ML20097G042
| ML20097G042 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/31/1992 |
| From: | Aadland J, Lowe A, Nana A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML15245A254 | List: |
| References | |
| BAW-2128, BAW-2128-R01, BAW-2128-R1, NUDOCS 9206160242 | |
| Download: ML20097G042 (58) | |
Text
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4 BAW-2128. Rev. I 1
May 1992 l
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ANALYSIS OF CAPSULE OCIll-D
- DUKE POWER COMPANY 4
OCONEE NUCLEAR STATION UNIT-3 j -- Reactor Vessel Material Surveillance Program --
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i lBSso2a2g2ososnoocaowo?8R' 951HB&WNUCLEAR P IDWSERVICE COMPANY
EAW:1118. Rev. 1 May 1992 ANALYSIS Of CAPSULE 0C111 D DUKE POWER COMPANY OCONEE NUCLEAR STATION UNIT-3
-- Reactor Vessel Material Surveillance Program --
by A. L. Lowe, Jr., PE J. D. Aadland A. D. Nana H. A. Rutherford W. R. Stagg B&W Document No. 77-2128 01 (See Section 11 for document signatures)
B&W Nuclear Service Company Engineering and Plant Services Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 B W il?v M i % r o
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SUMMARY
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This report describes the results of the examination of the third capsule (OCll!- [
D) of the Duke Power Company's Oconee Nuclear Station Unit 3 reactor vessel f surveillance program. The objective of the program is to monitor the effects of i neutron irradiation on the- tensile and fracture toughness properties of the reactor vessel materials by the testing and evaluation of tension and Charpy impact specimen. The program was designed in accordance with the requirements ;
of Appendix H to 10CFR50 and ASTM Specification E185 73. !
The capsule received an average fast fluence of 1.45 x 10 n/cm' (E > 1.0 Mev) and the predicted fast fluence for the reactor vessel T/4 location at the end of the eleventh cycle is 2.18 x 10 n/cm' (E > 1 MeV). Based on the calculated !
fast flux at the vessel wall, an 80% capacity factor, and the planned fuel
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management, the projected peak fast fluence that the Oconee Unit 3 reactor- f pressure vessel inside surface will receive in 40 calendar years of operation is 9.49 x 10 n/cm' (E > 1 MeV). [
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 behavior of 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 _
I irradiation are conservative.
The recommended operating period was ' extended to 15 effective full power years -
i as a result of the third capsule evaluation. These new operating limitations are ,
in accordance with the requirements of Appendix G of 10CFR50.- ,
This revision corrects reactor vessel fluence values and eliminates the pressure-temperature operating limits for 21 and 24' EFPY.
SW51131M00 v j y . _ _ , -_ _ _ _ - _ - . _ - . . , _ _ . . . _ _ _ _ , _. . . _ , . , , . . . _ _ . , , , . . . , - - - .
RECORD OF REVISIONS Date Revision Number Descriotion May 1991 0 Original Issue May 1992 1 Summary - Revision _ statement added, fluence value revised Table 6 3 Correct 21, 24, 32 EFPY fluence Table 7-5 Corrected 32 EFPY data Table 7-6 Corrected 32 EFPY data Section 8 Deleted 21 and 24 EFPY pressure-temperature operating limit curves Section 9 Fluence at EOL revised-Section 11 - Revision signatures.added-Appendix D'- Table D-2 fluence values corrected
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l 3 CONTENTS i
Page j 1. INTRODUCTION ...........................11.
l 2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
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l 3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . . . . . . 3-1 '
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- 4. PRE-IRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . . .
41 .
4.1. Tensinn Tests ...... . . . . . . . . . . . . . . . . -. 41
! 4.2. Impact Tests . . . . . .. . . . . . . . . . . . . . . . . . . 4 1-l S. POST-!RRADIATION TESTS ................._.....51 5.1. Thermal Monitors ......................51-5.2. Tension Test Results .................... 51 j 5.3. Chaipy V-Notch impar.t Test Results . . . . . . . . . . . . . 5-2 l 6. NEUTRON FLUENCE . . . .-. .-. . . . . . . . . . . . . . . . . . . . 6-1 l
6 .1 '. Introduction . . . . . . . . . . . .-. . . . . . . . . . . . .- 6-1 6.2. Vessel Fluence . . . . . . . . . . . . . . ... .. ... . . 64 6.3. Capsul e Fl uence . . . . . . . . . . .. . . . . . . . . . . . . . 65 6.4. Fluence Uncertainties . . . . . . . ... . . .=. . . . . . . ._. 6 5 .
- 7. DISCUSSION OF CAPSULE RESULTS . . ... . . . . . .-. . . . . . . . 7-1 [
7,1. Pre Irradiation. Property Data-. . . . . . . . . .:. . . . ._... 7-1 7.2. Irradiated Property Data . , . . . . . . . . . . . . .... . . . 7-1 7;2.1. Tensile Properties .. . ... . . . . . .....'. ... . . . 7-1 +
7.2.2._ Impact Properties . . . . . . .:. , . . . . . . . . . .L7-2 ,
7.3. Reactor Vessel Fracture Toughness . . . . . . . . . . . .. '.
7-4 l 8. DETERMINA110N OF REACTOR: COOLANT' PRESSURE BOUNDARY PRESSURE -
- j. TEMPERATURE LIMITS ......-...-...._,,....-....-..-..-81
- 9. SUMARY:0F RESULTS ..-.......-._._.....-.._.-..-9 ,
-10. SURVEILLANCE CAPSULE: REMOVAL SCHEDULE ~. . .:. . 10 1-
- 7. . . . . . .-. .
11 ;CERTIFICATIONi... . . . ... . . . . .,. . .'._. . . .c. .7. . . . -
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- t Contents (Cont'd)
APPENDIXLi Page A. Reactor Vessel Surveillance Program Backgrour.d Data a n d I n f o rm a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . A 1 i B. Pre Irradiation Tensile Data ....................B1 C. Pre lrradiation Char)y !mpact Data .................C1 D. Fluence Analysis Met 1odology ....................D1 i E. Capsule Dosimetry Data . . .....................E1 .
F. References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 List of Tables Table 3 1. Specimens in Surveillance Capsule 0C111 D . . . . . . . . . . . . .
3-3 3-2. Chemical Composition and Heat Treatment of .
Surveillance Materials . ......................-3-4 '
5 1. Tensile Properties of Cppsule OCill D Base Metal and Weld Metal Irradiated to 1.45 x 10 n/cm' (E > 1 MeV) . . . . . . . . . . . 53 '
5 2. Charpy impact Data From Capsule 0C111-0 Base Metal, ANK-191, Transverse Orientation, Irradiated to 145 x 10" n/cm' (E > 1 MeV; . . . . . . . . . . . . . . . . . . 5-3 5-3. Charpy Impact Data from Capsule 0Cill-D, Base Metal, AWS-192, Transverse Orientation, Irradiated to 1.45 x 10 n/cm' (E > 1 MeV) . . . . . . . . . . . . . . . . . . 5-4 5 4. Charpy impact Data From Capsule 00111-0 HAZ Metal, ANK 191, Longitudinal Orientation, Irradiated to 1.45 x 10 n/cm' (E > 1 MeV) . . . . . . . . . . . . . . . . . .
5 5 5. Charpy impact-Data from Capsule 0C111-D HAZ Metal, AWS 192, Longitudinal Orientation, Irradiated to 5 6.
1.45 x 10 n/cm' (E > 1 HeV)- ..................
Charpy Impact Data From Capsule 00111-D Weld Metal WF-209-1 55 Irradiated to 1.45 x 10 n/cm' (E > 1- MeV) . . . . . .:. . . . . . 5-5 5-7. Charpy impact Data From Capsule 00111 D Correlation Monitor Material,'* Heat No. A 1195 1 Irradiated to 1.45 x 10- n/cm' (E > 1 MeV) . . . . . .-.-... . . , . . . . . . . .
.-5-6 6-1. Surveillance Capsule Dosimeters.. . . ... . . . . . . . . . . . . .
6 6~
6-2. Oconee Unit-3 Reactor Vessel Fast Flux ........-......6-7.
6-3. Calculated Oconee Unit-3 Reactor Vessel Fluence . .. . . . . . . . 6 8 6-4. Surveillance Capsule 00111-0 Fluence, Flux,-and DPA . . . . . . . 6-9 6-5. Estimated Fluence Uncertainty . . . . . . . .- . .-.. . . . .'. . . . 6-9 7-1. Comparison of Capsule tilli-D Tensile Test Results . . . . . . .
7-6 7-2.. Summary of Oconee Unit 3 Reactor Vessel Surveillance Capsules -'
Tensile Test Results . . . . . . . . . . . . . . . . . . ...-. . . 7 7-3. Observed Vs.. Predicted Changes for Capsule OClll-0' Irradiated -
Charpy impact Properties - 1.45 x 10' n/cm' '(E -> 1 MeV) . - . .- . s. 8-
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I i lables (Cont'dl i
j Table Page 1
7-4. Summary of Oconee Unit 3 Reactor Vessel Surveillance Capsules l Charpy impact Test Results . . . . . . . . . . . . . . . . . . . 7-9 j 7-5. Evaluation of Reactor Vessel End of-Life Fracture. Toughness and Pressurize. Thermal Shock Criterion - Duke Power Company,
{ Oconee Unit-3 02 EFPY) . . . . . . . . . . . . . . . . . . . . . 7-10
. 7 6. Evaluation of Duke Power Company Oconee Unit-3 Reactor Vessel l End of Life, Upper Shelf Energy (32 EFPY) . . . . . . . . . . . . 711 j 8 1. Data for Preparation of Pressure-Temperature Limit Curves for -
j Oconee Unit-3 -- Applicable Through 15' EFPY . . . . . . . . . . . 8 4 i A-1. Surveillance Program Material Selection Data for Oconee 3 . . . . . A 3 i l
- j. A 2. Materials and Specimens in Upper Surveillance Capsules I OCill- A, 0C111-C, and OCIll-E . . . . . . . . . . - . . . . . . . . . A-4
. A 3. Materials and Specimens in Lower Surveillance Capsules
! OClli-B, 00111-D, and OCIll-F . . . . . . . . . . ._. . . . . .-. . A-4 i B 1. Pre-trradiation Tensile Properties of Shell Plate Material, j Heat ANK-191 . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-2 j B 2. Pre-!rradiation Tensile Properties of Weld Metal -
i Longitudinal, WF-209 1 . . . . . . . . . . . . . ... . . . . . . . B-2
- C-1. Pre-Irradiation Charpy impact Data for Shell forging i Material - Transverse Orientation, Heat ANK-191_. . ._. ._. . . . . C-2 i- C-2. Pre-Irradiation Charpy impact Data for Shell Forging j Haterial - Transverse Orientation, Heat AWS-192 . . . . . -. . . . . C-3 j
- C 3. Pre-Irradiation Charpy impact Data-for Shell Forging l Material - HAZ, Transverse Orientation. Heat ANK 191 ..._....C4 l
- C 4. Pre Irradiation Charpy impact Data for Shell Forging j Haterial - HAZ, Transverse Orientation, Heat AWS-192 .......C5 1
C-5 Pre-Irradiation Charpy impact Data for Weld Metal, WF-209-1 . . . . C 6
{ D 1. Flux Normalization Factor . . . . ... . . . . . . . . . . . . .-. . D-7 j D-2. Oconee Unit 3 Reactor Vessel Fluence by Cycle . . . . . . . . . . . D 8 j E-1. Detector Composition and Shielding . . . . . . . . . . . . . . . . E-2 4 E-2. Measured Specific Activities (Unadjusted) for Dosimeters in Capsule 00111 0 . . . . . . . . ... . . . . . . . . . . . . . . . . E-2
- E-3. Dosimeter Activation. Cross Sections, b/ atom . . . . . . . . . . . . E-3 4-l List of Fioures
,i Figure
- 3-1. Reactor Vessel Cross Section Showing Location of Capsule-OCill D Duke Power Company Oconee Unit-3 . . . . . . ._. ... . . _ 3-5 Reactor Vessel Cross Section Showing Location of Oconee
'-2.
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Unit.3 Capsule 0Clli-D in Crystal River Unit.3 Reactor - -. . . , . .- 3-6 3-3. Loading Diagram for Test Specimens in Capsule OClli-D . . . . . 3-7 5-1. Charpy-Impact-Data for Irradiated Base Material,.
Transverse Orientation, Heat'No.- ANK-191 . . . . . .-.=. . . . , . 5 7 4
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Fioures (Coatish Figure Page 5 2. Charpy Impact Data for Irradiated Base Material, Transverse Orientation, Heat No. AW5-192 . . . . . . . . . . . . . . , . . . 5-8 5 3. Charpy impact Data for Irradiated Base Material Heat-Affected Zone. Heat No. ANK-191 .. . . . . . . . . . . . . . . . . . . . 5-9 5 4. Charpy impact Data for Irradiated Base Material. Heat Affected Zone. Heat No. AWS-192 . . . . . . . . . . . . . . . .-. . . . . 5 10 5 5. Charpy impact Data for Irradiated Weld Metal, WF-209 .
. . . . 5-11 5 6. Charpy impact Data for Irradiated Correlation Material, Correlation Material, HSST PL-02, Heat No. A-11951. . . . . . . 5-12 6 1. General Fluence Determination Methodology . . . . . . . . . . . . 6-2 6-2. Fast Flux, Fluence and DPA Distribution Throu Vessel Wall . . . . . . . . . . . . . . . . .gh Reactor . . . . . . . . . . 6 10 6 3. Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 8-1. Predicted _ Fast Neutron Fluence at Various Locations Through Reactor Vessel Wall for 32 EFPY - Oconee Unit 3 . . . . . . . .- . - . 8-5 8 2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Heatup, Applicable for First 15 EFPY - Oconee Unit-3 .-. . . . . . . . . . . . . . . . . . B 6 8 3. Reactor Vessel Pressure Temperature limit-Curves for Normal Operation - Cooldown, Applicable for 8-4.
First 15 EFPY - Oconee Unit-3-. . . . . . . . . . . . . . . . . . .
8-7 Reactor Vessel Pressure-Temperature-Limit Curves for Inservice Leak and Hydrostatic Tests -Ap for First 15 EFPY - Oconee Unit-3 . . . . plicable
.....-...... 88 A-1. Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel . . . . . . . . . . . . . . . . . . . . A-5 C 1. Charpy Impact Data from Unirradiated Shell Forging Material
( ANK 191) Transverse Orientation . . . . . . . . . , . . . . . . . C-7 C-2. .Charpy Impact Data from Unirradiated Shell Forging Material
( AWS-192), Transverse Orientation . .- , . . . . . . . . . . . . . . C-8 C 3. Charpy impact Data from Unirradiated Shell Forging Material (ANK-191), Heat-Affected Zone, Longitudinal Orientation . . . . . . C-9 .
C-4. Charpy impact Data from Unirradiated.Shell Forging Material .
