ML20203P927
| ML20203P927 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/31/1986 |
| From: | Collins L, Ewing J, Lowe A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20203P928 | List: |
| References | |
| 77-1163819, 77-1163819-00, BAW-1901, NUDOCS 8605080364 | |
| Download: ML20203P927 (85) | |
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jI BAW-1901 March 1986 I
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I ANALYSIS OF CAPSULE TMIl-C I
GPU NUCLEAR THREE MILE ISLAND NUCLEAR STATION-UNIT 1
-- Reactor Vessel Material Surveillance Program --
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[88""888ggggg6 a ucoermott company PDR t_.................
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BAW-1901 March 1986 E
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ANALYSIS OF CAPSULE TMIl-C GPU NUCLEAR THREE MILE ISLAND NUCLEAR STATION-UNIT 1
-- Reactor Vessel Material Surveillance Program --
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by A. L. Lowe, J r., PE L. L. Collins I'
J. W. Ewing W. A. Pavinich I
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B&W Document No. 77-1163819-00 I
BABCOCK & WILCOX Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 I
Babcock & Wilcox J McDermott company
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SUMMARY
This report describes the results of the examination of the second capsule of the General Public Utilities Nuclear Three Mile Island Nuclear Station-Unit 1 reactor vessel surveillance program.
The capsule was removed and examined 18 2
after accumulating a fluence of 8.66 x 10 n/cm (E > 1 MeV), which is equivalent to approximately 35 effective full power years' operation of the reactor vessel.
The objective of the program is to monitor the effects of neutron irradiation on the tensile and f racture toughness properties of the reactor vessel materials by the testing and evaluation of tension and Charpy g
impact specimens.
3 The capsule received an average fast fluence of 8.66 x 10 n/cm2 (E > 1.0 18 Mev) and the predicted fast fluence for the reactor vessel T/4 location at 10 the end of the fifth cycle is 1.06 x 10 n/cm2 (E > MeV).
Based on the calculated fast flux at the vessel wall and 80% capacity factor, the p roj ected fast fluence at the Three Mile Island Unit I reactor pressure 18 vessel will receive in 32 EFPY operation is 7.98 x 10 n/cm2 (E > 1 MeV).
The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure.
The Charpy impact data exhibited the characteristic shift to higher temperature for the 30 and 50 ft-lb transition temperatures as a result of neutron fluence damage and a decrease in upper she.lf energy.
These results demonstrated that the current techniques used for predicting the increase in the RT and the decrease in NOT upper shelf properties due to irradiation are conservative.
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l CONTENTS l
Page 1.
INTRODUCTION...........................
1-1 2.
BACKGROUND............................
2-1 3.
SURVEILLANCE PROGRAM DESCRIF ?ON..
3-1 4.
eRE-1RRAD1A110N 1eS1S 4-1 4.1.
Tension Tests 4-1 4.2.
Impact Tests........................
4-1 5.
POST-IRRADIATION TESTS 5-1 5.1.
Thermal Monitors......................
5-1 5.2.
Tension Test Results....................
5-1 I
5.3.
Charpy V-Notch Impact Test Results.............
5-2 6.
NEUTRON DOSIMETRY 6-1 6.1.
Background.........................
6-1 6.2.
Vessel Fl uence...........
6-3 1
6.3.
Capsule Fluence 6-3 6.4.
Fluence Uncertainties 6-4 7.
DISCUSSION OF CAPSULE RESULTS 7-1 7.1.
Pre-Irradiation Property Data 7-1 7.2.
Irradiated Property Data..................
7-1 I
7.2.1.
Tensile Properties.................
7-1 7.2.2.
Impact Properties 7-2 7.3.
Reactor Vessel Fracture Toughness 7-4 8.
SUMMARY
OF RESULTS........................
8-1 I
9.
SURVEILLANCE CAPSULE REMOVAL SCHEDULE 9-1
- 10. CERTIFICATION 10-1 B
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I CONTENTS (Cont'd.)
APPENDICES Page A.
Reactor Vessel Surveillance Program -- Background Data and Information A-1 B.
Pre-Irradiation Tensile Data B-1 C.
Pre-Irradiation Charpy Impact Data C-1 D.
Fluence Analysis Procedures.................. D-1 g
E.
Capsulo Dosimetry Data E-1 E
F.
References F-1 List of Tables Table 3-1.
Specimens in Surveillance Capsule TMIl-C 3-2 3-2.
Chemistry and Heat Treatment of Surveillance Materials.....
3-3 5-1.
Tensile Properties of Capsule TMIl16C Base Metal and Weld Metal Irradiated to 8.66 x 10 n/cm2 (E > 1 MeV) 5-3 5-2.
Charpy Impact Data From Capsule TMIl1g Haz Metal, Longitudinal Orientation, Irradiated to 8.66 x 10 n/cm2 (E > 1 MiV) 5-3 Charpy Impact Data From Capsule TMIl1g,n/cmBasgMetal, Transverse 5-3.
Orientation, Irradiated to 8.66 x 10 (E > 1 MeV) 5-4 5-4.
Charpy Imoact Data From Capsule TMIlg Base Metal, Longitudinal Orientation, Irradiated to 8.66 x 10 n/cm (E > 1 MeV) 5-4 5-5.
Charpy Impact Data From Capsug TMIl-C, Weld Metal WF-25 Irradiated to 8.66 x 10 n/cm2 (E > 1 MeV) 5-5 5-6.
Charpy Impact Data From Capsule TMIl-C Correlation Monitor Mggerialg (eat No. A-1195-1, Irradiated to H
8.66 x 10 n/cm E > 1 MeV) 5-5 6-1.
Surveill ance Capsule Detectors.................
6-5 I
6-2.
Pressure Vessel Flux......
6-5 6-3.
Calculated TMI-1 Reactor Vessel Fluence 6-6 6-4.
Surveillance Capsule Fluence.................. 6-7 6-5.
Estimated Fluence Uncertainty 6-7 7-1.
Cc.nparison of Capsule TMIl-C Tensile Test Results 7-6 7-2.
Summary of Three Mile Island Unit 1 Reactor Vessel Surveillance Capsules Tension Test Results...........
7-7 7-3.
ObservedVs.PredictedChangesinIrgdiateg Charpy Impact Properties - 8.66 x 10 n/cm 7-8 7-4.
Summary of Three Mile Island Unit 1 Reactor Vessel Surveillance Capsules Charpy Impact Test Results........
7-9 7-5.
Evaluation of Reactor Vessel End-of-Life Fracture Toughness and Pressurized Thermal Shock Criterion - Three Mile Island Unit 1.
7-10 7-6.
Evaluation of Reactor Vessel End-of-Life Upper Shelf Energy 7-11 I
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l List of Tables (Cont'd.)
l Table Page A-1.
Unirradiated Impact t'roperties and Residual Element Content I
Data of Beltline Region Materials Used for Selection of Surveillance Program Materials -- Three Mile Island Unit 1 A-3 A-2.
Test Specimens for Determining l
Material Baseline Properties..................
A-4 5
A-3.
Specimens in Upper Surveillance Capsules A-5 (Designations A, C, and E)
A-4.
Specimens in Lower Surveillance Capsules l
(Designations B, D, and F)
A-5 B-1.
Pre-Irradiation Tensile Proporties for Shell Plate Material Heat C2789-2 B-2 I
B-2.
Pre-Irradiation Tensile Properties for Wel d Metal WF-25........................
B-2 C-1.
Pre-Irradiation Charpy Impact Data for Shell Plate Material -- Longitudinal Direction, Heat C2789-2........
C-2 i
C-2.
Pre-Irradiation Charpy Impact Data for Shell Plate Material, Transverse Orientation Heat C2789-2 C-3 C-3.
Pre-Irradiation Charpy Impact Data for Shell Plate I
Material, HAZ, Transverse Orientation, Heat C2789-2 C-4 C-4.
Pre-Irradiation Charpy Impact Data for Weld Metal, WF-25....
C-5 D-1.
Spectrum Averaged Cross Sections................
D-7 D-2.
Extrapolation of Reactor Vessel Fluence D-8 E-1.
Detector Composition and Shieldina...............
E-2 E-2.
Capsule TMIl-C Dosimeter Specific Activities E-2 E-3.
Dosimeter Activation Cross Sections, b/ atom E-3 List of Figures 3-1.
Reactor Yessel Cross Section Showing Location of Capsule TMIl-C in Three Mile Island Unit 1 3-4 3-2.
Reactor Vessel Cross Section Showing Location of I
Capsule TMIl-C in Crystal River Unit 3............. 3-5 3-3.
Loading Diagram for Test Specimens in Capsule TMIl-C...... 3-6 5-1.
Charpy Impact Data for Irradiated Plate Material, Longitudinal Orientation, Heat C2789-1 5-6 I
5-2.
Charpy Impact Data for Irradiated Shell Plate Material, Transverse Orientation, Heat C2789-1..............
5-7 5-3.
Charpy Impact Data for Irradiated Plate Material, I
Heat-Affected Zone Heat C2789-1 5-8 5-4.
Charpy Impact Data for Irradiated Weld Metal, WF-25 5-9 5-5.
Charpy Impact Data for Irradiated Correl ation Material, HSST PL-02................ 5-10 6-1.
Reactor Vessel Fast Flux, Fluence and DPA Distribution.....
6-8 6-2.
Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface 6-9 v
Babcock & Wilcox a McDermott company
Il List of Figures (Cont'd.)
I Figure Page A-1.
Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel A-6 A-2.
Location and Longitudinal Welds in Upper and lower Courses...
A-7 C-1.
Charpy Impact Data for Unirradiated Plate Material, Longitudinal Orientation, Heat C2789-1.............
C-6 C-2.
Charpy Impact Data for Unirradiated Plate Material, Transverse Orientation, Heat C2789-1........
C-7 E
C-3.
Charpy Impact for Unirradiated Plate Material, B
Heat-Affected Zone, Heat C2789-1..........,
C-8 C-4.
Charpy Impact Data for Unirradiated Weld Metal WF-25...... C-9 I
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INTRODUCTION This report describes the results of the examination of the second capsule of I-the GPU Nuclear Three Mile Island Nuclear Station Unit 1 (TMI-1) reactor vessel material surveillance program (RVSP).
The capsule was removed and evaluated after being irradiated in the Crystal River Unit 3 reactor.
This irradiation in Crystal River Unit 3 plus the previous irradiation in Three Mile Island Unit 1 is the equivalent of 5 years of operation of the TMI-l 18 reactor vessel.
The capsule experienced a fluence of 8.66 x 10 n/cm2 (E >
1 MeV), which is the equivalent of approximately 35 effective full power years' (EFPY) operation of the TMI-l reactor vessel.
The fi rst capsule (TMIl-E) from this program was removed and examined after the first year of operation; the results are reported in BAW-1439.1 The second capsule (TMIl-C) was removed and examined after irradiation in Florida Power I
Corporation Crystal River Unit-3 as part of the Integrated Reactor Vessel Materials Surveillance Program.
The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions.
The surveillance program for TMI-l was designed and furnished by Babcock & Wilcox (B&W) as described in BAW-10006A2 and conducted in accordance with BAW-1543A.3 The program was planned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel.
I The surveillance program for TMI-l was designed in accordance with E185-66 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 design.
Because of the di f ference, additional tests and evaluations were required to ensure
. meeting the requirements of 10 CFR 50, Appendixes G and H.
The recommenda-tions for the future operation of TMI-l included in this report do comply with these requirements.
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BACKGROUND The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water-cooled reactors.
The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation.
The general effects of fast neutron irradiation on the mechanical properties of such low-alloy ferritic steels as SA302, Grade B1, Modified, used in the fabrication of the Three Mile Island Unit 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 yiel d strength properties with a corresponding decrease in I
ductility after irradiation.
The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the Charpy upper shel f energy value.
I Appendix G to 10CFR50, " Fracture Toughness Requirements,4" specifies minimum fracture toughness requirements for the ferritic materials of the pressuro-retaining components of the reactor coolant pressure boundary (RCPB) of I
water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB.
The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal 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 10CFR50 became effective on August 13, 1973, the requirements ara applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the ef fective date.
