Text

Analysis of Capsule TMI-1E from Met Ed TMI-1:Reactor Vessel Matls Surveillance Program, Jan 1977.Prepared for Met Ed
	ML19260A122
Person / Time
Site:	Crane
Issue date:	01/31/1977
From:	Chulick E, Lowe A, Palme H; BABCOCK & WILCOX CO.
To:	;
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ML19260A119	List: ML19260A118; ML19260A121; ML19260A122;
References
	595-7029-43, BAW-1439, NUDOCS 7910290739
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{{#Wiki_filter:.. _. _ _... _ _. _ _ BAW-1439 January 1977 l in [R ANALYSIS OF CAPS 1'LE TMI-lE FROM METROPOLITAN EDISON COMPANY THREE MILE ISLAND NUCLEAR STATION-ilNIT 1

i

- Reactor Vessel Materials Surveillance Program - h by A. L. Lowe, Jr., PE E. T. Chulick H. S. Palme C. L. Whitmarsh 7 C. F. Zurlippe 5 A l l l 1 B&W Contract No. 595-7029-43 ) l P I BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260 <ly txecssers, vir81 1-asosy91o290 3g Babcock & Wilcox i 1480 040

l J CONTENTS 1 1 j Page 1 1. INTRODUCTION. 1-1 h 2. BACKGROUND. 2-1 3. SURVEILLANCE PROGRAM DESCRIPTION.. 3-1 0 4. PREIRRADIATION TESTS. 4-1 yj 4.1. Tensile Tests 4-1 4.2. Impact Tests. 4-1 5. POSTIRRADIATION TESTS 5-1 5.1. Thermal Monitors. 5-1 5.2. Chemical Analysis 5-1 5.3. Tensile Test Results. 5-2 5.4. Charpy V-Notch Impact Test Results. 5-2 e I 6. NEUTRON DOSIMETRY 6-1 6.1. Introduction. 6-1 6.2. Analytical Approach 6-2 y 6.3. Results 6-3 h 7. DISCUSSION OF CAPSULE RESULTS 7-1 r [ 7.1. Preirradiation Property Data. 7-1 ( 7.2. Irradiated Property Data. 7-1 r 7.2.1. Tensile Properties. 7-1 7.2.2. Impact Properties 7-1 d 8. DE'IT.RMINATION OF RCPB PRES 3URE-TI>tPTRATURE LIMi. 8-1 a Ia ( 9.

SUMMARY

OF RESULTS.. 9-1 ? h 10. SURVEILLANCE CAPSULE RD10 VAL SCHEDULE 10-1 0 11. CERTIFICATION 11-1 P 12. REFERENCES. 12-1 b 03 1480 041 ~ 'u - lii - Babcock 8.Wilcox a

Contents (Con t 'd ) Page APPENDIXES A. Reactor Vessel Surveillance Program -- Background Data and Information. A-1 B. Preirradiation Tensile Data Three Mile Island Unit 1.......................... B-1 C. Preirradiation Charpy Impact Data Three Mile Island Unit 1......... C-1 D. Threahold Detector Information. D-1 List of__Ta_b_1g Table 3-1. Specimens in Surveillance Capsule TMI-1E 3-2 ..s 3-2. Chemistry and Heat Treatment of Surveillance Materials 3-3 3-3. Chemistry and Heat Treatment of Correlation Material -- Heat A-1195-1, 'A533 Grade B, Class 1................ 3-4 5-1. Chemistry Data on Unitradiated TMI-1 RVSP Material 5-2 5-2. Tensile Properties of Capsule TMI-JE Base Metal and Weld Metal Irradiated to 1.07 x 10'B n/cm2............. 5-3 5-3. Charpy Impact Data for Capsule TMI-1E Base Metal Irradiated to 1.07 x 3010 n/cm2 5-4 5-4. Charpy Impact Data for Capsule idl-lE Weld Metal (WF-25) Irradiacci to 1.07 x 1018 n/cm2 5-5 5-5. Charpy Impact Data for Capsule TMI-lE Correlation Monitor Material Irradiated to 1.07 x 1018 n/cm2 (E > 1 MeV), Heat A-1195-1, Transverse Orientation 5-5 6-1. Surveillance Capsule Detectors 6-3 6-2. Flux Adjustment Factor 6--5 6-3. Dosimeter Activations 6-6 6-4. Normalized Flux Spectrn, E > 1 MeV 6-6 6-5. Neutron Fluence. 6-7 6-6. Predicted Fast Neutron Fluence in Pressure Vessel for 10 EFPY. 6-7 7-]. Comparison of Tensile Test Results 7-3 7-2. Observed Vs Predictad Cha c-in Irradjared ' harpy L pact Properties 7-4 8-1. Data for Preparation of Pressure-Temperature Limit Curves - Applicable Through Sixth Full-Powcr Year 8-4 A-3. Material Selection Data for Thl-1 Survel. lance Project A-3 A-2. Materials and Specimens in Upper Surveil..ance Capsules TMI-1A, TMI-lC, and TMI-]E 1-4 A-3. Materials and Specimens in Lower Surveillance Capsules TMI-1B, TMI-1D, and TMI-1F A-4 B-1. Preirradiation Tensile Properties of Shell Plate Material, Heat C-2789-2.. E-2 B-2. Preirradiation Tensile Properties of Shell Plate Material, RAZ, Heat C-2789-2 C-3 ,/g Q G GA^ 1 - iv - Babcock & \\Vilcox

i h its l Tables (Cont'd) Table Page ? B-3. Preirradiation Tensile Properties of Shell Plate Material, Heat C-3307-1............. B-4 B-4. Preirradiation Tensile Properties of Shell Plate Material, l HAZ, Heat C-3307-1 B-5 ,j B-5. Preirradiation Tensile Properties of Weld Metal, WF-25, [ Longitudinal B-6 C-1. Prairradiation Charpy Impact Data for Shell Plate Material, Longitudinal Orientation, Heat C-2789-2... C-2 C-2. Preirradiation Charpy Impact Data for ShelJ Plate Material, Transverse Orientation, Heat C-2789-2.... C-3 C-3. Preirradiation Charpy Impact Data fcr Shell Plate Material, HAZ, Longitudinal Orientation, Heat C-2789-2 C-4 C-4. Preirradiation Charpy Impact Data for Shell Plate Material, HAZ, Transverse Orientation, Heat C-2789-2 C-5 C-5. Preirradiation Charpy Impact Data for Shell Plate Material, Longitudinal Orientation, Heat C-3307-1.. C-6 C-6. Preirradiation Charpy impact Data for Shell Plate Material, Transverse Orientation, Heat C-3307-1 C-7 C-7. Preirradiation Charpy Impcet Data for Shell Plate Material, HAZ, Longitudinal Orientation, Heat C-3307-1 y C-8. C-8 Preirradiation Charpy Impact Data for Shell Plate Material, HAZ, Transverse Orientation, Heat C-3307-1 C9 C-9. Preirradiation Charpy impact Data for Weld Metal WF-25, Transverse Orientation C-10 ~ D-1. Detector Composition and Shielding D-2 D-2. TMI-1, Cycle 1 Neutron Dosi neters. D-3 J J List of Figures Figure 3-1. keactor Vessel Cross Section Showing Surveillance Capsule Locations 3-5 5-1. Impact Data From Irradiated Base Metal A, Longitudinal 1 Orientation i 5 2. 5-6 Impact Data From Irradiated Base Metal A, Transverse k Orientation 5-7 5-3. Impact Data From Irradiated Base Metal A, HAZ, Longitudinal i Orientation 5-8 5-4. Impact Data From Irradiated Weld Metal, Transverse ) Orientation 5-9 p 5-5. Impact Data From Correlation Monitor Material, Transverse Orientation 5-10 6-1. Predicted Fast Neutron Fluences at Various Locations Through Reactor Vessel Wall for First 10 EFPY 6-8 5 1 Of5 Babcock & Wilcox v-s

a h I I L Contents (Cont'd) } a Figure Page { l 7-1. Irradiated Vs Unirradiated Charpy Impact Properties of f i Base Metal, Longitudinal Orientation. 7-5 g I 7-2. Irradiated Vs Unirradiated Charpy Impact Properties of 2 l Base Metal Transverse Orientation 7-6 7-3. Irradiated Vs Unirradiated Charpy Impact Properties of Base Metal, HAZ 7-7 7-4. Irradiated Vs Unirradiated Charpy Impact Properties of [ Weld Metal, Transverse Orientation. 7-8 7-5. Irradiated Vs Unirradiated Impact Properties of Correlation Monitor Material, Longitudinal Orientation. 7-9 8-1. Fast Neutron Fluence of Surveillance Capsule Center Compared to Various locations Through Reactor Vesse) Wall for Firnt 10 EFPY 8-5 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Heatup, Applicable for First Six Effective Full-4 Power Years 8-6 ) 8-3. Reactor Vessel Pressure-Temperature Limit Curve for Normal { l Operation - Cooldown, Applicable f or First Six Effective J l Full-Power Years. 8-7 [ 8-4. Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Hydrostatic Tests, Applicable for First Six Effective Full-Power Years.. 8-8 l A-1. Location and Identifica: ion of Materials Used in Fabrication i l of TMI-1 Reactor Pressure Vessel. A-5 f A-2. Location of Longitudinal Welds in Upper t.nd Lower Shell g Courses A-6 C-1. Impact Data From Unitradiated Base Metal A, Longitudinal Orientation C-11 C-2. Impact Data From Unirradiated Base Metal A, Transverse Orientation C-12 I' C-3. Impact Data From Unitradiated Base Metal A, HAZ, Longitudinal Orienta ti on C-13 C-4. Impact Data From Unirradiated Base Metal A, HA2, Transverse Orientation C-14 C-5. Impact Data From Unirradiated Base Metal B, Longitudinal ] Orientation C-15 'i C-6. Impact Data From Unirradiated Base Metal B, Transverse Orientation C-16 g C-- 7. Impact Data From linirradiated kce Metal B, M2, Longitudinal Orientation C-17 C-8. Impact Data From Unirradiated Base Metal B, HAZ, Transverse Orientation C-18 C-9. Impact Data From Unirradiated Weld Metal, Transverse Orientation C-19 i l 1480 044 14WMDM i k i _ y1 _ Babcock & Wilcox } ]

k i 1. INTRODUCTION This report describes the results of the examination of the first capsule of the Met.ropolitan Edison Company's Three Mile Island Nuclear Station, Unit 1 (TMI-1) reactor vessel surveillance program. The objective of the program is i to monitor the effects of neutron irradiation on the tensile and impact prop-erties of reactor pressure vessel materir.Js under actual operating conditions. The curveillance program for TMI-l was designed and furnished by Babcock & Wilecx; it is described in BAW-10006A.I The program was plannc-d to monitor l the effects of neutron irradiation on the reactor vesse] materials for the 40-year design life of the reactor pressure vessel. l The surveillance program for TMI-l was designed in accordance with E185-66 h and thus is not in ce mpliance with Appendixes C and 11 to 10 CFR 50 since the requireraents did not *;xist at the time the program was designed. F.ecause of this difference, additional tests and evaluations were required to ensure meeting thc. requirements of 10 CFR 50, Appendixes C and 11. The recommenda-tions for the future operation of TM1-1 included in this report do comply with these requirements. I i S f 3 !B 1480 045 1_i Babcock & Wilcox

E o3 R@gp cun "- ep o 2. BACKGROUND The ability of the teactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water cooled re-The beltline region of the reactor vessel is eM most actors. critical region of the vessel because it is exposed to neutron irradiation. The general ef- .fects of f ast neutron irradiation on t he inechanical properties of such low-alloy ferritic steels as SA302B, Code Case 1339, used in the fabrication of the TMI-1 reactor vessel are well characterized and documented in the litera-ture. 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 mechanical property change is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the upper chelf impact strength. Appendix G to 10 CFR 50, " Fracture Toughness Requirements," specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of 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 anticipated c perational occurrences and cys tem hydrostatic tesrn, to which t he prenure Loundary 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 ap-plicable to all boiling end pressurized water-cooled nuclear power reactors, including those under conceruction or in operation on the effectEve date Appendix H to 10 CFR 50, " Reactor Vessel Materials Surveillance Program Re-quirements," defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the re-actor vessel beltline region of water-cooled reactors resulting f rom exposure } k hc Vilcox 2-1

.s,msmas =--2 -,cct=_wamp y f pethermalenvironment. Fracture toughness test hh hal specimens withdrawn periodically from the re- ] g] ,o m L1 permit determination of the conditions under ed with adequate safety margins against fracture t i Y 'n 9 l brittle fracture in reactor pressure vessels is primary AS!!E Boiler and Pressure Vessel Code, Section 4 ture mec'tanics concepts and the reference nil-oled re-c hich is defined as the greater of the drop al region l t enperature (per ASTH E-208) or the tempera-hich the material exhibirs 50 ft-Jb and 35

h low-w

{ D1, of 2 given caterial is used to index that

Jon of s

i ntensity factor curve (K cur m), which ap-IR reactor Lon III. The K s IR curve is a lower bound of t fracture toughness results obtained from a i ressure steel. When a given material is indexed to o in .ntensity factors can be obtained for this ma-mpanied r. All wable operating limits can then be j E stress intensity factors. -ing limits of a nuclear power plant, can be I s of radiation on the properties of the re-ti n embrittlement and the resultant ) of water-y changes S.a ing the 1 pressure vessel r. teel can be monitored by t ghness and 'Teillance capsule containing prepared speci-s is periodically removed from the operating argins i ested. rational The increase in the Charpy V-notch se in the 35 mils of leteral expatu. ion tem-odary nay E of Appendix arger temperature shift due to irradiation, just it for radiation embrittlement. L s are 3P-This

teactors, material to the K IR curve, which, in turn the nuclear power plant.

