ML20094J710
| ML20094J710 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/31/1984 |
| From: | Aadland J, Collins L, Lowe A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20094J702 | List: |
| References | |
| BAW-1836, NUDOCS 8408140433 | |
| Download: ML20094J710 (93) | |
Text
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L BAW-1836_
July 1984
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ANALYSES OF CAPSULE AN1-A ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT 1
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-- Reactor Vessel Material Surveillance Program --
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ANALYSES OF CAPSULE AN1-A ARKANSAS POWER & LIGHT COMPANY
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ARKANSAS NUCLEAR ONE, UNIT 1
-- Reactor Vessel Material Surveillance Program --
k by A. L. Lowe, Jr., PE J. D. Aadland
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L. L. Collins L. A. Hassler W. A. Pavinich l
l W. L. Redd
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l B&W Contract No. 582-7168 BABC0CK & WILC0X Utility Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505
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a McDermott company
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SUW4ARY This report describes the results of the examination of the third capsule of the Arkansas Power & Light Company's Arkansas Nuclear One, Unit i reac-tor vessel material surveillance program.
The. capsule was removed and examined after accumulating a fluence of 1.03 x 1019 n/cm2 (E > 1 MeV),
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which is' equivalent to approximately 30 effective full power years' opera-tion of the reactor vessel.
The objective of the program is to monitor the
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effects of neutron irradiation on the tensile and fracture toughness proper-ties of the reactor vessel materials by the testing and evaluation of ten-sion and Charpy impact specimens.
The capsule received an average fast fluence of 1.03 x 1019 n/cm2 (E > 1.0 MeV) and the predicted fast fluence for the reactor vessel T/4 location at the end of the fifth cycle is.1.2 x 1018 n/cm2 (E > 1 MeV).
Based on the
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calculated fast flux at the vessel wall and an 80% load factor, the pro-jected fast fluence that the Arkansas Nuclear One, Unit i reactor pressure vessel will receive in 40 calendar years' operation is 1.10 x 1019 n/cm2 (E
> 1 MeV).
The results of the tensile tests indicated that the materials exhibited nor-mal behavior relative to neutron fluence exposure.
The Charpy impact data results exhibited the characteristic behavior of shift to higher tempera-
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ture for both the 30 and 50 ft-lb transition temperatures as a result of neutron fluence damage and a decrease in upper shelf energy.
These results
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demonstrated that the current techniques used for predicting the change in both the increase in the RTNDT and the decrease in upper shelf properties due to irradiation are conservative.
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The reconnended operating period was extended to 21 effective full power years as a result of this third capsule evaluation.
These new operating limitations are in accordance with the requirements of Appendix G of 10 CFR 50.
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CONTENTS Page k
1.
INTRODUCTION..........................
1-1 2.
BACKGROUND...........................
2-1 3.
SURVEILLANCE PROGRAM DESCRIPTION................
3-1
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4-1 4.
PREIRRADIATION TESTS.
4-1 4.1.
Tensile Tests......................
4-1 4.2.
Imp ac t Tes ts.......................
5-1 5.
POSTIRRADIATION TESTS.....................
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5.1.
Thermal Monitors.....................
5-1 l'
5.2.
Te nsil e Tes t Re s ul ts...................
5-1 5.3.
Charpy V-Notch Impact Test Results............
5-2 6-1 6.
NEUTRON 00SIMETRY 6.1.
Background........................
6-1 6.2.
Yessel Fluence......................
6-2 6-3 6.3.
Capsule Fluence 6-4 6.4.
Fluence Uncertainties f
7.
DISCUSSION OF CAPSULE RESULTS 7-1 7.1.
Preirradiation Property Data...............
7-1 7.2.
Irradiated Property Data.................
7-1 7.2.1.
Te ns i l e P rope rti e s................
7-1 7.2.2.
Impact Properties 7-2 8.
DETERMINATION OF RCPB PRESSURE-TEMPERATURE LIMITS 8-1 9.
SUPNARY OF RESUI.TS 9-1
- 10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE 10-1 11-1
- 11. CERTIFICATION.........................
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CONTENTS (Cont'd)
Page APPENDIXES A.
Reactor Veesel Surveillance Program -- Background Data and Information................... A-1 1
B-1 j
B.
Preirradiation Tensile Data C-1 C.
Preirradiation Charpy Impact Data 0-1 D.
Fluence Analysis Procedures E.
Capsule Dosimetry Data.................. E-1 F. References........................ F-1 List of Tables Table 3-3 3-1.
Specimens in Surveillance Capsule AN1-A 3-2.
Chemistry and Heat Treatment of Surveillance Materials..... 3-4 3-3.
Chemistry and Heat Treatment of Correlation Material --
1 Heat A-1195-1, A533 Grade B, Class 1.............. 3-5 l
5-1.
Tensile Properties of Capsule AN1-A Base Metal and Weld
. Metal Irradiated to 1.03E19 n/cmz 5-3 5-2.
Charpy Impact Data From Capsule AN1-A, HAZ Metal,
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Longitudinal Orientation, Irradiated to 1.03E19 n/cm2 5-3 5-3.
Charpy Impact Data From Capsule AN1-A, Base Metal,2 Transverse Orientation, Irradiated to 1.03E19 n/cm 5-4 l
I 5-4.
Charpy Impact Data From Capsule AN1-A Base Metal, Longitudinal Orientation, Irradiated to 1.03E19 n/cm2 5-4 5-5.
Charpy Impact Data From Capsule AN1-A, Weld Metal-WF 193 Irradiated to 1.03E19 n/cmz 5-5 5-6.
Charpy Impact Data From Capsule AN1-A, Correlation Monitor Material, Heat No. A-1195-1, Longitudinal Orientation, Irradiated to 1.03E19 n/cmz 5-5 6-1.
Surveillance Capsule Detectors................. 6-4 6-5 6-2.
Reactor Yessel Flux......................
C-6 6-3.
Reactor Vessel Fluence Gradient 6-4.
Surveillance Capsule Fluence.................. 6-7 6-8 6-5.
Estimated Fluence Uncertainty 7-1.
Comparison of Tensile Test Results.............-.. 7-5 7-2.
Summary of ANO-19eactor Vessel Surveillance Capsules Tensile Test Results...................... 7-6 7-3.
Observed Vs Predicted Changes in Irradiated Charpy 7-7 Impact Properties 7-4.
Summary of ANO-1 Reactor Vessel Surveillance Capsules Charpy Impact Test Resul ts................... 7-8 8-1.
Data for Preparation of Pressure-Temperature Limit Curves for ANO-1 -- Applicable Through 21 EFPY......... 8-4 A-1.
Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of A-3 Surveillance Program Materials -- ANO-1 l
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Tables (Cont'd)
Table Page
.A-2.
Test Specimens for Determining Material Baseline Properties..
A-4 A-5 A-3.
Specimens in UTPer Surveillance Capsules A-5
.A-4.-
Specimens in Lower Surveillance Capsules B-1., Tensile Properties of Unirradiated Shell Plate Materi al, ' Heat No. C 51,14-1..................
B-2 B-2.
Tensile Properties of Unirradiated Weld Metal, WF 193.....
B-2 C-1.,Charpy Impact Data From Unirradiated Base ?iaterial, Longitudinal Orientation, Heat No. C 511a-1..........
C-2
.C-2. ;Charpy Impact Data From Unirradiated Base Material, Transverse Orientation, Heat No. C 5114-1...........
C-3 s
C-3.
Charpy Impact Data From Unirradiated Base Metal, HAZ, Longitudinal Orientattun, Heat No. C 5114-1..........
C-4 0-4.
Charpy Impact Data From Unirradiated held Metal, WF 193....
C-5 D-8 D-1.
Spectrum-Averaged Cross Seccions D-2.
Extrapolation of Reactor Vessel Fluence............
D-9 E-1 Detector Composition and Shielding...............
E-2 E-3 E-2.
Capsule AN1-A Dosimeter Activities -..............
E-3 Dosimeter Activation Cross Sections, b/ atom..........
E-3 List of Figures Figure 1
Reactor Vessel Cross Section Showing Surveillance 3-1.
3-6 Capsule Locations at ANO-1 3-2.
Reactor Vessel Cross Section Showing Location of ANO-1
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Capsule in Davis-Besse Unit 1 Reactor.............
3-7 3-3.
Loading Diagram for Test Specimens in AN1-A..........
3-8 5-1.
Charpy Impact Data for Shell Plate Material, HAZ, 5-6 Longitudinal Orientation 5-2.
Charpy' Impact Data for Shell Plate Material, Transverse 1
Orientation..........................
5-6 5-3.
Cnarpy impact Data for Shell Plate Material, Longitudinal Orien utico..........................
5-8 5-9 5-4.
Charpy Impcct Data for Weld Metal WF 193 5-5.
Charpy Impact Data for 0orrelation Monitor Material, HSST Pl ate 02, Heat No. A-1195-1...................
5-10 6-1.
Reactor. Vessel Radial ' 1ux/ Fluence Gradient..........
6-9 6-2 Azimuthal Fleence Gradient (E > 1.0 MeV) at the Inside Surface of the ANO-1 Reactor Vessel..............
6-10 i
Predicted Fast Neutren Fluence at */arious Locations Through 8-1.
Reactor Vessel Wall for First 21 EFPY.............
8-5 5
042. Reactor Vessel Pressure-Tempertture Limit Curves for Normal
. 3 Operation -- Heatup, Applicable for First 21 EFPY.......
8-6 8-3.
Reactor Vessei Presure-Temperatare Limit Curve for Normal
. ' Operation -- Cooldown, App 1 tcable for First 21 EFPY......
8-7 n
8-4. ; Reactor Vessel Pressure-Temperature Limit Curve for Inservice
- J Leak and Hydrostatic Tests, Applicable for First 21 EFPY 8-8 g
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Figures (Cont'd) 1
- Figure Page A-1.
Location and Identification of Materials Used in Fabrication of ANO-1, Reactor Pressure Yessel.........
A-6 J
A-2.
Location of Longitudinal Welds in Upper and Lower
- Shel l Cou rses.........................
A-7 C-1.
Charpy -Impact Data From Unirradiated Base Metal, 1
Longitudinal Orientation C-6 J
C-2.
Charpy Impact Data for Unirradiated Base Metal, Transverse Orientation C-7 C-3.
Charpy Impact Data From Unf rradiated Base Metal.
HAZ, Longitudinal Orientation.................
C-8
'C-4.
Charpy Impact Data From Unirradiated Weld Metal........
C-9 4
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1, D4Tr.000CTION This report describes the results of the examination of the third capsule of the Arkansas Power & Light, Company's Arkansas Nuclear One, Unit 1 (ANO-1) reactor vessel material surveillance program (RVSP).
The capsule was removed and examined after the equivalent of six years of operation of 19 the ANO-1 reactor vessel.
The capsule experienced a fluence of 1.03 x 10 n/cm2 (E > 1 MeV), which is the equivalent of approximately 30 effective full power years' (EFPY) operation of the reactor vessel.
The first cap-sule from this program was removed and examined af ter the first year of operation; the results are reported in BAW-1440.1 The second capsule was removed and examined after irradiation in Toledo Ediscn Company's Davis-Besse Unit 1 as part.of the integrated reactor vessel materials sur-veillance program; the results are reported in BAW-1698.2
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The objective of the program is to monitor the effects of neutron irradia-tion on the tensile and impact properties of reactor pressure vessel mate-fl rials under actual operating conditions.
The surveillance program for ANO-1 was. designed and furnished by Babcock & Wilcox (B5W) as described in
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.BAW-10006A.3 The program was planned to monitor the effects of neutron ir-radiation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel.
The surveillance program for ANO-1 was designed in accordance with E185-66 l
and thus is not in compliance with Appendixes G and H to 10 CFR 50 since the requirements did not exist at the time the program was designed.
Be-cause of this difference, additional tests and evaluations were required to ensure meeting the requirements of - 10 CFR 50, Appendixes G and H.
The recommendations for the future operation of ANO-1 included in this report do comply with these requirements.
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2.
BACKGROUND
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The ability of the reactor pressure vessel to resist fracture is the pri-mary factor ;in ensuring the safety of the primary system in light water-cooled reactors.. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation.
The general effects of fast neutron irradiation on the mechanical proper-
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ties of such low-alloy ferritic steels as SA533, Grade B, and SA508, C1. 2, used in the fabrication of the ANO-1 reactor vessel, are well characterized and documented in the literature.
The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation.
In reactor. pressure vessel steels, the most serious mechani-cal-property change is the increase in temperature for the transition from f
brittle to ductile fracture accompanied by a reduction in the upper shelf icpact toughness.
Appendix G to 10 CFR 50, " Fracture Toughness Requirements," specifies mini-mum: fracture toughness requirements for the ferritic materials of the pres-sure-retaining components of the reactor coolant pressure boundary (RCPB)
- of water-cooled power reactors, and provides specific guidelines for deter-mining the pressure-temperature limitations on operation of the RCPB.
The toughness and operational requirements are specified to provide adequate safety margins during any condition of nomal operation, including antici-pated 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 10 CFR 50 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 op-eration on the effective date.