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E Appendix H to 10CFR50, " Reactor Vessel Materials Surveillance Program Requirements,5" defines the material surveillance program required to monitor changes in the fracture toughness oroperties of ferritic materials in the reactor vessel beltline region of w ater-cool ed reactors resulting from exposure to neutron irradiation and the thermal environment.
Fracture toughness test data are obtained from material specimens withdrawn periodi-cally from the reactor vessel.
These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life.
A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code,Section III, "Nu cl e a r Power Plant Components."
This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RTNOT, which is defined as the greater of the drop weight nil-ductility transition temperature (per ASTM E-208) or the temperature that is 60F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion.
The RT f a given material is used to index that material to a reference NDT stress inter.si ty factor curve (K curve), which appears in Appendix G of IR ASME Section III.
The K curve is a lower bound of dynamic, static, and IR E
crack arrest fracture toughness results obtained from several heats of E
pressure vessel steel.
When a given material is indexed to the K
- cune, IR allowable stress intensity factors can be obtained for this material as a function of temperature.
Allowable operating limits can then be determined using these allowable stress intensity factors.
The RT and, in turn, the operating limits of a nuclear power plant, can NDT be adjusted to account for the effects of radiation on the properties of the reactor vessel materi al s.
The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which a surveillance capsule contain-ing prepared specimens of the reactor vessel materials is periodically removed from the operating nuclear reactor and the specimens are tested.
The increase in the Charpy V-notch 30 ft-lb temperature is added to the original RT to adjust it for radiation embrittlement.
This adjusted RT is used NDT NDT to index the material to the K IR curve which, in turn, is used to set Babcock & Wilcox 2-2 a McDermott company
!,I operating limits for the nuclear power plant.
These new limits take into account the effects of irradiation on the reactor vessel materials.
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SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for TMI-l comprises six surveillance capsules I
designed to monitor the effects of neutron and thermal environment on the materials of the reactor pressure vessel core region.
The capsules, which I
were inserted into the reactor vessel before initial plant startup, were positioned inside the reactor vessel between the thermal shield and the vessel wall at the locations shown in Figure 3-1.
The six capsules, originally designed to be placed two in each holder tube, are positioned 2
near the peak axial and azimuthal neutron flux.
BAW-10006A includes a full description of the capsule locations and design.
After the capsules were removed from THI-l in 1976 and included in the integrated RVSP, they were scheduled and irradiated in the Crystal River Unit 3 reactor 1escribed in 3
BAW-1543 A. During this period of irradiation, capsule TMIl-L.as irradiated in the top location in holder tube WX as shown in Figure 3-2.
Capsule TMIl-C was rmoved during the fifth refueling shutdown of Crystal River Unit 3.
This capsule contained Charpy V-notch impact test specimens fabricated from one base metal (SA302, Grade 81, Modified), one heat-affected-zone, a weld metal and correlation material. Tensile specimens were fabricated from the base metal and the weld metal only.
The specimens contained in the capsule are described in Table 3-1, and the location of the individual specimens within the capsule are described in Figure 3-2.
The
' I chemical composition and heat treatment of the surveillance material in capsule TMIl-C are described in Table 3-2.
I All test specimens were machined f rom the 1/4-thickness (1/4T) location of the plate material.
Charpy V-notch and tension test specimens were cut from the surveillance material such that they were oriented with their longitudi-nal axes either parallel or perpendicular to the principal working direction.
Capsule TMIl-C contained dosimeter wires, descrited as follows:
I 3-1 Babcock & Wilcox a McDermott company
I 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 0.66 wt % Co-Al alloy None Fe None Thermal monitors of low-melting metals and alloys were included in the g
capsule.
The metals and alloys and their melting points are as follows:
W Molting Alloy Point, F 90% Pb, 5% Ag, 5% Sn 558 94.5% Pb, 2.5% Ag, 3.0% Sn 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Cadmium 610 Lead 621 I
Table 3-1.
S. acirnens in Surveillance Capsule TMIl-C I
Number of Test Specirnens Material Description Tension CVN(a) Impact Weld Metal, WF 25 4
e HAZ g
Heat No. C2789-2, Longitudinal 0
8 5
Base Material Plato E
Heat No. C2789-2, Longitudinal 4
8 g
Transverse 0
4 Correlation Material, HSST Plate 02 0
8 Total Per Capsule 8
36 l
(*)CVN Denotes Charpy V-notch.
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rL Table 3-2.
Chmistry and Heat Treatment of Surveillance Materials Chemical Analysis, Weight Percent L
Heat No HSST WeldMgl g) gg Element C2789-2 Plate 02 WF-25 N
L C
0.24 0.23 0.090 j
Mn 1.36 1.39 1.62 P
0.010 0.013 0.014 S
0.017 0.0 13 0.015 Si 0.23 0.21 0.46 Ni 0.57 0.64 0.66 0.10
[
Cr 0.19 L
Mo 0.51 0.50 0.40 Cu 0.09 0.17 0.33 q
Heat Treatment y
l Heat No.
Temp, F Time, h Cooling I
{
C2789-2 1600-1650 9.5 Brine quench 1200-1225 9.5 Brine quench 1600-1650 9.5 Brine quench 1600-1650 9.5 Brine quench I
1510-1535 5.0 Brine quench I
1200-1225 5.0 Brine quench 1100-1150 27.5 Furnace cooled I
HSST PL-02 1600+75 4
Water quenched I
1225+25 4
Furnace cooled l
1125+25 40 Furnace cooled WF-25(c) 1100-1150 27.5 Furnace cooled i
i Per Certified Materials Test Repogt (c)Per Licensing Document BAW-150gP I
(d)Per Licensing Document BAW-1820 Per ORNL-44638 I
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I Figure 3-1.
Reactor Yessel Cross Section Showing Location of Capsule TMIl-C in Three Mile Island Unit 1
}
Surveillance Capsule Holder Tubes - Capsules TMI-lC, l
TMI-1D
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sules DtI-1A, TMI-1B l
Surveillance Capsule Holder Tube - Capsules Z
TMI-1E, TM1-1F, TMI-1G*,
TMI-lE*
- Thermal aging capsules.
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.I Figure 3-2.
Reactor Vessel Cross Section Showing location of Capsule TMIl-C in Crystal River Unit 3 I
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LOCATION CAPSLLE E!1-C SURVEILLANCE l
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SURVEILLANCE CAPSULE HOLDER TUBES SUPVElLL AMCE C AP SULE HOLDF2 TUBE I
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Figuro 3-3.
Loading Diagrara for Test Specimens in Capsule TMIl-C IHis SECil0N ROIAIED 4
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PRE-IRRADIATION TESTS J
Unirradiated material was evaluated for two purposes:
(1) to establish a H
baseline of data to which irradiated properties data could be referenced, and i
(2) to determine those materials properties to the extent practical from available material, as required for compliance with Appendixes G and H to 10CFR50.
l 4.1.
Tension Tests I
l Tension test specimens were fabricated from the reactor vessel shell course forging and weld rnetal.
The specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter.
They were tested on a 55,000-lb load capacity universal test machine at a crosshead speed of 0.050 I
inch per minute.
A 4-pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point.
Test conditions were in accor-dance with the applicable requirements of ASTM A370-77.9 For each material type and/or condition, six specimens in groups of three were tested at both room temperature and 580F.
The tension-compression load cell used had a l
certified accuracy of better than +0.5% of full scale (25,000 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-
.I 10 ments of ASTM Standard Methods A370-77 and E23-82 on an irrpact tester certified to meet Watertown standards.
Test specimens were of the Charpy V-notch type, which were nominally 0.394 inch square and 2.165 inches long.
Prior to testing, specimens were temperature-controlled in liquid immersion baths, capable of covering the temperature range from -85 to +550F.
Speci-mens were removed from the baths and positioned in the test frame anvil with tongs specifically designed for the purpose.
The pendulum was released I
4-1 Babcock & Wilcox J McDermott comparty 6.
I manually, allowing the specimens to be broken within 5 seconds from thef r removal from the temperature baths.
Impact test data for the unirradiated basel ine reference materials are presented 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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POST-IRRADIATION TESTS l
5.1.
Thermal Monitors I
Capsule TMIl-C contained three temperature monitor holder tubes, each containing five fusible alloy wires with melting points ranging from 558 to I
621F.
All the thermal monitors at 558, 580, and 588F locations had melted while those at the 610F location showed no signs of melting or slumping; the l
monitor at the 621F location melted in all three holder tubes.
It is there-fore assumed that the 610F and 621F monitors were placed in the wrong l
locations in the bolder tubes.
From these observations, it was concluded that the capsule had been exposed to a peak temperature in the range of 610 l
to 621F during the reactor operating period.
These peak temperatures are attributed to operating transients that are of short durations and are judged to have insignificant effect on irradiation damage.
Short duration operating l
transients cause the use of thermal monitor wires to be of limited value in I
determining the maximum steady state operating temperature of the surveil-I lance capsules; hcwever, it is judged that the maximum steady state operating temperature of specimens in the capsule was held within +25F of the 1/4T l
vessel thickness location temperature of 577F.
It is concluded that the capsule design temperature may have been exceeded during operating transients j
but not for times and temperatures that would make the capsule data unusable.
l 5.2.
Tension Test Results I
The results of the postirradiation tension tests are presented in Table 5-1.
Tests were performed on specimens at both room temperature and in the temper-ature range of 480 to 580F using the same test procedures and techniques used 6
to test the unirradiated specimens (Section 4.1).
In general, the ultimate strength and yield strength of the material increased with a corresponoing slight decrease in ductility as compared to the unirradiated values; both effects were the result of neutron radiation damage.
The type of behavior I
5-1 Babcock & Wilcox a McDermott company
I 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.
The results of the pre-irradiation tension tests are presented in Appendix B, 5.3.
Charpy V-notch Impact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor g
vessel beltline material are presented in Tables 5-2 through 5-6 and Figures W
5-1 through i-5.
The test procedures and techniques were the same as those used to test the unirradiated specimens (Section 4.2).
The data show that the materials exhibited a sensitivity to irradiation within the values to be expected from their chemical cornposition and the fluence to which they were exposed.
The results of the pre-irradiation Charpy V-notch impact tests are given in Appencix C.
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Weld Metal Irradiated to 8.66x10}-C Bage Metal and Tensile Properties of Capsule TM Table 5-1.
n/cm (E > 1 MeV)
Strength, psi Elongation, %
Red'n.
Specimen Test Temp, in Area, No.
F Yield Ultimate Uniform Total Base Metal, Longitudinal CC709 67 75,400 99,300 11 25 66 CC701 480 70,200 94,200 10 22 64 CC716 520 70,400 95,800 9
20 62 CC722 580 69,600 97,200 11 23 56 b
Weld Metal CC103 67 97,700 105,700 11 23 55 I
CC101 480 79,500 96,500 10 18 52 CCl22 520 78,000 95,900 10 18 50 l
CC110 580 80,000 97,100 9
17 49 i
Table 5-2.
Ct.arpy Impact Data From Capsule TMIl-C Haz Vetal, Longigdinal Orientation,2 (E > 1 MeV)
Irradiated to 8.66 x 10 n/cm Absorbed Lateral Shear l
Specimen Test Temp,
- Energy, Expagston,
- Fracture, No.
F ft-lb 10-in.
i CC417
-100 28.0 20.0 30 lI CC438
-70 44.0 25.0 40 CC424
-40 46.0 35.0 40 I
CC420
-20 78.0 51.0 50 I
CC406 0
52.0 35.0 40 CC436 40 92.0 59.0 50 CC430 69 83.0 60.0 90 CC423 200 110.0 82.0 100 6
)
5-3 Babcock & Wilcox J McDermott company
Il Ii I
Table 5-3.
CharpyImpactDataFromCapsuleTMIl-C,BaseMegl, Trangverse Orientation, Irradiated to 8.66 x 10 n/cm (E > 1 MeV)
Absorbed Lateral Shear Specimen Test Temp,
- Energy, Expa ston
- Fracture, aj No.
F ft-lb 10-in.
