ML20095A127
| ML20095A127 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/31/1983 |
| From: | Collins L, Lowe A, Pavinich W BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20095A084 | List: |
| References | |
| BAW-1794, NUDOCS 8408210387 | |
| Download: ML20095A127 (173) | |
Text
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BAW-1794 October 1983 1
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ANALYSIS OF CAPSULE V Virginia Electric & Power Company ,
North Anna Unit No. 2
-- Reactor Vessel Materials Surveillance Program --
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BAW-1794 3
October 1983 s
ANALYSIS OF CAPSULE V Virginia Electric & Power Company North Anna Unit No. 2 l
1 -- Reactor Vessel Materials Surveillance Program --
by
( A. L. Lowe, Jr., PE L. L. Collins W. A. Pavinich W. L. Redd i C. L. Whitmarsh e
B8W Contract No. 582-7116 BABC0CK & WILC0X Utility Power Generation Division i P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox
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CONTENTS Page
- 2. POSTIRRADIATION TESTING .................... 2-1 2.1. Vi sual Examination and Inventory . . . . . . . . . . . . . 2-1
- 2. 2. The rmal Moni to rs . . . . . . . . . . . . . . . . . . . . . 2-1 2.3. Tension Testing ..................... 2-1
- 2. 4. Charpy V-Notch Impact Testing .............. 2-2 2.5. Material Identification ................. 2-2
- 3. NEUTRON 00SIMETRY ....................... 3-1 3.1. Introduction . . . . . . . . . . . . . . . . . . . . . . . 3-1 I 3. 2. An al ytic al Model . . . . . . . . . . . . . . . . . . . . . 3-2 3.3. Results ......................... 3-4 i
4 CAPSULE RESULTS . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Tens il e Prope rti es . . . . . . . . . . . . . . . . . . . . 4-1
- 4. 2. Cha rpy Impact Properties . . . . . . . . . . . . . . . . . 4-1 5 DETERMINATION OF PRESSURE-TEMPERATURE LIMITS . . . . . . . . . . 5-1
- 6.
SUMMARY
OF RESULLTS ...................... 6-1
- 7. CERTIF IC ATION ......................... 7-1
- 8. R EF E R E NC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 APPENDIXES A. Reactor Vessel Surveillance Program Background Da t a a nd In fo rma t i o n . . . . . . . . . . . . . . . . . . . A-1 B. Preirradiation Tensile Data ............... B-1 C. Preirradiation Charpy Impact Data ............ C-1 D. Threshold Detector Information . . . . . . . . . . . . . . D-1 E. LRC-TP-78 (1-26-82) Tension Testing of Solid k Round Specimens ..................... E-1 F. LRC-TP-80 (1-26-82) Charpy Impact Testing of Metal l ic Ma teri al s . . . . . . . . . . . . . . . . . . . . F-1
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List of Tables Table Page 2-1. Tensile Properties of North Anna Unit 2 Capgule V, Base and Weld Metal Irradiated to 2.41x1018 n/cm' . . . . . . . . . 2-3 2-2. Charpy Impact Data for North Anna Unit 2, Capsule V, Base Metgl, Tangential Orientation, Irradiated to 2.41x10lo n/cm2 ....................... 2-3 2-3. Charpy Impact Data for North Anna Unit 2, Capsule V, Base Metal, Ax{al Orientation, Irradiated to ,
2.41x1018 n/cm ....................... 2-4 2-4. Charpy Impact Data for North Anna Unit 2, Capsule V, Base Metal, Hegt-Affected Zone, Irradiated to 2.41x1018 n/cm' ....................... 2-5 2-5 Charpy Impact Data for North Anna Unit 2 Ca Weld Metal, Irradiated to 2.41x1018 n/cm I .psule V,
......... 2-6 3-1. Surveillance Caps ul e Detectors . . . . . . . . . . . . . . . . 3-6 3-2. Calculation Model for North Anna Unit 2 ........... 3-7 3-3. Dosimeter Activations .................... 3-8 3-4. Neutron Fl ux and Fl uence . . . . . . . . . . . . . . . . . . . 3-8 3-5. Calculated Neutron Flux Spectra ............... 3-9 3-6. Dosimetry Results ...................... 3-10 3-7. Predicted Lifetime Fluence (32 EFPY) to Pressure Vessel for E > 1 MeV ........................ 3-11 3-8. Estimated Fluence Uncertainty . ............... 3-12 4-1. Comparison of Tensil e Test Results . . . . . . . . . . . . . . 4-4 4-2. Observed Versus Predicted Chan9es in Irradiated Ch a rpy Impact P rope rti es . . . . . . . . . . . . . . . . . . . 4-5 5-1. Data for Preparation of Pressure-Temperature Limit Curves for North Anna Unit 2 -- Applicable Through 10 EFPY ..... 5-4 A-1. Unirradiated Properties and Residual Element Content I Data of Beltline Region Materials Used for Selection of Surveillance Program Materials - North Anna Unit 2 .... A-3 A-2. Heat Treatment Hi sto ry . . . . . . . . . . . . . . . . . . . . A-4 A-3. Quantitative Chemical Analysis, wt % . . . . . . . . . . . . . A-4 A-4. Specimens in Surveillance Capsules Designated S, V, W, and Z . A-5 A-5. Specimens in Surveillance Capsules Designated T, U, X, and Y . A-5 Preirradiated Tensile Properties of Forging B-1.
Material (Base Metal) and Weld Metal . . . . . . . . . . . . . B-2 C-1. Preirradiated Charpy V-Notch Impact Data for North Anna Unit 2 Reactor Pressure Vessel Surveillance Base Metal, Heat No. 990496/292424, Axial Orientation .......... C-2 C-2. Preirradiation Charpy V-Notch Impact Data for North Anna Unit 2 Reactor Pressure Vessel Surveillance Base Metal .
Heat No. 990496/292424, Tangential Orientation . . . . . . . . C-3 C-3, Preirradiated Charpy V-Notch Impact Data for North Anna Unit 2 Reactor Pressure Vessel Surveillance Heat-Affected Zone Metal . . . . . . . . . . . . . . . . . . . . . . . . . . C-4 C-4. Preirradiated Charpy V-Notch Impact Data for North Anna Unit 2 Reactor Pressure Vessel Core Region Weld Metal .... C-5 D-1. Dosimeter Specific Activities . ............... D-2 0-2. Dosimeter Activation Cross Sections ............. D-7
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List of Figures Figure Page f 2-1. Charpy Impact Data From Irradiated Base Metal, Tangential
( Orientation ......................... 2-7 2-2. Charpy Impact Data From Irradiated Base Metal, Axial
- Orientation ......................... 2-8 2-3. Charpy Impact Data From Irradiated Base Metal, Heat-Affected Zone ........................ 2-9 j 2-4. Charpy Impact Data From Irradiated Weld Metal ........ 2-10 3 -1. Fl owchart for Fl uence Analysi s . . . . . . . . . . . . . . . . 3-13 3-2. Calculation Model of the North Anna Unit 2 Reactor . . . . . . 3-14 3-3. Surveillance of Capsule Geometry in North Anna Unit 2 .. .. 3-15 3-4. Relative Fast Flux at Specimen and Dosimeter Locations in Surveillance Capsule V .................. 3-16 3-5. Axial Shape of Fast Flux at the Pressure Vessel Surface ... 3-17 3-6. Radial Gradient of Fast Flux Through the Pressure Vessel . . . 3-18 3-7. Azimuthal Gradient of Fast Flux at the Pressure Vessel Inside Surface . . . . . . . . . . . . . . . . . . . . . . . . 3-19 5-1. Predicted Fast Neutron Fluences at Various Locations Through Reactor Vessel Wall for 32 EFPY ........... 5-5 5-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Heatup, Applicable for First 10 EFPY . . . . . . . 5-6 I 5-3. Reactor Vessel Pressure-Temperature Limit Curve for Normal Operation - Cooldown, Applicable for First 10 EFPY . . . . . 5-7 5-4. Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Hydrostatic Tests, Applicable for First 10 EFPY . . . 5-8 A -1. Location and Identification of Materials Used in the Fabrication of the Core Belt Region of North Anna Unit 2 Reactor Pressure Vessel ................... A-6 C -1. Charpy Impact Data From Unirradiated Base Metal, Axial Orientation ......................... C-6 C-2. Charpy Impact Data From Unirradiated Base Metal, i Tangential Orientation . . . . . . . . . . . . . . . . . . . . C-7 C-3. Charpy Impact Data From Unirradiated Base Metal, Heat Affected Zone ........................ C-8 C-4 Charpy Impact Data From Unirradiate Weld Metal . . . . . . . . C-9 i
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- 1. INTRODUCTION This report describes the results of the examination of the 'first capsul e frcan the Virginia Electric Power Company's North Anna Unit 2 reactor vessel surveillance program. Capsule "V" is a part of the continuing surveillance program that monitors the effects of neutron irradiation on the reactor pressure vessel materials under actual operating conditions.
The specific objectives of the program are to menitor the effects of neutron irradiation on the tensile and impact properties of the reactor pressure ves-sel materials under actLal operating conditions and to verify the fluence calculations to which the materials are exposed. The surveillance program for the North Anna Unit 2 reactor pressure vessel materials was designed and rer.ommended by Westinghouse Electric Company. The surveillance program and the pre-irradiation mechanical properties of the reactor vessel material s are described in WCAP-8772.1 The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and is based on ASTM E185-73, Annex A1, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels."
This report summarizes the testing and the post-irradiation data obtained from the testing and analysis of the tensile and Charpy specimens as well as f the evaluation of the thennal monitors. In addition, the dosimeters were measured and the fl uence val ues for both the capsule materials and the reactor vessel were calculated. The wedge-opening-loading (WOL) specimens were not tested at this time, but were placed in storage to be tested later if the need f6r the data develops.
The future operating limitations established after the evaluation of the surveillance capsule are in accordance with the requirements of 10 CFR 50, Appendixes G and H. The recommended operating period was extended to 10 effective full-power years (EFPY) as a result of the first capsule evalua-tion.
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- 2. POSTIRRADIATION TESTING 2.1. Visual Examination and Inventory All specimens were visually examined for signs of abnonnalities. The con-tents of the capsule were inventoried and compared with the program charac-terization report inventory. There was no evidence of rust or of the pene-
} tration of reactor coolant into the capsule.
The inventory of all specimens was consistent with Figure 2-6 of the Westing-house preirradiation repo rt.1 However, an inconsistency was found in this report. According to Table 2-1,1 capsule V should have 8 tangential and 12
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axial Charpy specimens, but Figure 2-61 shows 12 tangential (GT series) and 8 axial (GL series) specimens. Apparently some Charpy specimens were improp-erly identi fied. Those marked "GT" are actually axial specimens, and "GL" denotes tangential specimens. With these changes in identification, the information in Table 2-1 of the Westinghouse report is correct for the iden-tification of capsule V material. Test data presented in the following para-graphs support this conclusion.
2.2. Thermal Monitors Surveillance capsule V contained a tenperature monitor holder block contain-ing two fusible alloys with different melting points. The holder block was radiographed for evaluation. Neither of the two themal monitors had melt-ed. From these data, it was concluded that the irradiated specimens had been exposed to a maximum temperature of less than 579F during the reactor vessel operating period. This is not significantly greater than the nominal inlet temperature of 550F, and is considered acceptable. There appeared to be no significant signs of a temperature gradient along the capsule length.
