ML20107B271
| ML20107B271 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/30/1982 |
| From: | Abraham M, Harris C, Lowe A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20107B242 | List: |
| References | |
| BAW-1679, BAW-1679-R01, BAW-1679-R1, NUDOCS 8502200322 | |
| Download: ML20107B271 (84) | |
Text
{{#Wiki_filter:. BAW-1679, Rw, 1 June 1982 ANALYSES OF CAPSULE CR3-B
< FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3
- Reactor Vessel Materials Surveillance Program -
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BAW-1679, Rev. 1 June 1982 ANALYSES OF CAPSULE CR3-B FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 - Reactor Vessel Materials Surveillance Program - by . A. L. Lowe, Jr., PE M. S. Abraham C. E. Harris , J. K. Schmotzer C. L. Whitmarsh l l i B&W Contract No. 595-7030-65 l BABC0CK & WILCOX ' Nuclear Power Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox
J Revision 1 (6/30/82) f 4
- 1. INTRODUCTION This report describes the results of the examination of the first capsule of the Florida Power Corporation's Crystal River Unit 3 reactor vessel surveil-4 lance program. The capsule was removed and examined after the first year of cperation. (Capsule removed during March 1978 outage.) l1 The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under cetual operating conditions. The' surveillance program for Crystal River Unit 3 was designed and furnished by Babcock & Wilcox as described in BAW-10100A.1 The program was designed in accordance with the requirements of Appendix H to 10 CFR Part 50 and ASTM specification E185-73 and was planned to monitor the offects of neutron irradiation on the reactor vessel material for the 40-year design life of the reactor pressure vessel. The future operating limitations sstabl'ished after the evaluation of the surveillance capsule are also in ac-cordance with the requirement of 10 CFR 50, Appendixes G and H. The recom-mended operating period w:s extended to eight effective full power years as a result of the first capsule evaluation.
g This revision addresses the affects of the change in fuel management - a lon-
. ger fuel cycle and higher power level, which result in a lower estimated EOL 1 fluence.-
4 i 1-1 Babcock & Wilcox
- 2. BACKGROUND The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water cooled re-actors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general ef-facts of fast neutron irradiation on the mechanical properties of such low-alloy ferritic steels as SA533, Grade B, used in the fabrication of the Crystal River River Unit 3 reactor vessel are well characterized and documented in the liters-ture. The low-alloy ferritic steels used in the beltline region of reactor ves-
- sels exhibit an increase in ultimate and yield strength properties with a corre-sponding decrease in ductility after irradiation. In reactor pressure vessel steels, the most serious mechanical property change is the increase in tempera-ture for the transition from brittle to ductile fracture accompanied by a re-duction in the upper shelf impact toughness.
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 cperational requirements are specified to provide adequate safety margins i during any condition of normal operation, including anticipated operational ~ occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. A1'though the requirements of Appendix G to 10 CFR 50 became effective on August 13, 1973, the requirements are ap-t f plicable to all boiling and pressurised water-cooled nuclear power reactors, including those under construction or in operation on the effective date. Appendix H to 10 CFR 50, " Reactor Vessel Materiale Surveillance Program Re-quirements," defines the material surveillance program required to monitor i changes in the fracture toughness properties of ferritic materials in the re-I cetor vessel beltline region of water-cooled reactors resulting from exposure 2-1 Babcock &Wilcox
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- to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the re-actor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life.
A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix C to the ASME Boiler and Pressure Vessel Code, Section III. This method utilizes fracture mechanics concepts and the reference nil-ductility temperature RTg which is defined as the greater of the drop weight nil-ductility transition temperature (per ASTM E-208) or the temperature that is 60F below that at which the material exhibits 50 ft-lb and 35 mils lateral expansion. The RT g of a given material is used to index that material to a reference stress intensity factor curve (K curve), which appears in IR Appendix G of ASME Section III. The K IR """" " "*# "" I"*"'
- static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR """*'
allowable stress intensity factors can be obtained for this material as a func-tion of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors. The RT ET and, in turn, the operating limits of a nuclear power plant, can be 1 adjusted to account for the effects of radiation on the properties of the re-actor vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a j surveillance program in which a surveillance capsule containing prepared speci-mens of the reactor vessel materials is periodically removed f rou the operating nuclear reactor and the specimens tested. The increase in the charpy V-notch 50-ft-lb temperature, or the increase in the 35 mils of lateral expansion ten-perature, whichever results in the larger t'emperature shift due to irradiation, is added to the original RT to adjust it for radiation embrittlement. This ET adjusted RT "" ET is used to index the material to the KIR """** * '" ' '" is used to. set operating limits for the nuclear power plant. These new limits take into account the effect's of irradiation on the reactor vessel materials. 2-2 8abcock & Wilcox
- 3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program comprises six surveillance capsules designed to moni-tor the effects of neutron and thermal environment on the materials of the re-cctor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned inside the reactor l vessel between the thermal shield and the vessel wall at the locations shown in
! Figure 3-1. The six capsules, placed two in each holder tube, are positioned l near the peak axial and azimuthal neutron flux. BAW-10100A includes a full de-ceription of capsule locations and design.1 Capsule CR3-B was removed during the first refueling shutdown of Crystal River Unit 3. This capsule contained Charpy V-notch impact and tensile specimens fcbricated of SA533, Grade B Class 1, weld metal, and weld metal compact f rac-ture specimens. The specimen contained in the capsule are described in Table 3-1, and the chemistry and heat treatment of the surveillance material in cap-Cule CR-3B are described in Table 3-2. All test specimens were machined from the 1/4-thickness location 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 plate; specimens were also oriented transverse to the principal rolling direction. Capsule CR3-B contained dosimeter wires, described as follows: Dosimeter wire Shieldina U-Al alloy Cd-Ag alloy Np-Al alloy Cd-Ag alloy Nichal Cd-Ag alloy 0.56 wt % Co-Al alloy Cd 0.56 wt % Co-Al alloy None Te None 3-1 Babcock & Wilcox
Thermal monitors of low-melting eutectic alloys were included in the capsule. The eutectic alloys and their melting points are as follows: Alloy Melting point, F 90% Pb, 5% Ag, 5% Sn 558 97.5% Pb, 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Lead 621 Cadmium 610 Table 3-1. Specimens in Surveillance Capsule CR3-B Number of test specimens Material description Tension CVN impact 1/2 T compact tension
- Weld metal 2 12 8 Weld-HAZ Heat NN, transverse --
12 -- Base metal Heat NN, transverse 2 12 :- Total per capsule 4 36 8 (a) Compact tension specimens pre-cracked per ASTM E399-72. I i i l I \ 1 3-2 Babcock & Wilcox
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4 Ravision 1 (6/30/82) Table 3-2. Chemistry and Heat Treatment of Surveillance Materials Chemical Analysis i Heat Weld metal Weld metal Element C-4344-1 WF-209-1(a) WF-209-1(b) C 0.23 0.08 0.11 Mn 1.30 1.65 1.57 P 0.008 0.021 0.018 1 S 0.016 0.013 0.009 Si 0.22 1.00 0.54 Ni 0.54 0.10 0.61 Mo 0.55 0.45 0.43 Cu 0.20 0.39 0.36 Heat Treatment Heat Temp. Time . No. F h Cooling C-4344-1 1650-1700 9 Water quench 1180 4.5 Air cooled 1100-1150 60.0 Furnace cooled WF-209-1 1100-1150 48.0 Furnace cooled (* The weld metal used to fabricate the Charpy and tensile specimens has been identified as
" atypical" weld metal, as described in BAW-10144, February 1980.s The compact frac-ture specimens are fabricated from the desig-nated weld metal. g
( )The weld metal used to fabricate the compact fracture specimens has been identified as from the designated weld metal, as described in BAW-10144, February 1980.s 3-3 Babcock & Wilcox n- c--, , , . , , , ,e-. --
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- 4. PREIRRADIATION TESTS l
i Unirradiated material was evaluated for two purposes: (1) to establish a base-line of data to which irradiated properties data could be referenced, and (2) to determine those materials properties to the extent practical from available material, as required for compliance with Appendixes G and H to 10 CFR 50. 4.1. Tensile Tests Tensile specimens were fabricated from the reactor vessel she?.1 course plate l cnd weld metal. The subsize specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter. They were tested on a 55,000-lb-load capacity universal test machine at a crosshead speed of 0.050 inch per minute. A 4-pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accordance with the applicable requirements of ASTM A370-72. For esch material type and/ l cr condition, six specimens in groups of three were tested at both room tem- ! perature and 580F. The tension-compression load cell used had a certified cecuracy of better than ! 0.5% of full scale (25,000 lb). All test data for the prairradiation tensile specimens are given in Appendix B. 4.2. Impact Tests l l Charpy V-notch impact tests were conducted in accordance with the requirements l cf ASTM Standard Methods A370-72 and E23-72 on an impact tester certified to meet Watertown standards. Test specimena were of the Charpy V-notch type, which were nominally 0.394 inch square and 2.165 inches long. l Prior to testing, specimens were tesqperature-controlled in liquid immersion baths, capable of covering the temperature range from -85 to +550F. Specimens were removed from the baths and positioned in the test frame anvil with tongs specifically designed for the purpose. The pendulun (hammer) was, released manually, allowing the specimens to be broken within 5 seconds from their re-j moval from the temperature baths. i 4-1 Babcock & Wilcox
Impact test data for the unirradiated baseline reference materials are present-ed in Appendix C. Tables C-1 through C-5 contain the basis data which are plotted in Figures C-1 through C-5. 4.3. Compact Fracture Tests The compact fracture specimens fabricated from the weld metal, which were a part of the capsule specimen inventory, were not tested because of the lack of a recognized testing procedure. These specimens will be kept in bonded storage until an acceptable test procedure is developed. The,results of the testing of these specimens will be the subject of a separate report. 4-2 Babcock & Wilcox
i
- 5. POST-IRRADIATION TESTS ,
5.1. Thermal Monitors Surveillance capsule CR3-B contained three temperature monitor holder tubes, ecch containing five fusible alloys with different melting points ranging from 558 to 621F. All the thermal monitors at 558 and 580F had melted, while those - at 588, 610, and 621F remained in their original configuration as initially placed in the capsule. From these data it was concluded that the irradiated . . . specimens had been exposed to a maximum temperature in the range of 580 to _ less than 588F during the reactor vessel operating period. There appeared to be no significant temperature gradient along the capsule length. . jg - M 5.2. Tensile Test Results The results of the post-irradiation tensile tests are presented in Table 5-1. Tests were performed on specimens at both room temperature and 580F using the same test procedures and techniques used to test the unirradiated specimens (section 4.1). In general, the ultimate strength and yield strength of the material increased slightly with a 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 renge of changes to be expected for the radiation environment to which the cpecimens were exposed. m The results of the preirradiation tensile tests are presented in Appendix B. ' 5.3. Charpy V-Notch Impact Test Results . .k:
