desk), dated August 3, 20001.

8.4 Safety Evaluation by the Office of NRR Related to Amendment Number 197 to Facility Operating License Number DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated March 27, 2001.

8.5 Safety Evaluatior~by the Office of NRR Related to Amendment Number 199 to Facility Operating License Number DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated June 6, 2001.

8.6 USAR Section 4.5.3, Revision 3, dated May 29, 2002.

8.7 TR-0-MCD-001, "Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program," dated March 22, 1977.

8.8 TR-0-MCM-001, Revision 1, "Fort Calhoun Station Unit No. 1 Evaluation of Irradiated Capsule W-225," dated August 28, 1980 [Contained in Letter LIC-81-0011 from OPPD (W.C. Jones) to NRC (H.R. Denton), dated January 23, 19811.

8.9 TR-0-MCM-002, "Fort Calhoun Station Unit No. 1 Evaluation of Irradiated Capsule W-265," dated March 7, 1984 [Contained in Letter LIC-84-124 from OPPD (W.C. Jones) to NRC (D.G. Eisenhut), dated April 25, 19841.

8.10 BAW-2226, "Fort Calhoun Station Unit No. 1 Evaluation of Irradiated Capsule W-275,"

dated July 21, 1994 [Contained in Letter LIC-94-0250 from OPPD (T.L. Patterson) to NRC (Document Control Desk), dated December 9, 19941.

8.1 1 WCAP-11567, "Analysis of Capsule S from the Pacific Gas and ~lectricCompany Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," December 1987.

8.12 WCAP-13750, "Analysis of Capsule Y from the Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," July 1993.

8.13 WCAP-13440, "Supple.mental Reactor Vessel Radiation Surveillance Program for the Pacific Gas and Electric Company Diablo Canyon Unit No. 1," December 1992.

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 11 OF 69 8.14 CEN-636, Revision 2, "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline Materials," dated July 2000 [This document is located in the Attachment 1.I.

8.15 FC06876, Rev. 0, "Performance of Low Temperature Overpressure Protection System Analyses Using RELAPS: Methodology Paper."

8.16 FC06877, Rev. 1, "Low Temperature Overpressure Protection (LTOP) analysis, Revision 1."

8.17 Safety Evaluation by the Office of NRR Related to Amendment Number 207 to Facility Operating License Number DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated April 22, 2002.

8.18 Letter LTR-CI-01-25, Rev. 0 from WEC (S. T. Byrne) to OPPD (J. Jensen), "Assessment of Extended Beltline Limit for Fort Calhoun Station Reactor Pressure Vessel," dated December 18. 2001.

8.19 EA-FC-01-022, Rev. 0, "Pressure and Temperature Limit Curve for 40 EFPY."

8.20 10 CFR 50 Appendix G, "Fracture Toughness Requirements."

8.21 FC06863, Rev. 1, "LTOP Setpoint Instrument Loop Uncertainty and LTOP Trip Curve Development."

8.22 Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials."-

8.23 FC06799, Rev. 0, "40 EFPY Pressure and Temperature Limit Curve Inputs."

8.24 EA-FC-02-025, Rev. 0, "Development of the RCS P-TLR."

8.25 WCAP-15741, Rev. 0, "Reactor Vessel Surveillance Program Withdrawal Schedule Modifications," dated September 2001 [Contained in Letter LIC-01-0107 from OPPD (R.

L. Phelps) to NRC (Document Control Desk), dated November 8, 20011.

8.26 FC07187, Rev. 0, Low Temperature Overpressure Protection Analysis in Support of Shutdown Cooling Initiation Temperature Change

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 12 OF 69 Figure 5 FORT CALHOUN STATION UNIT 1 COMPOSITE PIT LIMITS, 40 EFPY RCS PRESSURE-TEMPERATURE LIMITS FOR HEATUP, COOLDOWN, AND INSERVICE HYDROSTATIC TEST

'TEMP, t .lhITT

~ ... ...-..-..--"-,".-,%

T,INLlrCA1'ED ZZEACT8R COOLANT SYSTJCSI TEMPERATUm (OF)

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 13 OF 69 Attachment 1 - CEN-636, Revision 2, "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline Materials," dated July 2000

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 14 OF 69 Report Preparedfor the Omaha Public Powver D i i Fort Calhoun Station Final Report dated July 19,2000 CEN-636, Revisii 02 V e r i k a t i i Status: Complete Evaluation of Reactor Vessel I Surveillance Data Pertinent to the Fort Calhoun Reador Vessel I Beltline Materials Basis for Prediction of RT, for the Fort Calhoun RPV Westinghouse Electric CE Nuclear Power Windsor, Connecticut

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK PAGE 15 OF 69 Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline Materials Basis for Prediction of RT, CEN-636, Revision 02 VERIFICATION STATUS: COMPLETE Prepared by:

C. L. Hoffmann Reviewed by: Date: 7 / i 9 / 20-C. J. Gimbrone Approved by: Date: ?/ ,? k fi.d p CEN-636,Revision 02 Page 2 of 56

TDB-IX FORT CALHOLlN STATION PAGE 16 OF 69 TECHNICAL DATA BOOK Table of Contents Paqe No.

Record of Revisions 2 Table of Contents 3 List of Tables 4 List of Figures 5 6

2.0 Infmdudionand B&ground 6 3.0 Description of Fort Calhoun Reactor Vessel Beltline Materials 8 4.0 Description of Surveillance Data Relevant to Fort Calhoun 9 5.0 Regulatory Position 2.1 Analysis of Relevant Surve~llanceData 10 5 I Credibility of Surveillance Data 10 5.2. Traceability of Mihama 1 Surveillance Data 13 5.3. Analysis Approach 5.4 Surveillance Data Analysis 6.0 Evaluation of Surveillance Data Credibility and Applicability to Fort Calhoun 26 7.0 Calculation of RT, 30 8.0 Conclusions References 3CEN-636, Revision 02

TDB-IX FORT CALHOLIN STATION TECHNICAL DATA BOOK REFERENCE USE PAGE 1.7OF 69 List of Tables No. Paqe No.

I Identificationof Reactor Vessel Plates and Welds in the Fort Calhoun Reactor Vessel Beltline 39 ldentification of Reactor Vessel Surveillance Program Welds Applicable to the Fort Calhoun Vessel Beltline Welds Test Results from the D.C. Cook Unit I Reactor Vessel Surveillance Program Test Results from the Diablo Canyon Unit 1 and Palisades Reactor Vessel Surveillance Program (Pre-Adjusted Data)

Test Results from the Diablo Canyon Unit 1 and Palisades Reactor Vessel Surveillance Program Test Results from the Salem Unit 2 Reactor Vessel Surveillance Program Test Results from the Mihama Unit 1 Reactor Vessel Surveillance Program (Pre-Adjusted Data)

Test Results from the Mihama Unit 1 Reactor Vessel Surveillance Program Derived Chemistry Factors for Reactor Vessel Surveillance Program Welds Applicable to Fort Calhoun Test Results from the Fort Calhoun Reactor Vessel Surveillance Program (SurveillanceWeld Wire Heat 305414)

Test Results from the Fort Calhoun Reactor Vessel Surveillance Program (Surveillance Plate Heat No. A1768-1)

Test Results from the Fort Calhoun Reactor Vessel Surveillance Program (Standard Reference Material)

~.... -.

CEN-636,Revision 02 Page 4 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 18 OF 69 List of Tables (cont'd) i No. Paqe No.

9 Derived Chemistry Factors for Fort Calhoun Reactor Vessel

'Surveillance Program Materials 10 Predicted RT, for the Fort Calhoun Reactor Vessel Beltline Plates and Welds A? Standard Reference Material Data from Combustion Engineering Designed Surveillance Capsules 55 A2 Analysis of Standard Reference Materials List of Figures No.

1 Effect of Tcold on SRM Data, HSST Plate 01 Results Normalized Paqe No.

I to 1E l 9 nlcmZ 53 2 Effect of T w l d on SRM Data HSST Plate 01 Results (CF=130.3 F) 54 CEN-636, Revision 02 Page 5 of 56 I

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 19 OF 69 1.0 Obiective This report evaluates surveillance data to demonstrate that the Fort Calhoun reactor pressure vessel will not exceed the Pressurized Thermal Shock (PTS) screening criteria (Reference 1) through the end of the current and renewal license terms (August 9, 2013 and August 9, 2033, respectively). This evaluation is based on the use of Position 2.1 of Regulatory Guide 1.99 (Reference 2) to calculate chemistry factors for the limiting weld wire heat combinations and justify reduction of the standard deviation for shift by one-half based on credible surveillance data. The PTS screening criteria projections are based on conservative values of neutron fluence that were calculated using the methods of the U.S.

Nuclear Regulatory Commission's Draft Regulatory Guide DG-1053, "Calculationat and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence". The approach used for calculating RTms complies with 10CFR50.61(b)(3). The objective of this report is to support NRC approval of the report's conclusions.

2.0 Introduction and Backclround The Fort Calhoun reactor vessel was fabr~cated by Combustion Engineering in Chattanooga, Tennessee during the time period 1966 to 1969. The vessel shell was fabricated using steel plates purchased to SA-533 Grade B, Class 1 requirements. The plates were joined together using automatic submerged arc welding using coppercoated electrodes. The primary coolant nozzles and the vessel flange were fabricated using forgings purchased to SA-508 Class 2 requirements. The forgings were joined to the II vessel shell using automatic and manual submerged arc welding.

I The reactor vessel shell, primary coolant nozzles and the vessel flange were designed to operate at high temperatures and pressures. The reactor vessel beltline materials were also designed for exposure to the fast neutrons generated in the reactor core. The material purchase specifications together with the forming, welding, and post-weld heat treatment processes were intended to provide for a high level of fracture toughness. The pre-service inspection and hydrostatic testing processes were intended to minimize the presence of fabrication-induced defects that could grow during the service lifetime. During the lifetime of the reactor vessel, periodic ~n-serviceinspections are conducted to look for defect indications in the vessel welds. In addition, a reactor vessel surveillance program is I

CEN-636, Revision 02 Page 6 of 56 i

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 20 OF 69 maintained throughout the life of the vessel to monitor the effect of neutron irradiation on the beltline materials.