2.3. Tension Testing
- All tension tests were conducted in accordance wi th Technical Procedure LRC-TP-78 (see Appendix E); four specimens were tested. One axial ' base 2-1 Babcock s.Wilcox
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metal specimen was tested at 550F. The weld metal specimen was damaged in a compression accident during testing and no useful data were obtained. The other weld metal and axial base metal specimens were tested at room tempera-ture. A constant displacement rate of 0.005 in./ min was imposed on each specimen until fracture. Percent total el ongation, pe rcent reduction in cross-sectional area, 0.2% offset yield strength, and ultimate tensile strength were detennined in accordance with ASTM E8-69 and A370-73. These f data are presented in Table 2-1.
2.4. Charpy V-Notch Impact Testing All impact tests were conducted in accordance with Technical Procedure LRC-TP-80 (see Appendix F). The four groups of specimens (axially oriented base metal, transversely oriented base metal, weld metal, and heat-af fected zone metal) were tested at tempe ratures between -100 and 550F. Absorbed energy, test temperature, percent shear fracture, and lateral expansion were detennined in accordance with ASIN Specifications E23-73 and A370-73. pl ots 7
of test temperature versus absorbed energy, percent shear, and lateral ex-pansion were prepared for each specimen. These data are presented in Tables I
2-2 through 2-5 and Figures 2-1 through 2-4 2.5. Material Identification The base metal is identified in the North Anna Unit 2 reactor vessel radia-tion surveillance program report as Intermediate Shell Forging 04, Heat No.
990496/292424 The Intermediate Shell Forging in the FSAR is identified as Heat No. 990496/292429. It is conceivable that the difference is due to an error in transc ribi ng the numbers (i.e., typographical error); however, because of not having the original documentation for the material the follow-ing procedure was followed. For purposes of this report, the material in-cluded in the surveillance capsule is identified by the heat number given in the surveillance program report. The heat number given in the FSAR is used to identify the material for development of pressure-temperature limits and for describing the materials used in the fabrication of the reactor pressure vessel. No effort was made to identify which of these two heat numbers is the correct one for the surveillance material.
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Table 2-1. Tensile Properties of North Anna Unit 2 Capsule, V, Base and Weld Metal Irradiated to 2.41x1018 n/cm2 Strength Test Elongation, %
i Specimen temp, Yi el d , Ultimate, RA, No. F psi psi Uniforn Total %
Base Metal - Axial Orientation GT-4 68 85,600 104,500 7. 5 18.4 45 GT-3 550 90,100 98,700 5.1 13.3 51 Weld Metal ,
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GW-3 68 78,600 91,600 6. 2 19.7 64 Table 2-2. Charpy Impact Data for North Anna Unit 2, Capsule V, Base Metal, Tangential Orientation, Irradiated 7
to 2.41x1018 n/cm2 Test Absorbed Laterial Shear Specimen temp, energy, expansion, fract ure, No. F ft-lb mils %
GL-10 -4 0. 24.0 18.0 0.
GL-12 0. 31.0 26.0 5.
GL-15 20. 47.0 36.0 10.
GL-9 68. 69.0 51.0 30.
GL-14 220. 121.0 94.0 100.
GL-13 330. 95.0 74.0 100.
GL-11 440, 107.5 67.0 100.
GL-16 550. 102.5 81.0 100.
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Table 2-3. Charpy Impact Data for North Anna Unit 2, Capsule V, Base Metal Axial Orientation, Irradiated to 2.41x1018 n/cm2 Test Absorbed Laterial Shear Specimen t emp , ene rgy , expansion, fracture, No. F ft-lb mils %
GT-19 -10 0,. 1. 5 2.0 0.
GT-14 -40. 5. 0 - 6.0 0.
GT-18 20. 11.0 11.0 5.
GT-21 49. 23.0 66.0 10.
GT-13 68. 31.0 26.0 5.
GT-15 98. 50.0 43.0 30. I GT-16 123. 37.0 36.0 20.
GT-17 162. 43.0 43.0 30.
GT-24 220. 58.0 60.0 100.
GT-23 330. 63.0 56.0 100.
GT-22 440. 58.0 60.0 100.
GT-20 550. 60.0 61.0 100.
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Table 2-4. Charpy Impact Data for North Anna Unit 2, Capsule V, Base Metal Heat-Affected Zone,
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Trradiated to 2.41x10 8 n/cm2 Test Absorbed Laterial Shear l Specimen temp, ene rgy , expansion, fracture,
} No. F ft-lb mils %
GH-14 -100. 3.0 5. 0 5.
GH-18 -4 0, 32.0 28.0 20.
GH-24 0. 64.0 41.0 50.
GH-20 20, 68.0 48.0 60.
GH-16 40. 62.0 49.0 10.
j GH-19 68. 50.0 44.0 50. -
GH-23 98 100.0 97.0 100.
GH-17 118, 88.0 66.0 100.
GH-22 220 87.0 60.0 100.
GH-21 330. 84. 0 56.0 100.
GH-15 440. 79.0 73.0 100.
GH-13 550. 71.0 63.0 100.
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Table 2-5 Charpy Impact Data For North Anna Unit 2, Capsule V, Weld Metal, Irradiated to 2.41x1018 n/cm2 Test Absorbed Laterial Shear Specimen temp, energy, expansion, fracture, No. F ft-lb mils %
GW-19 -100. 2.0 7. 0 5.
GW-16 -40. 16.5 18.0 20. f GW-18 -12. 29.0 40.0 20.
GW-21 0. 48.5 47.0 40.
GW-17 20, 58.0 51.0 50.
GW-14 20. 63.5 53.0 40.
GW-23 68. 74.0 62.0 100.
GW-24 121. 92.0 73.0 70.
GW-20 220. 92 . 0 83. 0 100.
GW-15 330. 102.5 84.0 100.
GW-22 440. 105.0 85.0 100.
GW-13 550. 103.0 91.0 100.
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Figure 2-3. Charpy Impact Data From Irradiated Base Metal, Heat-Affected Zone
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- 3. NEUTRON DOSIMETRY 3.1. Introduction Fluence data are required to (1) provide a correlation between radiation-induced property changes and fluence for surveillance specimens, (2) deter-mine exposure of the pressure vessel at the maximum flux location, and (3) predict long-term exposure of the pressure vessel at the location of maximum
- flux and at weld locations. Fluence to the surveillance specimens is mea-sured with passive dosimeters that are included in each capsule. The cor-responding reactions in these detectors are listed in Table 3-1.
Because of a long half-life of 30 years and an ef fective energy range of
>0. 5 MeV, the measurements of 137 C s production from fission reactions in 237Np and 2380 are most applicable for analytical determinations of fast (E > 1MeV) fluence during cycle 1. The other dosimeter reactions are useful as corroborating data for shorter time intervals and/or higher energy j fl uxes. Short-lived isotope activities are representative of reactor condi-tions only over the latter portion of the irradiation period (full cycle);
whereas reactions with threshold energy >2 or 3 MeV do not record a suffi-cient fraction of the total fast flux.
The energy-dependent neutron flux is not directly available from activation detectors because the dosimeters record only the integrated effect of the neutron flux on the target material as a function of both irradiation time and neutron energy. To obtain an accurate estimate of the time-averaged neutron flux incident upon the detector, the following parameters must be known: the operating history of the reactor, the energy response of the given detector, and the neutron spectrum at the detector location. Of these parameters, the definition of the neutron spectrum is the most difficult to obtain. Essentially, two means are available to obtain the spectrum: itera-tive unfolding of experimental data and analytical methods. Due to a lack 3-1 Babcock s.Wilcox
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of sufficient threshold detectors satisfyi ng both the threshold energy and hal f-li fe requirements necessary for a surveillance program, calculated spectra are used to convert activity to flux and fluence.
The approach used in this analysis is to calculate fast flux distributions in the capsule and the reactor vessel regi ons. These calculated fluxes are normalized at the capsule by comparison of measured to calculated dosimeter activi ties. This nonnalization factor is applied to all calcul ated fluxes in the capsule and the vessel. Fluence is obtained by a time integration of flux over the capsule irradiation period. Long-tenn fluence predictions are made by adjusting flux for future fuel cycle effects and then integrating over the time period of interest. This procedure is summarized in Figure 3-1.
3.2. Analytical Model Energy-dependent neutron fluxes at the detector locations were determined by a discrete ordinates solution of the Boltzmann transport equation with the two-dimensional code, DOT 3.5.2 The North Anna Unit 2 reactor was modeled from the core to the shield tank in R-Theta geometry (based on a plan view along the core midplane and one-eighth core symmetry in the azimuthal dimen-sion; see Figure 3-2 and Table 3-2). Also included was an explicit model of a surveillance capsule at the proper location (shown schematically in Figure 3-3). The center of Capsule V was positioned 191.69 cm (75.47 inches) from the core center and 15.0* off axis. The reactor model contained the follow-ing nine regions: core, liner, bypass coolant, core barrel, inlet coolant, thennal shiel d, pressure vessel, cavity, and shield tank. Input parameters to the code included a pin-by-pin time-averaged power distri.bution, CASK 23E 22-group mic rosc opic neutron cross sections 3 S8 order of angular quadra-ture, and P3 expansion of the scattering cross-section matrix.
Because of computer storage limitations, it was necessary to use two geo-metric models to cover the distance from the core to the primary shield. A boundary source output from the initial model A (core into the pressure ves-sel) was used to " bootstrap" a model B, which included the capsule.
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Flux output from the 00T 3.5 calculations in R-0 geometry required an axial distribution adjustment to account for axial ef fects. Thus, fluxes were mul-tiplied by an axial shape factor to correlate capsule elevation and axial flux distribution. Capsule V extended about 50 inches above and below the core midplane. The axial factor ranged from 1.06 to 1.19 for various dosi-meter locations in the capsule and was 1.20 at the maximum location in the pressure vessel.
The calculation described above provides the neutron flux as a function of energy at the detector position. These calculated data are used in the fol-lowing equations to obtain the cal culated activities used for comparison with the experimental values. The basic equation for the activity D (in uCi/gm) is given as follows:
Nf 4 Dj = 1 on (E)4(E) { Fj (1 - e-Ai tj)-\jT e
f (6-1)
Aj 3.7x10 4 E j :
where N = Avogadro's number, Aj = atomic weight of target material 1,
. ft = either weight fraction of target isotope in nth material I
or fission yield of desired isotope, on (E) = group-averaged cross sections for material n, listed in Table D-3, l
4(E) = group-averaged fluxes calculated by 00T analysis, Fj =. fraction of full power during jth time interval, tj, i Aj = decay constant of ith material, tj = time interval of reactor operation, Tf = decay time from end of jth interval.
C = D (neasured) (6-2) l D(cal cula ted)
! where O(calculated) is obtained from equation 6-1.
l I
\
3-3 Babcock & Wilcox
...o. a.... ,
}
Measured activi ty is detemined for each dosimeter using established ASTM procedures. Counting ra*es, which are obtained with a nultichannel Ge(Li) gamma spectrometer, are converted to specific activity at the time of renov-al from the reactor. Calculated activity is also referenced to reactor shut-down, as indicated in equation 6-1, with power history data (fractional ful l f power versus calendar time).