These new limits idiation on the reactor vessel materials. > gram Re' { monitor in the re- - exposut* 1 1480 047 cock & Wilcox 2-2 Babcock & Wilcox \\

m I ? - hc @gy o N , "MMk n en 3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for TMI-1 comprises eight surveillance capsules de-h signed to monitor the effects of neutron and thermal environment on the mate-rials of the reactor pressure vessel core region. The capsules, which were p 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. Six of the capsules, placed two in each .( holder tube, are positioned near the peak axial and azimuthal neutron flux. The remaining two capsules are thermal aging capsules and are placed in an area of essentially zero neutron flux. BAW-10006A includes a full description of capsule locations and design.1 Capsule TMI-lE was removed during the first refueling shutdown of TMI-1. This capsule contained Charpy V-notch impact and tens 33e specimens fabricated of SA302, Gr B Mod Steel, weld metal and correlation steel. The specimen contained in the capsule is described in Table 3-1, and the chemistry and heat treatment of the surveillance material in capsule TMI-1E are described in Table 3-2. The capsule also contained longitudinal Charpy V-notch specimens from correla-tion material obtained from plate 02 of the USAEC Haavy Section Steel Technol-ogy Program. This 12-inch-thick place of ASTM 533, Grade B, Class 1 steel was j produced by the Luken Steel Company (heat A 11.95-1) and heat-t reated by Cebus-tion Engineering. The chemistry ar.d heat tra: tment of the corr 4at f er. Ntcrial are dencribed in Table 3-3. All test specimens were machined from the 1/4-thickness Jocation of the plates Charpy V-notch and tensile specimens from the vessel material were oriented with their longitudinal axes parallel to the principal rolling direction of the platc; the specimens were also oriented transverse to the principal rolling direction. Capsule TMI-lE contained dosimeter wires, described as follows: 1480 048 3-1 Babcock & Wilcox

--me, mm. ~.. -u. n~ q,; w Dosimeter wire Shielding U-Al alloy Cd-Ag alloy Np-Al alloy Cd-Ag alloy Nickel Cd-Ag alloy 0.66% Co-Al alloy Cd

0. 66% Co-Al alloy None Fe None Thermal monitors of low-melting cutectic alloys were included in the capsule.

The eutectic alloys and t heir melting pointa are as follows: Alloy lie,ltjpg point, F 90% Pb, 5% Ag, 5% Sn $58 97.5% Pb, 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Lead 621 Cadmium 610 1 l Table 3-1. Specimens in Surveillance Capsule TMI-1E No. of specimens l Material description Tensile Charpy Wel d ne t al, WF-25 4 a Heat -af f ccted zone "A" (HAZ), Heat C-2789-2, Longitudina) 0 E I Baseline waterial, plate "A". Heat C-2789-2: Longitudinal 4 8 Transverse 0 4 Correlation, HSST, Plate 02 0 8 Total per capsule 8 36 I i i d i I 1480 049 i 1 3-2 Babcock & kVilcox !L

Table 3-2. Chenistry and Heat Treatment of Surved11ance Materials Chemical Analysis l Heat Heat Weld metal Element C-2789-2 C-3307-1 WF-25 C 0.24 0.21 0.088 Mn 1.36 1.24 1.50 g P 0.010 0.010 0.019 S 0.017 0.016 0.010 Si 0.23 0.27 0.45 Ni 0.57 0.55 0.71 Mo 0.51 0.47 0.33 Cu 0.09 0.12 0.34 Heat Treatment Ilea t

Temp, Tin,c,

No. F h Cooling C-2789-2 1600-1650 9.5 Brine quench M 3200-1225 9.5 Brine quench 1600-1650 9.5 Brine quench 1600-1650 9.5 Brine quench 1510-1535 5.0 Brine quench 1200-1225 5.0 Brine quench 1100-1150 40.0 Furnace cooled C-3307-1 1600-1650 9.5 Brjne quench 1200-1225 9.5 Brine quench 1225-1250 9.5 Brine quench 1100-1150 40.0 Furnace cooled l WF-25 1100-1150 27.5 Furnace cooled i 3 E HBO 050 3 3-3 Babcock & Wilcox

Table 3-3. Chemistry and Heat Treatment of Correlation Material - Heat A-1195-1. A533 Grade B, Class 1 (HSST Plate 02) _Ch'emical Analysis (1/4T) # 1 ) Element Wt % j C 0.23 h Mn 1.39 t P 0.013 S 0.013 Si 0.21 Ni 0.64 Ho 0.50 Cu 0,17 i 11 eat Treatment l 1. Normalized at 1675F 75P4 2. 1600F 3 75r for 4 h/ water-quenched. 3. 3225F 25F for 4 h/ furnace-cooled. 4. Il25F 25F for 40 h/ furnace-cooled. (* ORNL-44 63. 1 i 1480 051 3-4 Babcock & VVilcox

~ ~ ~ Figure 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule Locations X Surveillance Capsule Holder j Tubes - Capsules TMI-1C,

./

TMI-1D r f ~ / ~~ N f g N / l O l O O O s 0 0 O O O lO O / O 0 9 0 0 g j O 0 9 l 0 O O o e o ]i y w o ^ 3; g' s O 0 0 [/ . O O O . O O O 0 9 O g / O o o / y g O O O O / f \\ O / / g o l O m/ /'- c {j h Surveillance Capsule Holder Tubes - Cap-l sules TMI-1A, TMI-1B Survei]1ance Capsule s Holder Tube - Capsules Z TMI-1E, TMI-lF, nil-lG*, TMI-lH* 3

Thermal aging capsules.

I j 1480 052 3 3-5 Babcock & Wilcox

~ 7 i l i !a 4 g y 4. PREIRRADIATION TESTS I T p- ^ ,9 Unirradiated material was evaluated for two purposes: (1) to establish a baseline of data to which irradiated properties data could be referenced, 'nd b (2) to determine those materials properties to the extent practical f rom avail-1: able material, es required for compliance with Appendi::cs G and U to 10 CFR 50. o 4.1. Tensile Tests i Tensile specimens were fabricated f rom the reactor vecsel shcIl course plate and wcld metal. The subsize specimens were 4.25 inches long wit.h a reduced section J. 750 inches long by 0.357 inch in diameter. They were tested' on a j .i 20,000-3b-load capacity universal test mach'ne at a crosshead speed of 0.005 t r inch per minute. Test conditions were in accordance with the applicahic re- ? quirements of ASTM A370-72. For each material t.ype and/or condition, six 4 specimens in groups of three were t.ested at both room temperature and S70F. ) An LVDT-type clamp-end t crew-on ext.cusomet.er was used t o detc.rmine the 0.2% l yield point. Tbc t ension-compression load cell used had a certified necuracy h J of better than t 0.5% of full scale (10,000 lb). All test. data for the pre- [ irradiation tensile specimens are given in Appendix b. j 4.2. Impact Tests _ h Charpy V-notch impact t ests were conducted in eccordance with the requirements L 4 of ASTM "tandard Methods A170-72 and E23-77 on a remote controlled impact test-er certified t o meet Wat.ertown standards. 're s:t 5.pecirucnn s er e of t he Charpy ] V-notch type, which are 0.394 inch square and 1.165 luthec long. Prior to testing, specimens were temperatute-conditioned in a combination re-I sistance-heated / carbon dioxide-cooled chamber, designed t o cover the t empera-ture range from -85 to +550F. The specimen support arm, which is linked to g the pneumat ic transf er wrchanism, is instrumented with a contacting t hermo-couple allowing instantaneous specimen temperature determinations. Specimens were transferred from the conditioning chamber to the test fram anvil and pre-j cisely pretest-positioned with a fully automated, remotely controlled apparatus. 4 t Babcock & Wilcox 4_1 1430 093

, s_ 4 f 6 5 Transfer times were less than 3 seco. ids and repeat within 0.1 second. Once the specimen was positioned, the electronic interlock opened, and the pendulum was released from its preset drop height. After failing, the specimen, the g hammer pendulum was slowed on its return stroke and raised back to its start position. Impact test data for the unirradiated baseline reference materials are presented in Appendix C. Tables C-1 through C-9 contain the basis data which are plotted in Figures C-1 through C-9. s I l o 4 a i 1 J. 1! 1 Il 5 l 1, I 'l i I I i 1480 054 i i i i 4-2 Babcock & NVilcox I

l3 4 nm amm 5. POSTIRRADIATION TESTS c~ 5.1. Thermal Monitors j Surveillance capsule TMI-lE contained three temperature monitor holder tubes each containing five fusible alloys with different melting points ranging from l 558 to 621F. All the thermal wonitors at 558, 580, and 588F had melted, while those at 610 and 621F remained in tb<fr original confj uration ac initially g lf at ed in the capsule. From thecc data it was concluded that the irtadiated specimens had been exposed t o a maximum temperature in t he range of 588 to less i than 610F during t.he reactor vessel operating period. There appeared to be no 1, significant temperature gradient along the capsule length. _S.2. Chemical Analysic 4i Two broken impact specimen halves taken at. random from each of the two unirra-diated base. wetals (Heats C-3307-1 and C-2789-2) and the weld metal (WF-25) were analyzed for nickel, copper, phosphorus and sulf ur contents to verify original mill t est report data. To minindre possible scat ter in data due to surface conditions, r.pecimens were mechanically cleaned prior to analysis Four cets of copper analysis data were obtained from each of the six specimen halves, And t hree analysee were i.erformed for t he remaining clonents i (n ickel, phosphorus, and rulfin ). I The f ollowing aunlyt-ic.o1 t.echniques were emp3oyed: 1. fuckel and coppet content. by v-ray 1]uorescence nonly< in un ng B6W I Standard Analytical Method XR-2. 2. Phosphorus content by colorimetric absorption measurement using " molybdenum blue complex" solutions in accordance with ASTM E350 72 and B&W Standard Analytical Met hod P-3. 3. Sulfur content by gravimetric precipitation of bas 0% in accordance with ASTM E350-72. { 1_ 1480 055 i* 5-1 Babcock & Wilcox i

j a The data from chemical analyses were averaged, ar.d the results are reported i in Table 5-1 along with the tpplicable standare leviation values. ~ i 5.3. Tensile Test Results f 1 The results of the postirradiation tensile tests are presented in Table 5-2. Tests were performed on specimens at both room temperature and 570F. In gen-eral, the ultimate yield strength of the material increased slightly with a 4 corresponding slight 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 {t expected for the radiation environment to which the specimens were exposed. I The results of the preirradiation tensile tests are presented in Appendix B. 1 5.4. Charpy V-Notch Impact Test Results i The test results from the irradiated Charpy V-notch specitens of the reactor vessel beltline material and the correlation monitor material are presented in Tables 5-3 and 5-4 and Figures 5-1 through 5-5. The data show that the material exhibited a sensitivity to irradiation within the values predicted from its chemical composition and the fluence to which it was exposed. i The results of the preirradiation Charpy V-notch impact. test are given in Appendix C. I J Table 5-1. Chemistry Data on Unirradiated TMI-1 RVSP Material CVN-opecitar Material ~ type /- number heat No. Ni 0.01 Cu 1 0.02 P 1 0.003 S 1 0.004 s CC 738 Base metal /C-2789-2 0.42 0.10 0.018 0.007 DD 731 Base metal /C-3307-1 0.50 0.13 0.012 0.005 CC 018 Weld metal /WF-25 0.43 0.34 0.019 0.007 1480 056 5-2 Babcock s.Wilcox

3 1 Table 5-2. Tensile Properties of Capsule TMI-lE Base Metal ~ and Weld Metal Irradiated to 1.07 x 1018 n/cm2 (E > 1 MeV) Strength, psi Elongation, % Test Specimen

temp, Yield Ult.

Uniform Total Red'n of ID No. F (YS) (UTS) (UE) (TE) area, % Base Metal - Heat C-2789-2, Longitudinal CCE 721 RT 72,320 96,040 10.94 25.29 64.2 CCE 723 RT 72,900 95,750 11.06 28.80 67.8 g Mean 72,610 95,895 l?.0 27.05 66.0 45 Std dev'n 290 145 0.06 1.76 1.8 CCE 711 570 65,230 94,960 11.85 28.38 62.1 f CCE 713 570 67,840 96,190 31.53 26.43 57.7 Mean 66,535 95,575 11.69 27.41 59.9 Std dev'n 1,305 615 0.16 0.98 2.2 Weld Metal - WF-25 CCE 111 RT 83,320 98,200 10.20 23.94 55.9 CCE 124 RT 81,900 96,200 10.79 25.57 55.6 Mean 82,610 97,200 10.50 24.76 55.75 Std dev'n 710 1,000 0.30 0.82 0.15 l CCE 112 570 74,800 94,990 9.48 19.64 38.4 CCE 117 570 77,290 94,620 9.01 J8.53 (a) [ Mean 70,045 94,505 9.25 19.09 38.4 Std dev'n 1,245 185 0.24 0.56 3 (^} Measurement could not be obtained. 1 I I !I f 1480 057 l 5-3 Babcock & Wilcox

Table 5-3. Charpy Impact Data for Capsule TMI-1E Bas 2 Metal Irradiated to 1.07 x 1018 n/cm2 (E > 1 !!eV) Test Abs Lateral Shear Specimen

temp, energy,

exgans, fracture, ID No.

F ft-lb 10 in. Heat C-2789-2, Longitudinal CCE-739 399.6 118 60 100 709 279.6 118 68 100 736 198.9 121 57 98 713 140.6 83 53 55 735 70 66 43.5 35 751 60.9 47 36.5 5 743 41.2 48 35 6 750 0.1 22 13 2 Heat C-2789-2, Transverse CCE-608 278.3 88 61.5 100 618 140.1 59 41 20 627 101 56 40 25 626 70 53 36 6 HAZ - Heat C-2789-2, Longitudinal CCE-447 389 128 63.5 100 407 282.3 76 44.5 100 419 141.1 123 51 45 431 70 92 49.5 45 442 0.8 98 42.5 30 452 -39.8 72 42.5 40 448 -60.0 48 26 35 450 -60.5 40 20 12 1 4 E'y 1480 058 5-4 Babcock & Wilcox 9

A Table 5-4. Charpy Impact Data for Capsule TMI-1E Weld Metal e-(WF-25) Irradiated to 1.07 x 1018 n/cm2 (E > 1 MeV) e Test Abs Lateral Shear ) Specimen

temp, energy, 10gai.s,

fracture, ex ID No.

