ML20203K972
| ML20203K972 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/30/1985 |
| From: | Leverant G, Norris E SOUTHWEST RESEARCH INSTITUTE |
| To: | |
| Shared Package | |
| ML20203K961 | List: |
| References | |
| SWRI-7524, NUDOCS 8605010098 | |
| Download: ML20203K972 (140) | |
Text
- _ _ _ _ _.
SOUTHWEST RESEARCH INSTITUTE I
Post Office Drawer 28510, 6220 Culebra Road San Antonio, Texas 78284 1
I I
REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM g
FOR CALVERT CLIFFS UNIT 2 ANALYSIS l
OF 263 CAPSULE I
By l
E. B. Norris I
FINAL REPORT l
SwRI Project 06-7524 I
Prepared For Baltimore Gas & Electric Company Baltimore, Maryland September 1985 I
Approved:
MUS Gerald R. Leverant, Director Department of Materials Sciences I
I PP 28m =8ae PDR
ABSTRACT The first vessel material surveillance capsule was removed from the 263* position in Calvert Cliffs Unit No. 2 reactor pressure vessel during the 1982 refuelling outage.
The analysis of the data indicates that the pressure vessel beltline materials will retain adequate shelf toughness throughout tne 32 EFPY design lifetime.
Heatup and cooldown limit curves for normal operation have been developed for up to 12 and 16 effective full power years of operation.
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TABLE OF CONTENTS Page LIST OF FIGURES iv LIST OF TABLES v
I.
SUMMARY
OF RESULTS AND CONCLUSIONS 1
I L
II.
BACKGROUND 3
III.
DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM 6
IV.
TESTING OF SURVEILLANCE CAPSULE SPECIMENS 12 A.
Shipment, Opening, and Inspection of Capsule 12 B.
Neutron Dosimetry 13 C.
Mechanical Property Tests 23 I
D.
Chemical Analyses 32 V.
ANALYSIS OF RESULTS 38 VI.
HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL 44 OPERATION OF CALVERT CLIFFS UNIT NO. 2 l
VII.
REFERENCES 50 APPENDIX A - NUCLEAR PROJECTS OPERATING PROCEDURES A-1 I
APPENDIX B - HARDNESS TEST DATA, TENSILE TEST RECORDS AND TESTED B-1 SPECIMEN PHOTOGRAPHS APPENDIX C - PROCEDURE FOR THE GENERATION OF ALLOWABLE PRESSURE-C-1 TEMPERATURE LIMIT CURVES FOR NUCLEAR POWER PLANT REACTOR VESSELS I
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LIST OF FIGURES Figure Pace 1
Arrangement of Surveillance Capsules in the Pressure 7
Vessel 2
Vessel Material Surveillance Specimens 9
3 Arrangement of Specimens and Dosimeters in 11 the 263* Capsule 4
Calvert Cliffs Geometry used in Discrete Ordinates 18 Transport Calculation I
5 Azimuthal Dependence of Vessel Wall Fast Neutron 24 Flux (E > 1 MeV) Indexed to Surveillance Capsule Dosimetry Results 6
Radiation Response of Calvert Cliffs Unit No. 2 Vessel 28 Shell Plate D-8907-2 (Longitudinal Orientation) 7 Radiation Response of Calvert Cliffs Unit No. 2 Reactor 29 Vessel Heat-Affected Zone Material 1
8 Radiation Response of Calvert Cliffs Unit No. 2 Reactor 30 Vessel Weld Material 9
Radiation Response of Calvert Cliffs Unit No. 2 Reactor 31 I
Vessel Surveillance Correlation Material (HSST Plate 01 MY)
I 10 Effect of Fast Neutron Fluence on RTNDT Shift, Calvert 40 Cliffs Unit No. 2 B
11 Effect of Fast Neutron Fluence on Cv Upper Shelf Energy, 42 Calvert Cliffs Unit No. 2 12 Reactor Coolant System Pressure-Temperature Heatup 46 I
Limits For Up To 12 EFPY of Operation, Calvert Cliffs Unit No. 2 1