(AWS-192), Heat-Affected Zone, Longitudinal Orientation . . . . . . C-10 C-5. Charpy Impact Data from Unirradiated Weld Metal WF 209 . . . . . C-Il D-1. Rationale for the Calculation of Dosimeter Activities-and Neutron Flux in the Capsule . . ... . . . . . . . . . . . . . . D-9 D 2. Rationale for the Calculation of Neutron Flux-D 3.
in the Reactor Ves sel . . . .- . . . . . . . . . . . . . . . . . . . . D-10 Plan View Through Reactor Core Midplane (Reference R-e calculation Model) . , . . . . . . . -. . . . . . . . D- 1 1
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- 1. INTRODUCTION This report describes the results of the examination of the third capsule (0C111-D) of the Duke Power Company, Oconee Nuclear Station Unit 3 (Oconee Unit-3) reactor vessel material surveillance program (RVSP). The capsule was removed and evaluated after being irradiated in the Crystal River Unit 3 reactor as pr t of the Integrated keactor Vessel Materials Surveillance Program (BAW 1543A). This irradiation in Crystal River Unit 3 plus the previous irradiation in Oconee Unit-3 is the equivalent of 22.22 years of exposure in the Oconee Unit 3 reactor
- vessel. I8 8 The capsule experienced a fluence of 1.45 x,10 n/cm (E > 1 HeV).
-which is the equivalent to approximately 57-effective full power years' (EFPY) operation of the Oconee Unit-3 reactor vessel. The first capsule (OClli A) from this program was . removed and examined after the first' year of nperation; the results are reported in BAW-1438.8 The second capsule (0C111-B) was removed and examined after irradiation in Florida Power Corporation Crystal River Unit-3 as part of the Integrated Reactor Vessel Materials Surveillance Program; the results are reported in BAW 1697.8 The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillance program for Oconee Unit-3 was designed and furnished by Babcock &~Wilcox (B&W) as described in BAW 10006Ad and conducted in accordance with BAW-1543A.' The program was planned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel.
The surveillance program for Oconee Unit-3 ' was designed in accordance with E185 66and thus is ~not in compliance with:10CFR50, Appendixes G' and H' since the requirements did not exist at'the time the program was design. .Because of the difference, additional tests and evaluations were required to ensure meeting 1-1 SWARNIHbv i
i S the requirements of 10 CFR 50, Appendixes G and H. The recommendations for the future operation of Oconee Unit-3 included in this report do comply with these requirements.
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L j l j 2. BACKGROUND i
I 1 The ability of the reactor pressure vessel to res.st fracture is the primary i j factor in ensuring the safety of the primary system in light water cooled l l reactors. The beltline region of the reactor vessel is the most critical region !
I of the vessel because it is exposed to neutron irradiation. The general effects
- of fast neutron irradiation on the mechanical properties of such low. alloy- l j ferritic steels as SA508 Class 2, modified by ASME Code case 1332 4, used in the [
! fabrication of the Oconee Unit-3 reactor vessel, are well characterized and docu-
- mented in the literature. The low alloy ferritic steels used in the beltline l region of reactor vessels exhibit an increase in ultimate and yield strength
properties with a corresponding decrease in ductility after irradiation. 'The ,
a most significant mechanical property change in reactor pressure vessel steels I i is the increase in temperature. for the transition from brittle to ductile e
fracture accorapanied by a reduction in the Charpy upper shelf energy value.
I Appendix G to 10CFR50, " Fracture Toughness kequirements,'" specifies minimu'm '
$ fracture toughness requirements for the ferritic materials of the pressure-
- retaining components of the reactor coolant pressure boundary (RCPB) of l
, water-cooled power reactors,.and provides specific guidelines for determining the pressure temperature limitations on operation of the RCPB -The toughness and ;
operational requirements are specified to provide adequate safety margins during j any condition of normal operation, including anticipated operational occurrences-
- 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 to all' boiling and pressurized water-cooled-nuclear power reactors, including those ,
under construction or in' operation on the-effective date. '
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Appendix H to 10CFR$0, " Reactor Vessel Materials Surveillance Program Requirements,'" defines the material surveillahce 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. Fracture 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 pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code Section 111
" Nuclear Power Plant Components." This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RINDT, which is defined as the greater of the drop weight nil ductility transition temperature (per ASlH (
E-208) or the temperature that 'is 60f below that at which the material exhibit 50 ft-lbs and 35 mils lateral expansion. The RT NDT of a given material is used to index that material to a reference stress intensity factor curve (k gp curve),
which appears in Appendix G of ASME Section !!!. 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 ip curve, allowable stress intensity factors can be obtained for this material K
as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
The RT NDT and, in turn, the operating limits of a ruclear power plant, can be adjusted to account for the effects of radiation on tne properties of the reactor 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 to adjust it NOT for radiation embrittlement. This adjusted RT is used to index the material NDT to the K IR curve which, in turn, is used to set operating limits for the nuclear 2-2 13 W ita M i E n
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power plant. These new limits take into account the effects of irradiation on j the reactor vessel materials, j Appendix G,10CFR50, also requires a minimum Charpy V notch upper shelf energy f
- of 75 ft lbs for all beltline region materials unless it is demonstrated that l J 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 ft lb requirement provided the irradiation i
deterioration does not cause the upper-shelf energy to drop below 50 f t lbs. The -
regulations specify that if the upper-shelf energy drops below 50 ft lbs it must .
be demonstrated in a manner approved by the Office of Nuclear Regulation that the :
lower values will provide adequate margins of safety.
Wher a reactor vessel fails to meet the 50 ft-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, j The program must address the following: l 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. -l i
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- 3. SURVEILLANCE PROGRAM DESCRIPTION The surveiilanco program for Oconee Unit 3 comprises six surveillance capsules designei to monitor 1.he effects of neutron and thermal environment on the materials of Use reactor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned inside the reactor voss41 between the thermal shleid and the vessel wall at the locatiens shGe in figurc 34. . Tie -six capsules, originally designed to be placed two N cach holder tube, are positioned near the peak exial and aziinuthal neutron flux. WW-10006A int l)udes a-full description of the capsule locations and design. AftJr'thR Capuits were removed from Oconee Unit 3 in 1976 and -
included in the integrated RVSP, they were scheduled and feradiated in. the Crystal River Unit 3 reactor as described in BAW-1543A. During this period of irradiation, tapsule OCIII D was irradiated in the bottom location in holder tube YX as shown in Figure 3-2.
Capsule OCII) D was removed during the seventh refueling shutdown of Crystal River Unit 3. This capsule contained Charpy V notch impact test-. specimens fabricated from one base metal (SA508, Class 2), one heat affected o one, a weld metal and correlation material Tension test specimens were fabricated from the base metal and the weld metal only. .The specimens contained in the capsule are described in Table 31, and the location of the individual specimens within the capsule are described in Figure 3-3. The chemical composition and heat treatment of ths surveillance material in capsule 00111 0 are described in Table 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 surveillance material such that they were oriented with their longitudinal axes-either parallel or. perpendicular to the principal working direction.
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Capsule 0C111-D contained dosimeter wires, described as follows:
Dosimeter Wire ShielditLq.
U Al alloy Cd Ag alloy Np Al alloy Cd Ag alloy Nickel Cd-Ag alloy 0.66% Co Al alloy Cd 0.66% Co-Al alloy None fe None Thermal monitors of low melting alloys .nd metals were included in the capsule.
The alloys and metals and their melting points were as follows:
Allov Meltina Point,_[
90% Pb, 5% Ag, 5% Sn 558 97.5% Pb. 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Cadmium, 99.99+% 610 lead. 99.994% 621 l
1 3-2 S W !! E W a h r
Table 3-1. Soetimens in Surveillange Capsule 00111-0 No. of Specimens Material Description Ign}tig [hnp_y Weld metal. WF-209, longitudinal 2 12 Weld, HAZ Heat A ANK-191, longitudinal 0 12 Heat B - AWS-192, longitudinal 0 6 Baseline material Heat A - ANK-191. transverse 2 12 Heat 8 - AWS-192, transverse 0 6 Correlation, HSST plate 02 9 _1 Total per capsule 4 54 3-3 BWHrvEYii%w
i Table 3 2. Chemical Composition and Heat Treatment of Surveillance Materials Chemical _ Analysis Heat Ileat Weld Metal Correlation Material Element ANK- 191 AW5- 192 WF-209-1B" HSST-02 (A-1195-1)'*'
C 0.24 0.21 0.08 0.23 -l Mn 0.72 0.58 1.63 1.39 i P 0.014 0.011 0.017 0.013 !
S 0.012 0.015 0.012 0.013 l Si 0.21 0.24 0.61 0.21 !
Ni 0.76 0.73 0.58 0.64 !
Cr 0.34 0.30 0.10 --
Ho 0.62 0.60 0.39 0.50 ,
Cu 0.02 0.01 0.30 0.17 i i
Heat 1reatment Heat No. Temo.. F Time h Coolina ANK- 191 1620 1660 4.0 Water quench. !
1570 1610 4.0 - Water quench 1230-1270 10.0 Water' quench - ,
1100 1150 40.0 Furnace-cooled AWS-192 1620 1660 -4.0 Water quench !
1570 1610 4' 0 Water quench !
1220 1250 10.0 Water quench 3 1100 1150 40.0 Furnace-cooled WF-209-1B" 1100-1150 30.0 furnace-cooled .
A-119S l**' 1600175 4.0 Water quench 1225125 4.0 Furnace-cooled 1125 25 40.0 -Furnace cooled ,
Pe r BAW-1820.
"Per 'BAW-1500 and BAW-1820.
f
'*'ORNL 4463.*
'*Per plate section identification card.
Normal ized at 1675F i 75F.
l 3-4 '
L SWANt31Mby 1
L ,..m.. - , _ u. _ _... ,_ . _ _ _ _ _ _ ._ ._. _ __ _ -._. _ _
Figure 3-1. Reactor Vessel Cross Section Showing tocation of Capsule 00111-D Dub, Poyer Cp2Pln.Y_0cluee Unit-3 Surveillance Capsule lloider Tubes - Capsules DC117-C.
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l
, .._._-___.__.__.....__.._J
Figure 3-2. Reactor Vessel Cross Section Showing Location of Oconee Unit 3 Caosule 0C111 D in Crystal River Unit 3 Scactor--
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- 4. PRE-IRRADIATION TESTS Unirradiated material was evaluated for two purposes: (1) to establish a baseline of dat5 to which irradiated properties data could be referenced, and (2) l to detsrmine those materials properties to the exter.'. practical from available caterial, as required for compliance with 10CfR50, Appendixes G and H.
4.1. Tension Tests Tension test specimens were fabricated from the reactor vessel shell course forging and weld matal. The specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter. They were tested on a 55,0001b load capacity universal test machine at a crosshead speed of 0.050 inch per minute. A 4 pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accordance with the applicable requirements of ACTH A370-77.10 for each material type and/or condi-tion, six specimens in groups of three were tested at both room temperature and 580F, Yhe tension compression load cell used had a certified accuracy of better than 40.5% of full scah (25,000 lb). All test data for the pre irradiation tensile specimens are given in Appendix B.
L2. Imoact Tests Charpy V notch impact tests were conducted in accordance with the requirements of ASTM Standard Methods A370 77 10 and E23-82 II on an impact tester certified to meet Watertown standards. Test specimens were of the Charpy V notch type, which were nominally 0.394 inch square and 2.165 inches long.
Prior to testing, specimens were temperature-controlled in liquid immersion baths, capable of covering the temperature range from 50 to 4550F. Specimens-were removed from the baths and positioned in the test frame anvil with tongs specifically designed for the purpose. The pendulum was released manually, 4-1 S W l/ M a h r
I allowing the specimens to be broken within 5 seconds from their removal from the temperature baths.
Impact test data for the unirradiated baselin a reference inaterials are presented in Appendix C. Tables C-1 through C 5 contain the basis data that are plotted in Figures C-1 through C-5.
I l
l i
1 I
4-2 BWifMVSt%v
- 5. POST-IRRADIATION TESTS 5.1. Thermal Monitors Capsule OCIII-D contained two temperature monitor holder tubes, each containing five fusible alloy wires with melting points ranging from 558 to 621F. Four of the five wires in F.h set had the appearance of being melted. All the thermal monitors at 558, 580, and 588F had melted while those at '.he 610F location showed no signs of melting or slumping; the monitor at the 621F: location appeared to be melted in both holder tubes. Heretofore it was assumed that the 610F and 621F monitors were placed in the wrong locations i_n the holder tubes, and based on these observations, it was concluded that the capsule had been exposed to a peR temperature in the range of 610 to 621F during the reactor operating-period.
In the case of OCIII-D the original loading diagram was consulted. This drawing lists the five materials-used in the monitors, and showed the position in which-each wire was loaded. Both show the lead wire (621F melting poir.t) to be in the fourth position, with the cadmium wi_re (610F melting point) in the fifth-position. This means that the 610F monitor did not melt while the 621F- monitor appeared to melt. It is believed that the; lead wire softened and slumped and thereby presented the appearance of melting due tc long- exposure _- to elevated -
temperatures, which were not' sufficient to melt the cadmium wire. . Therefore, it is probable that the capsule was_ exposed to-temperatures in excess of 588F:but not as high as 610F, and that this was. sufficient to cause the-lead wires to O slump and appear to have melte3.
L2. Tension Test Results The results of_ the postirraaiation tension tests are presented in Table 5-1.
-Tests were performed on specimens at both room temperature and at the temperature of 500f using similar test _ procedures and techniques used to test the unirradiat -
ed specimens (Section 4.1). In general, the ultimate strength and yield.
]
5-1
~
SWADM!raiGL=v i
-- - a
strength of the material increased with a corresponding slight decrease Jin-ductility as compared to the unirradiated values; both-effects were the result-of neutron radiation damage. The type of behavior observed and the degr<e to which the material properties changed is within the range of changes to be expected for the radiation environment to-which the specimens were-exposed.
The results of the pre-irradiation tension tests are presented in Appendix B, 5.3. Charov V-Notch 1moact Test Results-The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 5-2 through 5-7 and Figures 51 through 5-6. The test procedures and techniques were similar to those used to test the unirradiated specimens (Section 4.2). The data show that the materials exhibited a sensitivity to -irradiation within the v'alues to be expected from their chemical composition and the fluence to which they were exposed, i
The results of the pre-irradiation Charpy V-notch -impact tests are given in Appendix C.
t 5-2 SWB!M!mhv D 1,
._ ____-___._i. ____i________ _.___._____o
Table 5-1. TensilePropertiesofCapsuleOC!lhDBaseMetaland' Weld Metal Irradiated to 1.45 x 10 n/cm' (E > 1 MeV)
Red'n.
