ML20086H472
| ML20086H472 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/30/1991 |
| From: | Devan M, Lowe A, Yoon K BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20086H467 | List: |
| References | |
| B-13971, BAW-2142, NUDOCS 9112090176 | |
| Download: ML20086H472 (144) | |
Text
{{#Wiki_filter:. ~ Docket No. 50-3 M B13971 Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Reactor Vessel Material Irradiation Surveillance Capsule W-104 November 1991 ?1120'90176 911127 {.'UR ALOCK 0G000336 PDR
l .. , BAW-214? l November 1991 I NORTHEAS kJCLEAR ElERGY COMPANY HILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 f
-- Roactor Vessel Material Surveillance Program --
1 I l > I l l l I hh. SERT / ICE COMPAIUY I a
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.I BAW-2142 November 1991 I
-I I ANALYSIS OF CAPSULE W-104 NORTHEAST NUCLEAR ENERGY COMPANY HILLSTONE NUCLEAR POWER STATION, UNIT NO. 2
-- Reactor Vessel Material Surveillance Program --
I by , E A. L. Lowe, Jr., PE E J. D. Aadland 4 A. D. Nana M. A. Rutherford W. R. Stagg 5 '
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'I I I B&W Document No. 77-2142-00 (See Section 11 for dacument signatures) B&W NUCLEAR SERVICE COMPANY Engineering and Plant Services Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 BW!!na?a%r I J
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SUMMARY
This report describes the results of the examination of the second capsule (Capsule W-104) of the Northeast Nuclear Energy Company, Millstone Nuclear Power Station, Unit No. 2 reactor vessel surveillance program and analysis of a supplemental dosimetry capsule. The objective of the program is to monitor the l effects of neutron irradiation on the tensile and iracture toughness properties of the reactor vessel materials by the testing and evaluation cf tension and Charpy impact specimens. The program was designed in accordance with the requirements of ASTM Specification E185-70. E 5 The capsule received an average fast fluence of 8.84 x 10 I8 n/cm' (E > 1.0 MeV) and the predicted fast fluence for the reacter vessel T/4 location at the end of the tenth cycle is 4.88 x 10 I6 n/cm2 (E > 1 HeV). Based on the calculated fast flux at the vessel wall, an 80% load factor, and the planned fuel management, the projected fast fluence that the Millstone Nuclear Power Station, Unit No. 2 reactor pressure vessel inside surface will receive in 40 calendar years of operation is 2.40 x 10 I9 n/cm' (E > 1 HeV) and the corresponding T/4 fluence is calculated to be 1.27 x 10 19 n/cm' (E > 1 MeV). The peak calculated RT NDT atT/4 vessel wall location is 156F at EOL per Regulatory Guide 1.99, Rev. 2 but appears 5 to be approximately 147F based on data from the surveillance capsules. Likewise, 3 the T/4 vessel wall upper-shelf energy is calculated to decrease to 55 ft-lbs per. E Regulatory Guide 1.99, Rev. 2 but based on data from this surveillance capsule, none of the beltline region materials should decrease below the 75 ft-lbs required by 10CFR50, Appendix G, prior to EOL. The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure. The Charpy impact data results exhibited the characteristic shift to higher temperature for the 30 ft-lb transition temperature and a decrease in upper-shelf energy. These results demonstrated that the current techniques used for predicting the change in both the increase in the RT NDT and the decrease in upper-shelf properties due to irradiation are conservative.
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I I I I CONTENTS l 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . Page 1-1
- 2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . . . . . . 3-1
- 4. PRE- I RRADI AT I ON TESTS . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Tension Tests . . . . . . . . . . . . . . . . . . . . . . . . 4-1
,I 4.2. Impact Tests . . . . . . . . . . . . . . . . . . . . . . . . 4-1
- 5. POST-IRRADIATION TESTING . . . . . . . . . . . . . . . . . . . . . 5-1 l 5.1. Visual Examination and Inventory . . . . . . . . . . . . . . . 5-1 5.2. Thermal Monitors . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3. Tension Test Results . . . . . . . . . . . . . . . . . . . . 5-1 5.4. Charpy V-Notch Impact Test Results . . . . . . . . . . . . . . . 5-2
- 6. NEUTRON FLUENCE . , , . . . . . . . . . . . . . . . . . . . . . . 5-1 6.1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2. Vessel Fluence . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.3. Capsul e Fl uence . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.4.
Flucace Uncertainties . . . . . . . . . . . . . . . . . . . . 6-6
- 7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . . . . . . . 7-1 I 7.1.
7.2. Pre-Irradiation Property Data . . . . . . . . . . . . . . . . Irradiated Property Data . . . . . . . . . . . . . . . . . . 7.2.1. Tensile Properties . . . . . . . . . . . . . . . . . 7-1 7-1
. 7-1 7.2.2. Impact Properties . . . . . . . . . . . . . . . . . . 7-2 I 7.3.
7.4. Reactor Vessel Fracture Toughness . . . . . . . . . . . . . . Neutron Fluence Analysis . . . . . . . . . . . . . . . . . . 7-4 7-5 I 8. DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY PRESSURE - TEMPERATURE LIMITS ........... . . . . . . . . . . . . . 8-1 i
- 9.
SUMMARY
OF RESULTS 9-1 I
- 10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . . . . . . . 10-1
- 11. CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 I - iii -
eswmuctran I WSERVICE COM13ANY
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I f_ontents (Cont'd) APPENDIXES Page A. Reactor Vessel Surveillance Program Background Data and Information . A 1 B. Pre-Irradiation Tensile Data . . . . . . . . . . . . . . . . . . . . B-1 C. Pre-Irradiation Charpy Impact Data . . . . . . . . . . . . . . . . . C-1 D. Fluence Analysis Methodology . . . . . . . . . . . . . . . . . . . . D-1 E. Capsule Dosimetry Data . . . . . . . . . . . . . . . . . . . . . . . E-1 F. Tension Test Stress-Strain Curves . . . . . . . . . . . . . . . . . F-1 G. Tabulation of Temperature Distribution in Reactor Vessel Wall . . . . G-1 H References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . H-1 I List of Tables 3-1. Specimens in Surveillance Capsule W-104 . . . . . . . . . . . . . 3-2 3-2. Chemical Composition and Heat Treatment of Surveillance Materials . 3 3 5-1. Conditions of Thermal Monitors . . . . . . . . . . . . . . . . . 5-3 5-2. Tensile Properties of Capsule W-104 Base Metal and Weld Metal Irradiated to 8.84 x 10' n/cm' (E > 1 MeV) ........... 54 5-3. Charpy Impact Results for Capsule W-104 Base Metal Longitudinal (LT) Orientation, Heat No. C5667-1, 8.84 x 10 n/cm2 . . . . . . 5-5
- 54. Charpy Impact Results for Capsule W-104 Base Metal Heat-Affected l Zone Material, Heat No. C5667-1, 8.84 x 10 n/cm 2 . . . . . . . 5-5 5 5-5. Charpy Imp'act Results for Capsule W-;04 Weld Metal, 10137/3999, 8.84 x 10 n/cm 2 . . . . . . . . . . . . . . . . . . . . . . . . 5-6 5-6. Charpy Impact Results for Capsule Y Correlation Monitor Material, HSST PL-01 Longitudinal (LT) Orientation, 8.84 x 10 n/cm' . . . 5-6 6-1. Surveillance Capsule Dosimeters . . . . . . . . . . . . . . . . . 6-7 6-2. Millstone Unit 2 Reactor Vessel Fast Flux . . . . . . . . . . . . 6-8 E 6-3. Calculated Millstone Unit 2 Reactor Vessel Fluence . . . . . . . 6-9 I 6-4. Calculated Millstone linit 2 Reactor Vessel DPA . . . . . . . . . 6-10 6-5. Fluence, Flux, and DPA for Surveillance Capsules at E 6-6.
104 & 97 Degrees . . . . . . . . . . . . . . . . . . . . . . . . 6-11 3 Surveillance Capsule Measurements . . . . . . . . . . . . . . . . 6-12 7-1. Comparison of Millstone Unit 2, Capsule W-104 Tension Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . 7-7 7-2. Summary of Millstone Unit 2 Reactor Vessel Surveillance Capsules
. Tensile Test Results . . . . . . . . . . . . . . . . . . . . . . 7-8 7-3. Observed Vs. Predicted Changes for Capsule W-104 Irradiated Charpy E Impact Properties - 8.84 x 10 n/cm' (E > 1 MeV) . . . . . . . . 7-9 5 7-4. Summary of Millstone Unit 2 Reactor Vessel Surveillance Capsules Charpy Impact Test Results . . . . . . . . . . . . . . . . . . . 7-10 7-5. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Fracture Toughness - Millstone Unit 2 . . . . . . . . . . . . . . . . . . 7-11 I
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I Tables (Cont'd) Table Page 7-6. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Upper-Shelf Energy - Millstone Unit 2 . . . . . . . . . . . . . . . . . . . . 7-12
'I 8-1. Data for Preparation of Pressure Temperature Limit Curves for Millstone Nuclear Power Station, Unit-2 -- Applicable Through 20 EFPY . . . . . . . . . . . . . . . . . . . . . . . . . 84 I A-1. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials - Millstone Unit No. 2 .............A3 I A-2.
B-1. Type and Quantity of Specimens Contained in Each Irradiation Capsule Assembly . . . . . . . . . . . . . . . . . . . . . . . . . A-4 Tensile Properties of Unirradiated Shell Plate Material, Heat No. C5667-1, Longitudinal . . . . . . . . . . . . . . . . . . B-2 B-2. Tensile Properties of Unirradiated Shell Plate HAZ Material, l Heat No. C5667-1, Transverse . . . . . . . . . . . . . . . . . . . B-2 B-3. Tensile Properties of Unirradiated Weld Metal 10137/3999. . . . . . B-3 C-1. Charpy impact Data from Unirradiated Base Material, Longitudinal Orientation, Heat No. C5667-1 . . . . . . . . . . . . . . . . . . . C-2 C-2. Charpy impact Data from Unirradiated Base Metal, HAZ, Longitudinal Orientation Heat No. C5667-1 . . . . , . . . . . . . . . . . . . . C-2 I C-3. C-4. Charpy Impact Data from Unirradiated Weld Metal, 10137/3999 . . . . C-3 Charpy Impact Data from Unirradiated Correlation Monitor Material, longitudinal Orientation, HSST ' late 01, Heat No. A1008-1. . . . . C-3 I D-1. Flux Normalization factor for the 104 Degree Capsule . . . . . D-1A. Flux Normalization Factor for the 97 Degree Capsule . . . . . . . . D-9 D-2. Millstone Unit 2 Reactor Vessel Fluence by Cycle . . . . . . .
. . D-8
. . D-10 E-1. Detector Composition and Shielding . . . . . . . . . . . . . . . . E-2 l- E-2. Measured Specific Activities (Unadjusted) for Dosimeters in the 104 Degree Capsule . . . . . . . . . . . . . . . . . . . . . . . . E-3 E-3. Measured Specific Activities (Unadjusted) for Dosimeters in the I E-4.
G-1. 9 7 Deg ree C ap s ul e . . . . . . . . . . . . . . . . . . . . . . . . . E- 3 Dosimeter Activation Cross Sect'ons, b/ atom . . . . . Normal Heatup Temperature Distribution Along Vessel Wall and
. . . . .. . E-4 I G-2.
Uncorrected Pressures . . . . . . . . . . . . . . . . . . . . . . . G-2 N::rmal Cooldown Temperature Distribution Along Vessel Wall and Uncorrected Pressures . . . . . . . . . . . . . . . . . . . . . . . G-3 E . List of Fiaures Figure Page 3-1. Reactor Vessel Cross Section Showing Location of Capsule W-104 in I 3-2. Millstone Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . Typical Surveillance Capsule Assembly Showing Location of 3-4 Specimens and Monitors . . . . . . . . . . . . . . . . . . . . . 3-5 I -v-l B WUREYfM L v
Fiaures (Cont'd) Figure Page 3-3. Typical Surveillance Capsule Tensile - Monitor Compartment Assembly (Three per Capsule) ..................3f6 3-4. Typical Surveillance Capsule Charpy impact Compartment Assembly (Four per Capsule) ....................,,3J 5 1. Charpy Impact Data for Irradiated Plate Material, Longitudinal Orientation, Heat No. C5667-1 . . . . . . . . . . . . . . . . , . 5-7 5-2. Charpy Impact Data for Irradiated Plate Material Heat-Affected Zone, Heat No. C5667-1 .. . . . . . . . . . . . . . . . . . . . 5-8 5-3. Ch:rpy Impact Data for Irradiated Weld Metal, 10137/3999. . . . . 5-9 5-4. Charpy Impact Data for Unirradiated Correlation Monitor Material, HSST PL-01, Heat No. A1008-1 . . . . . . . . . . . . . . . . . . 5-10 5-5. Photographs of Tested Tension Test-Specimens and Corresponding Fractured Surfaces Base Metal, Longitudinal . . . . . . . . . . 511 5-6. Photographs of Tested Tension Test Specimans and Corresponding Fractured Surfaces - Heat-Aftected Zone . . . . . . 5-12 5-7. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal 10137/3999 . . . . . . . . . . . 5-13 5-8. Photographs of Charpy impact Specimen Fracture Surfaces - Base Met al Heat No . C5667-1 . . . . . . . . . . . . . . . . . . . 5-14 5-9. Photographs of Charpy Impact Specimen Fracture Surfaces - Base Metal, Heat-Affected Zone . . . . . . . . . . . . . . . . . 5-15 5-10. Photographs of Charpy Impact Specimen Fracture Surfaces - Weld Metal 10137/3999 . . . . . . . ~. . . . . . . . . . . . . . 5-16 5-11. Photographs of Charpv Impact Specimen Fracture Surfaces - Correlation Material, HSST PL-01 ................517 6-1. General Fluence Determination Methodology . . . . . . . . . . . . 6-2 6-2. Fast Fluence and DPA Distribution Through Reactor Vessel Wall . . 6-13 6-3. Azimuthal Flux and Fluence Distributions at Peactor Vessel Inside Surface . . . . . . . . . . . . . . . . . . . . . . . . . 6-14 6-4. Relative Axial Variation of Fast Neutron (E > 1.0 MeV) E Flux and fluence . . . . . . . . . . . . . . . . . . . . . . . . 6-15 E 7-1. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Plate Material Longitudinal Orientation, Heat No. C5667-1 . . . . . . . . . . . . . . . . . . . . . . . . 7-13 7-2. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Heat-Affected Zone, Heat No. C5667-1. . . . . . . . . 7-14 7-3. Comparison of Unirradiated and Irradiated Charpy Impact Data for Weld Metal 10137/3999 . . . . . . . . . . . . . . . . . . . . 7-15
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7-4. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Correlation Monitor Material, HSST PL-01, E Heat No. A1008-1 . . . . . . . . . . . . . . . . . . . . . . . . 7-16 3 8-1. Predicted Peak Fast Neutron Fluence at Various Locations Through Reactor Vessel Wall for 32 EFPY - Millstone Nuclear Power S t a t i on , Un i t-2 . . . . . . . . . . . . . . . . . . , , . . . . 8- 5 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Heatup, Applicable' for First 20 EFPY - Millstone Nuclear Power Station, Unit-2 . . . . . . . . . . . . . . . . . . 8-6
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I E I$9ECLLCont'd) Figure Page 8-3. Reactot lessel Pressure-Temperature Limit Curves for Normal Operation - Cooldown, Applicable for First 20 EFPY - Millstone I 8 4. Nuclear Power Station, Unit-2 . . . . . . . . . . . . . . . . . . Reactor Vessel Pressure Temperature Limit Curves for Inservice 8-7 Leak and Hydrostatic Tests, Applicable for first 20 EFPY - I A-1. Millstone Nuclear Power Station, Unit-2 . . . . . . . . . . . . . Location and Identification of Materials Used in the Fabrication of Millstone Unit 2 Reactor Pressure Vessel . . . . . . . . . . . . A-5 0-8 g A 2. Location of Beltline Region Materials in Relationship to the 3 Reactor Vessel Core . . . . . . . . . . . . . . . . . . . . . . . . A-6 A-3. Location of Longitudinal Welds in Millstone Unit 2 Upper and Lowe r Shell Course s . . . . . . . . . . . . . . . . . . . . . . . . A-7 A-4. Location of Surveillance Capsule Irradiation Sites in Millstone Unit 2 .........................A8 C-1. Charpy impact Data from Untrradiated Base Metal (Plate), I C-2. Longitudinal Orientation, Heat No. C5667-1 ........... Charpy lapact Data from Unirradiated Heat-Affected-Zone Base Met al , He at No. C5667-1 . . . . . . . . . . . . . . . . . . . . . . C- 5 C-4 C-3. Charpy impact Data from Unirradiated Weld Metal, D137/3999 . . . . C-6 I C-4. Charpy Impact Data for Unirradiated Correlation Monitor Material, MSST PL-01 . . . . . . . . . . . . . . . . . . . . . . . C-7 9-1. Rationale for the Calculation of Dosimeter Activities and Neutron I' D-2. Fl ux in the Capsul e . . . . . . . . . . . . . . . . . . . . . . . . D-11 Rationale for the Calculation of Neutron Flux in the Reac'3r Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . D-12 I D-3. F-1. Plan View Through Reactor Core Midplane (Reference R-e Calculation Model) . . . . . . . . . . . . . . . . . . . . . . . Tension Test Stress-Strain Curve for Base Metal Plate Heat 0-13 C5667-1, Specimen No. IJ1, Tested at 70F . . . F-2 I F-2. Tension Test Stress-Strain Curve for Base Metal Plate Heat C5667-1, Specimen No. IJY, Tcited at 550F . . . . . . . . . . . . . F-2 F-3. Tension Test Stress-Strain Curve for Base Metal Plate Heat C5667-1, Specimen No. IK7, Tested at 550F , . . . . . . . . . . . . F-3 I F-4. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat C5667-1, Specimen No. 4KC, lested at 70F . . . . . . . . F-3 F-5. Tension Test Stress-Strain Curve for Base Metal Heat-Affected I Zone, Heat C5667-1, Specimen No. 4J1, Tested at 400F .. . . . . . F-4 F-6. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat C5667-1, Specimen No. 4J4, Tested at 550F .. . . . . . F-4 I F-7. F-8. Tension Test Stress-Strain Curve for Weld Metal 10137, Specimen No. 3KC, Tested at 70F . . . . . . . . . . . . . . . . . . . . . . F-5 Tension Test Stre,s-Strain Curve for Weld Metal 10137, Specimen I F-9. No. 3JT, Te sted at 400F . . . . . . . . . . . . . . . . . . . . . . F- 5 Tension Test Stress-Strain Curve for Weld Metal 10137, Specimen No. 3JL, Tested at 550F . . . . . . . . . . . . . . . . . . . . . . F-6 I , - vil - BW!!sMafaw
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- 1. INTRODUC110N I This report describes the results of the examination of the second capsule (Capsule W-104) of the Northeast Nuclear Energy Company, Millstone Nuclear Power I Station, Unit No. 2 (Millstone Unit 2) reactor vessel material surveillance program (RVSP). The capsule was removed and evaluated after being irradiated in the Millstone Unit 2 as part of the reactor vessel materials surveillance program Combustion Engineering (C-E) Report N-NLM-Oll.' The capsule experienced a fluence of 8.84 x 10 18 n/cm' (E > 1 MeV), which is the equivalent of approxi-mately ten effective full power years' (EFPY) operation of the Millstone Unit 2 reactor vessel inside surface. The first capsule (Capsule W-97) from this I program was removed and examined after the first year of operation; the results are reported in C-E Report TR-N-MCM-008.2 In addition, a supplemental dosimetry I ca;,sule, which had been inserted in the location of the previously analyzed Capsule W-97, was removed and analyzed as a part of the dosimetry evaluation in support of (,apsule W-104.