I All calc'ilated fluxes are then normalized as follows:
4 = C & (calculated). (6-3) 3.3. Results cal culated activities are conpared to dosimeter measurements in Table 3-3. k The fission wire data indicate about a 6% underprediction of fast flux (E >
1 MeV) by the analytical model described herein; non-fission wire data were overpredicted by 4 to 16%. As noted previously, non-fission wire results are used only to corroborate fission wire data. Fission wire data were also corroborated with additional fission product reactions not reported here. A value of 1.06 was selected for the nomalization factor and then applied to all calculated fl uxes. Thus, fast fl ux (E > 1 MeV) was determined to be 7.40 (+10) n/cm2-s in the capsule (center dosimeter location at core mid-plane) and 5.67 (+10) n/cm2-s at the pressure vessel maximun location (Table 3-4). Corresponding fluence values for cycle 1 (376.4 EFPD at 2175 Mw) were 2.41 (+18) n/cm2 and 1.84 (+18) n/cm2, respectively. Flux exposures for E >
0.1 MeV were greater by more than a factor of 2.5. The maximum location in the pressure vessel was at the inside surface along a major axis (across flats diameter) and about 80 cm below core midplane. The capsule lead fac-tor (ratio of fast flux in the capsule to maximum fast flux in the pressure vessel) was 1.31 based on a central location of Charpy specimens. To con-vert to the capsule center, multiply all capsul e fluxes and fluences by 1.058.
In order to translate dosimeter measurements to calculations and to specimen l ocations, it is necessary to know flux gradients inside of the capsule.
Sufficient details are included in the calculational model to show a radial gradient of about 20% between front and back Cha rpy specimens and an 3-4 Babcoc.k &
. . .Wilcox
azimuthal gradient of about 5% between side-by-side specimens (Figure 3-4)..
Ef fects of the axial position can be seen in Figure 3-5 Thus, midolane values that are listed in Table 3-4 can be converted to other locations with-in a capsule by a radial factor (Figure 3-4) and a ratio of axial factors, local to midplane (Figure 3-5).
Flux spectra were calculated in the capsule, at the vessel surface, and at two locations in the vessel,1/4T and 1/2T. The data listed in Table 3-5 indicate that, relative to the capsule, the flux energy spectrum is somewhat harder (higher average energy) at the vessel surface, quite similar at 1/4T, and softer (lower average energy) at 1/2T. These spectra imply that when E
> 1 MeV fluence is used to correlate material damage, caosule specimens rep-resent metal near the vessel 1/4T position. Because of dif ferent spectral shapes at other locations, E > 1 MeV flux correlations may be less accurate.
Also of note is the significant fraction of E > 1 MeV flux at energies < 2.5 MeV, an energy range where the 54Fe(n.p) and 58Ni(n.p) reactions have little response (Table D-2). Dosimeter reaction cross sections, averaged over the capsule spectrum, are listed in Table 3-6 The corresponding fast flux that was derived from the measured activities shows a range of values with the two fission reactions at the higher end, copper at the lower end, and the nickel and iron reactions near mid-range. This same trend has been observed in previous analyses. As noted previously, analytical results were normal-J ized to the fission reactions. Also included in Table 3-6 is the themal flux of 7. 5 (+10) n/cm2-s in the capsule that was derived from bare and j Cd-covered cobalt dosimeters. This value is sufficiently low so that nea-sured reaction rates in the non-shielded dosimeters are not affected.
I Cycle 1 fluence was extrapolated to the 32 EFPY vessel design 11fe by assum-l ing proportional i ty to fast flux escaping the core from fuel management j criticality analyses of cycles 2 and 3. Cycle 3, which utilized once-hurned fuel on the core periphery, was assumed to be representative of an equili-brium cycle. This procedure accounts for the changes in relative power dis-tribution resulting from fuel shuffling between fuel cycles and effectively translates cycle 1 results to equilibrium cycle results. Li fetime fluences (32 EFPY) at several pressure vessel locations are listed in Table 3-7, and .
as noted above are based on a low-leakage fuel cycle design. Fluence as a function of penetration through the pressure vessel is shown in Figure 3-6.
35 Babcock & Wilcox
. o.o.,e . .m
To facilitate estimation of flux (and/or fluence) at other capsule l oca-tions, the azimuthal variation of fast fl ux is pl otted in Figure 3-7.
Although the data were calculated at the vessel surface, they should be applicable at the capsule radius also. Since all other capsule locations are situated at angles >15' of f axis, the corresponding lead factors will be f less than the capsule V lead factor.
Uncertainties were estimated for the fluence values reported herein. These data, listed in Table 3-8, were based on comparisons to benchmark experi-ments when available, estimated and measured variations in input data, and '
engineering judgment. Because of the complexity of fluence calculations for reactor vessel surveillance, no comprehensive uncertainty limits exist for these results. The values in Table 3-8 represent a best-estimate based on considerable experience in this type of analysis.
Table 3 1. Surveillance Capsule Detectors Detector reaction Enerqy range, Mov !sntone half-life 54Fe(n.p)S4 tin >2.5 313 days 58Ni(n.p)S8Co >2.3 71.2 days 63u(n.a)60o C C >5.0 5.27 years 2380 (n.f)137cs >1.1 30.1 years 237Np(n,f)l37Cs >0.5 30.1 years 36 Babcock & Wilcox
. ..o . ......,
{ s Table 3-2 Calculation Model for North Anna Unit 2 i Component Material Outer radius, cm(a)
Core Homogenized mixture of 161.32 002 , coolant, cladding, structure Core liner SS304 164.18 Bypass coolant Water (583F, 2250 psia, 170.02 500ppmB)
Core barrel SS304 175.10 Inlet coolant Water (547F, 2250 psia, 181.13 500ppmB)
Thermal shield $$304 187.96 Inlet coolant (b) Water (547F, 2250 psia, 199.39 500 ppm B)
Pressurevessel(C) SA508 219.29 Cavi ty Air (130F,15 psia) 236.22 Shield tank SA508 237.49 Shield (d) Water (70F,15 psia) 245.00 j (a) Measured along a major axis (across core flats diameter).
(b) Capsule is in this region.
(c) Dimension includes 40 cm liner on inner surface.
(d)Model was attenuated; partial dimension only. .
1 4
3-7 Babcock & Wilcox
, m.o a . . ,
}
Table 3-3. Dosimeter Activations A B Measured Calculated C=A/B activity, activity, Nomalization Reaction uCi/9 uCi/g Constant 5%e(n.p)S4Mn 835(a) 893(b) o,93 58Ni(n.p)58Co 1583(a) 1650(b) 0.96 63Cu(n.a)60Co 1.97 2.28 0.86 2380 (n,f)l37Cs 2.86 2.73 ~1.05 237Np(n,f)137Cs 20.6 19.3 1.07 (a) Average of dosimeter wires from Table D-2.
(b) Average of dosimeter locations in calculational model.
Table 3-4 Neutron Flux and Fluence (a)
E > 1 MeV E > 0.1 MeV Fast fluence Fluence for Fast for cyc{e 1, cycle 1, n/cm n/cm2 flug Flug (376.4 EFPD) n/cm -s (376.4 EFPD) n/cm -s Capsule, central 7.40 (+10) 2.41(+18) 2.04(+11) 6.63(+18) dosirpeter loca-tiontDi ,
Pressure vessel 5.67(+10) 1.84(+18) 1.45(+11) 4.72(+18)
(a) Estimated uncertainties are 216% at the capsule and t24% at the vessel surface.
(b) Approximate geometric center of the Charpy specimens at core midplane; i to convert to geometric center of the capsule, multiply by 1.058.
l 3-8 kW & Mcox
... n. ,
I Table 3-5 Calculated Neutron Flux Spectra r
Group Flux Normalized to E > 1 MeV(a) '
Energy range, Capsul e Vessel Vessel Vessel MeV center Surface 1/4T 1/2T 12.2-15 6.9(-4) 1.04(-3) 8.7(-4) 7.9 (-4) 10.0-12.2 2.99(-3) 4.24(-3) 3.50(-3) 3.12(-3) 8.18-10.0 8.40(-3) 1.14(-2) 9.06(-3) 7.71(-3) 6.36-8.18 2.24(-2) 2.90(-2) 2.18(-2) 1.72(-2) 4.96-6.36 4.33(-2) 5.32(-2) 4.02(-2) 3.13(-2) 4.06-4.96 4.25(-2) 5.09(-2) 3.71(-2) 2.80(-2) 3.01-4.06 7.65(-2) 8.42(-2) 6.50(-2) 5.17(-2) 2.46-3.01 9.16(-2) 9.68(-2) 8.09(-2) 6.73(-2) 2.35-2.46 3.00(-2) 3.20(-2) 2.73(-2) 2.29(-2) 1.83-2.35 1.55(-1) 1.58(-1) 1.52(-1) 1.40(-1) 1.11-1.83 4.22(-1) 3.89(-1) 4.32(-1) 4.54(-1) 1.0 -1.11 1.06(-1) 9.17(-2) 1.32(-1) 1.78(-1)
(a) 9 , flux in energy group q flux wi th E > 1 MeV J
.I 3-9 Babcock & Wilcox
...o....,
d Table 3-6 Dosinetry Results Mic roscopic Derived fast cross secti9n fluxn/cn E g.g(Mgv 1
React ion 5. b/atontai bi 54Fe(n.p)S4Mn 0.0831(c) 6.53(+10) 58Ni(n,0)58Co 0.111(C) 6.71(+(10) 63u(n.n)60Co C 9.26(-4)(c) 6.05(+10) 2380 (n f)l37Cs 0.386 7.31(+10) 237Np(n.f)l37s C 2.61 7.43(+10) 59C o(n,y)60Co 37 7.5(+10)(d)
(a)3.o' o(E) 4 (E)dE
$(E)dE l'
(b) Values are referenced to geometric center of Charpy specimens.
(c) Average values for multiple dosimeters.
(d)Themal flux derived fron bare and Cd-covered cobalt dosimeters.
3 10 M & Wh*E
Table 3-7.
Pressure Vessel for E > 1 MeVta(iPredicted Lifetine Flue Vessel surface (b) 1/47 3/47 Average fait (C) 4.9(+10) 2.9(+10) 6.9(+9) flux, n/cm z -s Fast fluence. 4.9(+19) 2.9(+19) 6.9(+18) n/cm2 (a)At an elevation about 80 cm below core midplane and on a major axis (across flats core diameter).
(b) Estimated uncertainty of 128% for maximum value at vessel surface.
"t
+(t)dt (c);.'o . fluence t time I
i I
i l
I i
i i
i
! 3 11 2 Behoo.c.k . . & WHoom
Table 3-8.
Estimated Fluence Uncertainty Estimated Input Result uncertainty parameters
- 1. Fast fluence in the t16% Activity measurement. ,
capsule. Conversion from activity to fluence.
- 2. Fast fluence in the 124% Item I vessel, maximum flux location of oxide powder.
location, capsule Location of capsule.
irradiation time Radial flux extrapolation.
interval. Azimuthal flux extrapola-tion.
Capsule perturbation of ,
fl ux.
- 3. Fast fluence in the !28% Item 2 vessel, maximum flux Time extrapolation from location, end of life. capsule irradiation to 32 EFPY.
3-12 Babcock & Wilcox
. uco n c . ,
1 i Figure 3-1. Flowchart for Fluence Analysis I Transport Dosimeter Measured '
Cross Section Cross Sections Activity l
Fission Capsule Spectrum DOT Saturated Normal- Axial
= 4 Shape '
R-e Activities ization I
Geometry M Factov i
I l 5 Average Power Power Factor l Distribution Fluence -
'l
! Vessel - For Capsule &
~
l Axial
! Shape Vessel l
Factor
' I?
U
- I Core Fluence .
i
!,l Escape : Vessel
!E long i Flux l ji Term
- x
)
i
J M
6 o
O e
n s.
O 5,. -
- .o --
2
.u e o CY.