5 CC611 30 25.0 23.0 20 CC604 69 44.0 42.0 25 CC632 115 60.0 50.0 50 CC603 200 79.0 72.0 100 Table 5-4.
Charpy Impact Data From Capsule TMIl-C Base Metal,8 1
LongjtudinalOrientation,Irradiatedto8.66x10 e
n/cm (E > 1 MeV)
Absorbed Lateral Shear Specirnen Test Temp,
- Energy, Expagston,
- Fracture, No.
F ft-1b 10-in.
5 CC737
-20 18.0 13.0 5
CC731 0
22.0 20.0 10 CC711 15 44.0 40.0 20 CC720 30 58.0 44.0 30 CC740 40 51.0 39.0 20 CC718 69 55.0 40.0 15 CC728 200 115.0 87.0 100 CC734 300 113.0 80.0 100 I
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lI l
Table 5-5.
Charpy Impact Data From Capsule TMg-C, Weld Metal 2 (E > 1 MeV)
WF-25 Irradiated to 8.66 x 10 n/cm i
t Absorbed Lateral Shear Specimen Test Temp,
- Energy, Expagston
- Fracture, No.
F ft-lb 10-in.
I CCO29 69 19.0 17.0 5
l CC008 110 24.0 21.0 30 CC019 150 33.0 29.0 40 l
CC037 200 36.0 38.0 60 CC015 250 45.0 42.0 100 CC039 300 49.0 46.0 100 CC004 400 48.0 47.0 100 j
CC014 500 50.0 48.0 100 I
I Table 5-6.
Charpy Impact Data From Capsule TMIl-C Correlation I
Monitor Mater 1al, Heat No. A-1195-1, Irradiated 8
to 8.66 x 10 n/cm2 (E > 1 MeV) l Absorbed Lateral Shear Specirren Test Temp,
- Energy, Expansion,
- Fracture, I
No.
F ft-1b 10-3 in.
CC945 69 13.0 14.0 5
CC921 105 27.0 23.0 20 CC915 125 25.0 23.0 25 CC961 150 41.0 34.0 30 CC928 175 59.0 48.0 50 CC937 200 78.0 60.0 60 CC916 J00 97.0 74.0 100 CC955 400 97.0 81.0 100 iI I
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'I 5-5 Babcock & Wilcox J McDermott company
I Figure 5-1.
Charpy Impact Data for Irradiated Plate Material, Longitudinal Orientation, Heat C2789-2 W
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Figure 5-2.
Charpy Impact Data for Irradiated Plate Material, Transverse Orientation, Heat C2789-2 1%
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I Figure 5-3.
Charpy Impact Data for Irradiated Plate Material, Heat-Af fected Zone, Heat C2789-2 i
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Figure 5-4.
Charpy Impact Data for Irradiated Weld Metal, WF-25 100 I
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I Figure 5-5.
Charpy Impact Data for Irradiated Correlation Material, HSST PL-02 1%
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'I I
I 6.
NEUTRON DOSIMETRY 6.1.
Background
Fluence analysis as part of the pressure vessel surveillance program has three objectives:
a) determination of maximum fluence at the pressure vessel as a function of reactor operation, b) prediction of pressure vessel fluence in the future, and c) determination of the test specimen fluence 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 properties and fluence exposure.
Fluence data are obtained either directly or indirectly from flux I
distributions calculated with a computer model of the reactor.
The accuracy of calculated fast flux is enhanced by the use of a normalization factor based on measured activity data obtained from previous capsule analyses.
A significant aspect of the surveillance program is to provide a correlation between the neutron fluence above 1 MeV and the radiation-induced property changes noted in the surveillance specimens.
To permit such a correlation, activation detectors with reaction thresholds in tho energy rango of interest were placed in each surveillance capsule.
Properties of the detectors are given in Tables 6-1 and E-1.
I Because of a long half-life (30 years) and ef fective threshold energies of 0.5 and 1.1 MeV, the measurements of Cs production from fission reactions 237 2O in Np and U are more directly applicable to analytical determinations of the fast neutron fluence (E > 1 MeV) for multiple fuel cycles than are other dosimeter reactions.
Other dosimeter reactions are useful as cor-roborating data for shorter time intervals and/or higher energy fluxes.
Short-lived isotope activities are representative of reactor conditions only over the latter portion of the irradiation period (fuel cycle), whereas I
I 6-1 Babcock & Wilcox A M(Dermott company
I reactions with a threshold energy greater than 2 or 3 MeV do not record a significant part of the total fast flux.
I The energy-dependent neutron flux is not directly available for activation detectors because the cosimeters register only the integrated effect of neutron 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 operating history of the recctor, the energy response of the given detector, and the neutron spectrum at tee detector location.
Of these parameters, the definition of the neutron spectrum is the most difficult to obtain.
Essentially two means are available to obtain the spectrum:
iterative unfolding of oxperimental dosimeter data and/or analytical methods.
Because of lack of sufficient threshold reaction detectors satisfying both the threshold energy and half-life req ui rements of a surveillance program, calculated spectra have been used.
Neutron transport calculations in two-dimensional geometry are used to obtain energy dependent flux distributions throughout the reactor system.
Reactor conditions are selected to be representative of an average over the g
irradiation time period.
Geometric detail is selected to explicitly 5
represent the reactor system to provide the flux distributions in the reactor vessel.
The capsule flux distributions were obtained in previous analyses with an explicit surveillance capsulo assembly modeled.
The capsule energy-dependent flux distribution enabled generation of calculated dosimeter activities. Comparison of the measured to calculated activities provided the normalization factor applied to the calculated vessel flux and fluence.
Due to the consistency in both the normalization factors and calculated g
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 frotr 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 consistent with previously reported results.
A more detailed description of this calculational procedure is contained in Appendix D.
6-2 Babcock & Wilcox A M(Dermott tompany
l i
6.2.
Vessel Fluence, l
The maximum fluence (E > 1 MeV) in the reactor tessel during TMI-l cycle 2 17 2
l th rough 4 was determined to be 9.72x10 n/cm based on a maximum neutron 10 2
flux of 1.38x10 n/cm -s (Tables 6-2 and 6-3 ).
The location of the maximum j
fluence is a point at the cladding / vessel interface at an azimuthal location 0
of about 11 from a major core axis.
Fluenco data have been extrapolated to 32 EFPY of operation based in part on the promise that ex-core flux is proportional to fast flux escaping the coro and in part on the results of a I
fluence reduction study (Appendix D).
Core escape flux values are available l
from fuel management analyses of the current and future fuel cycles that have been designed.
For TNI-1, core escape flux values were used for cycles 5, 6 l
and 7 where cycles 6 and 7 were modified to represent low leakage cores.
Cycle 8, assumed to be the equilibrium cycle used for extrapolation to 32 l
EFPY, was based on very-lcw leakage cycles.
I Relative fluence and DPA (displacement por atom) as a function of radial l
location in the reactor vessel wall is shown in Figure 6-1.
Reacter vessel fluence lead factors, (clad interface flux /in-vessel flux), at the T/4, T/2, and 3T/4 locations are 1.8, 3.6, and 7.6, respectively.
DPA lead factcrs at the same locations are 1.6, 2.7, and 4.7, respectively.
Representative relative fluence as a function of azimuthal angle is shown in Figuro 6-2.
A 0
peak occurs at about 11 which roughly correspor.ds to a radius through a l
corner fuel assembly of the core and the position of the surveillance holder tube.
The ratio of fast flux at the maximum and minimum locations is about j
l.5.
I 6.3.
Capsule Fluence I
Cumulative fast fluence at the conter of the surveillance capsule was 18 2
calculated to be 8.66x10 n/cm 13% of which occurred in TMI-1, Cycle 1 and 87% in CR-3, Cycles 3 through 5 (Tablo 6-4).
fhese data roprosent an average value for the various lccations in the capsi le.
In TMI-1, Capsule TMIl-C was 0
located in an upper holder tube position at 11 of f the major axis and about 211 cm from the core contor for 466 EFPD.
It was then inserted in CR-3 in an 0
upper holder tube position at 11 of f axis and about 202 cm from the core 6-3 Babcock & Wilcox a M(Dermott company
I center for an additional 1032 EFPD.
Dbring the latter irradiation period, 0
the capsule was estimated to have been rotated approximately 90 counter-clockwise relative to its original design orientation (i.e.,
keyway facing the reactor core).
I 6.4.
Fluence Uncertainties I
Uncertainties were estimated for the fluence values reported herein.
These cata, Table 6-5, were based on comparisons to benchmark experiments when available, estimated and measured variations in input data, and engineering judgement.
Because of the compl exi ty of the fluence calculations, n 5
comprehensive uncertainty limits exist for these results.
The values in 5
Table 6-4 represent a best estimate based on past experience with these analyses.
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Table 6-1.
Surveillance Capsule Detectors Ef fective Lower I
Energy Limit Isotope Detector Reaction MeV Hal f-L i fe
- Fe(n,p)S4Mn4(a) 2.5 312.5 days 58Ni(n,p)S8Co 'I 2.3 70.85 days I
238 (n,f)137Cs(*)
1.1 30.03 years I
U 237Np(n f)l37 I
Cs *I 0.5 30.03 years 238 (n,f)106Ru 1.1 3 69 days U
237Np(n,f)l06Ru 0.5 369 days 238 (n,f)103Ru 1.1 39.43 days U
D Np(n,f)
Ru 0.5 39.43 days 238 (n,f)l44Ce 1.1 284.4 days U
237Np(n,f)144Ce 0.5 284.4 days 238 (n,f)95Zr 1.1 64.4 days U
237Np(n,f)95Zr 0.5 64.4 days
(*IReaction activities measured for capsule flux evaluation.
I Table 6-2.
Pressure Vessel Flux 2
Fast Flux (E > 1 MeV), n/cm.3 Inner Surface Flux (E > 0.1 MeV) 2 (Max Location)
T/4 3T/4 n/cm -s I
10 9
9 10 Cycle 1 1.45x10 8.1x10 1.9x10 2.95X10 (466 EFPD) 10 9
9 Cycle 2-4 1.38x10 7.6x10 1.8x10 (817 EFPD)
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I Table 6-3.
Calculated TMI-l Reactor Vessel Fluence 2
REACTCR VESSEL FAST FLUENCE (n/cm )
2 Fast Fluence, E > 1 MeV, (Max. Location), n/cm Cycle EFPY Inner Surface T/4 T/2 3 T/4 E001 1.28 5.90E+17 3.3E+17 1.6E+17 7.7E+16 E004 3.51 1.56E+18 8.7E+17 4.3E+17 2.0E+17 EOC5 4.28 1.90E+18 1.06E+18 5.3E+17 2.5 E+ 17 g
E0C6 5.41 2.31E+18 1.3E+18 6.4E+17 3.0E+17 3
EOC7 6.61 2.69E+18 1.5E+18 7.5E+17 3.5E+17 E0C8 7.81 2.94E+18 1.6E+18 8.2E+17 3.8E+17 a
10.00 3.40E+18 1.9E+18 9.4E+17 4.4 E+18 5
15.00 4.44E+18 2.5E+18 1.2E+18 5.8E+17 21.00 5.69E+18 3.2E+18 1.6E+18 7.4 E+17 32.00 7.98E+18 4.4E+18 2.2E+18 1.04E+18 Lead Factors (E > 1 MeV) 1.0 1.8 3.6 7.7 REACTOR VESSEL-CPA; DPA, (Max. Location)
Cycle EFPY Inner Surface T/4 T/2 3T/4 EOCl 1.28 5.90E+17 3.7E+17 2.3E+17 1.3 E+17 EOC4 3.51 1.56E+18 9.8E+17 6.0E+17 3.3E+17 EOC5 4.28 1.90E+18 1.2E+18 7.3E+17 4.0E+17 g
EOC6 5.41 2.31E+18 1.4E+18 8.9E+17 4.9E+17 m
EOC7 6.61 2.69E+18 1.7E+18 1.0E+18 5.7E+17 EOC8 7.81 2.94E+18 1.8E+18 1.1E+18 6.3E+17 g
10.00 3.40E+18 2.1E+18 1.3 E+18 7.2E+17 5
15.00 4.44E+18 2.8E+18 1.7E+18 9.4E+17 21.00 5.69E+18 3.6E+18 2.2E+18 1.2E+16 32.00 7.98E+18 5.0E+18 3.1E+18 1.7E+18 Lead Factors (DPA) 1.0 1.6 2.6 4.7 I
I E
I 6-6 Babcock & Wilcox a McDermott company
l Table 6-4.