The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact pru erties of reactor presrure vessel materials under actual operating conditions. The surveillance program for Millstone Nuclear Power Station Unit No. 2 was designed and furni wed by Combustion Engineering, Incorporated (C-E) as described in N-NLM-Oll' and conducted in accordance with The program was planned to monitor the effects of neutron 10CFR50, Appendix H.
= irraciation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel. .I I
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- 2. BACKGROUND I The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water cooled I reactors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of such low t!19y ferritic steels as SA533, Grade B, used in the fabrication of the Hillstone Unit 2 reactor vessel, are well characterized and documented in the literature. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding I decrease in ductility after irradiation. The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the Charpy upper shelf' energy value.
Appendix G to 10CFR50, " Fracture Toughness Requirements,"8 specifies minimum fracture toughness requirements for the ferritic materials of the pressure-l retaining compor'ents of the rrector coolar.t pressure boundary (RCPB) of water-cooled power reactors, ar.d provides specific guide'ines for determining the pressure temperature limitations on operr. tion of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences I and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 13, 1973, the requirements are applicable te all boiling and pressurized water-cooled nuclear power reactors, including those l under construction or in operation on the effective date.
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I Appendix H to 10CTR50, " Reactor Vessel Materials Surveillance Program Requirements,"' defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Tracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life. A method for guarding against brittle fracture in reactor prer,sure vessels is described in Appendix G to the ASME Doller ar.d Pressure Vessel Code, Section !!!, " Nuclear Power Plant Components."' This method utilizes fracture mechanics < concepts and the reference nil ductility temperature, RTNDT, which is defined a:, g the greater of the drop weight nil ductility transition temperature (per AS1H a E-208) or the temperature that is 60F below that at which the material exhibits 50 f t lbs and 35 mils lateral expansion. The RT NDT f a given material is used to index that material to a reference stress intensity factor curve (K curve), lR which appears in Appendix G of ASME Section III. The K curve is a lower bound IR of dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K IR curve, allowable stress intensity factors can be obtained for this material as a functicn of temperature. Allowable operating limits can then be determined , using these allowable stress intensity factors. 5 The RT NDT and, in turn, the operating limits of a nuclear power plant, can be g adjusted to account for the effects of radiation on the properties of the reactor W vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in'which a surveillance capsule containing prepared specimens of the reactor vessel materials is periodically removed from the operating nuclear reactor and the specimens are tested. The increase in the Charpy V notch 30 ft lb temperature is added to the original RT NDT to adjust it for radiation embrittlement. This adjusted RT NDT is used to index the material to the K IR curve which, in turn, is used in set operating limits for the nuclear l 22 I
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I power plant. 1hese new limits take into account the effects of irradiation on the reactor vessel materials. Appendix G, 10CfR50, also requires a minimum Charpy V notch upper-shelf energy of 75 ft-lbs for all beltline region materials unless it is demonstrated that I lower values of upper shelf fracture energy will provide an adequate margin for deterioration as the result of neutron radiation. No action is required for a l material that does not meet the 75 f t-lb requirement provided the irradiation l deterioration does not cause the upper shelf energy to drop below 50 ft lbs. The regulations specify that if the upper shelf energy drops below 50 f t-lbs, it must be demonstrated in a manner approved by the Office of Nuc1 car Regulation that the lower values will provide adequate margins of safety. When a reactor vessel fails to meet the 50 f t-lb requirement, a program must be submitted for review and approval at least three years prior to the time the predicted fracture toughness will no longer satisfy the regulatory requirements. The program must addrt:s the following: A, A volumetric examination of 100 percent of the beltline materials that do not meet the requirement. B. Supplemental fracture toughness data as evidence of the fracture toughness of the irradiated beltline materials. C. Fracture toughness analysis to demonstrate the existence of equivalent margins of safety for continued operation, if these procedures do not indicate the existence of an adequate margin of safety, the reactor vessel beltline may be given a thermal annealing treatment to recover the fracture toughness properties of the materials. I I I I I 2-3 I 13W # M E! L v
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- 3. SURVEILLANCE PROGRAM DESCRIPTION I The surveillance program for Hillstone Unit 2 comprises six surveillance capsules designed to monitor the effects ri nN.ron and thermal environment on the h materials of the reactor pressure core rrgir, T;.. ;apsules, which were inserted into the reactor vessel before inittJ timt sivtup, were positioned inside the l reactor vessel between the thermal shielu .ad the vessel wall at the locations shown in figure 31. The six capseles, d"igned to be placed in holders attached I to the reactor vessel wall are positioned near the peak axial and azimuthal neutron flux. C-E Report N NLM Oll' includes a full description of the capsule locations and design. During the ten cycles of operation, Capsule W 104 was I irradiated in the 104' position adjacent to the reactor vessel wall as shown in figure 3-1.
Capsule W 104 was removed during the tenth refueling shutdown of H111 stone Unit
- 2. The capsule contained Charpy V-notch impact test specimens fabricated from the one base metal (SA533, Grade B1), one heat-af fected-zone, a weld metal and a correlation monitor. Tension test specimens were fabricated from the base l metal and the weld metal only. The number of specimens of each material contained in the capsule are described in Table 31, and the location of the individual specimens within the capsule are described in Figures 3-2 through 3 4.
The chemical composition and heat treatment of the surveillance material in Capsule W-104 are described in Tabic 3-2. All test specimens were machined from the 1/4-thickness (1/4T) location of the plate material. Charpy V-notch and tension test specimens were cut from the I surveillance material such that they were oriented with their longitudinal axes either parallel or perpendicular to the principal working direction. lI lI 31 13W!i&WafL v
E The neutron dosimeters contained in Capsule W-104 are as follows: Threshold E tiitterial Shieldina Reaction Eneroy (Mev) _ Half-Life B Uranium None/Cd U"' (n. f) C s"' O.7 30.2 years Sulfur Nont- S (n. p) p*' 2.9 14.3 days Iron Nickel None Cd f e (n.p) Mn'" Ni (n.p) Co" 4.0 5.0 312.5 days 70.9 days l Coppe'c C4 Cu" (n.a) Co" 7.0 5.27 years Titanium None T i (n.p) Sc 8.0 83.8 days Cobalt None/Cd Co" (n,y) Co" Thermal 5.27 years four thermal monitors of low melting alloys were included in the W-104 capsule. The eutectic alloys and their melting points are as follows: 1 A11ov Comnosition wt?.' Melting Point. F l 80.0 Au, 20.0 Sn 536 90.0 Pb, 5.05 Sn, 5.0 Ag 558 97.5 Pb, 2.5 Ag 580 97.5 Pb, 0.75 Sn, 1.75 Ag 590 3 Table 3-1. Specimens in Surveillance Caolgle W-10.1 $ Number of Test SDecimens Material Description Tension CVN Impac1 , Base Metal (C5667-1) longitudinal 3 12 Heat-Affected Zone 3 12 Weld Metal (10173/3999) 3 12 Correlation Material - 12 l (HSST PL-01) ___ m Total Por Capsule 9 48 3-2 BWilRE?n!?om E
I I Table 3 2. Chemical Composition and Hoat Treatment of Surveillage_titi.crials Chemical Composition. wZo _. Heat N Correlagn Weld Metal C5667-1g Honitor I Element C 0.21 0.22 (10137/3999)Ic) 0.12 l Hn P 1.26
- 0. w 1.48 0.012 1.13 0.016 I S Si 0.014 0.12 0.018 0.25 0.013 0.16 Ni 0.61 0.60 0.06 Cr 0.10 ---
0.05 Ho 0.62 0.52 0.54 Cu 0.14 0.14 0.25 I Heat No. lemo. f Heat Treatment lim d Coolino Plate I C5667-1 1600 1225 1150 4 4 40 Water Quenched furnace Cooled furnace Cooled to 600f Weld Metal 1150 40 furnace Cooled to 600f I (10173/3999) Correlation I Monitor (HSST PL-01) 1700 1600 1225 4 4 4 Air Cooled Water-Quenched furnace Cooled 1150 40 furnace Cooled to 600f (*)Clemical analysis by Combustion Engineering of surveillance program test plate C5667-1 (b) Chemical analysis from NUREG/CR 4092." (c) Chemical analysis by Combustion Engineering of surveillance program l test weld metal.# I lI 3-3 GW$$NWE'o'Omv
I figure 3 1. Reactor Vessel Cross Section Showing location of Capsule R 104 in Mt11storte_Mt__2 l' I lid [~~~~~"' Outletibille Vessel /
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I Figure 3-3. Typical Surveillance Cap:ule Tensile Compartmant Assembly (Three per Caplule) Monitor I I Wedge Coupling End Cap /
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I I I l 4. PRE IRRADIATION TESTS Unirradiated material was evaluated for two purposes: (1) to establish a baseline of data to which irradiated properties data could be referenced, and (2) to determine those materials properties to the extent practical from available l material, as required for compliance with Appendixes G and H to 10CFR50. l The pre irradiated specimens were tested by Combustion Engineering as part of the development of the Hillstone Unit 2 surveillance program. The details of the testing procedures are described in C E Report N NLH-Oli and are summarized here to provide continuity. 4.1. Tension Tests Tension test specimens were fabricated from the reactor vessel shell plate and g weld metal. The specimens were 3.00 inches long with a reduced section 1.50 inches long by 0.250 inch in diameter. They were tested on a universal test machine. An extension device with a strain gaged extenscaeter was used to I. determine the 0.2% yield point. Test conditions were in accordance with the applicable requirements of ASTH A370 68.6 For each material type and/or condition, nine specimens in groups of three wert tested at room temperature,250 and 550F. All test data for the pre irradiation tensile specimens are given in Appendix B. l 4.2. Impact Tests Charpy V-notch impact tests were conducted in accordance with the requirements 7 of ASTM E23 72 on an impact tester certified to meet Watertown standards.8 Test specimens were of the Charpy V-notch type, which were nominally 0.394 inch square and 2.165 inches long. Impact test data for the unirradiated baseline reference materials are presented in Appendix C. Tables C-1 through C 4 contain the basis data that are plotted I I 4-1 IS W !! M a % v
l in Figures C-1 through C 4. These data were replotted and re evaluated to be 5l consistent with the irradiated Charpy curves and evaluations. i I I-I: I I I I a i E I I I l' I 4-2 13W!!nn?%%r E.
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!I I j 5. POST lRRADIATION TESTING 5.1. Visital Examination and Inventory The capsule was inspected and photographed upon receipt and confirmed that the 5 markings as those of Capsule W-104. The contents of the capsule were inventoried and found to be consistent with the surveillance program report inventory. All specimens were visually cramined and no signs of abnormalities were found. There was no evidence of rust or of the penetration of reactor coolant into the capsule. 5.2. Thermal Monitors Surveillance Capsule W 104 contained three temperature monitor holder blocks each containing four fusible alloys with different melting points. Each of the j thermal monitors was inspected and the results are tabulated in Table 51. From these data, it was concluded that the irradiated specimens had been exposed to a maximum temperature no greater than 55BF during the reactor vessel operating f period. This is not significantly greater than nominal inlet temperature of 550F, and is considered acceptable. The partly melted appearance is probably due to an irradiation induced creep mechanism and not actual riting. This behavior has been seen in other surveillance capsules. There appeared to be no signs of I a significant temperature gradient along the capsule length. 5.3. Tension Test Results The results of the post-irradiation tension tests are presented in Table 5-2. Tests were performed on specimens at room temperature, 400, and 550i . They were tested on a 55,0001b load capacity universal test machine at a cross-head speed of 0.005 inch per minute to yield point and thereafter 0.050 inch per minute. A 4-pole extension device with a strain gaged extensom.,ter was used to determine , the 0.2% yield point. Test conditions were in accordance with the applicable I 5-1 13WE!5EYf?00mr
I requirements of A51H A370 77.' for each material type and/or condition, ' specimens were tested at room temperature,400 and 550f. The tension compression l load cell used had a certified accuracy of better than 40.5% of full scale l (25,000 lb). Photographs of the tcnsion test specimen fractured surfaces are presented in Figures 5 5 through 5-7. In general, the ultimate strength and yield strength of the material increased with a corresponding slight decrease in ductility as compared to the unirradiated values; both effects were the result of ' neutron radiation damage. The type of behavior observed and the degree to which the material properties changed in within the range of changes to be expected for the radiation environment to which l the specimens were exposed. The results of the pre-irradiation tension tests are presented in Appendix B. 5,4. Charov V-Notch Impact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 5 3 through 5-6 and figures 51 through 5-4. Photographs of the Charpy specimen fracture surfaces are presented l in Figures 5-8 through 5-11. The Charpy V notch impact tests were conducted in g accordance with the requirements of ASTM E23-86 10 on an impact tester certified to meet Watertown standards,8 The data show that the materials exhibited a sensitivity to irradiation within I the values to be expected based on their chemical composition and the fluence to a which they were exposed. Detailed discussion of the results are provided in E Section 7. I l l The results of the pre-irradiation Charpy V-notch impact tests are given in Appendix C. I I i I 5-2 E j3ggfMy 7 r I
I I lable 5 1. Cond111gatof 1hermal Moniinn I Capsule
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(Bottom) 558F Softened but intact 580f Unmelted 590f Unmelted I I I lI E lI 5-3 ', maawa-
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c Table 5 3. CharpyimpactResultsforCapsuleW104BaseygtalLpngitudinal L (LT) Orientation, tient No. C56671, 8.84 x 10 n/cm Test Impact lateral Shear Specimen Temperature Energy Expansion, fracture 1D F fi.lh inch 9 127 70 13.0 0.018 20 12T 90 16.5 0.017 15 l llA 142 130 150 32.0 40.0 0.034 0.040 35 40 13B 175 43.5 0.040 40 140 185 40.0 0.046 50 1 112 200 74.0 0.069 75 12L 240 74.5 0.069 70 115 280 93.5* 0.079 100 I 116 124 340 400 95.0* 92.0* 0.084 0.082 100 100 14C 550 100.0* 0.088 100
- Values used to determine upper shelf energy value per ASTM E-185."
Table 5 4. Charpy Impcet Results for Capsule W-104 Ba g Metal Heat-Affected Zone Material, lisat No. C5667-1 8.84 x 10 n/cm' Test impact lateral Shear Specimen Temperature Energy Expansion, fracture ID F ft-lbs inch % 43J 20 52.5 0.036 50 457 70 27.5 0.024 30 43T 90 37.0 0.038 40 43A 110 83.5 0.055 70 452 130 88.0 0.067 80 450 150 51.0 0.043 50 437 200 145.0 0.095 100 43P 200 52.5 0.054 60 453 300 , 124.0* 0.081 100 45T 300 91.5* 0.073 100 45J 400 98.0* 0.080 100 45U 550 100.0 0.082 100
- Values used to /e' ermine upper shelf energy value per ASlH E 185."
5-5 B W i! M ale % r
I Table 5 5. 104 CharpyImpactResultsforCapsulegn/cm' Weld Metal, 10137/3999, 8.84 x 10 g Test Impact lateral Shear Spe:imen Tempersture Energy Expansion, Fracture ID (. _ 11-Jh1 inch % 353 0 13.5 0.013 15 35E ch 33.5 0.031 40 335 32B 3', 50 4 40.0 59.5 56.5 0.035 0.050 0.052 40 50 60 l 341 - 35P i!Y. 68.0 0.055 70 3 35B ?N 89.5 0.071 85 5 32C 'M
, 107.0* 0.087 100 35C #$ 106.0* 0.088 100 m 320 %0 109.5* 0.090 100 g, 32T 400 99.0 0.087 100 371 557 111.0 0.092 100
- Values used to determine upper-shelf energy value per ASTM E-185."