F ft-lb in CCE-040 395.4 59 41.5 100 002 279.9 64 42.5 100 1 033 200.2 55 34 90 ( 050 172 56 33 90 t 013 139.7 46 32.5 65 043 120.5 41.5 31 25 038 306.8 31 19.5 30 031 70 42 28 45 1' Table 5-5. Charpy Impact Data for Capsule TMI-1E Correlation Monitor Material Irradiated to 1.07 x 3018 n/cm2 (E > 1 MeV), Heat A-1195-3, Transverse Orientation Test Abs Lateral Shear Specimen

temp, energy, ex fracture

}[ pans, _1D No. Y .ft-3b in. 2 CCE-969 380.9 109 69 100 913 277.3 115 64.5 100 925 198.8 100 53 100 1 962 200.8 108 54.5 92 y 919 340.5 74 54 35 1 964 119.4 52 37 20 <g 948 105.8 34 24 20 g 922 70 28 20.5 12 1 1480 059 c I 5-5 Dabcock & Wilcox

1 1 ? Figure 5-1. Impact Data From Irradiated Base Metal A, Longitudinal Orientation i g 100 i 3 1

- 75

.'1 E 3 50 .m 5 e 25 ( Y m 1 ? ? O 4 3 .080 _e r N $.060 l 5 t w.040 5 E =

'.020 I

lll e ii .5 i e i i J s 200 DATA SU N RY T 180 nDT 52F T (35 n2) cy 55F 160 I (50 rr-u) cy 120F (-USE(ave) -5F (Best estimate) 3 140 - RTnot 5' 120 a 8 // 100 t V a w 80 O< 1 e ~~ 60 40 SA302 Gr E Mod hitRtaL 0mitniarie., Longitudinal 1.0E+18 nv: 20 Fwtxt gg, 1 g,y) HLAT Munga C-27 '9-Z i i i i i .g -80 -40 0 40 80 120 160 200 240 250 320 MO 400 Test Ttwenatung, F 480 060 Babcock & Wilcox 5-6

j Figure 5-2. Impact Data From Irradiated Base Metal A, Transverse Orientation 100 w 75 E i ys0 _. _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. _ _ _. _ _ _ _ _ _ _. _ _ _ _ _ _ E 3 x 2s - e .080 i i.060 5 0.040 5

=

lt.020 5 .5 i e e e i i e e i ,g__ 200 I. DATA SUIT %RY O 180 T,,, T (3s as)

/0F 3

05F g-J!,0 - !u (s0 ri-u) 8" (-USE(ava) 140 - RT,7 25F (Best estimate) 3 $ 120 [' .k 3 100 t] t, 5 l 0 80 - e-N I W - __-_.-_.-____e_ 40 krtAIAL SA302 Gr B Mod" 20 - Onismratrog _ Transverse k. Fturnet 1.07E+18 nyt ~. (E > 1 Mev) HEAT ltwega c 7, o i EO yo do yn 3lg 80 120 s 3a m itsr Tr & ERATURE, f a 5-7 Babcock 8. Wilcox 1480 061

8 4Lvi i Figure 5-3. Impact Data From Irradiated Base Metal A, g HAZ, Longitudinal Orientation 100 g s< 75 h J .E. y 50 - - _ - - - - - - - - - - - - - - - e 25 e 0 .080 i 3 060 e I e t o w.040 / 5 ~~T-e e

t.020 -e 3

i i e i e i i i ,g 200 DATA 'SumHY ] T NA f) 180 nr T (35 su) -18F cy 160 I (50 FT-u) ~W I cy C -USE (Avs) 103F y 140 - RT -78F (Best estimate) eT e b e $M ca 3 3 100 s-g ? .- t

i. <

Fw 80 isW e <g 1 ~ 60 d 40 i %rta utSA302 Gr D Mod. 20 Onnenration HAZ, Long. 1.07E+18 nyt j FwtNet (E > 1 MeV HEAT Nunata e 'm-2 f t i e e i e i N Q-80 -40 0 40 80 120 150 200 243 280 320 M0 O h It3TTEMPtaAtung,F 4 V E 5-8 Babcock & Wilcox h < 4nn n,, 11 N,.] , t,A j = l ~

f 1 i l j Figure 5-4. Impact Data From Irradiated Weld Metal, J Transverse Orientation i 100 75 l = ? 9 50 - - - - - - - - - - - - - - - - - - - - - - - - - - - g e Iw. a ?! I o f I s. s a 3 4 a a a a e f.060 M 5 O.040 .j A t; at

i.020 e

er U .5 I f I f I t t I f e y i = ,l DATA 'SUPfWRY ~ y NA (( 180 et Tcv (35 M.2) 180F h' 160 T (50 rr-in)1%F cy g.gg,(,ys) 64 ft-lb 120F (Best esti= ate) 140 - RTGT $^ f, 120 i J S .h 100 ? .T tg o = l I - gj g t ^ i 3 40 - e ,r ir a t u Weld Metal s w l 20 - OnstnT4 Tron 1.d m le nyt ""<E M > 1 Mev) HEAT Nuden WF-2 i. l 0 i e i e i e r -80 -40 0 40 80 120 160 200 240 250 320 MO 400 Itsr Tte taATuat, F 5-9 Babcock E. Wilcox 4 4., q .p.

Figure 5-5. Impact Data From Correlation Monitor Material, Transverse Orientation 100 = 75 .l J = E 1 y 50 _ _ _ _ _ _ _ _ _ _ _. _ _ _. _ _ _ .w 3 i W 3 a 25 g n ) 0 I .080 = e i I 5 D $.060 ] 8 i i O.040 3 E = i.020 ".1 I i .3 1 ,g, e i e e i i i e i 200 i j DATA SUPf%RY T NA 180 er i l 05 Ms) 119F a 160 I (50 n-ts) ll8F cy llSF (-USE(avs) 140 - RT,, 59F (Best estimate) 3 l b I E 120 C e & 100 I b I Ww M O I t I - g 40 hitaig HSST-PL-02 20 - Onioitarica Transverse Fww 1.07E+1Es nyt gg, 1 g,y HEATKussen A-i +ic. i i i i e i i 0-80 -40 0 40 83 120 ISO 200 243 280 320 360 400 itsr Ttwenarvar, F 1480 Of]4 3-10 Babcock a wiicox

k 1 2 h !] 6. NEUTRON DOSIMETRY h 6.1. Introduction A significant aspect of the surveillance program is to provide a correlation between the neutron fluence above 1 MeV and the radiation-induced property t changes noted in the surveillance specimen. To permit such a correlation, activation detectors with reaction thresholds in the energy range af interest were placed in each surveillance capsule. The properties of interest for the g detectors are given in Table 6-1. i Because of a long half-life (30 years) and an effective energy range of more than 0.5 MeV, only the measuren:ents of 137Cs production from fission reactions in 237 238 Np (and U) are directly applicable to analytical determinations of (. the fast neutron (E > 1 MeV) fluence during Cycles 1 and 2. The other dosim-cters are useful as corroborating data for shorter time intervals and/or higher energy fluxes. Short-lived isotope activitics are representas.ive of reactor conditions over the latter portion of the irradiation period (fuel cycle), 3[ only, whereas reactions with a high threshold energy do not record a signifi-g ; cant part of the total fast flux. The energy-dependent neutron flux is not directly available from activation 'I detectors beacuse the process provides the integrated effect of the neutron [ flux on the target inateria) as a function t,f both irradiation tiac nnd neutron u h, energy. To obtain an accurate estiuate of the average neut rea f J ux incident upon the detector, several parameters must be known: the operating history of the reactor, the energy response of the given detector, and the neutron spec-trum at the detector location. Of these parameters, the definition of the r neutron spectrum is the most difficult to obtain; essentially two taeans are ) available: iterative unfolding of experimental foil data and analytical f methods. Because of a lack of sufficient threshold foil detectors satisfying 2 both the threshold energy and half-life requirements necessary for a

]

1480 06r3 1 6-1 Babcock & Wilcox i.t.

e ? -^ surveillance program, iterative unfolding could not be used. This leaves the 3 specification of the neutron spectrum to the analytical method. 6.2. Analytical Approach Energy-dependent neutron fluxes seen by the detector were determined by a discrete ordinate solution of the Boltzmann transport equation. Specifically, 3 4 ANISN, a one-dimensional code, and DOT, a two-dimensional code, were used to calculate the flux at the detector position. In both codes, the Oconee system was modeled radially from the core out to the air gap outside the pressure vessel. The model included the core with a time-averaged radial power dis-tribution, core liner, barrel, thermal shield, pressure vessel, and water regions. Including the internal components enables the analytical method to account for the distortions of the required energy spectrum by attenuation in these components. The ANISN code used the CASKS 22-group neutron cross sec-tion set with an S6 order of angular quadrature and a P3 expansion of the ] scattering matrix. The problem was run along a radius across the core flats. Azimuthal variations were obtained with a DOT r-theta calculation that modeled- ~ a one-eighth plan view of the core and included a pin-by pin, time-averaged power distribution. The DOT calculation used S6 quadrature and a P1 cross v. section set derived from CASK. t 1,' s : Flu.ces calculated with this DOT model must be adjusted to account for lack of g P3 cross section detail in calculations of anisotropic scattering, a pertur-bation caused by the presence of the capsule, and the axial power distribution. D The first two items are both energy-and position-dependent. A P /P1 correc-3 tion factor was obtained by comparing two ANISN one-dimensional model calcula- ~ tions in which only the order of scattering was varied. The capsule pertur-bation factor was obtained from a co.aparison of two DOT x-y model calculations, ^ one with a capsule explicitly modeled - SS 304 cladding, aluminum filler region, hk 1: and carbon steel specimens and the other with water in those regions. The effect of axial power distribution was determined from a previous DOT r-z mod-el using an estimated average (axial) over three fuel cycles. The net result i from these parametric studies was a flux adjustment factor K (TC.:le 6-2), which Ns should be applicable to I he appropriate dosimeters in all 177-fuel rasecbly C.,, lNE9' surveillance programs in which the capsules are located at a radius of 211 cm '4 from the core center and at 11 degrees from a major axis. ~' O' 1480 066 s 5 6-2 Babcock & Wilcox

The calc *ation described in Table 6-2 provide the neutron flux as a function g of energy at the detector position. These calculated data are used in the following equations to obtah the calculated activities used for comparison I with the expe:-imental values. The basic equation 6 for the activity D (in { pC1/g) is as follows: 1 3 M -A t -1 (T-1 ) f 3.7f 4 Il I = 10 i "n ( ( ~" i j r i E j=1 i j C = normalizing constant, ratio of measured to calculated flux, H = Avagadro's number, {P Ag = atomic weight of target material 1, f = either weight fraction of target isotope in nth material f j or fission yield of desired isotope, o (E) = group-averaged cross sections for material n (listed in Table D-3), $(E) = group-averaged fluxes calculated by DOT analysis, F = fraction of full power during j th time interval, t), 13 = decay constant of ith material, h t) = intervel of power history, [ T = sum of total irradiation time, i.e., residual time in reactor, and wait time between reactor shutdown and counting, ga cumulative time from reactor startup to end of j th time

period, i.e.,

1 = t 3 k' k=1 { The normalizing constant C can be obtained by equating the right side of equation 6-1 to the measured activity. With C specifjed, the neutron fluence .j greater than 1 MeV can be calculated from a 15 MeV M y ((E s 1.0 MeV) = C [ 4(E) [Ft$$ (6-2) d I) E=1 'i = 1 f where M is the number of irradiation time intervals; the other values are defined above. 6.3. Results m Eh Calculated activities are compared to measuren.ents of the dosioners in Table d' 6-3. The 137 y Cs data show that f ast flux (E > 1 MeV) is so \\ 6-3 Babcock & Wilcox .r

29 I ?. e (s10%) by the analytical model described herein (if one assumes that the cal-l ? culated flux spectrum is correct). Such agreement is probably within the i 4 uncertainty limits of this snalysis. However, for conservatism, a flux normal-l ization factor of 1.1 is recommended for fast neutron calculations near the 4 pressure vessel. The 103Ru activities (because of a short half-life) indicate 1 that the core leakage flux over the latter part of the cycle was greater than + 3 the cycle average (the basis for the analytical model). An inspection of rela-f tive power distribution in the core showed that this is true. The St+Mn activity indicates an overprediction (s20%) of more than 2 MeV flux by the analytical model. The significance of this to the fast. flux calculation is tempered by i t the fact that approxtmately 40% of the neutrons with E > 1 MeV are in the 1- {f to 2-MeV range (Table 6-4). The 58Co activity is the result of bc:h of these effects overprediction of hign energy flux (high threshold energy) and high J leakage flux over the latter part of the cycle (short half-life). Sj l t The results of this analysis are consistent with a previous 177-fuel assembly I A surveillance specimen analysis.2 Future capsule data should add confidence q to the analytical procedure and pc,ssibly clarify variations in the normaliza-tion constant. g J# Besed on a normalization constant of 1.1, an average fast flux for Cycle 1 was 7 calculated at the capsule location and at the inside surface of the pressure 3 vessel wall. The data (Table 6-5) were converted to fast fluence values of f 1.07 x 1018 n/cm2 at the capsule center and 5.90 x 1017 n/cm2 at the pressure vessel wall for Cycle 1 at the full power rating of 2535 MWt. Extrapolated j values to a 40-year lifetime with a u s factor of 0.8 are also presented. It ,jt was noted that the maximm fluence at the pressure vessel occurred at an azi-muthal position of 8 degrees from a major axis (capsule located at 11 degrees); this is a function of power distribution in the core. Also, imcmse power i distributions can change significantly from one fuel cycle to ancther, it is inappropriate to compare extrapolated Cycle 1 data to the predicted 40-year fluence of 2.9 x 1019 n/cm2 (three-cycle average power distribution at 2772 MWt). The effect of extending the flux range down to 0.1 MeV was to approxi-mately double the fluence at the capsule and the pressure vesnel.