Surveillance Capsule Fluence Cummulattve Irradiation Flux (E > 1 MeV),
- Fluenge, 2
Time n/cm -s n/cm I
h TMI-1, Cycle 1 2.62x1010 18 1.07x10 3
(466.4 EFPD) l 10 10 n
CR-3, Cycles 3-5 7.76x10 7.59x10 g
(1132.4 EFPD) l Cumulative fluence 8.66x1018 I
l Table 6-5.
Estimated Fluence Uncertainty l
l Estimated l
Calculated Fluence Uncertainty Basis of Estimate In the capsule
+18%
Activity measurements, cross I
section fission yields, l
saturation factor, deviation y
from average fluence value I
In the reactor vessel
+24%
Activity measurements, cross I
at maximum location for sections, fission yields, cycles 4 of TMI-1 f actors, axial factor, capsule location, radial / azimuthal extrapolation, normaliza-zation factor i
In the reactor vessel
+26%
Factors in vessel fluence above at the maximum location pl u s uncertainties for extra-l for end-of-life polation to 32 EFPY extrapol ation I
i 1
6-7 Babcock & Wilcox A McDermott company
I Figure 6-1.
Reactor Vessel Fast Flux, Fluence and CPA Distributton 1
1.6
- "a e
L.F. =
=
.64
- a.. :
N o
u o.7
$ f
~
o s
c I
~
8 L.F. =
= 2.6 g 0.5 l
38 3 i T/4 e
0.4 4
g 223.00 cm DPA 3
w e4 c
1
~
g
.2 L.F. =
= 4.7 E > 1.0 MeV
.21 o
0.3 u
.e ev 1.8 M
L.F. = 1
=
57 2
e
<n N
6 4
228.36 cm
~~'
s x=
o V
6 fu t
am a
e m
L.F. = 1 = 3.6 o
.29 t._
m
(
M w
m e
3T/4 o
o 233.72 cm
.a 2.:3 L.F. = 1
= 7.7
- ,,7 13 2.:6
- .:5 (Typical B&W 177 F.A. Plant) 1 I
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z,:
,a=
g Radial Distance from Core Center (cm)
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6-8 Babcock & Wilcox a McDermott company
M M
M 0
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7.
DISCUSSION OF CAPSULE RESULTS 7.1.
Pre-Irradiation Property Data A review of the unirradiated properties of the reactor vessel core beltline region materials indicated no significant deviation frco expected properties except in the case of the upper shelf properties of the weld metal.
Based on the predicted end-of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of the weld metal, it was predicted that the end-of-service Charpy upper shelf energy (USE) will be below 50 f t-lb.
This weld was selected for inclusion in the surveillance program in accordance with the criteria in ef fect at the time the program was designed for Three Mile Island Unit 1.
The applicable selection criterion was based on the unirradiated properties only.
I 7.2.
Irradiated Property Data 7.2.1.
Tensile Proporties I
Table 7-1 compares irradiated and unirradiated tensile properties.
At both roca temperature and elevated temperature, the ultimate and yield strength changes in the base metal as a result of irradiation and the corresponding changes in ductility are within the limits observed for similar materials.
I There is some strengthening, as indicated by increases in ultimate and yield strengths and decreases in cuctility properties.
All changes observed in the I
base metal are such as to be considered within acceptable limits.
The changes at both room temperature and 580F in the properties of the weld metal are l a rge r than those obse rved for the base metal, indicating a greater sensitivity of the weld metal to irradiation damage.
In either case, the changes in tensile properties are insignificant relative to the analysis of the reactor vessel materials at this period in service life.
I 7-1 Babcock & Wilcox I
a McDermott company
I A comparison of the tensile data from the first capsule (Capsule TMIl-E) with the corresponding data from the capsule reported in this report is shown in Table 7-2.
The currently reported capsule experienced a fluence that is eight times greater than the first capsule.
I The ge neral behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction of area.
The most significant observation from these data is that the weld metal exhibited greater sensitivity to neutron radiation than the base metal.
7.2.2.
Impact Properties The behavior of the Charpy V-notch impact data is more significant to the calculation of the reactor system's operating limitations.
Table 7-3 compares the observed changes in irradiated Cnarpy impact properties with the predicted changes.
The 30 ft-lb transition temperature shift for the base metal is not in good agreement with the value predicted using either Regulatory Guide 1.99, Rev.
II 12 I
or Rev. 2 It would be expected that these values would exhibit better agreement when it is considered that a major portion of the data used to develop Regulatory Guide 1.99, Rev. 2, was taken at the 30 f t-lb temperature.
The transition temperature measurements at 30 ft-lbs for the weld metal is in poor agreement with the predicted shift using Regulatory Guide 1.99, Revision 1,
(R.G.
1.99, Rev. 1) but is in good agreement with the results using Regulatory Guide 1.99, Revision 2 (R.G. 1.99, Rev. 2).
At the 30 f t-lb l ev el, the shift is in good agreement with the predicted value which t
indicates that the estimating curves of Regulatory Guide 1.99, Rov. 2, are conservative for predicting the 30 ft-lb transition temperature shifts since the method requires that a margin be added to the calculated valve to provide a conservative value.
E The 50 ft-lb transition temperature shift for the base metal is not in good agreement with the shift that would be predicted according to Regulatory 7-2 Babcock & Wilcox a McDermott company
L E
Guido 1.99, Rev. 1.
The less than ideal comparison may be attributed to the
(
spread in the data of the unirradiated material combined with a minimum of data points to establish the irradiated curve.
Under these conditions, the I
comparison indicates that the estimating curves in RG 1.99 for icw-copper materials and at mid-range fluence level s are conservative for predicting the 50 ft-lb transition temperature shifts.
The data for the decrease in Charpy USE with irradiation shcwed good agree-ment with predicted values for both the base metal and the weld me tal.
However, the good comparison of the measured data with the predicted value is not expected in view of the lack of data for lower, medium, or high-copper-content materials at medium fluence values that were used to develop I
the estimating curves.
l A comparison of the Charpy impact data from the fi rst capsule (Capsule TMIl-E) with the corresponding cata from the capsule reported in this report is shown in Table 7-4.
The currently reported data experienced a fluence that is cight times greater than the first capsule.
l The base metal exhibited shifts at the 30 ft-lb and 50 ft-lb levels for the latest capsule that were similar to those of the fi rst capsule.
The I
corresponding data for the weld metal showed a further increase at the 50 ft-lb lev 91 as compared to the increase of the 30 ft-lb level.
This was due l
to a decrease ~ln the upper shelf energy with a corresponding increased shift at the 50 ft-lb level and an apparent increased shift at the 30 ft-lb level.
1 Both the base metal and the weld metal exhibited a decrease in the upper shelf values similar to the previous capsule. The weld metal in this capsule exhibited a greater decrease than the weld metal in the previous capsule.
II These data confirm that the upper shelf drop for this weld metal did not reach saturation as observed in the results of capsules evaluated by others.
This behavior of Charpy USE drop for this weld metal should not be considered indicative of a similar behavior of upper shel f region fracture toughness properties.
This behavior indicates that other reactions may be taking place within the material besides simple neutron damage.
Verification of I
I 7-3 Babcock & Wilcox a McDermott company
I this relationship must await the testing and evaluation of the data from compact fracture toughness test specimens.
I estimating Results from other surveillance capsules also indicate that RThDT curves have greater inaccuracies than originally thought. These inaccuracies are a function of a number of parameters related to the basic data available at the time the estimating curves are established.
These parameters may include inaccurate fluence values, poor chanical composition values, and variations in data interpretation.
The change in the regulations requiring the shift measurement to be based on the 30 f t-lb value has minimized the errors that result f rom u sing the 30 f t-lb data base to predict the shift I
behavior at 50 f t-lbs.
e The design curves for predicting the shift will continue to be modified ac more data become available; until that time, the design curves for predicting the RT shift as given in Regulatory Guide 1.99, Revision 2, are considered NDT shift of those materials for which data are adequate for predicting the RTNDT not available.
These curves will be used to establish the pressure-tempera-ture operational limitations for the irradiated portions of the reactor vessel until the time that new prediction curves are developed and approved.
The lack of good agreement of the change in Charpy USE is further support of 5
the inaccuracy of the prediction curves.
Although the prediction curves are 5
conservative in that they generally predict a larger drop in upper shelf than is observed for a given fluence and copper content, the conservatism can unduly restrict the operational limitations.
These data support the contention that the USE drop curves will have to be revised as more reliable data become available; until that time the design curves used to predict the decrease in USE of the controlling materials are considered conservative.
7.3.
Reactor Vessel Fracture Toughness An evaluation of the reactor vessel end-of-l i fe fracture toughness and the pressurized thermal shock criterion was made and the results are presented in Table 7-5.
I 7-4 Babcock & Wilcox a McDermott company I
I The fracture toughness evaluation shows that the controlling weld metal may have a T/4 wall location end-of-life RT f 244F based on Regulatory Guide NOT 1.99, Revision 2.
This predicted shift is excessive since data from the surveillance capsule exhibited a measured shift of 203F for a fluence of 18 2
8.65 x 10 n/cm.
Ratioing this measured shift to the T/4 wall location fluence, it is estimated that the end-of-life RT shift will be NDT s i gni ficantly less than the value predicted using Regulatory Guide 1.99, Revision 2.
This reduced shift will permit the calculation of less restric-tive pressure temperature operating limitations than if Regulatory Guide 1.99, Revision 2, was used.
The pressurized thermal shock evaluation demonstrates that the Three Mile Island Unit 1 reactor pressure vessel is within the screening criterion limits and, therefore, need not take any additional corrective action as I
required by the regulation prior to license expiration.
I An evaluation of the reactor vessel end-of-life upper shel f energy for each of the materials used in the fabrication was made and the results are presented in Table 7-6.
This evaluation was made because the weld metals used to fabricate the reactor vessel are Linde 80 flux, l ow-u p pe r-sh el f-energy, relative high copper and are expected to be highly sensitive to neutron radiation damage.
Two methods were used to evaluate the radiation induced decrease in upper shel f energy; the method of Regulatory Guido 1.99, Revision 2, which is the same procedure used in Revision 1, and the method 13 presented in BAW-1803 which was developed specifically to address the need I
for an estimating method for this class of weld metals.
The method of Regulatory Guide 1.99, Revision 2, show that all of the weld metals used in the fabrication of the reactor vessel will have an upper shelf energy belcw 50 ft-lbs at the 32 EFPY design life based on the T/4 wall location.
Regulatory Guide 1.99 method predicts a decrease below 50 f t-lbs for the controlling weld metal at the vessel inside wall.
Based on the Three Mile Island Unit 1 surveillance data and the prediction techniques presented in BAW-1803, it is calculated that none of the reactor vessel material upper shelf energies will decrease to below 50 f t-lbs during the vessel design life.
I 7-5 Babcock & Wilcox I
a McOcrmott company
I g
Table 7-1.
Comparison of Capsule TMIl-C Tensile Test Results 5
Elevated Tcep. Test Room Temp. Test Unirrad.
Irrad.
Base Metal -- C2789-2 Fluence, 10x18 n/cm2 (E > 1 MeV) 0 8.66 0
8.66 Ult. Tensile Strength, ksi 94.0 99.3 91.7 97.2 0.2% Yield Strength, ksi 71.1 75.4 62.7 69.6 Total Elongation, %
27 25 26 23 Reduction in Area, %
68 66 59 56 I
Weld Metal -- WF-25 18 Fluence, 10x n/cm2 (E > 1 MeV) 0 8.66 0
8,66 Ult. Tensile Strength, ksi 86.2 105.7 82.0 97.1 0.2% Yield Strength, ksi 69.2 90.7 64.3 80.0 Total Elongation, %
27 23 20 17 Reduction in Area, %
63 55 52 49 I
I I
I I
I I
7-6 Babcock & Wilcox J McDermott company
Table 7-2.
Summary of Threo Mile Island Unit 1 Reactor Yossol Surveillanco Capsules Tension Test Results Ductility, %
Rod.