Given the fact that the beltline.welds in the Fort Calhoun vessel were fabricated using copper coated electrodes, the copper content i n those welds is high (relative to vessel welds fabricated using noncopper coated electrodes). Such high copper welds have been shown to be more sensitive to the hardening effects of fast neutron irradiation than vessels fabricated during the mid- and late-1970s using noncopper coated welding electrodes.

Neutron irradiation causes a reduction of the fracture toughness in the reactor vessel beltline materials. This toughness reduction is manifested as a shift in the reference temperature, R ,T

,, to a higher value. The shift increases as a function of the fast neutron fluence and chemical content (specifically the copper and nickel content as used in Reference2). The magnitude of the shift is sensitive to the product form (e.g., plate or weld material).

The methodology for predictrng shift that is currently acceptable to the NRC is provided in References 1 and 2. These two documents plus a handout entitled "Evaluation and Use of Surveillance Data" (Reference 3) from a November 12, 1997 NRC-Industry Meetrng provide a set of NRC requirements and guidelines for using relevant and credible surveillance data to refine predictions of the shift in RT, and calculation of the adjusted reference temperature, ART. (Values of ART, or RT, in Reference 1, are obtained using the sum of the initial RTNm, the shift of RT

,, with irradiation, and a margin term.) In the longer term, work is proceeding on the development of an improved methodology for predicting values of ART. This longer term work entails an ASTM effort to revise ASTM Standard E900 and an NRC effort to revise Regulatory Guide 1.99. A recent report on that program is NUREGICRS551 (Reference 4).

The approach being taken in this document is to apply Position 2.1 of Regulatory Guide 1.99 (Reference 2) using surveillance data applicable to the limiting Fort Calhoun beltline welds. (Position 2.1 provides a procedure for adjusting the chemistry factor used to predict shift and for reducing the standard deviation for shift in the margin term.) Several weld wire heats in various combinations were used in the beltline welds for the Fort Calhoun vessel.

Therefore, numerous sources of surveillance data are being evaluated to give the broadest possible picture of the irradiation performance for the Fort Calhoun beltline welds. Data reviewed for applicability to Ft. Calhoun are Mihama Unit 1, Diablo Canyon Unit 1, D.C.

CEN-636, Revision 02 Page 7 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 21 OF 69 Cook Unit 1, Salem Unit 2, and a supplemental surveillance capsule from Palisades. Other welds that used one of the electrode heats in combination w~thanother to produce the surveillance weld were also reviewed. These are labeled in Table 2 as "not fully applicable" to the Fort Calhoun vessel limiting beltline welds. The applicable data were then analyzed in accordance with Position 2.1, chemistry factors were calculated, and data predictability assessed. The results of this Position 2.1 analysis were then used to calculate the adjusted reference temperature, R ,T

, applying the adjusted chemistry factor and the reduced standard deviation for shift from the analysis. The revised values of RT, are being reported to the NRC in accordance w~ththe requirements of 10CFR50.61 (b)(3).

3.0 Description of Fort Calhoun Reactor Vessel Beltline Materials The Fort Calhoun reactor vessel beltline materials and surveillance materials are described in Table 1. The first column gives the plate code or the weld seam identification. The second column gives the heat number for the plate or welding electrode. The third column gives the flux type and lot number for the welds. The fourth column gives the chemistry factor based on the best estimate copper and nickel content. (The material identification and the weld chemistry factor values are from Reference 5.)

The Fort Calhoun beltline consists of the intermediate and lower shell courses of the reactor vessel. Plates D-4802-1, D4802-2, and D-4802-3 comprise the intermediate shell course. Plates D-4812-1, D-4812-2, and D-4812-3 comprise the lower shell course. The plates and shell courses were joined together using automatic submerged arc welding using Mil B4 copper coated electrodes and Linde 1092 or Linde 124 flux. Weld seams 2-410 N C (where "NC" means seams A, 6,and C) are the axial welds between the plates to form the intermediate shell. Weld seams 3410 N C are the axial welds between the plates to form the lower shell. Weld seam 9-410 is the circumferential weld between the intermediate and lower shell course. Weld seams 2-410 N C and 9-410 were deposited usiqg the single arc process. Weld seams 3-41 0 N C were deposited using the tandem arc process.

Table 1 also p~ovidesa description of the Fort Calhoun surveillance program plate and weld material. The surveillance plate was obtained from plate 0-4802-2. The surveillance weld was fabricated using the same welding process as was used for weld seam 9-410 but with a different heat of wire.

CEN-636, Revision 02 Page 8 of 56

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK PAGE 22 OF 69 The beltline materials are evaluated using Reference 2 to identify the limiting material at end of the license period. The limiting material is the beltline plate or weld with the highest I

I RTms value. The limiting materials in the Fort Calhoun vessel beltline are from the lower shell course welds. As stated in the Introduction, the objective of this evaluation is to apply Position 2.1 of Reference 2 to surveillance data that are applicable to the limiting material, I the lower shell course welds. The results of this Position 2 1 analysis can then be used to I

calculate the adjusted reference temperature, R ,T

, at the end of the license per~od i applying the adjusted chemistry factor and the reduced standard deviation for shift from the analysis.

4.0 Description of Surve~llanceData Relevant to Fort Calhoun In Table 1, the weld wires used to fabricate the lower shell course welds (3410 AIC) in the For! Calhoun vessel were identified as heat numbers 12008, 13253, and 27204. The I

approach taken was to match up those heats or combination of heats with those used to 1 fabricate the surveillance welds in other reactor vessels manufactured by Combustion Engineering during a similar period of time.

The surveillance

- weld matches are identified in Table 2. A match is defined as having the same heat number in the surveillance weld as is in one of the welds in Table 1. In the case of a mixture of heats in the surveillance weld or For! Calhoun beltline weld, at least one of the two heats in the mixture had to match. 'The matches are based on CEOG Report CE NPSD-1119 (Reference 6) and similarly developed sources. (In all the matches cited, the traceability of the surveillance weld wire heat was established based on fabrication records as stated in Reference 6.) Data from five PWR surveillance programs (References 7 through 18) were identified as likely sources of information relative to the three heats from the Fort Calhoun weld seam 3410 AIC. Data determined to be applicable to Fort Calhoun are Mihama Unit 1, Diablo Canyon Unit 1, the weld from the Palisades supplemental surveillance program, the supplemental surveillance capsule for Fort Calhoun, Salem Unit 2, and D.C. Cook Unit 1. Data from three BWR surveillance programs were also identified using Reference 6. Only the Fitzpatrickweld was fully representative of the weld wire heats I used in weld seam 3-410 AIC. The remaining two BWR welds were either a mixture or i were representative of another weld (9410). Analysis of the Fitzpatrick surveillance weld j was not done given the limited number of measurements and the uncertainty regarding the

- CEN-636, Revision 02 Page 9 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 23 OF 69 effects of differences in irradiation environment between a BWR and the Fort Calhoun PWR vessel.

The data from four of the five PWR surveillance programs and from the Fort Calhoun surveillance program were compiled from the database assembled for the previously cited ASTM E900 effort (Reference 4). That database had been reviewed, updated and augmented by knowledgeable individuals from the Industry and, therefore, provides a credible source of information for each surveillance program. In addition the individual post-irradiat~on test reports were reviewed to the extent possible to assess the reasonableness of the data updates. The data from the Mihama Unit 1 surveillance program were obtained through a proprietary agreement between Kansai Electric Power Company and the Omaha Public Power District. [Note: Only the non-proprietary data are presented in this report.]

The surveillance program data sets are provided in Tables 3 through 6. The Fort Calhoun surveillance data (References 19 through 21) are provided in Tables 8A, 8B and 8C. Each table contains the surveillance capsule identity, the measured shift, the reported neutron fluence, and the irradiation temperature. [Note: The irradiation temperature for the surveillance specimens was taken as that of the reactor coolant cold leg. The temperatures were obtained from the E900 database and from Kansai for Mihama Unit 1.]

5.0 Reaulatow Position 2.1 Analvsis of Relevant Surveillance Data The objective of this section is to analyze the surveillance data in accordance with Position 2.1 of Reference 2. The Position 2.1 analysis will be augmented using -the guidance provided by the NRC (Reference 3). The guidance provides a set of NRC review requirements and guidelines for using relevant and credible surveillance data from other reactor vessels to refine predictions of the shift in RT, and calculation of the adjusted reference temperature, .R T

, Position 2.1 of Regulatory Guide 1.99 is applied to available surveillance data that were identified in the preceding section as relevant to the beltline welds in the Fort Calhoun vessel.

5.1 Credibility of Surveillance Data:

Regulatory Guide 1.99 presents five credibility criteria by which surveillance data P

CEN-636, Revision 02 Page 10 of 56

FORT CALHOLlN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE-USE PAGE 24 OF 69 from a given reactor are judged before the surveillance data can be used in place of Regulatory Position 1. The five criteria are discussed in turn below:

Criterion I : "Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement according to the recommendations of this guide."