- 3. on ,-
The test results from the irradiated Charpy V-notch specimens of the reactor p,
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v:ssel beltline material and the correlation monitor material are presented in hJ ,. A Tcbles 5-2 through 5-4 and Figures 5-1 through 5-3. The test procedures and y;.'y techniques were the same as those used to test the unirradiated specimens (sec- { tion 4.2). The data show that the material exhibited a sensitivity to irradia- W .. M,
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tion within the values predicted from its chemical composition and the fluence jj.[ to which it was exposed. h)f,,- lS%.. w . 5-1 Babcock & Wilcox
The results of the preirradiation Charpy V-notch impact test are given in Appendix C. Table 5-1. Tensile Properties of Capsule CR3-B Base Metal and Weld Metal Irradiated to 1.0E18 nyt (>l MeV) Elongation, % Strength, psi g , , , Specimen Test temp, No. F Yield Ult. Unif Total % Base Metal, Transverse NN-620 RT 78,125 100.000 E.S. 27 61 NN-615 580 69,375 95.625 10 21 49 Weld Metal e PP-007 RT 83,125 98,750 12 27 62 PP-017 580 69,375 90,625 12 25 53 Table 5-2. Charpy Impact Data From Capsule CR3-B Base Metal Irradiated to 1.0E18 (> 1 MeV) Absorbed ' Lateral Shear Specimen Test temp, energy, expansion, fracture. No. F ft-lb 10-3 in. % Base Metal, Transverse NN-689 20 24 16 10 NN-605 50 34 30 10 NN-658 70 33 36 30 NN-685 85 37 32 30 NN-641 100 43 39 30 NN-646 120 56 44 30 NN-650 140 59 54 70 NN-613 170 66 58 100 NN-608 190 93 73 100 NN-616 210 89 76 100 NN-630 260 89 73 100 NN-632 30 0 88 74 100 5-2 Babcock & Wilcox
Table 5-3. Charpy Impact Data From Capsule CR3-B Heat-Affected Zone Metal Irradiated to 8.4E17 (>l MeV) Absorbed Lateral Shear Specimen Test temp, energy, expansion, fracture. No. F ft-lb 10-3 in. % Heat-Af f ected Zone. Transverse NN-392 20 23 14 20 NN-372 50 28 21 30 NN-365 70 39 33 28 NN-366 100 60 50 88 l NN-377 120 51 43 100 NN-352 140 44 36 30 NN-381 170 60 56 100 NN-375 210 74 63 100 NN-389 260 66 62 100 NN-374 285 79 67 100 NN-383 ' 320 84 67
- 100 NN-380 360 74 65 100 Table 5-4. Charpy Impact Data From Capsule CR3-B Weld Metal Irradiated to 1.17E18 n/cm 2
(>l MeV) Absorbed Lateral Shear Specimen Test temp, energy, expansion, fracture, No. F ft-lb 10-3 in. % PP-092 0 13 11 10 PP-076 35 19 15 10 PP-069 55 27 24 20 PP-059 70 33 30 18 PP-082 95 35 31 30 PP-070 120 48 44 60 PP-091 130 40 38 45 PP-058 140 61 53 100 PP-089 190 70 66 100 PP-077 250 74 68 100 PP-088 285 66 66 100 PP-065 320 70 71 100 mmmi i
I Figure 5-1. Charpy Impact Data for Irradiated Base Metal, Transverse Direction im , , , , , _ . . _ _ , , a 75 - - a 1 y50 . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _.. E y e w 25 - - e g a e a e a I a e a e
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SUMMARY
180 T NA . NDT 73F TCV (35 HLE) 160 .T CV -L ) F . 37F TCV (30 FT-LB)
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.so -40 o to so in les 20o 240 zoo 320 uo non Tset Tevenatune, F 5-4 Babcock & Wilcox huuh u s i m'is
Figure 5-2. Charpy Impact Data for Irradiated Base tietal i Heat-Affected Zone im i i i . , _ _ (
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. 140 - Cy -USE (AVG) 79 ft-lb .
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-80 -40 o 40 so 120 160 200 24a 2so 320 sw am Test Teseenatune, F 5-5 Babcock s,Wilcox
Figure 5-3. Charpy Impact Data for Irradiated Weld Metal Im . . , , , . . .. .- -
" 75 -
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SUMMARY
180 T - NDT 108F TCV (35 MLE) 160
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- 6. NEUIRON DOSIMETRY s
6.1. Introduction = n A significant aspect of the surveillance program is to provide a correlation j between the neutron fluence above 1 MeV and the radiation-induced property
=4 changes noted in the surveillance speci:nen. To permit such a correlation, _
activation detectors with reaction thresholds in the energy range of interest were placed in each surveillance capsule. The properties of interest for the detectors are given in Table 6-1.
}
Because of a long half life of 30 years and an effective energy range of >0.5 i MeV, only the measurements of 187 Cs production from fission reactions in 237 Np i and 2:s U are directly applicable to analytical determinations of fast neutron 4 (E>l MeV) fluence during cycle 1. The other dosimeter reactions are useful as N corroborating data for shorter time intervals and/or higher energy fluxes. $ Short-lived isotope activities are representative of reactor conditions over the latter portion of the irradiation period (full cycle) only; whereas re- 7 actions with a high threshold energy do not record a significant part of the =J total fast flux. jR _ = The energy-dependent neutron flux is not directly available from activation detectors because the dosimeters record only the integrated effect of the neu- ; tron flux on the target material as a functica of both irradiation time and neutron energy. To obtain an accurate estimate of the average neutron flux , incident upon the detector, the following parameters must be known: the oper- E cting history of the reactor, the energy response of the given detector, and ] the neutron spectrum at the detector location. Of these parameters, the defi- 3 nition of the neutron spectrum is the most difficult to obtain. Essentially, j two means are available to obtain the spectrum: iterative unfolding of experi-mental foil data and analytical methods. Due to a lack of suf ficient threshold 35 1 foil detectors satisfying both the threshold energy and half-life requirements tecessary for a surveillance pronram, iterative unfoldin'g could not be used. 1 This leaves the specification of the neutron spectrtan to the analytical method.
'~I d Babcock & Wilcox
6.2. Analytical Approach 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 35.2 The CR3 reactor was modelled from the core out to the primary concrete shield in R-theta geometry (based on a plan view along the core midplane and one-eighth core symmetry in the azimuthal dimension). Also included was an explicit model of a holder tube and capsule assembly at the proper location. The CR3-B capsule was positioned 201.84 cm from the core center and 26.9* off-axis. The reactor model contained the following regions: core, liner, bypass coolant, core barrel, inlet coolant, thermal shield, pres-sure vessel, cavity, and primary concrete shield. Input parameters to the code included a pin by pin time-averaged power distribution, CASK 23E 22-group microscopic neutron cross sections,3 Se order of angular quadrature, and Ps expansion of the scattering cross section matrix. Because of computer storage limitations, it was necessary to use two geometric models to cover the distance from core to primary shield. A boundary source output from the initial model A (core through inlet coolant region) was used to " bootstrap" a model B, which included the capsule. Any " shadowing" effect of the pressure vessel .by the capsule was determined by running a third model, which represented the model B region of the reactor but without the capsule assembly. Flux output from the DOT 35 calculations required only an axial distribution adj ustment . Thus, fluxes were multiplied by an axial shape factor to account for the capsule elevation. CR3-B was in an upper location which extends from the core midplane upward about 60 cm. This factor was 1.19 for the capsule po-sition and 1.22 for the maximum location on the pressure vessel surface. The calculation described above provides the neutron flux as a function of energy at the detector position. These calculated data are used in the follow-ing equations to obtain the calculated activities used for comparison with the experimental values. The basic equation" for the activity D (in pCi/gm) is given as follows: M -A t Dg = Ai 3 x 10" i E [Fj n(E)HE) j=1 (1-e g d)e-AI(T-t$) (6-1) 6-2 Babcock & Wilcox _ ________ ______________-___:___ _ _ _ _ _ - _ = _ _ _ _ _ _ - _ _ .
Revision 1 (6/30/82) where C = normalizing constant, ratio of measured to calculated flux. N = Avogadro's ntsaber, A g = atomic weight of target material i, f g = either weight fraction of target isotope in nth material 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 jth time interval, t), A = decay cc,nstant of ith material, t) = interval of power history. T = sum of total irradiation time (i.e. , residual time in reactor) and wait time between reactor shutdown and counting, t) = cumulative time from reactor startup to end of jth time period, i.e., T = t j u g k=1 The normalizing constant C can be obtained by equating the right side of equa-tion 6-1 to the measured activity. With C specified, the neutron fluence great-er than 1 MeV can be calculated from 15 MeV M
$(E > 1.0 MeV) = C [, (6-2)
E-1 [ F)t) j=1 where M is the number of irradiation time intervals; the other values are defined above.
- 6. 3. Results Calculated activities are compared to dosimeter measurements in Table 6-2.
The fission wire data indicate about a 20% overprediction of fast flux (E > 1 MeV) by the analytical model described herein; non-fission wires showed a 30% overprediction. Several f actors which could account for this variance - capsule location, reaction cross section spectrum weighting, and measurement techniques - are presently under investigation. Additional dosimetry measure-ments (not reported here") corroborated the overprediction bias of the analyt- l1 ical-to-measured flux comparison. A safety factor of 1.1 was included in the fluence calculations to provide conservatism in the activity measurements (and therefore capsule and vessel fluence). 6-3 Babcock & Wilcox
Revisicn 1 (6/30/82) Based on a normalization constant of 0.81, an average fast flux for cycle 1 was calculated at the capsule location and the inside surface of the pressure vessel wall. The data (Table 6-4) were converted to fast fluence values of 1.05 x 10" n/cm2 at the capsule center and 2.83 x 10 " n/cm2 at the pressure vessel wall for 268.8 days of cycle 1 at the full power rating of 2452 MWt. l1 Values calculated for the pressure vessel wall (inside surface) refer to the maximum flux, which may be located at a different asinuthal and axial position than the surveillance capsule. In this analysis, the maximum fluence at the pressure vessel occurred at an asinuthal position 7' from a major axis (cap-sule located at 26.9*) and about 190 cm above the lower active fuel line. This is a function of power distribution in the core. The effect of extend-ing the flux range dow to 0.1 MeV was to approximately double the fluence at the capsule and the pressure vessel. Since the same normalization factor of 0.81 is used, additional uncertainty is introduced in this result because none of the dosimeter reactions are effective over this entire energy range. Based on the surveillance sample analysis for cycle 1, fuel management calcu-lations for reload cores through cycle 5, and estimated future fuel cycles, pressure vessel t'luence was predicted up to 32 EFPY (effective full-power years) of operation. These data, which are listed in Table 6-5, are based on the proportionality of vessel fast flux to calculated core leakage flux I for cycles IB through 5. Fuel cycle 5 was assumed to represent an equilibrium , cycle for post-cycle operation. This prediction procedure is described in reference 4. 6.4. Summary of Results
- 1. Calculated activities for fuel cycle 1 exceeded measured activities by 20 to 30%. For the flux range of greater than 1 MeV, a normalising con-stant of 0.81 was selected for application to analytical fast flux pre-dictions near the pressure vessel wall. *
- 2. Average fast fluence at the capsule location was 1.05 x 10 " n/ cat (E > 1 MeV) after 268.8 days of the first fuel cycle. The maximum value at the pressure vessel wall was calculated to be 2.8 x 10" n/cm8 .