Test Strength, ksi Total of 10{guonco2 n/cm Temp..F.
Ultimato
(*)
Ytold
% (*)
Elon.
% (*)
Area % (*)
Material 68 27 71.1 Baso Metal 0
RT 94.0 26 59 62.7 (C2789-2) 570 91.7 1.07 RT 95.9
+2.0 72.6
+2.1 27 0
66
-2.9 570 95.6
+4.3 66.5
+6.1 27
+3.8 60
+1.7 8.66 RT 99.3
+5.6 75.4
+6.0 25
-7.4 66
-2.9 580 97.2 46.0 69.6
+11.0 23
-11.5 56
-5.1
)
Weld Metal 0
RT 86.2 69.2 63 27 52 (WF-25) 570 82.0 64.3 20 1.07 RT 97.2
+12.8 82.6
+19.4 25
-7.4 56
-11.1 570 94.8
+15.6 76.0
+18.2 19
-5.0 38
-26.9 8.66 RT 105.7
+22.6 90.7
+31.1 23
-14.8 55
-12.7 5 80 97.1
+18.4 80.0
+24.4 17
-15.0 49
-5.8
(*) Change relative to unirradiated.
Os 6 m b$
?R g=
=D 8E I ::
A0 M
I' Table 7-3.
Observed Ys. Predicted Changes in Irrgiated2 Charpy Impact Proporties - 8.66 x 10 n/cm I
Material Predicted Increase in 30 ft-lb Trans. Temp., F Observed RG 1.99/2(a) RG 1.99/l(b)
Base Material (C2789-2)
Longitudinal 30 56 56 Transverse 13 56 56 Base Material (HAZ), Longitudinal
-12 56 56 Weld Metal (WF-25) 203 205 298 Correlation Material 77 124 144 Increase in 50 ft-lb Trans. Temp., F Base Material (C2789-2) l 56 f) 56f I
5 Longitudinal 8
g Transverse 17 56 56 Base Material (HAZ), Longitudinal 33 56(c) 56(c)
Weld Metal (WF-25) 301 205(c) 298 c)
Correl [tionVaterial 74 124(c) 744(c)
Decrease in Charpy USE, ft-lb Base Material (C2789-2) g Longitudinal 17 22 22 g
Transverse 19 17 17 Base Material (HAZ), Longitudinal 22 22 22 Weld Metal (WF-25) 31 34 34 Correlation Material 33 35 35
(*'Per draft of Regulatory Guide 1.99, Revision 2, dated February 10, 1986.
(b)Per Regulatory Guide 1.99, Revision 1, dated April,1977.
c) Based on the assumption that 50 ft-1 transition temperature is used to cQntrol the shift :1n RTNDT*1 l
?-8 Babcock & Wilcox a McDermott company
m.
r~ t_
rm rm q
m
.m r- \\.
m m_
rm
_ rm rm rm rm rm r
em v
Table 7-4.
Summary of Three Mile Island Unit 1 Reactor Vessel-Surveillance Capsules Charpy Impact Test Results t
Trans. Temp. Increase, F Decrease in Upper j
30 ft-lb 50 ft-lb Shel f Eneray, ft-lb.
Fgence n/cm Observ. Predicted (a)
Observ. Predicted (b) Observ.
Predictedsa) 2 Material 10 Base Metal (L) 1.07 29 25 29 20 11 13 (C2789-2) 8.66 30 56 8
56 17 22 Base Metal (T) 1.07 5
25 15 20 10 10 (C2789-2) 8.66 13 56 17 56 19 17 HAZ Metal 1.07 11 25 8
20 21 13 (C2789-2) 8.66
-12 56 33 56 22 22
[
Weld Metal.
1.07 124 9
121 105 17 24 (WF 25) 8.66 203 205 301 298 31 34 Corr. Mat'1.
1.07 44 56 34 51 15 20 (HSST PL-02) 8.66 77 124 74 144 33 35 l
- 'Per draft of Regulatory Guide 1.99, Revision 2, dated February 10, 1986.
(b)Per Regulatory Guide 1.99, Revision 1, dated April,1977.
.]
I h
is i
w I
ii M
i l
l l
Table 7-5.
Evaluation of Reactor Vessel End-of-Life Fracture Toughness and Pressurized Thermal Shock Critorion - Threo Milo Island Unit 1 Estimated PT5 Evaluation. F' '
NDI. F * '
I Natertal Estteated Fluence at End-Of4 tfe RT Matertal Description Chnmical Composition RV Insidejur f ace. n/cm Reactor Wessel Heat Material Copper Nic kel.
License inside 7/4 Wall License Screening Deltline Regten location Number Type w/o e/o 32 (IPY fwptration Surf ace t oc a t ion 32 (f PY ispiration Criteria tower No331e Delt ARY 059 SA508. C1.2 0.08 0.72 5.98fl0 4.26tl8 116 110 104 100 270 Upper Shell C2789-1 5A533.Cr.B1 0.09 0.57 7.98E18 5.96E18 117 110 107 103 270 Upper Shell C2789-2 SA533.Gr.Bl 0.09 0.57 7.98E18 5.96(18 117 110 107 10s 270 tower Shell C3307-1 5 A53 3.Gr.B1 0.12 0.55 7.98E18 5.%E le 140 130 125 120 2 70 tower Shell C3251-1 5A53 3.Gr.Bl 0.11 0.50 7.98E18
- 5. %[18 Ill 122 117 112 270 Leper Circum. Weld 11001) wF-70 ASA/tinde 80 0.35 0.59 5.98E18 4.26E18 243 219 256 239 300 Middle Circum. Wold (l005)
WF-25 ASA/Linde 80 0.35 0.68 7.98E 18 5.96(18 272 244 283 263 300 Lc=er Circum. Weld (1,0. 505)
WF47 ASA/ttnde 80 0.24 0.60 4.42f16 3.15E16 73 71 94 91 300 L cmer Circum. Weld (0.D. 5051 WF-70 ASA/tInde 80 0.35 0.59 N/D N/D N/D N/D N/A N/A
~~-
typer tongtt. Wold (1005)
WF-8 ASA/ttade 80 0.29 0.55 7.98E18 5.% Ele 234 211 231 215 270 tower Longtt. Weld (0.0. 635) 5A-1494 A5A/Linde 80 0.18 0.63 N/D N/D N/D N/D N/A N/A e
Lc=er tongit. Weld 5A-1526 ASA/Linde 80 0.35 0.68 7.48(18 5.33f18 268 240 2 79 260 270
(!.0. 375. 2005)
O Upper Circum. Weld (1001)
Atyptcal ASA/ttade 80 0.41 0.10 5.98E18 4.Mi18 226 211 236 214 300 Notes (a) Per Draf t Regulatory Guf de I.99. Feev tston 2. Dated February 10. 1986.12 (b) Per 10CTR50. Section 50.61. Fracture Toughness Requtyts for Protection Against Pressurtsed Thermal Shock (vents.
N/ A = Not Appitcable h/O
- Not Deteretned W
am EF
?n
~
R
!=
5D
$I i =
a a O
M M
M M
M M
M M
M m
M m
e e
r-----------
m m
e M
M M
M M
M M
M M
M M
M Table 7-6e Evaluation of Reactor Vossol End-of-Life Upper Shel f Energy No ter tal istleated ICL fluence E stleated lat 9 ( Est fested 10L,p U
istimated EFPT Esttmated (FPY Notertaf Descrtet ten fnest q1_[geggi 19 n
,lajt de Tfg_Pell latti.i __,.c R.o. i., / 2 i
t 9
_pe,es. 180 to 50 f t.in.
to 50 in.
Reactor venen Heat Noterial Copper.
httkel.
5=r f age L et etjoa U 5E.
lasise 1/4 self Tn td.
T/4 sell et T/4 sati BNf-1803 @ '
Boltit=e Regten Locat ton Number Fype e/o e/o ht e/(m
=/cm ft-lbt ivr f et e t ecet ten Surf ace l ocat ten R.G. 1.99/ 2 Lower Morale Etelt ARY 059
$A500. Cl.2 0.08 0.72 5.98t 18 3.32t 18 t i lli
113 116 N/A N/A 532 N/A tipper Shell C2789-1 5 4513.Gr.91 0.09 0.57 7.94l18 4.4 9t l8 (92t 76 78 N/A N/A Sl2 N/A Upper Shell C2 789-2 5As t ).Gr.8 L 0.09 0.57 7.98116 4.4 ?E 18 (92) 76 78 N/A II/ A
- 12 N/A I
Lower Shell C190 F-1 5 As l 3.Gr. Bt 0.12 0.55 7.98E tt 4.e lE 18 (921 'I 74 76 N/A N/A 792 N/A lower Shell C3251-1 5A51).Gr.81 0.11 0.50 F.98t !8 4.4)(18
( 92 ) '
F5 F'
N/A N/A
- 12 N/A I
Upper Circum. Weld (1001)
WF - 70 A5A/Linde 80
- 0. 3 %
0.59 5.98tle 3.12E 18
( FO s 41 45 53 54 4
>32 Niddle Circue. Weld 41005)
WT-25 ASA/Llade 80 0.15 0.68 F.9a( 18
- 4. 4 ]E 18
( 70 )I 41 43 53 55 4
732 Lower Ctecum. Wold II.D. 505 t 87 -6 7 ASA/t tade 80 0.24 0.60 4.42116 2.45(16 (70)I#3 N/A N/A N/ A N/ A N/A N/A Lower Ctecum. Weld (0.D. 5081 uf-70 ASA/Linde 80 0.35 0.59 1/A N/A N/A N/A N/ A WA GU A N/A N/ A Upper lengst. Wold (1005)
WF -0 ASA/Linde 80 0.29 0.55 7.98E 18 4.4 9t te (70)'#I of 46 55 56 7
>l2 L ower L ong s t. Wol d EO.D. 615) 54-1494 ASA/Llade 80 0.18 0.63 N/A N/ A N/ A N/A N/ A 01/ 4 N/ A N/A N/ A L a.or longi t. Weld
$4-1526 ASA/Linde 80 0.35 0.68
- 7. 48f 18 4.15f te 170 ) '
41 44 54 55 4
>32 (I.D. I T1. loot)
Upper Circus, weld (tons)
Atypecal ASA/Llade 80 0.41 0.10 5.98118 3.32118 79 48 51 63 64 32
- 32 Notes: (a) Per Draf t Regulatory Guide 1.99. Revisten 2. Deted f ebruary 10.1,a6.II W
(b) Per bas-leol. Deted Jameery 1984.II e.-*
(c) ist tested Noen Value Per BNf-10046P.II (d) Noen Valve Per BAW-1 pal.II N/ A = lept Appittehle N
- a. k EWn
^
On 2K bW
=o n
2=
EO M
I I
8.
SUMMARY
OF RESULTS I
The analysis of the reactor vessel material contained in the second I-surveillance capsule, TMIl-C, removed for evaluation as part of the Three Mile Island Unit 1 Reactor Vessel Surveillance Program, led to the fcllowing conclusions:
18 2
1.
The capsule received an average fast fluence of 8.66 X 10 n/cm (E > 1.0 MeV).
The predicted fast fluence for the regtor vgssel T/4 location at the end of the fifth fuel cycle is 1.06 X 10 n/cm (E > 1 MeV).
18 2.
The fast fluence of 8.66 X 10 n/cm2 (E > 1 MeV) increased the RT Of NDT the capsule reactor vessel core region shell materials a maximum of 203F.
3.
Based on the calculated fast flux at the vessel wall, an 80% capacity I
factor and the planned fuel management, the projected fast fluence that Three Mile Island Unit I reactor press r v sg,el inside surface will 8
receive in 32 EFPY operation is 7.98 X 10 n/cm' (E > 1 MeV).
4.
The increase in the RT for the shell plate material was not in good agreement with that p icted by the currently used design curves of RT versus fluence (i.e., R.G. 1.99/Rev. 2).
NDT 5.
The increase in the RT f ar the weld metal was in good agreement with NDT that predicted by the currently used design curves of RT versus I
fluence (i.e.,
R.G. 1.99/Rev. 2) and the prediction techhTques are conservative.
I 6.
The current techniques (i.e.,
R.G. 1.99/Rev. 2) used to predict the change in weld metal Charpy upper shelf properties due to irradiation are conservative.