N c
2 ._ -
x
= .x <
C 4 o a c
.C .a o - -
X u - a m, . - .- -
g z -a cm oo o s.
s=;" o
.E J v =
~
u .
m o l O
x . _ _1_ _
c _- ,
> m a a O g I l
- . .a 1 l
- l o
o a
--+--4 l u . . a I i 1
- - o -
- = n
= I I l I u " " '
--t I m T - T -- 0
- a - , .i .
w o l C x N. - o og l u
. s o
e
=
a
+ --t-4--+-- l c - - <+ , i l o o - s -
m A z l l +-
o
__+__9-_ "
l l -
o>'s>>sd -_l--
l
~
e
= +
3-14 Babcock s.Wilcox
. co....n ... ,
Figure 3-3. Surveillance of Capsule Geometry in North Anna Unit 2 3.71 =
CAPSULE SPECIMEN GulDE
_4 -3 L J 5.0o WE LD l l -
THERMA L Sh lELD
( o J
....
NOTE: ALL DIMEN31oNS ARE CM 3-15 Babcock s.Wilcox
...c....nc... ,
f Figure 3-4. Relative Fast Flux at Specimen and Dosimeter Locations in Surveillance Capsule V Specimen Specimen 1.06 0.86 Core *- $ A B C 3h 1.09 1.00 0.91 Specimen Specimen 4
Major
, Axis 1.11 0.91 A, B, C = dosimeter wire positions i
{
l l
)
l l
3-16 Babcock & Wilcox
. =co n c . ,
l Figure 3-5. Axial Shape of Fast Flux at the Pressure Vessel Surface
- 1. 3 -
1.2 -
1.1 -
> 1.0 -
fa 0.9 -
4 !*
i ,
- 0.8 -
+
- 0. 7 -
Y 5
" 4 06 e
a ttHt i N
0.5 - Dosimeter 1 2 34 5 6 y Positions I U 0.4 -
.e.
5 0.3 -
I 0.2 .
U
, f 0.1 -
, x 0 t I a a a a a a a I
! -180 -140 -100 -60 -20 0 +20 460 +100 +140 +180
- 5 g=
8 Distance From Core Hidplane, cm j Jx l
1 9'
+sgg
Figure 3-6. Radial Gradient of Fast Flux ,
Through the Pressure Vessel i
1.0
- LE AD FACTORS
.8 _
SURFACE TO T/4 1.7 3 !"
.6 _
_ i t
.4 _
T/4, R = 204.66
~
A ia 5
m j y .2 _
1 l
~
as l j i i
l 0.I _
3T/4, R = 214.41
.08 -
l
.06 !
1 1
l l
.04 I i l l '
0 5 10 is 20 l l I
' Vessel Thickness, en 3-18 Babcock s.Wilcox
. =co....n . ...,
i Figure 3-7. Azimuthal Gradient of Fast Flux at the Pressure Vessel Inside Surface i.0
.9 _
.8 -
.7 -
7 .6 -
A O.
.s q _
5
~
i
- l l
.4 -
5 E
I CAPSULE "V'
=
.3 -
.2 l I I l 0 le 20 30 40 l Angle free Major Arts, degrees l
[
i 3-19 Babcock & Wilcox
. =c o....n c ..,
i l
4 CAPSULE RESULTS 4.1. Tensile Properties Table 4-1 campares irradiated and unirradiated tensile properties. At both room and elevated temperatures, the ultimate and yield strength changes in the base metal as a result of irradiation and the corresponding changes in ductility are within the range expected. There is some strengthening, as indicated by increases in ultimate and yield strength and small decreases in ductility properties. Some of the changes observed in the data are so small as to be considered within experimental error. The relative changes in the properties of the base metal at room temperature are similar to those ob-served for the weld metal, indicating similar sensitivity as the base metal to irradiation damage. These observations are further supported by the fact that the base metal chemistry contains similar quantities of those elements contained in the weld metal that are believed to influence radiation sensi-tivity. In both cases, the changes in tensile properties are not signifi-cant relative to the analysis of the reactor vessel materials at this period in service life. Moreover the Charpy data govern the adjustments to pres-sure-temperature heatup and cooldown curves.
4.2. Charpy Impact Properties The behavior of the Charpy V-notch data is more significant to the calcula-tion of the reactor system's operating limitations. Table 4-2 compares the observed changes in irradiated Charpy impact properties to the predicted changes.
The 50 ft-lb transition shift for the base metal showed a dependence on orientation. The material properties transverse to the working direction l'xial) exhibited a greater sensitivity to radiation damage than the proper-les parallel to the working direction (tangential). In both cases, the
...ange in properties ,was poorly predicted. The weld metal shift was also 4-1 Babcock s.Wilcox
. =co ..n c . ,
I
/
significantly less than that predicted using Regulatory Guide 1.99. Simil a r ,
differences were observed for the 30 ft-lb transition temperature shift ex-cept that the magnitude of the difference between the observed and predicted increased relative to the differences at the 50 ft-lb level.
The less-than-ideal compa: . son may be attributed to a number of factors.
The spread in the data of the unirradiated material combined with a minimum i of data points to establish the irradiated material curve. Under these con-ditions, the comparison indicates that the estimated curves in Regulatory Guide 1.99 for low-copper materials and at low fluence levels are overly con-servative for predicting the tangential 50 ft-lb transition temperature shifts for all the materials.
The 30 ft-lb transition temperature shifts for the base metal is not in good agreement with the values predicted according to Regulatory Guide 1.99, al-though it would be expected that these values should exhibit better compari-
+ son when it is considered that a major portion of the data used to develop Regulatory Guide 1.99 was taken at the 30 ft-lb temperature. i The increase in the 35 mil lateral expansion transition temperature is com-pared to the shift in RTNDT curve data in a manner similar to the comparison made for the 30 ft-lb transition temperature shift. These data show a behav-i ior similar to that observed from the comparison of the observed and predict-ed 50 ft-lb and 30 ft-lb transition data.
The transition temperature measurements for the weld metal are not in agree-ment with the predicted shift. This can be attributed to the chemistry of the weld metal (low copper) as cmoared to the nominal chemistry of' normal weld metal (medium to high copper) for which the prediction curves were devel oped. This being the case, it would not be expected that the current prediction techniques would apply to this class of weld metal.
l The less than ideal comparison may be attributed in part to the combination of a spread in the data of the unirradiated material and the minimum number of data points available to establish the irradiated curve. These two varia-bles can contribute to increasing the error between the observed and pre-dicted val ues. Under these conditions, the comparison indicates that the estimating curves in Regulatory Guide 1.99 for low-copper materials at low 4-2 Babcock & Wilcox
. co n. ..,
{
fluence levels are overl'y conservative for predicting the transition tempera-ture shifts.
The data for the decrease in Charpy upper shelf energy (USE) with irradia-tion showed an interesting contrast when compared to the predicted values for the base metal . The decrease in upper shelf t.nergy for the axial orien-tation was underpredicted while the tangential properties were greatly over-predicted. By comparison, the weld metal value, although underpredicted, is in the same relative range with the predicted values. Again, in view of the lack of data for low-copper weldments at low fluence values that were used to develop the estimating curves, the predictive techniques should improve j as additional data are obtained from which better prediction curves can be devel oped.
Results from other capsules' evaluations indicate that the RTNDT estimating curves have inaccuracies at the l ow neutron fl uence l evel s (<5 x 1018 n/cm2 ). This inaccuracy is attributed to the limited data at the low flu-ence values and to the fact that the majority of the data used to define the curves in Regulatory Guide 1.99 are based on the shift at 30 ft-lb. For the materials used in reactor vessels, the shifts measured at 50 ft-lb/35 mils lateral expansion (MLE) are expected to be higher than those measured at 30 ft-lb. The significance of the shifts at 50 ft-lb and/or 35 MLE is not well understood at present, especially for materials having USEs that approach the 50 ft-lb level and/or the 35 MLE level. Fortunately, these materials, all of which exhibit this characteristic, no longer have to be evaluated at transition energy l evels of 5 J f t-lb.
The lack of consistent agreement of the change in Charpy USE 'is further sup-port of the inaccuracy of the prediction curves at the lower fluence levels.
Although the prediction curves are not always conservative in that they do not ' predict a larger drop in USE than is obser.ved for a given fluence and copper content, the lack of agreement is small and may be a function of mate-rial hi story. These data support the contention that the USE drop curves must be modified as more reliable data became available; until that time, the design curves used to predict the decrease in USE have adequate conserva-tism as currently used for regulations of the affected reactor pressure ves-s el s.
4-3 Babcock & Wilcox
. = c o....n . ..,
Ta bl e 4-1. Comparison of Tensile Test Results Room temo test Elevated temp test Unirr Irrad 550F 550F Base Metal Fluence, 1018 n/cm2 0' 2.41 0 2.41
(>l MeV)
Ult, tensil e strength, 102.0 104.5 97.2 98.7 ksi 0.2% yield strength, 84.9 85.6 75.8 90.1 ksi Uniform el ongation, % 13.2 7. 5 11.5 5.1 Total Elongation, % 19.0 18.4 18.1 13.3 RA, % 48.0 45.0 48.0 51.0 Weld Metal Fluence, 1018 n/cm2 0 2.41 0 2.41
(>lMeV)
Ult. tensil e strength, 86.0 91.6 80.1 --
ksi 0.2% yield strength, 76.1 78.6 63.9 --
ksi Unifonn elongation, % 14.2 6. 2 11.9 --
Total elongation, % 24.2 19.7 21.5 --
RA, % 69 64 62 --
4-4 Babcock & Wilcox
.uco ..n ... ,
(
Table 4-2. Observed Versus Predicted Changes in Irradiated Charpy Impact Properties f _
Material Increase in 30 ft-lb trans temp, F Observed Predicted (a)
Base material Axial 9 59 Tangential 9 59 Heat-af fected zone 10 59 Weld metal 2 46 Increase in 50 ft-lb trans temp, F Base material Axial 28 59 Tangential -4 59 Heat-affected zone 5 59 Weld metal 11 46 Increase in 35 MLE trans temp, F Base material Axial 18 59(b)
Tangential -1 59(b)
Heat-affected zone 16 59(b)
Weld metal -10 46(b)
Decrease in Charpy USE, ft-lb Base material .
Axial 13 10 Tangential 0 16 Heat-af fected zone 6 13 Weld metal 23 19 (a)These values predicted per Regulatory Guide 1.99, Revision 1.
(b) Based on the assumption that MLE as well as 30 ft-lb transi-tion temperature is used to control the shift in RTNDT-4-5 Babcock & Wilcox
.=co. a. ..,
(
l I
- 5. DETERMINATION OF PRESSURE-TEMPERATURE LIMITS The pressure-temperature limits of the reactor vessel shell course region of North Anna Unit 2 are established in accordance with the requirements of 10 CFR 50, Appendix G. The methods and criteria employed to establish operat-ing pressure and tempe rature limits are desc ribed in topical repo rt BAW-10046.4 The objective of these limits is to prevent nonductile failure during any normal operating condition, including anticipated operation occurrences and system hydrostatic tests. The l oading conditions of interest include the following:
- 1. Normal operations, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests.
- 3. Reactor core operation.
l The major camponents of the reactor vessel shell course region have been analyzed in accordance with 10 CFR 50, Appendix G. The reactor vessel outlet nozzle and the beltline region have been identified as the areas of the reactor vessel shell course region that regulate the pressure-tempera-ture limits. The reactor vessel outlet nozzle affects the pressure-tempera-ture 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 radi 6 t. ion exposure, the RTNDT of the beltl ine region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-tempe rature limits. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fl uid temperature is obtained through a i
point-by-point comparison of the limits imposed by the outlet nozzle and the beltline region. The maximum allowable pressure is taken to be the lower of two calculated pressures.