The chemistry factors for each of the three beltline welds (determined using Table 1 of Reference 2) range from 89 O F to 231 OF. [Note: The highest chemistry factor for the beltline plates is less than the lowest beltline weld, 89 OF. Therefore, the beltline plates will not limit vessel operation and are excluded from the subsequent discussion.] The surveillance weld was fabricated using weld wire heat 305414 with Linde 1092 flux lots #3947 and #3951. It was made from different welding consumables than those used for the Fort Calhoun beltline welds. The-surveillance weld is representative of but not identical to the beltline welds, so it does not meet Criterion 1. Therefore, it can not be used in a Position 2.1 analysis of the Fort Calhoun beltline welds. The focus of this report is on the use of data from surveillancewelds that were fabricated using the same weld wire heats as were used in the Fort Calhoun vessel limiting beltline weld; i.e., surveillance weld data that meet Criterion 1 for the Fort Calhoun beltline welds. The surveillance program welds listed in Table 2 include most of the weld heats listed in Table 1. The one not represented at all, weld wire heat #51989, has a chemistry factor of 89 OF and thus is not a controlling beltline weld. The surveillance welds in Table 2 include the individual heats of controlling beltline weld materials and, therefore, satisfy the first criterion for the most limiting combinations of weld wire heats. -

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-foot-pound temperature and the upper-shetf energy unambiguously."

As part of the effort to review the surveillance data for the ASTM E900 effort, all of the data were computer curve fit by Modeling and Computing Services as part of an effort sponsored by the U.S. Nuclear Regulatory Commission (Reference 4). The computer curve fit results (index temperature and transition temperature shift) were CEN-636, Revision 02 Page 11 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 25 OF 69 used for the E900 effort and reported in that database. Therefore, the individual test results for the materials data applied from Table 2 exhibited behavior consistent with pressure vessel materials, scatter was well within expected ranges, and there were no difficulties experienced in deriving the 30 foot-pound temperature. The second criterion is satisfied.

Criterion 3: 'When there are two or more sets of surveillance data from one reactor, 1

the scatter of RTNDT shift values about a best-fit line drawn as described in Regulatory i I

Position 2.1 normally should be less than 28 O F for welds and 17 O F for base metal.

Even if the fluence range is large (two or more orders of magnitude), the scatter shall not exceed twice thosevalues. Even if the data fail this criterion for use in shift i calculations, they may be credible for determining decrease in upper-shetf energy if the upper shelf can be clearly determined, following the definition given in ASTM I E l85-82."

i The weld metal shift measurements for the materials were evaluated individually against this criterion in Tables 3 through 6 and in Table 8. The results of that evaluation are provided in Section 5.4. In all but one case (Cook Unit I ) , the data scatter criterion was satisfied. [The November 1997 Guidelines (Reference 3) expanded on the use of this critqion. . Those guidelines were taken into consideration in this report.]

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the claddinglbase metal interface within

-+25"F."

This criterion could not be addressed using temperature monitor data because there was an inconsistent use of monitors among the various surveillance programs.

However, both NRC guidance (Reference 3) and the NRC sponsored work (Reference 4) used the reactor coolant inlet temperatures as a best 'estimate for the irradiation temperature of the Charpy specimens in the capsule. Implicit in the NRC sponsored approach is the assumption that Criterion 4 will be met. It is based on the premise that the reactor coolant will cool the vessel wall and the adjacent I

surveillance specimens the same. In the data analysis that follows, the reactor coolant inlet temperatures from the ASTM E900 database (Reference 4) were used 1 CEN-636, Revision 02 Page 12 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK PAGE 26 OF 69 to provide an estimate of the temperature of the Charpy specimens, and the differences in irradiation temperature were treated explicitly. Thus Criterion 4 is satisfied.

Criterion 5: "The surveillance data for the correlation monitor material in the capsule 1 should fall within the scatter band of the data base for that material."

There are limited sets of correlation monitor material (termed standard reference i

material in the Fort Calhoun vessel) data from the various surveillance capsules. For Fort Calhoun, the correlation monitor material measurements were addressed in Reference 20. For the other surveillance data, no such analysis could be performed.

Therefore, the Fort Calhoun correlation monitor material measurements satisfy I

Criterion 5. I In summary, the surveillance data are shown to satisfy the criteria above. The data are assessed individually for Criteria 3 and 4 in Section 5.4, Analysis of Surveillance !

Data. The plant specific Fort Calhoun surveillance data are assessed for Criterion 5 I also in Section 5.4. Therefore, the surveillance data are acceptable for use with Position 2.1 of Regulatory Guide 1.99, Revision 2. . 1 5.2 Traceability of Mihama 1 Surveillance Data In the specific case of the IMihama Unit 1 surveillance program, foreign data from a Westinghouse designed Pressurized Water Reactor (PWR) are being applied to a domestic Combustion Engineering designed PWR. In order to establish that the weld surveillance data from the Mihama Unit 1 reador vessel are applicable to the Fort Calhoun vessel, the following information was evaluated: a. Unirradiated and irradiated Charpy data for tandem weld wire heat 12008127204; b. Irradiation temperature of the capsule based on PWR cold leg; c. Neutron flux of capsules; d.

Gamma heating of capsules; e. Neutron spectrum of capsules; and f. Chemistry of surveillance data.

Each of these items is addressed below:

a. Unirradiatedand irradiated Charpy data for tandem weld wire heat 12008127204 I

0 CEN-636, Revision 02

TDB-IX FORT CALHOUN STATION TECHNICAL DATA BOOK PAGE 27 OF 69 The individual Charpy specrmen data for the unirradiated tandem weld wire heat I2008127204 are provided in Table 2 of Reference 15. Those data were used to establish the unirradiated Charpy curve. The Individual Charpy specimen data for the irradiated tandem weld wire heat I2008127204 were obtained from Kansai (Reference 17) and were used to establish the irradiated Charpy curve. Those data were checked agalnst the Charpy index temperatures cited by Kansai in Reference 16 for the Charpy shift values from each of the three surveillance capsules (V, R and S per Reference 15) and shown to be consistent.

b. Irradiation temperature of the capsule based on PWR cold leg-Kansai reported a value of 289 OC (552 OF) for the Mihama Unit 1 cold leg temperature (Reference 16). In an evaluation of the capsule configuration (Reference 22), it has been confirmed that that temperature is reasonable for similarly configured reactor vessels designed by Westinghouse.

c. Neutron flux of capsules-I I

The neutron flux correspond~ngto each lrradrated and tested capsule from Mlhama I Unlt 1 was reported by Kansai In Reference 17 together wlth their source reference and a descrrptron of the methodology used to calculate the neutron flux In Reference 22, ~thas been confirmed that the reported flux IS reasonable for s~mrlarly configured reactor vessels deslgned by Westlnghouse

d. Gamma heating of capsules-I In Reference 22, Westinghouse has confirmed that the design and constructton of the Mihama Unit 1 surveillance capsules are the same as that for other surveillance capsules that they fabricated during this timeframe. Therefore, it is reasonable to I conclude that the gamma heating in the Mihama Unit 1 surveillance capsules is the I same as that in similar'domestic Westinghouse capsules.

CEN-636, Revision 02 Page 14 of 56

TDB-IX FORT CALHOUN STATION REFERENCE USE PAGE 28 OF 69 TECHNICAL DATA BOOK

e. Neutron spectrum of capsules-In a CEOG sponsored program (Reference 23) it was demonstrated that surveillance data applicable to Combustion Engineering fabricated reactor vessel materials were equally predictable using Regulatory Guide 1.99, Revision 2.for plants designed by both Westinghouse and Combustion Engineering. lt was concluded from this that the irradiation environment was similar for the surveillance capsules from Westinghouse and Combustion Engineering plants. There was no definiive difference between the spectra such that one needs only to consider differences in the irradiation temperature and the neutron flux. Neutron spectrum was considered to be no more than a second order variable for embrittlement. (For example, embrittlement correlation development work reported in Reference 4 did not identify neutron spectrum as an independent or dependent variable.)

In Reference 24 no discernible differences were found between the neutron spectra for the surveillance capsules from Westlnghouse and Combust~onEngineering plants Reference 22 confirmed that the Mlhama Unit 1 neutron spectrum IS comparable to domestic Westinghouse PWRs. Therefore, the neutron spectra in the Mihama Unit 1 surveillance capsules is not expected to adversely affect the

-application of those surveillance data to the Fort Calhoun vessel.

f. Chemistry of surveillance data-Kansai reported copper and nickel contents of 0.19 and 1.08 wlo for the Mihama Unit 1 surveillance weld (Reference 16). Weld analyses by Combustion Engineering and the best estimate for the weld (Reference 6) for heat 12008 and 27204 yielded copper and nickel contents as follows:

WDC-351 (nta) Cu 0.98 Ni WDC-1817 0.19 Cu 0.98 Ni Best estimate 0.219 Cu 0.996 Ni The Kansai values are fully consistent with a weld deposit made using heats 12008 and 27204. Traceability of the Mihama Unit 1 surveillance weld has been established based on fabrication records from CE-Chattanooga.

2CEN-636, Revision 02

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 29 OF 69 5.3 Analysis Approach The analysis in the following section utilizes the ratio method of Reference 2. The ratio method is based on the relative chemistry factors. Regulatory Guide 1.99 (Reference 2) states that, "if there is clear evidence" of a difference in copper and nickel content, the measured shift should be adjusted by multiplying by the ratio of the chemistry factors for the vessel weld to that of the surveillance weld (i.e., the ratio method). For this evaluation, the ratio method was used to adjust the surveillance data from other programs to the best estimate chemistry for the Fort Calhoun reactor vessel. (This was done whether or not the copper and nickel contents were significantly different.) References 5 and 6 were used to obtain best estimate copper and nickel contents for the weld wire heats so that chemistry factors could be computed for the Fort Calhoun welds.

The effect of differences in the neutron irradiation environment is considered when applying surveillance data from another reactor pressure vessel. These differences have been addressed by the Combustion Engineering Owners Group, BGE, and Duke Power (see References 23, 24, and 25, respectively). The effect of neutron irradiation environment is taken to mean changes in measured transition temperature shift caused by differences in irradiation temperature, neutron flux and neutron energy spectrum. For the BGE and Duke evaluations (References 24 and 25), there was no expected influence of neutron flux or neutron energy spectrum given the use of only PWR surveillance data. The actual values of neutron flux and neutron energy spectrum were compared for the various plants being considered, and the values were within expected ranges for which no difference in irradiation behavior would be expected. The Duke evaluation entailed the comparison of data from two Westinghouse designed reactor vessels. The BGE evaluation entailed comparisons of data from a Combustion Engineering and a Westinghouse designed reactor vessel. For the CEOG evaluation (Reference 23), a statistical analysis of surveillance data from both Combustion Engineering and Westinghouse-designed reactor vessels demonstrated that there was no significant effect of differences in the irradiation environment for vessel materials fabricated by Combustion Engineering.