7.
The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.
8.
The capsule design operating temperature may have been exceeded during operating transients but it is believed not for times and temperatures that would make the capsule data unuseable.
I I
8-1 Babcock & Wilcox a McDermott company
[
[
[
9.
SUF.VEILLANCE CAPSULE REMOVAL SCHEDULE F
Based on the postirradiation test results of capsule TMIl-C, the f oll ow *,g schedule is recommended for examination of the remaining capsules in the Three Mile Island Unit 1 RVSP:
Evaluation Schedule (a)
I Est. Vessel Est. Capsule Fgence,2 Est. Date Capsule guencey 10 n/cm Data ID
_10 n/cm Surface 1/4T Available(b)
TMIl-B(
0.44 0.23 0.13 1988 TMIl-D(c) 1.30 0.34 0.19 1995 TMIl-F 1.30 0.38 0.21 1997 TMIl-A Located in TMI-2 holder tube ZY-condition unknown and status is not defined.
("'In accordance with BAW-10006A and ASTM E 185-82 as I
modi,fied by BAW-1543A, Rev. 2.
(b) Estimated date based on 0.8 plant capacity factor.
(c) Capsules designated as standbys and may not be evaluated when removed.
I I
I I
9-1 Babcock & Wilcox a McDermott company
I
- 10. CERTIFICATION I
The specirnens were tested, and the data obtained from Three Mile Island Unit I reactor vessel surveillance capsule TMIl-C were evaluated using accepted techniques and established standard rnethods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H.
I di&" "
aWidu
- 5. 'L. Lowe, J r., ME.
/ / Date
~
Project Technical Manager This report has been reviewed for technical content and accuracy.
I y!WG L. $. Gross, P.E.
/
' Date Chemistry, Materials,
& Structural Analysis I
I 3
10-1 Babcock & Wilcox t 5
' "c o"=o (o=9'ay
I i
f l
r APPENDIX A Reactor Yessel Surveillance Program Background Data and Information l
l i
l I
l A-1 Babcock & Wilcox a McDermott company
I 1.
Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with BAW-10006A, are shown in Table A-1.
The locations 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 Three Mile Island Unit I was defined in accordance with the data given in BAW-10006A.
3.
Capsule Identification The capsules used in the Three Mile Island Unit 1 surveillance program are identified belcw by identification number and type (Ref. 3).
Capsule Cross Reference Data Number Type TMIl-A I
TMIl-B II TMIl-C I
TMIl-D II TMIl-E I
TNIl-F II 4.
Specimens for Surveillance Capsule See Tables A-2, A-3 and A-4 I
I A-2 Babcock &Wilcox J McDermott company
7-(--
p Table A-1.
Unirradiated Impact Proporties and Residual Element Content Data of Boltlino Region Materials'Used for Selection of Surveillance Program Materials -- Threo Mile Island Unit 1 Charpy data, Cg Tr msve r me I"*
RT Onemistry. *
- ",**I*I w Aght 50 ft-lb, 35 M1.E.
- USE, NDT*
MDT'h ID heat reston T
l_on g ft-lb e 10F F
F fr-lb F
Cu P
1
_No.
Storial tyg location 101, 109, 117 (75)
(60) *I 0.08 0.006 0.005 0.72 ARY-59 SA508, C1 2 Nossle belt C-2789-1 SA302 Cr B NJ Upper shell 0
50, %, 13 (75)
(40) 0.09 0.010 0.017 0.57 C-2789-2 ' SA302 GR 3 md Upper shall 20 42, 37, 35(20) 90 72 98 30 0.09 0.010 0.017 0.57 C-3 30 7-1I ' SA302 Cr 8 24 Imer shell
-10 42, 41, 29(30) 80 60
!!2 20 0.12 0.010 0.016 0.55 C-3251-1 SA302 Cr B Nd lower shell
-10 43, 40, 29 (75)
(40) 0.11 0.012 0.013 0.50 VF-70 Weld Upper circ.
39, 35, 44 (66)
(20) 0.27 0.014 0.011 0.46 l
(1001)
WF-8 Weld Upper tons.
45, 38, 30 (66)
(20) 0.20 0.009 0.009 0.61 (100%/1001)
WF-25(*
Weld Mid-circ.
-20 38, 28, 49 46 0
81
-14
- 0. 34 0.015 0.013 0.71 7
(2001)
SA-1494 Wald Lower long.
54, 25, 44 (66)
(20) 0.14 0.015 0.012 0.45 (631)
SA-1526 Wald lower long.
33, 33, 33 (66)
(20) 0.36 0.016 0.012 0.60 (1003/371) j l
39, 35, 44 (66)
(20) 0.27 0.014 0.011 0.46 WF-70 Weld lower cire.
(50%)
WF-67 Weld lower circ.
29, 15, 30 (66)
(20) 0.27 0.014 0.017 0.57 (503)
I* Surveillance base metal A.
E Surveillance base metal B.
6 k I*I
${
Surve111ance weld metal.
?R (d)rro. mill,nd 9ealtric.tio. ie.t reporta.
- Items in () estimated per BAW-10046P.
sa 2=
anO M
l Table A-2.
Test Specimens for Determining Material Baseline Properties No. of Test Specimens Tension.
g Material Description 70F 600F'd>
CVN Impact g
Heat CC (Heat No. 2789-2)
Base Metal Transverse Direction 3
3 15 Longitudinal Direction 3
3 15 Heat-Affected Zone (HAZ)
Transverse Direction 3
3 15 Longitudinal Direction 3
3 15 Total 12 12 60 Heat DD (Heat No. 3307-1)
Base Metal Transverse Direction 3
3 15 Longitudinal Direction 3
3 15 Heat-Affected Zone (HAZ)
Transverse Direction 3
3 15 Longitudinal Direction 3
3 15 Total 12 12 60 Weld Metal (WF-25)
Longitudinal Direction 3
3 15
- Test temperature to be the same as irradiation temperature.
I I
A-4 Babcock & Wilcox a McDermott company
L FL Table A-3.
Specimens in Upper Surveillance Capsules
{
(Designations A, C, and E)
No. of Test Specimens Material Description Tension CVN Impact Weld Metal, WF-25 4
8 Heat-Affected Zone Heat C2789-2, Longitudinal 8
l Base Metal Heat C2789-2, Longitudinal 4
8 Transverse 4
Correlation Material, HSST PLO2 8
l Total Per Capsule 8
36 l
Table A-4.
Specimens in Lower Surveillance Capsules (Designations B, D, and F)
I No. of Test Specimens Material Description Tension CVN Impact g
Heat-Affected Zone g
Heat C2789-2, Longitudinal 4
10 Base Metal Heat C3307-1, Longitudinal 4
10 Transverse 8
Correlation Material, HSST PLO2 8
Total Per Capsule 8
36 I
I I
I A-5 Babcock & Wilcox I
J MCOffmott (OmpJriy
I Figure A-1.
Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel I
I I
L
(
I
.r r
ARY 059 Nozzle Belt 4-=--W F-7 0 I
C-2789-1 Upper Shell C-2789-2
+ WF-8 4=*- WF-2 5 SA 1494, 63%
C-3307-1 Shell
- -3251-1 SA 1526, 37%
C
_ G F-70. 50' E
(WF-67, 50%
B 122T229VA1 Dutchman I
I I
A-6 Babcock & Wilcox a McDermott company
I J.
Figure A-2.
Location of Longitudinal Welds in Upper and Lower Shell Courses W
I I
T11 Z
4 X
11 s
I Y
Upper Shell I
W I
I I
/=
a\\nAxis(
z x
O On Axis (
I I
Y Lower Shell I
I A-7 Babcock & Wilcox I
a McDermott company
...]
i I
I I
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APPENDIX B I
Pre-Irradiation Tensilo Data I
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I I
I I
I B-1 Babcock & Wilcox a McDermott company
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Table B-1.
Pre-Irradiation Tensile Properties of Shell Plate Material, Heat C2789-2 I
Strength, psi Elongation, %
Red'n.
Specimen Test Temp, of Area, No.
F Yield Ultimate Uniform Total CC-703 RT 70,030 93,910 13.6 26.4 67.3 E
CC-705 RT 69,930 94,810 10.41 27.1 68.3 5
CC-708 RT 73,430 93,210 11.99 26.4 68.3 Mean RT 71,130 93,980 12.0 26.6 67.97 1,630 655 1.30 0.33 0.47 Std. Dev'n.
CC-702 570 63,500 92,000 12.79 25.7 61.3 E
CC-712 570 62,250 91,000 12.48 26.4 60.6 3
CC-724 570 62,440 92,080 12.71 25.0 54.7 Mean 570 62,730 91,690 12.66 25.7 58.87 550 490 0.13 0.57 2.96 Std. Dev'n.
I I
Table B-2.
Pre-IrradiationJensilePropertiesof Wald Metal WF-25 I
Specimen Test Temp, Elongation, %
Red'n.
E Strength, psi of Area, 5
i No.
F Yield Ultimate Uniform Total CC-102 RT 69,240 85,300 11.52 25.7 64.9 l
CC-121 RT 67,680 84,920 13.15 28.6 62.0 CC-126 RT 70,780 88,350 13.22 25.7 61.6 Mean RT 69,230 86,190 12.6 26.7 62.8 l
Std. Dev'n.
1,270 1,535 0.79 1.37 1.47 CC-109 570 67,170 82,320 8.88 20.0 52.7 CC-116 570 61,620 80,660 10.33 21.4 53.5 CC-119 570 64,130 82,160 9.28 20.0 50.8 l
Mean 570 64,310 82,050 9.5 20.5 52.3 l
Std. Dev'n.
2,270 1,090 0.61 0.66 1.13 I
I l
B-2 Babcock & Wilcox J McDermott company L
IL b
r a
i E
r L
s i
EL I
m L
APPEt4 DIX C Pre-Irradiation Charpy Impac-Data E
E r
l c-1 sabcock a wiscox a McDermott company
I I
Table C-1.
Pre-Irradiation Charpy Impact Data for Shell Plage l
Material -- Longitudinal Direction, Heat C2789-2 Il Test Absorbed Lateral Shear Specimen Temp.,
- Energy, Expagston,
- Fracture, No.
F ft-lb 10~
in.
CC-719 361 1 23 71 100 716 361 127 71 100 El 742 358 121 67 100 g) i CC-764 281 122 73 100 l
755 280 125 70 100 753 278 122 72 100 CC-738 203 134 69 100 E
73 2 200 13 2 68 100 3
722 200 128 69 100 CC-761 139 120 69 90 760 139 116 69 88 757 138 133 72 100 CC-727 104 93 64 50 712 104 92 57 65 CC-741 80 73 58 12 745 80 96 61 10 703 80 74 52 8
I CC-721 41 56 45 14 724 41 68 51 25 715 41 66 52 18 CC-756 26 54 43 3
754 26 50 40 4
759 25 42 35 2
CC-717 1
37 29 1
726 1
45 34 2
714 0
36 25
<1 CC-763
-56 14 10 0
g 762
-56 10 9
0 m
758
-56 30 22 1
I I
C-2 Babcock & Wilcox a McDermott company
L E
Table C-2.
Pre-Irradiation Charpy Impact Data for Shell Plate Material, Transverse Orientation, Heat C2789-21 Test Absorbed Lateral Shear Specirnen Temp.,
- Energy, Expagston,
- Fracture, No.
F ft-lb 10-in.
[
CC-629 360 92 71 100 634 360 91 68 100 617 358 93 76 100 CC-638 280 101 63 100 640 280 98 67 100 647 280 90 66 100 CC-607 201 105 67 100 625 199 90 61 100 612 198 101 64 100 CC-646 160 98 61 100
{
644 160 94 67 100 641 159 92 61 94 f
CC-645 130 88 61 85 L
643 129 76 58 60 639 129 90 62 82
[
CC-648 101 57 47 35 642 101 68 51 40 i
637 101 75 53 30 CC-635 80 68 52 15 615 80 58 46 12 602 80 46 37 4
CC-633 41 38 32 2
619 40 33 29 2
{
601 40 32 28 3
CC-636 2
41 31 1
605 1
16 15 0
609 1
25 20 1
l C-3 Babcock & Wilcox a McDermort comparty
I I
Table C-3.