5-1 Babcock & Wilcox
. =co. a . ,
)
The limit curves for North Anna Unit 2 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at J the end of 10 EFPY. The 10th EFPY was selected because it is estinated that the second surveillance capsule will be withdrawn at the end of the refuel- )
ing cycle when the estimated fluence corresponds to approximately 10 EFPY.
The time difference between the withdrawal of the first and second surveil-
]
lance capsule provides adequate time for reestablishing the operating pres-sure and temperature limits for the period of operation between the second and third surveillance capsule withdrawals. '
The uni rradiated impact properties were dete nnined for the surveillance beltline region materials in accordance with 10 CFR 50, Appendixes G and H.
For the other beltline region materials for which the measured properties are not available, the unirradiated impact properties and residual el ements ,
as originally established for the beltline region materials, are listed in Table A-1. The adjusted reference temperatures are calculated by adding the predicted radiation-induced ARTNDT and the unirradiated RTNDT. The design curves of Regulatory Guide 1.99* were used to predict the radiation-induced ARTNDT values as a function of the material's copper and phosphorus content and neutron fl uence. Figure 5-1 illustrates the calculated peak neutron !
fluence at several locations through the reactor vessel beltline region wall as a function of exposure time. The neutron fluence values of Figure 5-1 are the predicted fluences, which have been demonstrated (section 3) to be conse rvati ve.
The neutron fluences and adjusted RTNDT values of the beltline region mate-rials at the end of the 10th EFPY are listed in Table 5-1. The neutron flu-ences and adjusted RTNDT values are given for the 1/4T and 3/4T vessel wall l ocations. The assumed RTNDT of the outlet nozzle steel fo rgi ngs is that shown in Table A-1.
Figure 5-2 shows the reactor vessel's pressure-temperature limit curve for i normal heatup. This figure also shows the core criticality limits as re- l qui red by 10 CFR 50, Appendix G. Figures 5-3 and 5-4 show the vessel 's pressure-temperature limit curve for normal cooldown and for heatup during
- Revision 1, January 1976.
5-2 Babcock s.Wilcox
. uco ..n c... ,
l
(
inservice leak and hydrostatic tests, respectively. All pressure-tempera-ture limit curves are applicable up to the 10th EFPY. Protection against nonductile failure is ensured by maintaining the coolant pressure below the f upper limits of the pressure-temperature limit curves. The acceptable pres-sure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to go criti-cal until the pressure-tempe rature combinations are to the right of the criticality limit curve. To establish the pressure-temperature limits for protection against nonductil e fail ure, the limits presented in Figures 5-2 through 5-4 must be adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel con-trolling the limit curves.
5-3 Babcock & Wilcox \
. =co., a c.....,
Table 5-1. Data for Preparation of Pressure-Temperature Limit Gurves for North Anna Unit 2 -- Applicable Through 10 EFPYta)
Weldert locatim Phstron fluence at Radiation irdral
- end of 10 EFPY. F Beltline to weld mjor ads, 1/4T Cgper Phosporus Type twton location CL. on degrees location RTET, F content 1 ccatert 1 At1/4T At 3/4T At1/4T At 3/4T At 1/4T At 3/4T Hest No.
$ SA508,C1.2 tbper shell - - - 9 0.08 0.010 8.SE18 2.1E18 46 23 55 32 SA5(B,C1.2 Internediate - - - 75 0.09 0.010 9.E18 2.I18 58 29 133 104 shell Loser shell - - - 56 0.13 0.013 9.E18 2.118 112 55 168 111
?STo.C1.2 yes -48 0.08B 0.017 9.4E18 2.118 91 45 42 -3
- Weld setal Mid. circum. -50.8 -
T A
(1001)
(8)A11 data for seterial identification, locattars, initial pr@erties, ard chenical cogosition mere obtained fron the Final Safety Analysis Report.
determined in scordance with Replatory Guide 1.99, Rev 1, 4ril 1977 (b)Adustant in RTWT i
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- 6.
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in North Anna Unit 2, Capsule V, led to the following conclusions:
- 1. The capsule received an average fast fluence of 2.41 x 1018 n/cm2 (E > 1 MeV). The predicted fast fluence for the reactor vessel inside wall location at the end of the first fuel ;ycle is 1.84 x 1018 n/cm2 (E > 1 MeV).
- 2. The fast fluence of 2.41 x 1018 n/cr.2 (E > 1 MeV) increased the RTNDT of the reactor vessel core region shell materials in the capsule to a maxi-mum of 66F based on the shell f orging axial orientation unirradiated data as shown in Figure C-1.
8
- 3. Based on the ratio of the fast flux at the surveillance capsule location to that at the vessel wall, the projected fast fluence that the North Anna Unit I reactor pressure vessel will receive in 32 EFPY operation is 4.9 x 1019 n/cm2 (E > 1 MeV).
4 The increase in the transition temperature for the base plate material was conservative compared to that predicted by the currently used design
, curves of ARTNDT versus fluence.
- 5. The increase in the transition temperature for the weld metal was not in good agreement with that predicted by the currently used design curves of ARTNDT versus fluence.
- 6. The current techniques used for predicting the change in Charpy impact upper shelf properties due to irradiation are not consistent when com-l pared with the changes observed for the various materials.
- 7. The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.
' 8. The thermal . monitors indicated that the capsule design was satisfactory for maintaining the specimens within the desired temperature range.
6-1 Babcock s Wilcox
. =co n ,
[
Power Company's No rth Anna Unit 2 surveillance Capsule V were eval ua ted using accepted techniques and established standard methods and procedures in accordance with the requirements of 10 CFR 50, Appendixes G and H.
+nA (M?b A. l.. l.oW, J r. , PE (/
PE %./W1 Date Project Technical Manager This report has been reviewed for technical content and accuracy.
G L D 0.AL1 J. p. Aadland
,a m Date Materials & Chemical Engineering l
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7-1 Babcock s.Wilcox
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APPENDIX A Reactor Vessel Surveillance Program Background Data and Infonnation
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- 1. Material Selection Datal The data used to select the materials for the specimens in the surveillance program, in accordance with E185-73 are shown in Table A-1. The locations of these materials within the reactor vessel are shown in Figure A-1,
- 2. Surveillance Materialsl The Rotterdam Dockyard Company supplied the Westinghouse Electric Corpora-tion with sections of SA508 Class 2 forging used in the core region of the North Anna Unit 1 reactor pressure vessel 'for the Reactor Vessel Material Surveillance Program. The sections of material were removed from a 10-inch lower shell course forging 04 of the pressure vessel heat treated as shown 1
in Table A-1 The Rotterdam Dockyard Company also supplied a weldment made 1 from sections of fo rging 04 and adjoining lower shell course foraing 03 using weld wire representative of that used in the original fabrication.
The forgings were produced by Rheinstahl Huttenwecke. The heat treatment history and quantitative chemical analysis of the pressure vessel surveil-lance material are presented in Tables A-1 and A-2, respectively.
- 3. Capsule Identificationi i
The capsules used in the North Anna Unit 2 surveillance program are identi- I fied and specimen tabulations are given in Tables A-3 and A-4 i b
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Table A-1. Unirradiated Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials - North Anna Unit 2*
Minimum temperatures Average for 50 ft-lb, F upper shelf, (ft-lb)
Chemist ry Parallel Normal Parallel Normal to major to major to major to major Heat Material Cu, P, NDTT working working RT NDT, Comp. No. type % % F direct. direct. F working direct. (a) working direct.
Upper A508,C1.2 0.08 0.010 +5 49 69(b) 9 86 shell Inter- A508,Cl .2 0.09 0.010 -49 49 135(C) 75(c) 85 74(c) mediate y shell w
Lower A508,C1.2 0.13 0.013 -13 31 116(c) 56(c) 92 80(c) shell Wel d Wel d 0.088 0.017 -67 12 -48 107 Heat- -49 -20 -49 125 a ffected zone g (d) Average energy at highest test temperatures (<68*F) -- % shear not reported.
,g (b) Estimated temperatures based on NRC Regulatory Standard Review Plan, Branch Technical Position MTEB 5-2.
g (c) Average transverse data obtained from surveillance program.
m e.
!I
,1 #
- Data obtained from the Final Safety Analysis Report.
3M
Table A-2. Heat Treatment History Temperature,
- Material F Time, h C ool ant Lower shell 1688-1697 2-1/2 Water-quenched forging 04, 1220-1229 6.0 Furnace-cooled Heat No. to 842 F 990496/292424 1130t25 14-3/4 Furnace cooled Weld metal 1130 t 25 -
13-1/2 Furnace-cool ed Table A-3. Quantitative Chemical Analysis, wt %
Forging 04 heat Rotterdam No990496/292324 dock W weld metal Element Westinghouse (a) analysis "" analysis C 0.19 0.19 0.08 S 0.011 0.015 0.011 N2 0.011 -
0.011 Co 0.003 0.011 <0.002 Cu 0.11 0.09 0.088 Si 0.25 0.21 0.25 Mo 0.60 0.63 0.49 Ni 0.86 0.80 0.084 Mn 0.76 0.67 1,82 Cr 0.35 0.34 0.042 V 0.031 0.02 0.002 P 0.018 0.010 0.017 Sn 0.016 - 0.004 Al 0.023 0.017 0.015 -
(a)All elements not listed are less than 0.010 wt %.
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Table A-4 Specimens in Surveillance Capsules Designated S.V,W, and Z No. of test specimens / capsule i Material description Tension CVN impact WOL
{ Forging 04 Tangential 8 t Axial 2 12 4 Heat-Affected zone 12 Weld metal 2 g Total per capsule 4 44 4 Table A-5 Specimens in Surveillance Capsules Designated T,U.X, and Y No. of test specimens / capsule Material description Tension CVN impact WOL Forging 04 Tangential 8 i Axial - 2 12 Heat-Affected zone 12 l
Weld metal 2 g 4 Total per capsule 4 44 4 I
i .
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Figure A-1. Location and Identification of Materials Used in the Fabrication of the Core Belt Region of North Anna Unit 2 Reactor Pressure Vessel
- b I[
z l N
fe -
r UPPER SHELL
- NEAT NO. 990598 I
- m 291396 INTERMEDI ATE SHE LL HEAT No. 990496 292429 WELD-NO. IDENTIFICATION LOWER SHELL NEAT NO. 990533 207355 i
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- Material identification and location obtained from the Final l Safety Analysis Report.
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APPENDIX B Pref rradiation Tensile Datal l
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Table B-1. Preirradiated Tensile Properties of Forging Material (Base Metal) and Weld Metal i
Strength, psi Elongation, %
Test temo, F Yield Ultimate Uniform Total RA, %
Base Metal RT 84,800 101,600 13.1 20.2 56 RT 85,000 102,500 13.3 17.8 40 ;
Average 84,900 102,000 13.2 19.0 48 300 77,700 95,500 10.9 16.8 46 300 77,600 94,800 10.1 16.6 53 Average 77,650 95,150 10.5 16.7 50 550 75,600 97,100 11.4 18.6 53 550 75,900 97,200 11.6 17.7 44 Average 75,750 97,150 11.5 18.1 48 Weld Metal RT 77,800 86,200 14.2 24.5 70 RT 74,400 85,700 14.2 23.9 68 Average 76,100 85,950 14.2 24.2 69 300 67,500 76,600 10.8 19.6 64 ;
300 69,500 77,600 10.6 22.2 69 Average 68,500 77,100 10.7 20.9 66 550 64,000 80,100 12.7 22.7 67 )
550 63,700 80,000 11.1 20.3 58 Average 63,850 80,050 11.9 21.5 62 !