Specimen Test Temp, Strenath. osi Elonaaticp. % in 1rea,
. No. F Yield Ultimate Uniform Total %
Ease Metal. Transverse. Heat ANK-191 JJ 602 70 66,900 _91,500 9.9 26.6 65.1 JJ 618 550 60,300 87,200 9.9 23.3 62,0
)[qld Metal . WF-209-1 JJ 017 70. 98,000 111,300-- 10.7 24.2 56.7 JJ 016 550 86,600 102,900- 9.0 17.7 -49.4 Table 5-2. Charpy impact Data From Capsule 0C111-D,' Base g al, ANK 191, Transverse Orientation, Irradiated to 1.45 x-10 n/cm' (E > 1 MeV)
Absorbed Lateral Shear-Specimen Test Temp., Energy, Expagsion. Fracture,-
No. F ft-lb 10 in. -%
JJ 634 0 9,0 02 10 JJ 693 25 23.0 23 10' JJ 646 30 39.0 30 10 JJ 683 30 63.0 50 20 JJ 659 40 58.0 -51 20 JJ 669 70 71.5 57 30 JJ 624 125 107.0 76 70 JJ 677 200 123.0*~ 83 100 JJ 627 250 124.0* 99 100- I JJ 615 350 128.5* 89- 100 JJ 628. 450 123.0 86 100-JJ 662 550 123.5 91 100-
- Values used to determine upper-shelf energy-value per ASTM E185.
5-3 SWB&TIMLv b .. . . . . . . __ ____ _ _ __ _ _ __i
Table 5-3. Charpy Impact Data From Capsule OCIII-D, Base Mgpal, AWS 192 Transverse Orientation, Irradiated to-1.45 x 10 n/cm' (E > 1 MeV) -
-Absorbed' Lateral- Shear Specimen Test Temp., _ Energy, .Expagsson, Fracture,-
- No. F ft-lb 10 in. %
KK 668 0 5.0 03 0' KK 615 40' 37.0 29 10-KK 649 70 50.5 43 20 KK 660 125 84.0 64 60 3 KK 617 200 90.0 67 90 KK 638 300 93.0* 77- 100
- Value used to determine upper-shelf energy value per ASTM E185.
Table 5-4. ANK 191 Charpy Longitudinal Impact Data -From Orientation, Capsule Irradiated to l' 45 OClll-D'HAI x 10 Metg 'n/cm' (E >:1 MeV)
Absorbed Lateral Shear Specimen Test Temp., Energy, Expagston, Fracture, No. F ft-lb 10 in. % ,
JJ 366 -50 23.5 13: l'0 JJ 349 0 25.0 16 20 JJ 329 15 31.5 20 30 JJ 326 70 136.5- 90 90-JJ 361 70 26.5 30- 20-JJ 378 70 138.0 97 100 JJ 367 100 63.0 40- 50 JJ 360 125- 62.0 46 .85' JJ 307 200 75.0* 51 100 JJ 368 250-- 64.0* '51 100 JJ 356- 350 66.0* 48 100 JJ 352 550 J73.5 59 100
- Values .used to determine upper-shelf energy value-per ASTM E185.
5-4
- S W 5 fwiM H h r-
i Table 5 5. Charpy impact Data From Capsule 0C111-0 HAZ Metal,' AWS-192 Longitudinal Orientation, Irradiated to 2.45 x 10 n/cm' (E > 1 MeV)
Absorbed Lateral ' Shear.
Specimen Test Temp., Energy, _Expagsion, Fracture, No. in.
F
_ lb ft 10 %
KK 328 50 36.0 22- 10 KK 314 0 79.5 50 40 KK 311 70 115.0 78 100 KK 331 125 64.0 51 60 KK 329 200 75.0* 50 100 KK 322 300 108.0* 80 100
- Values used to determine upper shelf energy value per ASTM E185.
Table 5-6. Charpy impact Data Fromggapsulp OCIII-D Weld _Hetal WF-209-1.
Irradiated to 1.45 x 10 n/cm (E > l MeV)
Absorbed Lateral Shear Specimen Test Temp., Energy, LFracture, Expagsion, No. F ft-lb 10 in. %
JJ 037 70 18.0- 18 20-JJ 011 125 16.5 16 10-JJ 045 160 30.0 27 40 JJ 058 175 27.0 31 70 JJ 051 185 22.0 19 30 JJ 052 200 41.0 38 90=
JJ 044 250 49.5* 100 JJ 025 300 39.0* 36 100 JJ 021 350 :39.5* 41 100 JJ 035 400 40.0* 40 100 JJ 012 450 47.0 47- 100 JJ 005 -550 39.0 45~- 100
- Values used to determine upper-shelf energy- value per ASTM E185.
i 5-5 SW#n51iid&w
,..g,
,.i
Table 5-7.- Charpy- Impact Data-From Capsule:0CIII-D' Correlation Monitor Mgerial -g Heat No.< A-11951,- Irradiated to-l'.45 x 10. n/cm - (E -> 1 MeV)-
Absorbed' l.ateral Shear-Specimen- Test-Temp., Energy, Expagston, Fracture,,
No. F ft-lb 10 in. -%
JJ'907 70 '14.0 11- 10-JJ 904 125 27.0 22 20 JJ 906 175 31.0 29- 30 JJ 912 250 79.0 66 -70 JJ 913 300 97.5 77 90 .
JJ 909 375 91.0* 82 100
!- d
- Value used to determine upper-shelf- energy- value per ASTM- E185.
L l
u _
6 SWANMt1 MAR-r
t 4
Figure 5-1. Charpy' Impact Data for Irradiated: Base Material, Transverse Orientation. Heat No. ANK 191 100 , , ,
ne -
, 75 -
, . v 3
~
yg -
- 25 -
a i i e i i e e 4 e
0.10 , , , , ,
l i
, g - - , _e 1
, e
[ g 0.08 -
it, k0,06 -
T =
1 -
- { 0.3
- 5 wa
) -
- S 0,02 -
i 5 a
0 22C' i i , ,
l'; i i
- DATA SumARY -
t
- k 200 "i No$*
9 Tcy (35 MLE) +"r l -
180 -icy (50 rr La) .wr Tey (30 rt-La) 99r l -
e IEC cv .USE (avr.) 125 ft-lbf i -j RT
- uor N "-
, 140
!- 8 .
- 4 l -5 120 - -
! 5!
, R 1; 100 -
1 E O. -
80 x , _
a 40 -
MATERI AL.1A508,CL.2 (IL)
- 1. H 1019n/cm2
- ~ 20 -
Fwtuct -
HEAT No. A Nx.191 4 ' ' ' ' ' '
0
-100 0 100 , 200 300 400 500 600 Test Temperature, F 5-7 B WfistrMIL-r t y -
a a3 , 4 A 4 a h+ai a M .A ie.1r.m.. A .eA- S 5-+s_s.
.s 1 - --
Figure 5 2. Charpy impact Data-for Irradiated Base Material, Transverse Orientation. Heat No. AWS-192 1 30 . . i . i
- " , 75 - -
o 5
- y 50 - -
a I25 - -
f ~f f 3 I 0 10 i e i i i a g 0.08 ,
5 e 50.06 - -
j0.04 - -
E a
-5 0.02 - --
2 0
220 i
.i i i i i DATA SU.TARY -
200 -T g, N.A. -
Tcy (35 met) +5?r 180 - T , (L rt-a) -61r -
ICV (30 ri-ts) +%F -
g 160 Cy -USE (Avo) 43 ft-lbf
$ RT,g7 N.A.
. 140 8
$120 a; 100 - -
.E ~ -
80 -
U r
- 60 -- -
40 _ . . .
MTERut $A508.CL.2 (IL) 20 -
FEutnct 1.45 1019n/cm2 _
HEAT No. AW5 102 0
!- -100 0 100- 200 300 400 - 500 600 Test Temperature, F 5-8 BW11:4!nMLr
e 4 4
I Figure 5-3. Charpy impact Data for Irradiated Base Material.
Heat-Affected Zone. Heat No. ANK-191 1
100 ; , ; ; ; , ,
e o H j , 75 - -
, e 6 .
en
@ 50
- e m
f 25 -
-e 5 1 t t t t t g
0.10 , , , , , , ,
~i e 1
g 0.08 - -
l 3 i
! !0.06 -
4 O _
_y0.0( - -
l
=
s y .
a 3, 0.02 - -
2 0
' ' ' ' ' 1 4
d.
3 4 , g g g
- DATA SGTARY -
200 - T,;, N.A. -
Tcy (35 mtt) .a t r 180 -T (50 FT-W *B 3r -
ICV (30 FT-LS) +4kI o 160 Cy -USE (avs) M f t-lbf -
h,A, S RT,37
, 1860 -
g 8 o
120 E
3 100 -
E O ;
80 -.
-g -
o _ e . _
40 -
- , MAttatAL SA508.Ct.2(HAl) 20 -
FLuencg 1.4510Hn/cm2. _
HEAT No. A N X-191 0
-100 0- 100 , 200 300 400 500 600 Test-Temperature, F-5-9 BWMAV!M%r l 4
Figure 5 4. Charpy Impact Data for--Irradiated Base Material.
Heat-Affected 70ne. Heat No. AWS-192 100 : . 7 , ,
.- i w
,. 75
] e g - _
_ sc f2 -
0 O.10 , , , , , ,
g _0.06 5
50.06 M
- e e -e
-.$0.04 .
' .E 5 0.02 - .
~,
s 0
20 i , , , , ,
- DATA SumARY -
200 -T,3, N.A. -
Tcy (35 not) -23f' 180 -Tg (50 rT-a) -28 t* -
-Tgy (30 FT-ts) -55F' g 160 Cy -USE (avo) 108 f t-lbf' RT,37 N.A.
. 140 - .
8 All data is suspect .
- due to scatter !
$120 e M 0 g 100 - .
s C
-- 80 -
o -
t w .
40 - _ .
MAftRIAL SA5083 CL2(H AI) 20 :FLuenct 1.45:1019n/cm2 .
' HEAT No. AWS-192 0
-100 0 100 - , 200- 300 400 500 600 !
Test.Temperaturei F 5-10 S W E N!Fi!M & .y.
Fioure 5-5. Charov impact Data- for irradiated Weld Metal. WF-2091 l
l'X i i , i .; : ,
75 -
3
! 50 I 25 a
o e
g I t i i 1 t
> 0.10 , , , , , ,
5, g 0.08 - -
2 e
$0.06 - -
O I g o a .
- 0.34
- e e -
t e 3
m 0.02 -
, . i
- e 2
0
' ' ' I i i 110 ,
r
- DATA S' J fMRY -
100 -T g7 4.A. -
I;y (35 mut) +170r M Tcy (50 FT-ts)het. Deter. -
g Igy (30 FT-L3) +18 5I 80 "t .USE y (AVG) b2 II"IDI "
g Ri g, R.A. l 2 70 -
E E
y so -
a y% -
e j e - e , .- , ,
30 , -
20 - ' -
- MAftniat Mn-Mo-N1/Linde 80 1 10 FLutact 1.45x1019 n/cm2 .
HEAT No. WF-200-1 0 ' ' ' ' . .
-100 0 100 200 300 400 500 600 Test Tencerature, F 5-11 SWALYe%Lv
1 Figure 5-6. Charpy Impact Data for Irradiated Correlation !
Material. HSST PL-02. Heat No. A-1195-1 -'
1 100 , , , . : , ,
B4
, 75 -
t 3
y.- -
f5 -
c 0,10 i i i i i i 0.03 - -' -
8
?
2 0.06 - -
_{0.04 -
5 5 0.02 -
5 x
0 2:0 . . , , i
- DATA
SUMMARY
200 - T,37 NA -
Tey (35 mtt) +181T 180
-Tcy (50 n-La) +1m -
Tcy (30 ri-La) + 1r,1 r 160 " Cy -USE ( Ava) 94 f t-lbf -
3 f RT ug, - R.A. '
, 140 -
g120 -
5 a 100 ...
E *
- 80 -
W
-r
, 60 -
w - -
,
- MATEntAL SA533,9 B16 I) 20 -
FLugner 1.45:1019n/cm2 ..-
HEAT No.. H33T-02 i f f I e ,
-100 0 100 200 300 400 500: -600 Test- Temperriture, F 5-12 S W ff a !! m h v i
4 l
- 6. NEUTRON FLUENCE j 6.1. Introduction i
j The neutron fluence (time integral of flux) is a quantative way of expressing the j cumulative exposure of'a material to a pervading neutron flux over a specific j period of time. Fast neutron fluence, defined as the fluence of neutrons having l energies greater than 1 MeV, is the parameter that is presently used to correlate l radiation induced changes 'in material properties. Accordingly, the fast fluence j must be determined at two locations: (1) in the test specimens located in the surveillance capsule,- and (2) in the wall of the reactor vessel. The former is
[
- used in developing - the correlation; between fasti fluence- and changes. in - the
- material properties of specimens, and the latter is used to ascertain the point-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 mater _ial properties. -
l The accurate determination of neutron flux is best accomplished through- the j simultaneous consideration of neutron dosimeter measurements _ and analytically L derived flux spectra. Dosimeter measurements alone cannot be used to' predict the -
e fast fluence in the vessel wall or in the test. specimens because (1) they cannot-l measure the fluence at the points. of interest, and. (2) they' provide (nly 4
rudimentary information about the neutron energy spectrum. Conversely, reliance on calculations alone to predict fast fluence i ts not prudent ' because of the
- length and complexity of the analytical procedures involved. LIn 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 pre-sented in general terms in Appendix D and in BAW-1485P.12 F
4 e
i-6-1 OWMMy-
- , - + - . , , ,-
Fiaure 6-1, General Fluence Determination Methodoloav MEASlRENENTS OF NEUTRON ANALYTICAL DETERMI)MTION OF 00SIMETER ACTIVITIES DOSIMETER ACTIVITIES APO NEUTRON FLUX ll ADJUSTED ENERGY DEPE}0ENT NEUIRON FLUX REACTOR OPERATING NEUIRON HISTORY N o PRE-FLUENCE DICTED FUTlRE OPERATION
^
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 discrete ordinates transport code, 00TIV.13 The calculations employ an angular quadrature of 48 sectors (S8), a third order LeGendre polynomial scattering approximation (P3), the CASK 23E cross section set I4 with 22 neutron energy groups and a fixed distributed source corresponding to the time weighted average power distribution for the applicable irradiation period.
In addition to the flux in the test specimens, the 00TIV calculation determines the saturated specific activity of the various neutron dosimeters located .in the surveillance capsule using the ENDF/B5 dosimeter reaction cross sections.15 The 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) 6-2 SWHEv'2a%r
1 irradiation history. Additional corrections are made to account for the following effects:
- Photon induced fissions in V and Np dosimeters (without this correction the results underestimate the measured activity).
- Fissile impurities in U dosimeters (without this correction the results underestimate the measured activity).
- Short half-life of isotopes produced in iron and nickel dosimeters (303- !
day Mn-54 and 71-day C0-58, respectively). (Withoutthiscorrection,the results could be biased high or low 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 neutton dosimeter measurements adjusted for the effects noted above. The neutron dosimeters located in the surveillance capsules are listed in Table 6-1. Both activation type and fission type dosimeters were used.
The ratio of measured dosimeter activity 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 an adjustment factor to the D0i-calculated flux at all points of interest.