I I Table 5 6. Charpy impact Results for Capsule Y Correlation Hongor Material, HSST PL Ol Longitudinal (LT) Orientation, 8.84 x 10 n/cm' _ l Test Impact lateral Shear Specimen Temperatare Energy Expansion, fracture ID ,,_. . f._ ft-1bs inch % l 74L 70 4.5 0.008 0 IR 74J 150 21.0 0.022 20 E 747 170 39,0 0.036 40 74D 18!5 41.0 0.038 40 a 751 200 46.5 0.047 50 5 741 210 60.5 0.051 60 754 270 60.0 0.054 60 74P 74A 240 275 60.0 94.5* 0.057 0.079 60 100 g 74K 300 90.5* 0.077 100 7/d 400 96.0* 0.085 100 742 550 87.0 0.078 100
- Values used to determine upper shelf energy value per ASTM E-185."
l I 56 13Wf!sefahr I
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- 6. NEUTRON FLUENCE 6.1. Introduction The neutron fluence (time integral of flux) is a quantitative way of expressing the cumulative exposure of a material to a pervading neutron flux over a specific period of time. Fast neutron fluence, defined as the fluence of neutrons having energies greater than 1 MeV, is the parameter that is presently used to correlate I radiation induced changes in material properties. Accordingly, the fast fluence must be determined at two locations: (1) in the test specimer: located in the surveillance capsule, and (2) in the wall of the reactor vessel. The former is used in developing- the correlation between fast fluence and changes in the material properties of specimens, and the latter is used to ascertain the point l
of maximum fluence in the reactor vessel, the relative radial and azimuthal distribution of the fluence, the fluence gradient through the reactor vessel wall, and the corresponding material properties. The accurate determination of neutron flux is best accomplished through the simultaneous consideration of neutron dosimeter measurements and analytically derived flux spectra. Dosimeter measurements alone cannot be used to predict the fast fluence in the vessel wall or in the test specimens because (1) they cannot measure the -fluence at the points of interest, and (2) they provide only rudimentary information about the neutron energy spectrum. Conversely, reliance on calculations alone to predict fast fluence is not prudent because of the length and complexity of the analytical procedures involved. In short, measurements and calculations are necessary complements of each other and together they provide assurance of accurate results. Therefore, the determination of the fluence is accomplished using a combined analytical-empirical methodology which is outlined in Figure 6-1 and described in the following paragraphs. The details of the procedures and methods are presented in general terms in Appendix D and in BAW-1485p." 6-1 1 BW#defaMB- l
I Ficure 6-1. General Fluence Determination Meth;dology MEASLRE}iENTS OF NEUTRON AMLYTICAL DETERMIMTION OF DOSIMETER ACTIVITIES DOSDiETER ACTIVITIES AT NElJiRON FLUX l Aa)USTED ENERGY DEPDOENT NEUTRON FLUX I REACTOR OPERATING NEUTRON HISTORY # D PRE-FLUENCE DICTED FURRE OPERATION I Analytical Determination of Dosimeter Activities and Neutron Flux The analytical calculation of the space and energy dependent neutron flux in the test specimens and in the reactor vessel is performed with the two dimensional a discrete ordinates transport code, DOTIV." The calculations employ an angular E quadrature of 48 sectors (SB), a third order LeGendre polynomial scattering approximation (P3), the BUGLE cross section set with' 47 neutron energy groups and a fixed distributed source corresponding to the ti_me weighted average power distribution for the applicable irradiation period. l_ In addition to the flux in the test specimens, the DOTIV calculation determines the saturated specific activity of the various neutron dosimeters located in the surveillance capsule using the ENDF/B5 dosimeter reaction cross sections. The l saturated activity of each dosimeter is then adjusted by a factor which corrects for the fraction of saturation attained during the dosimeter's actual (finite) I 6-2 13W!!na?aM8m g
irradiation history. Additional corrections are cade to account for the following effects:
- Photon induced fissions in V and Np dosimeters (without this correction the results underestimate the measured activity).
- Short half-life of isotopes produced in nickel, titanium, and iron dosimeters (71 day Co-58, 84 day Sc-46, and 312 day Mn-54 respectively).
(Without this correction, the results could be biased high or low I depending on the long term versus short term power histories.) Measurement of Neutron Dosimeter Activities The accuracy of neutron fluence predictions is improved if the calculated neutron flux is compared with neutron dosimeter measurements adjusted for the effects noted above. The neutron dosimeters located in the surveillance capsulet are listed in Table 6-1. Both activation type and fission type dosimeten were used. The ratio of measured dosimeter :.etivity to calculated dosimeter activity (M/C) is determined for each dosimeter, as discussed in Appendix D. These M/C ratios are evaluated on a case-by-case basis to assess the dependability or veracity of each individual dosimeter response. After carefully evaluating all factors known to affect the calculations or the measurements, an average M/C ratio is calculated and defined as the " normalization factor." The normalization factor is applied as 3.n adjustment factor to the DOT-calculated flux at all points of I interest. Neutron Fluenn The determination of the neutron fluence from the time averaged flux requires only a simple multiplication by the time in EFPS (effective iull-power seconds) I, over which the flux was averaged, i.e. l 5 )(00
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where f (AT) - Fluence at (i,j) accumulated over time T (n/cm'), g g - Energy group index, 6-3 13W!!senif&vv
p - Time average flux at (i,j) in energy group g, (n/cm'-sec), no AT - Irradiation time, EFPS. Neutron fluence was calculated in this analysis for the following components over the indicatad operating time: Iest Specimens: Capsule irradiation time in EFPS Fluence Monitors: Capsule irradiation time in EFPS Reactor Vessel: Vessel irradiation time in EFPS Reactor Vessel: Maximum point on inside surface extrapolated to 32 effective full power years The neutron exposure to the reactor vessel and the material surveillance specimens was also determined in terms of the iron atom displacements per atom of iron (DPA). The iron DPA is an exposure index giving the fraction of iron atoms in an iron specimen which would be displaced during an irradiation. It is l considered to be an appropriate damage exposure index rince displacements of atoms from their normal lattice sites is a primary source of neutron radiation damage. DPA was calculated based on the ASTM Standard E693-79 (reapproved g 1985). " A DPA cross section for iron is given in the ASTM Standard in 641 g energy groups. DPA per second is determined by multiplying the cross section at a given energy by the neutron flux at that energy and integrating over energy. DPA is then the integral of DPA per second over the time of the irradiation. In l the DPA calculations reported herein, the ASTM DPA cross sections were first collapsed to the 47 neutron group structure of BUGLE; the DPA was then determined l by summing the group flux times the DPA cross section over the 47 energy groups and multiplying by the time of the irradiation. 6.2. Vessel Fluence The maximum fluence (E > 1 MeV) exposure of the Millstone Unit 2 reactor vessel during Cycles 1-10 was determined to be 9.21 x 10 n/cm' based on a maximum neutron flux of 2.27 x 10' n/cm'-s during cycles 1 to 5 and 3.60 x 10' n/cm'-s during cycles 6 to 10 (2.94 x 10' n/cm'-s when time averaged over cycles 1 to 10; Tables 6-2 and 6-3). The maximum fluence occurred at the cladding / vessel interface at an azimuthal location of approximately I degree from a major 6-4 13W!!sNahr E
horizontal axis of the core (Figure 6-3). Cumulative DPA results were calculated
- at the quarter T positions and are presented in Table 6-4.
Fluencc data were extrapolated to 32 EFPY of operation based on two assumptions: j (1) the future fuel cycle operations do not differ significantly from the cycle i 10 design, and (2) the latest calculated (or extrapolated) flux remains constant !
. from that time through 32 EFPY. Tic extrapolation was carried out from E0C 10 I to 32 EFPY. TF tycle averaged fluxes for future cycles are assumed to be the flux experiencc, Juring cycle 10.
Fast fluence and DPA (displacement per atom) gradients relative to the inside surface of the reactor vessel wall are shown in Figure 6-2. Reactor vessel neutron fluence lead factors, which are the ratio of the neutre') flux at the clad interface to that in the vessel wall at the T/4, T/2 and 3T/4 locations, are 1.89, 4.14, and 9.60, respectively. DPA lead factors at the same locations are 1.62, 2.80, and 5.12, respectively. The relative fluence as a function of azimutha angle is shown in Figure 6-3. A peak occurs in the fast flux (E > 1 E
- a MeV) at two locations over two time periods. The peak average flux from cycles 1 to 10 occurred at about I degree with a corresponding value of 2.94 x 10' n/cm'- s . The peak average flux during the low-leakage cycle 10 occurred at about 27 degrees with a corresponding value of 2.70 x 10' n/cm'- s . Also note that the curves shown in Figure 6-3 ,re normalized to their respective absolute values at the O degree position, and therefore, their magnitudes cannot be compared to each other. However, the shapes are comparable.
The flux and fluence results were corrected using the final measured to l calculated activity ratio (M/C) derived from the 104 degree capsule (0.99) and ? were also corrected to account for an axial power peak (1,11). The M/C ratio is detailed in Appendix D. The axial fluence, which was normalized over the height of the core and assumed to be proportional to the axial power distributions in the pripheral assemblies, was averaged over cycles 1 to 10 and is shown in ll Figure 6-4. The relative axial fluence for cycle 10 is also shown in Figure 6-4. It is important to note the change in the distributions, since they will affect future fluence exposure to the vessel. Cycle 10 was the first low leakage cycle and, thus, provided a relatively flat profile over most of the core. The curves in Figure 6-4 are normalized to 1.0. Thus, peaking factors can be taken directly BWi!#NEM6
from the curves. However, the magnitudes of the curves cannot be compared to each other. Only the shapes are comparable. 6.3. Caosule Fluents Two capsules were installed and irradiated over two different time periods in Hillstone Unit 2. The first capsule, located at the 104 degree position - 218.04 cm from the core center, was irradiated for 3631.2 EFPD during cycles I to 10. This capsule contained fluence monito- (dosimetry) and charpy specimens. The cumult 'se fast fluence at the centar he 104 degree surveillance capsule was calculated to be 8.84 x 10 n/cr This fluence value represents an average value for the various % psule. The second c.apsule, located at the 95 e - also 218.04 cm from the . core center, was irradiated for 1828.4 49 C to 10. This capsule contained only fluence monitors. The cumr
- s iue.v.e at the center of the 97 degree surveillance capsule was calculated tc a 7.4 x 10 n/cm'. This fluence value represents an average value for the various locations in the capsule.
The flux and fluence results were corrected for each capsule using the final measured over calculated activity ratios derived for each capsule as detailed in appendix 0. 6.4. Fluence Uncertainties Surveillance capsules provide neutron dosimetry information as well as materials , data at various points during the lifetime of power reactors. The dosimetry 3 results, measured-to-calculated ratios, obtained from numerous analyses utilizing the same methodologies provide a measure of confidence in the analytical techniques and a benchmarking for the methodology used to determine ves:el l fluence. Table 6-6 presents a comparison of the results of fourteen surveillance capsule analyses which utilized the same analytical technique. I I I 6-6 sum ==_ ,
l 'I Table 6 1. Surveillance Capsule Dosimeters I Lower Energy Limit for Isotope I Dosimeter Reactions
- Reaction, MeV Half-Life 5'Ni(n p)6'Co 2.3 70.8 days
T i (n . p)**Sc 3.0 83.8 days
f e (n , p ) 6'Mn 2.5 312.5 days
Cu(n, y)* Co 3.7 5.27 years
8 U(n f)'Cs 1.1 30.0 years
**Co(n, y)* Co Thermal 5.27 years"
Np (n , f)'8'C s 0.5 30.0 years "The sulphur dosimeter was not analyzed, since at the time of measurement, too many half-lives had expired; subsequently not enough activity remained for a reliable measurement.
- Reaction activities measured for capsule flux evaluation.
I. "The Cobalt dosimeter was designed to measure flux in the thermal region, and therefore was excluded from the analysis. I I I I 6-7 BWllEEVaP%r
I Table 6-2. Millstone Unit 2 Reactor Vessel fast Flux Fast Flux (E > 1 MeV), n/cm'-s flux n/cm'-s (E > 0.1 MeV) Inside Surface Inside Surface Cycle (Max location) T/4 3T/4 (Max location) Cycles 1-5 2.27E+10** 1.19E+10 2.33E+9 5.60E+10 (1802.8 EFPD) Cycles 6-10 3.60E+10** 1.92E+10 3.79E+9 7.71E+10 (1828.3 EFPD) 15 EFPY 2.70E+10** 1.43E+10* 2.81E+9* ---- I 21 EFPY 2.70E+10 1.43E+10* 2.81E+9* ---- 24 EFPY 2.70E+10 1.43E+10* 2.81E+9* ---- 32 EFPY 2.70E+10 1.43E+10* 2.81E+9* ----
- Divide flux at inside surface by the appropriate lead factors on page 6-5 to obtain thase T/4 and 3T/4 fast flux values.
E
- The maximum flux results for cycles 1 to 10 are at the 1 degree position off g the major axis. Af ter cycle 10, the maximum location shifts to 27 degrees.
I I I I I 6-8 BWeinE?anLa g
l I l 1 lable 6 3. Calculated Millstone Unit 2.lcActor Vessel Flunce f ast Fluence, n/cm' (E > 1 MeV) Cummulative Inside Surface Irradiation Time (Max location) T/4 T/2 3T/4 End of Cycle 5 3.54E+18 1.86E+18 8.42E417 3.63E+17 (1802.9 EFPD) End of Cycle 10 9.21E+18 4.88E+18 2.22E+18 9.60E+17 (3631.2 EFPD) 15 EFPY 9.52E+18 5.04[+18* 2.30E+18* 9.92E+17* 21 EFPY 1.46E+19 7.72E+18* 3.53E418* 1.52E+18* 24 EFPY 1.72E+19 9.10E+18* 4.15E+18* 1.79E+18* 2.40E+19 2.50E418* I 32 EFPY 1.27E+19* 5.80E+18*
- Calculated using 1.0 1.89 4.14 9.60 these lead factors Conversion factors fluence (E > 1 MeV) 1.51E-21** 1.75E-21** 2.23E-21** 2,82E-21**
to DPA. I ** Multiply fast fluence values (E > 1 MeV) in units of n/cm' by these factors to obtain the corresponding DPA values. Note: Fast fluence results from cycle I through 10 represent the azimuthal peak I at one degree off the major axis. Fast fluence results after cycle 10 represent the azimuthal peak at 27 degrees of the major axis. l I 6-9 BW#sefahr g
I Table 6-4. Calculated Millstone Unit 2 Reactor Vessel DPA DPA. aisolacements/ atom (Total) Cummulative Inside Surface Irradiation Time (Max location) T/4 T/2 3T/4 End of Cycle 5 5.35E-3 3.26E-3 1.88E-3 1.02E-3 (1802.9 EFPD) End of Cycle 10 1.39E-2 8.54E-3 4.95E-3 2.71E-3 (3631.2 EFPD) 15 EFPY 1.44E-2* 8.82E-3* 5.13E-3* 2.80E-3* l 21 EFPY 2.20E-2* 1.35E-2* 7.87E-3* 4.29E-3* 24 EFPY 2.60E-2* 1.59E-2* 9.25E-3* 5.05E-3* 32 EFPY 3.62E-2* 2.22E-2* 1.29E-2* 7.05E-3*
- Calculated using these I
Conversion Factors Fluence (E > 1 MeV) 1.51E-21** 1.75E-21** 2.23E-21** 2.82E-21** to DPA. 2 I
** Fast fluence values (E > 1 MeV) in units of n/cm were multiplied by these factors to obtain the corresponding DPA values.
Note: DPA results from cycles 1 to 10 represent the azimuthal peak at one degree off the major axis. DPA results after cycle 10 represent the azimuthal peak at 27 degrees off the major axis. I I 6-10 gw==_ g
Table 6-5. Fluence.- Flux. and DPA for Surveillance Capsules at 104 & 97 Dearees Flux (E > 1 MeV), Fluence, Capsule Irradiation Time n/cm'- s n/cm' DPA
- I 104 Cycles 1-5 2.24E+10 3.49E+18 5.15E-3 (1802.8 EFPD)
> Cycles 6-10 3.40E+10 5.36E+18 7.69E-3 (1828.3 EFPD) Total -------- 8.84E+18 1.28E-2 97 Cycles 6-10 4.46E+10 7.04E+18 1.00E-2 (1828.3 EFPD) Total -------- 7.04E+18 1.00E-2
,N_gle: The results of each capsule were corrected with its respective measured to calculated activity ratio as detailed in appendix D.
I I I 6-11 B W!!s na h r g
Table 6-6. Surveillance Caosule Measurements Heasured/ Caosule Calculated AN1-C 1.04 RSI-F 1.03 CR3-F 0.99 OCl-C 1.01 OC2-E 0.98 DB1-LG1 1.08 CR3-LG) 1.06 0C3-D 1.00 TEl-D 1.03 STL-83 1.08 ## SH-U 0.88 Z10N-1 1.11 MS2-104 0.99 HS2-97 0.94 5 Average M/C for 14 surveillance data points = 1.02 E
'l Sigma standard deviation of data base - 0.06 l
l I 6-12 l B Wllnani W r 1 E'
=;
l
I I fioure 6 2. Fast Fluence and DPA Distribution Throuch Reactor Vessel Wall 1,000 ^ LF. = 1.62
'Q l LF. = 2 00
_ E '
, /4 '
0.500 - I h, , t+ .
%'f -
LF. = 5.12 2( I 0.300 - 221.54 cm inside Surface s c S,' o.d Bese u +.aiinterface ' I c. 2 0.200 A N.A
' t.
$ T/4 '
*C 227.10 cm O
$ LF. = 1.89 L' ',
d I 0.100 - ' A, ' -, , 3 m C
- . T/2 ,
@ 232.67 cm LL -
LF, = 4.14 0 , I $ 0.050 - e CC ~ J 0.030 - 35 238 23 cm LF. = 9 00 0.020 - Fluence DPA
- ; .../_,... Outside Surface 1
243.80 cm 0.010 I 220 225 230 235 240 245 Radial Distance From Core Contor (cm) l~I I I 6-13 eawsucLEAR WSERVICE COMPANY I
I. Figure 6-3. Azimuthal Flux and fluence Distributions I. at Reactor Vessel inside Surface 1.10 - 1.00 0.90 - h 0.00 - .I g E f 0.70 0.00 ~ 0.50 - 0 10 20 30 40 50 Degrees From Major Axis Basis Cys 110
.--+--
1.10 5 1.00 - 0.90 - g 5 C 0,80 - E 0.70 0.60 -
.~
/ E 3
w E 0.50 + o 10 20 30 40 50 Degrees From Major Axis Basis - Cy 10 Only g 6-14 G WitnEWa % r E
m m M M
~
M W W Fiqure 6-4. Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux and Fluence 1.2 _ 1.1 1,
- L' .,
"-n A - *,a . , ; , g7 - .