Appendix H to 10. CFR 50, " Reactor Vessel Materials Surveillance Program Re-quirements,". defines the material surveillance program required to monitor
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2-1 Babcock &WNeos a McDermott company
changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water-cooled reactors resulting from expo-sure to neutron irradiation and the thennal environment.
Fracture tough-ness test data are obtained from material specimens withdrawn periodically from the reactor vessel.
These data will pennit determination of the con-ditions. 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 i
is described in Appendix G to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components." This method utilizes fracture mechanics concepts and the ref-l erence nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition temperature [according to the I
American Society for Testing and Materials (ASTM) Standard E208] or the tem-perature that is 60F below that at which the material exhibits 50 ft-lbs
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and 35 mils lateral expansion (MLE).
The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve), which appears in Appendix G of ASME Section III.
The KIR curve is a lower bound of dynamic, static, and crack arrest fracture toughness re-suits obtained from several heats of pressure vessel steel.
When a given material is indexed to the KIR curve, allowable stress intensity factors x
can be obtained for this material as a function of temperature.
All owable operating limits can then be determined using these allowable stress in-tensity factors.
The RTNDT and, in turn, the operating limits of a nuclear power plant, can be adjusted to account for the effects of radiation on the properties of the reactor vessel material s.
The radiation embrittlement and the resul-tant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which a surveillance capsule con-taining 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 RTNDT to adjust it for radiation embrittlement.
This adjusted RTNDT is used to index the material to the KIR curve which, in turn, is used to set operating limits for the nuclear power plant.
These new limits
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take into account the effects of irradiation on the reactor vessel materi-2-2 Babcock &WHcom a W1cDermott company
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3.
SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for ANO-1 comprises six surveillance capsules de-signed to allow the owner to monitor the effects of the neutron and thermal environment on the materials of the reactor pressure vessel core region.
The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned inside the vessel between the thermal shield and the vessel wall at the locations shown in Figure 3-1.
The six cap-(
sules, placed two in each holder tube, were positioned near the peak axial and azimuthal neutron flux.
BAW-10006A includes a full description of the capsule locations and design.3 After the capsules were removed from ANO-1 and included in the integrated RYSP, they were scheduled for irradiation in the Davis-Besse Unit i reactor as described in BAW-1543.4 During this peri-od of irradiation, capsule AN1-A was irradiated in site YZ as shown in Figure 3-2.
Capsule AN1-A was removed from Davis-Besse Unit 1 after cycle 3 and an ac-cumulated fluence of approximately 1.0 x 1019 n/cm2 (E > 1 MeV).
This cap-sule contained Charpy V-notch impact and tensile specimens fabricated of SA533, Grade B Class 1 base metal, weld metal, Charpy V-notch impact speci-mens of heat-affected zone (HAZ) material and the correlation monitor ma-teri al.*
The specimens contained in the capsule are described in Table 3-1 and shown in Figure 3-3, and the chemistry and heat treatment of the surveillance material in capsule AN1-A are described in Table 3-2.
The chemistry and heat treatment of the correlation material are prescnted in Table 3-3.
All test specimens were machined from the 1/4-thickness (1/4T) location of the plates.
Charpy V-notch and tensile specimens from the vessel material were oriented with their longitudinal axes parallel to the principal roll-ing direction of the plate; Charpy V-notch specimens were also oriented
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- The correlation monitor material is from the Heavy Section Steel Tech-nology (HSST) Program.
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transverse to the principal rolling direction.
Weld metal specimens were oriented to measure the properties of the weld metal; the tensile specimens were oriented with their longitudinal axis parallel to the weld groove (all weld metal tensile specimens), while Charpy specimens were oriented with their notches oriented in the center of the weld in the longitudinal direc-tion.
The HAZ Charpy V-notch specirrens were oriented with their longitudin-al axes. transverse to the weld groove, with the notch in the base metal, and directed parallel to the axis of the weld.
Capsule AN1-A contained the following types of dosimeter wires:
Dosimeter wire Shielding U-Al alloy Cd-Ag alloy Np-Al alloy Cd-Ag alloy Nickel Cd-Ag alloy 0.66 wt % Co-Al alloy Cd-Ag alloy
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0.66 wt % Co-Al alloy None Fe None Thermal monitors of low-melting eutectic alloys and pure metal were in-cluded in the capsule.
The monitors and their melting points are as fol-lows:
Melting Alloy point, F 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 610 Lead 621 J
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Specimens in Surveillance Capsule AN1-A Number of test specimens Material description Tension CVN(a) impact Weld metal, WF 193 4
8 HAZ Heat No. C5114-1, longitudinal 0
8 Base material plate Heat No. C5114-1, longitudinal 4
8 transverse 0
4 Correlation, HSST plate 02 0
8 Total per capsule 0
36 (a)CVN denotes Charpy V-notch.
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Table 3-2.
Chemistry and Heat Treatment of Surveillance Materials Chemical Analysis Heat No.(a) Weld metal)
Weld metal)
C5114-1 WF 193(b WF 193lc Element C
0.21 0.065 0.09 Mn 1.32 1.50 1.49 P
0.010 0.016 0.016 S
0.016 0.008 0.016 Si 0.20 0.42 0.51 Ni 0.52 0.59 0.59 Cr 0.06 Mo 0.57 0.36 0.39 Cu 0.15 0.19 0.28 Heat Treatment
- Time, Heat No.
Temp, F h
Cooling C5114-1(a) 1550-1600 4.5 Brine-quenched 1200-1225 5.0 Brine-quenched 1100-1150 40.0 Furnace-cooled WF.193 1100-1150 29 Furnace-cooled (a)Per Certified Materials Test Report.
(b)Per Weld Procedure Qualification Test Record.
(c)Per Licensing Document BAW-1500.
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Table 3-3.
Chemistry and Heat Treatment of Correlation
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Element wt %
C 0.23 Mn 1.39
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P 0.013
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S 0.013 Si 0.21 Ni 0.64 Mo 0.50 Cu 0.17 Heat Treatment (b) 1.
Normalized at 1675F 75F.
2.
1600F i 75F for 4 h/ water-quenched.
3.
1225F i 25F for 4 h/ furnace-cooled.
4.
1125F 25F for 40 h/ furnace-cooled.
(a)0RNL-4463.
(b)Per is!entification card; heat treatment 3
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Figure 3-1.
Reactor Vessel Cross Section Showing Surveillance Capsule Locations at AN0-1 Surveillance Captule Holder Tube -- Capsules AN1-C, AN1-D t
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Reactor Vessel Cross Section Showing Location of r
ANO-1 Capsule in Davis-Besse Unit 1 Reactor L
X 1rradiation Site, h
Capsule AN1-A Capsule Holder -
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TOP Note: Arrow indicates position of orientation key.
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Loading Diagram for Test 7)
Specimens in AN1-A L
Charpy Specimens CGL CCA CCA C;A CCA CCA GG CCA CCA G
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. 414-604 421 919 037 9 32 71ll 720 026 92 033
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CCA 6M 443 702
,0;0 935 040 940 617 918 969 713 CCA,
0;A CCA Ch m
CCA G
CCA CCA GCL OGA 70>
439 008
'466 405 904 611 015 401
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4.
PREIRRADIATION TESTS L
.Unf rradiated mater ai l was evaluated for two purposes:
(1) to establish a baseline of data to which irradiated properties data could be referenced, and (2) to determine those materials properties to ' the extent practical from available material, as required for compliance with Appendixes G and H to 10 CFR 50.
4.1. -Tensile Tests Tensile specimens were fabricated from the reactor vessel shell course plate and weld metal.- The small-size 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 20,000-lb load capacity universal test machine at a crosshead speed of 0.005 inch per minute.
A linear voltage differential transformer (LVDT) -type clamp-end screw-on extensometer was used to determine the 0.2%
yield point.
Test conditions were in. accordance with the applicable. re-quirements of ASTM A370-72.
For each material type and/or condition, six specimens in groups of three were tested at both room-temperature and 570F.
The tension-compression load cell used had a certified accuracy of better
-than 0.5% of full scale (10,000 lb).
All test data for the preirradiation tensile specimens are given in Appendix B.
4.2.
. Impact Tests Charpy V-notch impact tests were conducted in accordance with the require-
.ments of ASTM Standard Methods A370-72 and E23-72 on an impact tester certi-fled to meet Watertown standards.
Test specimens were of the Charpy V-notch type, which were 0.394 inch square and 2.165 inches long.
Before testing,. specimens were temperature-conditioned in a combination re-sistance-heated / carbon dioxide-cooled chamber, designed to cover the temper-ature' range from. -85 to +550F.
The specimen support arm, which was linked to Lthe pneumatic transfer mechanism, is instrumented with a contacting 4-1
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k thermocouple allowing instantaneous specimen temperature determinations.
Specimens were transferred from the conditioning chamber to the test frame anvil and precisely pretest-positioned with a fully automated, remotely con-trolled apparatus.
Transfer times were less than 3 seconds and repeat with-in 0.1 second.
Once the specimen was positioned, the electronic interlock
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opened, and the pendulum was released fran its preset drop height.
After failing the specimen, the hamer pendulum was slowed on its return stroke and raised back to its start position.
Impact test data for the unirradiated baseline reference materials are pre-sented in Appendix C.
Tables C-1 through C-4 contain the basis data that are plotted in Figures C-1 through C-4.
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5.
POSTIRRADIATION TESTS 5.1.
Thermal Monitors I
Capsule AN1-A contained three temperature monitor holder tubes, each con-taining five. fusible alloy wires with melting points ranging from 558 to 621F.. All the thermal monitors at 558, 580, and 588F had melted, while those at the 610F location showed partial melting or slumping; the monitor l
at ~ the 621F location melted in all three holder tubes.
It is therefore assumed that the ' 610 and 621F monitors were placed in the wrong locations in the holder tubes.
From these observations, it was concluded that the capsule had been exposed to a peak temperature in the range of 610 to 621F during the reactor operating period.
These peak temperatures are attrib-uted to operating transients that are of short durations and are judged to
.have an insignificant effect on. irradiation damage.
Short duration operat-I ing transients cause the use of thermal monitor wires to be of limited val-ue in. determining the maximum steady-state operating temperature of the sur-veillance capsules; however, it.is judged that the maximum steady-state operating temperature of specimens in the capsule was held within 125F of tne 1/4T vessel thic'kness location temperature ' of 577F.*
It is concluded f that the capsule design temperature may have been exceeded during operating
~
transients but not for times and temperatures that would preclude the use of the capsule data.
4 5.2.
Tensile Test Results The results.of the postirradiation tensile tests are presented in Table 5-1..
Tests were performed on specimens at room temperature, 580F, and two intermediate temperatures.. The specimens were tested on a 55,000-1b load capacity universal test machine at a crosshead speed of 0.005 inch per p
l-l GI A. Wickstrom, " Thermal Analysis of Redesigned 177-FA SSHT," Babcock &
Wilcox Document 32-3728-00, May.28, 1976.
L 1 Bh&WHeoar a AncDermott company
minute up to yielding, and 0.050 inch per minute thereafter.
A four-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 ASTM A370-77.
The tension-compression load cell used had a certified accuracy of better than 0.5% of full scale (25,000 lb).
In gen-eral, the ultimate strength and yield strength of the material increased with a corresponding decrease in ductility; both effects were the result of neutron radiation damage.
The type of behavior observed and the degree 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.
5.3.
Charpy V-Notch Impact Test Results The test results f rom the irradiated Charpy V-notch specimens of the reac-tor vessel beltline material and the correlation monitor material are pre-sented in Tables S-2 through 5-6 and Figures 5-1 through 5-5.
The tests were conducted in accordance with the requirements of ASTM Standard Methods A370-77 and E23-82 on an impact tester certified to meet Watertown stan-dards.
Test specimens were of the Charpy V-notch type, which were 0.394 inch square and 2.165 inches long.
Prior to te. sting, specimens were tem-perature-controlled in liquid immersion baths capable of covering the tem-perature range from -85 to +550F.
Specimens were removed from tiie baths and positioned in the test frame anvil with tongs specifically designed for the purpose.
The pendulum was released manually, allowing the specimens to be broken within 5 seconds from their removal from the temperature baths.
The data show that the material exhibited a sensitivity to irradiation with-in the values predicted from its chemical composition and the fluence to which it was exposed.
l 5-2 Babcock &WHcom a McDermott company l
)
L Table 5-1.
Tensile Properties of Capsule AN1-A Base Metal and
(.
Weld Metal Irradiated to 1.03E19 n/cmz ( E > 1 MeV) ed Strength, psi Elongation, %
Specimen Test temp, 9,
No.
F Yield Ultimate Unif.
Total Base Metal, Longitudinal GG 713 76 78,700 99,700 11 24 55 GG 706 300 71,900 93,500 10 22 64 GG 716 400 70,100 92,200 10 22 64 GG 707 580 69,700 94,800 11 23 61 Weld Metal GG 107 76 84,100 99,200 12 24 54 GG 117 285 77,400 92,800 10 20 48 GG 104 400 76,100 90,800 10 20 53 GG 103 580 73,200 91,600 10 19 47 Table 5-2.