In this report, data from the Combustion Engineering and Westinghouse vessel designs were considered in the analysis. Therefore, prior work suggests thatttiere is CEN-636, Revision 02 Page 16 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 30 OF 69 no significant effect of neutron flux and neutron energy spectrum expected relative to the results in Table 7.

The effect of irradiat~ontemperature was explicitly considered in the BGE evaluation (Reference 24) using the rationale stated in Reference 3. That rationale assumes there is a 1.0 "F effect on the chemistry factor for each 1.0 O F difference in irradiation temperature. (The higher the irradiation temperature, the lower the chemistry factor would be, and vice versa, per Reference 3. Irradiation temperature is taken as the reactor coolant inlet temperature.) The analysis in the following sections utilizes a modified approach from that given in Reference 3 for adjusting surveillance data for d~fferencesin irradiation temperature. A description of the rationale and benefits for the ratio and Tcddadjustments for analysis of surveillance data follows.

The rationale and benefits of this approach were described at a March 13, 2000 meeting between the NRC and the Omaha Public Power District in regard to the application of Position 2.1 of Regulatory Guide 1.99, Revision 2 to two heats of surveillance welds applicable to the Fort Calhoun vessel. The chemistry factor calculation has traditionally been done by the NRC as described in Reference 3.

However, in order to analyze surveillance data from two separate programs it was necessary to first adjust for both CF dflerences and Td differences. Two issues were considered. The first is the viability of the Tcddadjustment method. The second is the appropriateness of adjusting the data prior to performing the data scatter analysis. -

a) Viability of the TcddAdjustment Method In November 1997, the NRC presented a set of guidelines (Reference 3) t o the industry that supplemented the guidelines contained in Regulatory Guide 1.99, Revision 02. The activities surrounding Generic Letter 92-01 and its antecedents prompted the need for the supplemental guidelines.

That Generic Letter had addressed some of the material variability issues including copper and initial RT, and the effect of irradiation temperature on the degree of embrittlement. In the November 1997 NRC-Industry meeting, the NRC presented ways they considered acceptable to treat each aspect:

CEN-636, Revision 02

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 31 OF 69 The "ratio methodnwas the prescribed way to treat differences in the copper and n~ckelcontent between the surveillance program weld being analyzed and the best estimate for the vessel weld.

Theuse of the q term was the prescribed way. to treat variabilityin initial RT., A value of ai = 17 OF was assigned for use with the generic initial RTN,

= -56 OF for welds fabricated by Combustion Engineering. A value of oi= 0 OF was assigned for use with a measured initial RT, Uust as is the case for plates and consistent with the practice for welds).

Posit~on2.1 of Reference 2 was the prescribed way to analyze surveillance data to der~vea chemistry factor (CF) using two or more sets of credible data.

The data are to be adjusted for chemistry d~fferencesusing the ratio method.

If the difference between the adjusted measured shift and the predicted shift using the derived CF is less than or equal to o, = 28 OF, data scatter is deemed acceptable and the derived CF as well as a reduced o, (2812 = 14 O F ) could be used for predicting future embrittlement of the vessel beltline weld.

The effect of irradiation temperature on the degree of embrittlement was considered initially in the credibility criteria for use of surveillance data (the capsule temperature was to be within 25 OF of the vessel wall) and in November 1997 in a post-CF derivation adjustment to the CF. The initial accounting was done to satisfy the applicability issue; i.e., for irradiation temperatures between 525 OF and 590 OF, the Regulatory Guide 1.99, Revision 02 embrittlement correlation was applicable without adjustment.

The adjustment suggested in November 1997 was done to satisfy the NRC concern that the irradiation temperature of the surveillance capsule in plant "X'was at a higher temperatures than that of vessel "Y' to which the data were to be applied. It was widely believed that higher irradiation temperatures would result in less shift than at lower irradiation temperatures.

The "rule-of-thumb was that the effect was on the order of 1.0 OF increaseldecrease in shift for each 1.0 "F difference in irradiation temperature.

CEN-636, Revision 02 Page 18 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 32 OF 69 At the March 13, 2000 meeting a method was presented for making the Tcdd adjustment at the same time as was done for the ratio method. The approach followed was to use the recommended equation from NUREGICR-6551 (Reference

4) to adjust the data for the effect of irradiation temperature differences. The method used was to compute the predicted shift at both temperatures of ~nterest. The temperature effect is then the difference in the two shifts that is added to or subtracted from the measured shift, whichever is appropriate.

The equation in Reference 4 takes into consideration both time and temperature in the computation, thus providing a more rigorous treatment than that afforded by the rule-of-thumb given in Reference 3. It also offers the benefit of the numerical analysis of 609 data points for defining the apparent effect of irradiation temperature differences. (That is, the coefficients for temperature, copper, etc., were developed from the data and refined by statistical analysis.) Finally, use of the recommended equation from Reference 4 to adjust the data before the sum-of-the-squares analysis is mathematically more desirable than making the rule-of-thumb adjustment after the sum-of-the-squares analysis. (The Position 2.1 analysis approach was specifically designed to give more weight to the surveillance data at the higher fluences in recognition of the fact that the higher fluence data were more indicative of the expected behavior than were the low fluence data. Adjusting the data for temperature differences after the sum-of-the-squares analysis would not provide the same significance weighting. The Reference 3 guidelines approach, therefore, diminishes the significance of the effect of temperature on the high fluence data which is in conflict with the intent of the Position 2.1 analysis approach.)

The approach described above fully adjusts the data for both of the Reference 3 issues. Those are the chemistry differences (i.e., using the ratio method) and the Tmld differences. The shift measurements are adjusted prior to deriving the chemistry factor and prior to analyzing the scatter in the data.

b) ~ ~ ~ r o ~ r i a t e of s s Adjustment Prior to Data Scatter Analysis - The third n eData credibility criterion of Regulatory Guide 1.99, Revision 02 is to ascertain that the scatter of the surveillance measurements about a best-fit line derived using Position 2.1 is no more than 28 'F for welds. If this can be shown, then the derived chemistry factor can be used together with a reduced value for prediction uncertainty (ad2=

CEN-636, Revision 02 Page 19 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 33 OF 69 i - surveillance program greatly reduces the uncertainty of the pred~ction,and the lack I t I of significant data scatter demonstrates that the material itself is not anomalous. In i

other words, the weld material is adequately represented by the embr~ttlement 1 correlat~on contained In Regulatory Guide 1.99, Revision 02.

i I

'The applicability of the irradiation temperature adjustment depends on the source of the data. In using Position 2.1 to evaluate plant-specific surveillance data, the only i data adjustment necessary is for the chemistry difference using the ratio method (if i

t there is a significant difference between the surveillance weld and the vessel weld).

I There is no need to adjust for irradiation temperature because the capsule temperature and the cold leg temperature are essentially the same (i.e., it is the same vessel).

In using Position 2.1 to evaluate surveillance data from another plant, both the ratio method and irradiation temperature adjustments must be considered. The Reference 3 guidance is to adjust the shift measurements by the ratio method, calculate the CF,and then adjust the derived CF for temperature differences. The analysis of data scatter is done on the ratio adjusted data, so it is not examining the scatter of the original measurements. The Reference 3 approach provides a temperature adjustment but is done without regard to the time dependence of the presumed temperature effect. In using Position 2.1 to evaluate surveillance data from two other plants, both the ratio method and irradiation temperature adjustments

. must be considered, and they need to be done prior to the sum-of-the-squares a analysis. Doing the analysis on data adjusted for both the ratio method and irradiation temperature accounts for the time dependence of the presumed temperature effect and permits the sum-of-the-squares analysis emphasis on the high fluence data. Doing the analysis without both initial adjustments coupled with the subsequent correction for a temperature effect is inconsistent with the intent of Position 2.1 and places an unrealistic burden on the user to demonstrate the data scatter criterion is met.

c) Illustration of the Td Adjustment Method - The Position 2.1 analyses were run two ways as shown in Tables 4 4 4B,6A and 6B. Tables 4A and 6A give the derivation for each surveillance set of CF based on the fully adjusted numbers (i.e.,

CEN-636, Revision 02 Page 20 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 34 OF 69 for both CF and T-I, differences). Tables 48 and 66 give the derivation for each surveillance set of CF based on the numbers adjusted for CF, followed by the Reference 3 suggested approach to address, ,T differences.

For the M~hama1 surveillance data analysis, Tables 6A and 6B, the derived CFs for weld wire heats 12008 with 27204 were as follows:

CF-WF = 206.6 OF based on shifts adjusted for FCS Tmld(543 OF) and best estimate chemistry (Table 6A)

CF= 200.9 "F based on shifts adjusted for best estimate chemistry, and CF-= 209.9 OF after adjustment for FCS Td (i.e., 552 "F -

543 OF= 9°F adjustment) (Table 6B)

Therefore, in the case of the Mihama 1 surveillance data, the difference in the derived CFs is small (3.3 OF), but the CF is larger using the rule-of-thumb approach of temperature adjustment. The data scatter is -identical for each because the adjustments used were the same in each case.

- For the Diablo Canyon 1 surveillance plus the Palisades supplemental capsule data analysis, Tables 4A and 46, the derived CFs for weld wire heat 27204 (tandem) were as follows:

C F ~= 21 F 5.5 OF based on shifts adjusted for FCS Tmld(543 OF) and best estimate chemistry (Table 4A)

CF= 220.2 OF based on shills adjusted for best estimate chemistry, and CFm= 210.2"F after adjustment for FCS TcoU(i.e., 543 "F -

533 OF= 10 "F adjustment) (Table 48)

The 10°F temperature difference corresponds to the data with the highest fluence exposure because that data has the greatest significance to the CF derivation. For the weld wire heat 27204 surveillance data, the difference in the two derived CFs is small (5.3 OF), but the CF obtained using the rule-of-thumb approach of temperature adjustment is smaller than the CF derived from the fully adjusted data.