Pre-Irradiation Charpy Impact Data for Shell Plate Material, HAZ, Transverse Orientation. Heat C2789-21 I
Test Absorbed Lateral Shear E
Specimen Temp.,
- Energy, Expagston,
- Fracture, 5
No.
F ft-lb 10-in.
CC-301 362 146 72 100 326 360 105 57 100 303 360 31 49 100 CC-327 201 91 50 100 325 201 13 7 70 100 324 200 89 46 100 CC-340 140 97 59 95 336 140 122 72 100 334 138 116 64 100 CC-316 81 122 61 100 321 80 114 54 55 E
302 80 123 64 85 E
306 80 120 61 85 CC-338 41 88 55 65 331 41 62 38 35 329 40 121 62 75 CC-307 2
60 39 15 322 1
56 35 6
311 1
65 39 12 CC-332
-20 60 44 6
339
-20 76 45 15 337
-20 112 61 60 CC-323
-39 80 46 45 308
-39 52 35 3
304
-40 54 33 10 CC-335
-80 36 22 2
g 333
-80 26 19 3
3 330
-80 42 32 5
I I:
.I:
C-4 Babcock & Wilcox a McDermott company
f L
Table C-4.
Pre-Irradiation Charpy Impact Data for Weld Metal, WF-251 f
Test Absorbed Lateral Shear Specimen Temp.,
- Energy, Expagston,
- Fracture, No.
F ft-lb 10-in.
5 CC-006 359 80 63 100 j
051 359 81 62 100 021 358 82 60 100 CC-034 200 82 59 100 032 199 78 61 100 001 198 84 55 100 CC-012 80 64 47 25 I
r 003 80 70 52 55 L
030 80 68 47 55 CC-024 61 58 50 55
{
009 61 62 45 75 CC-010 41 50 43 25 J
022 41 46 42 45 0 23 40 54 47 55 r-'
CC-018 2
42 36 30 On 1
40 36 10 007 1
40 35 8
020 1
42 36 10 E
E r
C5 Babcock & Wilcox i
4 M(Dermott company 1
I' lI Figure C-1.
Charpy Impact Data for Unirradiated Plate Material, Longitudinal Orientation, Heat C2789-2 100 i
i g
g i
[
i i
>< 75 J
5 0
50 - - - - - - - - - - - - -
m 2
Ji 25 I
0
^ '
/
.08
=.
g g
g g
g g
g g
g g
e i.06 2
i.04-9 5
=
j.02-
- 8 I
I I
3--
I I
I I
I I
I 0
200 i
i g
g g
g g
DATA
SUMMARY
Igow T,37 1or 14r Tgy (35 sg)
T (50 FT-u ) 26r 160 CV T, (30 rT-a) 7;r e
C -USE (Avs) 151 ft-h s y
= 140 gi 1er
-d NOT
$120-
=
8 3100-g j
8 5
5 80 -
l O
e 5
N 50-
--___--.2__.--._._________________-------
g_
e 8
MATERIAL tA'2?_Cr_!_Vad.
20-Ftutsct HEAT No.
C2739-2 I
I I
I I
I I
I 0
-00
-40 0
40 80 120 160 200 240 280 320 360 400 Test ItMeteATutt, F C-6 Babcock & Wilcox g
a McDermott company g
Figure C-2.
Charpy Impact Data for Unirradiated Plato f4aterial, Transverse Orientation, Heat C2789-2
^
i i
i 6
i 100 i
i 1
l l
i e
- 75 -
i.
.gg ___________
t 5 25 p
[u I
L t
1 I
I 1
I I
I O
(
. 08 g
1 1
i i
i I
4 e
i.06 2
e 2
e O.@-
x Gz 4=.02 e
I i
1 I
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I O
E i
i a
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DATA SUMFARY 130" T war T, (35 ate) 57F c
160- T (50 n-La)
- Cy 28F Tcy (30 FT-La)
C -USE (Avs) 99 f t-tu y
= 14C RT,
30F L
0 5 120 -
[
i
= 100 &
3
[
3 80-I e
z
[
~ g_
e
~
MArtRIAL 930? h.9.Wod OR i t NT Ai t om frans m se 20-FLutset 93"'
~
HEAT No.
C2787-2 I
I I
I I
I I
I 0
-C0
-40 0
40 80 120 160 200 240 280 320 360 400 Test TentRATunt, F C-7 Babcock & Wilcox a McDermott company
--- _ - _ - _ _ _ _ - _ - - - - _ _ _ _ _ _ - - _ - _ - _ - ~. - _ _. _ _. - - - _ - - - -. - _ _ - - - -. -
_ s, 4
g Figure C-3.
Charpy Impact for Unirradiated Plate Material, Heat-Affected Zone, Haat C2789-1 100 g
i 4 g
- j j
3 r
i I
" 75 I
e v* 50 I
5 5 25 -
I I
I I
I I
I I
I I
0-
.03 j
g g
g g
j g
g g
g g
1 Y
m.
\\,
a.0s n
e y
f ss
=
's T.00 -.
i-c I
n A
~
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1 I
I I
I I
I I
I g
I I
I I
I I
i DATA
SUMMARY
180b T,37 gn E
-40f T, (35 mLg) g c
-6v
_Tcy (5J FT-ts) 160
-77F Tcy (30 st-ta)
C -USE (Ave.)
132 f t-ns
- y
. 140 RT NA 7
NDT t
e 6 120-8 e 100 "
3 1
3 if 80-2
.a l
J l
t 60..
d E
t MATERIAL S A502.Gr.B.%d g,
L.
l 20 FLut=ct
- W
~
HEAT No.
C2781-2 f
I I
I I
I I
I 0-00
-40 0-00 80 120 160 200 240 280 320 360 400 Test TtretaArunt, F C-8 Babcock & Wilcox l
a McDermott company I
i
- I l
Figure C-4.
Charpy Impact Data for Unirradiated Wold Motal WF-25 100 l
l g
i j
i i
75 _
e l
I a
W* 50 i
,g 22, I
T l
l 1
1 1
1 1
I I
i 5.
I I
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I I
I I
I I
I i.Os-a I
~
f
" 04 3
L___
.I
\\
4 i.02-
=
l I
I I
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I I
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I O
100 i
i 4
g g
g
[
g g
i DATA SU W RY
-20F 90r T..,7 T (35 st)
-sr I
g.! Tcv (50 FT-La) 39F e
cs e
O Tcy (30 n-La)
C -USE (avs) 91 ft-Its I
y 70 RT
-20r e
nor C
e 1
L S,0 I
I 53 50 - - - - - - - -
I i
40l-w C
/
I L/_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
g x
20-I
".Atte r AL Veld "etal OnitNTAtton 10-TiurNCE D"'
~
Hur No.
Vf-25 I
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I I
I I
I 0-03
-40 0
40 80 120 160 200 240 280 320 360 400 Test Temagnatung, F I
C-9 Babcock &Wilcox I
a McDermott company
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I I
.I
.I I
I I
APPENDIX D Fluence Analysis Procedures I
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.I i
I l I I
ll D-1 Babcock & Wilcox g
a McDerrnott company
(
I 1.
Analytical Method The procedure employed in this analysis makes use of experience gained from previous analyses in two aspects:
the calculation of the capsule flux and the normalization factor.
Previous analyses calculated the capsule flux 20 using the DOT 4 code with an explicit model of the capsule.
This flux was then used 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 238 (n,f)l37Cs U
resulted in an average normalization factor of 0.95 from the 237Np(n,f)l37Cs reactions.
In addition, these analyses have produced and a consistent set of spectrum-averaged reaction cross sections as a function 0
0 of capsule position, i.e.,
11 or 27 az imuthally.
Based on the consistency of the normalization factor and the cross sections, an analysis procedure was developed.
Basically it involves calculating the capsule flux from equation D-6 using the measured activities and the spectrum-averaged cross sections.
A more detailed description of this procedure is given belcw.
1.1.
Capsule Flux and Fluence Calculation The DOT 4 code in previous analysis has been used to explicitly model the capsule assembly and 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 1, O, (pC1/gm) is:
g where:
N f
-A t
-A (T - td) i 9 = (3.7 x 10 )An I o" (E) : (E) r Fd (1-e I d) e i
D-1 D
4 E
j and Avogadro's number, N
=
atomic weight of target material n, An
=
either weight fraction of target isotope in n-th material or the f
=
g fission yield of the desired isotope, I
D-2 Babcock & Wilcox
- McDermott company
I I
n(E) group-averaged cross sections for material n
=
o l
(listed in Table E-3) group averaged fluxes calculated by DOT 4 analysis, 4(E)
=
fraction of full power during j-th time interval, tj l
F
=
j decay constant of the ith isotope, Ai
=
Sum of total irradiation time, i.e.,
residual time in reactor, T
=
l and the wait time between reactor shutdown and counting times Cumulative time from reactor startup to end of j-th time period.
j
=
A normalilzation factor was applied to the calculated flux obtained from the I
ratio of the measured to calculated activities, i.e.
D (measured) j D-2 C
=
j D (calculated) j l
Equation D-1 can be abbreviated as:
D =KP n (E) $
(E) 0-3 j
jIE where K contains the constant terms and P is the summation over j g
j representing 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:
Io (E) 4 (E) 0-4 n
r 9 (E)
Then, equation 0-3 becomes KPa $ (E > l MeV) 0-5 0
=
9 jj or, solving for the flux I
0i o(E > l MeV) =
0-6 Kpg iin D-3 Babcock & Wilcox I
a McDermott company
I Equation D-6 is used to calculate the capsule flux from the measured calculated from previous activities D and the average cross section, o n, E
j analyses (see Table D-1).
It should be noted that o is dependent upon the 5
n capsule azimuthal location since the calculated flux spectrum used to obtain 0
0 differs for the 11 and 27 locations due primarily to dif ferences in on distances from the core perimeter.
137 The measured activity (at least for the Cs activity) is representative of irradiations in all cycles previous to the measurement.
For Capsule TMIl-C this includes irradiation in cycle 1 of TMI-l and in cycles 3 through 5 of Crystal River Unit 3.
The irradiation history of the capsule is represented E
by the Pj term which accounts for production and decay of the product isotopo 5
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 P, the flux for the last cycle, or cycles, can be determined.
j Thus using Pj for TMI-1 cycle 1 and CR-3, 4 and 5, the average capsule flux can be determined.
1.2.
Vessel Flux and Fluence Energy-dependent neutron fluxes at the reactor vessel were determined by a discrete ordinates solution of the Boltzmann transport equation with the two-dimensional code DOT 4.
The TMI-l reactor was modeled from the core out to the primary concrete shield in R-0 geometry (based on a plan view along the core midplane and one-eighth core symmetry in the azimuthal dimension).
The reactor model contained in the f ollowing regions:
core, liner, bypass cool ant, core ba rrel, inlet cool ant, thermal shield, inlet cool ant (down-comer), pressure vessel, cavity, and concrete shield.
Input data to the codo include d a PDQ calculated pin-by-pin, time-averaged power distribution, CASK g
sections,21 3 rder of angular B
23E 22-group microscopic neu tron cross 8
quadrature, and P3 expansion of the scattering cross section matrix.
Reactor conditions, i.e. power distribution, temperature and pressure, were averaged g
over the irradiation period.
Two geometric models were used 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 (thermal shield to i
D-4 Sabcock & Wilcox 4 M(Octmott comparty
k I
primary shield).
In this way the effect of the specific power distribution b
of TMI-l Cycles 2-4 on vessel fluence was included in the calculation.
Flux output from the 00T4 calculations required only an axial distribution correction to provide absolute values.
An axial shape factor (local to average axi al flux ratio) was obtained from predicted fuel burnup distributions in the peripheral fuel assemblies nearest the vessel.
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 coro increases.
An axial factor of 1.17 was applied to the calculated flux for the TMI-l reactor vessel.
This axial
(
factor was time-averaged over the irradiation period.
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 equivalent DOT 4 models and constant terms in equation D-2 to obtain the calculated activity, a normalization factor of about 0.95 has been consist-ently 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 pressure vessel and capsule are separated by only 15 cm of water so that it is un-
{
likely that any significant change in accuracy would occur in this distance.