Neutron Fluence The determination of the neutron fluence from the time averaged flux requires only a simple multiplication by the time in EFPS (effective full-power seconds) over which the flux was averaged, i.e.
f ( AT) - E &% aT 9
where
~
fjj (AT) - Fluence' ~at (i,j) accumulated over time aT (n/cm'),
6-3 BWUunsLa
. . i g - Energy group index, d gjg - Time-average flux at (i,j) in energy group g, (n/cm'-sec),
ai - Irradiation time, EFPS. '
Neutron fluence was calculated in this analysis fo'r the following components over the indicated operating time:
Test Specimens: Capsule irradiation tims in EFPS -5 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 considered to be an appropriate damage exposure index since displacements of-atoms from their normal lattice sites _is a primary source' of neut'ron- radiation -
damage. DPA was calculated based on the ASTM Standard E693-79 (reapproved 1985).16 A DPA cross section for iron is given in the ASTM Standard in '641 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 the DPA calculations reported herein, the ASTM DPA cross _ sections -were first collapsed to the 22 neutron group structure of CASK-23E; .the DPA-was -then determined by summing the group flux times the DPA cross section over the 22 energy groups and multiplying by the time of the' irradiation.
6.2. Vessel Fluence The maximnm fluence (E > 1 MeV) exposure of the Oconee Unit '3. reactor vessel during Cycles-6-11 was determined to be 1.96.x 10 IO n/cm* based on a maximum neutron flux of 9.49 x 10' n/cm'-s (Tables 6-2 and 6-3) . The maximum fluence-occurs at the cladding / vessel interface at an azimuthal location of approximately-11 degrees' from a major-_ horizontal , axis 'of the- core.
l l
6-4 SWBM1lh -
1 4
e 4 Fluence data were extrapolated to 32 EFPY of operation based on two assumptions:
(1) the future fuel cycle operations do not differ significantly from their current designs, and (2) the latest calculated (or extrapolated) flux remains constant from that time through 32 EFPY, The extrapolation was carried out in two stages. (1) from EOC 11 to EOC 13, an0 (2) from EOC 13 to 32 EFPY In the first stage, cycle averaged fluxes are calculated based on the current designs for cycle 12 and 13, using DOT ajoint factors for assembly-averaged power distributions, in the second stage, the 32 EFPY fluence was calculated by assuming a constant flux over the period which was equal to the average flux for cycles 12 and 13.
Relative fluence and DPA (displacement per atom) as a function of radial location in the reactor vessel wall is shown in Figure 6-2. Reactor vessel neutron fluence lead factors, which are the ratio of the neutron flux at the clad inter-f ace to that in the vessel wall at the T/4, T/2 and 3T/4 locations, are 1.78, 3.53, and 7.25, respectively. DPA lead factors at the same locations are 1.59, 2.64, and 4.55, respectively. The relative fluence as a function of azimuthal angle is shown in Figure 6-3. A peak occurs in the fast flux (E > 1 MeV) at about 11 degrees with a corresponding value of 9.49 x 10' n/cm'-s.
6.3. Caosule Fluence The capsule wts irradiated for 2373.9 EFPD in the top holder tube position during cycles IB-7 of Crystal River Unit 3 located 11 degrees off the major horizontal exis at about 202 cm from the vertical axis of the core. The capsule was also irradiated for 477.9 EFPD in Oconee Unit 3, during cycle 1, located at the 11 degree position about 211 cm from the vertical axis of the core. The cumulative fast fluence at the center of the surveillance capsule was calculated to be 1.45 x 10 n/cm' of which 5.1% was accumulated during the Oconee cycle 1 irradiation, and 94.9% was accumulated during the Crystal River cycles IB-7 irradiation (Table 6-4), This fluence value represents an average value for the various locations in the capsule.
6.4. - Fluence Uncertainties Uncertainties were estimated for the fluence values reported herein. The results are shown in Table 6-5 and are based on comparisons to benchmark experiments, 6-5 B W s* ? s' ? a fe Lu w
when available; estimated and measured variations in input data; and on engineering judgement. The values in Table 6-5 represent best estimate values based on past experience with reactor vessel fluence coulysas.
Table 6-1. Surveillance Capsule Dosimeters Lower Energy Limit for Isotope Dosimeter Reactions (a) Reaction, HeV Half-Life .
547 ,(n,p)54Hn 2.5 312.5 days -
58Ni(n,p)58Co 2.3 70.85 days 238U (n,f)l370s 1.1 30.03 years 237Np(n f)I37Cs 0.5 30.03 years (8) Reaction activities measured for capsule flux evaluation.
6-6 B Ws* TEE?a % r
s Table-6-2. Oconee Unit'3 Reactor Vessel Fast Flux.
Fast Flux (E > 1 MeV),7 n/cm'-s1 Fl ux ' n/cm'- s (E > O'1 MeV)
Inside Surface . Inside Surface- -
Cycle (Max locr. tion) T/4- 3T/4 -(Maxlocationl- 1 Cycle 1 1.39E+10 7.73E+9 -1.85E+9 2.78E+10; (477.9 EFPD) .
Cycles 2-5 1.55E+10 8.61E+9 1.96E+9- 3.30E+10 ,
(1020.1 EFPD)
Cycles 6-11 9.49E+9 5.32E+9- 1,31E+9 1.99E+10 3 (2395.9 EFPD) g
.S Cycle 12** 8.92E+9 5.C'E+9* 1.23E+9* ----
(410 EFPD)
Cycle 13** 7.62E+9 4.28E+9* -1.05E+9*- ----
(410 EFPD) 15 EFPY 8.27E+9 4.65E+9* 1.14E+9* ----
21 EFPY 8.27E+9 :4.65E+9* 1.14E+9* ----
24 EFPY 8.27E+9 4.65E+9*- :1.14E+9* ------
32 EFPY 8.27E+9 4.65E+9^ 1.14E+9* ----
i
- Divide flux at inside surface by the appropriate. lead # factors or. page 6-8 to obtain these T/4 and 3T/4 fast flux values.:
o* Assumed cycle length of 410 EFPD-for _ flux extrapolation for Cycles'lc and 13, 3 .-
6 6 SW8MMLov .
1
s .
Table 6-3. Calculated Oconee Unit 3 Reactor Vessel Fiv3nce Fast Fluence, n/cm' (E > 1 MeV)
Cumulative Inside Surface .
Irradiation Time (Max location) T/4 T/2 3T/4 _
End of Cycle 1 0.58E+18 3.19E+17 7.73E+16 7.65E+16-(477.9 EFPD)
End of Cycle 5 1.94E+18 1.nSE+18 0.16E+17 2.50E+171 (1498 EFPD)
End of Cycle 11 3.91E+18 2.18E+18 1.17E+18 5.21E+17 ,
(3893.9 EFPD)
End of Cycle 12 4.23E+18 2.38E+18* 1.20E+18* 5.83E+17* I (4303.9 EFPD)
End of Cycle 13 4.50E+18 2.53E+18*- - 1.27E+18* I6.21E+17*
(4713.9-EFFD)--
15 EFPY 5.05E+18 2.84E+18* 1.43E+18*- 16.97E+17*
21 EFPY 6.62E+18 3.72E+18* - 1.88E+18* 9.'13 E+ 17
- 24 EFPY 7.40E+18 4.16E+18* - 2.10E+18* - 1.02E+18*
32 EFPY :9.49E+18 5.33E+18* 2.69E+18* '1.31E+18*
- Calculated using these 1.0 1.78 3.53 7.25 lead factors Conversion Factors-Fluence (E > 1 MeV) 1 45E-21**
. :1.63E-21** -1.94E-21** 2.31E-21** -
to DPA
- Multiply fast fluence values (E > 1 MeV) in' units of n/cm' by. these' factors:
to obtain the corresponding DPA values.
l 6-8 SWAIN!MHbv p
f' ar --'-
a w w- '- r 'y-w
- D-g +- = ht ---
}
- l s
j Table 6 4. Surveillance Caosule 0C111-0 Fluence. Flux. and DPJ j-1
- - flux'(E > 1 MeV), Fluence, j -- Irradiation-Time - n/cm'- s ; n/cm' DPA j OC3, Cycle 1 2.48E+10 7.39E+17 1.41E-3 l (477.9 EFPD) l CR3, Cycle 18_7, 6.65E+10 1.38E+19 1.90E-2 j (2373.9 EFPD)
[ Total --
1.45E+19 2.05E-2 i-a
! Table 6-5. Estimated Fluence Uncertainty
! Estimated . .
. , Calculated Fluence Uncertainty Basis of Estimate i In the capsule 15% . Activity measurements, cross
] section fission yields, satu-
- - ration factor,-deviation from
!- average fluence value
- In_the reactor vessel 21% Activity measurements,' cross I at maximum location for . sections, fission _ yields, fac -
l- cycles 1 through 11:of tors, axial: factor,' capsule j Oconee Unit-3 -location, . radial / azimuthal. ex-j- - trapolation,- normalization
- factor-In the reactor vessel-j 23% ' Factors-in vessel fluence above
- at the maximum location plus uncertainties for extra-j - for-end-of-life extra- polation to 32;EFPY i , polation.
3 j
n j
3.
b-4 i
. 6 r - aggst!Mgg, -
.r- v - + y -r---
Figure 6-2. Fast Flux, Fluence and DPA Distribution Throuah Reactor Vessel Vall
^
1.00 -
L.F. = 1.585
-_y g L.F. = 2.639 o : ,,'g
- s @ , L.F. = 4.545
'g 0.50 ". S
- , s S T/4 x O 3O 8 li 2
L.F. = 1.783 ls's A 3 : ',
5 $
~
% m
- O.20 - @) m '
L e
1 e v T/2 'q i
E L.F. = 3.534 's o u -
o 4 s C C'3 m .
'S $-
0.10 -
b, C<L -
o O -
3T/4 5 e -
L.F. = 7.248 5 O ~ Gr) -
N 5 0.05 -
5 o
5 .
O Z-223.01 cm. 228.37 cm 233.72 cm 0.02 -
Flux o-.-
DPA - - + --
0.01 ' 'I''I'I'!'!'
215 220 225 23C 235 240 245 Radius from Core Center (cm) 6-10 SW#saihr- i
Fiqure 6-3.
Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface ..
1.08 6 3 3 3 y 4 , ,
1 06 -
l.04 -
' t .02 -
l.00 -
a>
E 0.98 -
E c
en 7n 0.96 -
.'. C
~
' E 0.94 -
vi
' T, m
0.92 -
0.90 - -
0.88 - -
0.86 - -
0.84 - -
0.82 i ' ' ' i i ,
0* 5' 10* 15* '20* 25* 30* 35' 40* 45*
Degrees from Mapr Axis
- t. ._ ._._:.- .. . . _ ._ - . ._.. _ . _ . _ .
- 7. DISCUSSION OF CAPSULE RESULTS 7.1. Pre-Irradiation Property Data A review of the unirradiated properties of the reactor vessel core ' beltline region materials indicated no significant deviation from expected properties except in the case of the upper-shelf properties of.the weld metal. Based on the predicted end-of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of the weld metal, it was predicted that the 1 end-of-service Charpy upper shelf energy (USE) would be below 50 ft-lb.- - Weld -
metal representative of the controlling weld metal was selected for inclusion in the surveillance program in accordance with the criteria in effect at the time-the program was designed for Oconee Unit-3. . The applicable selection criterion -
tsas based on the unirradiated properties only.
7.2. Irradiated Property Data 7.2.1. Tensile Procerties Table 7-1 compares irradiated and 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 ultimate and yield strengths. and decreases in_ ductility properties. All changes observed in the base metal are I such as- to be considered within acceptable limits. The ' changes at both room temperature-and 580F in the properties of the weld metal are larger than those observed for the base metal, indicating a greater--sensitivity of the weld metal-to irradiation damage. In either case, the changes:in tensile properties are insignificant relative to the analysis of the reactor vessel materials at this i periodLin service life. ;
7-1 SW21Mll1Hhr
_ _ - - _ _ _ - _ __ a
A comparison of the tensile data from previously evaluated capsules (Capsules 0CIII-A and 0C111-B) with the corresponding data from the capsule reported in this report is shown in Table 7-2. The currently reported capsule experienced a fluence that is twelve 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 weld metal exhibited greater sensitivity to neutron radiation than the base metal.
7.2.2. Impact Propertiel 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 not in good agreement with the value predicted using either Regulatory Guide 1.99, Rev. 2 17 even without applying the margin which is required. It would be expected that.
these values would exhibit better agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken at the 30 ft-lb temperature.
The transition temperature measurements at 30 ft-lbs for the weld metal is not in good agreement with the results using Regulatory Guide 1.99, Revision 2 without the margin applied. With the addition of the margin the procedure is very conservative in predicting the irradiated values. The estimating curves of Regulatory Guide 1.99, Rev. 2, are conservative for predicting the 30 ft-lb transition temperature shifts since the procedure requires that a margin be added to the calculated value to provide a conservative value.
ihe data for the decrease in Charpy USE with irradiation showed good agreement with predicted values for the base metal and only fair agreement for the weld metal. However, a good comparison of the measured data with the predicted value is not expected in view of the lack of data for low , medium , or high-7-2 SW##enH3-
- _ - . - - . - - . - - - - . - .- . - - - - ~ . .- .-. . -
l 1
l copper-content materials _ at the fluence values that were used- to develop the estimating curves.
A comparison of the Charpy impact data from the previously evaluated capsu'es
-(Capsules 0Clll-A and 0C111-B) 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 _ eighteen times greater _than the first capsule.
The base metal exhibited shifts at the -30 ft-lb level for the latest capsule that were similar to those of the previous capsules. The corresponding data for the weld metal showed a further increase at the 30 ft-lb level. This may be due )
to a further decrease in the upper shelf energy with a corresponding ' increased shift at the 30 ft-lb level.
Both the base metal and the weld metal exhibited a decrease in the upper shelf values similar to the previous capsule. The weld metal in this capsule exhibited a slightly greater-decrease than the held metal-in the previous capsule. These data confirm that the upper. shelf drop for this weld metal did not reach satura-tion as observed in the results of capsules evaluated by others. This behavior of Charpy USE drop for this weld metal should not-be considered indicative of a similar behavior of upper shelf region fracture toughness properties. This behavior indicates that other reactions may be taking place within' the material besides simple neutron damage. Verification of this relationship must await the testing and evaluation of the data from compact fracture- toughness test specimens.
Results from other surveillance 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 time the estimating curves are established. These parameters may ' include inaccurate fluence. values, poor chemical composition values, and:: variations-in data interpretation. ' The change in the regulations requiring the shift measurement to be based on -the 30 -ft-lb- value has minimized the errors that result from using the 50 ft-lb data- base to predict the shift' behavior at 30
_ft-lbs.