Charpy Impact Data From Capsule AN1-A, HAZ Metal, Longitudinal Orientation, Irradiated to 1.03E19 n/cm2 (E > 1 MeV)
Absorbed Lateral Shear Specimen Test temp,
- energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
GG 421
-63 19.0 16 5
GG 414
-2 30.0 17 20 GG 436 32 38.0 26 30 GG 443 75 44.0 40 50 GG 401 153 66.0 47 100 GG 432 222 72.0 60 100 GG 405 378 74.0 52 100 GG 439 454 69.0 73 100 5-3 Babcock & WHeem a McDermott company
Table 5-3.
Charpy Impact Data From Capsule AN1-A, Base Metal, Tranjiverse Orientatico, Irradiated to 1.03E19 n/cm' (E > 1 MeV)
Absorbed lateral Shear Specimen-Test temp,
- energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
GG 617.
0 10.0 8
0 GG 601 75 33.0 29 10 GG 619 153 46.0 38 10 f
GG 604 222 83.0 77 100 GG 627 441 78.0 67 100 Table 5-4.
Charpy Impact Data From Capsule AN1-A, Base Metal, Longitudinal Orientation, Irradiated to 1.03E19 n/cm2 (E > 1 MeV)
Absorbed Lateral Shear Specimen Test temp,
- energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
GG 713
-2 22.0 17 0
f GG 723 32 35.0 30 5
GG 709 75 52.0 37 30 GG 710 115 57.0 47 45 GG /20 153 99.0 72 100 GG 702 268 108.0 80 100 GG'701 378 104.0 91 100 1
5-4 mg g gg, a McDermott company
WF 193 Irradiated to 1.03E19 n/cmgi-A, Weld Metal Charpy Impact Data From Capsule A Table 5-5.
r (E > 1 MeV) i Absorbed Lateral Shear Specimen Test temp,
- energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
GG 026 75 12.0 12 0
GG 040 115 22.0 22 10 GG 010 153 28.0 24 20 GG 008 222 43.0 41 100 GG 033 268 42.0 53 100 GG 003 326 47.0 47 100 GG 015 378 43.0 42 100 GG 037 454 45.0 47 100 Table 5-6.
Charpy Impact Data From Capsule AN1-A, Correlation Monitor Material, Heat No. A-1195-1, Longitudinal Orientation, Irradiated to 1.03E19 n/cm2 (E > 1 MeV)
Absorbed Lateral Shear Specimen Test temp,
- energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
GG 940 32 10.0 9
C GG 939 75 23.0 20 5
GG 942~
156 32.0 30 15 GG 918 192 72.0 56 80 GG 932 222 75.0 55 95 GG 904 302 87.0 78 100 GG 935 378 96.0 75 100 GG 949 454 94.0 73 100 5-5 mh&WHcom a McDermott company l
1 Figure 5-1.
Charpy Impact Data.for Shell Plate Material, HAZ, Longitudinal Orientation l
100 i
i g
i g
i i
i i
>< 75
!j, 25 -
I I
I I
I I
I I
I I
0
. 08 g
g g
g i
i g
g 5
i,os-E 2
O.04_
5
.re
!!.02
.5 I
I I
I I
I I
I I
I I
0 200 i
i g
i g
i i
g DATA SulW9Y 1150- T,,7 Tey (35 MLE) 60F
_Tey (50 FT-u ) 90F s
g Tcv (30 FT-u) 2F C -USE (avo) 74 ft-1bs y
3 140 RT,,,
$120-
$100-
.5%-
E o
~
454F t
60-bren At SA533. Gr B
~
1 OnteNTATION HAZ -- 1.ono-FLutact 1.03E19 n/crn2.
20 -
Hr.AT flo.
C 5114-1 I
I I
I I
I I
I 0
-C0
-40 0
90 80 120 160 200 240 280 320 360 400 Test Ttmetmatung, F 5-6 shAMhz a McDermott company
Figure 5-2.
Charpy Impact Data for Shell Plate Material, r
L Transverse Orientation 100 g
i i
g i
i g
i
" 15 -
=
E32 5
5 25 I
I I
I I
0
.08 3
g g
g g
i g
8.06 G
2 M
w.04-5 5
f.02-I I
I I
I I
I I
I I
I 0
200 g
g g
g g
i i
g
-- DATA SumARY 180- T,o7 127F T,(35 as) e
,Tcy (50 FT-ts) 144F g
T, (30 FT-ts) 83F c
C -USE (avo) 83 ft-1bs-y
~
g140 RT g7 3 120 8
3100-3 5
a 80-
=441F W
t 60
~
n r nig SA533, Gr B Onnentatto, Transverse ~~
20 Fmsect 1.03E19 n/cm2 -
HEAT No. _C 5114-1 I
'l l-1 I
I I
I i
i,
-40 0
40 80 120. 160 200 240 280 Jz0 360 400
-W Test it#wenatunt, F 5-7 IE=h,a,ar &WHcom a McDermott company
Figure 5-3.
Charpy Impact Data for Shell Plate Material, Longitudinal Orientation 100 i
g i
g i
i i
i
- 75
.3
!yw m
5 32s I
I I
I I
I I
0
,91
. 08 g
g g
g g
i i
i E
i.Os-i k
w.04 5
g.02
- ' 1
.5 I
I I
I I
I I
I I
I I
0 200 g
g g
g g
g g
g DATA SUNHRY 180- T,,,
73F Tcy (35 nu)
_Tey (50 FT-u) 78F gg Tcy (30 FT-u) 17F C -USE (Ave) 108 ft-1bs y
3140 RT,,,
C i120-E
- l. 100-5 b 80-9 k g
- ~
mienias, SA533, Gr B
~
OnIENTArt0N Longitudinal ~
Ftuence 1.03E19 n/cm2 -
20 -
14 :47No. C 5114-1 I
I I
I I
I I
I 0-80
-40 0
40 80 120 _ 160 200 240 280 320 360 400 Tsar Tenrenatune, F 5-8 m ean,.,ar & W U c o m a ucoermott company
9 Figure 5-4.
Charpy Impact Data for Weld Metal WF 193 r
2 100 g
i g
g g
g g
g g
g g
[ 75 Eg2 ________________ _____________
(
E 3 25-0
. 08 g
g g
g g
g g
i g
g l
5 s.Os 2
2 M
",04 5
4 g.02 5
I I
I I
I I
I I
I I
I O
100 i
i g
g g
g i
g g
g DATA SumARY 90- T,,,
T, (35 MLE) 194F c
Tcy (50 rT-La)
N.D.
80 Tcy (30 FT-La) 156F C -USE (Ave) 45 ft-lbs y
70 gy i E0-
%Oyg - - - - - _ _ - - - _ _ _ _ _ - - _ _ _ _ _ _ _ - - - - - - - -
o454F 5
sn-U<
t 30 ----------_
MArtatAL Weld Metal On:Entation 10-Fluence 1.03E19 n/cm2 -
l HEAT No. WF 193 I
I I
I I
I I
I 1
0-80
-40 0
40 80 120 160 200 240 280 320 360 400 Test TenrenArupt, F 5-9 Belmock&MIcom a McDermott company
n Figure 5-5.
Charpy Impact Data for Correlation Monitor Material, s
HSST Plate 02, Heat No. A-1195-1 J
100 i
g g
g i
g g
g g
i a 75 J
?
y2 2
3 25 I
I I
I I
I I
I 0
. 08 g
g g
g g
g g
g e
5 s.Os--
1 2
0.04-5 E
g.02 3
I I
I I
I I
I I
I I
I 0
200 g
g g
g g
i g
g DATA SUPNRY 180- T,,7 140F Tey (35 nLa)
Tcv (50 rt-ts) 140F Tcy (30 FT-ts) 94F C -USE (ava) 95 ft-lbs y
g140 RT,,7 O
$120-
=
5 3100-454F o
g
=
5 80-k s0 _
~
MArsa:AL HSST Plate 01 "
g.
20 FLugner 1.03F19 n/cm2 -
HEATIb. A-1195-1 I
I I
I I
I I
i 0J
-40 0
40 80 120 160 200 240 280 320 360 400
-G TestItnataarunt,I EEb8 5-10 a McDermott company
L 6.
NEUTRON DOSIMETRY
(
6.1.
Background
Fluence analysis as a part of the reactor vessel surveillance program has three objectives:
(1) determination of maximum fluence at the pressure ves-sel as a function of reactor operation, (2) prediction of pressure vessel fluence in the future, and (3) determination of the test specimen fluence l
within the surveillance capsule.
Vessel fluence data are used to evaluate changes in the reference transition temperature and upper shelf energy levels, and to establish a correlation between changes in material proper-ties and fluence exposure.
Fluence data are obtained either directly or indirectly from flux distributions calculated with a computer model of the reactor.
The accuracy of calculated fast flux is enhanced by the use of a nonnalization factor that utilizes measured activity data obtained from capsule dosimeters.
A significant aspect of the surveillance program is to provide a correla-tion between the neutron fluence above 1 MeV and the radiation-induced prop-erty changes noted in the surveillance specimens. To permit such a correla-
. tion, activation detectors with reaction thresholds in the energy range of interest were placed in each surveillance capsule.
The significant proper-
' ties of the detectors are given in Tables 6-1 and E-1.
Because of a long half-life (30 years) and effective threshold energies of 137 s production from fission reac-0.5 and 1.1 MeV, the measurements of C
tions in 237Np and 2380 are more directly applicable to analytical determin-ations of. the fast neutron fluence (E > 1 MeV) for multiple fuel cycles f
than are other dosimeter reactions.
Other dosimeter reactions are useful as corroborating data for shorter ' time intervals. and/or higher energy fluxes.
Short-lived isotope activities are representative of reactor con-ditions. only over the latter portion of the irradiation period- (fuel cycle), whereas reactions with a threshold energy greater than 2 or 3 MeV do not record a significant part of the total fast flux.
6-1 hah a McDermott company
v.
i The energy-dependent neutron flux is not directly available for activation detectors because the dosimeters register only the integrated effect of neu-tron flux on the target material as a function of both irradiation time and neutron energy.
To obtain an accurate estimate of the average neutron flux
)
incident upon the detector, several parameters must be known:
the operat-ing history of the reactor, the energy response of the given detector, and the neutron spectrum at the detector location.
Of these parameters, the definition of the neutron spectrum is the most difficult to obtain.
Essen-tially two means are available to obtain the spectrum:
iterative unfolding of experimental dosimeter data and/or analytical methods. -Because of a lack of sufficient threshold reaction detectors satisfying both the thresh-old energy and half-life requirements of a surveillance program, calculated spectra have been used.
]
Neutron transport calculations in two-dimensional geometry are used to ob-
)
tain energy-dependent flux distributions throughout the reactor system.
Re-actor conditions are selected to be representative of an average over the irradiation time period.
Geometric detail is selected to explicitly repre-sent the reactor system to provide the flux distributions in the reactor vessel.
The capsule flux distributions were obtained in previous analyses which mdeled an explicit surveillance capsule as,embly.
The capsule energy-dependent flux distribution enabled the generation of calculated dosimeter activities.
A comparison of the measured to calculated activi-ties provided the normalization factor applied to the calculated vessel flux and fluence.
Due to consistency in both the normalization factors and calculated average reaction cross sections from the previous analyses, the explicit calculation of capsule flux was not used in this analysis; rather an average capsule flux was obtained directly from the measured data using the average cross sections.
In addition, the average normalization factor from previous analyses was applied to the vessel flux to normalize it to measured values.
Use of this method has provided results that are consis-tent with previously reported results.
A more detailed description of this revised calculaticnal procedure is contained in Appendix D.
1 6.2.
Vessel Fluence The maximum fluence (E > 1.0 MeV) in the pressure vessel during ANO-1 cycle 5 was detennined to be 3.96 (+17) n/cm2 based on a maximum neutron flux of 6-2 a McDermott comparty
2 n/cm -s (Tables 6-2 and 6-3).
The location of maximum fluence 1.03 (+10) is a point at the cladding / vessel interface at an elevat. ion of about 103 cm dbove the lower active fuel boundary and at an azimuthal (peripheral) loca-
[
tion of abo'ut 8 degrees from a major axis (across flats diameter).
Fluence data have been extrapolated to 32 EFPY of operation based on the premise that excore flux is proportional to fast flux escaping the core (Appendix D).
Core escape flux values are available from fuel management analyses of the current and future fuel cycles that have been designed.
For ANO-1, fu-ture cycles 6 and 7 are available, and cycle 7 has been assumed to be the equilibrium cycle used for extrapolation to 32 EFPY.
Relative fluence as a function of radial location in the reactor vessel is shown in Figure 6-1.
Reactor vessel lead factors (clad interface flux /in-f vessel flux) for the T/4, T/2, 3T/4 locations are 1.8, 3.6, and 7.7, respec-tively.
Relative fluence as a function of azimuthal angle for cycle 5 is shown in Figure 6-2.
A peak occurs at about 8 degrees which roughly corre-sponds to a radius through a corner of the core and the position of the sur-veillance holder tube.
The ratio of fast flux at the neximum and minimum locations is about 1.5.
Previous analyses have shown this peak to be about 10 degrees with all fresh fuel in the flats assemblies. The 2-degree devia-tion in this case is due to a fresh fuel asssembly on a major axis (H-15) of the core flats and once-burned assemblies in the other two locations in the flats region.
6.3.