5 , ."
~n tu : ,#
O 1*0 I- - S' !
' '"~ a --Er- a--;.. s..s,--L . g ._p,_,.__ h**'
5 c __,.,_ 3..ar _.J : I ti. -
'3 R : e y 0.9 --
if g e o g G y 3 y Note: Both cu ves are based on axial power distributson data g tr : < ,1 supplied by the customer. The cyc'e 10 curve represents
= E the low-leakage core design. ,, 5 0.8 --
E n z - o e tu : n *_ 2 : $ H - 5 : tu 0.7 - a- : 0.6 u Cycles 1 to 10 Cycle 10 g } u
....,s....
. I ' ' ' ' I ' ' ' 'I ' I ' ' ' ' I ' ' ' '
' ' ' ' I ' ' ' ' I ' ' ' '
0.5
-150 -100 -50 0 50 100 150 200 ne -200 MC gqo DISTANCE FROM CORE MIDPLANE (cm)
Eh tra 4
.=
..-. - - -._ . .- _ - - . - . . . . . . . ~ . . _ . . - . . - . . - . . . . . . . - . . . - . . _ _ - . .
I 1 l l I l I I ! I i l Page Intentionally left Blank I! l f I' I E1 E' I I I I. I IS101tiSW NUCLEAR I3WSERVICE COMPANY
==
I I I
- 7. DISCUSSION OF CAPSULE RESULTS 7.1. Pre-Irradiation Property Data The weld metal and base metals were selected for inclusion in the surveillance program in accordance with the criteria in effect at the time the program was designed for Hillstone Unit 2. The applicable selection criterion was based on I the unirradiated properties only. A review of the original unirradiated properties of the reactor vessel core beltline region materials indicated no l significant deviation from expected properties except in the case of the upper-shelf properties of the base metals which were close to and in one case below the current required 75 ft-lbs. Based on the design end-of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of the base metals, it was predicted that the end-of-service Charpy upper-shelf energy (USE)
I will not be below 50 ft-lb. L2. Irradiated Property Data 7.2.1. Tensile Properties l Table 7-1 compares irradiated properties from Capsule W-104 with the unirradiated tensile properties. At both room temperature and elevated temperature, the ultimate and yield strength changes in the base metal as a result of irradiation and the corresponding changes in ductility are within the limits observed for similar materials. There is some strengthening, as indicated by increases in I ultimate and yield strengths and decreases in ductility properties. All changes observed in the base metal are such as to be considered within acceptable limits. The changes, at both room temperature and 550F, in the properties of the base metal are slightly larger than those observed for the weld metal, indicating a g slightly greater sensitivity of the base metal to irradiation damage. In either case, _ the changes in tensile properties are insignificant relative to the I I 7-1 B ill E! 5 E E EiI S m
I analysis of the reactor vessel materials at this time period in the reactor vessel service life. A comparison of the tensile data from the previously evaluated capsule (Capsule I W-97) with the corresponding data from the capsule reported in this report is g shown in Table 7-2. The currently reported capsule experienced a fluence that 3 is approximately two times greater than the first capsule. The general behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction of area. The most significant observation from these data is that the base metal exhibited slightly greater sensitivity to neutron radiation than the weld metal. 7.2.2. Impact Properties The behavior of the Charpy V-notch impact data is more significant to the calculation of the reactor system's operating limitations. Table 7-3 compares the observed changes in irradiated Charpy impact properties with the predicted changes. The 30 ft-lb transition temperature shift for the base metal is in relatively good agreement with the value predicted using Regulatory Guide 1.99, Rev. 2 18 and when the margin is added the predicted value is conservative. It would be expected that these values would exhibit good agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken at the 5 30 ft-lb temperature. E The transition temperature measurements at 30 ft-lbs for the weld metal is not g in good agreement with the predicted shift using Regulatory Guide 1.99, Revision u 2 but the predicted value is conservative. The shift being over estimated with the predicted value which indicates that the estimating technique based on the Regulatory Guide 1.99, Rev. 2, is overly conservative for predicting the 30 ft-lb transition temperature shifts. Since the method requires that a margin be added to the calculated value to provide a conservative value, the final shift values l using Regulatory Guide 1.99, Revision 2 should be based on Position 2 which will help to account for some of the over-conservatism in the application of 4 Regulatory Guide 1.99, Position 1. 7-2 BW#sME%v I
t 'I The data for the decrease in Charpy USE with irradiation showed poor agreement with predicted values for the base metal and was under predicted by 22 percent. I The weld metal decrease in Charpy USE was over predicted by 108 percent. However, the poor comparison of the measured weld metal data with the predicted value is to be expected in view of the lack of data for low , or medium-copper-content materials at medium fluence values that were used to develop the estimating curves. A comparison of the Charpy impact data from the previously evaluated capsule from Millstone Unit 2 with the corresponding data from the capsule reported in this report is shown in Table 7-4. The currently reported data experienced a fluence that is two times greater than the first capsule. The base metal exhibited transition temperature shifts at the 30 ft-lb level for the latest capsule that were similar in magnitude relative to fluence to those I of the previous capsule (Capsule W-97). The corresponding data for the weld metal also showed a further increase at the 30 ft-lb level as compared to the previously reported increase at the 30 ft-lb level. This is in contrast to the fact that no further decrease in the upper-shelf energy was observed. in fact, the weld metal exhibited a small increase in upper-shelf energy values while the base metal exhibited no further change. Both the base metal and the weld metal exhibited decreases in the upper-shelf values similar to the previous capsule. The weld metal in this capsule exhibited a small increase as compared to the weld metal in the previous capsule. This small apparent increase may be an artifact attributable to the variations in the testing techniques between two separate test sites. These data confirm that the I upper-shelf drop for both the base metal and the weld metal may be reaching a stabilized condition, or Saturation" as observed in the results of capsules evaluated by others.I9 This bahavior of Charpy USE drop for this weld metal should not be considered indicative of a similar behavior of upper-shelf region fracture toughness properties. Although the Charpy USE appears not to be decreasing, there is no evidence at this time that the USE fracture toughness properties are not continuing to degrade. The observed behavior indicates that other reactions may be taking place within the material besides simple neutron I I 7-3 13W!!#aafLv
l I! damage. Verification of this relationship must await the evaluation of the data from other surveillance capsules. Results from other surveillcqce capsules also indicate that RT NDT estimating curves have greater inaccuracies than originally thought. These inaccuracies are a function of a number of parameters related to the basic data available at the tie the estimating curves are established. These parameters may include inaccurate fluence values, inaccurate chemical composition values, and variations in data interpretation. The change in the regulations requiring the shift measurement to be based on tne 30 ft-lb value has minimized the errors that resulted from using the 30 ft-lb data base to predict the shift behavior at 50 ft-lbs. The design curves for predicting the shift will continue to be modified as more data become available; until that time, the design curves for predicting the RT NDT shif t as given in Regulatory Guide 1.99, Revision 2, are considered adequate for predicting the RT NOT shift of those materials for which data are not I available. These curves will be used to establish the pressure-temperature 3 operational limitations for the irradiated portions of the reactor vessel until the time that improve / rediction curves are developed and approved. The relatively gor cement of the change in Charpy upper-shelf energy is in support of the acy of the prediction curves for medium copper content materials. Howe. , for low copper content m:.terials such as wold metal the predicted values ay oe too conservative. Although the prediction curves are conservative in that they generally predict a larger decrease in upper-shelf energy than is observed for a oivan fluence and copper content, the conservatism can unduly restrict the operational limitations. These data support the contention that the upper-shelf energy drop curves will have to be revised as more reliable data become available; until that time the design curves used to predict the decrease in upper-shelf energy of the controlling materials are considered conservative. 7.3. Reactor Vessel Fracture Touchnes1 An evaluation of the reactor vessel end-of-life fracture toughness was made and the results are presented in Table 7-5. l 1 l l
$$!$IYO ANY i
E
I The fracture toughness evaluation shows that the controlling base metal may have a 1/4 wall location end-of-life Rino, of 156f based on Regulatory Guide 1.99, I Revision 2, including a margin of 34f. lhe controlling weld metal may have a 1/4 wall location end-of life RTnn, of 141f based on Regulaitory Guide 1.99, Revision 2, including a margin of 56f. These predicted shifts are excessive since data from the suricillance capsules exhibit measured Rlnn, values significantly less for comparable fluence values, it is estimated that the end of-life RTuo, shif t for both the controlling base metal and weld metal will be significantly less l than the value predicted using Regulatory Culde 1.99, Revision 2. 1his reduced shif t will permit the calculation of less restrictive pressure temperature operating limitations than if Regulatory Guide 1.99, Revision 2, was used. An evaluation of the reactor vessel end-of-life upper-shelf energy for each of the materials used in the reactor vessel fabrication was made and the results are presented in Table 7-6, 1his evaluation was made because the base metals used . to fabricate the reactor vessel are characterized by relatively low-upper-shelf I energy and moderate copper contents; and consequently, are expected to be sensitive to neutron radiation damage. The method used to evaluate the radiation induced decrease in upper-shelf energy is the method defined in Regulatory Guide 1.99, Revision 2, which is the same pro:edure used in Revision 1. Thi method of Regulatory Guide 1.99, Revision 2, shows that the base metals used in the fabrication of the beltline region of the reactor vessel will have an upper-shelf energy greater than 50 ft lbs through the 32 Ef PY design life based on the T/4 wall location. Regulatory Guide 1.99 method also predicts an upper-shelf energy above 50 ft-lbs for the controlling base metal at the vessel inside I wall, The weld metal t,pper-shelf energies unirradiated values are so high as to preclude any chance of the values decreasir.g below 50 f t-lbs during the 32 EfPY design ife. Based on surveillance capsule data, it is estimated that the controlling vessel base metal upper-shelf energy will remain above the reoutred l 50 ft-lbs during the vessel design life. 7.4. MRutran f1unnctlnAhsh The neutron fluente analysis shows a sharp reduction in the neutron flux as the result of improved fuel management schemes to lower core leakage. These new I 13W#EMTJL
I analysis calculated an end-of-life fluence value of 2.40 x 10" n/cm' (E > 1 MeV) at the reactor vessel inside surface peak location. The corresponding value for g the vessel wall T/4 location it alculated to be 1.27 x 10" n/cm' (E > 1 MeV). E These values represent a 19 percent reduction compared to the values calculated based on the dosimeters and analysis as reported for Capsule W-97.2 I I I I I I I ni E I I I I 7-6 B W!!nEnt!eLur E
I If ' a 7-1. Comparison of Millitone Unit 2. Capsule W-104 Tension Test RentLt1 I Room Temo Test Unitt** 1rtitd Elevated Tgmo Test
- Unirr** 1rrld Base Metal -- C5667-1. lonaitudinal I Fluence, 10 18 n/cm' (E > 1 MeV) 0 8.84 0 8.84 l Ultimate tensile strength, ksi 0.2% yield strength, ksi 85.7 63.5 103.0 79.4 82.9 55.3 95.5 64.2 Uniform elongation, % 11.7 12.4 10.1 10.4 Total elongation, % 29.0 25.9 25.7 21.4 Reduction of area, % 70.7 62.1 69.4 58.0 Base Metal -- Heat-Affected..lgng Fluence, 10 18 n/cm' (E > 1 MeV) 0 8.84 0 8.84 Ultimate tensile strength, ksi 87.6 100.0 83.1 96.0 l
0.2% yield strength, ksi 63.7 79.7 61.7 73.5 Uniform elongation, % 8.9 9.6 7.3 9.8 Total elongation, % 23.7 23.5 20.7 20.1 Reduction of area, % 70.1 66.8 67.8 54.8 Weld Metal -- 10137/3999 Fluence, 10 18 n/cm' (E > 1 MeV) 0 8.84 0 8.84 Ultimate tensile str<.gth, ksi 85.8 99,0 84.9 95.3 0.2% yleid strengtt, ksi 73.0 85.3 66.7 77.0 Uniform elongation, % 11.2 11.2 9.9 8.5 Total elongation, % 28.0 26.3 24.7 20.8 , Reduction of area, % 74.2 64.5 65.3 52.7 I
- Test temperature is 550F.
** Average of the lower yield strength data in Appendix B.
7-7 y sw==_
Tabl e_7-2. Suminary of Millstone t' nit 2 Reactor Vessel Sur<eillance Crosules Tensile Test Results
. Ductility. %
'trength. ksi Cap. F Test Total Reduction Material I.D. 10}gence,2 n/cm Temp, F Ultimate MI *I Yield W.I"} Elon. M I*) of Area M I*)
Base metal --
- 00. 71 83.7 53.5 --
29.0 -- 10.7 -- Longitudinal 550 82.9 55.3 -- 25.7 -- 69.4 -- (C5667-1) W-97 3.78 72 96.0 +12 73.3 +15 26.0 -10 67.0 -5 550 86.7 +1 61.7 +12 27.0 +5 64.0 -8 W-104 8.84 70 403.0 +20 7. ' ; +25 25.9 -11 62.1 -8 550 95.5 +15 64.2 +16 21.4 -17 58.0 -16
~ Base metal --
- 00. 71 87.6 --
63.7 -- 23.7 -- 70.1 -- E Heat-affected 550 83.1 -- 61.7 -- 20.7 -- 67.8 -- Zone (B7835-1) W-97 3.78 72 100.1 +14 75.8 +19 27.0 +1A 68.0 -3 550 87.8 +6 63.4 +11 22.0 +6 66.0 -3 W-104 8.84 70 100.0 +14 79.7 +25 23.5 -1 66.8 -5 550 96.0 +16 73.5 +15 20.1 -3 54.8 -19 Weld metal --
- 00. 71 85.8 --
73.0 -- 28.0 -- 74.2 -- (10137/3999) 550 84.9 -- 66.7 -- 24.7 -- 55.3 -- W-97 3.78 72 102.1 +19 86.8 +19 27.5 -2 70.0 -6 g 550 90.4 +6 72.8 +9 22.5 -9 66.0 +1 O W-104 8.84 70 99.0 +15 85.3 +17 26.3 -6 64.5 -13 sE 550 95.3 +12 77.0 +15 20.8 -16 52.7 -19 R$ h E"
" Change relative to unirradiated.
5 sum uns ums num ums sua muu em 1mse uma aus uma imme uma sus met sur imms
M M M M M M M M M M M M M M M M M M Table 7-3. Observed Vs. Predicted Changes for C gsule W-104 Irradiated 2 Charpy Impact Properties - 8.84 x 10 n/cm (E > 1 MeV) Difference Predicted Per R.G. 1.99/2 Observed
- WithogAI With Material Unirrad. Irrad. Diff. Margin Margin (b)
Increase in 30 ft-lb Trans. Temo.. F Base Material (C5667-1) Lorgitudinal . +38 +137 99 96 130 Heat-Affected Zone (C5667-1) -30 + 15 45 9s 130 Weld Metal -30 + 28 58 Ill 167 Correlation Material (HSST PL-01) +24 +165 141 98 132 Decrease in CharDY USE. ft-lb Base Material (C5667-1) Longitudinal 131 94 37 N.A. 29'" Heat-Affected Zone (C5667-1) 129 104 25 N.A. 29'd Weld Metal 132 108 24 N.A. 50'd D Correlation Material (HSST PL-01) 142 94 48 N.A. 32'd 6 m. kk* I*)Mean value per Regulatory Guide 1.99, Revision 2, May 1988.
"3 E (b)Mean value per Regulatory Guide 1.99, Revision 2, May 1988, plus margin.
{g (c) Bounding value per Regulatory Guide 1.99, Revision 2, May 1988 (includes margin). k N. A. - Not applicable.
Table 7-4. Summary of Millstone Unit 2 Reactor Vessel Surveillance Capsules Charpy Imact Test Results Transition Temperature Upper-Shelf Energy, Increase. F ft-lb F 30 ft-lb Predicted 30 ft lb Ot, served Predicted Material 10]gence,2 n/cm Observed W/0 Margin" W/ Margin" USE tESE USE odSE" Base material (C5667-1) longitudinal 3.78 70 73 107 94 37 107 24 8.84 99 96 130 94 37 102 29 Heat-affected zone 3.78 94 73 107 91 38 105 23 8.84 45 96 130 104 25 100 29 [ Weld metal 3.78 76 75 131 98 34 91 41 o 8.b4 58 111 167 108 24 82 50 5"}Mean value per Regulatory Guide 1.99, Revision 2, May 1988. (b)Mean value per Regulatory Guide 1.99, Revision 2, May 1988, plus margin. (c) Bounding value per Regulatory Guide 1.99, Revision 2. l N EE
==
b
- s l iii l
E" m m m m e e m BB W W W W W m m m W W
W M M M M W M M M M M M M M M M M M M Table 7-5. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Fracture Toughness - Millstone Unit 2 Material Estimated EOL Fluence
- Chemical End-of-Lt fe RT., F'"
Material LVscrietten Ceeposition w/o** Inside T/a Wall Feb. Mat *1. Reacter vessel Heat Surface Locatics initial leside T/s Wall Code Beltilr.e Location Amber Type Cepper Ricket n/cz' n/cs' RT . , F Surface Location C-505-1 Intemed. Shell C5843-1 SA533, Gr.B 0.13 0.64 2.29E+19 1.21E+19 +5 152 136 C-505-2 Interwed. Shell 05843-2 SA533. Gr.B 0.13 0.64 2.29E+19 1.21E+19 +25 ITZ 156 C-505-3 Intened. Shell 05843-3 SA533, Gr.B 0.13 0.65 2.52E+19 1.33E+19 0 149 133 C-506-1 Lower Shell C5667-1 SA533. Gr.B 0.14 0.61 2.40E*19 1.27E*19 +6 164 147 C-506 2 Lower Shell C5667-2 SA533. Gr.B 0.14 0.61 2.40E+19 1.27E+19 -30 128 111 C-500 3 Lower Shell C5510-1 SA533. Gr.B 0.13 0.70 2.40E*19 1.27E+19 0 150 134 8-203 Upper Circum. Weld 33A277/ ASA Weld / 0.30 0.18 3.55E+18 1.8?E+18 M 79 56 3922 L1ade 0091 8-203 Upper Circum. Weld 10137/ ASA beld/ 0.23 0.06 3.55E+18 1.88E*18 -60 72 55 3999 Linde 0091 y 9-203 Mid. Circus. Weld 10117/ ASA Weld / 0.23 0.06 2.40E+19 1.27E+19 -55 132 114 1 3999 Linde 0091
~
9-203 Mid. Cirtum. Weld 9PIM/ ASA Weld / 0.30 0.06 2.40E+19 1.27E*19 -60 164 141 3998 Llode C091 2-203-A.-8.-C Intermed. Longit. Weid #B746/ ASA Weld / 0.12 0.20 1.95E+19* 1.03E+19 -50 91 79 3878 Linde 124 3-2 1-A.-B.-C Lower Longit. Wald A3746/ ASA Weld / 0.12 0.20 1.95E+19* 1.03E 19 -50 91 79 3878 Llode 124
'"Per Regulatory Guide 1.99. Revision 2. N y 1938.