~ f i % 3 6 Revision

, 02 Page 21 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 35 OF 69 The data scatter criterion is met in the case of the CF derived using the fully adjusted data. This is justified because the analysis entails the use of data from two different vessels and three unique MT

, values. It would be unreasonable to expect I test results that are presumed sensitive to irradiation temperature to be predictable I without first removing the bias due to irradiation temperature. As was expected, the data scatter criterion was not met with the data that were corrected only for CF differences.

This method of analyzing surveillance data using both a chemistry factor and irradiation temperature adjustment is seen to result in comparable values to those obtained using the NRC guidelines in Reference 3. Use of the NRC guidel~nes resulted in a larger adjustment (positive or negative) in the two cases considered because that approach does not take into account time-at-temperature. The approach using the fully adjusted data provides the capability to analyze data irrad~atedat multiple temperatures.

5.4 ' Surveillance Data Analysis D.C. Cook Unit 1- The Cook surve~llanceweld was fabricated using weld wire heat 13253 (Reference 6). The chemistry factors for the Cook surveillance weld and the Fort Calhoun vessel weld are 206.4 "F and 189.05 OF, respectively. The Cook shift measurements in Table 3 (References 7 through 9) were adjusted for chemistry I factor differences using the ratio 189.1 OF 1206.4 OF= 0.916. The shifts were adjusted to the Fort Calhoun irradiation temperature, 543 OF, using the approach I outlined in the preceding section. The computed adjustments were -3.2 OF, -5.1 OF, I -6.1 OF, and -7.2 OF for capsule T, X, Y and U, respectively. The fully adjusted shift measurements are shown in Table 3.

1 The chemistry factor derived based on the four capsule results is 116.9 OF. The predicted shifts based on this chemistry factor were compared to the adjusted Charpy shifts. The adjusted minus predicted shifts for capsules Y and U are well in excess of o,,for welds (28 OF). The chemistry factor was re-derived based on three capsule results, where capsule U was excluded because it was the most overpredicted value. The resultant chemistry factor value based o n capsulesT, X CEN-636, Revision 02 Page 22 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 36 OF 69 and Y is 137.4 OF, which is higher than the chemistry factor value based on all four capsules. The adjusted minus predicted shifts for those three capsules are within 0, for welds (28 OF). The adjusted minus predicted shift for capsule U is greater than o,,

but is negative (i.e., conservative). Therefore, the Cook Unit 1 surveillance data are predictable when the capsule U results are excluded. The derived chemistry factor of 137.4 OF IS much lower than the values for the surveillance weld (206.4 OF) from Table 1 and for the Fort Calhoun vessel weld (189.05 OF).

Diablo Canyon Unit 1- The Diablo Canyon surveillance weld was fabricated using weld wire heat 27204 (Reference 6). The chemistry factors for the Diablo Canyon surveillance weld and the Fort Calhoun vessel weld are 221.8 OF and 226.81 OF, respectively. The analysis included the use of data for weld heat 27204 irradiated in the Palisades reactor vessel in a supplemental capsule. The chemistry factor for the Palisades supplemental surveillance weld is 229.04 OF. The ~ i a b l oCanyon (References 10 and II ) and Palisades (Reference 18) s k i measurements in Table 4 were adjusted for chemistry factor differences using the ratio 226.81°F /221.8OF=

1.022 for the Diablo Canyon data and 226.81 OF1229.04 "F = 0.990 for the Palisades data. The shifts were adjusted to the Fort Calhoun irradiation temperature, 543 OF, using the approach outlined in the preceding section. The computed adjustments were -1.6 OF, -2.0 OF, and -9.0 OF for capsules S and Y from Diablo Canyon and for capsule SA-60-1 for Palisades, respectively. The fully adjusted shift measurements are shown in Table 4A. A comparative analysis is provided in Table 48 in which the shift measurements were adjusted only for the chemistry factor differences.

The chemistry factor derived in Table 4A based on the three capsule results is 215r5 OF. The predicted shifts based on this chemistry fadar were compared to the measured Charpy shifts. The measured minus predicted shifts for the three capsules are all less than oA.The chemistry factor derived in Table 4B based on the three capsule results is 220.2 OF before adjusting for irradiation temperature differences. The adjusted chemistry factor is 210.2 OF using the guidelines of Reference 3. The predicted shifts based on the Table 48 chemistry factor were compared to the measured Charpy shifts. The measured minus predicted shift for capsule S (fluence of 2.84E18 nlcm2) is in excess of oAfor welds (28 OF), but the difference is negative (i.e., conservative). The derived chemistry factors of 215.5 and 220.2 OF are slightly lower than the values for the surveillance welds (221.8"F CEN-636, Revision 02 Page 23 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 37 OF 69 and 229.04 OF) from Table 1 and for the Fort Calhoun vessel weld (226.81 OF). The weld heat 27204 surveillance data are predictable when the data are fully adjusted to account for the differences in both chemical content and irradiation temperature.

Salem Unit 2- The Salem surveillance weld was fabricated using weld wire heat 13253 (Reference 6). The chem~stryfactors for the Salem surveillance weld and the Fort Calhoun vessel weld are 198.1 OF and 189.05 OF, respectively. The Salem shift measurements in Table 5 (References 12 through 14) were adjusted for chemistry factor differences using the ratio 189 1 OF 1198 OF= 0.955. The shifts were adjusted to the Fort Calhoun irradiation temperature, 543 OF, using the approach outlined previously. The computed adjustments were -1.7 OF, -2.2 OF, and -3.0 O F for capsules T, U, and X, respectively. The fully adjusted shift measurements are shown In Table 5.

The chemistry factor derived in Table 5 based on the three capsule results is 190.4"F. The predicted shifts based on this chemistry factor were compared to the measured Charpy shifts. The measured minus predicted shifts for the three capsules are all less than cia. The derived chemistry factor of 190.4 OF is very similar to the values for the surveillance weld (198.1 OF) from Table 1 and for the Fort Calhoun vessel weld (189.05 OF). Therefore, the Salem Unit 2 surveillance data are predictable.

Mihama Unit 1- The Mihama Unit 1 surveillance weld was fabricated using weld wire heats 12008 and 27204. The chemistry factors for the Mihama surveillance weld and the Fort Calhoun vessel weld are 227.2 OF and 231.06 OF, respectively. The Mihama shift measurements in Table 6 (Reference 16) were adjusted for chem~stry factor differences using the ratio 231.06 O F 1227.2 OF= 1.017. The shifts were adjusted to the Fort Calhoun irradiation temperature, 543 OF, using the approach outlined in the preceding section. The computed adjustmentswere +4.3 OF, +5.3 OF, and +7.4 OF for capsules 1, 2 and 3, respectively. The fully adjusted shift measurements are shown in Table 6A. A comparative analysis is provided in Table 6B in which the shift measurements were adjusted only for the chemistry factor differences.

CEN-636, Revision 02 Page 24 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 38 OF 69 The chemistry factor derived in Table 6A based on the three capsule results is 206.6 OF. The predicted shifts based on this chemistry factor.were compared to the measured Charpy shifts. The measured minus predicted shifts for the three capsules are all less than a,. The chemistry factor derived in Table 6 6 based on the threecapsule results is 200.9 OF before adjusting for irradiation temperature differences. The adjusted chemistry factor is 209.9 "F using the guidelines of Reference 3. The predicted shifts based on the Table 6B chemistry factor were compared to the measured Charpy shifts. The measured minus predicted 'shifts,for the three capsules are all less than o ~ .The derived chemistry factors of 206.6 and 209.9 O F are lower than the values for the surveillance weld (227.2 OF) from Table 1 and for the Fort Calhoun vessel weld (231.06 OF). The Mihama surveillance data are predictable when the data are fully adjusted or partially adjusted to account for the differences in both chemical content and irradiation temperature.

Fort Calhoun - The Fort Calhoun surve~llanceweld was fabr~catedusing weld wire heat 305414 (Reference 6) The chemistry factor for the Fort Calhoun surveillance weld IS 212 OF. The shift measurements in Tables 8A, 8B and 8C are from References 19 through 21). No chemistry factor adjustment was made because the 1 data are not being related to any vessel weld. The data are being used only to assess predictability of the Fort Calhoun surve~llance weld data.

'The chemistry factor derived in Table 8A based on the three capsule results is 229.0 OF. The predicted shifts based on this chemistry factor were compared to the measured Charpy shifts. The measured minus predicted shifts for the three capsules are all less than o,. Therefore, the Fort Calhoun weld surveillance data are predictable. The derived chemistry factor of 229.0 OF is higher than the value for the surveillance weld (212 OF) in Table 1.

The Fort Calhoun surveillance plate was fabricated using heat A1768-1. The chemistry factor for the Fort Calhoun plate is 65 OF based on Table 2 of Reference 2). No chemistry factor adjustment was made because there is no difference between the surveillance plate and the vessel plate chemistry. The data are being used to assess the predictability of the Fort Calhoun surveillance plate data.

CEN-636, Revision 02 Page 25 of 56

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 39 OF 69 if The chemistry factor derived in Table 86 for the surveillance plate based on the three capsule results (where the longitudinal and transverse measurements were

! combined) is 72.0 OF. The predicted shifts based on this chemistry factor were i compared to the measured Charpy sh~fts.The measured mlnus predicted shifts for the five measurements are all less than oA.Therefore, the Fort Calhoun plate I surveillance data are predictable. The der~vedchem~stryfactor of 72 0 OF is similar to the Table 2 value (65 OF).

f The standard reference material in the Fort Calhoun surveillance program was from HSST Plate 01. The chemistry factor for the plate is 131.7 OF using the reported chemical content from the E900 database with Table 2 of Reference 2. No chemistry factor adjustment was made because there is no corresponding vessel plate chemistry. The data are being used to assess the predictability of the Fort Calhoun standard reference material data.