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APPENDIX C Preirradiation Charpy Impact Datal l
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Table C-1. Preirradiation Charpy V-Notch Impact Data -
for North Anna Unit 2 Reactor Pressure Vessel Surveillance Base Metal, Heat Nc.
990496/292424, Axial Orientation Test Absorbed Lateral Shear {
Specimen t emp , ene rgy , expansion, fracture, l No. F ft-lb mils %
WL1 -50. 14.0 6.0 <5.
WL2 -50. 4.0 0.0 <5.
-50. 4.0 <5.
WL3 0. 0 WL4 15. 15.0 7.0 5.
WL5 15. 14.5 4.0 5.
I I WL6 15. 22.0 14.0 10.
WL7 75 53.0 38.0 55.
'WL8 75. 36.5 31.0 40.
WL9 75. 31.0 23.0 40.
WL10 120. 53.0 49.0 65.
WL11 120. 44.5 46.0 55.
WL12 120. 50.5 43.0 55.
WL13 210. 80.0 64.0 100.
WL14 210. 72.0 63.0 100.
! WLIS 210. 71.0 63.0 100.
WL16 300. 64.5 49.0 100.
l WL17 300. 67.0 64.0 100.
WL18 300. 76.0 63.0 100.
l C-2 Babcock & Wilcox
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Table C-2. Preirradiation Charpy V-Notch Impact Data for North Anna Unit 2 Reactor Pressure Vessel Surveillance Base Metal, Heat No.
990496/292424, Tangr_ntial Orientation Test Absorbed Lateral Shear Specimen temp, ene rgy , exoansion, fracture, No. F ft-lb ,_ qils %
WT1 -50. 33.0 2,4.0 14.
WT2 -50. 16.0 9. 0 10.
WT3 -50. 30.0 22.0 14.
WT4 15 35.0 23.0 15.
WT5 15 36.0 25.0 15.
WT6 15. 61.5 45.0 28.
WT7 50. 68.0 49.0 35.
WT8 50. 90.0 66.0 55.
WT9 50. 62.0 45.0 30.
WT10 100 37.0 35.0 45.
WT11 100. 109.0 78.0 100.
WT12 100 102.0 72.0 88.
WT13 150. 102.0 75.0 100.
WT14 150. 126.0 83.0 100.
WT15 150. 125.0 84.0 100.
WT16 212. 113.0 81.0 100.
WT17 212. 102.0 75.0 100.
WT18 212. 129.0 80.0 100.
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I Table C-3. Preirradiation Charpy V-Notch Impact Data for North Anna Unit 2 Reactor Pressure Vessel Surveillance Heat-Affected Zone Metal Test Abso rbed Lateral Shear Specimen temp, ene rgy , expansion, fracture, No. F ft-lb nils %
WH1 -115. 5.0 1.0 8.
WH2 -115 8. 0 3.0 9.
WH3 -115. 15.0 9. 0 15 WH4 -3 5. 51.0 34.0 57.
WH5 -35. 50.0 32.0 52.
WH6 -3 5. 52.0 32.0 52.
WH7 32. 27.0 26.0 30.
WH8 32. 62.0 47.0 72.
WH9 32. 90.0 59.0 81.
WH10 68. 95.0 63.0 100.
WH11 68. 94.0 67.0 100.
WH12 68. 104.0 69.0 100.
WH13 140. 76.0 59.0 100.
WH14 140. 92.0 62.0 100.
WH15 140. 113.0 70.0 100.
WH16 212. 106.0 66.0 100.
WH17 212. 122.0 71.0 100.
WH18 212. 77.0 58.0 100.
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l I Table C-4 Preirradiation Charpy V-Notch Impact Data for North Anna Unit 2 Reactor Pressure Vessel Core Reaion Weld Metal Test Absorbed Lateral Shear Specimen temp, ene rgy , expansion, fract ure, No. F ft-lb mils %
WW1 -115. 11.0 7.0 13.
WW3 -115 7. 0 3.0 12. .
WW2 -115. 6.0 1.0 9.
WW4 -55. 19.5 16.0 23.
WW5 -55. 18.5 12.0 20.
WW6 -55. 18.0 14.0 23.
WW7 -15. 43.5 39.0 41.
WW8 -15. 30.5 27.0 40.
WW9 -15, 43.0 , 35.0 42.
WW10 15. 76.5 59.0 72.
WW11 15. 69.0 55.0 70.
WW12 15 77.0 63.0 73.
WW13 68. 83.0 64.0 77.
WW14 68. 91.0 73.0 85.
WW15 68. 88.0 74.0 85.
WW16 140. 109.0 87.0 100.
WW17 140. 116.0 89.0 100.
WW18 140. 108.0 82.0 100.
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Figure C-1. Charpy Impact Data From Unirradiated Base Metal, Axial Orientation E I I I I I I J l 1 3 i L
- 75 - --
0 5
e em
= 50 -
3 en 5
5 25 - -
I i I t t i I f i C
. 08 3 g i ; g 3 g g- i i g
=
i.06 - -
E e o M
y.04-5 *
- .02- -
5
- 1 I I I I I I I I O ,-
200 i . . g g i g g i g g DATA SurtWlY 180- T,,7 -48 F Te , (35 not) *64 F g ,Tcv (50 n-La) +117F _
Tey (30 n-ta) 460F
(.USE(avs) 75 FT-L8S -
. 140 RT +57 F
- nor 2 '
gg- -
8 j100- -
3 5E-e e
W N I t 3 - 1 so 1
.-____?-. _ -_.-- -_____------ i i
e 40- l
,,, g g SA508.CL2
-____ .4- - _ _ _ - ~~
Onnentario, AX1AL 20 -
Ft.vence fl001E ,
Heat 110, 990496/292424
- i i i i i i i , , ,
o
-C0 -40 0 40 80 120 _ 160 200 240 280 320 360 400 ;
Test Tennemarunt, F l l
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r Figure C-2. Charpy Impact Data From Unirradiated Base Metal, Tangential Orientation U i i i I il 13 I I i 1
. is -
I m
yw- _ _ _ _ _ _ _ . _____________________
I 33 -
es I I I I I I I I I I I O
h
- g h
. I I I I * .I .- I I I I s .
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.!. 06 3 .
t . .
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._____ ____u_________________-
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I I I I I I I I I I I 0
' ' ' I I i l l 1 I I DATA SUMERY i 180- T,,, -48 F _
Te , (35 nu) +24F gg _ Te , (50 n-u) +30F _
Te , (30 n-u) -13F C,-USE (ave) 120 FT-LBS alW RT,,, -22F -
. 8 5 120 -
m.
I - . . .
4 100 _
i .
E 80 -
W t
60 -- * *
- 1 l
p %- . 1 e
- krunng SA508,CL2 l
- __-_-------_ --.--~
Onstatation TAllGENTI AL *" l 20 , pu ,,ec nogg .
{
) lisar flo. 990406/292424 I ' I I I I l I 0 i i s
-00 -4 0 4 80 120 _ 160 200 2W 280 320 360 400 Test foetuarunt, F C-7 Babcock s.Wilcox
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Figure C-3. Charpy Impact Data From Unirradiated Base Metal, Heat Affected Zone
^
U i i I 4 i 3 l l l l l [
- 75 I.- .
yw -- __________________________
5 5 25 -
I I I I I I I I I I I 0 .
I I I I i i l I I I I
". , e
_a g .Os_
5
- t
- .04 -
b g.02 -
=
=
J I I I I I I I I I I I 0
200 , , . g i g g g i g g f\
DATA $1N N Y 180- T,c7 IlsF _
Te , (35 nor) -6 F
-10F 3g _ Te , (50 n-u)
Te , (30 n-a) -54F c,-ust ( v.) 95 FT-L8 -
a140 RT,,, -48F
-gm- .
.:100-g .
b 80 _
W t e _
g -
l
. - - .s - ---________-.___-______-__---
~
- - hunig 8ASE METAL-HAZ FLutace N001E 20 Maar llo. 990496/292424 I I I I I I I ' ' '
I 0 360 400
-C0 -40 0 40 80 120 , 160 200 240 280 320 Trot Tampenatues, F C-8 Babcock a,Wilcox
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I Figure C-4. Charpy Impact Data From Unirradiated Weld Metal 100 g ; g , ,g g , ; , , !
.2 w
.n - .
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l APPENDIX D Threshold Detector Infortnation
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Table D-1. Dosimeter Specific Activities Plant: North Anna Unit 2 Dosimeter: Cobalt-Aluminum Dosimeter Post irrad. Radio- Nuclide Specific Act. ,(a) Activity (b,c) 1.ocation wt, g Reaction nuclide activity, uCi uCi/g pCi/g of target Bare:
Top-Co 0.00884 5%o(n,y) 60Co 1.986 224.7 149800 B ottom-Co 0.00828 5%o(n,y) 6%o 1.921 232.0 154700 Cd Shielded:
Top-Co 0.00870 5%o(n,y) 60Co 0.6770 77.82 51880 o B ottom-Co 0.00740 5%o(n,y) 60C0 0.6826 92.24 61490 h
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Table 0-1. (C ont 'd)
Plant: North Anna Unit 2 Dosimeter: Iron Dosimeter Post irrad. Radio- Nuclide Specific act.,(a) Activity (b,c)
Location wt, a Reaction nuclide activity,uCi uCi/g pCi/g of target Top 0.05553 54Fe(n.p) 54Mn 2.710 48.80 838.5 5%e(n,y) 59Fe 6.829 123.0 37270 Mid-Top 0.05531 54Fe(n,p) 2.590 54Mn 46.83 804.6 5%e(n,Y) 59Fe 6.790 122.8 37210 3
Mid 0.05238 54Fe(np) 54Mn 2.571 49.08 843.3 5%e(n,Y) 59Fe 6.132 117.1 35480 Mid-Bottom 0.05601 54Fe(n,p) 54Mn 2.691 48.04 825.4
]
5%e(n,Y) 59pe 6.868 122.6 37150 Bottom 0.05215 54Fe(n p) 54Mn 2.612 50.09 860.7 4
5%e(n,y) 59Fe 6.272 120.3 36450 i
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Table D-1. (C ont 'd)
Plant: North Anna Unit 2 Dosimeter: Copper and Nickel Dosimeter Post irrad. Radio- Nuclide Specific act..(a) Activity (b,c) location wt, o Reaction nuclide activity,uCi u Ci/q uCi/g of target C opper:
Mid-Top 0.06459 63Cu(n,a) 60Co 0.08504 1.317 1.825 Mid 0.06637 63Cu(n.a) 60Co 0.09286 1.399 2.045 Mid-bottom 0.05936 63Cu(n,a) 60Co 0.07879 1.327 1.940 Nickel:
o Mid-Top 0.06532 58Ni(n.p) 58Co 69.22 1060 1564 60Ni(n.p) 60C0 0.1247 1.909 7.297 Mid 0.05526 58Ni(n,p) 58Co 60.36 1092 1611 60Ni(np) 60Co 0.1095 1.982 7.576 Mid-botton 0.05658 58Ni(n.p) 58Co 60.30 1066 1573 60Ni(n,p) 60Co 0.1086 1.919 7.336
.K sR e x-i s=
a
Table D-1. (Cont'd)
Plant: North Anna Unit 2 Dosimeter: Uranium and Neptunium Dosimeter Post irrad. Radio- Nuclide Specific act. ,(a) Activity (b,c) 1.ocation wt, o Reaction nuclide activity, aCi pCi/g pC1/g of target 238 0.00533 2380 (n,F) 95Zr 66.2 U308 0.3531 78.2 103Ru 0.4207 78.9 93.2 106Ru 0.09546 17.9 21.1 137Cs 0.01289 2.42 2.86 144Ce 0.2149 40.3 47.6 237Np02 0.00718 237Np(n,F) 95Zr 3.846 536 609 7 103Ru 3.870 539 612 106Ru 0.7762 108 123 137Cs 0.1296 18.1 20.6 144Ce 1.840 256 291 1
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Table D-1. (C ont 'd)
Dosimeter Material Data (a)These data are the disintegration rates per gram of wire as of 1200 hrs, 7 March 1982.