*Per Section 6 of this report using neutron transpoet calculation methods.
Waterials Chestcal Compositions Per Certified Paterial Test Eeperts" and Pressurfred Thersal Shock Reports."
** Fluence value for longitudinal weld with maximus value.
N EE 4 EE P4 Bn s B"
=
s
Table 7-6. Evaluation of Reactor Ve55el End-of-Life (32 EFPY) Upper-Shelf Eneruy - Millstone Unit 2 P'aterial Estimated Chemical EOL Fiume" Estimated EOL-USL Estimated EFPY to Composition, Per RG 1.99/2" 50 ft-Ibs fL41 trial _[33cr f etion w/e" Inside T/4 Wall Initial Fab. Mat *1. Reactor Yessel Heat Surface locatten 05E" != side T/4 Wall Inside T/4 Wali Cod
- Beltilne locatics Nuotwr Type Copper Michel n/ca' n/ca' It-1bs Surface Location Surface location C-505-1 Intenned. Shell C5843-1 SA533 Cr.8 0.13 0.64 7.29E+19 1.21E+19 76 56 59 >32 >32 C-505-2 Intermed. Shell C5843-2 SA533, Gr.B 0.13 0.64 2.29E+19 1.2iE+19 19 58 61 >32 >32 C-505-3 Intermed. Shell C5843-3 5A533 Gr.8 0.13 0.65 2.52E+19 1.33E*19 TT $5 59 >32 >32 C-506-1 tower Shell C5667-1 5A533, Sr.8 0.14 0.61 2.40E+19 1.27E*19 73 53 55 >32 >32 C-506-2 tower Shell C5667-2 SA533, Gr.B 0.14 0.61 2.40E+19 1.27E 19 86 62 65 >32 >32 C-506-3 tower shell C5518-1 SA533, Gr.B 0.13 0.70 2.40E+19 1.2TE+19 88 64 68 >32 >32 6-203 Upper Circun. Weld 33A277/ ASA Weld / 0.30 0.18 3.55E+18 1.88E+18 (130)** 85 91 >32 >32 3 22 Linde 0091
% 8-203 ttpper Circus. Weld 10!37/ ASA Weld / 0.23 0.06 3.55E*18 1.88E+19 130 12 97 >32 >32 3999 Linde 0091 N
9-203 Mfd. Circue. Weld 10137/ ASA Weld / 0.23 0.06 2.40E+19 1.27E*19 130 72 T9 >32 >32 3999 Linde 0091 9-203 Mid. Circus. Weld 90136/ ASA Weld / 0.30 0.06 2.40E+19 1.27E*19 130 60 70 >32 >32 3998 Linde 0091 2-203-A-B-C Intermed. Longit. Weld AS746/ ASA Weld / 0.12 0.20 1.95E*19 1.03E+19 ( 93)" 65 69 >32 >32 3878 Linde 124 3-203 A-B-C tower lengtt. Weld A8746/ ASA Weld / 0.12 0.20 1.95E+19 1.03E+19 ( 93)" 65 69 >32 >32 3878 Linde 124 "Per Regulatory Golde 1.99. Revision 2, May 1988."
"Per Section 6 of this report. " Materials cheetcal compositions per Certified Material Test Reports" and Pressurtred Thermal Shock Reports."" '" Estimate based en values given fee the other weld metals fabricated with Linde 0091 **1d fler.
N " Estimate based en values listed in Embrittlement Data Base'* fer Linde 124 weld metals. N b 4 F 6 5
r-~ l I I Figure 7 1. Comparison of Unirradiated and Irradiated Charpy impact Data Curves for Plate Material I U. . ~ Lonaitudinal Orientation. Heat flo. C56671 i i i i
", n I Unitradiated - -
~
g g ,
- Fluence, B 84 X 10 n/cm' _
R ,. .
/
/ , i i i i i I
0 .: i i i i i . 5
~
s I 2 0.:8 E Unitradiated - k0,06 -
-*- Fluence 8 64 x 10 n/cm'
}o.a. -
_ 110F 3 s2 - i i ! i i
- 0 i i i i 6 i 200 -
18; - 3 IfC 7
~
I
, 14; -
?
100 Unitradiated - SUSE = 37tt-)bs I 3 100 E
~
j- -
~
I 80 W - Fluence, 8.84 X 10 n/ cm' 5 ~ 6*. 120F I 40 - l,99F
= tg t s4533c,ayt3 2; - Foutwet See Above .
II I t i ! HtAt ha. i C6667-1 i 1)Q Q 100 200 300 400 500 60) Test Traperature, F 7-13 G W U 5N NE0 k r
I Figure 7-2. Comparison of Unirradiated and Irradiated Charpy impact Data Curves for Heat Affected Zone. Heat No. C5667-1
- L: - , .
1 f . 1 Unitradiated
~
75 "' g - Fluence. 8 84 X 10' n/cm' 3 y 5; f:! - - I t f 1 1 1 A i I . I I I
~, g';g _ Unitra diated -C _ !e [ fluence. 8 84 X 10* n/cm' E
g R0.;s .-2 0.04 - 3 -#- 69F
/
3 - s 0 :- i i t i t i g
;20 . . i i i i
~
200
? 83 g 160 g
s E
. 1u: -
E -
- Unitradiated ~
g 120 AUSE = 25f t-Ib8 g 100 - C j gg . Fluence. 8.84 x 10 n/cm' , y 3
-v C -
5 L- 71F 40 - ep 45F y,,,, g g s4533 c,siman 20 - Fouemeg See Above . HEAT No. C5667-1 0 300 400 saa (,,y)
*100 0 100 200 Test interatu re, F 7-14 BW!!sefSshr E
l I I Figure 7 3. f.omparison of Unirradiated and Irradiated Charpy Imoact Data Curves for Weld Metal 10137/3999 L. . , i i i i
- Unirradiated - _
I J
) - Fluence. 6 64 X 10 n/ cm' gn I I :s a
i e i i i I 0.10 i i i i i i g,;g - Unirradiated - l 3,3 - - Fluence. 8 64 X 10 n/cm' - I
}0.04 -
- 67F 3 ,
I 5 0,02 - E
, i i i .i i 3
I ::t . . i i i i 200 lI 180 - o ist - I [140 2
. Unirradiated - oUSE = 24tt-ibs -
l120 l $ g, 100 E
- Fluence, 8.84 X 10* n/cm'
$ 80 -
I $ I 6w. 61F I 40 -
- :- SbF mtgang M.ID-4 /Lmde 0091 See Above -
00 - FLutact I 0
- 2M 100 0 100 200 HEAT 143.
300 10137 400 500 Test Terceroture, F 7-15 13W!!f4Feb%v I
figure 7-4. Comparison of Unirradiated and Irradiated Charpy g: l Impact Data Curves for Correlation Monitor Material. HSST PL 01. Heat No. A1008-1 i
- .. , , , i i i 3 . Unittadiated -
n 55; -
- - Fluence. 8 84 x 10 n/cm' x ~
25 - g I I ' C D 10 i i i ' ' i 0A8
~ ~ -
E Unitradia ted - 5 - Fluence. 8.84 x 10,, n/cm, _ 50,06 - o c.au - 135F --
~
5 0.02 ~ a , , i 1 ' ' 20 ' ' ' ' i ' 200 g E
~
180 E g
~. It0
~
$ 140 -
f Unirradiated AUSE = 48tt-ibs - 5 j12: g 3 5 a_ 100 N W q J
~
r - 60 147F e - 141F
' runut asst P' 01 Fluter S #DO*'
20 -
/' gg, g, A1008-1
~
l-3 0 2 Tes T rperyture F 7-16 13W!!&Fc2=v E
I I I 8, DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY PRESSURE - TEMPERATURE LlHITS The pressure temperature limits of the reactor coolant pressure boundary (RCPB) of M111stanc Unit-2 are established in accordance with the requirements of l 10CfR50 Appendix G. The methcds and c,*lteria employed to establish operating pressure and temperature limits are described in topical report BAW-10046A.' g The objective of these limits is to prevent nonductile failure during any normal operating condition, including anticipated operation occurrences and system hydrostatic tests. The loading conditions of interest include the following:
- 1. Normal operations, including heatup and cooldown.
l 2. 3. Inservice leak and hydrostatic tests. Reactor core operation. The major components of the RCPB have been analyzed in accordance with 10CfR50, Appendix G. The closure head region, the reactor vessel outlet nozzle, and the l beltline region have been identified as the only regions of the reactor vessel (and consequently of the RCPB) that regulate the pressure-temperature limits. Since the closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure temperature limits of the first several service I periods. The reactor vessel outlet nozzle also affects the pressure temperature limit curves of the first several service periods. This is due to the high local stressos at the inside corner of the nozzle, which can be two to three times the membrane stresses of the shell. After the first several years of neutron .l radiation exposure, the RT NDT of the beltline region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature limits of the RCPB. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by point comparison of the limits I 8-1 OWS$5tM0$my
I: ! imposed by the closure head region, the outlet nozzle, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressures. The limit curves for Millstone Unit 2 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of twenty EfPY. The twenty EFPY was selected as the time period based on beltline region material properties because it represents a logical sequence from the pi nious analysis. The current surveillance capsule (Capsule W-104) was scheduled to be withdrawn at the end of the refueling cycle when the estimated capsulo fluence corresponded to approximately one fourth the inside surface end-t,f-life value. The removal of the thermal shield, in spite of the use of low leakage fuel cycles, caused the capsule fluence to ex eed the original target fluence value and the new value approaches the reactor vessel inside surfacs 15 EfPY fluence or the reactor vessel T/4 fluence at 24 EfPY, Thus, the capsule results provide support data for calculation of operating limits to 70 EfPY. The time difference between the withdrawal of this surveillance capsule and future operating requirements provides adequate time for re establishing the operating l pressure and temperature limits for subsequent periods of operation beyond the g current surveillance capsule withdrawal. The unirradiated impact properties were determined for the surveillance beltline region materials in accordance with 10CFR50, Appendixes G and H. for the other l beltline region and RCPB materials for which the measured properties are not 5 available, the unirradiated impact properties and residual elements, as E originally established for the beltline region materials, are listed in Table g A-1. The adjusted reference temperatures are calculated by adding the predicted W radiation induced RT NDT and the unirradiated RT NDT. The predicted RT I5 l NOT calculated using the respective neutron fluence and copper and nickel contents. Figure 8-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel beltline region wall. The peak fluence for the intermediate shell plate, C5843-3, is greater than the other two intermediate l shell plates because of holes drilled in the core barrel, to blunt cracks, allows fluence streaming to this plate, The supporting information for figure 8-1 is desuibed in Section 6. The neutron fluence values of figure 8-1 are the g a l 8-2 l SWffaMh l
I predicted fluences that have been demonstrated (Section 6) to be conservative. The design curves of Regulatory Guide 1.99, Rev. 2, were used to predict the radiation-induced RTNDT values as a function of the material's copper and nickel content and neutron fluence. l The neutron fluences and adjusted RTNDT values of the beltilne region materials at the end of the twenty full power year are listed in Table 81. The neutron fluences and adjusted RT values are given for the 1/4T and 3/4T vessel wall NDT locations (T - wall thickness). The assumed RTNDT of the closure head region and the outlet nozzle steel forgings is 60F, in accordance with BAW 10046. The I RTuor values selected for calculation of the pressure temperature limit curves are those values which exhibit the highest values at the T/4 and 3T/4 locations. Figure 8-2 shows the reactor vessel's pressure temperature limit curves for normal heatup. This figure also shows the the core criticality limits as required by 10CFR50, Appendix G. Figure 8-3 shows the vessel pressure tempera-ture limit curves for norm 11 cooldown. Figure 8 4 shows the vessel heatup and l cooldown limitations during inservice leak and hydrostatic tests. pressure-temperature limit curves are applicable up to the twenty EFPY as All indicated. Protection against nonductile failure is ensured by maintaining the coolant pressure below the upperlimits of the pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation I are below and to the right of the limit curve. The reactor is not permitted to go critical until the pressure-temperature combinations are to the right of the l criticality limit curve. To establish the pressure temperature limits for protection against nonductile failure of the RCPB, the limit 5 presented in l Figures 8-2 through 8-4 are adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel con-trolling the limit curves. This is necessary because the reactor vessel is the most limiting component of the RCPB. I I I I B-3 [EWSWv&bsv
t\ Table 8-1. Data for Preparation of Pressure-Temperature Limit Curves for Millstone Nuclear Power Station. Unit-2 -- Applicable Throuch 20 EFPY a dsets ledete4 Ad3ested dt , Si, he Iearnflatatles et (ed of** hterlet ident ific al ten ** Core toc at tom testeo et led of b eplane from Serface thee+ set (sespositses* g g ,, g.e g~g g- ygg f eb h t'1. Seit t ene to held hjer Asts sold 1/4 f losere # **"[ 'Y* I I#* * *g Insetten g/cn*'* C*"*** "'*"'I*** Meet Be. Ret ten tetet ten ** <a Degrs+* toes Tree 8 64 92 e5 th 59 34/34 IIS 9e 3.3gg,gg 8.13 C 505 1 (M43 I *tS33, Ga . 3 letecued hell eu g2 +25 06 59 M/M (let j"{ lle g** g g, 3.13 (-505 2 tus3 2 bA533. Gr.S lateceed. hell go e 65 92 e as 62 34/34 fu g .g g, s.33 C 5e5-3 (5843-3 5A533. 6r.3 laterned. Shell . see e6 9e M M/34 334 g.3ag,;g 9 39 8 63 les (.506 1 (Suf l SA533, Gr 3 1***r 5 hell . S le S 61 POS - 30 9e u M/M 9e Fe g _3sg.g g C 506-2 (SMF 2 SA533, gr.3 go er 3,3s - - 8 89 62 M/M 323 96 9 13 9.79 to 5A533, Gr.3 teser %egl - 3. 3af
- 19 C-506 3 (5518 5 "I I
# 38 O 38 385 88 M II M/II 8 203 33A2Fff ASA geld / gyyer (grgeon, geH --
fn 2 Seiel$ 3922 Ltese ce91 lot 44 44 25 44/28 M 3 yn y geg gg a 23 0 36 8 203 18131/ AM 3 eld / lapper (.lczee. sold . 3999 iSeen 0091 FI i 8.23 9 96 19e -55 ISS FS %/M 898 i ges g _ jeg ,gg 9 203 1811F/ ALA Weld / P&d (tet ess teeld i g 4lede oms 1999 e 8 8e 3M to 824 e9 %/% 824 SS b yes g,3gg,gg 9 38 9 203 901 h/ ASA held / Rid. Itrten aneld 3998 1tese 3091 M el 43 %/48 67 32 yes gg , gg=* $ 12 8 29 12 2 203 A, 9 ( aatso/ ALA gel (/ 3eg, M . Leag64. Weld 8*/30*/ #
'110 tiese 124 %/e n 67 32 9 20 72 Se el el g Sw,tg= 9 Il 3 203 A, - 8, -( aare6/ A u m id/ teser gen,,t E Ids tf/ 38'/60 in 3478 t seee 124
*'RI, caltstated per Segolatory Golde 1.99. Greilles 2. dated Raf IDES."
"'See %ett le 6. Taisle 4-3.
"gY, e. lees eM to taltelate peessere-tempeestere leotts.
"*Por (ertified matertal Test Reports" ansf Pressortaed inereal hect Soports.** **
"Ilseme eales for eeld letatee et signest flamen e 1metten.
re. t-,e.l ..ls i s.e , _ A 3, . g s.at to. e, ,,,, teel e.l.s .l -t s. d= M 8**C 00 0
=
3 m W L I L_J L___J l J Figure 8-1. Predicted Peak Fast Neutron Fluence at Various Locations Through Reactor Vessel Wall for 32 EFPY - Millstone Nuclear Power Station. Unit-2 . 2.5 2.40 X 10 n/cm' 9e Cycles 1 thru 10 - i_ Cycle 11 to EOL + y - Standard Core 1-9
~ '
Low-leakage Cores
- 2.0 -
n Low-leakage Core 10 e. E g G*
& 1.5 -
sd 1.27 X 10 n/cm' c o r /p p x@x60 r
- o 4
h 1.0 - l C toGg\\oD o & J U Tl 5.80 X 10" nicm'- E 413 S c McS *5 - 2 0.5 - T/2 Locag3on 2.50 X 10'* n /cm' 5 vesse\W II Q Z Vessel Wall 3 4T Location
= 0 16 20 24 28 32 !gg 0 4 8 12 nn EFPY $5
Figure 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Heatup, Applicable for First 20 EFPY - Millstone Nuclear Power Station. Unit-2 2400 pres w e. Assurred RT.erf Po.nt ps g Tamp J H L Battiene Rega %T 145 A 487 70 Bett1rne Regvsm MT 118 8 503 105 C 546 140 2000 - o 638 m E 829 225 00 F 1050 250 m 3gn9 _ c 1440 zw C. H 2250 322 J 1050 290
@- K 1440 320 3 2250 3s2 1600 - t T be acceptance pressure temperature cominnator's are teiow ar'd
@ to the right of the !,rrnt cunre(s) The timet curves include the pressure Q- *"*""'#t"'""'*'5"'"'''"'"'*"*"""'** G K c- y 1400 -
the pressure on the reactor vessel reg *on controllmg the istrut cum.