7 SWEEYah u
An evaluation of the reactor vessel end-of-life upper-shelf rgy for each of ene the materials used in the fabrication was Table 7-6. s are made presented andin the resu the reactor vessel are Linde 80y, flux, low-upper-s ae copper and are expected to be highly sensitive to neutron radiation danrela age. Two methods were used to evaluate the radiation energy. 5duced decrea upper shelf The method of Regulatory Guide 1.99, Revision 2,same which is the procedure as used in Revision 1, and the method presented in BAW 1803" which was developed of weld metals.
specifically to address the need of an o
estimating meth d f or this class The methods of Regulatory Guide 1.99, Revision 2, indicate that t wo of the welds may be expected to decrease below 50 f t-lb level prior to EOL However, BAW-1803 shows that none of the. materials used in the fabrication will have an upper-shelf energy below 50 ft-lbs through or vessel 32 EFPY d on the T/4 wall location. esign life based below 50 ft-lbs for the controlling weld metal at the vessel nside wall.
1 i
7-5 (SW??5NEEYN5m
~
copper-content materials 3t the ficc..ce values that were used to develop the estimating curves. l i A comparison of the Charpy impact data from the preitously evaiuated capsules (Capsules Otlll-A and OC111-B) 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 eighteen times greater than the first capsule.
The base metal exhibited shifts at the 30 ft-lb level for the latest capsule that were similar to those of the previous capsules. The corresponding data for the weld metal showed a further increase at the 30 ft-lb level. This may be due j to a further decrease in the upper shelf energy with a corresponding increased shift at the 30 ft-lb level.
Both the base metal and the weld metal exhibited a decrease in the upper shelf values similar to the previous capsule. The weld metal in this capsule exhibited a slightly greater decrease than the weld metal in the previous capsule, These data confirm that the upper-shelf drop for this weld metal did not reach satura-l tion as observed in the results of capsules evaluated by others. This behavior of Charpy USE drop for this weld metal should not be considered indicative of a similar behavior of upper shelf region fracture toughness properties. This behavior indicates that other reactions may be taking place within the material besides simple neutron damage. Verification of this relationship must await the testing and evaluation of the data from compact fracture toughness test specimens.
Results from other surveillance capsules also indicate that RT estimating NDT 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 time the estimating curves are established. These parameters may include
, inaccurate fluence values, poor chemical composition values, and variations in data interpretation. The change in the regulations requiring the shift measurement to be based on the 30 ft-lb value has minimized the errors that result from using the 50 ft-lb data base to predict the shift behavior at 30 ft-lbs, f
7-3 BWitnEYk _.-
l I
l The design curves for predicting the shif t will continue to be modified as more !
data become available until thet time, the design curves for predicting the !
RTNDT shift as given in Regulatory Guide 1.99. Revision 2, are considered adequate for predicting the RTNDT shift of those materials for which data are not l available. These curves will be used to establish the pressure temperature i operational limitations for the irradiated portions of the reactor vessel until l
the time that new prediction curves are developed and approved.
l; The lack of good agreement of the change in Charpy USE is further support of the !
inaccuracy of the prediction curves. Although the prediction curves are i conservative in thst they generally predict a larger drop in upper shelf than is '
observed for a given fluence and copper content, the conservatism can unduly. [
restrict the operational limitations. These data support the contention that the l USE drop curves will have to be' revised as more reliable data become available; I until that time the design curves used to predict the decrease in USE of the i contmiling materials'are considered conservative.
J 7.3.
Reactor Versel fracture Touchness i
An evaluation of the reactor vessel end-of life fracture toughness and the pressurized thermal shock criterion was made and the results are presented in ,
Table 7 5.
l
, The fracture toughness evaluation shows that the controlling weld metal may have
- an end-of life Riuor of 228F based on Regulatory Guide -1.99, Revision 2. This
( predicted shift is excessive since the latest capsule _(Capsule 00111-0) exhibited
! a measured shift of 140F for a fluence of 1.45 x 10" n/cm'. Ratioing this l measured shift to the T/4 wall location fluence, it is estimated that the end of-life RTc shift will be significantly less. than the value .erdirNd using- a Renulatory Guide 1.99, Revision 2. This reduced shift permD e the cahulation L
of iess restrictive pressure temperature operating limitations tun if Regulatory.. l Guide 1.99, Revision 2, was used.
The pressurized thermal shock evaluation: demonstrates that the Oconee Unit 3' reactor pressure vessel is well below the screening criterion limits and, therefore, need not take any additional ~ corrective _ action 'as required by the regulation.-
L 9
7-4 S W 51121H & - ,
c . -.- - - - -..a - -..- . .- .a u - :
An evaluation of the reactor vessel end of i,r'e upper shelf energy for each of the materials used in the fabrication was made and the results are presented in Table 7 6. This evaluation was made because the weld metals used to fabricate the reactor vessel are Linde 80 flux, low upper shelf energy, relative high copper and are expected to be highly sensitive to neutron radiation damage. Two uethods were used to evaluate the radiation induced decrease in upper shelf energy. The method of Regulatory Guide 1.99, Revision 2, which is the same procedure as used in Revision 1, and the method presented in BAW 1803"' which was developed specifically to address the need of an estimating method for this class of weld metals.
The methods of Regulatory Guide 1.99, Revision 2, iridicate that two of the welds may be expected to decrease below 50 f t lb level prior to [0L. However, BAW 1803 shows that none of the, materials used in the fabrication of the reactor vessel etil have an upper-shelf energy below 50 ft-lbs through 32 ErpY design life based on the T/4 wall location. Regulatory Guide 1.99 method also predicts a decrease below 50 ft L s for the controlling weld metal at the vessel inside wall.
7-5 B W!!?aY% % r
litbl e 7 - 1. Comn rison of Cansule 0C11_l:D lension Test Results Clevated Temp.
Room Temn. Test- _lest (580f)
Unirr Iraj Unin JrQL i Base Metal -- Ati!LLtl. Transverse fluence, 10 I9 n/cm' (> 1 MeV) 0 1.45 0 1.45 Ultimate tensile strength, Asi 85.4 91.5 35.0 87.2 0.2% yield strength, ksi 63.1 (6.9 55.7 60.3 Uniform elongation, % 13.8 9.9 11.9 9.9 Total elongation, % 30.4 26.6 28.6 23.3 Reduction of area, % 66.8 65.1 66.5 62.0 Weld Metal Wi-209-1 Fluence, 10 I8 n/cm' (> 1 MeV) 0 1.45 0 1.45 Ultimate tensile strength, ksi 90.5 111.3 87.9 102.9 0.2% yield strength, ksi 75.0 98.0 67.4 86.6 Uniform elongation, % 12.5 10.7 10,9 9.0 Tott' elongation, % 28.1 24.2 21.4 17.7 Reduction of area, % 62.9 56.7 52.1 49.4 i
l l
l 7-6 B W !!K E?i %
Table 7-2.- Suseary of Oconee Unit 3 Reactor Vessel Surveillanca capsules Tensile Test Results .
Strenoth. ksi Ductility. %
F Test Total Reduction Material -10]gence-n/cm 2
Temp, F Ultimate % I.I Yield %g.) Elon. %g.) of Area %g.)
Base Metal 0 70 85.4 -
63.1 30 -
67 -
(ANK-191) 580 85.0 -
'55.7 -
29- - 67 -
0.81 70 87.5 + 2.5 63.0 - 0.2 30 00 68 + 1.5 580 86.4 + 1.6 57.2 + 2.7 29 00 67 0.0 3.12 69 105.6 +23.7 91.3 +44.7 25 -16.7 56 -16.4 585 86.9 + 2.2 56.0 + 0.1 37 +27.6 64 - 4.5 14.50 70 91.5 + 7.1 66.9 + 6.0 27 -10.0 65 - 3.0 550 87.2 + 2.6 60.3 + 8.3 23 -20.7 62 - 7.5
~ Weld Metal 0- 70 90.5 -
75.0 -
28 -
63 -
4 -(WF-209-1) 580 87.9 -
67.4 -
21 -
52 -
0.81 70 93.9- + 3.8 79.3 '+ 5.7 27 - 3.6 63 0.0 580 93.7 +-6.6 72.6 '+ 7.7 23 + 9.5 60 +15.4 3.12~ 69 105.0 +16.G' 90.6 +20.8 25 -10.7 57 - 4.8 585 85.6 - 2.6 57.8 -14.2= 26 +23.8 ~56 + 7.7 14.50 70' 111.3 +23.0 98.0 .+30.7 24 -14.3 57 - 9.5 550 102.9 +17.1 86.6 +28.5 18 -16.7 49 - 5.8 9-e (.)chan9e r.1 tive to unirradiated.
1 Table 7-3. Observed Ys. Predicted Changes for Cggsule 9C111 D Irradiated ,
Charpy impact Properties - 1.45 x 10 n/cm (E > 1 MeV) i l
Predicted i' 1.99/2 ,I RG 1.99/2+M IbI Predictef Material Observed RG Increase in 30 ft 'b Trans. Temo.. r l Base material (ANK-lh')
i Transverse 31 22 33 ;
Heat affected zone 74 22 33- l Base material (AWS-192)
Transverse 45 22 33 Heat affected zone. .1 22 33 l Weld metal (WF 209 1) 140 196 252 Correlation material (HSST plate 02) 119- 141 197 Decrease in Charov USE. ft-lb -
Base material (ANK-191)
Transverse 19 17 N.A.
Heat affected zone 22 11 N.A. .I i
Base material (AWS-192)
Transverse 19 12 N.A.
Heat affected zone 6 13 N.A.
Weld metal (WF 209 1) 24 30 N.A.
Correlation material (HSST plate 02) 36 36 N.A.
i
(*)PerR.G.-1.99, Revision 2,May1988. .)
(b)PerR.G.1.99, Revision 2,May1988,'plus2xmargin. l j
7-8 )
13 sana 1 MAL- 1 t1
_ _ . . _ .___,,,.u.,__,_.__._.__. _ . _
- -- J
Table 7-4. Sumary of Oconee Unit 3 Reactor Vessel Surveillance Capsules Charpy Irpact Test Results .
Transition Temperature Increase. F Upper-Sheli 30 ft-lb Predicte Enercy. ft-lb Fgence,r Material 10 n/cm Observed 30ft-lbgy Predicte4) 30 ft-lbi- Observed Predicted'4 Base material (ANK-191)
Transverse 0.81 9 8 12 130 136 3.12 32 14 ZI 123 131 14.50 31 22 33 125 127 Heat-affected zone 0.81 0 8 12 145 85 3.12 Ind. 14 21 67 82 14.50 74 22 33 68 79 Base material (AWS-192)
Transverse 0.81 63 8 12 100 108 3.12 19 14 21 99 103 14.50 45 22 33 93 100 i Heat-affected zone 3.12 Ind. 14 21 123 105
$ 14.50 1 22 33 108 101 Weld metal (WF-209-1) 0.81 48 70 I05 54 " 50 3.12 64 121 177 49" 44 14.50 140 196 252 42" 36 l 96 104 Correlation material 3.12 39 87 131 (HSST plate 02) 14.50 119 141 197 94 94 g (a)Per R.G. 1.99, Revision 2, May 1988.
(b)Per R.G. 1.99, R evision 2, plus margin.
"Per 8AW-1803, Revision 1, dated March 1991, plus margin.
on h " Upper-shelf energy value re-defined per ASTM E185.
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- 8. DETERMINATION OF P.EACTOR COOLANT-PRESSURE BOUNDARY PRES $URE . TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Oconee Unit-3 are established in accordance with the requirements of 10CFR50, Appendix G. The methods and criteria employed to establish operating pressure and temperature limits are ' described in topical- report BAW 10046A." The objective of these limits is to prevent nonductile failure during any normal operating condition, including anticipated- operation' occurresnces and system hydrostatic tests. The loading conditions of-_ interest include-the following:
- 1. Normal operations, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests.
- 3. 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 beltline region have been identified as the only regions of the reactor vessel (and consequently of the RCPB)~ that regulate the pressuretemperature 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-periods. The reactor vessel outlet no:zle also affects the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses 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 radiation exposure, the RT NDT of the ' beltline region materials will be high enough that= the_ belt 1.ine region of the reactor ve'ssel 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.llmits 8-1 BWRR8NQLa
_ _ _ . _ . . _ . . _ _ . . - - ~ . . _ _ _ _ _ _ . -
_._-_.._..m _ _ _
i imposed by the closure head region the outlet nozzle, and the beltline region..
l The maximum allowable pressure is taken to be the lowest of the three calculated pressures.
l The limit curves for Oconee Unit-3 are based on the predicted values of the t adjusted reference temperatures of all the beltline region materials at the end :
of fifteenth EFPY. The fifteenth EFPY was selected as the last time period because it represents a logical sequence from the previous analysis and the ;
survelliance capsule was scheduled to be withdrawn at the end of the refueling l cycle wher the estimated capsule fluence corresponded to approximately the inside [
surface end of-life value. However, the use of low leakage fuel cycles caused 1 the capsult fluence to exceed the reactor vessel inside st'rface end of-life fluence. The time difference between the withdrawal of this surveillance capsule i and future operating requirements provides adequate time for re establishing the operating pressure and temperature limits for subsequent periods of operation-_ '
beyond the current surveillance capsule withdrawt.1.
The unirradiated impact properties were determined for the surveillance beltline region materials in accordance with 10CFR50, Appendixes G and H. For the other seltline region and RCPB materials for.which the measured properties are not available, the unirradiated impact properties and residual.. elements, as originally established for the beltline region materials, are listed in-Tabic A-1. The adjusted reference temperatures are calculated by adding the predicted _
radiation induced RT NDT and the unirradiated RT The predicted RT IS NDT. NDT 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' supporting information for Figure 8-1 is described in Section 6. The neutron fluence-values of Fire 8-l' are the predicted fluences that- have been demonstrated (Section 6) w 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.
The neutron fluences and adjusted RTNDT values of the beltline region materials :
at the end of the _ fifteenth full-power year are listed.in Table 8-1. -The neutron -
fluences and adjusted RT values are given for the:1/4T_ and 3/4T vessel wall NDT l
8-2 S W B MIFIM & nr
. _. ._ .._, . _ . _ _ _ . . __. _ __._a_. u,.,_ 2 n _ ;_ -
locations (T wall thickness). The assumed RT NDT f the closure head region and the outlet nozzle steel forgings is 60F, in accordance with BAW-10046. The RTuo, 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.
In the case of the 3T/4 location, one weld metal made up the outside 25% of the weldment and another made up the remainder. It was assumed that the theoretical T/4 thickness flaw used to calculate the pressure-temperature curves layed such that its forward most crack front was posed to enter the material composing the major portion at the weldment and, therefore, that material is controlling.
I 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's pressure-tempera-ture limit curves for normal cooldown. Figure 8-4 shows the vessel's heatup and cooldown limitations during inservice leak and hydrostatic tests. All pres-sure-temperature limit curves are applicable up to the fifteenth EFPY as indicated. Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits of the pressure temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation 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 criticality limit curve. To establish the pressure-temperature limits for protection against nonductile' failure of the RCPB, the limits presented in figures 8-2 through 8-4 must be adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel controlling the limit curves. This is necessary because the reactor vessel is the most limiting component of the RCPB. !