Due to the consistency of this factor at the capsule and the above considera-tions, it will be applied to the TMI-l Cycles 2-4 calculated vessel flux with 137 only a minor correction to account for DOT 4 model change and revised Cs yield values which have been implemented since the previous analyses.
The model change incorporated a more accurate water temperature in the core former region.
The previous model assumed this water region was at the
(
average core water temperature, recent information indicates an average of the core and downcomer temperature is more realistic. This decrease in water
{
temperature increases its density.
From DOT 4 studies, this density increase causes a reduction in the vessel flux by ~4%.
The second correction is due 137 to use of recently recommended ASTM Cs yieW values from the fission of
[
238 237 0 and Np.
Use of the recommended values reduce the calculated activity b
D-5 Babcock & WHcom a M(Dermott company
_______________.__________________._______j
I used to obtain the normalization factor by -5%.
Since both changes reduce the calculated activity, i.e. the denominator in equation D-2, the 0.95 normalization factor must be increased.
Thus the ef fective 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.
Yessel Fluence Extrapolation For up-to-date operation, fluence values in the pressure vessel are calcu-g lated as described above.
Extrapolation to future operation is required for a
prediction of vessel life based on minimum upper shel f energy and for calculation of pressure-temperature operation curves.
Three time periods are considered in the extrapolation:
- 1) to-date operation for which vessel fluence has been calculated,
- 2) designed future fuel cycles for which PDG calculations have been performed for fuel management analysis of reload cores, and 3) future cycles for which no analyses exist.
Data f rom time period 1 are extrapolated through time period 2 based on the premise that ex-core flux is proportional to fast flux that escapes the core boundary.
Thus for the vessel, W
$e,c
=
,V,c
$e.V I
where: v = vessel e = core escape r = reference cycle c = future fuel cycle Core escape flux is available from PDQ output.
Extrapolation from time period 2 through 3 is based on the last designed cycle in period 2 having the same rel ati ve power distribution as an " equilibrium" fuel cycle.
Generally, the designed fuel cycles include several cycles into the future, cycle 5 for TMI-1.
Data for TMI-l are listed in Table D-2.
This table indicates for cycles beyond 5 how the cycle flux and fluence were extrapo-D-6 Babcock & Wilcox a McDermott company
[
lated using factors to account for anticipated new fuel cycle designs leading to a very-low-leakage fuel cycle design by cycle 8 and beyond.
This procedure is considered preferable to the alternative of assuming that lifetime fluence is based on a single, " equilibrium" fuel cycle because this
(
procedure accounts for all known power distributions.
In addition, errors that may result from the selection of a " equilibrium" cycle are reduced.
[
F L
Table D-1.
Spectrum Averaged Cross Sections i
0 Reaction 11 Average Cross Section, (barns) 58Ni(n,f)58Co 0.1222 54Fe(n,1)S4Mn 0.09283 238 (n,f)137Cs 0.4037 U
l 23 7Np(n,f)137Cs 2.416 I
ll ll
!I I
I I
D-7 Babcock & Wilcox a M(Dermott company
Table D-2.
Extrapolation of Reactor Vessel Fluence 2
Vessel Fl uence, n/cm CoreEscapg
- Time, Cumulative Time, Vessel Flux, Time 2
Cycle Flux, n/cm s EFPY EFPY n/cm s Interval Cumul ative 1
0.418E+14 1.28 1.28 1.45E+10 5.90E+17 5.90E+17 2-4 0.569E+14 2.24 3.51 1.38E+10 9.72E+17 1.56E+18 5
0.578E+14 0.77 4.28 1.40E+10 3.38E+17 1.90E+18 6
0.82x.578E+14 1.13 5.41 1.15E+10 4.08E+17 2.31E+18 l
7 0.72x 578E+14 1.2 6.61 1.01E+10 3.81E+17 2.69E+18 i
8 0.656x.416E+14 1.2 7.81 0.66E+10 2.50E+17 2.94E+18
- 0. 656x.416E+ 14 2.19 10 0.66E+10 4.56E+17 3.40E+18 0.656x.416E+14 5
15 0.66E+10 1.04E+18 4.44E+18
?
0.656x.416E+14 6
21 0.66E+10 1.25E+18 5.69E+18 l
0.656x.416E+14 11 32 0.66E+10 2.29E+18 7.98E+18 l
W
- i. h b$
?R
!=
ED BE=
Ail
\\
l
=
l M
M M
M M
M M
M M
M M
M M
M M
M M
M M
rL J
L FL IL F
L r
L EL e
L APPENDIX E Capsulo Dosimetry Data E
E E
E-1 Babcock & Wilcox a McDermott company
Table E-1 lists the characteristics of the threshold detectors.
Table E-2 shows Capsule TMIl-C measured activity per gram of target material (i.e., per E
gram of uranium, nickel, etc. ).
Activation cross sections for the various 5
235 materials were flux-weighted with a U fission spectrum (Table E-3).
I Table E-1.
Detector Composition and Shielding Detector Material _
% Target Shielding Reaction 238 (n,f)137Cs 238 U
U-Al 10.38%
U Cd-Ag Np-Al 1.44%
Np Cd-Ag Np(n,f)137Cs 237 237 58 58Ni(n,p)58Co Ni 67.77%
Hi Cd-Ag 59 59 Co-Al 0.66%
Co Cd Co(n,Y) Co 59 59 Co-Al 0.665 Co None Co(n,Y)
',o Fe 5.82% Fe None Fe(n,p)5'An 54 54 Table E-2.
Capsule TMIl-C Dosimeter Specific Activities Dosimeter Activity, (u C1/gm of Target) g Detector Dosimeter 5
Material Reaction 001 DD2 DD3 DD4 58Hi(n,p)00Co 1299.7 1349.6 917.9 1591.2 Ni 54Fe(n,p)54Mn 944.6 969.7 669.1 1137.3 Fe 238 (n,f)137Cs 8.960 9.290 6.786 11.90 U-Al U
Np-Al Np(n,f)137 237 Cs 51.46 52.65 34.61 72.42 I
I I
I I
I E-2 Babcock & Wilcox a McDermott company
I TABLE E-3 00SIMETER ACTIVATION CROSS SECTIONS,(a) b/ atom I
Energy Range 237 238 (n,f) 58 54 G
MeV Np(n,f)
U Ni(n,p)
Fe(n p) 1 12.2 - 15 2.3 23 1.050 4.830E-1 4.133E-1 2
10.0 - 12.2 2.341 9.851E-1 5.735E-1 4.728E-1 3
8.18 - 10.0 2.309 9.935E-1 5.981E-1 4.772E-1 4
6.36 - 8.18 2.093 9.110E-1 5.921E-1 4.714 E-1 5
4.96 - 6.36 1.541 5.777E-1 5.223 E-1 4.3 21E-1 6
4.06 - 4.96 1.532 5.454E-1 4.146E-1 3.275E-1 7
3.01 - 4.06 1.614 5.340E-1 2.701E-1 2.193E-1 8
2.46 - 3.01 1.689 5.272E-1 1.445E-1 1.080E-1 9
2.35 - 2.46 1.695 5.298E-1 9.154E-2 5.613 E-2 I
10 1.83 - 2.35 1.677 5.313E-1 4.856E-2 2.940E-2 11 1.11 - 1.83 1.596 2.608 E-1 1.180E-2 2.948E-3 12 0.55 - 1.11 1.241 9.845E-3 6.770E-4 6.999E-5 13 0.111 - 0.55 2.34 E-1 2.432E-4 1.174E-6 1.578E-8 14 0.0033 - 0.111 6.928E-3 3.616E-5 1.023 E-7 1.389E-9 (a) ENDF/B5 values that have been flux weighted (over CASK onergy groups) 235 based on a U fission spectrum in the fast energy range plus a 1/E shape in the intermediate energy range.
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APPENDIX F References I
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1 REFERENCES 1.
A. L. Lowe, Jr., et al., Analyses of Capsule TMI-1E Metropolitan Edison Company, Three Mile Island Nu clear Station Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1439. Babcock & Wilcox, Lynchburg, Virginia, January 1977.
I 2.
G. J. Snyder and G. S. Carter, Reactor Vessel Material Surveillane Program, Revision 3, BAW-10006 A, Rev. 3, Babcock & Wilcox, Lynchburg, Virginia, January 1975.
3.
A. L. Lowe, Jr., et al., Integrated Reactor Vessel Material Surveillance Program, B AW-1543 A, Rev. 2 Babcock & Wilcox, Lynchburg, Virginia, May 1985.
4.
Code of Federal Regul a t io n, Title 10, Part 50, Fracture Toughness Requirements for Light-Water Nuclear Power Reactors, Appendix G, Fracture Toughness Requirements, Federal Register, Vol. 48, No.104, May 17, 1983.
5.
Code of Federal Regulation, Title 10, Part 50, Fracture Toughness Requirements for Light-Water Nuclear Power Reactors, Appendix H, Reactor Vessel Material Surveillance Program Requirements, Federal Register, a
Vol. 48, No.104, May 27,1983.
I 6.
K. E. Moore and A. S. Heller, Chemistry of 177-FA B&W Owners Group Reactor Vessel Beltline Welds, _BAW-1500P, Babcock & Wilcox, Lynchburg, Virginia, September 1978.
7.
J. D. Aadiand, Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor Vessel and S u rveil l ance Program Materials Information, BAW-1820, Babcock & Wilcox, Lynchburg, Virginia, December 1984.
m 8.
Heavy Section Steel Technology Program, Semlannual Progress Report for Period Ending Federal 28, 1969, ORNL-4463, Oak Ridge National Laboratory, Oak Ridge, Tennessee, January 1970.
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9.
American Society for Testing and Materials, Methods and Definitions for 7
Mechanical Testing of Steel Products, A370-77, June 24,1977.
[
- 10. American Society for Testing and Materials, Methods for Notched Bar Impact Testing of Metallic Materials, E23-82, March 5,1982.
Iu 11.
U.S.
Nuclear Regulatory Commission, Effect of Residual Elements on
[
Predicted Radiatico Damage to Reactor Vessels, Regulatory Guide 1.99, Revision 1, April 1977.
I' 12.
U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Draf t Regulatory Guide 1.99, Revision 2, February 10, 1986.
13.
A.
S.
Heller and A.
L.
Lowe, J r., Correl ations for Predicting the
(
Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds, BAW-1803, Babcock & Wilcox, Lynchburg, Virginia, January 1984.
14.
U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Section p
61,,
" Fracture Toughness Requirements for Protection Against Pressurized L
Thermal Shock Events," First Published - Federal Register, Vol. 50, No.
141, July 23,1985.
15.
H. S. Pal me, H. W. Behnke, and W. J. Keyworth, Methods of Compliance
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With Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G, BAW-10046P, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia,
{
March 1976, 16.
H. S. Palme, et al., Methods of Compliance With Fracture Toughness and Operational Requirements of Appendix G to 10CFR50, BAW-10046A, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, July 1977.
17.
A.
L.
Lowe, J r., et al.,
Pressurized Thermal Shock Evaluations in
(
Accordance With 10CFR50.61 for Babcock & Wilcox Owners Group Reactor Pressure Vessesi, BAW-1895, Babcock & Wilcox, Lynchburg, Virginia,
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Jc.nuary 1986.
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F-3 Habcock & Wilcox J McDermott company
18.
K. E. Moore and A.
S. Heller, BAW 177-FA Reactor Vessel Beltline Weld Chanistry Study, BAW-1799, Babcock & Wilcox, Lynchburg, Virginia, July 1983.
19.
K. E. Moore, et al., Evaluation of the Atypical Weldment, BAW-10144A, Babcock & Wilcox, Lynchburg, Virginia, February 1980.
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- 20. B&W's Version of "DOTIV Version 4.3, One-and Two-Dimensional Transport Code System," Oak Ridge National Laboratory, Distributed by the Radiation Shielding Information Center as CC = 429, November 1,1983.
21.
"C ASK-40-Group Coupled Neutron and Gamma - Ray Cross Section Data,"
Radiation Shielding Information Center, DLC-23E.
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F-4 Babcock & Wilcox a McDermott company
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