Capsule Fluence Cumulative fast fluence at the center of the surveillance capsule was calcu-lated to be 1.03 (+19) n/cm2, 7% of which occurred in ANO-1 and 93% in Davis-Besse (Table 6-4).
These data represent an average location in the capsule.
In ANO-1, capsule AN1-A was located in an upper holder tube posi-tion at 11 degrees off the major axis and about 211 cm from the core center for 345 effective full power days (EFPD).
It was then inserted in Davis-Besse in an upper holder tube position at 11 degrees off axis and about 202 cm from the core center for an additional 943 EFPD.
During the latter ir-radiation period, the capsule was estimated to have been rotated approxi-mately 90 degrees counterclockwise relative to its original design orienta-tion (keyway facing the reactor core).
6-3 Babcock & Wilcox a McDermott company
6.4. -Fluence Uncertainties Uncertainties 'were estimated for the fluence values reported herein.
These data, Table 6-5, were based on comparisons to benchmark experiments when available, estimated and measured variations in input data, and engineering
)
judgment. Because of the complexity of the fluence calculations, no compre-hensive uncertainty limits exist for these results.
The values in Table 6-5 represent a best estimate based on experience with these analyses.
Table 6-1.
Surveillance Capsule Detectors
]
Effective lower energy limit, Isotope
]
Detector reaction MeV half-life 54 e(n,p)S4 n(a) 2.5 312.5 days F
M 58Ni(n.p)58 o(a) 2.3 70.85 days C
238 (n,f)137 s(a) 1.1 30.03 years U
C 237Np(n,f)137 s(a) 0.5 30.03 years C
238 (n,f)106 u 1.1 369 days U
R 237Np(n,f)106 u 0.5 369 days
)
R 238 (n,f)103 u 1.1 39.43 days U
R 237Np(n,f)103 u 0.5 39.43 days R
238 (n,f)l44 e 1.1 284.4 days U
C 237Np(n,f)l44 e 0.5 284.4 days C
238 (n,f)95 r 1.1 64.4 days U
Z 237Np(n,f)95 r 0.5 64.4 days Z
(a)Due to the revised procedure for capsule flux de-termination, only these reactions were measured.
6-4 mgg%
a McDermott company
Table 6-2.
Reactor Vessel Flux L
[
Fast flux, n/cm -s (E > 1 MeV)
"('
2 O
V Inside surface Inside surface (max location)
T/4 3T/4 (max location)
Cycle 1A(a) 1.39(+10) 7.9(+9) 1.8(+9) 2.68(+10)
(345 EFPD) f-Cycle 18-4(a) 1.46(+10)(b) 8.4(+9)(b) 1,9(+g)(b) 3.11(+10)(b)
(1047 EFPD)
Cycle 5 1.03(+10) 5.9(+9) 1.3(+9) 2.15(+10)
(447 EFPD) l (a) Capsule AN1-B analysis, B&W-1698.
(b)Due to 110 mag contamination of the fission dosimeter data, the measured activity of the detectors for the AN1-B capsule were determined to be 10% too large. Correcting for this contamination effect has reduced the normalization factor from1.07 to 0.95.
Thus, the flux values listed for cycles 18.4 above are about 11% lower than previously reported in BAW-1698.
6-5 BabcodsAWHcom a McDermott company
.\\
J J
Table 6-3.
Reactor Vessel Fluence Gradient Fast fluence, n/cm2 (E > 1.0 MeV)
Cumulative Inside surface 1
irradiation time (max location)
T/4 3T/4 J
End of cycle 1A(a) 4.14(+17) 2.4(+17) 5.4(+16)
(345 EFPD)
]
End of cycle 4(a) 1.73(+18)(b) 9.8(+17)(b) 2.2(+17)(b)
(1392 EFPD)
End of cycle 5 2.13(+18) 1.2(+18) 2.8(+17)
(1839 EFPD) l-8 EFPY 3.38(+18) 1.9(+18) 4.7(+17)
]
15 EFPY(c) 5.61(+18) 3.2(+18) 7.3(+17) 21 EFPY(d) 7.5(+18) 4.2(+18) 9.8(+17) i l
32 EFPY 1.10(+19) 6.3(+18) 1.4(+18) 9 (a) Capsule AN1-B analysis, BAW-1698.
l (b) Reduced from values reported in 8AW-1698, see (b) on l
Table 6-2 of this report.
(c)15 EFPY values are needed for T/2 and the outer surface; these values are 1.6(+18) and 2.9(+17), respectively.
l (d)21 EFPY values are needed for T/2 and the outer surface; these values are 2.1(+18) and 3.9(+17), respectively.
l l
l 6-6 a MCDermott company
.. _ ~, -
. ~..
~
Table 6-4.
Surveillance Capsule Fluence f
L Cumulative Flux, n/cm2-s
- Fluencg, fluenge, (E > 1 MeV) n/cm' n/cmZ Method 1
[
ANO-1, cycle 1A(a) 2.44(+10) 7.27(+17) 7.27(+17) t (34" EFPD)
Dav.. Besse Unit 1, cycle 1(a) 9.79(+10) 3.16 (+18) 3.89(+18)
(374 EFPD)
Davis-Besse Unit 1, cycle 2(b) 9.67(+10) 2.61(+18) 6.50(+18)
(296 EFPD)
Davis-Besse Unit 1, cycle 3 1.54(+11) 3.62(+18) 1.01(+19)
(273 EFPD)
Method 2 ANO-1, cycle 1A 2.44(+10) 7.27(+17) 7.27(+17)
(345 EFPD) j Davis-Besse Unit 1, cycle 1 9.79(+10) 3.16(+18) 3.89(+18)
(374 EFPD)
Davis-Besse Unit 1, cycles 2 and 3 1.30(+11) 6.38(+18) 1.03(+19)
(569 EFPD)
Method 3 ANO-1, cycle 1A 2.44(+10) 7.27(+17) 7.27(+17)
(345 EFPD)
Davis-Besse Unit 1, cycles 1, 2, and 3 1.20(+11) 9.75(+18) 1.05(+19)
(943 EFPD)
AVERAGE CUMULATIVE FLUENCE FROM 3 METHODS = 1.03(+19)
(a) Capsule AN1-B analysis, BAW-1698.
(0) Capsule RSI-D analysis, BAW-1792.6
[-
m SMhz j
6-7 a McDermott company
=____ - --___ _.
]
Table 6-5.
Estimated Fluence Uncertainty Estimated Calculated fluence uncertainty Basis of estimate In the capsule 18%
Activity measurements, cross sec-tions, fission yields, satura-tion factor, and deviation from 1
average fluence value.
J In the reactor vessel at 124%
Activity measurements, cross sec-1 maximum location for tions, fission yields, satura-
]
cycle 5 of ANO-1 tion factors, axial factor, cap-sule location, radial / azimuthal extrapolation, and sigma of
]
average normalization factor.
In the reactor vessel at i26%
Items in vessel fluence above
)
the maximum location for plus time / flux extrapolation to 3
end-of-life extrapolation 32 EFPY.
l l
I l
l l
6-8 Babcock &WHcom a McDermott company l
V N
Figure 6-1.
Reactor Vessel Radial Flux / Fluence Gradient
-s 0
10 F.
L T/4 L. F. = 1.7 5
{
8 T/2 O
L.F=3.57 f
VESSEL T
l. D.
X l-ST/4
=
L.F.=7.65 I
{
E G
S Ia O
10-I E
VESSEL 0.D.
L. F.=19. I 4 x
C I
I I
I I
I I
216 220 224 228-232 236 2 40 Distance From Core Center, em N A Micos 6-9 a McDermott company
1 J
Figure 6-2.
Azimuthal Flux Gradient (E > 1.0 MeV) at the Inside Surface of the ANO-1 Reactor Vessel 1.10 j
]
1.05 a
1.00 E
1 e
?
0.95 1
i E
il n
% 0.90 C
c
- 0. 85 3
.?_
% 0.80 l
T I
=
5 l
C
, 0.75 E
4 w
0.70 l
0.65 e
i i
i i
a i
l 0.0
- 5. 0 10.0 15.0-20.0 25.0 30.0 35.0 40.0 45.0 j
Azimuthal Angle, degrees l
1 i
i l
l 6-10 N&Wilcom a M.Dermott cc.i.pany
s 7.
DISCUSSION OF CAPSULE RESULTS 7.1.
Preirradiation Property Data A review of the unirradiated properties of the reactor vessel core beltline region indicated no significant deviation from expected properties except in the case of the upper shelf properties of the weld metal.
Based on the prelcted end-of-service peak neutron fluence value at the 1/4T vessel wall f
location and the copper content of the weld metal, it was predicted that the end-of-service Charpy upper shelf energies (USE) will be below 50 l
ft-lb.
The weld metal selected for inclusion in the surveillance program was selected in accordance with the criteria in effect at the time the pro-gram was designed for ANO-1.
The applicable selection criterion was based on the unirradiated properties only and before it was known that the weld metals are more sensitive to radiation damage than the base metal.
7.2.
Irradiated Property Data 7.2.1.
Tensile Properties 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 range of anticipated change. There ap-pears to be some strengthening, as indicated by increases in ultimate and yield strength and similar decreases in ductility properties.
All changes observed in the base metal are such as to be considered within acceptable limits.
The change in the room teaiperature properties of the weld metal is greater than those observed for the base metal, indicating a greater sensi-tivity 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 period in service life.
7-1 NEh a McDermott company
)
A comparison of the tensile data from the first capsule (capsule AN1-E, fluence = 7.3E17 n/cm ) with the corresponding data from the capsule re-
)
2 ported in this report is shown in Table 7-2.
The currently reported data
]
came from specimens which experienced a fluence approximately 14 times greater than those from the first capsule.
The general behavior of the tensile properties as a function of neutron irradiation is an increase in q
both ultimate and yield strengths and a decrease in ductility as measured by both total elongation and reduction in area.
The most significant ob-q servation from these data is that the weld metal exhibited much greater J
sensitivity to neutron irradiation than did the base metal.
n 7.2.2.
Impact Properties The behavior of the Charpy V-notch impact data is more significant to the calculation of the reactor system's operating limitations.
Table 7-3 com-
[
pares the observed changes in irradiated Charpy impact properties with the
(
predicted changes.
The shift in transition temperature was calculated for the surveillance ma-terials using the predictive models in both Regulatory Guide (RG) 1.99, Re-vision 1, and the currently proposed revision to RG 1.99.
Neither version specifies at which toughness level the prediction should be applied.
The latest revision of 10 CFR 50, Appendix G specifies that the 30 ft-lb level of impact toughness should be used in detennining the transition tempera-ture shift, although past revisions specified the 50 ft-lb level.
l l
The observed 30 ft-lb transition temperature shifts for all the surveil-lance materials are not in good agreement with the predicted shift values, with all the observed shifts being significantly less than the correspond-ing predicted shifts.
The large degree of inaccuracy shown is due largely to the conservatisms built into each predictive model.
The observed 50 ft-lb transition temperature shifts for all the surveil-lance materials are not in good agreement with the predicted shift values, although the observed shifts were closer to the predicted values than the 30 ft-lb transition temperature shifts. The observed 35 MLE transition tem-perature shif ts show no pattern from one material to another.
7-2 Batscock &WHcom a McDermott company
Enk k
The observed decrease in Charpy USE with irradiatien agreed with the L
predicted decrease fairly well for the longitudinal-oriented base metal, the weld metal, and the correlation material.
The predicted decrease was
{
very conservative in the case of the transverse-oriented base metal, and was non-conservative in the case of the HAZ material.
The disagreement for the transverse base metal may be explained by a lack of transverse-oriented specimen data at the time the predictive model was developed, while HAZ material data often behaves in an erratic manner, probably due to the difficulty in machining the specimen notch in the same HAZ location from specimen to specimen.
The agreement in USE decreases for the other three
[_
material s indicates that the predictive model is representative, not conservative.
k.
A ' comparison of the Charpy impact data from the first two capsules (AN1-E and AN1-B) with the corresponding data from the capsule reported herein is
(-
'shown in Table 7-4.
The currently reported capsule experienced a fluence that is 14 times greater than capsule AN1-E, and 2-1/2 times greater than
{
capsule AN1-B.
The base metal and HAZ metal exhibited shifts at the 30 and 50 ft-lb levels for the current capsule that were larger than those of AN1-E. ~ The 30 ft-lb levels had to be estimated for capsule AN1-E, but it appears that the 50 ft-lb shift increases were greater than the 30 ft-1b shift increases.
For the current capsule, the weld metal exhibited a shift
(
at the 30 ft-lb level which was larger than that of AN1-E.
Interestingly, the observed shift was greater than the predicted shift for capsule AN1-E, ll whereas the reverse was true for the current capsule.
The predictive model is now conservative for this weld.
h For the current capsule, the correlation me.terial exhibited a shift at the 50 ft-lb level which was larger than that of the first capsule and moderate-
'ly-larger than that of the second capsule. At 30 ft-lbs, the shift for the current capsule was slightly less than that of the second capsule.
This
. indicates a saturation effect in this material at these fluence levels.
The longitudinal base metal, the HAZ material, and the weld metal all showed decreases in USE from capsule AN1-E.