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Figure 8-3.
Cooldown. Applicable for First 15 EFPY - Oconee Nuclear Station. Unit-3 2400 Assomed ITRDg,I Foint Pressere, psig feep., f F
2200 seitti.e aegie. 1/51 is0 a 435 70 Beltline Beglen 3/5T 139 3 W) 150 Closure need Regles 60 C 75 5 725 g 2000 o.tline sente 60 e 1555 rso E E 2040 320
- I 7250 370 g
g
- 1800 Ibe acceptable pressure-teeperature coebinations are below and is the b right of the llelt curve (s). The lielt cerves da set laclude she U pressere differential between the peint of system pressere g 1600 ee.s.reseet e.d the pressere e. the reacter eesset re,ien co.tretti.,
g the llelt curve, ser da they laclade any ad(Itional margin of safety r ' for possible instrosent errar.
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5 50 100 150 200 250 300 350 400 Reactor Vessel Coolant Temperature. F L---__... . . . _ . . . . . .
1 1 ~
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i Figure 8-4. Reactor Vessel Pressure-Temperature Limit Curves for Inservice leak and i Hydrostatic Tests. Applicable for First 15 EFPY - Oconee Nuclear Station. Unit-3 2600 Assomed Eigg, I Penst - Pressure, psig f rep., f O Beltline Region 1/41 18 0 A 577 70 I
, e_
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'25 #"5 d
u 2000 II" ' '
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- 9.
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in the third surveillance capsule, OCill-D, removed for evaluation as part of the Duke Power Company, Oconee Nuclear Station Unit 3, Reactor Vessel Surveillance Program, led to the following cnnelusions:
- 1. The capsule received an average fast fluence of 1.45 x 10"' n/cm' (E >
1.0 MeV) . The predicted fast fluence for the reactor vessel'T location at the end of the eleventh fuel cycle is '2.18 x 10 n/cm' /4(E
> 1 MeV).
- 2. The fast fluence of 1.45 x 10 n/cm2 (E > 1 MeV) increased the RTN of the capsule reactor vessel core region shell materials a maximum Of-140F,
- 3. Based on the calculated fast flux at the vessel wall, an 80% capacity factor and the planned fuel management, the projected fast fluence that the Oconee Unit-3 reactor pressure vessel inside surface will receive in 32 EFPY operation is 9.49 x -10 n/cm2 (E > 1 MeV).
- 4. The increase in the RT for the shell forging material was~ not in good agreement with thakDfredicted by the currently used design curves of RT NDT versus fluence .(i.e., R.G.1.99/Rev. 2).-
- 5. The increase in the RT for the weld metal was significantly _less than that predicted by tNeurrently used design curves of RT versus fluence (i.e., R.G.1.99/Rev. 2), however, the prediction Mhniques are conservative.
- 6. values decreased for 32 EFPY because of a decrease in the The estimaRT[N E0L fluence values and are below the PTS screening criteria.
- 7. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used 6 to predict ~ the change in weld metal Charpy upper-shelf properties-due to irradiation are conservative.
8, The analysis of the neutron dosimeters demonstrated'that the analytical techniques used to predict the neutron flux and f_luence were- accurate..
9-1 SW5iMMAL,
- 9. The capsule design operating temperature may have been exceeded during operating transients but not 'qr times and temperatures that would make the capsule data unusable. ihe response of the surveillance capsule was representative of the conditions to which the reactor vessel experienced.
9-2 SW!/44faf4 =r
_..._..m_.- . . _ _ _ _ _ _ . _ _ _.__ _ . _.__ __._ _ . -. _ _
l .
I 1
l
- 10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE !
l Based on the postirradiation test results of capsule OCl!! 0, the following ;
schedule is recommended for examination of the remaining capsules in the Oconee !
Unit 3 RVSP:
Evaluation Schedule ("I Estimated Vessel t Capsule Estimated ca EOL Fluence, n/cm' ID Fluence, 10"' psulen/cm Surfece 1/4T Estimated Data Datt Available\ 0 )
i l
OCIII C"' O.86- 8.lE18 4.5E18 St'andby OC l l i - E"' -1.60 8.lE18 4.5E18 Standby OC l l l- F"' l.60 8.lE18 4.5E18 Standby l (a)ln accordance with BAW-10006A and ASTM E185 82 as modified by BAWil543A, f Rev. 3. t (b) Estimated date based on 0.8 plant capacity factor.
(c) Capsules designated as standbys and wil'l not be evaluated when removed. ,
t 10-1 BY M 1
i 4 I l
- 11. CERTiflCAT1014 The specimens were tested, and the data obtained from Duke Power Company Oconee Nuclear Station Unit 3 reactor vessel surveillance Capsule 0Cill-D were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H.
f ~
A. L. Lowe, Jr., P(E.
3/ P6 8/frlW
/ Date Project Technical Manager This report has been reviewed for technical content and accuracy.
-D.O , i[
H. J. Devan.(Material Properties) d/d/Y/
Date M&SA Unit t ) .h. kAhd N. L. Snidow (fluence Analysis) k/dl'il Date Performance Analysis Unit'
/,$.D W S//fo/ff K. R. Yoon, N.E. (Fracture Analysis) Date H&SA Unit Verification of independent review.
/ &'
' / ;)C ;/$f C64. Si/s,,/
'K. E. Moore, Manager - Date M&SA Unit-This report has been approved-for release.
?/ A& P/
T. L. Baldwin Date Program Manager
.11 1 DWit481 MAR =v
0 4
Revision 1 The revisions to this report were made as stated in accordance with the standard methods and procedures for the original report.
1 f)) 1 e ,% N ,5 lArYWYlNZ A~. 'L Lowe, Jr. /- ' Date ProjectTechnicalMana(ger Revisions have been reviewed for technical content and accuracy.
NM M. J. Dedn (Material Properties
/ I !/2 '92 Date M&SA Unit-t/H n # r.3!9JL L. Petrusha (fluence Analysis) I ~/- Date Performance Analysis Unit
)
Y K. K. Y n P.E. (Fracture Analysis)
WI2/92-Date M&SA Un Verification of independent review.
$$bD ' 642-Q%
K. E. Moore, Manager Dite M&SA Unit This report approved for release.
1..L. Baldwin W 6"b.9 .L Date Program Manager 11-2 SWBNNhv
U APPDlDIX A Reactor Vessel Survei_llance Program Background Data and information A-1 SW!!bnFe%
_x........ .
- - - - - -a
-. ..-. - -. ..-.. - _ .-. - - - . - . - . - . ~ - - . . - . - - - . ... . . .-.
l
- 1. Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E-185 66, are shown in Table A 1. The locations of these materials within the reactor vessel are shown in figure A 1.
I
- 2. Definition of Beltline Reaion i l
The beltline region of Oconee Nuclear Station Unit 3 was defined in accordance !
with the data given in BAW-10006A.'
- 3. Caosule identification .
The capsules used in the Oconee Unit +3 surveillance program are identified below by number and type.
Caosule Cross- Reference Data 10 No. lyM '
OClll- A : V 00111 8 VI q OCIII C V .
0C111-0 VI-OClli-E V OCill-F VI-
- 4. Specimens Per Surveillance Caosule See Tables- A 2 and A-3.
8 A-2 ,
S W 5 M rli h r1 .
, .+a.n,--- ,vae,,,-- ,,,,,,,-,.,,n., ,. ,. - ,, ,.,,,,.-J,,-m.,.,..n,._., ..,.-,,,-,._,,.,,,,.gA--, .,,,,.,.,,.,.,-+,,-,,,>y.
l Table A-1. Surveillance Program Material Selection Data for Oconee 3 -
Charry Data, C p Beltilne Midplane Drop TranJverse Maleria_1_id nt_tf_igal_f_qr!
t i Region to Weld Weight longitudinal Heat No.
._50 ET Ch p istry. %
Type location CL, cm TET, F 9 10F, ft-lb ft-lb 35 MtE USE bT. Cu P 5 Ni AMX 77 5A508 C1 2 Lowr nortle --
90, 121, 106 0.06 0.006 C.009 0.74 belt 103, 91, 128 AAW 163 SA508 C1 2 Upper shell --
20 62, 77, 40 0.04 0.006 0.012 0.78 AWG 164 5A503 C1 2 Lower shell --
20 82, 83, 90 -- -- -- --
0.02 0.010 0.010 0.78 WF-154 Weld Circum seam 123 --
41, 37, 43 -- -- -- --
0.20 0.015 0.021 0.59 WF-25 Weld Circum seam - 63 --
33, 28, 49 -- -- -- --
0.29 0.019 0.010 0.71 WF-182 Weld Clrtum seam -249 --
- 35. 40, 30 -- -- -- --
0.22 0.024 0.006 0.58 L Wr-209-1A Weld Surv. weld -- --
29, 30. 32 -- -- -- --
0.34 0.013 0.010 0.48 G
B sE R*
4
Table A 2. Materials and Specimens in Upper Surveillance Caosules OClll A. 00111-C. and OClli-E Number of Soecimens Material Description Tension Charov Weld metal, WF 209-1 2 12 Heat affected zone (HAZ)
(
Heat A ANK 191, longitudinal 0 12 Baseline material Heat A - ANK-191, longitudinal 0 9 Heat A - ANK 191, transverse 2 12 Heat B - AWS 192, transverse Q t Total per capsule 4 54 Table A 3. Materials and Specimens in Lower Surveillance Caosules 0C111 B. 0C111 0. and OClli-F Nismber of Specimens Material-Descriotion Tension [hnny Weld metal,- WF 209-1 -2 12.
'd, HAZ Heat A - ANK-191, longitudinal 0 12 Heat B - AWS-192, longitudinal 0 6 Baseline material - -
Heat A - ANK 191, transverse 2 12 Heat B - AWS-192, transverse 0- 6 Correlation HSST plate 02- Q 1 Total per capsule 4 54 A-4 SWhy
__ -_ - - - - = = - - - - - - O
f;gure A-1. Location and identification of Materials Used in Fabrication of Reactor Pressure Vessel 1
- I m
l
\
t
( .,
y r
RDH-4680 (Lower Nortle Belt)
_ : WF- 200 g AVS-192 (Upper Shell) l
_ g_ W-67 (75% ID)
W 70 (25% OD)
AhK-191 (lever Shell)
I WF-169 4
i )
- 1. 417525-1 (Dutchman)
)
A5 SWAWN9 h r.
l
1 b
APPENDIX B Pre-Irradiation Tensile Data B.1 BW!!K We % r
Table B 1. Pre trradiation Tensile Properties of Shell Plate Material llent ANK-191 Test Strenath. osi Elonaation. % Red'n of Specimen No. Temo. F Yield. Ultimate Uniform lotal attam_1_
Transverse JJ 603 RT 70,070 85,060 13.4 30.7 66.9 604 RT 60,940 86.520 14.8 31.1 68.4 600 RT 58,310 84,730 13.3 29.3 65.1 Hean RT 63,110 85,440 13.83 30.37 66.8 Std. dev'n. 5,040 780 0.68 0.77 0.55 JJ 607 508 56.200 85,300 12.26 28.6 66.1 S10 580 56,220 85,580 13.82 28.6 61.1 611 580 54,600 84,150 9.6 28.6 67.5 Hean 580 55,670 85,010 11.89 28.6 66.57 Std. dev'n. 5 700 620 1.74 0 0.66 Table B 2. Pre-trradiation Tensile Properties of Weld Metal .Lonoitudinal . WF-209-1 Test Strenath, osi Elonaation. % Red'n of Specimen _th Temo. F Yield Ultimate 1)ltilplJ 191]t1 Area. %
JJ-002 RT 74,630 90,460 12.5 29.3 63.2 004 RT 73,540 89,110 13.1 27.1 63.6 018 R1 76,720 91,910 11.9 27.9 62.0 Mean RT 74,960 90,490 12.5 28.1 62.9 Std. dev'n. 1,320 1,140 0.49 0.91 0.68 JJ 001 580 67,540 87,700 10.83 20.7 52.9 011 580 66,480 88,300 11.54 22.1 53.0 013 580 68,210 87,620 10.33 21.4 50.3 Mean 580 67,410 87,870 10.9 21.4 52.07 Std. dev'n. 15 712 303 0.50 0.57 1.25 i
B-2 BWltKMW
1 APPENDIX C Pre-Irradiation Charpy Impact Data C-1 BW!! nan"o Lur
___m____.___ __.J
Table C-1. Pre-Irradiation Charpy Impact Data forLShell Forging-Material - Transverse Orientation. Heat ANK-191
-Asborbed Lateral;. Shear Specimen: Test Temp., Energy, Expagsion, Fracture, No, -F ft-lb 10 in, %
JJ-651 362 172 100-640 361 144= 66- 100 621: 357 128- 72 100 JJ-701 284 136 73 100 705 283 140 73 100 700 277 135 71 100 JJ-635- 200- 150 100 625 200 153 .67 100 637 200 141 73. 100.
JJ-704 140 130 74 88-703 140. 126- 74 85 702 140 118 71 70 JJ-676 80 123- 71 55 611 79 130 78 65-657 79 :129 74- 60 JJ-670 41 77 62 5 685 41- 71- 58 4-612 41 86 66 10 .,--
JJ-708 25 68 55 5 706 24 -74: 62 -- 10 707 ~28 57 46 8-
> 75 21- 58: 48- 6 43 20- 47 41- 4 JJ 22 0.9 35 26 1 --
'2 - 0.9 56 44-- - 2. 1 580 0.5 18 15- 0; JJ-698 -39 3 2 -0 697- -40 8 8 0-699 :14 12 0 C-2 ,
S W BlW!l11ML y: U a - _ _ - - - - -
i
- . l Table C-2. Pre Irradiation Charpy Impact Data forIShell forging ,
Material - Transverse Orientation. Heat AWS-192 Asborbed lateral Shear .
Specimen Test Temp.,- Energy, Expagsion, fracture, No. F ft-lb _10 in. -
KK-665 360- 98 77 .100=
641 360 103 70 100 634 359 109 72 100-KK-676 282 +
111 71 100 677 280 106 71- 100 674 279 104 70 100- r KK-619 200 115 71 100 627 199 108 68 100 669 199 115- 72 100 KK-683 140 119 73 100 659 140 92 90 '
679 139 113 67 -100 684 139 114 71 100 KK-622 81- 64 51 ;8 651 80 83 62 -- 2 5 648 80 99 71 35 KK-675 61 80 65 40 673 60 80 64 35 '
678 60 64 54 25 610 60 50 44 15.
KK 614 40 51 43 6-630 40 40 36 3-639 40- 59 50' 12~
KK-657 0 37 30 <1 633 0 53 42 4 666 0 23 18 <1 KK-681 -59 2- -
1- 0-682 -60 30 1 -26 J 680 -60 3 2' O C-3 S W J E M M H & n y-.