However, f

additional uncertainty is introduced in this result because none of the dosim-eter. reactions are effective over this entire range 4 8 0 0 6 8 i, Y 6-4 Babcock 8.Wilcox i

g A more meaningful prediction can be obtained by extrapolating Cycle 1 flux, ( which has been modified by the calculated relative core leakage flux based j, on predicted fuel reload and burnup conditions in Cycles 2 and 3. These data are presented in Table 6-6 and Figure 6-1 for various locations in the pressure vessel. The major assumptions used are that fast flux at the pressure vessel is proportional to core leakage flux and that power distributions for Cycles 1, 2, and 3 (2.7 EFPY) are applicable out to 10 EFPY. This will require that no significant changes occur in fuel reload procedures and length of cycle { and that cycles subsequent to Cycle 3 be similar to that cycle. Thus, there is considerable uncertainty that should be considered in using Table 6-6. These fluence values are somewhat Jower ihan the generic design curve because g of lowe'r reactor power, less severe power distribution, and no allowance for uncertainties in fuel reload procedures. 3 u Table 6-1. Surveillance Capsule Detectors U Threshold Isotope Detector reacticq _enere,y, MeV half-life 54Fe (n. p) S4Mn 2.0 303 days 58Ni(n, p) 58Co 2.5 71.3 days 238U(n f)l37Cs 1.5 30 years 237Np(n,f)137Cs 0.5 30 years i 238U (n, f)103Ru 1.5 39.5 days f 237 l. Np (n, f ) 10 3Ru 0.5 39.> days i til f 'lable 6-2. Flux Adiustment Factor t [_ Energy Axial power Capsule p jp rangc, MeV factor 3 1 perturb'n ,_ L >0.1 1.1 1.22 1.40 1.88 h >l J.1 1.23 1.20 1.62 >2 1.1 1.24 1.04 1.42 >2.5 1.1 1.25 0.96 1.32 0M I 4;* d 6-5 Babcock t.Wilcox

)1 N i Table 6-3. Dosimeter Activations ~ A B measured calculated C = A/B activity, (a) activity, normalization i Reaction pCi/g pCi/g constant } 54Fe (n,p) S4Mn 446 537 0.83 ij 58Ni(n.p)58Co 832 868 0.96 ) 238U (n, f)137Cs 1.42 1.24 1.15 237Np (n, f)137Cs 7.23 6.40 1.13 238U (n, f)10 3Ru 47.9 37.1 1.29 237Np (n, f)103Ru 205 173 1.18 1 I (* Average of four dosimeter wires from Table D-2. 1 Table 6-4. Normalized Flux Spectra. E > 1 MeV I 5 In water Energy range, near pressure 235g MeV _ vessel wall fission 1 12.2-15.0 0.0015 0.0002 10.0-12.2 0.0063 0.0013 8.18-10.0 0.0181 0.0052 l 6.36-8.18 0.0499 0.021 't i i 4.96-6.36 0.0906 0.051 } 4.06-4.96 0.0784 0.052 3.01-4.06 0.1159 0.159 I: 2.46-3.01 0.1200 0.132 2.35-2.46 0.0389 0.034 1.83-2.35 0.1506 0.178 1 1.11-1.83 0.2832 0.323 1.0 -1.11 0.0466 0.044 1.000 1.000 i 1480 070 4 6-6 Babcock t.Wilcox

namen: Table 6-5. Neutron Fluence j .)

Flux, Cycle 1 Lifetime (8) n/cm -s 2

(_471 EFPD) (_32 years) Fast fluence. E > 1 MeV 4 / Capsule center 2.62+10 1.07+18 2.65+19 Pressure vessel wall 1.45+10 5.90+17 1.46+19 P e Fluence. E > 0.1 MeV 9 ^r Capsule center f 5.25+10 2.14+]8 5.30+19 Pressure vessel wall 2.95+10 1.20+18 2.98+19 b' 8 Extrapolation of Cycle 1 average flux. Table 6-6. f Pressure Vesse] for 10 EFPY(a) Predicted Fast Neutron Fluence in i Locatior. in pressure vessel inside Outside surface T/4 _3T/4 surface g Fast neutron flux, 1.68+10 9.33+9 2.24+9 9.60+8 2 n/cm -s, E > ) MeV

i Fast fluence, n/cm2 5.3+18

3. 0+] 8 7.1+17 3.0+17 6I i

Effective full power years.

)

I L, il 4 1 1480 071 i I 6-7 Babcock & Wilcox

Figure 6-1. Predicted Fast Neutron Fluences at Various Locations Through Reactor Vessel Wall for First 10 EFPY 6.0 I I l g g 8 5 5.6 5.3 x 1018 5.2 ^n 4.8 g h 4.4 4.0 A 3.6 ~ y

p" C

a m b 3.2 f>6* 3.0 x 1018 2.8 2.4 E 2.0 got gNE A 1.6 a b k 1.2 ~ g 7.1 x 1017 0.3 ~ o 00 o Q. O 3/4T Incation 3.0 x 1017 0.4 e Outside location O N - = - t I 5 0 o 2 3 4 5 6 9 'O x EFPY wsy __a

} } t> 4i 7. DISCUSSION OF CAPSULE RESULTS 3 4 ) 7.1. Preirradiation Property Data j A review of the unirradiated properties of the reactor vt asel core belt region indicated no significant deviation from expected properties except in the case 'of the upper nheJf properties of the weld metal. Based on the predicted end-F of-service peak neutron fluence value at the 1/4T vessel wall location and the A copper content of this weld, it is predicted that the end-of-service Charpy upper shelf energy (USE) wj)) be below 50 fe-lb. This weld was selected for inclusion in the surveillance program in accordance with the criteria in ef- ) feet at the time the program was designed for TMI-J. The npplicable selection criterion was based on the unitradiated properties only. 7.2. Irradiated Property Data [0 7.2.1. Tensile Properties Table 7-1 compares irradiated and unirradiated tensile properties. At *ooth room temperature and 570F, the ultimate and yic]d ctre.ngths incteased slightly as a result of irradiation accompanied by a corresponding decrease in ductil-ity. The elongation of the base metal appears to have increased at the 570F l test temperature; however, the naall increase may be attributed to the anntal-ing out of irradiation ef fects during the time the furnace temperature was j s tabili zing. In either case, the small change in tensile properties is insig- ) nificant. relative to the analysis of the scactor vessni materials at die pe-riod in service life. 4 7.2.2. Impact Properties I The behavior of the Charpy V-notch impact data is more significant to the cal-i culation of the reactor systems' operating limitations. Table 7-2 compares the observed changes in irradiated Charpy impact properties with the predicted changes as shown in Figures 7-1 through 7-5. } i,.l 1480 073 y_1 Babcock & Wilcox 3

The increase in the shift of the data at the 50-ft-lb transition temperature was greater than the predicted shift for the base metal on the values given in Regulatory Guide 1.99. However, when these observed values are compared with the predicted values given in BAW-10056, it is seen that the values are in good agreement.10 Comparison of the shift of the weld metal and correlation material to the Regulatory Guide 1.99 predicted values shows good agreement. This indicates that the prediction curves for low-copper materials at low flu-ence levels are not conservative. The prediction curves for medium-and high-copper materials are in good agreement with the observed values. The increase in the 35-mil lateral expansion transition temperature is compared, with the shift in RTg curve data in a manner similar to the comparison made for the 50-ft-lb transition temperature shift. These data show a behavior similar to that observed from the comparison of the observed and predicted transition data. The significant difference is the larger shif t exhibited by the weld metal, which is due primarily to the fact that the data curve dropped close to the limit curve. Therefore, the large shift can be attributed to the method of measuring the MLE shift. Since the medium-and high-copper materials have the tendency to exhibit the greater irradiation damage, and, in turn, are the materials that ultimately control the reactor operation limitations, the use of the prediction curves of Regulatory Guide 1.99 for establishing the operation limitations is conserva-tive. The data for the decrease in Charpy USE with irradiation showed a good compari-son for the base metal, which had a low copper content. The weld metal indi-cates a very conservative predictive value, which is a reflection of the lack of data for high-copper-content weldments at low fluence values. The RT shifts shown are higher than predicted from Regulatory Guide 1.99 at NDT the fluence level of t his capsule. This indicates that the predicted curves are not accurate for the very low neutron fluence level (%1 x 1018 2 n/cm ), This inaccuracy is a result of the limited data at the low fluence values and of the fact that the majority of the data used to define the curves in Regula-tory Guide 1.99 are based on the shif t at 30 f t-]b as compared to the current requirement of 50 ft-lb. For most materials the shif ts measured at 50 ft-lb/ 35 MLE are expected to be higher than those measured at 30 ft-lb. The signif-icance of the shifts at 50 ft-lb and/or 35 MLE is not well understood at Babcock & Wilcox 7-2 54qq q73

i present, especially for materials having USEs that approach the 50-ft-lb level ,5 and/or the 35-MLE level. Materials with this characteristic should be evaluated at transition energy levels lower than 50-ft-lb. j The design curves for predicting the shift at 50 ft-lb/35 MLE will probably be modified as data become available; until that time, the design curves for pre-dicting the RT shift as given in Regulatory Guide 1.99 are considered ade-ET quate for predicting the RT shift of those materials for which data are not g available and will continue to be used to establish the pressure-temperature i operational limitations for the irradiated portions of the reactor vessel. -,li f Table 7-1. Comparison of Tensile Test Result,s, l1* Room temp test 570F test i Unirr Irrad Unirr Irrad ~ Base Metal - C-2789-2, Longitudinal Fluence, 1018 n/cm2 (> 1 MeV) 0 1.07 0 1.07 Ult. tenaile strength, kei 93.9 95.9 91.7 95.6 0.2% yield strength, kai 71.1 72.6 62.7 66.5 Elongation, % 26.6 27.1 25.7 27.4 h RA, % 67.9 66.0 58.8 59.9 3 Weld Metal - WF-25 \\ Fluence, 1018 n/cm2 (> ) MeV) 0 .i.07 0 1.07 Ult. Lensile strcngLh, ksi 86.2 'i/ 2 ti2.0 94.6

0. 2% y ield s t.r ength, ksi 69.?

82.6 64.3 4.0 ,1 Elongation, I 26.7 24.7 20 S 19.] RA, % 62.8 55.7 52.3 38.4 3 i; 3 t 3i 1480 075 h t y_3 Babcock & Wilcox

.a 4 7 Table 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Properties Material Observed Predicted

Increase in 50-ft-lb trans temp, F Base material (C-2789-2)

Longitudinal 29 10 1 Transverse 15 10 j Base material (HAZ), longitudinal 8 10 l Weld retal (WF-25) 117 110 1 Cor: elation material 44 48 Inc.rease in 35-MLE trans temo, F { Base material (C-2789-2) 4 i Longitudinal 38 30( ) i Transverse 13 10(b) g Base material (HAZ), longitudinal 22 10(b) l Weld metal (WF-25) 190 110(b) l Correlation material 64 48(b) { s Increase in Charpy USE, ft-lb I t Base material (C-2789-2) Longitudinal 11 13 Transverse 10 9 ' Base material (HAZ), longitudinal 23 13 i, Weld metal (WF-25) 17 24 I Correlation material I 15 21 I (" These values predicted per Regulatory Guide 1.99, Revision 1 Based on the assumption that MLE as we)) as 50-ft-lb transition temperature is used to control the shift in 1:TET' I i i I 1480 076 7-4 Babcock 8. Wilcox

i 4 1 t Ij Figure 7-1. Irradiated Vs Unirradiated Charpf Impact Properties J of Base Metal, Longitudinal Orientation I i 100 } w 75 Unirradiated 1 E .3 y 50 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 5 1 5 25 1.07 x 1018 2 n/cm t 0 e f f 3 f ,080 j Unitradiated ,5.060 5 ) i$,040 1.07 x 1018 n/m2 } j

y g.020

/ r AT = 38F .3 g f .5 i i e i .000 200 r i 380 p il 160 ij [g 3 140 'i ? J 120 ,A i 9 Unirradiated - AUSE " 11 ft-lb .h JM ~ i [2 ,$ go AT = 29F - + -+ i j J.07 x 10l# n/cmi i g E 60 / T

[------_--________________________

J ArtalALSA302 Gr B Mod 20 -

CastNTAriOft Longitudinal FLutser see above Haar flutste c-2789-2 b h 80 -40 0 80 120 160 200 240 280 320 MO %) ,g Test Ttw tRAruas, F E14 s 1 7-5 Babcock & Wilcox )4 '4 1480 077

Figure 7-2. Irradiated Vs Unirradiated Charpy Impact Properties of Base Metal, Transverse Orientation 100 a 75 J y 50. _ _ _ _ _. _ _ _ _ _ _. _ _ .w i r 25 a 0 .080 i. $.060 Unirradiated E $.040 1.07 x 1018 n/m2 g t =

t.020 aT = 13F e

.5 i i e ,rg M i 180 - 160 . 140 T' 5 120 i 8 .lP 100 3 Unirradiated o }M w t, .USE = 10 ft-lb aT = ISP r .I [ 4 1.07 x 1018 2 n/cm 60 40 - skrEAinSA302 Gr B Mod 20 - O"'I"'**' " Transvezse Futac;_see above p2Ar Nupera C-2789-2 0 -80 -40 0 40 80 120 ISO 200 240 280 320 360 400 itsrTenet=Aruar,F 78 14 8 0 "6 cock 8.Wilcox 5$ 7-6

Figure 7-3. Irradiated Vs Unirradiated Charpy Impact Properties of Base Metal, HAZ

100,

" U ~ Unirradiated Cy so 48. [ 1.07 x 10 n/m l25 18 2 = 0 6 .080 R.%0 ~ Unirradiated 5 $.040 1.07 x 10 n/m 18 2 g .020 aT = 23 1 e) ' i I I I f f I I I f i e 3l XM DATA

SUMMARY

I i 180 _ T,,, Ty (35 az) ]60 - I. (50 n-La)~ n II t USE (Ave) - aOSE = 23 ft-lb y 140 - RT,7 t o $ 120 Unirradiat.ed l g Fr S

100 2

- L O7 x 1010 n/cm h W w tt) U. t -- 50 ho - AT = 8F "#I" 3 *l ^ OnitNTATION!#U9' *$iAZ 20 pggg,See above HEAT Numsta C-2789-2 4 i e i e i i e i e e i g '80 -40 0 40 80 120 150 200 240 280 320 360 400 Tor 1 ~ nA m t. r i 480 079 a 7-7 Babcock 8.Wilcox i

Figure 7-4. Irradiated Vs Unirradiated Charpy Impact Properties of k' eld Metal, Transverse Orientation 1n0 I Unirradiat ,3 f y50 1 j 5 2 18 n/cm 3 25 1.07 x 10 i i O .080 i Unirradiated g E 5 aT = 190F _ g _____ g' [____________ ~ e _ _ _ __ _ _ U.040 2 l .020 1.07 x 1018 n/m s I 9 1 9 I t t t a 9 i 180 ISO - 140 - g t $ 120 3 100 ~ AUSE = 3 7 f t-)b g 5 Unirradiated -5 .5 80 A s o e t I o 1 AT = 117F r/ 18 n/cm2 ~ - (,o._________c---______' b 1.07 x 0 tu %TttIAL Weld Metal (WileMTATIoss QI,.} see above pwg,,c, 20 WF-25 h nut Nuxsta V -8 -40 0 40 80 120 160 200 240' 280 320 M0 FA $? 0'0 M Test itet=ATunt, F 1480 080 Babcock 8.Wilcox P 7-8 I s