- 3. Extension of the flux range to E > 0.1 MeV tended to approximately double the calculated fluence values.
4 Based on specimen locations within the capsule, the specific fluence was determined for each group of materials (Table 6-6). I 6-4 Babcock & Wilcos
I Table 6-1. Surveillance Capsule Detectors Detector reaction Energy range, MeV Isotope half-life s"Fe(n.p)5"Mn >2.5 312 days s sNi(n.p) seco >2.3 71.3 days 2 sU(n.f)18'Cs >1.1 30.2 years 2:7 Np(n,f)187Cs
>0.5 30.2 years 888 U(n.f)18'Ru >1.1 365 days 237 Np(n.f)188Ru
>0.5 365 days Table 6-2. Dosimeter Activations C=A/B Measured activity, (*} Calculated )
activity,(b normalization Reaction pCi/g uCi/g constant s"Fe (n.p) s "Mn 454 668 0.68 ssNi(n.p)seco 1056 1540 0.69
'88U(n.f)18'Cs 1.21 1.50 0.81 2s7 Np (n . f) 18 'Cs 6.89 9.12 0.76 2:aU(n.f)1Ru 10.3 15.1 0.68 as7 Np(n,f)lRu 49.3 62.8 0.78
("} Average of four dosimeter wires from Table D-2.
) Average of four dosimeter locations in calculational model.
Table 6-3. Normalized Flux Spectra, Flux per MeV for Range (E > MeV) T/4 in assy Energy range. At center pressure MeV of Capsule vessel fission 12.2-15.0 0.0003 0.0005 0.0002 10.0-12.2 0.0016 0.0024 0.0009 8.18-10.0 0.0057 0.0079 0.0035 6.36-8.18 0.0158 0.0196 0.0126 4.96-6.36 0.0391 0.0441 0.0369 4.06-4.96 0.0594 0.0610 0.0775 3.01-4.06 0.0857 0.0786 0.145
, 2.46-3.01 0.185 0.169 0.232 2.35-2.46 0.307 0.279 0.279 1.83-2.35 0.312 0.296 0.331 1.11-1.83 0.529 0.517 0.444 1.0 ~1.11 0.729 0.856 0.412 I")Ustd to evaluate reaction cross sections (1/E tail added to lower energy groups).
Table 6-4. Neutron Fluence E > 1 MeV E > 0.1 HeV Fast fluence Fluence Fast flux, for cycle 1A Flux, for cycle 1A n/en8 -s (268.8 EFFD) n/ca -s (268.8 EFPD) Capsule center 4.51(+10) 1.05(+18) 1.02(+11) 2.37(+18) Pressure ves- 1.22(+10) 2.83 (+17) 2.51(+10) 5.83(+17) sel wall (max) 6-6 Babcock &WHcom
Revision 1 (6/30/82) 4 Table 6-5. Predicted Fast Fluence in Pressure Vessel at Maximum Location (a) Power, 1 g) Cycle MWt Interval Total n/cm2 -s Interval Total 1A 2452 0.74 0.74 1.2(+10) 2.8(+17) 2.8(+17) IB 2452 0.47 1.21 1.5(+10) 2.2(+17) 5.0(+17) 2 2452 0.46 1.67 1.7(+10) 2.5(+17) 7.5(+17) g 3 2452 0.88 2.55 1.44(+10) 4.0(+17) 1.2(+18) 4 2544 0.94 3.49 1.29(+10) 3.8(+17) 1.6(+18) 5 2544 1.09 4.58 1.24(+10) 4.3(+17) 2.0(+18)
>5 2544 3.42 8.0 1.24(+10) 1.4(+18) 3.4(+18)
>5 2544 24.0 32.0 1.24(+10) 9.4(+18) 1.3(+19)
(*)1nside surface of the base metal at an azimuthal location about 7 to 11* from a major core axis and an elevation near the core midplane.
) Maximum fluence can be trarislated to T/4, 3T/4, and outer surf ace lo-cations with the factors 1/1.8,1/7.7, and 1/21, respectively.
Table 6-6. Calculated Neutron Fluence for Material Specimens (n/cm2 ) Tensile Specimens Weld metal 1.0 x 10 18 Base metal 1.0 x 101 ' Charpy Specimens Weld metal 1.17 x 101 ' Base metal 1.0 x 101 ' Ileat-affected zone 8.4 x 10 17 Compact Fracture Specimens 1 Weld metal 1.0 x 10 ' 1 6-7 Babcock a, Wilcox
]
wg{" ~8 = T
%8 V
T T T N V V V N N N 7 1 ' 7 8 7 0 8 1 1 1 I 0 0 x 0 1 6 1 1 hg X X X 1 u 4 9 4 o r 3 1 4 h ' 6 T s N O F n I C i o T A A F t C l a O u l c L S o L s u Mq T F 4
/
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* - - - * ~ *
- 0 2 4 9 0 6 2 8 4 0 6 *
. 1 0 4 3 3 2 2 2 1 C" _ A 5 G ~ _ . =5 m c$E* .
- Fgh gW
!llll
l d 1
- 7. DISCUSSION OF CAPSULE RESULTS .
7.1. Preirradiation Property Data A review of the unirradiated properties of the reactor vessel core belt region i indicated no significant deviation from expected properties except in the case i of the upper shelf properties of the veld metal. Based on the predicted end-of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of this weld, it is predicted that the end-of-service Charpy upper shelf energy (USE) will be below 50 ft-lb. This weld was selected for . inclusion in the surveillance program in accordance with the criteria in ef-fect at the time the program was designed for Crystal River Unit 3. The ap-plicable selection criterion was based on the unirradiated properties only. 7.2. Irradiated Property Data 7.2.1. Tensile Properties Table 7-1 compares irradiated and unirradiated tensile properties. At both room temperature and elevated temperature, the ultimate and yield strength changes in the base metal as a result of irradiation and the corresponding changes in ductility are negligible. There appears to be some strengthening, as indicated by increases in ultimate and yield strength and similar decreases - in ductility properties. All changes observed in the base metal are such as . . to be considered within acceptable limits. The changes at both room tempera- g -Q
- a. .~
ture and 580F in the properties of the base metal are greater than those ob- h;)..if served for the weld metal, indicating a gr' ester sensitivity of the base metal id / [d au .. to irradiation damage. In either case, the changes in tensile properties are i.p d, insignificant relative to the analysis of the reactor vessel materials at this .v} o%;
?;;,rjs, period in service life.
?.,]1-{
;3:p 7.2.2. Impact Properties 4 rj
% .,a The behavior of the Charpy V-notch impact data is more significant to the cal-Y.
y
++*y culation of the reactor system's operating limitations. Table 7-2 compares [1 .
b[.4 7-1 Babcock & Wilcox
l . the observed changes in irradiated Charpy impact properties with the predicted changes as shown in Figures 7-1 through 7-3. The 50-f t-lb transition temperature shif t for the base metal was in good agree-ment with the shift that would be predicted according to Regulatory Guide 1.99. The less-than-ideal comparison may be attributed to the spread in the data of the unirradiated material combined with a minimum of data points to establish the irradiated curve. Under these conditions, the comparison indicates that the estimating curves in RG 1.99 for medium-copper materials and at low fluence
- 1evels are reasonably accurate for predicting the 50-f t-lb transition tempera-ture shifts.
The 30-ft-lb transition temperature shift for the base metal is not in as good agreement with the value predicted according to Regulatory Guide 1.99, although it would be expected that these values would exhibit better comparison 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. The increase in the 35-mil lateral expansion transition temperature is compared with the shift in RT curve data in a manner similar to the comparison made NDT for the 50-ft-1b transition temperature shif t. These data show a behavior sin-ilar to that observed from the comparison of the observed and predicted 50-ft-lb transition data. All the transition temperature measurements for the weld metal are in poor agreement with the predicted shift. This can be attributed to the chemistry of the weld metal as compared to the nominal chemistry of normal weld metal for which the prediction curves were developed. The abnormal silicon content and low nickel content combina to make the weld metal less sensitive than the base metal. This being the case, it would not be expected that the current prediction techniques will apply to the weld metal. The data for the decrease in Charpy USE with irradiation showed a poor agree-ment with predicted values for both the base metal and the weld metal. How-ever, the poor comparison of the measured data with the predicted value is not unexpected in view of the lack of data for medium- to high-copper-content ma-terials at low to medium fluence values that were used to develop the estimat-ing curves. 7-2 Babcock & Wilcox
Revision 1 (6/30/82) l i Results from other capsules indicate that the RT, estimating curves have greater inaccuracies at the very low neutron fluence levels (s1 x 10" n/cm2 ), j This inaccuracy is attributed to the limited data at the low fluence values and of the fact that the majority of the data used to define the curves in RG 1.99 are based on the shif t at 30 f t-lb as compared to the current require-ment of 50-ft-lb. For mest materials the shifts measured at 50 f t-lb/35 MLE cre expected to be higher than those measured at 30 ft-lb. The significance of the shif ts at 50 f t-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. Materials with this characteristic may have to be evalu-cted at transition energy levels lower than 50 ft-lb. he design curves for predicting the shif t at 50 f t-lb/35 MLE will probably l be modified as data become available; until that time, the design curves for 1 predicting the RTg shift as given in Regulatory Guide 1.99 are considered Edequate for predicting the RT g shift of those materials for which data are not available and will continue to be used to establish the pressure-tempera-t kure operational limitations for the irradiated portions of the reactor vessel. The lack of good agreement of the change in Charpy USE is further support of the inaccuracy of the prediction curves at the lower fluence levels. Although the prediction curves are conservative in that they predict a larger drop in upper shelf than is observed for a given fluence and copper content, the con-ssrvatism can unduly restrict the operational limitations. These data support the contention that the USE drop curves will have to be modified as more re-liable data become available; until that time the design curves used to pre-dict the decrease in USE are conservative. In evaluating the Charpy data from the CR3-B capsule it is important to note that the weld metal from which the Charpy specimens were fabricated has been l identified as " atypical" as described in BAW-10144A s. In accordance with the NRC review of BAW-10144A the behavior of the weld metal in Crystal River 3 [ RVMSP will be monitored to ensure the correct predictive methodology for those 1 ( weldsents that could have been fabricated with " atypical" weld wire. The data from the CR3-B capsule indicate that the atypical weld metal is significantly less sensitive to neutron radiation damage than normal weld metal of the same
-type.