The transverse base metal did not decrease from AN1-E, possibly indicating a saturation effect, but more likely showing inaccuracy due to the limited number of specimens available in both cases.
The-correlation ' material did not show a significant de-crease - from capsule AN1-B, and.this probably shows a saturation effect.
7-3 EhW8com a McDermott comp:1y 1-
l The lack of saturation in the other materials may be due to the large
]
increase in fluence from capsule AN1-E to the current capsule; saturation may have been attained at an intermediate fluence.
Results from other surveillance capsules indicate that the RTNDT estimating curves have greater inaccuracies than originally thought.
These inaccura-cies are a function of a number of parameters related to the basic data
)
available at the time the estimating curves were established.
Some of these include inaccurate fluence values, poor chemical composition values, and variations in data interpretation.
The change in the regulations re-quiring the shift measurement to be based on the 30 ft-lb value may help to' minimize errors that result from using the 30 ft-lb data base to predict the shift behavior at 50 ft-lbs.
The design curves for predicting the shift will be modified as more data be-come available; until that time, the design curves for predicting the RTNDT
]
shift as given in RG 1.99 are considered adequate for ;; dicting the RTNDT shift of those materials for which - data are not available and will continue to be used to establish the pressure-temperature operational limitations for the irradiated portions of the reactor vessel until the time that new prediction curves are developed and approved.
)
)
l l
J l
J 7-4 Babcock &WIlcox a McDermott company
0 Table 7-1.
Comparison of Tensile Test Results Elevated temp Room temp test test L
Unirr Irrad Unirr Irrad Base Metal -- C 5114-1
(
Fluence, 1018 n/cm2 (E >l MeV) 0 10.3 0
10.3 Ult tensile strength, ksi 94.8 99.7 91.d 94.8 0.2% yield strength, ksi 72.0 78.7 64.5 69.7 Uniform elongation, %
10 11 12 11
(
Total elongation, %
27 24 24 23 Reduction of area, %
68 55 64 61 Weld Metal -- WF 133 Fluence, 1018 n/cm2 (E >l Mey) 0 10.3 0
10.3 Ult tensile strength, ksi 84.6 99.2 81.4 91.6 0.2% yield strength, ksi 67.6 84.1 60.4 73.2 f
Uniform elongation, %
12 12 11 10 Total elongation, %
28 24 22 19 Reduction of area, %
64 54 52 47 1
l 7-5
, w o,, moi,co,,,,ny
Table 7-2.
Sunenary of ANO-1 Reactor Vessel Surveillance Capsules Tensile Test Results Ductili ty,1 Strength, ksi Material 1018 n/cm2 temp, F Ultimate
% A(a) Yield
% A(a) elongation
% a(a) Reduction%A(a)
F1 ence Test Total of area 68 72.0 27 Base metal 0
72 94.8 64 64.5 24 570 91.8 0.73 70 96.9
+2.2 75.6
+5.0 26
-3.7 67
-1.5 570 95.3
+3.8 68.4
+6.0 22
-8.3 57
-10.9 10.3 76 99.7
+5.2 78.7
+9.3 24
-11.1 55
-19.1 580 94.8
+3.3 69.7
+S.1 23
-4.2 61
-4.7 64 28 67.6 Weld metal 0
72 84.6 52 60.4 22 570 81.4 0.73 70 91.3
+7.9 76.2
+12.7 24
-14.3 59
-7.8
?
570 89.1
+9.5 68.8
+13.9 19
-13.6 47
-9.6 10.3 75 99.2
+17.3 84.1
+24.4 24
-14.3 54
-15.6 580 91.6
+12.5 73.2
+21.2 19
-13.6 47
-9.6 (a) Change relative to unirradiated.
2 5 D
'I
m Table 7-3.
Observed Ys Predicted Changes in Irradiated Charpy Impact Properties Transition Temperature Increase Observed shift in property, F Predicted )
Predicted )
shift, F(b shift, Fta Material 30 ft-lb 50 ft-lb 35 MLE Base material (C5114-1)
Transverse 66 70 63 122 140 Longitudinal 48 70 60 122 140 HAZ 45 90 18 122 140 Weld metal 151 ND(c) 129 284 243 Correlation material 46 66 85 157 163 i
(HSST plate 02)
Charpy USE Decrease
{
Decrease, ft-lb Material Observed Predicted (d)
Base material (C5114-1)
Transverse 11 23 Longitudinal 24 32 HAZ 41 28 Weld metal 28 31 Correlation material 35 34 (HSST plate 02) o E
Pj (a)Per RG 1.99, Revision 1.
5D (b)Per draft revision of RG 1.99, based on shift plus 20 margin.
f$
(c)ND denotes not determinable.
E (d)Per RG 1.99, Revision 1; no change of USE model in draft revision.
.e -
Table 7-4.
Sununary of ANO-1 Reactor Vessel Surveillance' Capsules Charpy Impact. Test Results Decrease
- USE, Transition temperature increase,- F 30 ft-lb 50 ft-lb 10 2 n/cm observed observed.
Predicted (a) Observed Predicted (a)
F1gence, 2 Material Base material. (C 5114-1)
Longitudinal 0.73 30(b) 19 32 14 24 10.30 48 70 122 24 32 Transverse 0.73 10(b) 22
-32 14 19 10.30 66 70 122 13 23 l
HAI 0.73 13(b) 42 32 33 21 10.30 45 90 122 41 28 76 15 15
?
Weld metal 0.73 115 137(c) 10.30 151' ND 284 28 31 Correlation material 0.73
.12(b) 40-42
- 13 18 (HSST plate 02) 4.28 50 57 101 34 28 10.30 46 66 157 35 34 (a)Per RG 1.99, Revisfon 1.
- (b) Estimated.
(c)ND denotes not deteminable.
i 5
F 25D 1s ag i
M
L 8.
DETERMINATION OF REACTOR COOLANT PRESSURE B0UNDARY PRESSURE-TEMPERATURE LIMITS
(
The pressure-temperature Ifmits of the reactor coolant pressure boundary (RCPB) of ANO-1 are established in accordance with the requirements of 10 CFR 50, Appendix G.
The inethods and criteria employed to establish operat-ing pressure and temperature limi ts are described in topical report BAW-10046A.6 The objective of these limits is to prevent nonductile fail-h ure during any normal-operating condition, including anticipated operation occurrences and system hydrostatic tests.
The following loading conditions
(-
are of interest:
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 10 CFR 50, Appendix G.
The closure head region, the reactor vessel outlet noz-zi e, and the beltline region have been identified as the only regions of the reactor vessel, and consequently of the RCPB, that regulate the pres-sure-temperature limits.
Since the closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first several service periods.
The reactor vessel outlet nozzle also affects the pressure-temperature limit curves of the first several service periods 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 RTNDT of the beltline region materials will be high enough that the belt-line region of the reactor vessel will start to control the pressure-tem-perature 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 8-1 Babcock &WHcom a McDermott company
)
l limits imposed by the closure head region, the outlet nozzle, and the belt-line region.
The maximum allowable pressure is taken to be the lowest of three calculated pressures.
The limit curves for ANO-1 are based on the predicted values of the ad-justed reference temperatures of all the beltline region materials at the end of the twenty-first EFPY; this time was selected because it is within the fluence value for the third capsule that confirmed the material be-havior to be conservatively predicted by RG 1.99, Revision 1.
Also, it is estimated that a fourth surveillance capsule will be withdrawn at the end of the refueling cycle when the estimated fluence corresponds to approxi-mately 32 EFPY, which will confim material behavior to reactor vessel end of life (E0L).
The time difference between the withdrawal of the third and fourth surveillance capsules will be adequate for re-establishing the op-erating pressure and temperature limits for the final period of reactor vessel operation.
)
The unirradiated impact properties were detemined for the surveillance beltlir;e region materials in accordance with 10 CFR 50, Appendixes G and H.
For the other beltline region and RCPB materials for which the measured properties are not_ available, the unirradiated impact properties and resid-ual elements, as originally established for the beltline region materials, are listed in Table A-1.
The adjusted reference temperatures are calcu-lated by adding the predicted radiation-induced ARTNDT and the unirradiated RTNDT.
The predicted ARTNDT is calculated using the respective neutron fluence and copper and phosphorus contents. Figure 8-1 illustrates the cal-culated peak neutron fluence at several locations through the reactor ves-sel beltline region wall as a function of exposure time.
The supporting information for Figure 8-1 is described in section 6.
The neutron fluence values of Figure 8-1 are the predicted fluences that have been demonstrated (section 6) to be conservative. ' The design curves of RG 1.99* were used to predict the radiation-induced ARTNDT values as a function of the material's copper and phosphorus content and neutron fluence.
- Revision 1, April 1977.
8-2 maswa,ar agyue,og a McDermott company
The neutron fluences and adjusted RTNDT values of the beltline region ma-
[
terials at the end of the twenty-first EFPY are listed in Table 8-1.
The neutron fluences and adjusted RTNDT values are given for the 1/4T and 3/4T
(
vessel wall locations.
The assumed RTNDT of the closure head region and the outlet nozzle steel forgings is 60F, in accordance with BAW-10046A.
Figure 8-2 shows the reactor vessel's pressure-temperature limit curve for normal heatup.
This figure also shows the core criticality limits as re-quired by 10 CFR 50, Appendix G.
Figures 8-3 and 8-4 show the vessel's
(
pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively.
All pressure-tempera-
[
ture limit curves are applicable up to the twenty-first EFPY.
Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits of the pressure-temperature limit curves.
The ac-ceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve.
The reactor is not per-mitted to go critical until the pressure-temperature combinations are to the right of the criticality limit curve.
To establish the pressure-tem-perature limits for protection against nonductile failure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be adjusted by the pres-sure differential between the point of system pressure measurement and the pressure on the reactor vessel controlling the limit curves.
This adjust-ment is necessary because the reactor vessel is the most limiting component of the RCPB.
8-3 Babcock &WHcom a McDermott company
Table 8-1.
Cata for Preparatfort cf Pressure-Temperature Limit Curves for ANO-1 -- Applicable Through 21 EFPY wi+ent totation
,,,,,,,,,,,c,,,
,,,,,g,,,,,,,,,,
ch,,g,t,y end of 21 ETPY ^ RT t end of ' Adjusted Rig). at Core Location mi41ane fran Weld Comer rhosporus (E > 1 Mev). n/an? 21 a,b). F end of 21 El?Y, F 4terial im Beltline to wid nejor axis, 1/4T thirr concert,
- cordent, wat m.
Type
___ region location CL, cm degrtes location RTmT. F 5
At 1/4T At 3/4T At 1/4T At 3/4T At 1/4T At 3/4T
(+60)(C) 0.03 0.009 3.2E18 7.4l17 25 '
12 -
85 72 A1N 131 SMIB,C1.2 Nonle belt
(+40)(Cl 0.17 0.014 4.2E18 9K17 104 -
W 144 90 C 512D.2 SA533, Cr B tDper stell
+10 0.15 0.010 4.2E18 9.8E17 78 2
88 48 C 5114-2 SA513, Cr B Lpper shell C 512D.1 SA533, Gr B tmer stell
+10 0.17 0.014 4.2E18 9 K 17 104 50 114 60
+30 0.15 0.010 4.2E18 9K17 78
-2 1(B
' 68 C 5114-1 SA533, Gr B Laer shell W 182-1 Weld -
typer cire. sem (1005) 123 Yes
+19 0.24(d) 0.014(d) 3.2E18 7.4E17 130 0
149 82 W 112 Weld Middle circ, sean (2001) 63 Yes 0
0.31(d) 0.016(d) 4.2E18 9 K 17 201 97 2D1 57 l
SA 172 Weld Laer cire. sean (2005)
-249 Yes
(+20)(C) 0.25(d) 0.017Id)
- 2. 1 16 5.5E15 12 6
32 ai 27.6 Yes
(+20)(C) 0.29(d) 0.010(d) 3.1E18 7.3E17 145 70 -
165 90 W 18 Weld t$per long. (both 1001) 56 Yes
(+20)(C) 0.29(d) 0olo(d) 3.3E18 7.6E17 140 72 169 92
?
W 18 Weld Lower long. (both 1001) l l
(al er Replatory (hide 1.99, Revision 1, April 1977.
P (b)Replatory Cu161.99 not valid for shifts less than 5(F.
(c)Per BM-10046A, Revision 1, July 1977.
IS er BM-1799, July 1983.
P o
E 55D 8E i43 M
~
Figure 8-1.
Predicted Fast Neutron Fluence at Various Locations Through Reactor Vessel Wall for First 21 EFPY 8
l8 7.5xi0 nvt
)
x 7
A w
6 N
S**g c*
57 5
6e oo
\\D o
18 5*
4.1x10 nyt q,5 y
8 aO *"
E c
gg Lo
=
3 NeS
- 18 S
CD 2.lx10 nvt 5
8 2
vessel Tl2 Location U
g 9.8x1017nyt Vessel 3/4T Location I
17 Vessel Outside Surf ace
- 3. 9x 10 nvt i
i i
I I
i I
I I
I 0 O 1
2 3
4 5
6 7
8 9
10 ll 12 13 I4 15 16 17 18 19 20 21 EFPY a
2 5
=, D 3E ii M
Figure 8-2.
Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation --
Heatup, Applicable for First 21 EFPY 21000 Pressure.