5 1 4 Figure 7-5. Irradiated Vs Unirradiated Impact Propertie: of Cor-1 relation Monitor 'brerial, Longitudinal Orit ntation 100 1 k* u 15 J Unirradiated a s B yg ______________ _ .is "J 5 25 18 2 1.07 x 10 n/cm 1i D' e 8A0 r~ g i;) $.060 ITn3rradiated 3 \\ E e.** l _ 1.07 x 1018 2 w.040 n/cm l g __ _.. _ _ _ _ _j - _ _ 7 e /. AT = 64F 'd.

i'.020 5

.'-5 i i e i e i e i e i I W i. E 180 160 [ . 140 _.sUSE = 15 ft-lb ? ? { F 120 l 7 El Unitradiat.ed m% 3300 h f

t na 50

._ J.07 x 1016 2 r/co p, t -- g _ AT = 44F i.p ,f t:0 - + i gngpgHSST-PL-02 j 0 - OnitarAf ten Ler.gitudinal Ftuence see above HEAT Nuestm A1195-1 ] i i i i e i i g -80 -40 0 40 80 120 160 200 2:43 280 320 MG 471 Tesi itw inATunt, F 7-9 Babcock & Wilcox tl. .i. e 1480 081

~. 1 8. DETERMINATION OF RCPB PRESSURE-TEMPFRATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) i l of TMI-l have been established in accordance with the requirements of 10 CFR 50, i Appendix G. The methods and criteria employed to establish operating pressure i and temperature limits are. described in topical report BAW-10046.8 The objec-f tive of ther.e limits is to prevent nonductile failure during any normal operat-h ing condition, including anticipated operation occurrences and system hydro-f(3 static tests. The loading conditions of interest include the following: O 1. Normal operations, including heatup and cooldown. 2. Inservice leak and hydrostatic testc. ? 3. Reactor core operation. The major components of the RCPB have been analyzed in accordance with 10 CFR F 50, Appendix G. The closure head region, the reactor vessel outlet nozzle, o and the beltline region have been identified as the only regions of the reactor vessel, and consequently of the RCPB, that regulate the pressure-temperature limits. Since the closure head region is significantly stressed at relatively J low temperatures (due to mechanical loads resulting from bolt preload), this y region largely controls the pressure-temperature limits of the first several service periods. The reactor vessel outlet nozzle also affects the pressure-j l temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle, which can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure, the RTNDT of the beltline region materials will be high enough that the beltline region of the reactor vessel will start I te control the pressure-temperature limits of the RCPB. For the service period for which t-he limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, the outlet nozzle, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressures. 8-1 Babcock & Wilcox l, 1480 082

The sixth full-power year was selected because the estimated second surveil-lance capsule will be withdrawn at the end of the refueling cycle, which cor-k responds to approxims'ely five full-power years. The time difference between 8 fif th and sixth full-power years provides adequate time for re-establishing the operating pressure and temperature limits for the period of operation be-tween the second and third surveillance capsule withdrawals. The limit curves for '1MI-l are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the sixth full-power year. The unirradiated impact properties were determined for the surveillance beliline region materials in accordance with 10 CFR 50, Ap-pendixes G and H. For the other beltline region and RCPB materials, the un-3 rradiat ed $2:tpact properties were estiwited using the methods described in 4 BAW-10046P.8 The unirradiated impact properties and residual elements of the 4C belt.line region mat.erials are listed in Table A-J. The adjusted reference j temperatures are calculated by adding the predicted radiation-induced ARTET and the unirradiated RTg7 The predicted ARTET is calculated using the j respective neutron fluence and copper and phosphorus contents. Figure 8-1 ( illustrates the calculated peak neutron fluence at several locations through y .4 l the reactor vessel beltline region wall and at the cent er of t.he surveillance 'j capsules as a function of exposure time. The supporting information for Figure j 8-1 ie described in BAW-10100.9 The neutron fluence values of Figure fl-1 are M the predicted fluences, which have been demonstrated (section 6) to be conser-vative. The design curves of Regulatory Guide 1.99* were used to predict the radiation-induced ARTET values as, a f unction of the mat.crial's copper and x phosphorus content and neutron fluence. The neutron fluences and adjust.ed RTET values of t.he beltline tegion materials at the end of the sixth f ull-pcm.r yem are 31sred :In Tabic fi-J. The neutron ~ f 3uenecc and adjusted RTNDT valaes are given for t he 1/47.ind 3/4T vem] wal i. M loca ti ons (T = wall thickness). The assumed RTET of the closure head region and the outlet nozzle steel forgings is 60F, in accordance wi d RAW-10046P.0 9 Figure 8-2 shows the reactor vessel's pressure-temperature lfwit curve for f normal heatup. This figure also shows the core criticality limits as required 3

Revision 1, January 1976.

1480 083 8-2 Babcock & Wilcox =

I by 10 CFR 50, Appendix G. Figures 8-3 and 8-4 show the vessel's prescure- [ cemperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively. All pressure-temperature limit curves are applicable up to the seventh effective full-power year. Protection against nonductile failure is ensured by maintaining the coolant pressure be-i low the upper limits of the pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to go crit-ical until the pressure-temperature combinations are to the right of the crit-icality limit curve. To establish the pressure-ramperature limits for protec-tion against nonductile f ailure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel control-ling the limit curves. This is necessary because the reactor vessel is the ) most limiting component of the RCPB. l 4 l i, 1 l l 1480 084 8-3 Babcock & Wilcox

O v,. p 2 S A S S A R i ui t I T* i-1,4 5 2 2 2, i H I : 3 2. k i n V* A o o o W .e o *. o I -n a .i g=p t= I i:E cm

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=.

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2 -

R R. :T R, 4 l 't 0 U ) ^ > ~ ~ m n 2 2 a4 sak k 3 k h 2 3 3 8-4 Babcock 1.Wilcox

1 Figure 8-1. Fast Neutron Fluence of Surveillance Capsule Center Compared to Various Locations Through Reactor I Vessel Wall for First 10 EFPY 10 ) I 9.5 x 1018 m nyt 9 8 7 E l E H A d ~ 1 / c e 5.3 x 1018 3 of nvt; l 8 g 5 c b 5 f 4 8 5 g ef 3 x 1038 3 60 nyt 9 40 2 u t,c W g6 g eS

S 7.1 x 1017 1

nvt ation vessel 3/47 L 7 vessel cutside Surface O i 0 1 2 3 4 5 6 7 6 9 jo Time, EFPY h 8-5 Babcock & Wilcox

Figure 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Norr.al Operation - lient.:p, Applicable for First Six Effective Full- ?ower Years 2400 8 h 4 4 e I I 2200 NDT' O G 3eltline Region 1/4T 145 2000 ~ Closure 3ead Region 60 Outlet Nozzle 60 ? g 1000 - -e ?rcssure,

Temp, 3

1600 Peint __ esi F O Applicable for A 480 60 o 4 D 625 181 up to 100F/n. g 1400 C 625 273 D ~250 302 N E 625 273 1200 F 525 313 g G 2750 342 \\ criticality ~ g 1000 Limit Ee> 800 - u 3 E 600 3 C F A g The acceptelle pressure-temerature conbine.tions are below and 400 m$ to the right of the limit curve (s). The limit cu vee do nt.t in-b g ptag clude the pressure differential between the point of system pres-0 g r s'are masumet and the nessm on ne reactor vessel ugion (gij@;[jd'b' i i i.i o 200 - n y controllire the limit cu ve, nor de they include any additional ( w nargin of.sa_'c;y for pessible instrument error. 0 ~- 40 80 c 160 200 240 2U 5 O O 320 30 400 O Reactor Vessel Coolent Temperature. P N -p MM k

. -y m - . Ww--:--mm _-,x-y w Figure 8-3. Reactor Vessel Pressure-Temperature Limit Curve for Normal Operation - Cooldown, Applicable for First Six Effective Full-Power Years 2400 I I I I I I Assumed RT 2200 NDT' Beltline Region 1/4T 145 Peltline Region 3/4T 82 2000 Closure IIead Region 60 Outlet Nozzle 60 1800 m .aa*

Pressure, Temp, f

1600-Point psi F u 3 A 180 60 B 570 110 Applicable for ~ y C 625 136 cooldown rates co D 625 205 ,8, o E 2250 230 up to 100F/h 1200 aooO T 100( m mW> 800 - w 0 0 C D E 600 B A The acceptable pressure-temperature combinations are below 9 y CD 400 do not include the pressure differential between the point - o-CD of system pressure measurecent and the pressure on the o reactor vessel region controlling the limit curve, nor do CD 200 A they include any additional margin of safety for possible p. CD CD instrument error. N I I I c o 40 80 120 160 200 240 280 320 Reactor Vessel Coolant Temperature, F . _ __ w _r.:- .._ __.,.., m - m, ,j_y-

I i Figure 8-4. Reactor Vessel Pressure-Temperature Limit Curve for l Inservice Leak and Hydrostatic Tests, Applicable for First Six Effective Full-Power Years f 2600 6 i e a i Assumed RT NDT' Beltline Pagion 1/4T 145 Beltline Region 3/4T 82 ~ 2200 - Closure Head Region 60 p Outlet Nozzle 60 2000 - l 5 Applicele f or licatup 1800 Point l to and cooldown rates up w "n A 250 70 to 50F/h j '(I ( E 1600 D 480 100 L C 625 185 6 p 1> 62h 245 a 3 1400 E 2500 268 h h 8 { U 3200 ) 1 The acceptable pressure-temperature combinations a2.e i I below and to the right of the limit curve (s). The _. limit curves do not include the pressure diff erential 1000 r' between the point of system pressure measurement and j o U the pressure on the reactor vessel region controlling I the limit curve, nor do they include any additional j[ j; 800 margin of safety for possibic int.trument error. C D 600 400 1 A 200 1,. f[ 100 10 180 220 ?O 7 00 2M i

) l heactor Vassel Coolant Temperature, F e

4 ll l l ~ d e u80 089 Babcock & Wilcox h 8-8 E5 ay

i$

I i i f li 5 9.

SUMMARY

OF RESULTS i The analysis of the reactor vessel material contained in the first surveillance capsule (TMI-1E) removed from the TMI-1 pressure vessel led to the following 3 conclusions: J 'l ) 1. The capsule received an average fast fluence of 1.07 x 1038 n/cm2 (E > 1 MeV). The predicted fast fluence for tce reactor vessel 1/4T Jocation at f the end of the first fuel cycle is 3.2 x 3017 n/cm2 (E > 1 MeV). 2. The fast fluence of 3.2 x 1017 n/cm2 (E > 1 MeV) increased the RT of g the pressure vessel core region shell materials by a maximum of 14SF. N i 3. Based on a ratio of 1.6 between the fast flux at the surveillance capsule location to that at the vessel wall and an 80% load factor, the projected i fast fluence that the TMI-1 reactor pressure vessel will receive in 40 I calendar years' operation is 1.46 x 1019 n/cm2 (E > 1 MeV). i 4. The increase in the RT for the base plate material was greater than g that predicted by the currently used design curves of ART versus fluence g because of inaccuracies of the prediction curves resulting from lack of ~ irradiation data for low-copper materials at low fluences. 5. The increase in the RT for the weld metal was greater than that predicted g by the currently used design curves of ART ecause of the inaccuracies NDT in the method of measuring the shif t in transition temperature as the data curves approach the specified limit curves. P 6. The current techniques used for predicting the change in Charpy impact upper shelf properties due to irradiation are conservative. 7. The analysis of the neutron dosimeters demonstrated that the analytical e techniques used to predict the neutron flux and fluence were cccurate. 8. The thermal monitors indicated that the capsule design was satisfactory for c:sintaining the specimens within the desired temperature range. 1480 090 ) 9-1 Babcock & Wilcox a is

r~ 1 ) ,1 I } m ii 10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE i L Based on the postirradiation test results of capsule TMI-lE, the following ~ schedule is recommended for examination of the remaining capsules in the i TMI-l reactor vessel surveillance program: J il 1, i Evaluation schedule Est. capsule Est. date(c) st. N Capsule

fluence, data ID n/cm2 Surface 3/4T available TMI-1A(a) 8.4 x 1018 13 23 1982 i

TMI-1B 1.4 x 10 J 21 38 .1985 P TMI-1C(*} 2.15 x 1019 32 55 1990 .i TMI-1D Standby f. TMI-lF Standby g TMI-lG " ' Standby k TMI-lH Standby h I, (* Capsules ontain weld metal specimens. h Capsules designated thermal aging capsules, y (")These dates do not represent the earliest. dates that datn will he available for the ma terials 1. hat control the operating limit.ations. Similar materials are included as part of the B&W Integrated Reactor Surveillance Program, which will provide necessary data on a timely basis. The earliest date that t'tese data will be available is 1980. 1 -ij 1480 091 i ik 10-1 Babcock & Wilcox C +. EY

I 1 i 11. CERTIFICATION The specimens were tested, and the data obtained from Three Mile Island Unit 1 surveillance capsule TMI-E were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements fI of 10 CFR 50, Appendixes G and 11. k ll?? A. LT'Lowe, Jr.. Date i l'roject Technica nager l This report has been reviewed for technical content and accuracy. i 68ta-2/ L 1977 K. E. Moore Date Technical Staff i : 1 1480 092 ] I 11 1 Babcock & Wilcox a(

i l 1 L f I 12. REFERENCES d l I G. J. Snyder and G. S. Carter, Reactor Vessel Material Surveillance Pro-gram, BAW-10006A; Rev. 3, Babcock & Wilcox, Lynchburg, Virginia, January 1975. g I 2 A. L. Love, Jr., et al., Analysis of Capsule OCl-E From Duke Power Company E Oconee Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1426, Babcock & Wilcox, Lynchburg, Virginia (to be published). 1 3 User's Manual for ANISN, a One-Dimensional Discreto Ordinates Transport ~ Code With Anisotropic Scattering, K-3 693 (RSIC-CCC-82), Union Carbide Corp., Nuclear Division, March 1967. J 4 ] ' User's Manual for the D6T-IIW Discrete Ordinates Transport Computer Code, l t WANL-TME-1982, December 1969. 5 CASK Group Coupled Neutron and Gamma-Ray Cross Section Data, _RSIC-DLL-23, Radiation Shielding Information Center. L 6 Draf t - New Standard E482-00, " Recommended Practice for Neutron Dosinetry e for Reactor Pressure Vessel Surveillance," October 10, 1974. h 7 W. L. Zi.ip, _ Review of Activation Methods for the Determination of Fast Neutron Spectra, Reactor Centrum nederland, Petren, May 1965. i 8 .j H. S. Palme and H. W. Behnke, Methods of Compliance With Fracture Tough-4i utss and Operalional Requi.rementa of Appendix G to 3 0 CFR 60, EAW-100MP, liabcock & Wilcox, Lynchburg, Virginia, Oct ober 1975. 7 9 H. S. Palme, G. S. Car ter, and C. L. Whitmarsh, Reactor Vessel Material Surveillance Program -- Compliance With 10 CFR 50, Appendix H, for Oconce-Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February 1975. 10 H. S. Palme, Radiation Embrittlement Sensitivity of Reactor Pressure Vessel + f . Steel, BAW-10056A, Babcock & Wilcox, Lynchburg, Virginia, August 1973. 1480 093 12-1 Babcock 8. Wilcox