7-3 Babcock &Wilcox
Ravision 1 (6/30/82) Table 7-1. Comparison of Tensile Test Results Elevated Room temp i:est temp test (580F) Unirr Irrad Unirr Irrad l ! Base Metal - 4344-1 Transverse l l 1 Fluence, 10 ' n/cm2 (> 1 MeV) 1 0 1.0 0 1.0 l l Ult. tensile strength, kai 92.2 100.1 90.6 95.6 i i l 0.2% yield strength, kai 69.3 78.1 63.9 69.4 i Elongation, % 24 27 23 21 RA, % 62 61 56 49 Weld Metal - WF-209-1(*) Fluence, 101s n/cm 2 (> 1 MeV) 0 1.0 0 1.0 l Ult. tensile strength, kai 93.8 98.7 88.5 90.6 0.2% yield strength, kai 77.0 83.1 67.5 69.3 Elongation, % 29 27 20 25 i RA, % 62 62 48 53 t (a) Atypical weld metal. i i l
- i l
I i l i i 7-4 Babcock &Wilcox l
.p---- - - - - - - a c. ,. . , ,.,,-,.n,,n._,. a., an,-,,._,- _ . .
Rsvision 1 (6/30/82) Table 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Properties Material Observed Predicted (*} Increase in 30-ft-lb trans temp, F Base material (C-4344-1) Transverse 21 57 Heat-affected zone (C-4344-1) 70 57 Weld metal (WF-209-1)(C} 28 105 Increase in 50-ft-1b trans temp, F Base material (C-4344-1) Transverse 51' 57 Heat-affected zone (C-4344-1) 64 57 Weld metal (WF-209-1)( } 13 105 Increase in 35-MLE trans temp, F Base material (C-4344-1) Transverse 35 57(b) Heat-affected zone (C-4344-1) 74 57(b) Weld metal (WF-209-1) 34 105(D} Decrease in Charpy USE, ft-lb Base material (C-4344-1) Transverse 5 16 Heat-affected zone (C-4344-1) 5 14 Weld metal (WF-209-1) (C) 9 24 (a)These values predicted per Regulatory Guide 1.99, Revision 1. (b) Based on the assumption that MLE as well as 50 ft-lb transi-tion temperature is used to control the shift in RTET * (c) Atypical weld metal. I i 7-5 Babcock &Wilcox
Figure 7-1. Irradiated Vs Unirradiated Charpy Impact Properties of Base Metal 1m , , , , , , , , n 15 - J Unitradiate
.E.
y 50 _-_____________ _ 1.0 x 10 1 ' nyt W - w 25 - 0
.080 , , , , , , , _
i i.060 - Unirradiated y 1.0 x 10 1 ' nyt AT=35F - w .040 3
=
5 i" .020 -
*li a I I $ t f I I I I 1 200 , , , , , , , , , , ,
180 160
. 140 -
f C I-$120 3 AUSE = 5 ft-lb e 100 - y a g Unitradiated 9 - w 80 - r, AT=51F 4 I I - 1.0 x 10 18 nyt . 60 40 - AT=21F
.ignyenig SA533 Gr 31 7-_-------___-__.- ,
Foutact See above litat llussen C-4344-1 a i a t t i e a e e a 80 -40 0 40 80 120 160 200 240 280 320 3% 400 TestTesernarums,F 7-6 Wd & cox
- - - . ~ , - - - , - . , . , . , , _ , , , , , . _ _ , _
Figure 7 -2. Irradiated Vs Unirradiated Charpy Impact Properties of Heat-Affected Zone Material l 100 , , , , , , , , , l
, ** 75 - -
j - Unirradiated 1 5 y so .__ _ _ _ _ ____ _ _ _ _ _ _ -. _ __ _ _ _ _ _ _ _ _ - __.
! 8.4 x 10 1 ' nyt ll5
& 25 - -
1 0
.080 i . . . . . i i s i -
$ .060 - Unirradiated ~
; a AT=74r c
, , r I 8. 4 x 10 17 nyt O .040
[_ _ _ _ '_ -. E
=
i .020
,, i e i e e i i e i . .
E a 5 s s B u u s a e i 1 180 - - 160 - - 140 - - T' E 120 - - I i 8 AUSE = 5 ft-lb
# 100 - -
l 0 y 1r w 80 - - y Unirradiate . T I 8.4 x 1017 nyt 60 - AM4r _
-______- c ]- - . - - - - - - - - - - - - - - - - - - - - - - - - - - .
40 - AT=70Fd y - hungg SA511-B/MA7 Onstwntion HAZ-Tranav. 20 - See above
~
Funner br ha C-4344-1 9 f f I I a a e e e a Eso -40 o 40 so 120 1so 200 240 2so 320 360 om i Test Teetnaruns, F 7-7 Babcock 8.Wilcox
Figure 7-3. Irradiated Vs Unirradiated Charpy Impact l Properties of Weld Metal I l 100 , , , ,
** 75 - -
Unirradiated I E y 50 ._ _ _ _ _ _ _ _ _ _. _ __ ________________.____
- 1 g 1.17 x 101 ' ave 1 5 25 -
f - I O
- 5 5 5 5 5 3 __
B ' i j .%0 - Unitradiated we h AT=34F 1.17 x 10 1 ' nyt w .040 - 5 -___ ______ _ l' 5
=- ,
i .020 e-b ( l I r f f f i I I e s e a a u a a e y a s 180 150
. 140 - -
t O l , E 120 p 1 3 ' 3 100 - - l
, AUSE = 9 ft-lb '
8 \ l w l w 80 - u - 1 AT-13F----@ 4.-_ 3
- 60 -
Unitradiated. - 40 . AT=28F 4--9 1,17 x 10 1 ' nyt . l gy,,, g Weld Metal 9tasistarios NA ' 20 - y,, , u,, 7,,,,,,, 1 mar w w-2oo-i l t I f f I e a a e e 0
.so -40 o 40 so 120 1eo 200 2eo m 520 3so 400 Test Taw anarums, F 7-8 Babcock siWilcox
I. i i i 4 .: N i i 8. DETERMINATION OF RCPB PRESSURE-TEMPERATURE LIMITS i . The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Crystal River 3 are established in accordance with the requirements of 10 CFR 50, Appendix C. The methods and criteria employed to establish operating pressure and temperature limits are described in topical report BAW-10046. 5 1bf-The objective of these limits is to prevent nonductile failure during any nor-
- mal operating condition, including anticipated operation occurrences and system hydrostatic tests. The loading conditions of interest include the following
- 1. Normal operations, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests.
- 3. Reactor core operation.
The major components of the RCPB have been analyzed in accordance with 10 CFR 50, Appendix G. The closure head region, the reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of the reac-tor vessel, and consequently of the RCPB, that regulate the pressure-tempera-ture limits. Since the closure head region is significantly stressed at rela-
- tively low temperatures (due to mechanical loads resulting from bolt preload),
this region largely controls the pressure-temperature limits of the first several service periods. The reactor vessel outlet nozzle also affects the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle which can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure, the RT, of the beltline region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature limits of the RCPB. For the service period for which the limit curves are established, the maximum allowable pressure as c function of fluid temperature is obtained through a point-by-point comparison cf 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
- three' calculated pressures.
8-1 Babcock & Wilcox
l l
! The limit curves for Crystal River 3 are based on the predicted values of the adjusted reference temperatures of all the beltline regien materials at the )
end of the eighth full-power year. The eighth full-power year was selected because it is estimated that the second surveillance capsule will be withdrawn at the end of the refueling cycle when the estimated fluence corresponds to approximately the ninth full-power year. The time difference between the with-drawal of the first and second surveillance capsule provides adequate time for re-establishing the operating pressure and temperature limits for the period of operation between the second and third surveillance capsule withdrawals. The unirradiated impact properties were determined for the surveillance belt-line region materials in accordance with 10 CFR 50, Appendixes G and H. For the other beltline region and RCPB materials for which the measured properties are not available, the unirradiated impact properties and residual elements, 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 ART and the unirradiated RT The predicted ET ET. ART is calculated using the repective neutron fluence and copper and phos-ET phorus contents. Figure 8-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel beltline region wall and at the center of the surveillance capsules at each of two locations as a function of exposure time. The supporting information for Figure 8-1 is described in BAW-10100.1 The neutron fluence values of Figure 8-1 are the predicted flu-ences, which have been demonstrated (section 6) to be conservative. The de-sign curves of Regulatory Guide 1.99* were used to predict the radiation-in-duced ARTg values as a function of the material's copper and phosphorus content and neutron fluence. The neutron fluences and adjusted RT values f the beltline region materials ET at the end of the eighth full-power year are listed in Table 8-1. The neutron fluences and adjusted RT ET values are given f r the 1/4T and 3/4T vessel wall , locations (T = wall thickness). The assumed RTET f the ci sure head region and the outlet nozzle steel forgings is 60F, in accordance with BAW-10046P.s l i 1 I
- Revision 1. January 1976.
8-2 Babcock &Wilcox i
1 Figure 8-2 shows the reactor vessel's pressure-temperature limit curve for normal heatup. This figure also shows the core criticality limits as required by 10 CFR 50, Appendix G. Figure 8-3 and 8-4 show the vessel's pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively. All pressure-temperature limit curves I are applicable up the the ninth effective full-power year. Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits of the pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to go critical until the pressure-temperature combinations are to the right of the criticality limit curve. To establish the pressure-~emperature c limits for protection against nonductile failure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel controlling the limit curves. This is necessary because the reactor vessel is the most limit-ing component of the RCPB. l S-3 Babcock & Wilcox
Table 8-1. Data for Preparation of Pressure-Temperature Limit Curves for Crystal River Unit 3, Applicable Through 8 EFPY Weldment location peutros fluence et **I
- I Untre end of 8 EFFT AST,,,, et end of Adjusted RTuot **
_ Nates tel identif'a .,,gggg,, g,,, ,gg,g,,, g ,, , y,gg, ,g RT NDT. eat H>I . n/cm 0 W. e of 8 Em. F Heat No. Type _ resten lecet tee to weld %. cm es s e .