Temp.,
g g
g
. Assumed RTMDT, F Point PSIG F
8eltline Region 1/4T 201 A
457 70
=
~
Beltlins Region 3/4T 97 8
,457 135
{
Closure Head Region 60 C
625 219 Applicable l
Outlet Nozzle 60 0
625 250 For Heatup E
730 253 Rates up to F
I410 340 100F/h 1600 n.
G 2250 390 M
1320 373 F
~
h gg i
2250 430 The acceptable pressure-temperature combinatloc.s are below 3
and to the right of the limit curve (s). The Limit curves m
g
- de not include the pressure differential between the point 8
S of the system pressure measurement and the pressure on the C ri ticality reactor vessel region controlling the limit curve, ner Limit uo gg
- do they include any additional margin of safety E t
,_ for possible instrument error, C
ee 600 j
400
-A li
~
200 0
I I
I I
I I
I I
I E
60 100 140 180 220 260 300 3 100 380 420 460
?
Reactor Vessel Coolant Temperature, F g
3D 354 DE m - _,
<-__ m m __,
<- r t
c-Figure 8-3.
Reactor Vessel Pressure-Temperature Limit Curve for Nomal Operation -- Cooldown, Applicable for First 21 EFPY 2400 The acceptable pressure-temperature combinations are below and to the right of the limit curve (s). The limit curves do not include p
2200
-the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve, nor do they include any additional margin of safety cm N00 3
for possible instrument error, cL 1800 Assumed RTNDT, F g
h Beltline Region 1/4T 201 g
_ Beltline Region 3/4T 97 I
_ Closure Head Region 60 p
Outlet Nozzle 60 co l200 O
- Pressure, Temp..
m Point PSIG F
i 1000 Applicable for g
A
-300 70
{
B 648 225 Cooldown Rates g
C 1390 305 Up to 100F/h D
2250 353 g
600 B
u l
E l
400 A
200
=
0 I
I I
I I
I I
E 60 100 140 180 220 260 300 340 380 0
Reactor Vessel Coolant Temperature, F ED if iA$
M
j Figure 8-4.
Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Hydrostatic Tests, Applicable for First 21 EFPY 2600 Assumed RTNDT, F G
2160 0 settline aegion 1/gT 20s Beltline Region 3/4T 97 2200 Closure Head Region 60 Outlet Nozzle 60 se*
Pressure Temp.,
Point PSIG T
g 1800 5
A 400 70 3 1600 a
521 iso F
C 425 375 Applicable For at 1160 0 D
625 225 Heatup and Cool-5 E
871 230 down Rates up to fl00F/h (s50F in m
e, j
1200 F
1563 320 G
2500 371 any 1/2-h o
Perioq)
"E 1000 E
The acceptable pressure-temperature S
200 combinations are below and to the right of the limit curve (s). The limit curves 6
O do not include the pressure differential r
600 C
D between the point of system pressure o
g measurement and the pressure on the B
m reactor vessel region controlling the 560 0 A
limit curve, nor do they include any l
additional margin of safety for possible instrument error.
5 200 O
0 I
I I
i i
i i
j oO 100 li40 180 220 260 300 3t60 380 tiOO fD Reactor Vessel Coolant Temperature, F
,3 l 3Pg x
M
L g
7 k
9.
SUPMARY OF RESULTS
(
The analysis of the reactor vessel material contained in the third surveil-k lance capsule, AN1-A, removed for evaluation as part of the ANO-1 reactor vessel surveillance program, led to the following conclusions:
1.
The capsule received an average fast fluence of 1.03 x 1019 n/cm2 (E >
1.0 MeV).. The predicted fast fluence for the ANO-1 reactor vessel 1/4T
[
location at the end of the fifth fuel cycle is 1.2 x 1018 n/cm2 (E > 1 MeV).
h
- 2.
The fast fluence of 1.03 x 1019 n/cm2 (E > 1 MeV) increased the RTNDT of the AN1-A capsule reactor vessel core region shell materials a maxi-mum of 151F.
3.
Based on the calculated fast flux at the vessel wall, an 80% load fac-tor, and the planned fuel management, the projected fast fluence that the ANO-1 reactor pressure vessel will receive in 40 calendar years' operation is 1.10 x 1019 n/cm2 (E > 1 MeV).
'4.
The increase-in the RTNDT for the base plate material was not in good agreement with that predicted by the currently used. design curves of
'ARTNDT versus fluence, but the prediction techniques are conservative.
5.
The increase in the RTNDT for the weld metal was not in good agreement with that predicted by the currently used design curves of ARTNDT ver-sus fluence, but the prediction techniques are conservative.
6.
The current techniques used for predicting the change in weld metal Charpy _ impact upper _ shelf properties due to irradiation are conserva-tive.
y 7.
The comparison of changes in transition temperature and USE with those of previous surveillance capsules indicates that a saturation effect with fluence may be - occurring.
Further data are required before this can be stated definitively.
L 9-1 Babeeck&WHsque a MCDermott Company
]
]
8.
The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.
9.
The capsule design operating temperature may have been exceeded during operating transients but not for times and temperatures that would pre-clude the use of the capsule data.
1 9
/
h e.Micou a McDermott company
- 10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE Based on the postirradiation test results of capsule AN1-A, the following
(
schedule is recommended for examination of the remaining capsules in the ANO-1 RVSP:
Evaluation schedule (a)
Est. vessel Est. capsule thg*he'2 Est. date Capsule fluence data ID 1019n/cmI Surface 1/4T available(b)
.AN1-C 1.4 0.32 0.18 1986 AN1-D(c) 1,4.
0.45 0.25 1990 AN1-FIC) 1.4 0.52 0.29 1992 (a)1n accordance with BAW-10006 and E 185-82 as f
modified by BAW-1543, Rev. 2 (b) Estimated date based on 0.8 plant operation fac-tor.
(c) Capsules designated as standbys which may not be evaluated.
10-1 Eh & WHeem a McDermott tompany
rif,-
y.
, l e.
j
(
Y r
- s...
[.^
e t4 l',
o e
h f,
( :-
l 4
g.
f-I; I
V 2
d s
5
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l i
t L
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l
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~ I f
b
- f, k.
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a l
i l
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1 0
{
- 11. CERTIFICATION The specimens were tested, and the data obtained from Arkansas Nuclear One, h_
Unit 1 surveillance capsule AN1-A were evaluated using accepted techniques and established standard methods and procedures in accordance with the re-
[
quirements of 10 CFR 50, Appendixes G and H.
2Y okt/VW E. E. E0we, J r.~, P.E.//
Date Project Technical Manager This report has been reviewed for technical content and accuracy.
y W~
% S$r Y
0 L. B. Gross, P.E.
U Date Materials and Chemical t
Engineering Services I
L 11-1 Babcock &WHcom a McDermott company
l L~
I s
(
l APPENDIX A Reactor Vessel Surveillance Program --
Background Data and Infomation A-1 Batscock&WIleOE a McDermott company
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 loca-tions of these materials within the reactor vessel are shown in Figures A-1 and A-2.
2.
Definition of Beltline Region The beltline region of ANO-1 was defined in accordance with the data given in BAW-10006A.3 3.
Capsule Identification The capsules used in the ANO-1 surveillance program are identified below by identification number, type, and location.
Capsule Cross Reference Data Number Type AN1-A 1
AN1-B II AN1-C I
AN1-D II f
AN1-E I
AN1-F II 4.
Specimens per Surveillance Capsule See Tables A-2, A-3, and A-4.
l A-2 Babcock &WHcom a McDermott company l
O
v,
r Il Table A-1.
Unirradiated Impact Properties and Residual Element Content Data of-Beltline Region Materials Used for Selection of Surveillance Program Materials -- ANO-1 Charpy data, CVN(a)
- Distance, Transverse core midplane 50 35 Chemistry,1(b; Material Beltline to weld Longitudinal RT oy,
- ident, Material region centerline, Drop wt, ft-Ib,
- MLE, USE, N
heat No.
type location cm TunT. F At 10F ft-1b F
F ft-Ib F
Cu P
S Ni 0.05 0.013 0.020 0.68 (74,83,611 AYN131 SA508, C1 2 Nozzle belt
}41,58,68f 67,63,59 0.17 0.014
- r. 013 0.55 C 5120-2 SA533, Gr B Upper shell 0.15 0.010 0.016 0.52 31,34,35 C 5114-2(CISA533, Gr B Upper shell
,C 5120-1 SA533, Gr B Lower shell 53,56,54 0.11 0.014 0.013 0.55 l
0.15 0.010 0.016 0.52 I
21,57,21 C 5114-1ldISA533, Gr B Lower shell 0.18 0.014 0.015 0.59 WF ;82-1 Weld Upper circus 123 36,33,44 0.22 0.024 0.006 0.58 Middle circum
-63 35,40,30 WF 112 Weld 0.29 0.017 0.012 0.47 I
p SA1788 Wel d Lower circum
-249 40,38,36 0.105 0.004 0.017 0.45 W
WF 18 Weld U er longitl.
45,46,38 0.19 0.016 0.008 0.59 29,30,37 l
WF 193(el Weld Surveillance weld
[
(a)CVN denotes Charpy V-notch.
(b)from mill and qualification test reports.
(c) Surveillance base metal A.
(dl urveillance base metal B.
S l
l (e) Surveillance weld metal.
e E
55o O
E M
z
f Table A-2.
Test Specimens for Determining Material Baseline Properties No. of test specimens Tensile Material description 70F S70F(a) Charpy Heat GG (C 5114-1)
Base metal Transverse direction 3
3 26 l
Longitudinal direction 3
3 27 HAZ Transverse direction 3
2 27 Longitudinal direction 3
3 26 Total 12 11 106 Heat HH (C 5114-2)
Base metal Transverse direction 3
3 27 Longitudinal direction 3
3 27 HAZ Transverse direction 3
3 27 Longitudinal direction 3
3 26 f
Total 12 12 107 Weld metal Longitudinal direction 3
3 27 l
7 A-4 Eh &WHcom a McDermott company
Table A-3.
Specimens in Upper Surveillance Capsules (Designations A, C, and E)
No. of test specimens e
Material description Tensile Charpy Weld metal, WF 193 4
8 HAZ A, heat No. C 5114-1, longitudinal 0
8
(
Base material -- Plate A, Heat No. C 5114-1: longitudinal 4
8 transverse 0
4 Correlation, HSST plate 02 0
8 Total, per capsule 8
36 Table A-4.
Specimens in Lower Surveillance Capsules (Designations B, D, and F)
(
No. of test specimens Material descriptior.
Tensile Charpy HAZ B, Heat No. C 5114-2, longitudinal 4
10 Base material -- Plate B, Heat No. C 5114-2: longitudinal 4
10 transverse 0
8 Correlation, HSST plate 02 0
8 Total per capsule 8
36 ESIDCOCit &MICOE a MtDermott company
Figure A-1.
Location and Identification of Materials Used in Fabrication of ANO-1, Reactor Pressure Vessel A
\\
l
(
/
W AYN 131 Nozzle Belt 4+ WF 182-1 C 5120-21 Upper Shell
+- WF 18 C 5114-2J I
I 4* WF 112 i
j WF-18 -*
C 5120-1 L wer Shell
/
C 5114-1 l
-SA1788 l
125W609VA1 Dutchman I
f A-6 Babcock & M Icom a McDermott company
Figure A-2.
Location of Longitudinal Welds in Upper and Lower Shell Courses L
W E
(
27.6*
f lz' X
f Upper Shell
/ 27.6*
E I
Y W
S'6 '
E Z
X J
Lower Shell 56' E
Y A-7 Babcock &Wilcox a McDermott company
M
(
l APPENDIX B Preirradiation Tensile Data B-1 Babcock & Wilcox a McDermott company l
Table B-1.
Tensile Properties of Unirradiated Shell Plate Material, Heat No. C 5114-1
)
Strength, psi Elongation, %
Specimen t
Red'n of No.
F Yield Ultimate Uniform Total area, %
Longitudinal GG-708 RT 71,430 94,500 10.3 27.9 68.8 GG-711 RT 72,870 94,930 8.3 26.4 68.5 GG-717 RT 71,590 94,120 11.3 26.7 67.3 Mean RT 71,%0 94,850 10.0 26.7 68.2 Std dev'n 790 320 1.5 1.1 0.8 GG-704 570 64,110 91,160 11.2 22.9 63.6 GG-705 570 64,220 92,070 12.8 25.7 63.6 GG-715 570 65,290 92,270 12.0 24.3 64.9 Mean 570 64,540 91,830 12.0 24.3 64.0 Std dev'n 650 590 0.8 1.4 0.8 Table B-2.
Tensile Properties cf Unirradiated f
Weld Metal, WF 193 Strength, psi Elongation, %
Specimen t
Red'n of No.