7*. b J f i I f {! APPENDIX A Reactor Vessel Surveillance Program - -i Background Data and Information l I i e L v il i l1 0 1480 094

l A-1 babcock & WilCOX 1

? 1. Material Selection Data 1 The data used to select the materials for the specimens in the surveillance program, in 4.ccordance with E-185-o6, are shown in Table A-1. The locations j of these mat.erials within the reactor vessel are shown in Figures A-1 and A-2. i 2. Definition of Beltline Region The heitlit.. region of Three Mile Island Unit I was defined in accordance with i the data given in BAW-10100A. 4 3. Capsule Identification The capsules used in the Three Mile Island Unit 1 surveillance program are identified below by identification number, type, and location. i Capsule Cross Reference Data Nurber Tm 1.oca tion 4 i l YMI-1A A Upper E l TMI-1B B Lcver [. 1MI-1C A Upper THI-lD B Lower TMI-lE A Upper TMI-1F B I.ower TMI-lG A Thermal aging L TMI-lH B Thermal aging , I, 4. f,pecimens per Surveillance Capsule See Tables A-2 and A-3. 1' f i r:6. . }. i I l .I nS5 D l, i 1480 093 y . 7 A -2 Babcock 8, Wilcox D l~ 7 \\

- - g g --

sgr p g =amar g g g g;;;;g,q Table A-1., Material Selection Data for TMI-l Surveillance Project Charpy data, C, Material Bs.. re Drop Trsnsverse weight. RT ID, heat IPEIM So gg lb* 35 HLE. M E. MDT* Oienistry. 2 tNDT' ? Im t.. f t-lb @ ICF F F fr-lb F Cu F S NI wo, m tertal type _locatter ARY-59 SA508, C1 2 possie belt 101, 109, 117 (75)I*I (60)I*I 0.08 0.006 0.008 0.72 C-2789-1 SA302 Cr 8 md Dpper shell 0 50, 36, 33 (75) (40) 0.09 0.010 0.017 0.57 I C-2789-2 *I SA302 GR B Mod Upper s'e11 20 42, 37, 35(20) 90 72 98 30 0.09 0.010 0.017 0.57 I C-3307-1 SA?O2 Cr B Nd 1mer steell -10 42, 41, 29(30) 2G 60 112 20 0.12 0.010 0.016 0.35 C-3251-1 SA302 Cr B Had Emer shall -10 43, 40, 29 (75) (40) 0.11 0.012 0.013 0.50 WF-70 Weld Upper circ. 39, 35, 44 (64) (20) 0.27 0.014 0.011 0.46 (1002) WF-8 Weld Upter long. 45, 38, 30 (66) (20) 0.20 0.009 0.009 0.61 (2001/100%) WF-25(*I Weld ) tid-:1rc. -20 38, 28, 49 46 0 $1 -14 0.34 0.015 0.013 0.71 (1001) SA-1494 Weld Imer lost. 54, 25, 44 (65) (20) 0.14 0.015 0.012 0.45 (631) I S A-1526 Wald tower long. 33, 33, 33 4.66) (20) 0.36 0.016 0.012 0.60 (106t/37%) WF-70 Weld lower cire. 39, 35, 44 (66) (20) 0.27 0.014 0.0 11 0.46 (50%) WF-6 7 Weld Imer cire. 29, 35, 30 (SS) (20) 0.27 0.014 0.017 0.57 (50%) I*'Survalliance base metal A. (b)Surve111ance base set.31 B. I*ISurveillance veld metal. II to g OMh lhh From mill and quelfilcation test reports. / . os p. Items in () estimated per BAW-1M46P. 4 4 h beui idd k.c31a o O X* e s _6_. CO Ox CW Ch - - - - - ~ ~

~- Table A-2. Materials and Specimens in Upper Surveillance Capsules TMI-1A, TMI-1C, and TMI-lE No._of specimens Mat _erial description Tensile _ Charpy Weld metal, WF-25 4 8 IIAZ A, heat C-2 789-2, longitudinal 0 8 Baseline material plate A, h<.:at C-2 789-2: Longitudinal 4 Transverse 8 0 4 Correlation, HSST plate 02 0 8 Total per capsule 8 36 1 1 } Table A-3. Materials and Specimens in Lower Surveillaare Capsules TMI-1B, TMI-1D, nno TMT--lF i i 3 No. of specimens t Material description Tensil e Charpy HAZ B, heat 3307-1, longitidinal 4 30 Baseline material plate B, heat C-3307-1: Longitudinal j 4 Transverse 10 0 8 3 Correlation. HSST plate 02 0 8 Total per capsule 0 B 36 y i 6 r s. i, i 1480 097 E A-4 w Babcock & Wilcox fi L$$

Figure A-1. Location and Identification of Materials Used in Fabrication of TMI-l Reactor Pressure Vessel .a h x ( ?! Ii b " ARY 059 Nozile Belt T c WF-70 il C-2789-1 Upper Shell

- WF-8 C-2789-2 il

$+ WF-25 -= 2 SA 1494, 63% j + C-3307-1 J SA 1526, 37%

-3251-1 L wer Shell C

i4 E1 (iJF-70,50% [WF-67,50% 122T229 val Dutchman rl I J ?.] 1480 098 A-5 uabcock & Wilcox

1 1

Figure A-2. Location of Longitudirial Welds in Upper and Lower Shell Courses W s k E 11* 7 4 -X 9 Y Upper Shell V i.i Z-G + I y x i OnAxis(j, fin Axis q I I i Y Lower Shell [l 1480 099 1 A-6 Babcock & Wilcox i 1 3

l I APPENDIX B l Preirradiation Tensile Data Three Mile Island Unit 1 s 6 1[ 1480 100 Babcock & Wilcox B-1

W h

N6

Table B-1. Preirradiation Tensile Properties of Shell Plate Material, Heat C-2789-2 Specimen

temp, Strength, psi Elongation. %

8d " No. F _ Yield _ Ult. Unif. _Totti g, _ Longitudinal CC-703 RT 70,030 93,910 J3.6 16.4 67.3 705 RT 69,930 94,810 10.41 27.1 68.3 708 RT 73,430 93,210 11.99 26.4 68.3 M* " RT 71,130 93,980 12.0 26.6 67.97 "td dev,n J,630 655 J.30 0.33 0.47 CC-702 570 63,500 92,000 12.79 25.7 61.3 712 570 62,250 91,000 12.48 26.4 60.6 724 570 62,440 92,080 12.7) 25.0 54.7 Hean 570 62,730 91,690 J2.66 25.7 58.87 Std dev,n 5 550 490 0.13 0.57 2.96 Transverse CC-601 RT 70,180 93,710 11.33 18.6 62.7 602 RT 68,260 91,960 13.3 27.1 63.3 606 RT 66,830 90,810 J2.51 26.4 63.4 ,160 12.38 24.03

63. J 3 dev,n 1,370 3,190 0.81 3.85 o,31 CC-603 570 67,840 91,460 12.65 25.7 53,0 604 570 65,000 92,000 12.35 25.0 57.4 605 570 62,560 91,460 13.31 25.7 55.7 Mean 570 65,130 91,640 J2.77 25.47 55.37 1

Std dev,n 5 2,J60 260 0.40 0.33 ),g1 i I 't si i.. ,1 1 i f l 1 4 h B-2 Babcock e. Wilcox g

g. I k.' Table B-2. Preirradiation Tensile Properties of Shell Plate Material, HAZ. Heat C-2789-2

- "*E E8 "E**"*

Specimen

temp, of area, No.

F Yield Ult. Unif_._ Total Longit.udinal CC-403 RT 63,960 83,340 10.0 21.4 63.9 405 RT 65,730 82,690 7.7 18.6 62.9 h 408 RT 66,620 83,450 8.16 17.9 63.4 i Mean RT 65,440 83,160 8.6 19.3 63.4 I Std dev'n 1,110 340 0.99 1.51 0.41 i CC-401 570 62,310 79,340 '6.79 15.7 55.7 402 570 61,370 79,450 6.45 14.3 40.6 406 570 61,060 00,080 7.03 17.9 52.1 Mean 570 61,580 19,620 6.76 15.97 49.5 Std dev'n 15 530 330 0.24 1.48 6.44 l Transverse CC-303 RT 66,160 89,810 12.8 20.0 62.9 L 304 RT 73,130 97,280 11.71 26.4 63.7 i 306 RT 69,670 95,233 13.26 26.4 61.2 Hean RT 69,650 94,110 12.6 27.1 62.6 Std dev'u 2,850 3,150 0.65 1.03 1.04 CC-301 570 67,430 96,400 11.94 24.3 51.7 5 1 302 570 67,640 94,890 11.56 25.7 45.6 305 570 62,290 93,380 11.13 25.7 55.8 Mean 570 65,790 94,890 11.5 25.2 51.0 Std dev'n 15 2,470 1,230 0.33 0.66 4.19 l I il 1 J 1480 102

l

) L B-3 Babcock & Wilcox

Table B-3. Preitradiation Tensile Properties of Shell Plate Mr erial, Heat C-3307-1 Specicien p, - Streng. Psi _ Elongation % No. F Yield Ult. Unif. Total o a a, 1,on gitudinal i DD-707 RT 60,640 83,170 15.13 31.4 71.0 711 RT 60,140 83,420 13.73 30.0 70.4 716 RT 60,400 83,1:0 11.67 32.1 71.0 Mean ( RT 60,390 83,240 13.51 31.2 70.8 Std dev'n 204 130 1.42 0.87 0.28 UTF/05 570 68,360 82,340 11.06 28.6 66.4 710 570 53,660 82,250 13.89 28.6 67.9 72) 570 53,34G 82,500 13.92 29.3 67.1 Mean 570 58,450 t:2,360 12.96 28.8 67.1 Std dev'n 5 7,010 103 1.34 0.33 0.61 Transverse [ DD-601 RT 59,540 82,220 15.85 30.7 64.0 s, 607 RT 59,740 82,300 34.0 27.9 67.2 3 609 RT 58,930 82,810 14.44 29.3 5C.1 ( Mean RT 59,400 82,440 14.8 29.3 60.4 Std dev'n 350 260 0.78 3.14 7,42 f DD-602 570 54,270 82,910 13.65 27.1 57.2 604 570 55,220 82,830 13.53 25.0 59.7 608 570 54,110 83,170 14.92 29.3 61.5 Mean 570 54,530 82,970 14.0 27.1 59.5 Std dev'n 5 490 145 0.63 3.76 3.76 6 l 4 iit ii 1480 l03 I .I,. B-4 Babcock & Wilcox z ?

1 I. Table B-4. Preirradiation Tensile Properties of Shell Plate Material, HAZ, Heat C-3307-1 i

- - "E

' E" Specimen o a ea, No. F Yield Ult. Unif. Total Longitudinal i ] DD-412 RT 59,290 83,160 10.42 22.9 70.1 '] 420 RT 58,990 83,100 10.71 23.6 70.9 424 RT 58,440 83,040 11.07 23.6 70.6 l Mean RT 58,910 83,100 10.7 23.4 70.5 '1 Std dev'n 350 50 0.27 0.33 0.33 l l DD-415 570 57,060 82,080 9.47 21'.4 65.0 ai 417 570 57,770 81,670 8.72 20.7 68.7 i 419 570 77,580 82,880 5.43 21.4 67.6 Mean 570 64,140 82,210 7.9 21.2 67.1 Std dev'n !5 9,510 500 1.75 0.33 1,55 l Transverse DD-302 RT 61,260 85,510 15.0 32.1 65.0 303 RT 63,760 87,860 J 3. 64 28.6 67.9 305 RT 61,240 85,120 14.77 30.0 64.0 Mean RT 62,090 86,160 14.5 30.2 65.6 Std dev'n 1,180 1,210 0.59 1.44 1.65 DD-301 570 54,110 82,670 13.6 25.7 56.5 304 570 54,530 82,390 13.2 25.7 60.0 306 570 55,580 84,270 13.67 27.1 62.8 Mean 570 54,740 83,110 33.5 26.2 59.8 Std dev'n 5 620 830 0.21 0.66 2.58 t I l i l l 1480 104 B-5 Babcock & Wilcox 1

? Table B-5. Preirradiation Tensile Properties of Weld Metal, WF-25, Longitudinal Psi Elongation, % reng Specimen

emp, o

r a, No. F Yield Ult. Unff. Total CC-102 RT 69,240 85,300 11.52 25.7 64.9 121 RT 67,680 84,920 13.15 28.6 62.0 126 RT 70,780 88,350 13.22 25.7 61.6 Mean RT 69,230 86,190 12.6 26.7 62.8 Std dev'n 1,270 1,535 0.79 1.37 1.47 CC-109 570 67,170 82,320 8.88 20.0 52.7 116 $70 61,620 80,660 10.33 21.4 53.5 119 570 64,130 82,160 9.28 20.0 50.8 Mean 570 64,310 82,050 9.5 20.5 52.3 Std dev'n 15 2,270 1,090 0.61 0.66 1.13 Er I t, 1 1480 105 n-6 Babcock & Wilcox I

I l v i I I I j i s APPENDIX C Preirradiation Charpy Impact Data Three Mile Island Unit 1 !.( + .ll . I .I ll 1 00 106

[

C-1 Babcock & Wilcox

4 e Table C-1. Preirradiation Charpy Impact Data for Shell Plate Material, Longitudinal Orientation, L Heat C-2789-2 7 Test Absorbed Lateral I Specimen Shear i

teep, energy, expansion,

fracture, No.