- I/479 F Co. I P. 2 At l/4 T At 3/4 T At 8/4 T At 3/4 T At 1/4 T At 1/4 T 823Vl90 SA500. C12 Leuer measle belt - - - 410 0.05 0.000 1.Sil8I *I 3.4El7 16 7 26 17 C-4 344-1 SA1318. C1 1
- Upper shall -- - - +20 0.20 0.000 1.9 18 4.4 El f 70 - 34 90 S4 C-4 344-2 SA$330. C1 I Upper shell - - - +20 0.20 0.008 1.9Eis 4.4817 70 34 90 S4 C-414 7-1 SA533a. Cl I temer shell - - - -20 0.12 0.013 8.9tle 4.4417 46 22 26 2 C-4347 2 SA5335. C1 1 Lower shell - - - +45 0.12 0.013 8.9E18 4.4517 46 22 91 67 WF-169 Weld Upper etre (602) +123 - No (+20) (-) (-) -- 3.4tl7 - 33 -- 53 sA-1769 Weld Upper etre (401) +B23 - Tee (+20) (-) (-) 1.5:16 - los - las -
WF-8 Weld Upper long. (l002) - 17 Tee (620) (-) (-) 3.8510 4.2B17 880 53 330 73 WF-lO Weld Upper lens. (l005) - Il Tee (+20) (-) (-) 1.8518 4.2E17 ISO 53 530 73 WF-70 Wald Middle etre (l001) -6 - Tee (+20) (-) (-) 8.9EIS 4. 4 E17 IS7 76 177 96 sA-1580 Weld Lauer long. (l001) - 20 fee (+20) (-) (-) 5.6Els 3.7887 114 55 134 75 WF.IS4 Wold Lower cire (1001) -249 - Tea (+20) (-) (-) 1.1886 2.5815 10 mes. 30 20 Atypical Wold Middle etre (l001) 6 - Tee +90 I == = l.9818 4.4E17 44( 22I *I 134 112 g M ( ) = Estimated value per SAS-80046A. Rev 1. July 1977.* I (-)
- Velves per BAW-15ttr. Orteber 1990.g F
- For Regulatary Cutde I.99. Rev. l.'
Per SAW-101444 Rebroery 1980.* DO tv 4 ha 93 ba. O Q e 57 A S O I to D o
- E.
- s
- me- m O to
(-) v M e-.
4 i Revision 1 (6/30/82) l j Figure 8-1. Fast Neutron Fluence of Surveillance Capsule Center j Compared to Various Locations Through Reactor l , Vessel Wall for First 8 EFPY n m
^ 2 w
1 e'
=
= ,e 6 , sto1*
- I ~
g CEME' gs@ C N
=
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E
*CR 3B CAPSULE LOCATIQN i
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=
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i Revision 1 (6/30/82) , t l 4 l 1 l I l 9. l
SUMMARY
OF RESULTS i 4 The analysis of the reactor vessel material contained in the first surveil-lance capsule CR3-B removed from the Crystal River Unit 3 pressure vessel
! led to the following conclusions:
- 1. The capsule received an average fast fluence of 1.05 x 10 18 n/ca z
, (E > 1 MeV). The predicted fast fluence for the reactor vessel T/4 location at the end of the first fuel cycle is 1.58 x 10 17 n/cm 2 (E > 1 MeV). l
- 2. 2 The fast fluence of 1.05 x 10 e n/cm (E 2
> 1 MeV) increased the RT ET f 2
.the capsule reactor vessel core region shell materials a maxistan of 70F.
l 3. Based on a ratio of 5.4 between the fast flux at the surveillance capsule l1 location to that at the vessel wall and an 80% load factor, the projected fast fluence that the Crystal River Unit 3 reactor pressure vessel will 1 receive in 40 calendar years' operation is 1.3 x 101 ' n/cm 2(E > 1 MeV). l1 3
- 4. The increase in the RTg for the base plate material was in good agree-ment with that predicted by the currently used design curves of ART ET
- versus fluence.
- 5. The increase in the RTg for the weld metal was not in good agreement with l
that predicted by the currently used design curves of ARTET **#8"* f1"- \ ence because of the difference in chemical composition.
- 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 l 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.
i ?. l I l t 9-1 Babcock & Wilcox i _ _ _ . _ _ . _ _ , _ _ _ .
l Revision 1 (6/30/82) e
- 10. l SURVEILLANCE CAPSULE RDt0 VAL SCHEDULE ;
Based on the post-irradiation test results of capsule CR3-B, the following schedule is recommended for examination of the remaining capsules in the Crystal River Unit 3 reactor vessel surveillance program: ! Evaluation schedule (a) s Est. capsule 1{1 g. Est. date Capsule fluence, data ID 101 ' n/cn2 Surface 1/4 T available(b) CR3-A Standby - - -- h CR3-C 0.7 0.13 0.07 5,-1983 # CR3-D(C) 1.1 0.21 0.11 L ff 1986 CR3-E 1.7 0.32 0.18 fr 1989 f CR3-F(*) 3.0 0.56 0.31 1994 ' (a)In accordance with BAW-10100A and E-185-79. (b) Estimated date based on 0.8 plant operation factor.
" Capsules contain veld metal compact fracture specimens.
(:.. :J W 10-1 Babcock & Wilcox
l Revision 1 (6/30/82) I
- 11. CERTIFICATION l l
The specimens were tested, and the data obtained from Crystal River Unit 3 surveillance capsule CR3-B were evaluated using accepted techniques and es-tablished standard methods and procedures in accordance with the requirements of 10 CFR 50 Appendixes G and H. J, fS .70 sine $N
- f. L. Lowe, Jr.,4.E. Date Project Technical Manager This report has been reviewed for technical content and accuracy.
$h ha) -
7bc/F/ J.g.Aadland Date Confponent Engineering Certification, Revision 1 EE / TcLa1972
.'L.'Lowe, Jr/( P.E. Date Project Technical Manager 1 aY __
drb7lfr2
. Aadlancf
~~
J. Dat'e C onent Engineering 11-1 Babcock &Wilcox
Revision 1 (6/30/82)
- 12. REFERENCES 1
H. S. Palme, G. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material Sur-veillance Program -- Compliance With 10 CFR 50 Appendix H, for Oconee-Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February 1975. 2 DOT 3.5 - Two-Dimensional Discrete Ordinates Radiation Transport Code, CCC-276 WANL-TME-1982, Oak Ridge National Laboratory, December 1969. 8 CASK Group Coupled Neutron and Gamma-Ray Cross Section Data, RSIC-DLC-23, Radiation Shielding Information Center. C. L. Whitmarsh, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAW-1485 June 1978. 3 K. E. Moore, et al., Evaluation of the Atypical Weldment, BAW-10144A, Babcock & Wilcox, Lynchburg, Virginia, February 1980.
- H. S. Palme, et al., Methods of Compliance With Fracture Toughness and Op-erational Requirements of 10 CFR 50, Appendix G, BAW-10046A, Rev. 1 Babcock & Wilcox, Lynchburg, Virginia, July 1977.
7 1 A. L. Lowe, Jr., et al., Irradiation-Induced Reduction in Charpy Upper Shelf Energy of Reactor Vessel Welds, BAW-1511P, Babcock & Wilcox, Lynchburg, Virginia, October 1980. s Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Pre-dicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regula-tory Commission, Washington, D.C., April 1977. 12-1 Babcock & Wilcox
l l l l APPENDIX A Reactor Vessel Surveillance Program - Background Data and Information A-1 Babcock & WilCOX
- 1. Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E-185-73, are shown in Table A-1. The locations of these materials within the reactor vessel are shown in Figures A-1 and A-2.
- 2. Definition of Beltline Region The beltline region of Crystal River Unit 3 was defined in accordance with the data given in BAW-10100A.
- 3. Capsule Identification The capsules used in the Crystal River Unit 3 surveillance program are iden-tified below by identification number, type, and location.
Capsule Cross Reference Data Number M Location CR3-A A Upper CR3-B B Lower CR3-C A Upper CR3-D B Lower CR3-E A Upper CR3-F B Lower
- 4. Specimens per Surveillance Capsule See Tables A-2, A-3 and A-4.