The chemistry factor derived in Table 8C for the standard reference material based on the two capsule results is 1'38.3 OF. The predicted shifts based on this chemistry factor were compared to the measured Charpy shifts. [Note: This exceeds the requirements of Regulatory Guide 1.99, Revision 2, Criterion 5 in which it is necessary only to show the data are within the scatterband of available measurements.] The measured minus predicted shifts for the two measurements are both less than o ~ .The derived chemistry factor of 138.3 OF is similar to the Table 2 value (131.7 OF). Therefore, the Fort Calhoun standard reference material data are predictable.

6.0 Evaluation of Surveillance Data Credibilitv and Applicabilitv to Fort Calhoun The results of the preceding analysis are summarized in Tables 7 and 9. The derived chemistry factors are provided in Table 7 for each of the surveillance program welds that are applicable to the Fort Calhoun beltline welds. The derived values correspond to the best estimate chemistry for the weld wire heat(s) used to fabricate the surveillance welds. The ratio method was applied to adjust the chemistry of the specific surveillance program weld to the best'estimate chemistry for the vessel weld. Also shown in Table 7 are the chemistry factors obtained using Table 1 o f Reference 2 for the surveillance weld and the best estimate chemistry for the weld wire heat.

I CEN-636, Revision 02 Page 26 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 40 OF 69 All of the surveillance materials analyzed in Tables 3 through 6 are cred~blewith respect to being applicable to the llmiting materials in the Fort Calhoun reactor vessel beltline. This applicability is with respect to weld wire heat number, welding flux type, and welding process. Any differences in copper and nickel content between a surveillance weld and the Fort Calhoun reactor vessel beltline weld with the same weld wire heat(s) were addressed through use of the ratlo method in accordance with Reference 2. Any difference in irradiation temperature between the surveillance weld and the Fort Calhoun reactor vessel beltline weld was addressed through use of the Tcouadjustment method described in Section 5.3. The data were evaluated for scatter using the criterion that the surveillance measurements were to be predictable within one a, of the predicted shift using the derived chemistry factor in accordance with Reference 2.

In the case of heat 23253 from D.C. Cook Unit 1, Table 3, there are measurements from four surveillance capsules. The high fluence measurement, capsule U, is significantly overpredicted. The derived chemistry factor based on capsules T, X, and Y from D.C.

Cook Unit 1 is 137.4 OF. In the case of heat 13253 from Salem Unit 2, Table 6, all three measurements are predictable within one ( 3 but ~ the derived chemistry factor (190.4 OF) is higher than obtained from the D.C. Cook Unit 1 data (137.4 OF). Therefore, a conservative chemistry factor adjusted for the Fort Calhoun weld irradiation temperature and chemical c o n k t and made with heat 13253 is 190.4 OF. It is based on the fully credible surveillance data from Salem Unit 2. The derived chemistry factor and the vessel weld bestestimate chemistry factor from Table 1 of Regulatory Guide 1.99, Revision 2 are very similar (190.4 OF and 189.1 OF, respectively).

In the case of heat 12008 and 27204 from Mihama Unit 1 (Table 6A), all three surveillance measurements are predictable within one o ~ The . derived chemistry fador is 206.6 OF and includes adjustments for differences in irradiation temperature and chemical content between the Mihama Unit 1 surveillance weld and the Fort Calhoun beltline weld. It is based on the fully credible data from Mihama Unit I.The derived chemistry factor, 206.6 O F is less than the vessel weld bestestimate chemistry factor, 231.06 OF from Table 1 of Reference 2.

In the case of heat 27204 (tandem) from Diablo Canyon Unit 1 and the Palisades supplemental capsule (Table 4A), all three surveillance measurements are predictable within one (3,. The derived chemistry factor is 215.5 OF and includes adjustments to the CEN-636, Revision 02

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FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 42 OF 69 the predicted chemistry factors. In other words, the surveillance weld data that correspond to the weld wire heats used in the Fort Calhoun beltline welds are conservatively predicted. There is .no immediate explanat~on available for the observation that the Fort Calhoun surveillance weld material (i.e., heat #305414) data were underpredicted by Reference 2, whereas the 0.75% and 1.00% nickel specification heats were conservatively predicted. There are no Fort Calhoun beltline welds with a 0.60% nickel content. Therefore, this issue is not applicable.

The data in Table 7 encompass three of the five most limiting weld wire heat combinations used in the Fort Calhoun reactor vessel beltline. The surveillance data coverage by weld seam is as follows:

Welds 3410 NC: D.C. Cook 1 heat 13253, Diablo Canyon 1 heat 27204, Palisades supplemental capsule heat 27204, and Salem 1 heat 13253.

Weld 9 4 10: No applicable data. [Note: The chemistry factor associated with the best estimate copper and nickel content for heat 20291 is 188.41 OF.

This weld is unlikely to be limiting because it is a circumferential weld for which the PTS screening criterion is 300 OF.]

Welds 2410 NC: No applicable data. [Note: The chemistry factor associated with the best estimate copper and nickel content for heat 51989 is 89.03 OF.

These welds will not become limiting for the Fort Calhoun vessel.]

Position 2.1 of Reference 2 allows one to use credible surveillance data to determine the adjusted reference temperature.

~~. his

. . ~ . is done by deriving a value for the chemistry factor (CF). If the 'data scatter is within prescribed limits, then the derived CF may be used with half the normal value for oato calculate the adjusted reference temperature. Based on the preceding, there are credible surveillance data for three of the limiting heats used in the Fort Calhoun reactor vessel beltline. For each surveillance weld, a chemistry factor was derived using the ratio method together with an adjustment for irradiation temperature. As shown in Table 7, the derived chemistry factors obtained were less than or equal to the value obtainable from Table 1 of Reference 2. Position 2.1 states that "if this procedure gives a higher value of adjusted reference temperature than that given by using the procedures of Regulatory Position 1.1 (i.e., Table 1 of Reference 2), the surveillance data CEN-636, Revision 02 Page 29 of 56

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 43 OF 69 I should be used. If this procedure gives a lower value, either may be used." Given the i availability of credible surveillance data that show the Regulatory Position 1.1,chemistry factors to be conservative, those chemistry factors may be used. In the calculation of the margin, If the data 'scatter is within prescribed limits one may use half the normal value for I

o, when determining the adjusted reference temperature.

7.0 Calculation of RT, The limiting beltline material for the Fort Calhoun vessel is that from the lower shell axial I I

welds, 3-410 N C . The preceding analysis has demonstrated that there are credible surveillance data available for three of the four most limiting weld wire heat combinations used to fabricate those axial welds. These three sets of credible data pertain to each of the 1

heats used for the lower shell axial welds, although not for each possible combination of heats. Given the availability of credible and predictable surveillance data,for the three weld I

wire heat combinations, it is justified to use the derived CF and to use half the normal value for o~to calculate the margin when determining the adjusted reference temperature. For the one weld wire heat combination for which surveillance data are not yet available, the CF from Table 1 of Reference 2 and the normal value for o~will be used to calculate the adjusted reference temperature, R .T 1

Provided below is the determination of the RT, for the limiting beltline materials predicted for the end of the current license for Fort Calhoun (August 9, 2013). The neutron fluence was conservatively determined to be 1.728 x10T9nlcm2 (E>'IMev) for that date using an unbiased estimate (see Reference 26). This was projected out to the end of a renewed license period, August 9, 2033, using the same unbiased estimate. (The projected.value actually corresponds to the end of that fuel cycle, March 2034 and, therefore, contains an i

! added conservatism.) The projected neutron fluence value is 2.431 XIO'~ dm2(E>1Mev)

I (Reference 26). The fluence was calculated in a manner consistent with the methods of the U.S. Nuclear Regulatory Commission's Draft Regulatory Guide DG-1053 (Reference 27).

i The RT, calculation was performed as follows:

I.

RT, = Initial R,T, + Shift + Margin CEN-636, Revision 02 Page 30 of 56 I

I -

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 44 OF 69 1 Following are the calculations for each of the three heats combinations for which credible 1 and predictable surveillance data are available and for the fourth limiting heat combination for which surveillance data are not yet available.

a. Heat 13253 Initial RTN, = - 56 OF (generic value for CE welds)

Shift = Chemistry Factor X Fluence Factor Chemistry Factor (CF) = 190.4 OF (based on Salem 2 surveillance data)

Fluence factor (FF) is a function of neutron fluence, f, in units of 1x10'~n/cm2

!i ~~,+28-01~1opf) f Margin = 2(02 + =:)IR oa= 28 OF12 = 14 OF (half the value for welds) oi= 17 OF (for generic CE welds) 2(o~+a~)1n=2(170F2+140F2)'R=44.00F RTprs= - 56 OF + 190.4 OF X f(.28-0.' X10g4 + 44.0 "F For the end of the current license for Fort Calhoun (August 9,2013), the RT, is:

For the end of the renewed license period for Fort Calhoun (August 9, 2033), the RT, is:

These projected values are less than the PTS screening criterion value of 270 OF for axial welds. Thus the vessel weld will remain below the PTS screening criterion for a period exceeding 20 years beyond the current 40 year license term.