$ C and a% mclude addetional marge of safety for posvble m<trument errw.
_m o 1200 - O tocat on Ad.vstment -33 wg F J _8 1000 - u> a)
> -Criticality Limit g 800 -
E D G o 600 -- c, C)
- _ _
EE c 1 6 R hj 400 - 20F/Hr. 30F/Hr. m gj 200 - 70F 110F i k= _ _ \_140F _ 50F/Hr.
$ 0 ' ' ' ' ' ' '
50 100 150 200 250 300 350 400 Reactor Vessel Coolant Temperature, F i IllE e M M M M M5 m M M m e e e e m a e
M M M M M M M M M M M M M M M M M M M Figure 8-3. Reactor Vessel Pressure-Temperature Limit Curves for Norma) Operation - Cooldown, Applicable for First 2n EFPY - Millstone Nuclear Power Station. Unit-2 2400 Pressure, Temp ,F E
.F Point psig yg _ Assumed RT Bert!ine Rege %T I45 A 434 70 Berthne Regic : MT 118 8 587 150 2000 c 898 210 D 1420 260 g E 2250 303 3 1800 - The acceptance pressure temparatire combinatens are below and to the right of the limit curve (s). The lim t curves include the pressure c) differential between the point of system pressure measurement and it 5
u) 1600 the pre 55ure a the 'e*ctor "'55e! 'esen controrang the lim.t curve. and also include additonal margm of safety for possible instrument I
& vu.
(E 1400 - 1.ocats Adarstment = -33 psis D e u
, y sostrument Error = +6F o 1200 -
O o 3, 1000 u ca
> c
' 800 -
2o cc c) 600 - G
- B EE
- 400 - A EE 58 8? 200 - 70F 200F 3OF/Hr.
300F 80F/Hr.
$h _
20F/Hr. ,_ i i
,li g i i i 50 100 150 200 250 300 350 400 Reactor Vessel Coolant Temperature, F
Figure 8-4. Reactor Vessel Pressure-Temperature Limit Curves for Inservice Leak and Hydrostatic Tests. Applicable for First 20 EFPY - Millstone Nuclear Power Station. Unit-2 Pressure, f 2400 - Assumed RT m.F Point psig Temp,F Bettline Region %T 145 A 667 80 2200 - Bertrine Region %T 118 8 800 140 t20 C 1115 200
'ui D 1687 250
- o. 2000 -
E 2056 270 E af F 2500 290 y 1800 T be acceptance pressure temperature combinations are so. ana us to the right of the limit curve (s). The limit curves include the pressure E d'tferential between the po.nt of system pressure measurement and D CL 1600 the pressure on the reactor vessel region controllitig the limit curve.
~
and also include addRional margin of safety for possible instrument 1400 - error.
? -5 " o O 1200 t=* tea ^om -33 p5's sostrument tuer - +6r e
c e
$ 1000 -
m 800 - O B o - ro 600 - A e m 400 -
.D 200 -
80F SF/Hr. Q . QR 0 I I I I I ' d8 50 100 150 200 250 300 350 400
;;S i Reactor Vessel Coolant Temperature, F ==
lg g % M E E EE
f I G I
- 9.
SUMMARY
OF RESU!.TS The analysis of the reactor vessel material contained in the second surveillance capsule (Capsule W 104) removed for evaluation as part of the Millstone Nuclear Power Station Unit No. 2 Reactor Vessel Surveillance Program, led to the following conclusions:
- 1. The capsule received an aveng,e fast fluence of 8.84 x 10 18 n/cm'(E>
1.0 MeV). The predicted fast fluence for the reactoyg vessel T/4 location at the end of the tenth fuel cycle is 4.88 x 10 n/cm' (E > I 1 MeV).
- 2. The fast fluence of 8.84 x 10 18 n/cm' (E > 1 MeV) increased the RT I of the capsule reactor vessel core region shell materials to a maxib of 137F.
I 3. Based on the calculated fast flux at the vessel wall, an 80% load fact v and the planned fuel management, the projected fast fluence that the Millstone Nuclear Power Station Unit No. 2 reactor pressure vessel ingde surf ace will receive in 40 calendar year's operation is 2.40 x l 4. 10 n/cm (E > 1 MeV). The increase in the RT for the shell plate material was in good agreemtntwiththatpreNtedbythecurrentlyuseddesigncurvesof I RT versus fluence (i.e., Regulatory Guide 1.99, Revision 2), and the pr$Ictiontechniquesareconservative.
- 5. The increase in the RT for the weld metal was not in good agreement withthatpredictedanbhepredictiontechniquesareconservative.
I 6. Neither the base metal nor the veld metal upper-shelf energies at the T/4 location, based on surveillance capsule results, are predicted to decrease below 50 ft-lbs prior to 32 EFPY.
- 7. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in the base metal and the weld metal RTm proper-ties due to irradiation are conservative.
I 8. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in the base metal Charpy upper-shelf properties
.I I
IBW!!nM?! L v I -
I due to irradiation are not conservative but for the weld metal are conservative.
- 9. The analysis of the neutron dosimeters demonstrated that the analytical ,
techniques used to predict the neutron flux and fluence were accurate. I I I I I I I 5 E ! I I' l ( I I 9-2 awama- y
I I I
- 10. SVRVEILLANCE CAPSULE REMOVAL Stil[DULE I Based on the postirradiation test results of Capsule W 104 and the recommended withdrawal schedule of Table 1 of E185" the following schedule is recommended for the examination of the remaining capsules in the Millstone Nvcicar Power Station Unit No. 2 RVSP:
Evaluation ScheduleI ") Capsule locationg{ Lead Removal ExpectedCapgu Identification Capsulesg factorIbI Time fluence (n/cm )g) g W 83 83* 0.92 Cycle 13 1. 3 x 10""'" W 263 263' O.92 Cycle 28 2. 5 x 10**' W-277 277* 0.92 Cycle 43 3. 7 x 10"'" W 284 284* 0.87 Sp are 3.7 x 10"
Reference reactor vessel irradiation locations, figure 31.
'd ihe factor by which the capsule fluence leads the vessels maximum inner wall fluence.
id Estimated fluence values based on current fuel cycle designs.
'd' Approximate fluence at 1/4 wall thickness at 32 EfPY.
'I Approximate fluence at vessel inner wall at 32 EfPY, g '" Approximate fluence at vessel inner wall at 48 EiPY,
Spare capsule to be irradiated and available for an intermediate evaluation, I if data needed, to support licensing requirements. Estimated withdrawal at Cycle 48 to have expected fluence.
I
'I 13W#fM!Lv I
l 1. I I I I I l I I l I l l I Page Intentionally left Blank I I E E I' I g I I I 95 HIB&W NUCLWAR , I3 W SERVICE CUMPANY ame
I I I I 11. CERTif! CAT 10t1 The specimens were tested, and the data obtained from flortheast fluclear Energy Company, Millstone fluclear Power Station, Unit flo. 2, reactor vessel surveillance Capsule W-104 were evaluated using accepted techniques and established standard l methods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H. l /d%A e ~ W:2
~, L( lowe, Jr. , P4.
. . E. 10lV 9)
/ D' ate' Project Technical Manager This report has been reviewed for technical content and accuracy.
l ???n b)C tom 91 M. J. ifevan (Material Analysis) Date M&SA Unit flX 4 m - K. K. Yoon, />.E. (Fracture Analysis) m/M9/ Date M&SA Unit f tIIIo7G ID 'l 9 I L. Petrusha (Fluence Analysis) ' Date Performance Analysis Unit Verification of independent review. I $$4 e /$ h S) ' K.' E. Moore, Manager ~Date M&SA Unit This report is approved for release. (/hhy m' /0lV ff l' L. Baldwin, P.E. Dr.te Program Manager 1 11-1 l BW!!?vMEa%v l 4
I I I I I I I I I Page Intentionally left Blank I I a E I I I I I I 13W#sefaMLw g'
I I I I I I I I APPDiDIX A Reactor Vessel Surveillance Program I Background Data and Information - I I I I I ' 'I I I A-1 13W!!$5EEf$$av
l I L._tMerial Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E185 70, are shown in Table A-1. The lucations of
' these materials within the reactor vessel are shown in figures A-1 and A 2.
2 Definition of Beltline Reaipa The beltline region of Millstone Unit 2 was defined in accordance with the definition given in ASTM E185-73. l L fa nule Identification The capsules used in the Hillstone Unit 2 surveillance program are identified l below by identification, location, and original target fluence.' Target l Capsule Capsule Capsule Approximate Removal Identification loc at i on' Refueling Fluence, n/cm' 1 W 97 97* 8 3.2 x 10 2 W-104 104' 16 5.7 x 10 3 W-284 284* 23 8.3 x 10 4 W 263 263* 30 1.2 x 10 5 W 277 277* 35 1.4 x 10 l 6 W 83 83' 40 1.7 x 10 B E L._Soeciment Per Surveillance Caosule The type and quantity of each material contained in each surveillance capsule is shown in Table A-2. I I I I A-2 awamm. l
M M M M M M M M M M M M M M M M M M M Table A-1. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of 2 Surveillance Wram Materials - Millstene Unit No. 2 '## #' Chagy.,,liggLData Transverse Material 50 35 g y* Chesistry, Fabricator w t*, Material Ident., Beltilne Orep wt Longitudiaal ft-Ib, Mt E. USE, Material Type Region Location T ,F At 10F, ft-1b F F ft-Ib F Cs es t Code Heat Iso. 5A533, Gr S Intermed. Shell -20 --- -- -- 76 +5 0.13 0.64 C-505-1 C5343-1 C5383-2 5A523 Gr B Internad. Shell -10 --- -- -- 79 +25 0.13 0.64 C-505-2 C5843-3 SA533, Gr B Intermed. Shell -10 --- -- -- 77 0 0.13 0.65 C-505-3 SA533 Gr B Lower Shell +10 --- -- -- 73 +10 0.14 0.61 C-506-1 C5667-1 Lewer Shell -40 -- - 66 -30 0.14 0.61 54533. Gr B C-506-2 C5667-2 SA533 Gr 8 Lower Shell -30 --- -- - 83 0 0.13 0.70 C-506-3 C5518-1 Upper circum. -- 151,121.123 -- -- ---
+10 0.30 0.18 8-203 33A277/3922 ASA Weld /Linde 0091 Upper circira. 101.108.107 -- -- --- +10 0.23 0.06
> S-203 10137/3999 ASA Weld /Linde 0091 W Middle circus. --- 101.!08,'.07 -- -- --- +10 0.23 0.06 9-203 19137/3999 ASA Weld /Linde 0091 Middle circum. --- 110,116.107 -- -- ---
+10 0.30 0.06 9-203 90136/3998 ASA Weld /Linde 0091
--- 65.75,78 -- -- ----
+10 0.12 0.20 2-203-A,-8.-C A8746/3878 ASA Weld /Linde 124 Ir.terned. log it.
tower longtt. --- 65.75,78 -- -- --- +10 0.12 0.20 3-203-A.-B.-C A8746/3878 ASA Weld /Linde 124 Survett. weld -60 --- -- -- 130 -55 0.30 0.06 l 90136/3998 ASA Weld /Linde 0091 Surve11. weld -60 -- -- -- 130 -55 C.21 0.06 10137/3999 ASA Weld /Linde 0091 SA533, Gr B Survell. plate -10 --- -- -- 107 +5 0.14 0.61
--- C5667-1 l
N EE .
==
RE 55 82 6
==
Table A-2. Type and Quantity of Speciraens Contained in Each Irradiation Capsule Assembly Base Metal Weld Metal Correl. Target (Heat No CS667-11 (10137/3999)'d HAZ (Heat Material ** Capsule Fluence'd Incact No. C5667-1) - Total Specimns Location (n/cm') T+ Tensile Impact Tensile Impact Tensile Impact Impact Tensile { Vessel 97* 3.2 x 10 12 12 3 12 3 12 3 -- 48 9 Vessel 104* 5.7 x 10 12 -- 3 12 3 12 3 12 48 9 Vessel 284* 8.3 x 10 12 12 3 12 3 12 3 -- 48 9 Vessel 263* 1.2 x 10 '5 12 -- 3 12 3 12 3 12 48 9 Vessel 277* 1.4 x 10 12 12 3 12 3 12 3 -- 48 9 Vessel 83* *7v 10 R H 3 R _3 R 3 -- 48 9 l'TALS 72 48 18 72 18 72 18 24 288 54
'd Adjusted to nearest value attainable during scheduled refueling.
" Reference material correlation amitors.
'd Weld wire / weld flux lot combination.
g L* - Longitudinal ke T+ - Transverse a k$ 83 d a x lE E E EE
I I Figure A 1. Location and Identification of Materials Used in the fabrichtion of Hillstone Unit 2 Reactor Pressuro Vessel" IIE ACTOR VESSEL BELTLINE M ATt RI At S NOTSHOWN INTERMEDIATE SHELL prnnm "*-----"
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WELD $E AM No. 2-203C
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/ INTERMEDIATE CHELL
/ LONGITUDlPJALU.'E LD LE AM No. 2-203-B I If4TERMEDIATE SHELL-PLATE No, C-505 1 INTE RMEDIATE SHE
'PL AT E No, C-605- J INTERMEDI ATE SHELL # '
-N . tillE RMEDI AT .TO t. OWE R I
LONGITUDINAL WELD %:'ff- SHE WELDLL GingtJ,(,AM Ho, e-SE AM No. 2-203- A INTERMEDI ATE LH:'LL PLATE No. C-505-2 --
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No. C-506-3 LOWER CHELL # , g% 4 . LONGITUDINAL WELD # # j<- SEAM No. 3-203-A ._ 1 ,
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I figure A-2. Location of Beltline Region Materials in Relationship to thL Reactor Vessel Core I
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s I I A-6 ' BWitnEM%r E
l I figure A-3. Location of Longitudinal Welds in Hillstone Unit 2 Upper and Lower Shell Courses2 ' I O I [ I I 270 - Y : 90 Middle Shell I I ' q3 180 g 0 i N I 270 Y : 90 I Lower Shell I l 180 I _,
"*'"""~
E
I Fioure A-4. Location of Surveillance Caosule Irradiation Sites in Millstone Unit 2 0 130 Outlet Nozzle
/f ~ ~ ~ ~ ~' ~ ~ 'l\ ;
( . Vessel / 0 104 ,_,,,
/ Id I \ Nozzle 5 Core Sh'roud )
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97 ~ y eactorR Vessel y
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I I !E e 9 >a 1 2 lI k l !I , APPENDIX B Pre-Irradiation Tensile Data 4 s I i 1 e 4 1 E d. N h J i a^
- B-1 rssneswsuctran 1.3 WSERVICE COMPANY ii
I Ii i 1 Table 8-1. Tensile Properties of Unirradiated Shell Plate Haterial. Heat No. C5667-1. lonaitudinal Test Specimen Temp, Strenath. ksi Elonaation. % Reduction of No. F Yield
- Ultimate Uniform Total Area. %
1K3 71 61.8/66.1 84.2 12.3 30.0 71.4 IJJ 71 64.9/68.0 87.4 11.5 28.0 71.4 IKY 71 63.7/67.4 85.4 11.3 29.0 69.4 IK5 250 58.2/60.0 78.0 10.6 28.0 70.0 IJD 250 61.2/62.5 80.9 9.7 25.0 69.4 IKP 250 58.8/61.8 78.9 10.2 26.0 69.4 IJE 550 56.9/ -- 84.6 10.2 26.0 69.4 IKT 550 53.3/54.5 81.7 10.5 25.0 69.4 IKK 550 55.7/ -- 82.4 9.7 26.0 69.4
- Lower and upper yield strengths.
I Table B-2. Tensile Properties of Unirradiated Shell Plate HAZ Material. Heat No. C5667-1. Transverse Test Specimen Temp, Strenath. ksi Elonaation. % Reduction of No. F Yield
- Ultimate Uniform Total Area. %
4K2 71 63.7/66.1 87.6 9.5 23.0 69.4 E E 4JM 71 63.7/68.6 87.3 8.2 23.0 69.4 4JB 71 63.7/68.6 87.8 9.0 .25.0 71.4 4KT 250 60.0/61.8 81.1 7.5 24.0 71.4 4K4 250 59.4/61.2 80.8 8.3 24.0 73.4 4K3 250 57.6/58.2 79.2 7.7 26.0 72.0 4KL 550 62.5/ -- 82.6 6.7 20.0 69.4 4JE 550 62.5/ -- 81.6 6.9 21.0 67.3 4KM 550 60.0/ ~ 85.2 8.2 21.0 66.7
- Lower and upper yield strengths.
I B-2 BW##NiifLv I
I T able B-3. Tensile Properties of Unirradiated Weld Metal 10137/3912 Test Specimen Temp, Strenath. ksi Elonaation. % Reduction of No. .F Yield
- yltfmgit .t)ni f orm Ipigl Area. %
3JU 71 74.7/75.9 87.0 10.8 27.0 73.5 3J3 71 72.2/76.5 85.3 12.2 31.0 75.5 3K6 71 72.2/75.9 85.1 10.5 26.0 73.5 3KB 250 67.8/72.9 79.1 9.7 25.0 71.4 3JA 250 69.8/73.5 80.8 9.2 26.0 71.4 3JK 250 69.8/74.7 81.7 9.0 25.0 71.4 3JM 550 67.4/ -- 84.0 9.3 22.0 63.2 I 3KL 550 66.1/ -- 83.8 86.9 10.6 9.7 26.0 26.0 67.3 65.3 3L4 550 66.7/67.4
- Lower and upper yield strengths.