viz., 238 0, 237Np, 58Ni, 60Ni, 59Co, (b) hese 4 Fe, and data S hre e, the disintegration rates per gram of target nuclide:
(c)The following abundances and weight percents were used to calculate the disintegration rate per gram of target element:
2380 - 84.8 wt %; 99.9% target nuclide 237Np - 88.1 wt %; 99.9% target nuclide Ni - 100 wt %; 67.77% 58Ni target nuclide 26.16 60Ni target nuclide o Co - 0.15 wt %; 100% 59Co target nuclide Fe - 100 wt %; 5.82% 54Fe target nuclide 0.33% 58Fe target nuclide Cu - 100 wt %; 68.4% 63Cu target nuclide F
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Table D-2. Dosimeter Activation Cross Sections (a)
Energy range, i G MeV 237Np 2380 58Ni 53pe 63Cu 1 13.3 -15.0 2.323 1.050 0.4830 0.4133 4.478 (-2) 2 10.0 -12.2 2.341 0.9851 0.5735 0.4728 5.361 (-2) 3 8.18 -10.0 2.309 0.9935 0.5981 0.4772 3.378 (-2) 4 6.36 -8.18 2.093 0.9110 0.5921 0.4714 1.246 (-2) 5 4.96 -6.36 1.541 0.5777 0.5223 0.4321 3.459(-3) 6 4.06 -4.96 1.532 0.5454 0.4146 0.3275 6.348 (-4) 7 3.01 -4.06 1.614 0.5340 0.2701 0.2193 7.078 (-5) 8 2.46 -3.01 1.689 0.5272 0.1445 0.1080 3.702 (-6) a 9 2.35 -2.46 1.695 0.5298 9.154 (-2) 5.613 (-2) 6.291 (-7) 0 10 1.83 -2.35 1.677 0.5313 4.856 (-2) 2. 940 (-2) 1.451 (-7) 11 1.11 -1.83 1.596 0.2608 1.180 (-2) 2.948 (-3) 1.317 (-9) i 12 0.55 -1.11 1.241 9.845 (-3) 6.770 (-4) 6.999 (-5) 0 l 13 0.111 -0.55 0.2341 2.432 (-4) 1.174 (-6) 1.578 (-8) 0 14 0.0033 -0.111 0.0069 3.616 (-5) 1.023 (-7) 1.389 (-9) 0 j (a)ENDF/BV values flux-weighted with a fission spectrum combined with a 1/E intermediate energy distribution.
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l APPENDIX E LRC-TP-78 (1-26-82)
Tension Testing of Solid Round Specimens l
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ac 373-3 . I sing RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE DATE I-8-82
- 1. Introduction This Technical Procedure describes the requirements for tension I testing of solid, circular cross-section, metal specimens at room and elevated temperatures. ,Unless otherwise indicated, the room temperature tests conform to ASTM E8-81, " Standard Methods of Tension Testing of Metallic Materials," and the elevated temperature tests conform to ASTM E21-79, " Standard Recommended Practice for Elevated Temperature Tension Tests of Metallic Materials."
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l E-2 enoccouna uo. LRC-TP PAGE NO. lof 9
RC 372-3 11ns RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE DATE
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- 2. Operator Qualification Tension test operators shall be qualified by training and experience and shall have demonstrated competence to perform the tests in accordance with this procedure to the satisfaction of the LRC Project Leader.
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E-3 enoccount no. LRC-TP-78 rAct No. 2 or 9
sc'aia-2 t u?5 RESEAHCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE
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- 3. Test Specimen Test specimens shall be solid, of circular cross-section, and generally conform to the shapes described in ASTM E8-81. Figures 8 and 9. For elevated temperature testing, specimens shall conform to the additional requirements of ASTM E21-79 Paragraph 7.6. Specimen length and diameter may be varied in accordance with project needs, however, the ratio of gage length to diameter shall be 4:1. Specimen ends beyond the gage length shall be of a shape and size compatible with the intended gripping device.
The gage length shall be bounded by fiducial marks such that the fiducial marks may be measured after test to determine extension of the gage length.
The method of placing fiducial marks shall be determined by the LRC Project Leader.
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E-4 rnocrount No. LRC-TP-71 PAGE No.
3 of <
ac an-3 l ts/n RESEARCH AND DEVELOPMENT DIVISION l TECHNICAL PROCEDURE DATE 1-8-82 j l
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- 4. Equipment 4.1 Tension Testing Machine 4.1.1 The tension testing machine shall be in conformance with the require-ments of ASTM E8-81 Paragraph 5.1.
4.1.2 The machine shall have sufficient capacity to load specimens to ,
i failure. ;
4.1.3 Gripping devices shall generally conform to ASTM E8-81. Figure 2 or 3 except that clevis-type grips to accommodate square-end pin-loaded specimens are permitted.
4.2 Extensometry Extensometers shall be of class B-2 or better (ASTM E83-67).
4.3 Heating Apparatus 4.3.1 For elevated temperature testing, a furnace or oven shall be provided to heat specimens uniformly such that test temperature will be maintained to within 1 SF along the gage length of the specimen.
4.3.2 Temperature measurement shall be by thermocouples suitable for the temperature of the test. Thermocouple output shall be measured by potentiometers. The complete temperature measurement system shall be of sufficient accuracy and sensitivity to monitor the test temperature within the specified limits.
t E-5 PMoctDURE No.LRC-TP-78 raos No. 4 of 9
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tum RESEARCH AND DEVELOPMENT DIVISION
. TECHNICAL PROCEDURE DATE
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- 5. Calibration 5.1 Load The load cell and its associated electronics shall be calibrated against standards traceable to the National Bureau of Standards (NBS) at least
. annually by a manufacturer's representative or similarly qualifed person.
5.2 Stroke The stroke (displacement) monitoring portion of the tension test machine shall be calibrated at least annually against standards traceable to the NBS by a manufacturer's representative or similarly qualified person.
5.3 Strain -
Extensometers, signal conditioners, and output recorders shall be cali-brated at least annually against NBS traceable standards.
5.4 Temperature ,
(, 5.4.1 Thermocouples or thermocouple wire shall be purchased with cali-bration traceable to NBS standards. Thermocouples shall be re-placed at least after one year of service.
5.4.2 Potentiometers shall be calibrated at least annuallysagainst stand-ards traceable to the NBS.
5.5 Dimensional Measurement 5.5.1 Micrometers used for dismetral measurement shall be calibrated at least annually in accordance with LRC-TP-44 (2/7/78).
i i 5.5.2 Calipers used for measurement of gage length changes in displacement
! shall be calibrated at least annually against NBS traceable standards.
5.5.3 Remotely operable measurement devices, such as video-micrometers, shall be calibrated at least annually against NBS traceable standards.
E-6 enoesouna mm.LRC-TP-78 paos no. 5 of 9
3 312-3 M5 CESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE 1-8-82 out
- 6. Tension Test Procedure 6.1 Specimen Measurement The minimum cross-sectional area of the reduced section of the specimen shall be determined. The diameter at the ends of the reduced section shall not be less than the diameter at the center of the reduced section.
Diameters shall be d'etermined with a micrometer or comparable instrumenta-tion.
I 6.2 Evaluated Temperature Test Set-up For elevated temperature testing, the test set-up shall be in accordance with ASTM E21-79, paragraphs 9.3, 9.4, and 9.5.
6.3 Strain Rate .
Except if designated otherwise by the LRC Project Leader, the strain rate shall be maintained at 0.005 2 0.002/ min to the 0.2% offset yield strength.
The strain rate shall then be increased to 0.05 2 0.01/ min.
6.4 Temperature Test temperature shall be as designated by the LRC Project Leader. Speci-mens shall be held at test temperature for at least 20 minutes before testing to assure achievement of thermal equilibrium.
E-7 emoctount No. LRC-TP-78 raos No. 6 of 9
08 DDD D sins RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE ong 1-8-82
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- 7. Calculations 7.1 Yield Strength 7.1.1 Yield strength shall be determined by the 0.2% offset method.
7.1.2 The 0.2% offset is defined es a displacement equal to 0.002 times the specimen, gage length.
7.1.3 Yield strength shail be determined from the load-displacement chart by drawing a straight line parallel to th'e elastic portion of the load-displacement curve extending from 0.2% offset on the zero load axis. The load at the intersection of this straight line with the load-displacement curve shall be taken as the " load at yield." This value shall then be divided by the original cross-sectional area to obtain the yield strength.
7.2 Tensile Strength 7 Tensile strength shall be determined by dividing the maximum load carried by
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the specimen during the tension test by the original cross-sectional area of the specimen.
7.3 Total Elongation 7.3.1 The change in length of the " gage length" due to testing shall be determined by fitting together the broken ends of the specimen and measuring the distance between the gage marks to, at least, the nearest 0.01 inch and subtracting the gage length from this value.
7.3.2 Total elongation shall be expressed as the percentage increase in gage length and obtained by dividing the change in length of the
" gage length" by the gage length and multiplying by 100.
7.4 Uniform Elongation The displacement between the onset of plastic deformation and the maximum load shall be determined from the load-displacement chart. Uniform elongation shall be expressed as a percentage and obtained by dividing this displacement by the gage length and multiplying by 100.
E-8 rnocrovnt No. LRC-Tp-5 raus No. 7 of
hG 37J.3 s gafa RCSEARCH AND DEVELOPMENT, DIVISION
, , TECHNICAL PROCEDURE ong 1-8-82 I
L 7.5 Reduction of Area r 7.5.1 The change in specimen diameter due to testing shall be determined by fitting together the broken ends of the specimen and measuring
, the diameter at the smallest section. This measurement shall be
- made at three different angular orientations and the mean of the three taken as a valid measurement. The value shall then be used to determine a cross-sectional area that shall be subtracted
)
from the original cross-sectional area to obtain the decrease in
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cross-sectional ares due to testing. ,
l 7.5.2 Reduction in area shall be expressed as the percentage decrease in cross-sectional area and obtained by dividing the decrease in cross-sectional area by the original cross-sectional area and multiplying
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E-9 emoceouns wo.LRC-TP-78 paos no. 50t 9
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RC.372-3 .
W 75 RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE oug 1/26/82 i
- 8. Test Record
- 8.1 The test record shall contain the'following results or information:
(a) Project title and account number (b) Test machine identifier (c) Date of test (d) Operator (e) Specimen identifier (f) Specimen dimensions i (g) Gage length I (h) Test temperature (i) Speed of testing (j) Specimen dimensions after test (k) Change in gage length ;
(1) 0.2% offset yield strength (m) Tensile strength (n) Total elongation (o) Uniform elongation l
(p) Reduction in area 8.2 The test report shall include any autographic, load-displacement charts obtained in the test.
( 8.3 The designation of the procedure followed, including revision identification, shall be recorded on the test record and the test record shall be countersigned by the LRC Project Leader.