CEN-636, Revision 02 Page 31 of 56

FORT CALHOUN STA1-ION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 45 OF 69

b. Heat 12008 and 27204 ln~tialRT,, = - 56 OF (generic value for CE welds) [Note: A measured value of initial RT,

= -58 OF is available for this weld. For purposes of this calculation the more conservative generic value and its associated margin was used.]

i Shift = Chemistry Factor X Fluence Factor Chemistry Factor (CF) = 206.6 "F (based on Miharna 1 surveillance data) o Fluence factor (FF) is a function of neutron fluence, f, in units of 1x10 nlcm2 FF= 6 2 8 - 0 . 1 x b g 0 Margin = 2(az + a"):'

a, = 28 OF12 = 14 OF (half the value for welds) ai = 17 OF (for generic CE welds) 0 2(a: +a:)"= 2(17 OF' + 14 " ~ ~ ) " = 4 4 . "F 0

For the end of the current license for Fort Calhoun (August 9, 2013), the RTmsis:

For the end of the renewed license period for Fort Calhoun (August 9,2033), the RT, is:

i RTm = - 56 OF + 256.0 OF + 44.0 OF = 244 OF I

I These projected values are less than the PTS screening criterion value of 270 OF for axial t welds. Thus the vessel weld will remain below the PTS screening criter~onfor a period

.II exceeding 20 years beyond the current 40 year license term.

c Heat 27204 Initial RT, = - 56 OF (generic value for CE welds)

Shift = Chemistry Factor X Fluence Factor

-- CEN-636, Revision 02 Page 32 of 56

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 46 OF 69 0 Chemistry Factor (CF) = 215.5 OF (based on Diablo Canyon 1 and Palisades !

surveillance data)

Fluence factor (FF) is a function of neutron fluence, f, in units of 1x10" nlcrn2 FF, f(28-01xbgO Margin = 2(0? + 02)

OA = 28 OF12 = 14 OF (half the value for welds) o,= 17 OF (for generic CE welds) 2((3,2+o~)*=2(17°F2+140F2)1R=44.~0F RT, = - 56 OF + 215.5 O F ~ f ( ~ +~44.0 - OF~ . ~ ~ ' ~ ~

For the end of the current license for Fort Calhoun (August 9, 2013), the RT, is:

RT, = - 56 OF + 247.9 OF + 44.0 OF = 236 OF For the end of the renewed license period for Fort Calhoun-(August .. .

9, 2033), the RT, is:

RTms = - 56 OF + 267.0 "F + 44.0 OF = 255 OF These projected values are less than the PTS screening criterion value of 270 OF for axial welds. Thus the vessel weld will remain below the PTS screening criterion for a period exceeding 20 years beyond the current 40 year license term i d. Heat 12008 and 13253 Initial RTm = - 56 OF (generic value for CE welds)

Shift = Chemistry Factor X Fluence Factor Chemistry Factor (CF) = 208.68 OF (from Table 1, Reference 2 for weld heats 12008 and 13253)

Fluence factor (FF) is a function of neutron fluence, f, in units of 1x10 nlcm2

~~,f(.ZB-O.lxbg9 i

1 CEN-636, Revision 02 Page 33 of 56

TDB-IX FORT CALHOUN STATION TECHNICAL DATA BOOK REFERENCEUSE PAGE 47 OF 69 Margin = 2 ( d + 02)"

OA = 28 OF (value for welds) 4 = 17 OF (for generic CE welds) 2 ( d + 02)'"= 2(17 OF' + 28 OF') In= 65.5 OF For the end of the current license for Fort Calhoun (August 9, 2013), the RTmSis:

For the end of the renewed license period for Fort Calhoun (August 9, 2033), the RT ,, is:

These projected values are less than the PTS screening criterion value of 270 OF for axial welds. Thus the vessel weld will remain below the PTS screening criterion for a period exceeding 20 years beyond the current 40 year license term.

e. Plate Code D4802-2 (Heat A1768-11 Initial RTNDT = 18 OF (measured value)

Shift = Chemistry Factor X Fluence Factor Chemistry Factor (CF) = 72.0 OF (based on Fort Calhoun surveillance data)

. Fluence factor (FF) is a function of neutron fluence, f, in units of 1x10'~n/crn2 FF= $.28 - 0.1 x log f) where f= 2 . 4 5 ~ 1 0 'n/crn2

~

renewed license period, respectively (Reference 26).

and 3 . 4 5 ~ 1 0 'n/crn2

~ for the current and Margin = 2(0? + 0h2)'~ i ob = 17 OF12 = 8.5 OF (half the value for plates)

O, = 0 OF (for measured value) 2(42 + 02)" = 2(0 OF' + 8.5 OF^) In= 17.0 OF 1

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 48 OF 69 For the end of the current license for Fort Calhoun (August 9, 2013), the RT, is:

I For the end of the renewed license period for Fort Calhoun (August 9,2033), the RT s, is:

i RTprs= 18 "F + 95.3 "F + 17.0 OF = 130 "F I These projected values are less than the PTS screening criterion value of 270 OF for plates.

Thus the vessel plate will remain below the PTS screening criterion for a period exceeding 20 years beyond the current 40 year license term.

8.0 Conclusions i

1) The Fort Calhoun surveillance program data are credible and predictable as summarized in Table 9.

2) There are four.sets of credible surveillance weld data available from other plants that are applicable to the Fort Calhoun reactor vessel beltline welds. The derived chemistry factor given in Table 7 for each set was less than or equal to the value obtainable from Table 1 of Regulatory Guide 1.99. ~-

3) Given the availability of credible and predictable surveillance weld data, it is justi6ed to use half the normal value for oh to calculate the margin when determining the adjusted reference temperature for the Fort Calhoun vessel beltline materials.

4) The highest projected value of RTmsis 250 OF at the end of the current license. This was determined using the normal value for a, (28 O F ) and the limiting material chemistry factor of 208.68 OF from Table 1 of Regulatory Guide 1.99, Revision 02. It corresponds to weld wire heats 12008 and 13253 for Fort Calhoun weld 3-410 NC.

The highest projected value of RTmsat the end of the renewed license term is 268 OF for that same weld material as shown in Table 10. These projected values are less than the PTS screening criterion value of 270 OF for plates and axial welds and CEN-636 Page 35 of56

TDB-IX FORT CALHOLIN STATION PAGE 49 OF 69 TECHNICAL DATA BOOK Thus the vessel plates and welds will remain below the PTS screening criterion for a period exceeding 20 years beyond the current 40 year license term.

Ii

5) In the analysis of the surveillance data, the data were adjusted for both differences I in copper and nickel content and for differences in irradiation temperature. It was necessitated by the fact that the data available for one of the heats was from two I different reactor vessel surveillance programs that in turn had to be adjusted for the Fort Calhoun vessel. The irradiation temperature adjustment method was based on the use of NUREGICR-6551 (Reference 4). In the M o cases evaluated, the 1 adjustment method resulted in a derived chemistry factor that was comparable to that obtained using guidelines (Reference 3) developed previously. The proposed method with its dual adjustments was successfully used to reconcile surveillance I

data from two different-plants. i References iI 1. 10CRF50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", Federal Register, Vol. 60, No. 243, December I

2. US Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 02, 1!

"Radiation Ernbrittlernent of Reactor Vessel Materials", May 1988.

3. "Evaluation and Use of Surveillance Data", Handout from NRC-Industry Meeting on Status of Generic Letter 92-01, Supplement 1, Rockville, MD, November 12, 1997.

4. E.D. Eason, et al., "Improved Embrittlement Correlations for Reactor Pressure Vessel Steels", NUREGJCR-6551, dated November 1998.

5. "Response to Request for Additional Information Related to Generic Letter 92-01, Revision 1, Supplement 1", OPPD Letter LIC-980124, dated September 28, 1998.

6. "Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Contenr, CEOG Report CE NPSD-1119, Revision 1, dated July 1998.

0CEN-636, Revision 02

FORT CALHOLIN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 50 OF 69 1 7. D.C. Cook Unit 1, Capsule T, SWRI-024770 i 8. D.C. Cook Unit 1, Capsule X, SWRI-02-6159 1 9 D.C. Cook Unit 1, Capsule Y, SWRI-06-7244-001

10. "Analysis of Capsule S from the PGE Diablo Canyon 1 Reactor Vessel Radiation I

Surveillance Program", December 1987, WCAP-11567 1

11. "Analysis of Capsule Y from the PGE Diablo Canyon 1 Reactor Vessel Radiation Surveillance Program", July 1993, WCAP-13750. 1

12. "Analysis of Capsule T from the Public Service Electr~c& Gas Company Salem 2 Reactor Vessel Radrat~on Surveillance Program," March 1984, WCAP-10492 1

13 "Analysis of Capsule U from the Publlc Servrce Electric & Gas Company Salem 2 I Reactor Vessel Rad~at~on Surverllance Program," September 1987, WCAP-11554

14. "Analysis of Capsule X from the Public Service Electric & Gas Company Salem 2 Reactor Vessel Radiation Surveillance Program," June 1992, WCAP-13366.

I I

I

15. S.E. Yanichko, "Kansai Electric Power Co., Mihama Unit No. 1 Reactor Vessel Radiation Surveillance Program", Westrnghouse Report WCAP-7374, January I

16. Yasunobu Nashida, Kansai Electric Power Co., to J.K Gasper, Omaha Public Power District, "Mihama Unit No. 1 Reactor Vessel Material Information", dated December 7,1999.

17. Katsuhiko Shigemune, Kansai Electric Power Co., to J.K Gasper, Omaha Public Power District, "Reactor Vessel Data of Mihama Unit I " , dated April 17,2000.

18. Personal telephone communication, J.R. Kneeland, Consumers Energy, January 7, 2000; and T.C. Hardin letter to J.R. Kneeland, 'CVGFWPH Analysis of Charpy Energy Data from Capsule SA-60-I", dated August 9, 1999.

~2 CEN-636, Revision 02 . -

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 51 OF 69

19. "OPPD Fort Calhoun Station, Evaluation of lrradiated Capsule W-225, August 1980, TR-0-MCM401 , Revision I .

20. "OPPD Fort Calhoun Station, Evaluation of Irradiated Capsule W-265, March 1984, TR-0-MCM402.

21. "OPPD Fort Calhoun Station, Evaluation of Irradiated Capsule W-275, November 1994, BAW-2226.