I I I I I I I - I I I B-3
I I I I I I I I I Page Intentionally left Blank I I E I I I I I I B WUEE?a % v g
I I I I I I I I I APPENDIX C Pre-Irradiation Charpy impact Data" I I I I . I I I I C-1 B WE?5EEES$$aur
Tabla C -1. Charpy impact Data From Unirradiated Base Material, Lonoitudinal Orientation. Heat No. C5667-1 Absorbed Lateral Shear Specimen Test Temp, Energy, Expagsion, Fracture, ID F ft-lb 10 in. % llM -80.0 4.5 3 0 161 -40.0 7.0 9 10 12E 0.0 13.5 16 20 163 40.0 32.0 33 35 13U 60.0 45.5 44 45 15K 60.0 56.0 43 45 12C 70.0 45.0 42 45 110 70.9 77.0 66 60 12D 80.0 62.5 54 67 13E 80.0 79.5 67 70 13K 80.0 89.0 67 70 15L 120.0 116.0 87 85 11E 120.0 119.0 86 90 14Y 160.0 124.5 89 100 15P 160.0 134.5 91 100 13J 210.0 130.0 90 100 152 210.0 136.5 89 100 Table C-2. Charpy Impact Data from Unirradiated Base Metal, Heat-Affected Zone. Heat No. C5667-1 Absorbed Lateral Shear Specimen Test Temp, Energy, Expagsion, Fracture, 10 F ft-lb 10 in. % a E 43C -150.0 10.0 10 0 44U -120.0 16.0 10 10 441 - 80.0 11.5 7 10 434 - 40.0 31.5 23 35 44T 0.0 45.0 35 45 45B 20.0 31.5 29 45 431 ~20.0 96.0 60 65 442 30.0 34.5 31 45 465 40.0 62.5 53 70 3 41A 40.0 101.0 67 70 5 430 40.0 117.5 68 70 465 80.0 111.5 77 85 424 80.0 150.0 88 100 44P 120.0 123.0 86 100 42M 160.0 113.0 81 100 45P 160.0 130.0 88 100 C-2 aw==_ g
~.
I I Table C-3. Charov Imoact Data from Unirradiated Weld Metal. 10137/39]Lt I Specimen ID Test Temp, F Absorbed Energy, ft-lb Lateral Expagsion, 10 in. Shear Fracture, 32L -120.0 5.5 4 0 35A - 80.0 10.0 10 10 3 33P - 40.0 22.0 25 30 E 341 0.0 34.0 34 40 361 0.0 94.5 69 70 355 10.0 77.0 68 65 332 10.0 79.0 66 75 333 10.0 80.5 65 75 33M 20.0 91.0 76 75 E 32M 40.0 99.0 76 80 5 31J 60.0 115.0 88 90 36C 80.0 131.5 96 100 362 80 0 134.5 94 90 I 34Y 33T 12 .o
- 12. 0 129.5 132.5 92 92 100 100 356 160.0 127.0 91 100 31K 160.0 140.5 96 100 Table C-4. Charpy impact Data from Unirradiated Correlation Monitor Material, Lonaitudinal Orientation. HSST Plate 01. Heat No. A1008-1' I Specimen ID Test Temp, F
Absorbed Energy, ft-lb Lateral Expagsion, 10 in. Shear Fracture, 755 -80.0 3.5 3 0 75U -40.0 5.5 7 0 757 0.0 13.5 15 15 .I 75J 75M 0.0 40.0 14.5 39.0 16 35 15 20 75B 40.0 43.5 36 25 I 750 75C 80.0 80.0 69.5 73.5 53 60 45 45 75A 120.0 114.5 79 80 75P 120.0 114.5 85 80 75T 160.0 133.5 92 90 75K 160.0 138.0 89 100 180.0 140.5 89 I 75L 100 756 210.0 142.0 89 100 75E 210.0 145.0 92 100 I I C-3 BWuna?a!!am
Figure C-1. Charpy impact Data From Unirradiated Base Metal (Plate). Lonoitudinal Orientation. Heat No. C5667-1 12 i . i: 75 I g- -
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I I Figure C-2. Charpy impact Data From Unirradiated Heat-Affected Zone Base Metal Heat No. 05667-1 10: ,
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I Fiaure C-3. Char _py Imoact Data From Unirradiated Weld Het.it1 . 10137/3999 10: I
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I I figure C-4. Charpy impact Data for Unirradiated (orrelation Monitor Mderial'. HSST PL-01
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83 WSERVICE COMPANY
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I I l-I APPENDIX D l . Fluence Analysis Methodology i I l5 LI I I I D-1 99 BSW NUCLEAR I 13 WSERVICE COMPANY n._,. _ .-_ . ._ _ _ , . . _ _ _.
I
- 1. Analytical Method A semi-empirical method is used to calculate the capsule and vessel flux. The method employs explicit modeling of the reactor vessel and internals and uses an average core power distribution in the discrete ordinates transport code DOTIV, g version 4.3. DOTIV calculates the energy and space dependent neutron flux for p, the specific reactor under consideration. This semi-empirical method is conven-iently outlined in Figures 0-1 (capsule flux) and D-2 (vessel flux). l The two dimensional transport code D0TIV was used to calculate the energy- and l space-dependent neutron flux at all points of interest in the reactor system. l DOTIV uses the discrete ordinates method of solution of the Boltzmann transport equation and has multi-group and asymmetric scattering capability. The reference calculational model is an R-O geometric representation of a plan view through the reactor core midplane which includes the core, core liner, coolart, core barrel, ,
thermal shield, pressure vessel, and concrete. The material and geometry model, represented in Figure D-3, uses one-eighth core symmetry. In order to include I reasonable geometric detail within the computer memory limitations, the code parameters are specified as P3 order of scattering, S, quadrature, and 47 energy groups. The F3 order of scattering adequately describes the predominately forward scattering of neutrons observed in the deep penetration of steel and water media, as demonstrated by the close agreement between measured and calculated dosimeter activities. The S, symmetric quadrature has generally produced accurate results in discrete ordinates solutions for similar problems, E and is used routinely in the B&W R-O DOT analyses. E Flux generation in the core was represented by a fixed distributed source which the code derived based on a combined '50 and Pu fission spectrum, the input relative power distribution, and a normalization factor to adjust flux level to the desired power density. Geometrical Conflauration for modeling purposes, the actual geometrical configuration was divided into three parts, as shown in Figure D-3. The first part, Model "A," was used to generate the energy-dependent angular flux at the inner boundary of Model "B," which began at the outer surface of the core barrel. Model A included a detailed I D-2 B Wllsu M a un I
E I representation of the core baf fle (or liner) in R-O geometry that has been checked for both metal thickness and total metal valume to ensure that the DOT I approximation to the actual geometry was as accurate as possible for these two very important parameters. The second, Model B, contained an explicit represen-tation of both surveillance capsules and associated components for the applicable time periods. The B&W Owners Group's flux Perturbation Experiment" verified that the surveillance capsule must be explicitly included in the DOT models used for capsule and vessel flux calculations in order to obtain the desired accuracy. Detailed explicit modeling of the capsule, capsule holder tube, and internal components were therefore incorporated into the DOT calculational models. The third, Model "C " was similar to Model B except that no capsule was included. I Model C was used in determining the vessel flux in quadrants that did not contain a surveillance capsule; typically these quadrants contain the azimuthal flux peak on the inside surface of the reactor vessel. An overlap region of approximately 56 cm was specified between Model A and Modeis B or C. The width of this overlap region, which was fixed by the placement of the Model A vacuum boundary and the Model B boundary source, was determined by l an iterative process that resulted in close agreement between the overlap region flux as predicted by Models A and B or C. The outer boundary was placed sufficiently far into the concrete shield (cavity wall) that the use of a
" vacuum" boundary condition did not cause a perturbation in the flux at the points of interest.
Macroscopic Cross Sections Macroscopic cross sections, required for transport analyses, were obtained with the mixing code GIP. Nominal compositions were used for the structural metals. Coolant compositions were determined using the averao- boron concentration over I a fuel cycle and the bulk temperature of the regio .. The core region was a homogeneous mixture of fuel, fuel cladding, structure, and coolant. The cross-section library presently used is the (47-neutron group and 20-gamma group) BUGLE coupled set. The dosimeter reaction cross sections are based on the ENDF/B5 library, and are listed in Table E-3. The measured and calculated dosimeters activities are compared in Table D-1. I I D-3
I Distributed Source The neutron population in the core during full power' operation is a function of g neutron energy, space, and time. The time dependence was accounted for in the analysis by calculating the time-weighted average neutron source, i.e. the g neutron source corresponding to the time-weighted average power distribution. 3 The effects of the other two independent variables, energy and space, were accounted for by using a finite but appropriately large number of discrete intervals in energy and space. In each of these intervals the neutron source was assumed to be invariant and independent of all other variables. The space and energy dependent source function can be considered as the product of a discretely expressed " spatial functien" and a magnitude coefficient, i.e. Sv33,= \ ' P )p x lRPD,,X,) magnitude spatial I Sv%
- Energy and space-dependent neutron source, n/cc-sec, v/K = fission neutron production rate, n/w-sec, Po
- Average power density in core, w/cc, RPD 4 = Relative power density at interval (1,j), unitless, X, = Fission spectrum, fraction of fission neutrons having energy 3 in group "g," g i - Radial coordinate index, j = Azimuthal coordinate index, I g - Energy group index.
The spatial dependence of the flux is directly related to the RPD. Even though the entire (eighth-core symmetric) RPD was modeled in 'the analysis, only the peripheral fuel assemblies contributed significantly to the ex-core flux. A pin-by-pin RPD was not available for Millstone Unit 2, however, a RPO by fuel g assembly was supplied. To obtain a pin-by-pin distribution, it was assumed that 5 the shape of the RPD by fuel assembly was the same as the shape for a similar I D-4 BW!!nEWSi!Lv
I
- g cycle in St. Lucie Unit 2. This was a valid assumption for the following reasons:
- l. The RPD for St. Lucie 2 was on a pin-by-pin basis and representative of fuel cycles with fresh fuel in the peripheral assemblies. Millstone I Unit 2 also used fresh fuel in the peripheral assemblies for cycles 1 to 9. Cycle 10 was the first low leakage cycle at Hillstone Unit 2.
Since, the magnitude of the power in the fresh fuel assemblies is usually higher than the magnitude in once burned fuel assemblies (all I- other factors being equal), the use of the St. Lucie fresh fuel RPDs for cycle 10 of Hillstone Unit 2 was a conservative assumption for approximately 11.5% (cycle 10) of the total EFPDs (cycles 1 to 10). .I 2. St, Lucie 2 is a sister plant of Millstone 2. They are both Combustion Engineering designs and for all practical purposes, core and capsule sizes are identical. Thus, using the St. Lucie-2 power shape and the Hillstone Unit 2 RPDs by assembly for the peripheral fuel assemblies, pin-by-pin RPDs were calculated for Millstone Unit 2 for three time periods, (1) cycles 1 to 5, (2) cycles 6 to 10, and (3) cycle 10. These time-weighted average RPDs were used to generate the normalized I space and energy dependency of the neutron source. Calculations for the energy and space dependent, time-averaged flux were performed for the midpoint of each DOT interval throughout the model. Since the reference model calculation produced results in the R-O plane at the average axial power position, axial correction factors were calculated and used to adjust the results to the capsule elevations. The correction factors were calculated using axial power data for cycles 8 to 10 and then adjusted to represent the axial power distribution for cycles 1 to ** 1.1. Capsule Flux and Fluence Calculation As discussed above, the DOTIV code was used to explicitly model the capsule assemblies and to calculate the neutron flux as a function of energy within the capsules. The calculated fluxes were used in the following equation to obtain calculated activities for comparison with the measured data. The calculated I activity for reaction product 0,, in ( Ci/gm) is: I I I D-5 BWunEW1?!Lr
I I { o,,(E) 4 (E) { Fj (1-e-1 's) e
- d *> >
D= j 2 (3.7x108) An E z , N - Avogrdro's number, Ao - Atomic weight of target material n, fi Either weight fraction of target isotope in n-th material or the fission yield of the desired isotope, o,(E) - Group-averaged cross sections for material n (listed in Table E-3)
$(E) - Group avereged fluxes calculated by DOTIV analysis, I F, - Fraction of full power during j-th time interval, tj g Ai - Decay constant of the ith isotope, T - Sum of total irradiation time, i.e., residual time in reactor, and the wait time between reactor shutdown and counting times, r - Cumulative time from reactor startup to end of j-th time period.
3 t - Length of the j-th time period l 3 Adjustments were made to the calculated dosimeter activities to correct for the effects listed below: N Short half-life adjustments to Ni, Ti, and Fe dosimeter activities E Photofission adjustments to 238 237 0 and Np dosimeter activities Axial correction factor to adjust for ar.ial power distribution After making these adjustments the calculated dosimeter activities were used with the corresponding measured activities to obtain the measured to calculated' l activity ratios or flux normalization factors for each capsule, 104 and 97 degrees: C, - D (measured) 3 D, (calculated) I 0-6 BW!!na?aWr y N
I I These normalization factors were evaluated, averaged, and then used to adjust the calculatcd test specimen flux and fluence for each apsule to be consistent with I- the dosimeter measurements. The maximum normalization factor (104 degree capsule) was then used to adjust the calculated vessel flux and fluence. The flux normalization factors are given in Table D-1,
- 2. Vessel Fluence Extrapolation I for past core cycles, fluence values in the pressure vessel were calculated as described above. Extrapolation to future cycles was required to predict the I useful vessel life. Two time periods were considered in the extrapolation: 1) operation to date for which vessel fluence has been calculated, 2) future fuel cycles which no analyses exist, for the Millstone Unit 2 analysis, time period I was through cycle 10, and time period 2 covered cycles from the end of cycle 10 through 32 EFPY, The flux and fluence for time period 2 was estimated by assuming that the flux at the inside surface of the pressure vessel (PVIS) for future cycles was the same as that calculated for cycle 10. This was a conservative assumption because the cycle 10 fluence calculation was based on the fresh fuel power distribution (shape I only) taken from St. Lucie (a sister plant) while cycle 10 was actually the first low leakage cycle, it was found in the Millstone Unit 2 analysis and is shown in figure 6-12 that the peak fluence at the PVIS occurred at approximately 1 degree off the major axis for cycles I to 10. However, when the cycle 10 RPD l alone was appiled to calculate the fast flux, the peak flux values at the PVIS shifted from 1 degree to approximately 27 degrees. If future cycle designs are similar to that of cycle 10, the peak fluence value would occur at approximately 27 degrees at 32 EfPY, For this reason, the flux used to extrapolate from E0C 10 to 32 EFPY was the flux calculated at 27 degrees and not at I degree. The I cycle 11 and 12 designs are similar to cycle 10 in that tHre is all burned fuel in the peripheral assemblies except for one fresh ass ,1y approximately 25 degrees off the major axis. It is evident that the fresh fuel assembly is responsible for the shift of the flux peak at the PVIS from 1 to 27 degrees. If cycle designs past cycle 12 utilize all burned fuel, the flux peak at the PVIS may shift back towards the 1 degree location. Future analyses will ascertain the actual ef fects.