A-2 Babcock &Wilcox l
T;bla A-1. Unirradinted Imp ct Propertics cnd Re91du21 Element Centent D tc of Beltline Region Materials Used for Selectica of Surveillance Prograai Materials - Crystal River Unit 3 C1.arpy data. CVN Distance. core 7, _ ,,,, Material Battline midplane to Drop wt. Lonattudinal RT Cheeletry. 2 ident. Material region veld centerline. 50 ft-lb. 35 MLE. US E. NDT. heat No. type location em TleT' F At IDF ft-lb F F ft-lb F Cu y S N1 ABM96 5A-508. C1 2 Mossle belt - 20 11.12.13 - -- - 10 0.054 0.008 0.006 -- C-4 344-1 SA-533. Cr 3 Upper shell - -10 39(F) . 32.32 80 -- 88 20 0.20 0.008 0.016 - C-4 344-2 SA-533. Cr a Upper shall - -10 20(F) . 33 80 - 88 20 0.20 0.008 0.016 - C-4 347-1 SA-533. Cr 3 1.awer shell - -20 42.36.61 - -- -- -20 0.12 0.013 0.015 - C-4 347-2 SA-533. Cr B Lower shell - -20 9.9.24 100 - 119 40 0.12 0.013 0.015 -- WF-8 Weld Upper long. - - 45.38.30 - - - - 0.20 0.009 0.009 -- seen (1003) WF-18 Weld Upper toes. - - 45.46.38 - - - - 0.105 0.019 0.004 -- seas (1001) W-169 Weld Upper cire 123 - 36.43.42 - - - - 0.106 0.014 0.013 - eeam (60E OD) 8 5A-1769 Wold Upper cire 123 - 36.35.38 - - - - 0.19 0.021 0.016 - W - sean (40E ID) WF-70 Weld Middle cire -62 - 39.35.44 - - - - 0.27 0.014 0.011 -- essa (1008) 49.41.40 - - - 0.22 0.015 0.013 - SA-1580 Wald lower lens. - - - seam (1001) WF-154 Weld tower long. - - 41.37.43 - - - - 0.20 0.015 0.021 - seas (1005) WF-209-1 Weld Surveillance - -50 29.30.32 103 - 63 43 0.30 0.020 0.005 -- F k. Se Y
=
O O M
4 Table A-2. Test Specimens for Determining Material Baseline Properties 4 No. of test specimens Tension Material description 70F 600F
- CVN impact Compact-tension i
. Heat NN Base metal Transverse direction 3 3 15 - l Longitudinal direction 3 3 15 --
Heat-affected zone (HAZ) ! Transverse direction 3 3 15 -- Longitudinal direction 3 3 15 - stal 12 12 60 - Heat PP Base metal Transverse direction 3 3 15 -- Longitudinal direction 3 3 15 - Heat-affected zone (HAZ) Transverse direction 3 3 15 - Longitudinal direction J 3 _15, -- Total 12 12 60 - 4 Weld metal Longitudinal direction 3 3 15 8 1/2 TCT 4 1 TCI (a) Test temperature to be the same as irradiation temperature. ! ( } Test temperature to be determined from shift in impact transition curves after irradiation exposure. l l ? ' l i I l A-4 Babcock & Wilc0X i l
Table A-3. Specimens in Upper Surveillance Capsules (Designation A C, and E) No. of test specimens Material description Tension CVN impact Weld metal 2 12 Weld, HAZ Heat NN, transverse -- 12 Heat PP, transverse -- 6 Base metal Heat NN, transverse 2 12 Heat PP, transverse -- 6 Correlation material -- 6 Total per capsule 4 54 Table A-4. Specimens in Lower Surveillance Capsules (Designation B, D, and F) No. of test specimens 1/2 T Material description Tension CVN impact compact tension * , Weld metal 2 12 8 Weld HAZ Heat NN, transverse -- 12 -- , Base metal Heat NN, transverse 2 g -- Total per capsule 4 36 8 (a) Compact tension specimens precracked per ASTM E399-72. l A-5 Babcock &Wilcox l
Figure A-1. Location and Identification of Materials Used in Fabrication of Crystal River 3 Reactor Pressure Vessel m I p' - 1 L l f N . i I , - l Cl/ 123V190 - Nozzle Belt i (ABM96)
,_ WF-169 - 60%(OD)
SA-1769 - 40%(ID)
^ E C-4344-1 C
o $ # C-4344-2 upper Shell i O E i T WF-70(100%) 1 n G O C-4347-Il y e% C C-4347-2[LwerShell 1 0 f N $ c - WF-154(100%) Y 124W295 val-Dutchman l l 1 A-6 Babcock & Wilc0X l
I Figure A-2. Location of Longitudinal Welds in Upper and Lower Shell Courses W
)
17* i
'l X Z
Upper Shell 17' k Y W 20' Z II Lower d 3 Shell 20' Y A-7 Babcock & WilCOX
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+; m ..*
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\l' l
E i APPENDIX B Preitradiation Tensile Data B-1 Babcock & WilCOX
i Table B-1. Preirradiation Tensile Properties of Shell Plate Material Heat C-4344-1 Strenath, psi Elonnation, % Specimen emp, , ar a. No. F Yield Ult. Unif. Total % Longitudinal NN-801 RT 70,000 91,875 ES 27 69 NN-802 RT 70,000 91,875 ES 25 69 NN-803 RT 69,375 91,875 11 28 70 Mean RT 69,790 91,875 11.0 26.6 69.3 Std dev'n 360 0 0 1.5 0.6 NN-804 580 63,750 89,375 14 26 61 NN-805 580 64,375 90,000 12 25 64 NN-806 580 64,375 90,625 11 27 63 Mean 580 64,160 90,000 12.3 26 62.6 Std dev'n 360 625 1.5 1.0 1.5 Transverse NN-602 RT 70,000 92,500 ES 28 64 NN-603 RT 69,375 92,500 11 29 61 NN-606 RT 68,750 91,875 16 16 62 Mean RT 69,375 92,290 13.5 24.3 62.3 Std dev'n 625 360 3.5 7.2 1.5 NN-609 580 63,125 90.625 10 22 51 NN-610 580 64,375 90,625 10 24 60 NN-616 580 64,370 90,625 10 23 58 Mean 5 80 63,950 90,625 10.0 23.0 56.3 Std dev'n 720 0 0 1.0 4.7 ES = Extensiometer slipped. B-2 Babcock &Wilcox
4 Table B-2. Preirradiation Tensile Properties of Shell Plate Material. HAZ, Heat C-4344-1 i Specimen emp, ***"I *" I *E**I "" o r a,
- No. F Yield Ult. Unif. Total %
i' Lonnitudinal i NN-501 RT 70,000 92,500 10 29 69 , NN-502 RT 69,375 92.500 10 28 69 NN-503 RT 67,500 89,375 ES 28 70 j Mean RT 68,960 91.460 10.0 28.3 69.3 i 4 Std dev'n 1,300 1,805 0 0.6 0.6
- NN-505 580 62,500 88.750 10 27 63 NN-506 580 61,250 88.750 14 26 60 NN-507 580 61,200 88.750 10 24 65 Mean 580 61,650 88,750 11.3 25.6 62.6 Std dev'n 740 0 2.3 1.5 2.5 '
Transverse NN-301 RT 69,375 91.875 9 22 63 NN-302 RT 70,000 91,875 8 23 65 NN-303 RT 68,750 91,875 10 22 62 Mean RT. 69,375 91.875 9.0 22.3 63.3 Std dev'n 625 0 1.0 0.6 1.5
. NN-304 580 63.750 85,000 8 17 56 NN-305 580 63,125 86,875 9 18 57 NN-306 580 62,500 86,250 9 17 54
! Mean 580 63,125 86,040 8.6 17.3 55.6' Std dev'n 625 955 0.6 0.6 1.5 L ES = Extensiometer alipped. 3
\
k B-3 Babcock & WilCOX
Table B-3. Preirradiation Tensile Properties of Weld Metal, WF-209-1 Longitudinal Specimen e , Strenath, psi Elongation, % ", No. F Yield Ult. Unif. Total % PP-015 RT 75,625 92.500 ES 28 64 PP-016 RT 78,125 95,000 12 28 62 PP-019 RT 77,500 94,063 12 31 62 Mean RT 77,080 93,855 12 29.0 62.6 Std dev'n 1,300 1,260 0 1.7 1.1 1 PP-008 580 69,375 90,000 12 21 50 PP-009 580 66,250 87,500 12 22 53 PP-013 580 66,875 88,125 10 17 43 Mean 580 67,500 88,540 11.3 20.0 48.6 Std dev'n 1,650 1,300 1.1 2.6 5.1 ES = Extensiometer slipped. 1 l l 1 l B-4 Babcock & Wilc0X
APPENDIX C Preirradiation Charpy Impact Data C-1 Babcock & Wilcox
l l Table C-1. Preirradiation Charpy Impact Data for Shell Plate Material -- Longitudinal Direction, Heat C-4344-1 Test Absorbed Lateral Shear Specimen teep, energy, expansion, fracture, l No. F ft-lb 10-3 in. % l l NN-805 -20 40 35 10 NN-809 -10 41 33 10 NN-803 0 39 34 20 4 NN-817 10 41 34 20 NN-818 15 57 46 30 NN-812 25 54 46 20 NN-802 55 12 10 10 NN-811 55 64 54 30 NN-804 70 77 55 40 NN-801 85 79 66 50 NN-813 90 102 75 80 NN-810 100 121 83 80 NN-814 120 116 84 80 NN-806 160 124 86 100 NN-815 330 121 90 100 l l 1 l C-2 Babcock &Wilcox l
l i i Table C-2. Preirradiation Charpy Impact Data for Shell Plate Material -- Transverse Direction, Heat C-4344-1 l l Test Absorbed Lateral Shear Specimen temp, energy , expansion, fracture, I No. F ft-lb 10-3 in. % NN-631 -20 20 18 0 NN-621 0 24 22 10 NN-683 20 34 30 10 NN-682 40 37 34 20 NN-619 60 50 47 20 NN-663 80 59 53 40 NN-680 100 68 60 50 NN-654 110 69 58 50 NN-611 125 73 64 80 NN-671 140 81 72 100 NN-684 155 93 81 100 NN-679 175 95 79 100 NN-659 200 92 74 100 NN-686 260 95 79 100 NN-603 325 93 78 100 I i C-3 Babcock & Wilcox
Table C-3. Preirradiation Charpy Impact Data for Shell Plate
- Material - HAZ, Longitudinal Direction, Heat C-4344-1 Test Absorbed Lateral Shear Specimen temp, energy, expansion, fracture, i
No. No. ft-lb 10-3 in. % NN-516 -40 41 30 30 l NN-507 -20 27 21 20 l NN-504 0 48 41 40 l NN-503 40 60 46 55 NN-513 60 65 52 90 NN-511 80 59 44 65 NN-501 90 77 62 70 ! NN-512 100 82 62 100 l NN-506 140 102 73 100 I NN-509 170 87 71 100 ) l NN-505 200 93 73 100 l NN-510 260 96 69 100 l NN-508 320 89 70 100 i i l I l l C-4 Babcock &Wilcox
i Table C-4. Preirradiation Charpy Impact Data for Shell Plate Material -- HAZ, Transverse Direction, Heat C-4344-1 Test Absorbed Lateral Shear Specimen temp, energy, expansion, fracture, No. F ft-lb 10-3 in. % NN-331 -60 16 13 10 NN-304 -40 16 14 20 NN-308 -20 31 28 30 NN-309 0 38 34 30 NN-334 20 55 41 60 NN-313 40 43 42 50 NN-336 60 51 43 100 NN-326 80 67 57 100 NN-324 100 53 50 100 NN-343 110 60 55 100 NN-342 140 83 67 100 NN-305 170 86 73 100 NN-307 200 75 63 100 NN-332 260 77 63 100 NN-302 330 87 73 100 C-5 Babcock & Wilcox
Table C-5. Preirradiation Charpy Impact Data for Weld Metal Test Absorbed Lateral Shear Specimen temp, energy, expansion, fracture, No. F ft-lb 10~8 in. % l PP-047 0 21 20 20 PP-050 15 27 26 30
- PP-030 30 31 34 30 l
PP-042 40 34 32 30 l 1 PP-023 55 30 30 30 , PP-027 75 36 36 40 PP-048 90 44 46 100 PP-040 105 46 45 100 l l PP-009 120 56 63 100 I PP-012 135 39 42 100 PP-013 150 63 72 100 i PP-025 200 68 69 100 PP-045 230 78 77 100 PP-020 260 79 77 100 PP-019 300 79 78 100 C-6 Babcock & Wilcox
Figure C-1. Charpy Impact Data From Unirradiated Base Metal, Longitudinal Orientation Irn , , , , eo e en 15 - . J
.B. <
4 s0 ___________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ j aT 25 -
# e n
' ' ' l ' ' ' ' ' '
.080 , , , , .
i
$ .%0 - -
5 O .040 E a
- t .020
; e a
f l l l l t I t I a e 200 , , , , , , , , , , , DATA SUM!'ARY Ign T -10F . NDT SF TCV (35 MLE) 160
-TCV (50 FT-LB) 14F ND TCV (30 FT-LB) . 140 - CV-USE (AVG) 124 ft-lb -
f RT -10F NDT _ E 120 - * ~ - l
-- e
! 1 l * - i s 100 - 3 w W - 80 -
,. e V
i
~
t ! M - t
~ ~
kroeng SA533, Cr B Onsewrarian Loneitudfn=1 20 - Fu4mes None
- litaf Rween C-4344-1 f f f f $ t I i t a 3 0
-so -40 0 40 so 120 1so 200 240 2s0 320 3rio a Test Tosianarvas, F C-7 Babcock & Wilcox
Figure C-2. Charpy Impact Data From Unirradiated Base Metal. Transverse Orientation 100 , , , , , _ , , , _ ,_ ,
.* 75 - -
J 5 yso - - - - - - - - - - - - . - - - - - - . . - - - - - - - - - - - - - - - - l s
- T -
in 25 - 0
'E s e, s a s r e i
- l o
5 .060 - 5 1 , w .040 - - 1 5 - - - - - - - - - :._------------------------__. e i
# .020 - - i 5
a
, i .