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E-10 PmocEouRE No.LRC-TP-78
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f APPENDIX F LRC-TP-80 (1/26/82) l Charpy Impact Testing of Metallic Materials i
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Mc 372-3 gens RESEARCH AND DEVELOPMENT DIVISION 3 10/2U81 TECHNICAL PROCEDURE oars
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- 1. Introduction ]
This Technical Procedure describes a method for notched-bar impact testing of metallic materials by the Charpy apparatus and conforms to ASTM E23-81,
" Standard Methods for Notched Bar Impact Testing of Metallic Materials."
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emoctount No. LRC-TP-80
' "' 1 Of 11 F-2
RC 373-3 t uts RESEARCH AND DEVELOPMENT DIVIslON TECHNICAL PROCEDURE oan 10/27/81 L .
- 2. Operator Qualification Charpy impact test Operators shall be qualified by training and experience and shall have demonstrated competence to perform the tests in accordance with this Procedure to the satisfaction of the LRC Project Leader.
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l rnoctount No. LRC-TP-RO
- a - 2 o f 11 F-3
NC 312 3 isn5 RESEARCH AND DEVELOPMENT DIVISION f TECHNICAL PROCEDURE oare 10/27/81 i
- 3. Specimen 3.1 The specimen shall be a Charpy impact test specimen. Type A (V-notch),
as shown in ASTM E23-81 Figure 6.
3.2 Subsize specimens may be used if so designated by the LRC Project Leader. If subsize specimens are used, specimen size and the justi-fication for such use shall be noted on the test report.
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paocroune No. LRC-TP-80 paos no. 3 of 11 F-4
RC 37203 11/75 C'ESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE DATE 10 /27 /R1 4.. Equipment 4.1 Testing Machine The Charpy impact test machine shall be a pendulum type of rigid construction and of capacity at least 125% of that required to break each specimen. The machine shall be rigidly mounted to either a concrete floor not less than 6 inches thick or a base of a weight at least 40 times that of the pendulum. The machine shall conform to ASTM E23-81, Section 4.
4.2 Self-centering Tongs A set of tongs, as shown in ASTM E23-81. Figure 14 shall be used to. center the specimens within the testing machine's anvils.
4.3 Instrumented Data Acquisition The impact machine shall be equipped with a "Dynatup" instrumented tup, signal conditioning unit, and data storage unit suitable for recording both load and energy versus time information.
4.4 Temperature Conditioning Heating and cooling chambers shall be of the liquid immersion type.
The bath temperature measuring device shall be placed near the cen-ter of a group of immersed specimens. The fluid shall be agitated continuously. The heating and cooling chambers shall have a grid raised at least 1 inch from the bottom to provide a specimen sup-port. The fluid shall be of sufficient depth so that'the specimens when immersed shall be covered by the liquid by at least 1 inch.
Desired temperature shall be maintained within 22F (elC) and
- measured with mercury-in-glass thermometers conforming to ASTM specification El.
PmoCEouRE No. I.RC-TP A0 PAGE NO. 4 Of 11 F-5
RC 3 U-2 '
sing RESEARCH AND DEVELOPMENT DIVISION TECHNI. CAL PROCEDURE o,rg 10/27/81 i
- 5. Inspections and calibrations 5.1 Impact Machine 5.1.1 Initial Inspection At installation, the test machine shall be inspected to assure conformity with the requirements of ASTM E23-81.
5.1.2 Annual Qualification The impact tester shall be qualified by testing standardized specimens and obtaining evaluation from the Army Materials and Mechanics Research Center (Watertown, Massachusetts) at l
. least annually. If the machine is moved, repaired, adjusted, or if there is any reason to doubt the accuracy of the results, or if the pendulum is stopped by a specimen during test, or if the test value obtained exceeds 80% of the test machine's capacity, then the tester shall be requalified without re-gard to the time interval.
5.1.3 Daily check i
' (a) At least once daily, the machine shall be checked by a free swing of the pendulum. With the indicator at the maximum l
energy position, a free swing of the pendulum shall indicate l
zero energy.
(b) At the discretion of the LRC Project Leader, an aluminum calibration specimen shall be tested at room temperature in a manner otherwise identical to subsequent tests.
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Paoceouma No. LRC-TP-RO F-6 **os ao. 5 of 11 s
RC 372-3
,, m RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE DAM 10/27/R1 5.2 Instrumented System The data sto: age unit (s) of the instrumented impact system shall be calibrated at least annually against standards traceable to the National Bureau of Standards (NBS). The system as a whole shall be checked daily by comparison of an energy value recorded by the instrumented system with that indicated on the impact tester.
5.3 Temperature Measurement Thermometers used for temperature measurement shall be calibrated against NBS traseable standards within 3 months before the test.
5.4 Lateral Expansion Measurement The dial indicator used for the measurement of lateral expansion shall be calibrated at least annually against NBS traceable standards.
5.5 Percent Shear Fracture Determination If a dial caliper is used in the determination of percent shear fracture, it shall be calibrated at least annually against NBS traceable standards.
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PROCEDURE No. LRC-TP-80
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F-7 t) of 11
RC 377-3 11/75 RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE oug 10/27/81 i
- 6. Test Procedure 6.1 Test Redundancy The number of " identical" specimens to be tested at each temperature shall be as determined by the LRC Project Leader.
6.2 Specimen Heating and Cooling Nominal test temperatures shall be designated by the LRC Project Leader. Specimens shall remain immersed in the liquid medium for not less than.10 minutes.
6.3 Specimen Placement Seff-centering tongs shall be used for alignment of the specimen within the machine anvils. The specimen-gripping portion of the tongs shall be cooled or heated with the specimens.
6.4 Machine Preparation
. The pendulum shall be placed in its latched position. The energy indicator shall be set at its maximum energy position. The "Dynatup" instrumented system and the computer data ac,quisition system shal'1 be prepared to receive data.
l 6.5 Test Operation
! 6.5.1 The specimen sball be removed from the immersion bath with the tongs and centered between the machine's anvils. The pendulum shall then be released from its latched position.
l This entire operation shall be completed within 5 seconds.
The energy valve shall be read from the indicator prior to i . locking the pendulum for the next test.
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Pnocrount No. LRC-TP-80 PAGE No. 7 of 11 l
RC 372 3 H/75 RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE oars 10/27/81 6.5.2 If the specimen fails to break, the test shall be terminated for that specimen. The LRC Project Leader shall be immedi-ately notified. The fact shall be recorded indicating whether the break occurred through extreme ductility or lack of sufficient energy in the blow. The results of such tests shall not be used for material evaluation.
6.5.3 If the specimen jams in the machine, the test should be terminated for that specimen. The LRC Project Leader shall be immediately notified. The results of that test shall be disregarded and the machine shall be checked for damage or maladjustment and the requirements of Section 5.1.2 shall be followed.
6.5.4 After testing, specimens shall be carefully removed from the-tester or specimen catcher and maintained as speci=en pairs by placing in plastic bags, by taping together, or as other-vise convenient in accordance with requirements established by the LRC Project Leader.
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Paoctount so. LRC-TP-RO
'^ ** - 8 of 11 F-9
RC 372 3 it/m RESEARCH AND DEVELOPMENT DIVISION
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TECHNICAL PROCEDURE oarg 10/27/81 l
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- 7. Procedures for Obtaining Test Results 7.1 Impact Energy Charpy Impact Energy shall be read to the nearest integer on the tester's energy indicator. The maximum energy signal on the oscilloscope (E, max) shall be read to the nearest determinabla value.
7.2 Lateral Expansion ;
Lateral expansion shall be determined in accordance with the following procedure.
7.2.1 Measurements shall be taken on a gage consisting of a dial indicator rigidly mounted to a fixture. The fixture shall provide reference supports for locating broken specimen with respect to the dial indicator. The dial indicator shall be equipped with a flat anvil to provide proper con-tact with the specimen. This gage shall conform in principle to ASTM E23-81. Figures 17 and 18.
, 7.2.2 The broken specimen halves shall be r,etrieved and paired.
I The sides perpendicular to the notch shall be visually inspected to ensure that they are free from burrs (as would be formed during impact testing). If such burrs are present, they shall be removed (as by rubbing on emery cloth), making sure that the prcerusions to be measured are not disturbed. l The fracture surface shall be visually examined to ascertain l that the protrusions have not been damaged as by contacting !
the anvil or machine mounting surface. Damaged specimens'
. shall not be measured for lateral expansion except if l decided otherwise by the Project Leader.
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"^ 9 of 11 l
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RC 372 3 m 75 [ESEARCH AND DEVELOPMENT DIVISION TECHNIC'AL PROCEDURE oars 10/27/81 7.2.3 The halves shall be placed together so that the compression sides are facing one another. One half shall then be taken and pressed firmly against the reference supports with the protrusion against the gage anvil. The reading shall be noted and the measurement repeated with the other broken half, ensuring that the same side of the specimen is measured. The larger of the two readings shall be taken as the expansion of that side of the specimen. This procedure shall then be repeated to measure the protrusion on the opposite side. The larger values for each of the two sides shall be added together and this sum, expressed in mils, shall be recorded as the lateral expansion for that specimen.
7.3 Fracture Appearance Percentage of shear fracture shall be determined in accordance with the methods of 7.3.1 or 7.3.2, below. The LRC Project Leader shall select the method to be followed.
7.3.1 Measurement Method -- The length and width of the cleavage portion of the fracture surface shall be directly measured with a dial caliper and the percentage of shear fracture determined therefrom. " Length" and " width" are defined in ASTM E23-81. Figure 14. Determination of percentage values shall be by use of ASTM E23-81. Table 1 or Table 2.
7.3.2 Comparison Method -- The appearance of the fracture surface shall be visually compared with those in ASTM E23-81, Figure 15, and the values obtained therefrom shall be taken j as the percentage of shear fracture.
emocrouat NctLRC-TP-80 F-ll PAGE NO. 10 Of 11
RC 372-3 ti m RESEARCH AND DEVELOPMENT DIVISION TECHNICAL PROCEDURE oarg 1/26/82
- 8. Test Record 8.1 The teat record shall contain the following results or information in columnar form:
(a) Specimen identifier (b) Test temperature, F (c) Impact energy, dial, in ft-lbs (d) Impact energy, Dynatup. (Eamax) , in f t-lbs (e) Lateral expansion, in mils (f) Percentage of shear fracture (g) Oscilloscope readings (i) Load, in lbs/div (ii) Sweep speed, in msec /div 8.2 The test record shall also contain the following information:
(a) Project title and account number (b) Specimen size and justification for use if subsize specimens were used (c) Date of each test (d) Operator for each test ,
8.3 The designation of the procedure followed, including revision identification, shall be recorded on the test record and the test record shall be countersigned by the LRC Project Leader.
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- 8. REFERENCES
- 1. J. A. .Davidson, J. H. Phillips, and S. E. Yanichko, Virginia Electric &
Power Compa ny North Anna Unit 2 Reactor Vessel Radiation Program, WCAP-8772, Westinghouse, Pittsburgh, Pennsylvania, November 1976.
- 2. 00T 3.5 -- Two-Dimensional Discrete Ordinates Radiation Transport Code, (CCC-276), WANL-TME-1982, Oak Ridge National Laborato ry , Oak Ridge Tennessee, December 1969.
- 3. CASK -- 40-Group Coupl ed Neutron and Gamma-Ray Cross Section Data, RSIC-DLC-23, Radiation Shielding Infonnation Center, Oak Ridge Tennesse, March 1975.
4 H. S. Palme and H. W. Behnke, Methods of Compli ance With Fracture Toughness and Operational Requi rements of Appendix G to 10 CFR 50, BAW-10046A, Babcock & Wilcox, Lynchburg, Virginia, July 1977. -
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