Table C-3, Pre-Irradiation Charpy Impact Datt for Shell Forging-Material - HAZ Transverse Orientation. Heat ANK-191
-Asborbed . Lateral Shear Specimen Test' Temp.,- Energy. Expagston, Fracture, No.= F
-ft-lb 10 in. %-
JJ-313 360 86 53 100 315- 359 -106 63 100 335 358 73 54 100 JJ-328 201- 102 54: 100 323 201 92 59 100 308 -200- 83- 49 100 JJ-322 80 76' 42 319 80- 98 51 100 344 80 74 45 80-JJ-357 40. 66- 44 85-332 40- '
57 33 45 339 40 55' 30- 55 JJ-301- 20 46 29 70:
304 .20 36 127- 30 JJ-331 0 35 29- 35 324- 0 33 -26 23 337- 0 47 ~33 13 5 JJ-424 -59 27 20 2 420 -60 56 44 4-421 -61 67 35 ;6.
C+4 j SW2f4171HORany j
4 ;
. Table C 4.- Pre-Irradiation Charpy_ Impact Data-for Shell Forging- ,
i Material HA7. Transverse Orientation. Heat AWS-192
-l l
i- Asborbed Lateral- Shear I i Specimen Test Temp., Energy, Expagsion, Fracture, !
in. u
~
No.-
F ft-lb 10 _%
~
l
} KK'-325 361 135- 65. 100 :
- 336 -360 153: 77 100 '
l 318 358 128- 63 -
100 l
KK-343 278. .118 76- -100-
. 339 278 125 76. - 100-i 344 277 132 76 100
! KK-333 199 65 37 -100 310 190 122 66 100 l; 313 197 78 53 100 3,
i KK-346 139 131 77 100-337 139 132 671 100.
348 139 121 71 100
! KK-312 80 91 46 94 3 307 79 76 41- '85 l
335 79 63- 36 98 '
KK-308 40 141 67- 100- '
! 316 40 72 42 35
- 327 40 ~66. 40 45 3_ KK-341 20 93 70 40 i- 340 20 104 77- ' 65-e 345 19 82 64 30-1 l' KK-317 0.5 106
- 64 45-
! 324 0.1 43 31 18-326 0.1 ~43 32 40 KK-309 30 22 20 j 330 -39 80 51 28-KK-342 -39 27 21 :2 347 -40 52- 35 5--
338 38- 28 3 4
4 b 'l
< u 1
[ C-5 L SWB*amidbr ,
'l
, .. ,. - - . . - . , . . . - - - - -. ._ - , , ,..,..-.,-..a,
I?ble C-5. Pre-Irradiation Charov Imonet Data for Weld Metal WF-209-11 Asborbed lateral . Shear.
Specimen Test Temp., Energy, Expayston, Fracture, in.
No. F .ft-lb 10 .%
JJ-063 359 -57 49 100-083~ 357 58: 48 100-089- 357- 65 53 100 JJ 091 201 57 46- 100 053 201 '63 50 100 084 199 78 57 100 JJ-093. 140 60. 46 100 060 2140- 69 52- 1100 065 140 61 49 100 JJ-092 110 64- 46 30 038 110 58 46. 100-077 109 47 37 96-JJ-064 80 61- 40 98-095 80 38 32 95 081 79 48 35 85 JJ-090 40 28 28 10 069 40 33 -29 15=
07 5- 40 21 20 12 JJ 061 17 16 5-050. -40 16 14 -2 057 -39 19 18 6 i
l C-6 S W B N!r M L ev
- , . =. . ..
5 1
Figure C-1. Charpy impact Data _From Unitradiated Shell Forging j Material (ANK-1911. Transverse Orientation E . i ;i : . i n
4 , 75 - -
4 5,
@ 50 -
- ,{ 25 - -
a e e e e e n
0.10 i i i i , i 4
g 0.05 - -
- o e e .1 i e s -
e 1
f0.06
~
i 5 0.04 -
e -
t 4
3 c 0.02 - -
2 e g - e i t I e e
{
i 220 i i i i i i
--- DATA SumARY --
200 -T g, +20r -
Tgy (35 met) +st 180 -
-Tcy (50 rt-La) +1?F Tgy (30 FT-ta) -9F o lE0 ~Cy -USE (Avr,) ikk ft-lbf -
RT +20r I
. 140 - ot Is 8
- ; I $ e 5 120 ,
Q 4
g 100 - .
l C z -
E o
. sc -
1 e e
40 -- -
MATERI AL ' S A500,tt.2(IL) -
20 -
NONE e Ftuence -
o HEAT No, ' ANK-191 a 1 i t t , ,
100 0 100 200 300 400 500 600 Test Temperature, F C-7 SW#N&v -
Figure C-2. Charpy Impact' Data from Unirradiated Shell Forging Material (AWS-192). Transverse Orientation 10c , , :- ; :, : , ,
o 75 - -
J w
.f 50 m
- e e/o e -
~z 2s 0 ;
d, , , , ,
0,10 , , , , , i 0 06 , -
5 * *
- 0 i e g e e R 0.06 -
- e 0.04 -
2
= 0.01 -
a:
0 -
220 . ,
i i i i
- DATA
SUMMARY
200 -T,37 +20r -
Tgy (35 mug) 19 r 180 -T (50 ri- a) +37r -
T,y (30 FT ts) 4f 160 Cy -USE (Avo) 112 ft-lbf -
e
~
a +?0i 140 -RT" ' -
8 S
lM -
e ,
e a 100 -
e ; -
E *
- 80 -
e -
u l ~ 60 -
e -
- e 40 -
e -
e MAign AL. $A508.Cl.2 (IL) 20 -
- Ftuencg NONE -
HEAT No. AWS-192 0
-100 0 100 , 200 300 400 500 600 Test Temperature, F C-8 B Wit &5"cW h y
J
'i i Figure C-3. Charpy Impact Data from Unirradiated Shell Forging Material (ANK-191) Heat-Affected Zone. Lonoitudinal Orientation l
- _. i i
- i i 4
75 -
g , -
E 1
sM
~
~ e E 25 - -
- l M
0 l
9 1 0.10 i e i i i i
, d.
g 0.08 - -
O i
k0.06 -
o -
0.04' - ,
t
_e, e 5 0.02 -
]
O --
220 . . i e i i
- DATA SumARY -
j 200 - T,3, +20f -
Tcy (35 m.t) .t?r 180 -T y (50 FT-ts) + 32r -
- icy (30 ri-La) -tor g IE0 -
, g.USE (Avn) 90 f t-hf
+20F i_ RT,7
, 10 - -
l 8 a
gm -
4 $ ,
. g 100 -
, -e -
se :
I g --s
,g - e .
l g _
I e e
-i. e e
Matta:AL S A508,Cl.2 (HAl) 20 -
FLugnet NONE -
HEAT No. A N K-191 0
, -100 0 100- , 200 300. 400 500 600 Test- Temperature, F C-9 BWs*tamv
Figure C-4. Charpy impact Data from Unirradiated Shell Forg'ing Material
( AWS-192) Heat-Affected Zone. Lonoitudinal Orientation 100 , : , : i ;, ; , ,
w
- s - -
.t.i y 50 -
- 2 E 75 - -
5 e e g i I e i t 1 0.10 , , , , , ,
g0.08 , , , ,
. e
-k0.06 - I -
O e e e
{0,04 -
, I
- 3 e c 0.02 - 8 -
4
=
0 220 . . i i i
- DATA SumARY -
200 - T,37 +20r .
ICV ( 5 met)
-W 180 -T Cy m pr. W -W -
Tcy (30 FT-La) '%f g 160 Cy -USE (ava) 114 ft-lbf -
a *
+20F
~, 140 -RT" * * -
8 *
=
g- .
120 ~
o * , -
E g 100 - e' -
y e e 2 80 -
e *
-g .
o W . .
g $7:niat S AS06,Ct.2 (HAl)-
20 -
FLUENct NONE -
HEAT No. ._Ayt-19 2 0
-100 0 100 , 200- 300 400 500 600 Test Temperature, F C-10 SW#Ahv
Fiaure C-5. Charov Impact Data From Unirradiated Weld Metal. WF-209-1 2 3 2 2 100 , -
w
, 15 -
t 3
y 50 -
m t.
S 25 - -
a g i t t t t t 0.10 i i i , , i g 0.08 - -
2m kD.06 -
a 2e L
- 3 e 0.04 -
e , -
,3 s 0.02 -
- p. . -
E a
0 IM , , , , , ,
- DATA SumARY -
100 -T oy - -
Tey (35 nt.t) +81I 90 Tey G n-ts) 6 -
g Tey (30 FT-ts) +45r 80 s ~Cy-USE (avo) 66 ft-lbs .
g lit e, -
o 70 -
o e
- e E% -
!.o e 5 w
E . 50 -
e l ' 40 -
.?.
e 30 -
20 - e -
MATERIAL Veld Metal-llnde 80 10 -
FLutset None -
HEAT No. VF-209-1 0 ' ' ' I - '
-100 0 100 - 200 300 400 500 600 Test Temperature, F C-11 SW.'ts,1!mhv
APPENDIX D Fluence Analysis Methodology D-1 S Ws?sifauiimv
- 1. Analytical Method A semiempirical 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,. _
version 4.3. 00TIV calculates-the energy-and space dependent neutron fluxfor the specific _ reactor under consideration. This semiempirical method is conven-iently outlined in Figures D-1 (capsule ' flux)- and D 2 (vessel flux).
The two-dimensional transport code DOTIV was used to calculate the erergy- and space-dependent neutron flux at all points of interest in the reactor system.
DOTIV uses the discrete ordinates method of solution of the Boltzmann transpor.t equation and has multi-group and asymmetric-scattering capability. The reference calculational model is an R-e geometric representation of a plan view through the reactor core midplane which includes the core, core liner, coolant, core barrel,
~
thermal shield, pressure vessel, and concrete. The material _ and geometry model',
represented in Figure D-3, uses one-eight core symmetry. In order to include reasonable geometrie detail within the computer memory limitations, the code parameters are specified as P3order of sr.attering, Se quadrature, and 22 energy _ -
groups. The P3 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 Se symmetric quadrature has generally
- - produced accurate 'results in discrete ordinates solutions for similar problems, and is used routinely in the B&W R-e DOT analyses.
Flux generation 'in the core was represented by a fixed; distributed source which the -code derives based on a 235 0-fission spectrum, the input relative power distribution, and a normalization _ factor to adjust : flux level- to the desired 1 power density.
Geometrical Confiouration
- For modeling purposes, the actual geometrical configuration is divided into three-
_ parts, as shown in Figure D-3. The first -part, Model "A,".' is used to generate-the energy-dependent - angular flux. at the inner boundary 'of Model "B," which begins at the- outer surface of the core barrel, Model A includes a detailed l
l l' D-2 SW."fMiMLv v
.-_-s -
representation of the core baffle (or liner) in R-e geometry that has been checked for both metal thickness and total metal volume to ensure that the DOT approximation to the actual geometry is as accurate as possible for these two very important parameters. The second, Model B, contains an explicit represen-tation of the surveillance capsule and associated components. The B&W Owners Group's Flux Perturbation Experiment 22 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. The magnitude of the perturbations in the fast flux due to the presence of the capsule was determined in the Perturbation Experiment to be as high as 47% at the capsule center and as high as 10% at the inner surface of the reactor vessel. Detailed explicit modeling of the capsule, capsule holder tube, and internal components is therefore incorporated into the DOT calculational models. The third, Model "C,"
is similar to Model B except that no capsule is included. Model C is used in determining the vessel flux in quadrants that do 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 approximatcly 32.5 cm or 17 radial intervals is specified between Model A and Models B or . The width of this overlap region, which is fixed by the placement of the Model A vacuum boundary and the Model B boundary source, was determined by 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 does not cause a perturbation in the flux at the points of interest.
Macroscopic Cross Sections Macroscopic cross sections, required for transport analyses, are obtained with the mixing code GIP. Nominal compositions are used for the structural . metals.
Coolant compositions were determined using the average boron concentration over a fuel cycle and the bulk temperature of the region. The core region is a-homogeneous mixture of fuel, fuel cladding, structure, and coolant.
D3 B W ilE M ale % r
The cross-section library presently used is the (22-neutron group and 18-gamma group) CASK 23E 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.
Distribute'd Source The neutron population in the core during full l power operation is a function of neutron energy, space, and time. The time dependence is accounted for in the-analysis by calculating' the - time-weighted average _ neutron source, i.e. the neutron source corresponding'to the_ time-weighted average power distribution.
The. effects of -the other two independent variables, energy and- space, are accounted- for by usir.g a finite but appropriately large number of discrete intervals in energy and space. In each of these intervals the neutron source is assumed to be invariant and independent of all other variables.: lhe space and energy dependent source function can be considered as the product of a discretely expressed " spatial function" and a magnitude coefficient, i.e.
Svg3g = [v/K Po] x (RPDgX,]~ (D1)- -j magnitude spatial where:
Sv g3g
=
Energy-and space-dependent neutron source, n/cc-sec, v/K -
Fissic neutron production rate, n/w-sec, Po Average power density in core, w/cc, RPD jj
= Relative power density at interval (i,j), unitiess, X, = . Fission spectrum,-- fraction of fission neutrons having energy in group "g,"
i - Radial coordinate index, j = Azimuthal coordinate index, g = - Energy _. group -_ i ndex.- ;
The spatial dependence of the flux-is directly related to the RPD distrib'ution. I Even though the entire-(eighth-core symmetric) RPD distribution is modeled in the D-4 BWHLW18%
. _.- __ _ - . _ _ ~._ _ __ .- _. _
i l analysis, only the peripheral fuel assemblies contribute significantly to the ex-l core flux. The axial average pin by-pin RPD distribution is calculated on a
- quarter-core symmetric basis for 8 to- 12 times -during each core -cycle for the 4
entire capsule irradiation period. The time-weighted average RPD distribution j is used to_ generate the normalized space and energy dependency of the neutron j 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 fluxes in the R-o plane that are averaged
- over the core height, an axial correction factor of 1.17 was required to adjust j these fluxes to the capsule elevation.
i 1.1. Capsule Flux and Fluence Calculation i
j As dist.issed above, the- DOTIV code was used to explicitly model the capsule l assembly and to calculate the neutron flux as a function of energy within'the . .
capsule. The calculated- fluxes were used in the following equation to' obtain l -
calculated activities for comparison with the measured data. The calculated l activity for reaction product 0,, in (pCi/gm) is:
1 D, = Ea o(E) 4(E)j EF, (1-e'*03) e W (T - ri) (0-2) l (3.7 x 10') An E -j ,
where:
N - Avogadro's number, 4 An - Atomic weight of target material n,
! f i - Either weight fraction of target _ isotope in n-th material or the j fission yield of the desired-isotope,-
I
- o,(E) - Group-averaged cross sections for material n-(listed in Table
- E-3)-
, 4(E) - Group averaged fluxes calculated by D0TIV analysis, l Fi_ -- Fraction-.of full power during j-th time interval, t, F
- Ai = Decay constant of the ith isotope, 1
.T - Sum of total. irradiation time, i.e., residual time in reactor, i
e i
D-5 BWifMin3 v
and the wait time between reactor shutdown and counting-times', ,
rj - Cumulative time.from reactor startup to end of j-th time period.-
7 t 3- Length of the j-th time period Adjustments were made to the calculated dosimeter activities to correct for the effects listed below:
Short half life adjustments to Ni and Fe dosimeter activities 238 Photofission adjustments to 0 and 237 Np dosimeter act'ivities Fissile impurity adjustments toL 2380 ' dosimeter activities After making these adjustments the calculated dosimeter activities were used with the corresponding measured activities to obtain the flux normalization factors:
D, (measured)l C, = ,
D, (calculated) .