F Yield Ultimate Uniform Total area, %
GG-101 RT 67,800 84,850 13.3 28.6 64.7 GG-105 RT 67,790,
84,980 10.8 27.1 62.9 GG-114 RT 67,080 84,040 12.7 28.6 64.3
{
Mean RT 67,560 84,620 12.2 28.1 64.0 Std dev'n 410 510 1.23 0.9 1.0 GG-102 570 62,390 82,400 9.9 21.4 51.1 l
GG-109 570 59,110 80,850 11.4 22.1, 50.4 i
GG-112 570 59,640 80,870 1:. 2 22.1 54.7 Mean 570 60,380 81,370 10.8 21.9 52.1 Std dev'n 1,760 890 0.8 0.4 2.3
(
B-2 maan,,ar&WI8com a McDermott company imamm.mm. mum..
e APPENDIX C Preirradiation Charpy Impact Data C-1 Babcock &Wilcom a McDermott company
Table C-1.
Charpy Impact Data From Unirradiated Base Material, 1]itudinal Orientation, Heat No. C 5114-1 Test Absorbed Lateral Shear
)
Specimen
- temr, energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
GG-707 319 132 68 100 708 319 131 68 100 741 319 124 67 100 GG-703 199 136 66 100 3
735 199 130 60 100 717 198 124 64 100 GG-748 142 136 69 100 1
1 747 140 129 65 100
(
756 140 136 68 100
(
GG-744 86 108 64 80 1
J 732 85 108 61 82 706 85 104 57 75 GG-745 50 70 48 25 751 50 82 57 30 752 50 102 63 45 GG-749 36 68 48 18 750 35 73 49 15
[
755 35~ '
71 51 12 t
GG-754 20 21 16 2
753 20 30 21 3
746 19 53 38 5
GG-73'3 0.9 48 32 4
724 0.3 43 30 2
704 0.3 60 43 8
GG-736
-39 46 31 3
739
-39 37 24 2
722
-39 15 8.5 0
C-2 Babcock &WHcom a McDermott company
t
('
Table C-2.
Charpy Impact Data From Unirradiated Base Material, Transverse Orientation, Heat No. C 5114-1
(-
Test Absorbed Lateral Shear
(
Specimen
- temp, energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
f GG-603 320 88 62 100 605 319 84 60 100 618 322 94 64 100 f
GG-632 259 91 61 100 6 38 259 94 63 100-640 258 92 61 100 GG-610 200 86 58 100 616 200 96 58 100 624 200 101 60 100 GG-634 141 89 60 94 639 140 90 61 100 633 138 100 61 97 GG-631 111 66 46 75 f
636 111 53 43 45 630 110 80 55 70 GG-626 80 54 40 35 612 80 56 42 45 607 79 48 32 30 GG-637 61 60 42 30 635 61 38 29 20 629 60 64 44 45, GG-625 30 37 32 2
GG-614 30 32 24 1
GG-608
-38 28 16 1
620
-40 28 16 1
613
-40 14 5
0 C-3 Balueck&WW8com a McDermott company
]
Table C-3.
Charpy Impact Data From Unirradiated Base Metal, HAZ, Longitudinal Orientation, Heat No. C 5114-1 Test Absorbed Lateral Shear Specimen
- temp, energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
GG-423 321 83 62 100 434 320 96 58 100 4 38 320 83 57 100 GG-447 205 104 56 100 445 202 116 59 100 449 198 123 63 100 GG-453 142 107 63 100 450 140 95 54 100 448 140 107 58 100 GG-426 81 70 44 100 420 80 74 44 94 403 80 84 46 100 GG-452 56 79 43 100 446 55 77 42 90 455 54 73 45 92 GG-451 46 86 44 94 456 45 69 43 98 i
GG-412 31 52 33 35 417 30 66 35 85 472 30 46 33 65 GG-411
-10 43 21 65 444
-10 40 21 30 418
-10 53 31 18 GG-433
-38 32 18 18 419
-39 32 17 15 425
-39 44 25 12 C-4 R h &WWhos A McDermott tompany
k
(
L Table C-4.
Charpy Impact Data From Unirradiated
(
Weld Metal, WF 193 Test Absorbed Lateral Shear
(-
Specimen
- temp, energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
(
GG-011 320 77 59 100 l
044 320 73 61 100 035 319 70 60 100 GG-023 201 76 49 100 006 200 72 52 100 031 200 66 55 100 GG-055 141 67 55 96 031 140 67 46 97 049 139 11 53 100 GG-050 109 51 97 056 109 b7 59 82 045 108 64 46 85 GG-001 87 63 52 75 021 82 56 48 65 034 82 50
- 38 55 GG-053 60 35 32 20 047 60 28 25 25 046 60 49 38 55 GG-005 30 42 35 28 012 29 38 29 40 024 29 42 42 38 GG-052 1
16 14 8
054 1
33 25 10 048 0
37 31 6
GG-028
-38 22 15 4
013
-39 12 9
2 027
-39 20 12 3
C-5 Sabeosk&WUNeem a AecDermott company
Figure C-1.
Charpy Impact Data for Unirradiated Base Metal, Longitudinal Orientation J
W I
I I
I
- I I
i i
T I
75 I32 1
3 25 -
)
i i
. 08 g
g y
g g
g g
g g
i I
t L
i.06 I
e s
a,og b
f.02
{
(
.3 I
I I
I I
0 200 g
g g
i i
g g
DATA $LfRWtY 180-i
+10F no, T, (35 mLa) 13F c
,Tgy (50 rt-La) 8F gg 73 (30 rt La) 31F C,-USE (ave) 132 ft-ths p140 ny,,
+10F
~
2.
_?
$120 l
{
,k m
3m -
W t W 1
40 SA533. Gr B
. - h Matentat Catentation Lonoitudinal' g2 FLutace None 20 Hu t No. C 5114 1 I
I l
I I
I I
0 C0
-40 0
40 80 120, 160 200 240 780 320 360 400 Test Tenetaatune, F C-6 Babcock &WWIfcom a McDermott company
k
?
s Figure C.2.
Charpy Impact Data for Unirradiated Base Metal,
[
Transverse Orientation L
t@
i i
i i
1 3 -
i 7
i
" 75 -
J!
(
g2 m
25 -
I I
I I
I I
I I
0
. 08 g
g g
g g
g g
g g
g E
~
t s
.!. 06 s
z.
r 3
'd.04 -
5 a
g.02 3
I I
I l
I I
I I
I 0
I I
I I
I I
I I
DATA Su n utY 180-I
+10F nst -
T,(35 mLa) 64F g
, T,(50 at ts) 74F e
g T, (30 at-u) 17F e
C U$E (ave) 96 f t.lbs v
140 9
RT,,,
+30F
$120 Ia l@-
E 3m -
E e
60 -
t
" ~
~
Matentat SA533. Gr B
.g__
arMrnsvuw **
20 Fluence None Peat h.
C 5114 1 I
I I
I I
I I
I 0W
+40 0
40 80 120, llo 200 240 200 320 360 400 Test Temptsatune, F C-7 h &WI8ess a MtDermott tempany
___.__._____._.J
Figure C-3.
Charpy Impact Data for Unirradiated Base Metal, HAZ, Longitudinal Orientation
]
100 g
i i; -
i - i i
i i
i I
j j.
E 32s I
I I
I I
I O
l I
I I
I I
I I
I I
I E
5.06 r
]
2.,
.i
'I g.02 Et I
I I
l I
t i
t l
t t
0 I
I i
i i
i i
i DAT4 $UNWlY --
40F 180 7,,,
T, (35 mLa) 42F c
T, (50 et.La) 0F c
g
, T, (30 et-ts).43F c
C,-USE (Ave) 115 ft-lbs p14 ti,,,
OF 0
$120 g
2 100 _
2 3 80 2
9 e
60 s _
g Matsenac SA533. Gr B g,,
20 Ftuence Nana he no. C 51141 I
I I
I I
I 0.00 4
0 4
80 120, 160 200 24 290 320 360
- 40 0 Test Teneenarvet, F C-8 maimpggeou a M<Ottmott tompany
s Figure C-4.
Charpy Impact Data for Unirradiated Weld Metal k
i g i
i 100 g
g g
g e
(
_,s _
3, ________-
(
Ls 0
. 08 g
g l
l l
g g
i i
g i
5 8.0s y
I O.04 5
-__--_w
-~~~~~_--_-------~~
A e
g.02
=
t t
t t
t t
t t
t t
t O
100 i
g g
i g
DATA SUMMRY
-20F 90-T, T, (3s not) 65F c
T, (50 n-ts) 65F c
g
, T, (30 n-La)
EF c
C,-U$E (ave) 73 f t-lbs 10 gi 30F not
[
.e
$FO
,I 50 _________. _.___________________
E 3 40 30 _ _ _ _
r n _
lkrantg Weld metal Onitatations 10 pove,,c: None we h. WF 193 I
I I
I I
I I
I 0-80 40 0
80 120 160 200 240 200 320 360 400 Test Teseenatune, F C-9 m a gues, a A4(Dermott tompany
/
\\
l APPENDIX D Fluence Analysis Procedures D-1 Ba4H:OCk & WllCOM a McDermott compJny
1.
Analytical Method The procedure used in this analysis differs from previous procedures in two significant aspects:
the calculation of the capsule flux and the normaliza-
)
7 tion factor.
Previous analyses calculated the capsule flux using the 00T4 code with an explicit model of the capsule.
This flux was then used in equation D-1 to obtain a calculated activity for each dosimeter. A factor,
)
defined as the measured activity divided by the calculated activity, was then calculated and used to normalize the calculated capsule and vessel
]
flux to measured results.
Use of this procedure in a series of capsule analyses has resulted in a fairly consistent average normalization factor Np(n,f)l37 s reactions.
In addi-cf 0.95 from the 238 (n,f)137 s and 237 C
U 0
tion, these analyses have produced a consistent set of spectrum-averaged re-action cross sections as a function of capsule position, i.e.,11 or 26 de-grees azimuthally.
Based on the consistency of the normalization factor and the cross sections, a revised analysis procedure was developed.
Basic-ally, the procedure involves calculating the capsule flux from equation D-6 using the measured activities and the spectrum-averaged cross sections.
The vessel flux still is calculated with DOT 4, but the flux is normalized using the average normalization factor from previous analyses.
A more de-tailed description of this revised procedure is given below.
1.1.
Capsule Flur and Fluence Calculation Preivous capsule analyses employed the 00T4 code, including an explicit model of the capsule assembly, to calculate the neutron flux as a function of energy in the capsule.
The calculated fluxes were used in the following equation to obtain calculated activities for comparison with the measured data. The calculated activity for reaction product i, Dj, (uC1/gm) is:
- A tj )e-Aj (T-tj ) (0-1)
M t
N 1
Dj = 7n 3.7 x 10 fj en(E)4(E)
Fj(1 - e 4
j1 where N = Avogadro's number, An = atomic weight of target material n, fi = of ther weight fraction of target isotope in nth material or the fission yield of the desired isotope, o-2
.-. wn
{
a wonmois comp.ny
)
s i
L
.'n(E) = group-averaged cross sections for material n (listed in
(
Table E-3),
f(E) = group-averaged fluxes calculated by 00T4 analysis, Fj = fraction of full power during jth time interval, tj, tj = decay constant of the ith isotope, tj = interval of power history, T = sum of total irradiation time, i.e.,
residual time in
- reactor, and wait time between reactor shutdown and
- counting, tj = cumulative time from reactor startup to end of jth time
- period, The normalization factor that was applied to the calculated flux was ob-tained from the ratio of the measured to calcklated activities, i.e.,
Dj (measured)
(0-2)
C = Dj (calculates)
Equation D-1 can be abbreviated as:
Dj = KjPj o (E)((E)
(D-3) n I
where Kj contains the constant terms and Pt is the sunnation over j repre-senting the power (saturation) term.
Given the energy-dependent flux from DOT 4 and the energy-dependent cross sections, an average reaction cross section for E greater than 1 MeV can be found from
[o(E)$(E) n (D-4)
Un"
[4(E)
Then, equation D-3 becomes Dj = KjPj5n (E > 1 MeV)
(D-5) or, solving for the flux (D-6) 4(E > 1 MeV) =
KPG ggn D-3 Babeest 4WUNees a momommy
The revised method essentially uses this last equation to calculate the cap-sule flux from the measured activities 01 and the average cross section.
)
E, calculated from previous analyses (see Table D-1).
Note that 3n IS d**
n pendent upon the capsule azimuthal location since the calculated flux spec-
]
trum used to obtain 3 differs from the 11-and 26-degree locations due to n
differences in distances from the core.
]
The measured activity (at least for the 137 s activity) is representative C
of irradiations in all cycles previous to the measurement.
For the AN1-A
]
x capsule, this includes irradiation in cycle 1A of ANO-1 and cycles 1, 2, and 3 of Davis-Besse.
The irradiation history of the capsule is repre-sented by the Pj tem which accounts for production and decay of the prod-uct isotope by cycle as a function of time at a particular power level.
By accounting for production and decay during previous irradiation cycles and
)
the use of an appropriate Pj, the flux for the last cycle, or cycles, can be determined.
Previous capsule analyses have determined cycle-by-cycle
)
capsule activities and fast fluxes for cycle 1A of ANO-1 and cycles 1 and 2 of Davis Besse for syn. metric capsule locations.
Thus, the cycle 3 activity can be detemined from 3=Ay-[Aje"Ati
.(0-7)
)
A i
wh're A3. cycle 3 activity, e
AT = the measured activity at the end of cycle 3 (EOC 3),
tj = decay period from EOCi to EOC 3.