F ft-lb 10-3 in. b-CC-719 361 123 71 100 f. 716 361 127 71 100 742 358 121 67 100 T= CC-764 281 122 73 100 755 280 125 70 3 753 278 122 100 i. 72 100 CC-738 203 J34 69 100 732 200 132 68 100 722 200 128 69 100 CC-761 139 120 69 760 139 116 90 69 88 757 138 133 72 100 CC-72 7 104 93 64 50 742 104 92 57 65 CC-741 80 73 58 745 80 96 12 61 703 80 74 10 52 8 CC-721 41 56 i 45 724 41 68 14 51 715 41 66 25 52 18 CC-756 26 54 43 754 26 50 3 40 759 25 42 4 35 2 CC-717 1 37 29 726 1 45 1 34 g 714 0 36 2 ~ 25 <1 CC-763 -56 14 10 r. 762 ~56 10 o i. 9 758 -56 30 o 77 1 'c W 4 3 1480 107 g. C-2 Babcock & Wilcox n.g w

l Table C-2.

Preirradiation Charpy Impact Data for Shell Plate ltterial, Transverse Orientation, Heat C-2789-2 Test Absorbed Lateral Shear Specimen tec:p,

energy, expansion,

fracture, No.

F ft-lb 10-3 in. f j CC-629 360 92 71 100 I 634 360 91 68 100 617 35 8 o' 76 100 CC-638 280 101 63 100 640 280 98 67 100 4 647 280 90 66 100 o

I CC-607 201 105 67 100 625 199 90 61 100 612 198 301 64 100
l CC-646 160 98 61 100 644 160 94 67 100 a

641 159 92 61 94 h CC-645 130 88 61 85 i 643 129 76 58 60 039 129 90 62 82 CC-648 101 57 47 35 4 642 101 68 51 40 637 101 75 53 30 CC-635 80 68 52 15 615 80 58 46 12 602 80 46 37 4 3 CC-633 41 38 32 2 619 40 33 29 2 601 40 32 28 3 l l 7 CC-636 +2 41 31 <1 a 605 +1. 0 16 15 0 P 609 +1 25 20 1 I 1480 108 l l D C-3 Babcock 8.Wilcox

]i 1 Table C-3. Preirradiation Charpy Impact Data for Shell Plate Material, HAZ, Longitudinal }p Orientation, Heat C-2789-2 t l Test Absorbed Lateral Shear i f i Specimen

temp, energy, expansion,

fracture, N.

l No. F ft-lb 10-3 in. j i f CC-418 361 115 58 100 433 361 111 62 100 440 360 85 58 100 { L' CC-453 283 128 76 100 I. l 459 283 140 70 100 l 457 282 165 74 100 CC-401 201 126 62 100 f 422 201 140 66 100 427 198 130 64 300 g CC-451 140 107 60 300 446 140 94 54 300 { CC-l.02 80 112 62 95 i 404 80 149 70 100 ~ 425 80 12 8 60 90 t i g CC-463 41 120 60 94 455 41 98 59 55 458 39 330 65 88 CC-441 1 83 43 65 425 1 80 51 60 421 1 86 43 98 CC-416 - 39 '/S 44 45 439 -39 74 45 15 411 -40 79 30 22 CC-408 -47 71 43 35 445 -49 84 45 60 461 60.6 95 61 40 464 60.4 24 18 2 CC-462 -61 64 38 25 456 ~79 12 9 3 ~ 460 -80 22 J4 7 454 -80 23 36 <1 i I i i f 1480 109 C-4 Babcock & Wilcox 1

p l l Table C-4. Preirradiation Charpy Impact Data for Shell Plate Material, HAZ, Transverse d Orientation, Heat C-2789-2 Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture,

} No. F ft-lb 10-3 in. CC-301 362 146 72 100 326 360 105 57 100 1 303 360 81 49 CC-327 201 91 50 100 1 325 201 137 10 100 324 200 89 46 100 CC-340

40 97 59 95 336 140 122 72 100 334 138 116 64 100 CC-316 81 122 61 100 321 80 114 54 55 302 80 123 64 85 306 80 120 61 85 CC-338 41 88 55 65 I

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 l [ 339 -20 76 45 15 I 337 -20 112 61 60 00-323 -39 S0 46 45 308 -39 52 35 3 [ 304 40 54 33 10 ! l CC-335 -80 36 22 2-333 -80 26 19 3 330 -80 42 32 5 1 lv = H i [] 1480 110 LI C-5 Babcock & Wilcox p 4

l j Table C-5. Preirradiation Charpy Impact Data for Shell Plate Material, Longitudinal Orientation, Heat C-3307-1 Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture, No.

F ft-lb 10-3 in. DD-732 361 213 65 100 704 360 186 63 100 757 358 160 73 100 DD-770 281 195 65 100 771 280 368 68 100 767 278 185 66 300 lip-74 2 201 368 65 300 I. 718 200 163 76 100 731 200 169 65 100 j ) DD-768 141 152 74 300 761 139 148 76 90 l 765 139 336 75 85 l DD-730 80 108 70 25

i 720 79 130 79 45 741 79 306 62 30 DD-769 56 95 74 18 l

777 56 53 46 6 764 55 56 49 4 ItD-760 40 57 46 2 754 40 53 44 2 702 40 60 49 4 DD-762 20 50 42 3 766 20 64 52 5 l 763 20 28 27 3 DD-712 +1 45 37 ) 743 +0 37 17 <1 746 +0 18 17 <3 i i i 1480 ill Babcock & Wilcox c-6

U I Table C-6. Preirradiation Charpy Impact Data for Shell Plate Material, Transverse Orientation, Heat C-3307-1 Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture, No.

F ft-lb 10-3 in. DD-650 361 106 69 100 632 359 113 72 100 646 359 102 72 100 DD-661 282 120 72 100 662 282 110 73 100 656 277 101 74 100 DD-635 201 105 68 100 617 201 114 69 100 6ri 200 116 73 100 l DT-653 149 117 72 88 654 149 103 71 92 655 149 104 68 65 i DD-664 100 63 51 25 663 99 68 52 20 657 99 70 52 20 l DD-602 81 51 43 5 633 80 66 51 12 622 80 64 43 10 l DD-658 65 47 42 6 659 65 44 40 5 660 65 46 40 8 UD-641 40 36 34 4 636 40 28 27 2 616 39 43 36 2 DD-648 1 31 27 2 612 1 36 30 <1 647 0 22 21 2 l 1480 112 C-7 Babcock & kVilcox 1

I i l i t Table C-7. Preirradiation Charpy Impact Data for Shell Plate Material, HAZ, Longitudinal Orientation, Heat C-3307-1 I Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture, No.__

F ft-lb _10-3 in. t DD-443 362 206 75 100 446 360 220 457 359 220 -t DD-460 205 179 77 100 451 200 184 68 300 423 200 187 75 100 DD-461 141 142 78 300 466 140 165 83 100 472 140 180 68 100 DD-412 80 364 67 100 410 80 124 71 65 432 79 178 76 85 DD-467 45 1. 68 50 47] 45 132 80 55 469 44 101 76 65 DD-414 2 120 75 45 408 3 122 63 35 438 0 134 67 50 DD-470 -38 52 40 15 468 -39 52 39 15 465 -40 62 42 32 DD-464 59 62 43 25 463 ~60 50 33 32 462 -61 54 35 30 DD-416 -79 125 68 45 458 -81 324 64 25 409 -80 302 63 35 I i i 1480 ii3 e i t C-8 Babcock & \\Vilcox

l. Table C-8. Preirradiation Charpy Impact Data for Shell Plate Material, RAZ, Transverse Orientation, Heat C-3307-1 Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture, No.

F ft-lb 10-3 in. DD-319 361 175 69 100 311 350 184 76 100 DD-321 282 134 80 100 327 280 148 79 100 328 278 154 66 100 DD-312 200 141 62 100 l 310 200 159 69 100 DD-326 140 126 74 80 330 140 150 76 88 I DD-313 80 167 66 85 316 79 118 58 90 317 79 166 77 100 DD-325 43 92 60 50 331 40 70 55 28 329 40 69 56 35 DD-305 1 141 66 85 301 0 116 63 40 320 0 154 76 50 DD-332 -38 88 57 18 323 -39 40 34 2 322 -40 42 33 3 3 DD-315 -79 76 50 15 3 314 -81 92 59 30 [ 304 -81 65 39 3 i l l 1480 114 I t 4 h C-9 Babcock 5 \\Milcox

I 1 I t I, Table C-9. Preirradiation Charpy Impact Data for Weld t Metal WF-25, Transverse Orientation t Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture, No.

F ft-lb 10-3 in. CC-006 359 80 63 100 051 359 01 62 100 021 358 82 60 100 CC-034 200 82 59 100 i 032 199 78 61 100 001 198 84 55 100 CC-012 80 64 47 25 003 80 70 52 55 030 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 45 023 40 54 47 55 CC-018 2 42 36 30 027 1 40 36 10 1 007 1 40 35 8 /g 020 1 42 36 30 y 4 i 1480 115 T' C-10 Babcock 8.Wilcox Q. W-(

Figure C-1. Impact Data From Unirradiated Base Metal A, Longitudinal Orientation 100

- 75 J

E O 50 - - - - - - - - - - - - - 2 E 25 w e ee e O _ p-e .080 =, a e $.060 e e e y .M0 g

e, l

".020 e ~ 8 ~ .3 8 8 e i e i e i ,9qn 200 1 DATA SUPTIARY i 180 T,37 _ 10F Ta (35 et.r) 14F/25F 160 I, (50 rr-tr.) 20F/42F c (.USE(avs) 131 It-Ibs . 140 - RT,,37 10F L e t -g 120

Ne

+ e a 8 .J 100 ) 8 Y w 80 - O t - so. / -Z----...---..-------... I e 40 - e %TERIAL SA)2 Gr D Mod. 20 - OnitarAtton ungitudinal (. pgggg None HEAT Num t, C-1789-2 0 -80 -40 0 40 80 120 ISO 200 240 280 320 360 FA IE57 IEwtaA7uar, F I 1480 116 c_11 eescocu s wiicox 4

Figure C-2. Impact Data From Unirradiated Base Metal A, Transverse Orientation 300 i e << 75 e O e 50 - - - - - ( .w 2 Y** 25 - o i e i } .080 4 A e e a 3 060 6 e Q g o e 'd.M0 xo / f-_.-__-._.___.,_'. '4 se .n*

t.020 5

.5 e e i e i ,9no I 200. DATA'SUMiUtY 380Tor - In (35 RI). 57P/72F 100T (50 n-ta). _70F/9CP y (.USE(ave) 98 ft-lb 140 - RT 30F e s WT =~ b $ 120 \\ b i .7 100 3-s e. 6 i E F e N B0 - c S b U o e t E' 00 j I. / - _ _ _ _. -.. _ _ _ _ /j i ? 40 - c i s %TrarqSA302 Gr B Mod 20 - DeathTATIO8e Transverse FLutsect None 't O + i HEAT Nunata C-2789-2 -80 -40 0 40 80 120 150 20] 243 28') 320 s om 6 list It w aparuar, F h ] [ g C-12 Babcock & Wilcox F f .r

g i 1 I Figure C-3. Impact Data From Unirradiated Base Metal A, HA2, Longitudinal Orientation o 100, e n 75 - g e e y 50 u. 3 Ym 25 e 0 .080 = l c I = e -E.060 t 8 e 8 { 3 e 4 3 -_e_f 8 .(AO F a s 5 'r-___.________________________. 1 l 020 e q ~<a t e e i e e e i i e ,p 1 200 i i DATA

SUMMARY

ygg. T,gy _Not determined 6 T (35 ru) -40F/-22F _ i :; 3 ISO To (50 FT-ts) -65F/-38P 2 g.USE (AVG) 132 ft-lb i si . 140 - RT Not datermined 1 NOT _ y e E h s 5 120 - e 3 e e .7 100 - i i i 52 k Q 3 / 60 --/-------_______________________________} / / SitalALSA302 Gr B Med 20 / DarserArron Lengitudinal Ftuenca _None NtAr Nueets C_-2783-2 0 -80 -40 0 f.0 80 120 150 200 240 230 320 50 400 i Test k w u Arunt, F 1480 118 1 c-13 Babcock 8. Wilcox

4 Figure C-4. Impact Data From Unirradiated Base Metal A, HAZ, Transverse Orientation 100 w 75 a gg s. 4 25

I 0

.030; b e e E.060 ^ e e ? O.tWO [ t - - -.g g g, -y---._..__.. a

.020 l

.5 L .000 i 200 i a DATA SU TARY yg,.1,,_Not determined la (35 n.1) -58F/+30F 160.In (50 FT La) ~48F/-15F (.(ISE(Avs) 108 ft-lb ,, 140 - RT Not determined C er f [120 5 8 i e e 6 e .t., .V 100 -( b e W iw to o e L 6 k e ~M. e le 40 E / / ~ Vita AL bb.302 Gr B Mod / 0aismiation Transverso g FLut*cr None Ntat Numta C-2789-2 0 ~, -80 -40 0 40 80 120 lbo e00 243 280 320 360 4m itsr Trvtaatunt, F C-14 Babcock 8 Wilcox 1480 119

Figure C-5. Impact Data From Unirradiated Base Metal B,- Longitudinal Orientation 100 i e } w n J E y so u. e 5 Yvi 25 e e 0 .080 4 g 3N $.060 I i .040 w 5 3 e = i.MO a i e i e e i i e e 200 DATA SUTARY ~10F =T o e 180 NOT in (35 n.r) 28F/42F 160.T (50 FT ts) 38F/50F e cy 182 ft-lb e g.USE (Avs) -10F 140 - RT,7 1 5 120 a 8 3 4 ]M i 3 i;w 80 I O 4I, e 1 60 40 - / / % taiatSA302 Gr B Moci OnitaratioN Longitadinal 20 3 FLUENCg None Hur Numsta C-3307-1 O -80 -40 0 40 80 120 ISO 200 240 2&J 320 360 400 Test TimetaArung, F c-15 a cock & Wilcox l

.~ l 1 i Figure C-6. Impact Data From Unirradiated Base Metal B, Transverse Orientation im t u 75 Y iy' w_______________ u. 5 Y 25 e m 0 .080 I I - g $.GF>0 u E tw.040 3 A- ,5 ,/

.020 a

ed i .000 2m OATA SUVARY i J20 T,,, -10F T (35 Pt.1) 50F/60F 160 I (50 FT-La) '/3F/80F cy g.USE (ava) 112 ft-lb 11 0 - RT 20F \\ 1 wor t ,aM I c I 8 i O .? 3 00 e ~e e c t. C. i i '5a Q) t; k g. f / e,0 - e p %r!RI ALSA302 Gr B Mod e / Catturation Transverse 20 e FLut=ct Hone Hrar 'tu-san _C-3307-1 i 0 i t i 80 41 0 0 46 0 83 123 150 200 20 280 329 !O c IFst ltartaArV4t, F 1480 121 C-16 Babcock & Wilcox

c' f/ I ( Figure C-7. Impact Data From Unirradiated Base Metal 3, HAZ, Longitudinal Orientation L IM

75 J

e E 4 50 - - - - - - - - 2 4 f 25 > e a e i i e i i e i i o fM ~' a 4 s. 6 i i $.060 / / i -,k/ O.040 E W.020 2 .3 i e e i ,rg i 2D i DATA SumARY y Not determined 220 NDT in (35 eu) -55F/-43F 200 In (50 tT-LA) -58F/-39P >200 ft-lb 4.USE (Avs) Not determined i . 180 - RT,7 L 3 160 I 1 a 5 3 14 s E r + 120 w U. ~ t ' 100 p / / 80 j MrtalAL SA302 Gr B Mod / FLutact None l Oni t.=rarte, longitudi nal / 60 -

gl.