These normalization factors were evaluated, averaged, and then used to adjust the calculated _ test specimen flux,and fluence to be consistent with the dosimeter-measurements. Additionally, the norma'lization factor was used to--update the:
average normalization factor which had been derived from previous analyses. The updated normalization factor was then used-to adjust the calculated vessel' flux and fluence. The flux normalization -factors are given in Table. D-1. L
- 2. Vessel Fluence Extrapolation For past core cycles, fluence: values in the pressure vessel are calculated as L described above. Extrapolation- to future cycles is . required to predict the useful vessel life. Three time periods-are considered inithe extrapolation: 1) operation to date for which vessel fluence has been calculated, 2) future fuel cycles for which PDQ calculations have been performed, and 3) future cycles for which_no analyses exist.
For the Oconee Unit 3. analysis, time period 1 is through cycle 11, time peri 6d 2 covers cycles 12 and 13, and time period 3Jcovers from the end of cycle 13 through 32' EFPY. The flux and fluence 'for time period 2 was estimated- b; calculating the vessel flux using an adjoint-DOT calculational procedure with the'-
appropriate assembly-average power distributions and _ integrating these values D-6 SWMJtthy L
, _ _ ~, _
over time period 2. The extrapolation of the fluence through time period 3 was accomplished by assuming that the average flux during period 3 was equal to the average flux for period 2 (cycles 12 and 13), i l
Table 0-1. Flux Normalization Factor Measured Calculated Flux Activity,I") Activity,(b) Normalization uti/a uCi/a factor 1
54Fe(n,p)S4Hn 935.27 1016.73 0.92(C) 58 Ni(n,p)58Co
- 1363.12 1397.86 0.98(d) 238U (n,f)l37Cs 18.35 17.29 1.06 237Np(n,f)137Cs 101.50 97.94 1 14 Averaged: 1.00I ')
< (8) Average of four dosimeter wires.
(b) Average at four calculated activities.
(c) Average at four ratios (one for each dosimeter wire) corrected by short half-life factor of 0.83.
(d) Average of four ratios (one for each dosimeter wire) corrected by short half-life factor of 0.77.
(*) Average of all four dosimeters was selected as the normalization constant.
D-7 BWs*t? vent %%m
i l
Table D-2. Oconee Unit 3 Reactor Vessel Fluence by Cycle
=
Vessel Fluence, n/cm Jc) incremental Cumulative Vessel lux, ,
Cycles Time, EFPY Time, EFPY ti/cm s : Incremental Cumulative 1 1.31 1.31 1.39E+10 '0,58E+18: 0.58E+18 2-5 2.79 4.10 1.55E+10 1.37E+18 1.94E+18 6-11 6.56 10.66 9.49E+9 1.96E+18J 3.91E+18 12 1.12 11.78 8.92E+9 3.16E+17 4.23E+18 13 1
- 12.91
-7.62E+9(*) 2.70E+17(b) 4.50E+18(b). ,
2.09 15.00 8.27E+9(") 5. 47 E+ 1'7 ( b)-5.05E+18(b) 6.00 21.00 8.27E+9(a) 1.57E+18(b) 6.62E+18(b)_ ,
3.00 24.00 8.27E+9(a) 7.83E+17(b) 7.40E+18(b) f 8.00 32.00- 8.27E+9(a) 2.09E+18(b)- 9.49E+18(b)
(*)The fuel cycle designs for future cycles-12 and-13.were used to estimate the maximum neutron flux at the inside surface of reactor vessel.
(b) Extrapolated values. -
(c) Peak fluence at inside surface of reactor vessel.
J D8 ,
BWWJ!NLv.
, r - y <, g 9- ,- # r w 4 -.e-- , - ,- - - - -
Figure D 1. Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Capsule
-l DEF/84 0055 SECTIONS _ GEDETRY &'GMDRATUtE PGet DISTRI- 1 DEF/85 00$DETUt RfAC- FUt KEE. A DDT OLffEN5 SDG 1 TUR 0055 SECTIONS .CAPSIAE INSER-
, TEN -(P00)
EIP y
- SGutEL +
0105$ SECTIONS QEE 3 r
" TDE AVE DISTRI-GEDETRY &. DDT 4-- BUTED SOL 81CE Sv mW3tATt5tE --+ KEEL A (E R, e )
KDEL 8-ir ir ir DDT 4 #G1AR FIBK MDEL B c AT 845tEL PGet HIS15ty I DOSDETER -(F CAPStLE
- 5 ACTIVITIES (PIDEST (IEE) i i
- ir t
FINAL <
CALOLATED ACTWITIES L a 1 EASWtED !
(- 00SDETER i ACTIVITIES !
AXIAL 1 QIIRECTEM FACTUt u
ir- r i
CAPS 11E FIDX l IGMALIZATEN FACTGt
- M/C RATID ;
l_ D-9
- SWEIMIPMLv
Figure D-2. Rationale for the Calculation of Neutron flux in the Reactor Vessel GEDETRY & a%DRATlRE FIR M] DEL 'A' DOT POER DISTRI-MICROSCDPIC CROSS BlfTIONS SINCE SECTIONS E!0F/B4 STARTUP (PDQ DOF/B5 (R EIU1VALEND 3r 3r ir MACROSCDPIC CROSS 53REL CIDE SECTIONS BY E-G OLP
" GIP" CIDE TM AVE DISTR.3UTED SOLRCE SV (E, R,e )
ir y +
DOT 4 GCOEL A)
GEDORY NO EMDRATIRE FCR DDT KDEL B CR C ir u DOT 4 KDEL B ANGLAR Ft1R AT 4 #0/0R MIEL C 4- BARREL SLRFACE PCRMALIZATIIN FACTER FRM AXIAL CIRRECTION CAPSlLE FilBCE ANALYSIS FACitR (FRCH 11E DIMRAM (N llE PREVIOUS PAGE)
<r <r v TI)E-AVERAGE VESSEL FUK AT MXIMN VESSEL LDCATICH (E, R,e)
D-10 BWUna M ar
l i
Figure D-3. Plan View Through Reactor Core Midplane -
l (Reference R-0 Calculation Model) t Y ~
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.-6 8
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o :
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i tit i L15
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3.
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- wie 8 nis Mi
$u s o . - e- M.jor A.is o Model A Models 5/C x
s APPENDIX E Capsule Dosimetry Data E-1 SW!!s!F5hr ,
Table E 1 lists the characteristics of the neutron-dosimeters. Table E-2 shows-the measured activity per--gram of target material (i.e., per gram of uranium, nickel, etc.) for the capsule dosimeters. Activation cross sections-for the various materials were flux-weighted with the 235 0 fission spectrum shown in Table E-3.
Table E-1. Detector Comoosition and Shieldino Detector Material - % Taraet Shieldina Reaction 238 238U (n,f)I37Cs U Al 10.38% 0 Cd Ag _
23 237 Np(n,f)l37Cs Np-Al 1.44% Np Cd Ag 68 Ni 67.77% Ni Cd-Ag Ni(n p)58Co' 59 59 Co(n, y)60Co Co-Al 0.66% 00 Cd Co-Al 0.66%5900 None 59Co(n,y)60Co 54 54 Fe 5.82% Fe None Fe(n.p)S4Mn-Table E-2. Measured Specific Activities (Unadjusted) for Dosimeters in Caosule OC3-D '
Dosimeter Activity. (uti/am of Taraet)
Detector Material Dosimeter-Reaction DD5 DD6 DD7 DDB Ni 58Ni(n,p)SBCo 1252.84 1725.77 1196.49 1640.06:
Fe 54Fe(n,p)S4Mn '802.15 1091,91 812.36 1034.64-U-Al 238U (n,f)137Cs 15.44_ 22.20- 14.81 20.94 237 Np(n,f)l37Cs Np-Al 89.98- -120.64 85.60- 109.74 1 l
E-2 BWhiMNW
Table E 3. Dosinster Activation Cross Sections, b/ atom (a) l 1
237 58 54 G Energy Range, MeV Np(n.f) 238U (n,f) Ni(n p)- Fe(n p) -
1 12.2 - 15 2.323 1.051E+0 4.830E-1: 4.133E-1 2 10.0 - 12.2 2.341 9.851E-1 5.735E-1 4.728E-1 3 8.18 - 10.0 2.309 9.935E 5.981E 4.772E-1 4 6.36 -
8.18 2.093. 9.110E-1 5.921E-1 4.714E-1 5 4.96 -
6.36 1.542 5.777E-1 5.223E-1 4.321E-1 6 4.06 -
4.96 1.532 5.454E-1 4.146E-1 3.275E-1 7 3.01 - 4.06 1.614 5.340E-1 2.701E-1 2.193E-1 8 2.46 -
3.01 1.689 5.325E-1 1.445E-1 1;080E 9 2.35 -
2.46 1.695 5.399E-1 -9.154E-2 5.613E-2 l 10 1.83 -
2.35 1.676 5.323E-1 4.856E-2 2.940E-2 11 1.11 -
1.83 .l.594 2.608E-1 1.180E-2 2.948E .
12 0.55 -
1.11 1.241 9.845E-3' l.336E-3 . 6.999E-5 13 0.111 - -0.55 2.352E-1 2.436E-4. 5.013E-4 6.419E-8 1.
l 14 0.0033 0.111 1.200E-2 6.818E-5. -1.512E-5 0 i -
-(*)ENDF/85 valygg that have been fluv weighted -(over CASK energy: groups).
based on a U fission spectruo in the fast energy range plus a 1/E shape in the intermediate energy range.
E-3 SWAfA31WLv d d, = W D e W ++ e w, -w s g 4W-*=-w
+- h*:
1*"-rT 1
E 1
l APPENDIX F References l
i l
l t
1 F-1 nureswsuctzan EdwSERVICE COMPANY
a
- 1. . A. L. Lowe, Jr.. et_ al., Integrated Reactor Vessel Material Surveillance Program, BAW-1543A. Rev. 2, Babcock - & Wilcox, - Lynch' urg, Virginia, May 1985, and Addendum 1, July 1987. -
- 2. A.- L. Lowe, Jr., et al., Analyses of- Capsule OCIII-A from- Duke Power Company Oconee Unit-3, Reactor Vessel Materials Surveillance Program, SAR-lila, Babcock & Wilcox, Lynchburg, Virginia, July 1977.
- 3. A. L. Lowe, Jr., s.t_al., Analyses of Capsule OCIll-B from Duke Power Company Oconee Nuclear Station Unit-2, Reactor Vessel Materials Surveil-lance Program, BAW 1697, Babcock & Wilcox,- Lynchburg, Virginia, October _
1981.
- 4. G. J. Snyder and G. S. Carter, Reactor Vessel Material Surveillance Program, Revision 3, BAW-10006A. Revision 3, Babcock & Wilcox Lynchburg,_
t- Virginia, January 1975.
- 5. American Society- for To ting and Materials, Recommended Practice for Surveillance Tests on Structural Materials in Nuclear-Reactors. - E185 66, November 1966.
- 6. Code of Federal Regulation, Title 10, Part 50 Fracture Toughness Re-quirements for Light-Water Nuclear Power Reactors, Appendix -G, Fracture-Toughness Requirements.
- 7. Code of Federal Regulation, Title 10, Part'-50, Fracture Toughness Re- '
quirements for Light-Water Nuclear -Power Reactors, Appendix _ H Reactor-Vessel Material Surveillance Program Requirements.
- 8. American Society of Mechanical Engineers (ASME) . Boiler and Pressure Vessel Code Section III, _- Nuclear Power- P1 ant L Components, ' Appendix G, Protection Against Nonductile Failure-(G-2000).
9.
~
Heavy .Section Steel Technology: Program, Semiannual - Progress Report for Period Ending February 28,.1969, ORNL-4461, Oak Ridge National Labora-tory,'0ak Ridge, Tennessee, January 1970.-
- 10. American Society for Testing and Materials, . Methods and Definitions for- q Mechanical Testing of Steel Products, A370-77, June:24,1977. l F-2 B W NA W1M L w
, m ..
-11. American Society for Testing and Materials, Methods for Notched Bar impact Testing of Metal' c ttaterials, E23-82, March 5,1982.
- 12. S. . Q. King, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAW-1485P. Revision 1, Babcock & Wilcox, Lynchburg, Va., April.1988.
- 13. B&W's Version of DOTIV Version 4.3, One. and Two-Dimensional Transport Code System," Oak Ridge National Laboratory, Distributed by the Radiation Shielding Information Center as CC-429, November 1, 1983,
- 14. " CASK-40 Group Coupled Neutron and Gamma-Ray Cross Section Data," Radia-tion Shielding Information Center, DLC-23E.-
- 15. Dosimeter File ENDF/B5 Tape 531, distributed March 1984, National Neutron Data Center, Brookhaven National Laboratory, Upton, Long Island,-NY.
- 16. American Society of Testing Materials, Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Fer Atom (DPA), E693-79 (Reapproved 1985).
- 17. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Reaulatory Guide 1.99. Revision 2, May 1988.
- 18. A. L. Lowe, Jr. and J. W. Pegram, Correlations for Predicting the Effects of Neutron Radiation on Linde 50 Submerged-Arc Welds, BAW-1803. Revision 1, B&W Nuclear Service Company, Lynchburg, Virginia, March 1991.
- 19. H. W. Behnke, et al., Methods of Compliance With Fracture Toughness and Operational Requirements _ of Appendix G to 10CFR50, BAW-10046A. Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, June 1986,
- 20. J. D._ Aadland, Babcock & Wilcox Owner's Group 177-Fuel- Assembly Reactor Vessel and Surveillance Program Materials Information, BAW-1820, Babcock
& Wil::ox, Lynchburg, Virginia, December 1984.
- 21. K. E. Moore and A. S, - Heller, B&W 177-FA Reactor Vessel Beltline Weld -
Chemistry Study, BAW-1799, Babcock & Wilcox, -Lynchburg, Virginia, July 1983.
F-3 SW#EWiHLv
- 22. N. L. Snider and L. A. Hassler, B&WOG flux Perturbation Experiment at ORNL, Heasured and Calculatt.1 Dosimeter Results, DAW 1816, Dabcock &
Wilcox, Lynchburg, Virginia September 1985.
3 F-4 S W ff 4 &"Si h r
- _ _ _ . _ _ _ _ _ - . . . - . _ - - , - - - .