,g , e e i e i e .
200 , , , , , , , , , , , DATA
SUMMARY
180 .T NDT
-10F ,
TCV (35 MLE) 38F 160
-TCV (50 FT-I2) 61F -
TCV (30 FT-LB) 16F 140 - Cy-USE (AVG) 94 f t-Ib - t
- RT NDT IF 5 120 -
c 3 J 100 -
- l > .a - I g e .
W w so - t O. ! k
-- 60 - -
W - - l l hTenig SA533, Cr B l I biStfaTICII Tra n ava rme ] l N - ham Fueles I Ilmar lhsean C-4344-1 1 f f i i f n a a e e a 0
-so -40 o 40 so no Iso 200 2eo 2so 320 yo 400 Test To esearvas, F C-8 Babcock 8 Wilcox
l Figure C-3. Charpy Impact Data From Unirradiated Heat-Affected Zone Base Metal. Longitudinal Orientation Inc , , , , . _ , , e q 75 -
.E.
y ,0 .__ _ - - - _
=
3M - 0
*E a s I s s g a u a e *
- =
9 5 060 e f - 3 O .040 - 3 5 W i .020 - - S E I_ t t f I I t t i e e a e a s a a u a a a e DATA
SUMMARY
180 T NA . NUr TCV (35 MLE) -35F 160 28F -
-TCV (50 FT-LB)
TCV (30 FT-LB) -72F
. 140 - g-USE (AVC) 95 f t-lb
- T' O RT -32F
, NDT 5 120 -
c. 3 100 - a W E e e W ia 80 - - I O. k ,
~M - e e -
j l 40 . e - ! hisang SA533. Cr B l i e Ontentation HAZ-Lonnit.
~ "
l Funner None l kar Nuesta C-4344-1 I t I f f I t e a e i e 80 -40 0 40 to 120 160 200 240 280 320 MO 400 Test focenaruns, F C-9
Figure C-4. Charpy Impact Data From Unirradiated Heat-Affected Zone Base Metal. Transverse Orientation W i . 3 . __ _ ._ - - .- s
** 75 - -
y 50 __- _ _ _ _ _e.__________-._______________. . 5 y * * - I w 25 - g i t f f I e a e e a n
.080 : , , , , , s , s , , 1
* * )
_4 , e e - l E' .060 -
, )
- e 5 e O .040 -
5
=
5 < 1
# .020 .3 t f f f f i e e i n e y
25 s : e e a a e e a DATA
SUMMARY
l T NA . 180 NDT-3F TCV ( 5 EE) 160 TCV (50 FT-LB) SOF TCV (30 FT-LB) -19F a 140 - CV-USE (AVC) 84 f t-lb - \ l i RT NDT
-10F
' -5 120 8 3 100 - W e ia 80 - -
* , e
'A t e
- g . , .
- ___t.____________________________.
w .
.%7salai, n -911. ce n bleffaflesi MA7-Trahav._ ~
20 - Fumeca None litar liuseen C-4344-1 0
-so -40 0 40 so 12o Iso 200 240 2s0 320 560 vm Test femenarvas, F c 10 Babcock & Wilcox
Figure C-5. Charpy Impact Data From Unirradiated Weld Metal Inn a 75 - -
.E.
- y 50 . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _. _ _ _ _ _ _ _ _ _ _ ._.
m 5 Ji 25 - - i i i e i e i e i . n
.080 , , , , , ,
5 e { ,%0 - - E i
- e
- 6's .040 - -
3 ...-- _. _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . E i .020 - - e i e
,g i i i i e i i i 200 , , , , , , , , , , ,
DATA
SUMMARY
ggn .TNDT -20F , 74F TCV (35) MLE) 160 .TCV (50 FT-LB) 112F . TCV ( LB) 36F
. 140 . C..-USE v (AVG) 79 ft-lb -
A RT 52F NDT E 120 - 8 -
.:' 100 -
3 Y w N - V t e
/
so .
. . . ., / . . . . . . _ . . . . . _ . . . . . . . . . .
hrsataa. Weld Metal Ibtiswfafion NA 20 - p None - - 10:47 liusosa UF-209 e i e i e i e i e i i o
-80 -40 0 40 80 120 160 200 240 280 320 360 W)0 fast Tesessanos, F c-11 Babcock & Wilcox
APPENDIX D Threshold Detector Inforumtion D-1 Babcock & Wilcox
Table D-1 lists the composition of the threshold detectors and the thickness of cadmium used to reduce competing thermal reactions. Table D-2 shows the cycle 1 measured octivity per gram of target material (i.e., per gram of ura-nium, nickel, etc.) corrected for the wait time between irradiation and count-ing. Activation cross sections for the various materials were flux weighted with a 2ssU spectrum (Table D-3). Table D-1. Detector Composition and Shielding Monitors Shielding Reaction Cd-Ag 0.02676 in. Cd " 88 10.38 wt % U-Al U(n f) 2s7 1.61 wt % Np-Al Cd-Ag 0.02676 in. Cd Np(n,f) Ni Cd-Ag 0.02676 in. Cd ssNi(n.p)5'Co 0.56 vt % co-Al cd-0.040 in. Cd "Co(n,y)Co 0.56 wt % Co-Al None ssCo(n,y)C0 s6 y,(3,p)s"Mn Fe None i l I D-2 Babcock & Wilc0X
Table D-2. Crystal River 3 Cycle 1 Neutron Dosimeters (Irradiation Ended March 3, 1978) Nuclide activity, pCi/gm of pCi/gm of Monitor material Wt, gm Reaction Nuclide uCi material (s) target (b.c) Set ED-1 ase U-Al 0.0800 23eU(n.f)Fp '3 Zr 0.414 5.18 50.3 188 Ru 0.477 57.8 5.% lRu 0.0904 1.13 11.0 187 Cs 1.27 0.0105 0.131 242 Ce 50.1 0.413 5.16 1""Ce 0.202 2.53 24.6 2s7 Np-Al 287 Np(n f)Fp ssZr 0.212 3.74 260 y 0.0566
" 183 Ru 0.225 3.98 276
. lRu 0.0371 0.656 45.6 187 Ce 0.00461 0.0815 5.66 1""Ce 0.0860 1.52 106 se co 1,090 Ni 0.133 seNi(n.p)seCo 98.0 737
Ni(n.p)soCo Co 0.141 1.06 4.05 Co-Al (5/8 in.) 0.0189 8 'Co (n ,y) ' 'Co Co 1.45 76.8 13,700 in Cd g
Co-Al (1/2 in.) 0.0151 5'Co(n,y)Co Co 6.28 416 74,300 h Fe 0.156 s" Fe (n .p) s =g, s"Nn 4.35 27.9 479 g se Fe(n e y)5'Fe Fe 11.9 76.5 23,200 8 M
Table D-2. (Cont'd) Nuclide activity, pCi/gm o pCi/gm of Monitor asterial Wt. am Reaction Nuclide pCi material {a) target (b c) Set 3D-2
*3'U-Al 0.0766 2se U(n,f)Fp ss Zr O.367 4.79 46.5 188 Ru 0.433 5.65 54.8 lAu 0.0715 0.933 9.05 187 Cs 0.00988 0.129 1.25 1"3 Ce 0.363 4.74 46.0
***Ce 0.167 2.18 21.2 237 Np-Al 0.0541 as7yp (,,g)pp ss Zr 0.233 4.31 299 l'8 Ru 0.234 4.33 301
+ 1"Ru 0.0381 0.704 48.9 187 Ce 0.00539 0.09 % 6.92 1"'Ce 'O.100 1.85 128 seygg,,p)seCo se co 1,040 Ni 0.132 93.5 708 seygg,,p)soCo Co 0.136 1.03 3.94 Co-Al (5/8 in.) 0.0190 "Co(n,y)Co Co 7.01 74.7 13,300 -
in cd co-Al (1/2 in.) 0.0155 seco (n,y)Co Co 1.16 369 65,900 m Fe 0.154 s*y,(,,p)ssy , s*Mn 3. % 25.7 442
**Fe(n,y)ssy, ssFe 10.3 67.2 20,400 I
M
T-bla D-2. (Cont'd) Nuclide activity, pCi/gm of pCi/gm of Monitor material Wt, am Reaction Nuclide pCi material (a) target (b.c) Set B-3 2se U-Al 0.0792 2sey (,,g)pp , ssZr 0.367 4.63 44.9 1
. '8Ru 0.588 7.43 72.1 lRu 0.108 1.36 13.2 187 Cs 0.0116 0.147 1.43 1" Ice 0.501 6.33 61.4 1""Ce 0.247 3.12 30.3 as7 Np-Al 0.0528 2s7pp (n,g)pp ssZr 0.296 5.60 389
& l'3 0.317 6.01 417 Ru 188 Ru 0.0491 0.929 64.5 187 Ca 0.00713 0.135 9.38 1""Ce 0.120 2.28 158 ss Ni(n.p)"'Co se co 1,310 Ni 0.133 118 887
Ni(n.p)Co Co 0.162 1.22 4.66 Co-Al (5/8 in.) 0.0191 s'Co(n,y)Co Co 1.22 98.3 17,600 in cd F se co(n,y)Co Co 0.197 482 86,100 Co-Al (1/2 in.) 0.0151
""Fe(n.p)5"Mn ss Mn 4.99 32.4 557 st- Fe 0.154 P se Fe(n,y)"'Fe ss Fe g 13.6 88.1 26,700 W
E
. . _ __ _ _ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ . . _ _ _ _ _ __ __ _ . . . __ .m. _ _ _. _ Table D-2. (Cont'd) Nuclide activity, pCi/gpa of pCi/p of Monitor material Ut, en Reaction Nuclide pCi material (s) taraet(b.c) Set ED-4 88'U-Al 0.0480 2seU(n.f)Fp Zr 0.177 3.68 35.7 188 Ru 0.222 4.63 44.9 lRu 0.0391 0.814 7.90 187 Cs 0.00445 0.0928 0.901 l'1 Ce 0.177 3.68 35.7 1Ce 0.0787 1.64 15.9 27 as7y a Np-Al 0.0752 p (,, g) y, ssZr 0.252 3.35 233 b 1Ru 0.271 3.61 251 l"Ru 0.0411 0.547 38.0 1s7 Cs 0.00605 0.0804 5.58 1"Ce 0.103 1.37 95.1 Mi seygg,,p) seco se 0.139 co 73.7 530 782
' 'Ni (n . p) * 'Co Co 0.105 0.757 2.89 f Co-Al (5/8 in.) 0.0195 ssCo(n,y)Co Co 0.710 49.6 8,860 in cd Co-Al (1/2 in.) 0.0155 s'Cc(n,y)Co **Co 0.107 262 46,800 3r Fe 0.159 Fe(n.p)Mn s=Mn 3.12 19.6 -
337 s= g s eFe(n,Y) Fe 5'Fe 7.54 47.4 14,400 h M
Tcb19 D-2. (Cont'd) (* These data are the disintegration rates per gram of wire.