Mt ar bar, C3307-1 I 40 -20 -40 0 40 80 120 150 2m 240 280 320 350 4% Test itmetaatuas, F C-17 Babcock 8. Wilcox

i Figure C-8. Impact Data From Unirradiated Base Metal B, HAZ, Transverse Orientation I IM i a e e w 75 5 E ,g50 t e 5 25 I B O' 8 . 0S0 g o 9 $.060 6 / j s e ? / j O.040 g -- _ y. /_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _.. _ _ _ _ _ _ _ _ _. ' l r 1

.020 5

5 .5 .h ,g 240 DATA 'S'JTARY g ,7 Not determined .S in (35 m) <-80F/-26F I pm.In (50 n ts) -62F/-10F q.USE bys) 150 ft-lb 180 - RT Not determined 3 g7 r i 8 i r o 6 16c-( 5 e y e g .y 140 - j. T i-1# I' 6;* 120<- C W E ? t Y " 100 - 6 e e CO - S % ira At SA302 Gr B Mod Onit=Tatieg Transverse 4 ~ / L None - - _ _ /- - _ _ _ _. g,Ut NCg ,,,,, g g, C3307-1 ,p v e r 40 4 -80 -40 0 40 80 120 150 2'v) 243 231 320 W) Mo l[] Trst iteraArung, F 1480 123 g c-18 Babcock & Wilcox <r'

I i 3 ql Figure C-9. Impact Data From Unirradiated Weld Metal, }g Transverse Orientation 100 1 -e 2 i ~ e

y. so - - _ ____ _ _ _ _

3 25 0 i i i .080 i i a s a i _a k MO r j u e 1 5 e a w.040 ). .#.020 .= .5 1 .000 i i 200 MTA MU" u i -20F t 180 mor - m I s i (35 ru) -5F/0F h cy d 160 I (50 Fr-La) 39F/46F i cy i (-USE(Avo) _81F i ~, ,, 140 - RT -14# er y h I _$120 - 5 -) 1 8 i

100 -

t 3 ~ [. 'I e 80 I .a 3 a !.i e .I - w I i 40 / I hita nt Weld Metal I 20 Oa 8 a t^r t o,, Transverse Fw act None HEAT Numeta _ WF-25 1 0 i -0 -40 0 80 120 150 200 243 280 320 MG 400 itsr TtwtaAruat, F 1480 124 c-19 Babcock 8.Wilcox

~ ~ .{. 't i l i l' i 4 APPENDIX D Threshold Detector Information Y it 1 N I 1R !} l 4 I' : .k \\480 hN 1 $1 D-1 Babcock & Wilcox

f k

Table D-1 lists the composition of the threshold detectors and the cadmium thickness used to reduce competing thermal reactions. Table D-2 shws the Cycle 1 measured activity per gram of target material (i.e., per gram of ura-nium, nickel, etc.) corrected for the wait time between irradiation and counting. Activation cross sections for the various materials were flux-weighted with a 235U spectrum (Table D-3). , Table D-1. Detector Composition and Shielding &nito rs Shielding Reaction 11.87 U-Al Cd-Ag 0.02676" Cd 238U(n, f) 1.61% Np-Al Ld-Ag 0.02676" Cd 237Np (n, f) Ni Cd-Ag 0.02676" Cd 58Ni(n p)58Co 0.66% Co-Al Cd-0.040" Cd 59Co(n,y) 60Co 0.66% Co-Al None 59Co (n,y) 6 0Co Fe None 54Fe (n.p) S4Mn = fg a

t ce ep.--k n

1480 126 4 ~~[ i i

{ ~

D-2 Babcock & Wilcox g n I

a i 1 Table D-2. TMI-1, Cycle 1 Neutron Dosimeters I Total Postirrad.

activity, pCi/g,(*)

pCi/g, , c) Monito r ut, mg Nuclide pCi total Reaction target v g. 1 A. 23aU 58.29 137Cs 9.2-3 1.58-1 238 U(n. f) FP 1.53 i 95Zr 2.26-1 3.88 3.77+1 95Nb 4.95-1 8.49 u.24+1 t 103Ru 2.93-1 5.03 4.88+1 f D'0 Ba 2.19-1 3.76 3.65+1 141Ce 2.27-1 3.89 3.78+1 i 106Ru 7.08-7 1.21 3. 3 7+1 B. 237Np 48.04 137Cs 5.28-3 1.10-1 237Np f u,1) FP 7.64 Ip 952r 1.04-1 2.16 1.50+2 i 95Nb 1.58-1 3.29 2.28+2 '1 IU3Ru 9.07-2 1.89 1.31+2 pe ol.a 2.20 4.58+1 3.18+3 140Bs 2.30-2 4.79-1 3.33+1 18 1Cc S.26-2 1.72 1.19+2 10GRu 2.76-2 5.75-) 3.99+1 136.69 5BCo 8.00+1 5.85+2 58Ni(n p)58Co 8.63+2 C. Ni t +.. 60Co 1.80-1 1.32 60Ni(n,p)60Co 5.05 l ] D. Co(Cd) 19.95 00Co 1.0) 5.06+1 59Co (n, y) 6 0Co 7.67+3 I $i E. Co 3 6.04 6000 5.40 3.37+2 59 Co(n,y)c0Co 5.13+4 h p F. Fe 155.91 5 511 4.15 2,G-1 "Fe (n t>) LD'.n 4.574'/ 59Fe B ?C 5. 7 f.-t ] MFe(n,3)De

1. '>'.H 4 v

e ED2 i A. 238U 58.53 137Cs 1.01-2 1.73.I 2 36 (n, f)FP 3.68 U 95Zr 2.82-1 4.82 4.68+1 ,I l 95Nb 5.80.1 9.91 9.62:2 103Ru 3.62-1 6.18 6.0041 t'ioBa 2.53-1 4.32 4.19+1 1480 127 D-3 Babcock & Wilcox I l

? Table D-2. (Cont'd) Total Po s.-irrad.

activity, pCi/g,(*}

pCi/ g, (b,c) Monito-wt, mg Nuclide uCi total Reaction target l '* lCe 2.76-1 4.72 4.58+1 lost.o 7.' 2 1.27 1.23+1 Np(n, f) FP 9.17 l Np 41.07 137Cs 5.41-3 1.32-1 237 B. 237 95Zr 1.47-1 3.58 2.49+2 95Nb 3,10-1 7.55 5.24+2 103Ru 1.62-1 3.94 2.74+2 L_ 140Ba 1.23-1 2.99 2.08+2 g 141Ce 1.70-1 4.14 2.88+2 106Ru 2.65-2 6.45-1 4.48+1 C. Ni 130.50 58Co 8.99+1 6.89+2 58H1(n.p)S8Co 1.02+3 60Co 2.00-1 1.53 60Ni(n.p)s0Co 5.85 D. Co(Cd) 20.35 60Co 1.26 6.14+1 59Co(n,y)60Co 9.30+3 E. Co 16.70 60Co 7.10 4.25+2 59Co (n,y) 60Co 6.44+4 F. Fe 154.62 5"Mn 4.94 3.19+1 54Fe(n p)S4Mn 5.48+2 59Fe 1.02+1 6.60&1 58Fe(n,y)S9Fe 2.00+4 ED3 A. 238U 66.47 137Cs 6.95-3 1.05-1 238 U(n,f)FP 1.02 95Zr 1.86-1 2.80 2.72+1 95Nb 3.81-1 5.73 5.56+1 103Ru 2.40-1 3.61 i 3.50+1 ~ li+ 0Ba 2.21-1 3.32 3.22+1 l ice 1.84-] 2.77 2.69+1 luGRu 5.07-2 7.63-1 7.40 Np 51.96 137Cs 3.68-3 7.13-2 237Np(n,f)FP 4.95 B. 237 95Zr 1.28-1 2.43 1.72+2 95Nb 2.61-1 5.05 3.51+2 103Ru 1.41-1 2.73 1.90+2 140Ba 1.08-1 2.09 1.45+2 1480 128 l I

( ( 'a h Table D-2. (Cont'd) l l Total 3 Postirrad. activity, pCi/g,(* pCi/g,( 'C} Monitor wt, cg Nuclide uCi total Reaction target 141Ce 1.46-1 2.83 1.97+2 c 106Ru 2.48-2 4.80-1 3.33+1 C. Ni 134.95 saco 5.68+1 4.21+2 58Ni(n.p)58Co 6.21+2 60Co 1.28-1 9.49-1 60Ni(n p)EOCo 3.63 D. Co(Cd) 20.28 60Co 7.98-1 3.93+1 59Co(n,y)60cc 3,93+3 1 E. Co 17.05 60Co 3.19 1.87+2 59Co(n,y)60Co 2.83+4 I. Fe 156.02 Si+Mn 2.99 1.92+1 54Fe (n, p) S'+Mn 3.30+2 i 59Fe 4.26 2.73+1 58Fe(n,y)S9Fe 8.27+3 A. 238U 61.43 137Cs 9.14-3 1.49-J 238U(n,f) FP 1.45 95Zr 2.39-1 3.89 3.78+1 a 95Nb 4.81-1 7.83 7.6 01-1 103Ru 3.03-1 4.93 4.78+1 140Ba 2.38-1 3.87 3.76+1 i i 1teIce 2.27-1 3.70 3.59+1 I 106Ru 5.23-2 8.51-1 8.26 i B. 237Np 51.96 137Ca 5.35-3 3.03-3 237Np(n,f)FP 7.15 95Zr 1.57-3 3.02 2.10+2 95Nb 3.41-1 6.56 4.56+2 f 103Ru 1.68-1 3.23 2.24+2 d 140Ba 9.54-2 1.84 1.28+7 i l 141Ce 1.76-1 3.39 2.35+2 l 106Ru 5.42-2 1.04 7.22+1 C. Ni 131.40 Seco 7.35+1 5.59+2 58Ni(n,p)b8Co 8.25+2 60Co 1.74-1 1,32 00Ni(n.p)60Co 5.05

l l

D. Co(Cd) 20.20 60Co 9.97-1 4.94+1 59Co(n,y)60Co 7.48+3 3 1 E. Co 16.43 60Co 5.10 3.10+2 59Co(n,y)60Co 4.70+4 I i 1480 129 j li o D-5 Babcock & Wilcox 1 il

Table D-2. (Cont'd) Total Postirrad.

activity, pC1/g,(*)

pC1/g, 'C) Monitor wt, mg Nuclide uCi total Reaction target F. 7e 151.79 54Mn 3.96 2.61+1 54Fe(n,p)S4Mn 4.48+2 59Fe 7.37 4.86+1 58Fe(n,y)S9Fe 1.47+4 (* This column is the disintegration rate per gram of wire using the post-irradiation weight. This column is the disinte ration rate per gram of target nuclide, viz. 238U, 237Np, LONi, GONi, Co, 54pe, Saye, 5

The following abundances and w2ight percents were used to calculate the disintegration rate per gram of target nuclide:

23aU 10.38 wt %, 99.27% isotopic 237Np 1.44 wt %, 100% isotopic Ni 100 wt %, 58Ni 67.77% isotopic 60Ni 26.16 isotopic Co 0.66 wt %, 59Co 100% isotropic Fe 100 we %, 5"Fe, 5.82% isotopic Safe 0.33% isotopic I . I 9 } l i i )s 1480 130 i L 1 li Babcock & Wilcox I D-6 l i

-l f I Table D-3. Dosimeter Activation Cross Sections (a) i Energy f

range, 238U 58Ni Steye G

MeV 237Np f 1 13.3 -15.0 2.231 1.073 0.460 0.425 2 10.0 -12.2 2.34 0.981 0.622 0.537 ( 3 8.18 -10.0 2.31 0.991 0.659 0.583 4 6.16 -8.18 2.09 0.917 0.638 0.572 I 5 4.96 -6.36 1.54 0.60 0.54 0.473 i 6 4.06 -4.96 1.53 0.562 0.403 0.325 d t. 7 3.01 -4.06 1.616 0.553 0.264 0.206 [ 8 2.46 -3,01 1.69 0.550 0.139 0.096 9 2.35 -2.46 1.695 0.553 0.089 0.0524 10 1.83 -2.35 1.676 0.535 0.051 0.022 11 1.11 -1.83 1.593 0.229 0.0128 0.0115 1 [ 12 0.55 -1.11 1.217 0.008 0.00048 } 13 0.111 -0.55 0.1946 0.00013 f 14 0.0033-0.111 0.0410 l 'w (*)ENDF/4 values flux-weighted with a fission spectrum. i i I 2 l 6 5 i 2 1 1 60 13I l' [ ,1 1: O p_7 Gabcock & Wilcox I, E .um,}}

ML19260A122

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