Using Pi for cycle 3, the cycle 3 capsule fluxes can be detemined.
By ap-propriate modification of At and Pj, capsule fluences for any combination of cycles can be determined.
For capsule AN1-A, capsule fluxes for cycle 3, 2 and 3, and 1, 2, and 3 were detemined (see Table 6 4).
Capsule fluence is obtained from the product of capsule flux and irradia-tion time period.
For this analysis, integrated capsule fluences were ob-tained from the sum of fluences for the Davis-Besse and the ANO-1 irradia-tions.
Three such values were obtained, i.e.,
from the three methods of calculating Davis-Besse fluxes (see Table 6-4).
The reported capsule fluence represented the average of these values.
This was done to reduce 0-4 BabeestsaWIIess a McDermott tempany
(
[
dependence on any particular cycle measurement to mitigate et f ects due to
(
errors associated with that cycle's measurement.
1.2.
Vessel Flux and Fluence lhe vessel flux calculation essentially uses the same method as previous analyses.
Energy-dependent neutron fluxes at the reactor vessel were de-termined by a discrete o.'dinates solution of the Boltzmann transport equa-tion with the two-dimensional code 00T4.
The ANO-1 reactor was modeled f
from the core out to the primary concrete shield in R-9 geometry (based on a plan view along the core midplane and one-eighth core symmetry in the
(
azimuthal dimension).
The reactor model contained the following regions:
core, liner, bypass coolant, core barrel, inlet coolant, thermal shield, inlet coolant (downcomer), pressure vessel, cavity, and concrete shield.
Input data to the code ficluded a PDQ calculated pin-by pin, time-averaged power distribution, CASK 23E 22-group microscopic neutron cross sections,8
(
S8 order of angular quadrature, and P3 expansion of tqo scattering cross section matrix.
Reactor conditions, i.e., power distribution, temperature and pressure, were averaged over the irradiation period.
Because of com-puter storage Ilmitations, it was necessary to use two geometric models to
{
cover the distance from core to primary shield.
A boundary source output from Model A (core to downcomer region) was used as input to Model B (ther-mal shield to primary shiel d).
In this way, the effect of the specific power distribution of ANO-1 cycle 5 on vessel fluence was included in the calculation, k
Flux output from the 00T4 calculations required only an axial distribution correction to provido absolute values.
An axial shape factor (local to average axial flux ratio) was obtained from predicted fuel burnup distribu-tions in the peripheral fuel assemblies nearest the vessel, i.e., the core flats region.
This procedure assumes that the axial fast flux shape at the core edge and the pressure vessel are equivalent.
In the 177-FA reactor geometry, this is considered to be a conservative assumption because axial shape should tend to flatten as distance from the core increases.
An axial factor of 1.15 was applied to the calculated flux for the ANO-1 reactor ves-sol. This axial factor was time-averaged over the irradiation period.
0-5 Babcock & Wilcom A WDetmott rompany
In addition to the axial shape factor, one other factor is applied to the flux to normalize it to measured values.
From previous analyses, using equ! valent D0T4 models and constant terms in equation D-2 to obtain the cal-culated activity, a normalization factor of about 0.95 has been consistent-ly obtained.
Although this factor is strictly correct only at the capsule location, it is assumed to be applicable to all locations in the reactor model.
This assumption is based on the following considerations:
(1) B&W 177-FA reactors have essentially the same configuration and (2) the pres-sure vessel and capsule are separated by only 15 cm of water so that it is unlikely that aq/ significant change in accuracy would occur in this dis-tance.
Due to the consistency of this factor at the capsule and the con-siderations above, it will be applied to the ANO-1 cycle 5 calculated ves-sel flux with only a minor correction to account for DOT 4 model change and 137 s yield values which have been impicmented since the previous revised C
analyses.
The model change incorporated a more accurate water temperature in the core fomer region.
The previous model assumed that this water re-gion was at the average core water temperature; recent information indi-cates that an average of the core and downcomer temperature is more realis-tic.
This decrease in water temperature increases its density.
From DOT 4 studies, this density increase causes a reduction in the vessel flux by ap-proximately 4%.
The second correction is due to use of recently recommend-137 s yield values from the fission of 238U and 237Np.
Use of the ed ASTM C
recommended values reduce the calculated activity used to obtain the normal-ization factor by approximately 5%.
Since both changes reduce the calcu-lated activity, i.e., the denoninator in equation 0-2, the 0.95 normaliza-tion factor must be increased.
Thus, the effective normalization factor becomes 1.04 x 1.05 x 0.95 = 1.04.
This normalization factor was applied to the calculated fluxes for this analysis.
2.
Vessel Fluence Extrapolation For up-to-date operation, fluence values in the reactor vessel are calcu-lated as described above.
Extrapolation to future operation is required for the prediction of vessel life based on minimum USE and for the calcula-tion of pressure-temperature operation curves.
Three time periods are D-6 Babcock & WHcom a Mc Dermott (ompany
s f(
considered:
(1) to-date operation for which vessel fluence has been calcu-(
lated (2) designed future fuel cycles for which P0Q calculations have been performed for fuel management analysis of reload cores, and (3) future fuel
(
cycles for which no analyses exist.
Data from time period 1 are extrano-lated through time period 2 based on the premise that excore flux is prop 3r-(
tional to the fast flux that escapes the core boundary.
Thus, for the ves-
- sel,
(
- e,C l
- v C X vR
- e,R where the subscripts are defined as y = vessel, e = core escape, R = refer-ence cycle, C = a future fuel cycle.
[
Core escape flux is available from P0Q output.
Extrapolation from time pe-riod 2 through time period 3 is based on the last designed fuel cycle in period 2 having the same relative power distribution as an " equilibrium" fuel cycle.
Generally, the designed fuel cycles include several cycles into the future, 6 and 7 for ANO-1.
Therefore, the last cycle in time period 2 should be representative of an " equilibrium" cycle.
Data for ANO-1 are listed in Table 0-2.
This procedure is considered preferable to the alternative of assuming that lifetime fluence is based on a single, hypothetical " equilibrium" fuel cycle because this procedure accounts for all known power distributions.
In addition, errors that may result from the selection of a hypothetical
" equilibrium" cycle are reduced.
0-7 Babcock 4 WIfcom a M(t)ermot t iompany
Table D-1.
Spectrum-Averated Cross Sections Average cross Reaction section. Un (barns)_
"Ni(n.f)58Co 0.1222 54,(,,f)5h 0.09283 7
238(nf)137Cs 2.416 U
237Np(n.f)137Cs 0.4037 0-8 magggees a WOermeeiaespany
~
Table 0-2.
Extrapolation of Reactor Vessel Fluence
- " "*'D "#
Core escape C a lative flux
- Time, time, Vessel flux, Time Cycle n/cm2,s EFPY EFPY n/cm -s inte. val Cumulative 2
1A 0.482(+14) 0.95 0.95 1.39(+10) 4.17(+17) 4.17(+17)
IE-4 0.509(+14) 2.88 3.83 1.46(+10) 1.32(+18) 1.73(+18) 5 0.406(+14) 1.22 4.10 1.03(+10) 3.96(+17) 2.13(+18) 6 0.408(+14) 1.10 5.20 1.04(+10)(a) 3.59(+17) 2.49(+18) 7 0.397(+10) 1.15 6.35 1.01(+10)(a) 3.67(+17) 2.86(+18)
>7 0.397(+10)(b) 1.65 8.0 1.01(+10)(a) 5.26(+17) 3.38(+18)
>7 0.397(+10)(b) 7.0 15.0 1.01(+10)(a) 2.23(+18) 5.61(+18)
>7 0.397(+10)(b) 17.0 32.0 1.01(+10)(a) 5.42(+18) 1.10(+19)
?
(a) Predicted value based on proportionality of core escape flux.
(b)After cycle 7, core escape flux is expected to be that of cycle 7, i.e., cycle 7 is "equilibefum" cycle.
E E
1 2:D
1 L
APPENDIX E Capsule Dosimetry Data E-1 Babcock &Wilcox a McDermott company
Table E-1 lists the composition of the threshold detectors and the equiva-
)
lent cadmium thickness used to reduce competing thermal reactions.
Table E-2 shows capsule AN1-A measured activity per gram of target material (i.e., per gram of uranium, nickel, etc. ).
Activation cross sections for
]
235U fission spectrum the various materials were flux-weighted with a (Table E-3).
Table E-1.
Detector Composition and Shielding Monitors Shielding Reaction 238 (n,f) 10.38% U-Al Cd-Ag 0.03" Cd U
1.44% Np-Al Cd-Ag 0.03" Cd 237Np(n,f)
Ni 100%
Cd-Ag 0.03" Cd 58Ni(n,p)58Co 59 o(n,y)60 o 0.66 wt % Co-Al Cd-Ag 0.03" Cd C
C 59 o(n,y)60 o C
C 0.66 wt % Co-Al None 54 e(n,p)54 n F
M Fe 100%
None E-2 Babcock &WUHcom a McDermott company
Table E-2.
Capsule AN1-A Dosimeter Activities f
Dosimeter activity,,pCi/gm cf target Dosimeter Dosimeter material reaction AD)
AD2 AD3 AD4 Ni 58Ni(n,p)58 o 2990 2508 2429 3778 C
54 e(n,p)S4 n 1805 1484 1398 2201 F
M Fe 238 (n,f)137 s 13.4 11.19 9.68 15.04 238 -Al U
C U
237Np-Al 237Np(n,f)137 s 78.54 64.64 53.54 89.81 C
Table E-3.
Dosimeter Activation Cross Sections, b/ atom (a)
Energy range, 54 e(n p) 238 (n,f) 58Ni(n,p)
G_
MeV 237Np(n,f)
F U
1 12.2 - 15 2.323 1.050 4.830(-1) 4.133(-1) 2 10.0 - 12.2 2.341 9.851(-1) 5.735(-1) 4.728(-1) 3 8.18 - 10.0 2.309 9.935(-1) 5.981(-1) 4.772(-1) 4 6.36 - 8.18 2.093 9.110(-1) 5.921(-1) 4.714(-1) 5 4.96 - 6.36 1.541 5.777(-1) 5.223(-1) 4.321(-1) 6 4.06 - 4.96 1.532 5.454(-1) 4.146(-1) 3.275(-1) 7 3.01.- 4.06 1.614 5.340(-1) 2.701(-1) 2.193(-1) 8 2.46 - 3.01 1.689 5.272(-1) 1.445(-1) 1.080(-1) 9 2.35 - 2.46 1.695 5.298(-1) 9.154(-2) 5.613(-2) 10 1.83 - 2.35 1.677 5.313(-1) 4.856(-2) 2.940(-2) 11 1.11 - 1.83 1.596 2.608(-1) 1.180(-2) 2.948(-3) 12 0.55 - 1.11 1.241 9.845(-3) 6.770(-4) 6.999(-5) 13 0.111 - 0.55 2.34(-1) 2.432(-4) 1.174(-6) 1.578(-8) 14 0.0033-0.111 6.928(-3) 3.616(-5) 1.023(-7) 1.389(-9)
(a)ENDF/B5 valp. e have been flux-weighted (over CASK energy groups) based on a 50 fission spectrum in the fast energy range plus a 1/E shape in the intermediate energy range.
E-3 Babcock &WHcos a McDermott company
r l
APPENDIX F References l
F-1 Babcock & Wifcom a McDermott company
l 1 A. L. Lowe, Jr., et al., Analysis of Capsule AN1-E From Arkansas Power &
Light Company, Arkansas Nuclear One -- Unit 1, Reactor Vessel Material Surveillance Program, BAW-1440, Babcock & Wilcox, Lynchburg, Virginia, f
April 1977.
2 A. L. Lowe, Jr., et al., Analysis of Capsule AN1-B From Arkansas Power &
Light Company's Arkansas Nuclear One, Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1698, Babcock & Wilcox, Lynchburg, Vi rginia, November 1981.
3 G. J. Snyder and G. S. Carter, Reactor Yessel Material Surveillance Pro-gram, BAW-10006A, Rev. 3, Babcock & Wilcox, Lynchburg, Virginia, January 1975.
4 A. L. Lowe, Jr., K. E. Moore, and J. D. Aadland, Integrated Reactor Ves-sel Material Surveillance Program, BAW-1543, Rev.
2, Babcock & Wilcox, Lynchburg, Virginia, February 1984.
5 A. L. Lowe, Jr., et al., Analyses of Capsule RSI->D, Sacramento Municipal Utility District, Rancho Seco Unit 1, Paactor Vessel Material Surveil-lance Program, BAW-1792, Babcock & Wilcox, Lynchburg, Virginia, October 1983.
6 H. S. Palme and H. W. Behnke, Methods of Compliance With Fracture Tough-ness and Operational Requirements of Appendix G to 10 CFR 50, BAW-10046A, Rev.1, Babcock & Wilcox, Lynchburg, Virginia, July 1977.
7 00T4 Two-Dimensional Discrete Ordinates Radiation Transport Code, NPGD-TM-559, Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, August 1981.
8 CASK 40-Group Coupled Neutron and Gamma-Ray Cross Section Data, RSIC-DLC-23E, Radiation Shielding Information Center, March 1975.
1 J
F-2 Babcock &WHcom A MCDermott Company
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