22. S.L. Anderson, "Mihama Unit 1 Irradiation Environment", Westinghouse Report LTR-REA-00618, June 22,2000.

23. "Application of Reactor Vessel Surveillance Data for Embrittlement Management",

Combustion Engineering Owners Group Report CEN-405-P, Revision 3, September 1996.

24. Robert E. Denton, Baltimore Gas and Electric Company, "Request for Approval of Updated Values of Pressurized Thermal Shock (PTS) Reference Temperatures (RT), Values (10CFR50.61)", letter dated July 21, 1995.

25. "Duke Power Company, Evaluation of McGuire Unit 1, Surveillance Weld Data Credibility", Technical Report No. Al-I-98412-T005, revision 1, November 1998, transmitted by Duke Energy Corporation letter, H.B. Barron to U.S. Nuclear Regulatory Commission, "Reactor Vessel Radiation Surveillance Program", dated January 7,1999.

26. S.L. Anderson, "Fast Neutron Fluence Evaluations for the Fort Calhoun Unit 1 Reactor Pressure Vessel", Westinghouse Report WCAP-15443, July 2000.

27. U.S. Nuclear Regulatory Commission's Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure vessel Neutron Fluence".

CEN-636, Revision 02 Page 38 of 56

TDB-IX FORT CALHOUN STATION REFERENCE USE PAGE 52 OF 69 TECHNICAL DATA BOOK Table 1 Identification of Reactor Vessel,Plates and Welds in the Fort Calhoun Reactor Vessel Beltline Plate or Weld Plate or Weld Weld Flux Type and Chemistry Identification Electrode Heat No. Lot No. Factor ("F)"

Plate D4802-1 C2585-3 NIA 82.2 Plate D4802-2 A1 768-1 NIA 65 Plate 04802-3 A1768-2 NIA 73.1 1 Plate D4812-1 1 C3213-2 1 NIA I 8 3 1 I Plate D4812-2 1 C3143-2 1 NIA I I

65 I I I I I Plate D4812-3 C3143-3 NIA 65 I Surveillance Plate D4802-2 2-41 0 N C 51989 NIA Linde 124, #3687 1 72.0c 89.03 1

3-410 N C 12008 & 13253 ( T ) ~ Linde 1092, #3774 208.68 3 4 10 N C 13253 ( T ) ~ Linde 1092, #3774 189 05 3 4 10 N C 12008 & 27204 ( T ) ~ Linde 1092,#3774 231.06 e 3 4 10 N C 6410 Surveillance Weld 27204 ( T ) ~

20291 305414 Linde 1092, #3774 Linde 1092, #3833 Linde 1092, #3947 226.81 188.41 212 and #3951 Notes:

a) Chemistry Factor from Table 1 or 2 of Reference 2.

b) "T" denotes a tandem arc weld; other welds are single arc.

c) Chemistry Factor as derived based using surveillance measurements in Table 8B of this report.

0 CEN-636, Revision 02

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 53 OF 69 Table 2 Identificationof Reactor Vessel Surveillance Program Welds Applicable to the Fort Calhoun Vessel Beltline Welds Linde 1092 #3714 These are not fully applicable to the Fort Calhoun vessel limiting beltline welds.

CEN-636, Revision 02

Table 3 Test Results from the D.C. Cook Unit 1 Reactor Vessel Surveillance Program (SurveillanceWeld Wire Heat No. 13253)

" Capsule

-" +.

Idebtity _:

- -- 2: 4 C h a ~ Qhiff,

> -- y "F . Adjusted'"'Charpy Neutron Fluence, Irradiation Temperature, OF Shift, "F -

A

>, ipxi z --,$

4nza*

" " " 6 A

,"* nlcm2 ,

5 A

_ * - I ,

T 70 60.9 2.69E18 537 X 146 128.7 8.13E18 537' Y 184 162.5 123E19 537 u a 09 92.6 1.77~19 537

'not reported; assumed to be same as other reportedvalues Capsule Identity Adjusted'='Charpy (FF) x Adjusted Fluence I (w2 -

Adjusted predictedb Shift, "F Shift Factor (FF) Shift, "F (a) Shift adjusted for FCS Tmld(543OF) and best estimate chemistry '

(b) Predicted using CFw,oU,= 137.4"F CEN-636, Revision 02 Page 41 of 56

I Table 4A Test Results from Diablo Canyon Unit I and Supplemental Capsule with Tcoldand CF Pre-Adjustment for Weld Heat 27204 C =577.6 CF=577.612.6808= 215.5 O F (a) Shift adjusted for FCS Tmld(543 O F ) and best estimate chemistry CEN-636, Revision 02 Page 42 of 56

Table 4B Test Results from Diablo Canyon Unit 1 and Supplemental Capsule with Separate Adjustment for Tcold and CF for Weld Heat 27204 3k DCn q-0 iio Z

DC1-S 115.5 75.8 ,6562 ,4306 115-14 4 ~ - 2 9 DC1-Y 238.1 234.0 ,9830 ,9662 238-216= 22 SA-60-1 247.5 280.4 1.1331 1.2840 247-249= -2 C =590 2 C =2.6808

, CF=590.2/2.6808=220.2 O F

~ ~ =OF + (533 OF - 543 OF) = 210.2 C F T ~ 220.2 O F

\

(a) Shifl adjusted for best estimate chemistry CEN-636, Revision 02 Page 43 of 56

Table 5 Test Results from the Salem Unit 2 Reactor Vessel Surveillance Program (Surveillance Weld Wire Heat No. 13253)

- Capsule Identity Charpy Shift, "F ~djusted'"Charpy Neutron Fiuence, Irradiation Temperature, "F z 1 I 1

Shift, O F ntcm2 T 145 136 8 2.75E18 539 U 180 169.7 5.50E18 539 X 188 176 6 1.07E19 539 (a) Shift adjusted for FCS Tc& (543 O F ) and best estimate chemistry

-0 CEN-636, Rewsion 02 Page 44 of 56 9 0

m I cn

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 58 OF 69

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 59 OF 69

I Table 7 Derived Chemistry Factors for Reactor Vessel Surveillance Program Welds Applicable to Fort Calhoun Vessel Weld 3-410 a) Adjusted to Best Estimate CF and T, for Fort Calhoun (543 O F ) ; value in parentheses was determined by adjusting for Tcddafter deriving chemistry factor.

b) Chemistry Factor (CF) from Table 1 of Reference 2 based on the copper and nickel content for the surveillance weld.

C) Chemistry Factor (CF) from able 1 of Reference 2 based on the best estimate copper and nickel content for the weld wire heat or combination of heats.

CEN-636, Revision 02 Page 4 7 of 56

Table 8A Test Results from the Fort Calhoun Reactor Vessel Surveillance Program (Surveillance Weld Wire Heat No. 305414)

Capsule Identity> Charpy Shift, O F Neutron Fluence, Irradiation Temperature, O F W225 210 5.53E18 527 W265 225 7 71E18 534 W275 219 1.28E19 538 C =2.6975 CEN-636, Revision 02 Page 48 of 56

+

m07-l 9

z0 q

0 9 Table 8B

?Z Test Results from the Fort Calhoun Reactor Vessel Surveillance Program (Surveillance Plate Heat No. A1768-1)

Capsule Identity Charpy Shift, OF Neutron Fluence, (Lg,TrIa n/cmz W225 60,NIA 5.53E18 527 W265 74,70 7.71E18 534 I W275 73,72 1.28E19 a) "Lg"is longitudinaland "TI" is for transverse orientation Charpy data 538 l

Capsule Identity Charpy Shift, "F (FF) x Shift Fluence (FF~ -

Measured Predicted (Lg,Tr) Factor (FF) Shift, :F' CEN-636, Revision 02 Page 49 of 56

Table 8C Test Results from the Fort Calhoun Reactor Vessel Surveillance Program (Standard Reference Material)

Capsule Identity Charpy Shift, OF Neutron Fluence, Irradiation Temperature, OF

, - $.- n/cm2 I'shift perW275 141' Surveillance Program test report 1.28E19 538 I

Capsule Identity' ' Charpy&Shift,

~,

OF (FF) x Shift Fluence (FV2 -

Measured Predicted

),

Factor (FF) Shift, O F CEN-636, Revision 02 Page 50 of 56

Table 9 Derived chemistry Factors for Fort Calhoun Reactor Vessel Surveillance Materials CEN-636, Revision 02 Page 51 of 56

Table 10 Predicted RT, for the Fort Calhoun Reactor Vessel Beltline Plates and Welds Notes:

l a) Chemistry Factor from Table 1 or 2 of Reference 2 or derived using surveillance measurements in this report.

b) Chemistry Factor derived using surveillance measurements in this report.

c) Prediction based on fluence of 2 . 4 3 ~ 1 0 n/cm2

'~ for axial welds and 3 . 4 5 ~ 1 0dcm2

'~ for plates and weld 9-410.

I

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCEUSE PAGE 66 OF 69

r I

Figure 2 Effect of Tcold on SRM Data HSST Plate 01 Results (CF=130.3 F)

-7 I 520 525 530 535 540 545 550 555 Tcold, F

Table A1 Standard Reference Material Data from Combustion Engineering Designed Surveillance Capsules

Shift per surveillance report CEN-636, Revision 02 Page 55 of 56

. . -. - ---.-.___ _ _ ~. .- -. . ..- ;;I; 92

_-r ,

--7'---.- ..-.I..-----.- _____-____

~

zO oD Table A2 Analysis of Standard Reference Materials FZ 00 55 DCn m -l gi i 5o Z

Shift per surveillance report

/FFl x Shift m2 C=l379.43 C=l0.5882 CF=(1379.43)1(10.5882)=130.3OF CEN-636, Revision 02 Page 56 of 56 1

I!

-.... . . .. --- -

LIC-08-0071, Transmittal of Fort Calhoun, Unit 1, Reactor Coolant System Pressure and Temperature Limits Report, Revision 4

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