I D-7 ,g 13W!bsefafLv i
I Table 0-1. Flux Normalization Factor for the 104 Dearee Caosule Measured Calculated Flux E Dosimeter Act i v i ty ,' Activity,*' Normalization"' E Reaction uti/a uti/a Factor 68 Hi (n.p)Co 796.1 922.8 0.863
T i ( n , p) **Sc 157.1 163.2 0.963
F e ( n . p )Mn 790.4 938.S 0.842
Cu(n,y)* Co 9.120 8.190 1.113 assV(n,f)*Cs 13.18 11.40 1.156 Averaged: 0. 99'#
(*) Average of three dosimeters wires. (b) Average at three calculated activities. (c) Ratio of average measured activity to average calculated activity. (d) Average of all five dosimeters was selected as the normalization constant. I E E I I I I I D-8 BWUnE?ti!Lr 1
l i Flux Normalization f actor for the 97 D19r1C_fdplult Table D-1 A. I Dosimeter Reaction Measured Activity," vCi/o Calculated Activity,"d vCi/a flux Normal ization'd factor
N i ( n , p )C o 975.0 1246 0.782
T i ( n , p )Sc 202.2 213.1 0.949
f e ( n , p )Hn 1080 1264 0.854
Cu(n, y)" Co 9.210 8.993 1.024 2U(n , f)C s 11.04 10.44 1.057
N p ( n , f) C s 45.98 47.76 0.963 Averaged: 0 . 9 4'*
(a) Average of three dosimeters wires. (b) Average at three calculated activities. (c) Ratio of average measured activity to average calculated activity. (d) Average of all six dosimeters was selected as the normalization constant. I However, the normalization constant for the 104 degree capsule was larger and therefore more conservative. Thus, it was applied to all vessel flux and fluence results. I . I E I I I I D-9 g I3W!!sefaf42-
I Table 0-2. Millstone Unit 2 Reactor Vessel Fluence by Cycle Vessel Fluence, n/cm* Incremental Cumulative Vessel Flux, Cycle (s) Time, EFPY Time, EFPY n/cm'- s incremental Cumulative l-5 4.94 4.94 2.27E+10 3.52E+18 3.52E+18 I 6-10 5.00 9.94 3.60E+10 5.69E+18 9.21E+18 5.06 15.00 1.881'+10(b) 4.31E+18(c) 0.52E+18(c) 6.00 21.00 1.88E 10(b) 5.llE+18(c) 1.46E+19(c) 3.00 24.00 1.88E+10(b) 2.55E+18(c) 1.72E+19(c) 8.00 32.00 1.88E+10(b) 6.82E+1P.(c) 2.40E+19(c) (a)The normalization constant derived from the 104 degree capsule has been applied. l (b)The fuel cycle design for cycle 10 was used to estimate the maximum neutron E flux at the inside surface of reactor vessel after the E0C 10. 5 (c) Extrapolated values. (d) Peak fluence at inside surface of reactor vessel. I 5 I I I I I D-10 BWitnEniLr E
l I Figure D-1, Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Cansy.11 I ENDF/B4 Cross Sections Goometry & Quadraturo Power Distri. ENDF/85 Dostmotor Roac- for Model A DOT butions Since tion Crots Soctions Capsu!o Insor-tion I Y GIP I v Cross Sections U T U I l DOT 4.3 Model A j* sJ Radial Power Shapo Appiled Goometry & Quadrature Model B V U p. DOT 4.3 Angular Flux
, Model B 4 At Barrel g v Power History -
Activitios I Final
- of Capsulo (PRHIST Codo)
I > Calculated Activltios w I
, Measured Dosimotor Axlal Activities I Correction Factur Y y > Normalization I > Capsulo Flux 4 Factor -
M/C Ratio I I 13W!!sME23=v
figure D-2. Rationale for the Calculation of Neutron Flux in the Reactor Vessel I ENDF/B4 Cross Sections Geometry & Quadrature Power Distri. ENDF/B5 Doslmeter Reac- for Model A DOT butions Since tion Cross Sections Capsule Inser-tion GlP I
" y y 1r Cross Sections l DOT 4.3 Radial Power ,
*
- Shape Appiled Model A Geometry &
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, Model B l
V V y DOT 4.3 Angular Flux Model C 4 At Barrel I Normalization Factor from E Capsule Fluence Analysis (from the Diagram on the g Previous Page) 5 Axla! Cor- tion
, Factor Time Avera ed Vessel Flux
> Maximum Vessel Location c (E R,0)
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l-lH I I I I I I I I APPEt10lX E Capsule Dosimetry Data g i I I I . I I I I E-1 13 W ife effif L v
4 I Table E-1 lists the characteristics of the neutron dosimeters. Tables E 2 and E 3 show the measured activity per gram of target material (i.e., per gram of g uranium, nickel, etc.) for each capsule's dosimeters. Activation cross sections 5 for the various materials were flux weighted with the 235 0 fission spectrum shown in Table E 4. l Table E-1. Detector Composition and Shieldina Detector Material Shieldina . Reaction Ni Wire Cd lli ( n . p)C o Ti Wire Bare T i ( n , p ) *'Sc Fe Wire Bare ** f e ( n , p )"Hn Cu Wira Cd Cu(n, y)' Co U foil Cd '"U ( n , f ) C s l Np foil Cd ilp (n , f)C s . I 5 I I I I I E-2 IB W !!n nY5 % - E
I I labic E 2. Measured Specific Activities (Unadjusted) f.pI_P,giletters in titLLQ4_ Degree _ Ccnig1L Dosimeter Activity, htLil3mDAr9tt) Ettector Main 111 Dailgicr Reittien Unnen Lenin L0xet Ni Wire "fli ( n , p)"Co 852.6 715.8 819.8 l l Tl Wire T i (n . p) **Sc
**f e(n p)"Hn 168.1 147.8 155.5 fe Wire 817.3 771.4 782.5 Cu Wire "Cu (n , y)*Co 8.86 9.04 9.46 V foil "U ( n , f ) "'C s 12.90 12.89 13.76 I Table E-3. Measured Specific Activities (Unadjusted) f oLQqLimtitr.1_ittthtRZ_uesr_c e_LonsalL_
I Detector MainlAl Qosimeter Reati_i.0D Dosimeter Activity, UPTer_. f uC1/anLpl_lgIgt1.) EtnicI LREcL Ni Wire "Ni (n . p)"Co 1028 949.6 947.3 Ti Wire '"T i ( n , p ) **S c 211.5 189.6 205.b fe Wire "fe(n.p)"Mn 1108 1022 1109 Cu Wire "Cu(n, y)" Co 10.19 9.33 8.11 U foil "U ( n , f ) "'C s 11.04 10.38 11.70 Np foti 'Np(n , f)"'Cs 44.85 48.30 44.78 I I I lI I E-3 13 W #JEeFa!# m y
Table E-4. Dosimeter Activation Cross Sections, b/ atom" Group Upper No. Energy (eV) "'Np(n,f) 2U (n, f) "Fe(n,p) **Ni(n,p) Cu(n,a) Ti(n.p) 1 1.733+1 2.507+0 1.215+0 2.803-1 3.215-1 3.641-2 2.407-1 2 1.419+7 2.310+0 1.033+0 4.260-1 4.980-1 4.535-2 2.667-1 3 1.221+7 2.340+0 9.851-1 4.728-1 5.734-1 5.360-2 2.600-1 4 1.000+7 2.326+0 9.933-1 4.769-1 5.971-1 3.842-2 2.356-1 5 8.607+6 2.243+0 9.898-1 4.759-1 5.988-1 1.926-2 2.043-1 6 7.408+6 1.935+0 8.240-1 4.687-1 5.845-1 9.389-3 1.555-1 7 6.065+6 1.516+0 5.588-1 4.266-1 5.141-1 2.956-3 9.645-2 8 4.966+6 1.547+0 5.452-1 3.041-1 3.847-1 4.568-4 3.766-2 9 3.67945 1.633+0 5.292-1 1.998-1 2.424-1 3.600-5 5.573-3 10 3.012+6 1.680+0 5.282-1 1.371-1 1.674-1 5.844-6 4.747-4 11 2.723+6 1.698+0 5.365-1 8.061-2 1.232-1 1.692-6 6.816-6 12 2.466+6 1.695+0 5.398-1 5.715-2 9.340-2 6.645-7 1.100-6 q' 13 2.365+6 1.694+0 5.404-1 5.134-2 8.278-2 4.712-7 3.770-7 e= 2.s46+6 14 1.692+0 5.410-1 4.564-2 7.227-2 3.305-7 3.427-7 15 2.23146 1.677+9 5.358-1 2.892-2 4.600-2 1.124-7 2.326-7 16 1.921+6 1.646+0 4.799-1 8.181-3 2.440-2 1.500-8 8.518-8 17 1.653+6 1.605+0 3.154-1 2.933-3 1.206-2 0 0 18 1.353+6 1.538+0 4.480-2 6.824-4 3.758-3 0 0 19 1.003+6 1.394+0 1.296-2 5.308-5 1.362-3 0 0 20 8.209+5 1.207+0 3.820-3 4.367-6 1.156-3 0 0 21 7.427+5 9.913-1 1.553-3 6.842-7 9.891-4 0 0 22 6.081+5 6.497-1 6.233-4 1.097-7 7.958-4 0 0 23 4.393+5 2.961-1 2.846-4 8.051-8 6.086-4 0 0 24 2.688+5 1.221-1 1.635 4 5.615-8 4.483-4 0 0 U3 25 2.972+5 5.615-2 1.001 1 3.448-8 3.058-4 0 0 EE 26 1.835r5 3.347-2 7.720-5 1.197-8 1.577-4 0 0
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I I I I I I I I APPENDIX f Tension Test Stress Strain Curves I ~ I I I I I I I P., I
I figure F-1. Tension Test Stress Strain Curvr. for Base Metal Plate Heat C5667 1. Specimen No. IJ1. Tested at 70F I Specimen: 1J1 Strength W eld: 79379. - """- UTS! 103033 _ see. I a 2. .
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I Figure F-4. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Tone. Heat C5667-1. Specimen No. 4KC. Tested at 70F I Specimen: 4KC Strength 700-I Yield: 79716. -
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I Figure F-5. Tension Test Stress-Strain Curve for Base Metal Heat-Affected ' Zone. Heat C5667-1. Specimen No. 4J1. Tested at 400f Specimen: 4J1 t i e. - - i Strength Yleid: 71324 - 70s-UTS: 92704 ee. - ([' f see. 0 vt tL x 2
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Strength Yield: 73473 . 700. I UTS: 95980. E en. ,
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I I I I I I I I APPENDlX G Tabulation of Temperature I Distribution in Reactor Vessel Wall I I I I I I I I G1 13W!!?v&f!Ay
I Table G 1. Normal Heatup Temperature Distribution Along Vessel Wall and Uncor.tfLied Pressures Shell Inside 10 of 1/41 3/4T Uncorrected Time Fluid Surface Vessel Location, location, Pressures, .[liini Temo.. F of Clad, f Wall. F F F osia 1.2 70.4 70.4 70.3 70.0 70.0 523 45.2 85.1 85.1 84.0 80.0 75.8 526 62.6 90.8 90.8 89.7 85.2 80.3 527 E 77.7 95.9 95.9 94.6 89.9 84.6 532 5 94.0 101.3 101.3 100.0 95.1 89.6 539 109.0 106.3 106.3 105.0 99.9 94.3 547 124.1 112.1 112.1 110.5 105.0 99.1 555 135.7 117.9 117.9 116.1 109.8 103.1 557 147.3 123.7 123.7 121.8 115.0 107.6 562 157.8 128.9 128.9 127.0 119.8 112.0 569 168.2 134.1 134.1 132.1 124.7 116.6 579 178.6 139.3 139.3 137.3 129.7 121.3 591 187.9 146.6 146.6 144.1 135.0 125.6 595 194.9 152.4 152.4 149.7 139.7 129.3 598 203.0 159.2 159.2 156.3 145.5 134.0 606 208.8 164.0 164.0 161.0 149.7 137.7 614 215.8 169.8 169.8 166.7 155.0 142.3 627 222.7 175.6 175.6 172.4 160.2 147.0 643 228.5 180.4 180.4 177.1 164.7 151.1 659 235.5 186.2 186.2 182.9 170.1 156.1 680 3 241.3 248.2 191.1 196.8 191.1 196.8 187.7 193.4 174.7 160.4 701 3 180.1 165.6 727 254.0 201.7 201.7 198.2 184.8 170.0 753 261.0 207.5 207.5 204.0 190.4 175.4 786 266.8 212.3 212.3 208.8 195.1 179.9 816 272.6 217.2 217.2 213.6 199.8 '184.4 849 285.4 227.8 227.8 224.2 210.1 194.5 931 E 297.0 237.5 237.5 233.8 219.6 203.7 1018 E 309.7 248.1 248.1 244.3 229.9 213.9 1127 321.3 257.7 257.7 254.0 239.4 223.2 1244 334.1 268.4 258.4 264.6 250.0 233.6 1393 346.8 279.0 279.0 275.2 260.4 243.9 1549 358.4 288.7 288.7 284.8 270,0 253.4 1710 370.0 298.3 298.3 294.5 279.6 262.9 1894 382.8 309.0 309.0 305.1 290.2 273.4 2127 394.4 318.7 318.7 314.8 299.7 282.9 2357 407.2 329.3 329.3 325.4 310.3 293.4 2647 I I G-2 13W!!MM&v
~
I I lable G 2. Normal Cooldown Temperature Distribution Along Vesse.1 Wall and _Uncorrecied Prenges _ Shell inside ID of 1/4T 3/4T Uncorrected Time Fluid Surface Yessel location, location, Pressures, Illini lead of Clad. E Wall. F f ._. f osio 194.2 311.1 311.1 317.6 343.2 372.1 2666 I 198.1 202.1 305.9 300.5 305.9 300.5 312.4 307.1 338.0 332.6 366.9 361.5 2517 2375 212.7 294.9 294.9 300.2 321.3 347.6 7236 221.9 290.3 290.3 295.0 313.6 336.7 2' 232.5 285.0 285.0 289.1 305.6 325.7 ...J l 243.0 252.3 279.8 275.1 279.8 275.1 283.5 278.5 298.2 292.1 316.1 308.4 1896 1803 262.8 269.9 269.9 273.0 285.6 300.5 1706 272.1 265.2 265.2 268.2 280.0 294.0 1626 259.9 259.9 273.9 287.0 I 282.7 291.9 255.3 255.3 262.7 258.0 268.8 281.2 1540 1471 302.5 250.0 250.0 252.6 263.0 274.9 1396 313.0 244.8 244.8 247.3 257.3 268.8 1328 322.3 240.1 240.1 242.6 252.4 263.6 1272 l 332.8 342.1 234.9 230.2 234,9 230.2 237.3 232.6 246.9 242.1 257.9 252.9 1212 1163 I 352.7 361.9 224.9 220.3 224.9 220.3 227.3 222.7 236.7 232.0 247.3 242.5 1111 1069 372.5 215.0 215.0 217.4 226.6 237.0 1024 383.0 209.8 209.8 212.1 221.3 231.6 979 402.9 199.9 199.9 202.2 211.2 221.4 896 431.9 190.2 190.2 192.1 199.3 207.8 829 447.8 184,9 184,9 186.6 193.4 201.2 797 462.3 180.1 180.1 181.7 188.2 195.6 771 476.8 175.2 175.2 176.8 183.1 190.2 746 492.7 169.9 I 507.2 165.1 169.9 165.1 171.5 166.6 177.6 172.6 184.5 179.4 721 699 523.0 159.8 159.8 161.4 167.2 173.9 677 'I G3 IBW.!!nEYa!Mm I - -__ _ A -.
I Table G 2. Normal Cooldown Temperature Distribution Along Veisel Wall and Untarmird Fressures (Cont'd) Shell Inside 10 of 1/4T 3/4T Uncorrected Time fluid Surface Yessel location, location, Pressures. Diini Temn.. F of Clad. F hlL_1 F F .,__ psia 537.6 155.0 155.0 156.5 162.3 168.9 658 552.1 150.1 150.1 151.6 157.4 163.9 640 568.0 144.8 144.8 146.3 152.1 158.5 622 l 582.5 140.0 140.0 141.5 147.2 153.6 607 597.0 135.2 135.2 136.7 142.3 148.7 592 612.9 129.9 129.9 131.3 137.0 143.3 578 627.4 641.9 125.0 120.2 125.0 120.2 126.5 121.7 132.1 127.2 138.4 133.5 566 555 l 657.8 114.9 114.9 116.4 121.9 128.1 543 : 672.3 110.1 110.1 111.5 117.0 123.2 533 688.2 104.8 104.8 106.2 111.7 117.8 523 702.7 99.9 99.9 101.4 106.8 112.9 515 717.2 95.1 95.1 96.6 101.9 108.0 507 731.7 747.6 90.3 85.0 90.3 85.0 91.7 86.4 97.1 91.7 103.1 97.8 500 492 l 762.1 80.1 80.1 81.6 86.9 92.8 486 778.0 74.8 74.8 76.3 81.5 87.5 480 792.5 10.0 70.0 71.4 76.7 82.6 474 3 5 I I I I I G4 BWlinMi'sem
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I . t I I I I APPENDIX 11 l Rofcrences I I I I l I li 1 trG W NUCtiAll W88 RVICt CUMPANY _ _ _ . _ . - - . . - - _ . - . . _ , - . . - _ _ . - . . - . . ~ . .
. _ . - - - - _ , . . , - - - - - _ . - - - . - - - - - . . - ~ . . - . - , - , . - . - . - -- --.- -
I
- 1. John J. Koziol, Program for Irradiation Surveillance of Millstone Point Unit 2, Reactor Vessel Materials, N NLM Oll, Combustion Engineering, Inc., g Windsor, Connecticut, October 15, 1970. E
- 2. S. T. Byrne, Post Irradiation Evaluation of Reactor Vessel Surveillance g Capsule W 97, _TR-N MCH 0M , Combustion Engineering, Inc., Windsor, u Connecticut April 1982.
- 3. Code of federal Regulation, Title 10 Part 50, Domestic Licensing of Production and Utilization f acilities, Appendix G. Fracture Toughness Requirements.
- 4. Code of federal Regulation. Title 10 Part 50, Domestic Licensing of Production and Utilization facilities, Appendix M, Reactor Vessel Material Surveillance Program Requirements.
- 5. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Ill, Nuclear Power Plant Components, Appendix G Protection 6.
Against Nonductile failure (G 2000). ASTM Designation A370 68,
- Methods and Definitions for Mechanical Testing l
of Steel Products," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA. l
- 7. ASTM Designation E23 72, " Method for Notched Bar impact Testing of Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelohta, PA. E 5
- 8. Standardized Specimens for Certification of Charpy impact Specimens from the Army Haterials and Mechanics Research Center, Watertown, Mass. 02172, Attn: ORXHR HQ.
l
- 9. ASTM Designation A370 77, " Methods and Definitions for Mechahical Testing of Steel Products," in ASTM Standards, American Society for Testing and l
Materials, Philadelphia, PA.
- 10. ASTM Designation E23-86, " Methods for Notched Bar impact Testing of l
[ Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
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.E s
l I 11. AS1H Designation E185-XX (to be released), Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, in ASTM Standards, American Society for Testing and Haterials, Philadelphia, PA.
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' Bugle 80 Coupled 47 Neutron, 20 Gamma Ray,3 P , Cross Section Library for I 15.
LWR Shielding Calculations," Radiation Information Shielding Center, DLC-75.
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I 17. ASTM Designation E693-79, Characterizing Neutron Exposures in ferritic Steels in Terms of Displacements Per Atom (DPA)," in ASTM Standards. American Society for Testing and Materials, Philadelphia, PA.
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Subject:
Millstone Nuclear Power Station, Unit 2, Reactor Pressure Vessel I I " 13W!!ssfafLv
I (RPV) Haterials Surveillance Program Docket No. 50 336 889, December 9, 1977.
- 22. Letter from J. F. Opeka to Offic of Nuclear Reactor Regulation,
Subject:
Haddem Neck Plant, Hillstone Nuclear Power Station, Unit Nos. 2 and 3, 10CFR50.61 Compliance, Docket No. 50 336, January 23, 1986.
- 23. Letter from E. J. Hroczka to Office of Nuclear Reactor Regulation,
Subject:
Millstone Nuclear Power Station, Unit No. 2 Request for Additional Information,10CFR50.61 Compliance. Docket No. 50 336, December 19, 1986.
- 24. F. W. Stallman, gL11., PR-EDB: Power Reactor Embrittlement Data Base.
Version 1, Oak Ridge National Laboratory, Oak Ridge, Tennessee, July 1991.
- 25. Based on information furnished by Northeast Nuclear Energy Company.
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