}These data are the disintegration rates per gram of target nuclide; viz., '8'U. a s7pp ,
seN i, Ni, Co. stFe and Fe.
- The following abundances and weight percents were used to calculate the disintegration rate per gram of target element:
assU - 10.38 wt %; 99.27% target nuclide 237 Np - 1.44 wt %; 100% target nuclide Ni - 100 wt %; 67.77% seNi target nuclide. 26.16% Ni target nuclide Co - 0.56 wt %; 100% 5'Co target nuclide Fe - 100 we %; 5.82% stFe target nuclide. 0.33% seFe target nuclide. 7
~
F I E w R
Table D-3. Dosimeter Activation Cross Sections (*} Energy range, MeV as7 Np 2se g seNi st y, _G, I 1 13.3-15.0 2.321 1.073 0.4598 0.425 . I 2 10.0-12.2 2. 340 0.981 0.622 0.537 i 3 8.18-10.0 2.308 0.991 0.659 0.583 4 6.36-8.18 2.09 0.9165 0.638 0.572 5 4.96-6.36 1.54 0.600 0.540 0.473 6 4.06-4.96 1.533 0.562 0.403 0.325 7 3.01-4.06 1.616 0.553 0.264 0.206 8 2.46-3.01 1.691 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 12 0.55-1.11 1.217 0.008 0.00048 -- 13 0.111-0.55 0.1946 0.00013 -- -- 14 0.0033-0.111 0.0410 -- -- -- (* ENDF/4 values flux weighted with a fission spectrum from memo, L. A. Hassler to Distribution " Documentation for Constants and Procedures used to Calculate Dosimeter Activities for RVSP Cap-sules," LR-845-7854-06, November 27, 1978. D-8 Babcock & WilCOX
CONTENTS Page
- 1. INTRODUCTION . . ......................... 1-1
- 2. RACKGROUND . . . . . . ...................... 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . . . . . . 3-1
- 4. PREIRRADIATION TESTS . . . . . .................. 4-1 4.1. Tensile Tests ....................... 4-1 4.2. Impact Tests . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.3. Compact Fracture Tests . . . . . . . . . . . . . . . . . . .
. 4-2
- 5. POST-IRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Thermal Monitors . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2. Tensile Test Results . . . . . . . . . . . . . . . . . . . . 5-1 5.3. Charpy V-Notch Impact Test Results . . . . . . . . . . . . . 5-1
- 6. NEUTRON DOSIMETRY ........,................ 6-1 6.1. Introduction . . . . .................... 6-1 6.2. Analytical Approach .................... 6-2 6.3. Results .......................... 6-3 6.4. Summary of Results . . ................... 6-4
- 7. DISCUSSION OF CAPSULE RESULTS .................. 7-1 ,
7.1. Prairradiation Property Data . . . . . . . . . . . . . . . . 7-1 7.2. Irradiated Property Data . . . . . . . . . . . . . . . . . . 7-1 i 7.2.1. Tensile Properties . . . . . . . . . . . . . . . . . 7-1 1 7.2.2. Impact Properties ................. 7-1
- 8. DETERMINATION OF RCPB PRES 8URE-TMPERAftRE LIMITS ,....... 8-1 StBSWlY OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . .
- 9. 9-1
- 10. SURVIILLANCE CAPSULE REDW7AL SCEEDULE .............. 10-1 i
- 11. CERTIFICATION .......................... 11-1
- 12. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . .' . . 12-1
- 111 - BeM &hx
R:vicion 1 (6/30/82) CONTENTS (Cont'd) Page APPENDIX A. Reactor Vessel Surveillance Program - Background Data and Inf o rmat ion . . . . . . . . . . . . . . . . . . . . . . A-1 B. Prairradiation Tensile Data . . . . . . . . . . . . . . . . B-1 C. Prairradiation Charpy Impact Data . . . . . . . . . . . . . C-1 D. Threshold Detector Informatiou .............. D-1 List of Tables Table 3-1. Specimens in Surveillance capsula CR3-B . . . . ........ 3-2 3-2. Chemistry and Heat Treatment of Surveillance Materials .... 3-3 5-1. Tensile Properties of Capsula CR3-B Base Metal and Weld Metal Irradiated to 1.0E18 nyt ................ 5-2 5-2. Charpy Impact Data From Capsula CR3-B Base Metal Irradiated to 1.0E18 . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 5-3. Charpy Impact Data From Capsula CR3-5 Heat-Affected Zone Metal Irradiated to 8.4E17 .................. 5-3 5-4. Charpy Impact Data From Capsula CR3-B Weld Metal Irradiated to 1.17E18 n/cm 2 . . . ....... ............. 5-3 6-1. Surveillance capsule Detectors ................ 6-5 6-2. Dosimeter Activatione . . . . . . . . . . . . . . . . . . . . . 6-5 6-3. Normalized Flux Spectra. Flux per MeV for Range . . ... ... 6-6 6-4. Neutron Fluence . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6-5. Predicted Fast Fluence in Pressure Vessel at g Maximum Location . . . ........ ............ 6-7 6-6. Calculated Neutron 71uence for Material Specimens . . . . . . . 6-7 7-1. Comparison of Tensile Test Results .............. 7-4 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Properties . . . . . . . . . . . . . . . . . . . . . . . 7-5 d-1. Data for Preparation of Pressure-Temperature Limit Curves for Crystal River Unit 3. Applicable Through 8 IFPY . . . . . . 8-4 A-1. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials -- Crystal River Unit 3 . . . . . A-3 A-2. Test Speelmens for Determining Material Baseline Properties . . A-4 A-3. Specimens in Upper Surveillance Capsules ........... A-5 A-4. Specimens in Lower Surveillance Capsules ........... A-5 B-1. Prairradiation Tensile Properties of Shell Plate Material. Hea t C-4 34 4- 1 . . . . . . . . . . . . . . . . . . . . . . . . . B-2 B-2. Prairradiation Tensile Properties of Shell Plate Materials. MAZ. Heat C-4344-1 . . . ... ................ B-3 B-3. Prairradiation Tensile Properties of Weld Metal. WP-209-1 Longitudinal ................... ...... B-4 C-1. Prairradiation Charpy lapact Data for Shell Plate Material -- Longitudinal Direction. Heat C-4344-1 . . . . . . . . . . . . . C-2
- iv - Babcock & Wilcox
i i J Rovision 1 (6/30/82) l Table (Cont'd) t Table Fase i ! C-2. Prairradiation Charpy Impact Data for Shell Plate Meterial - ! Transverse Direction. Heat C-4344-1 . . . . . . . . . . . . . . C-3 i C-3. Freirradiation Charpy Impact Data for shell Plate Material - MAZ. Longitudinal Direction. Best C-4344-1 . . . .. ..... C-4 . C-4. Prairradiation Charpy Impact Data for Shell Plate Material - l MAZ Transverse Direction. Meat C-4344-1 . . . . ....... C-5 C-5. Preirradlation Charpy Impact Deta for Wald Metal ....... C-6 D-1. Detector Composition and Shielding . . . . . . . ....... D-2
- D-2. Crystal River 3 Cycle.1 Neutron Dosimeters . . . ....... D-3 D-3. Dosimeter Activation Cross Sections . . . . . . . ....... D-8 2 i l
I List of Flaures , Figure 1 3-1. Reactor Vessel Cross Section . . . . . . . . . . .. ..... 3-4
; 5-1. Charpy Isract Data for Irradiated Base Metal. Transverse 1
Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 3 5-2. Charpy Impact Data for Irradiated Base Matal. Heat- i i Affected Zone . . . . ..................... 5-5 !
. 5-3. Charpy Impact Data for Irradiated Wald Metal . . ....... 5-6 F j 6-1. Predicted Fast Neutron Fluence at Various Locations a through Reactor Vessel Wall for First 8 EFPY . . ...... . 6-8 l1 l ! 7-1. Irradiated Vs Unirradiated Charpy Impact Properties of i Base Metal ............... . . . . ....... 7-6 I
7-2. Irradiated Vs Unitradiated Charpy Impact Properties of
- Heat-Affected Zone Material . . . . . . . . . . . . . . . . . . 7-7 "
7-3. Irradiated Vs Unitradiated Charpy Impact Properties of l Wald Metal .......................... 7-8 j 8-1. Past Neutron Fluence of Serve 111ance Capsule Center Compared , 4 to Various Locations Through Anactor Vessel Us11 for First ' 8 EFFY .. .......................... 8-5 l1 i i 8-2. Reactor Vessel Pressure-Temperature Limit Osrves for Normal ! ! Operation Heatup Applicable for First 4 EFtY .. . . . . . . . . 4-6 i 8-3. Rasetor Vessel Pressure-Temperature Limit Curve for Normal i Opetation - Cooldown Applicable for First 8 IFFT ....... 8-7
- 8-4. Reactor Yessel Pressure-Temperature Limit Curve for Inservice Leak and Rydrostatic Tests. Applicable for First 8 IFFT . . . . 8-8
, A-1. Imcatama and Identif Acatise of Materiale Used la Fabrication i of Crystal River 3 Essetor Pressure Vessel . . . ....... A-6 l A-2. Iscation of Imagitudinal Molds la Upper and Imuer shall i Courses . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-7 ' A-3. Imading Diagram for Test Speciases la Immer Surveillance - ' l Capsula CR3-5 . . . . . . . . . . . . . . . . . . . . . . . . . A-8 l C-1. Charpy Impact Data From Unitradiated Base Metal. Longitudinal , Orientation . . . . . . . . . . . . . . . . . . . . . . . . . . C-7 i C-2. Charpy Impact Data Free Unitradiated Base Metal. Traaeverse j, Orientation . . . . . . . . . . . . . . . . . . . . . . . . . . C-8 l -y. M & Wilcox _ ____.- _. _ . _ _ _ _ _ _ _ _ _ _.- _ _ _ _J
1 4 Finures (Cont'd) Figure Page 1 C-3. Charpy Impact Data From Unirradiated Heat-Affected Zone ) Base Metal. Longitudinal Orientation . . . . . . . . . . . . . . C-9 l C-4. Charpy Impact Data From Unirradiated Heat-Affected Zone l Base Metal. Transverse Orientation . . . ............ C-10 C-5. Charpy Impact Data From Unirradiated Wald Metal ........ C-11 4 l ( I i
- vi - Babcock &Wilcox L ,}}