13 Reactor Coolant System Pressure-Temperature Cooldown 47 Limits For Up To 12 EFPY of Operation, Calvert Cliffs Unit No. 2 14 Reactor Coolant System Pressure-Temperature Heatup 48 Limits For Up to 16 EFPY of Operation, Calvert Cliffs Unit No. 2 15 Reactor Coolant System Pressure-Temperature Cooldown 49 Limits For Up to 16 EFPY of Operation, Calvert Cliffs Unit No. 2 iv
I LIST OF TABLES Table Page I
Reactor Vessel Surveillance Materials, 8
Calvert Cliffs Unit No. 2 II Summary of Reactor Operations, Calvert Cliffs 15 Unit No. 2 III Results of Discrete Ordinates Sn Transport Analysis, 19 263* Capsule from Calvert Cliffs Unit No. 2 IV Summary of Neutron Detector Measurements, 21 263* Capsule from Calvert Cliffs Unit No. 2 V
Summary of Neutron Dosimetry Results, 263* Capsule 22 from Calvert Cliffs Unit No. 2 VI Charpy Impact Properties of Longitudinal Plate 26 Calvert Cliffs Unit No.2, 263* Capsule VII Charpy Impact Properties of HAZ Material 26 Calvert Cliffs Unit 2, 263* Capsule I
VIII Charpy Impact Properties of Weld Metal 27 Calvert Cliffs Unit 2, 263* Capsule IX Charpy Impact Properties of Reference Plate 27 I
Calvert Cliffs Unit 2, 263* Capsule X
Constants for Tanh-Fit C Transition Curves 33 y
XI Effect of Irradiation on Surveillance Materials 34 263* Capsule - Calvert Cliffs Unit 2 XII Tensile Properties of Surveillance Materials 35 263* Capsule - Calvert Cliffs Unit No. 2 XIII Check Chemical Analysis Results 37 for Calvert Cliffs 41 XIV Projected Values of RTNDT I
Unit No. 2 I
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SUMMARY
OF RESULTS AND CONCLUSIONS The analysis of the first material surveillance capsule removed from the Calvert Cliffs Unit No. 2 reactor pressure vessel led to the following conclusions:
(1)
Based on a calculated neutron spectral distribution, the 263*
capsule received a fast fluence of 8.06 x 10 neutrons /cm2 (E > 1 MeV) at 18 its radial center line in 4.58 effective full power years (EFPY).
(2)
The surveillance specimens of the reactor vessel beltline materials experienced shif ts in RTNDT ranging from 104*F for the heat af-fected zone material to 69*F for the weld metal as a result of neutron radiation exposure up to the 1982 refuelling outage.
(3)
The heat affected zone material exhibited the largest shift in RT Since this material also has the highest initial (unirradiated)
NDT.
RT it is projected to control the heatup and cooldown limitations
- NDT, throughout the design lifetime of the pressure vessel.
(4)
The calculated estimate of maximum neutron fluence of 5.5 x 1018 neutrons / cm2 (E > 1 MeV) received by the vessel wall accrued in 4.58
- EFPY, Therefore, the projected maximum neutron fluence af ter 32 EFPY is 19 3.8 x 10 neutrons /cm2 (E > 1 MeV).
This estimate is based on a lead factor of 1.48 between the center of the 263* capsule and the point of maximum pressure vessel flux.
(5)
Based on Regulatory Guide 1.99 trend curves and procedures, for the Calvert Cliffs Unit 2 vessel core the projected values of RTNDT beltline region heat affected zone material at the 1/4T and 3/4T positions after 12 EFPY of operation are 118*F and 64*F, respectively.
These values were used as the bases for computing heatup and cooldown limit curves for up to 12 EFPY of operation.
I
I (6)
Based on' Regulatory Guide 1.99 trend carves and procedures, the values of RT for the Calvert Cliffs Unit 2 vessel core beltline re-NDT gion heat affected zone material at the 1/4T and 3/4T positions af ter 16 EFPY of operation are projected to be 134*F and 72'F, respectively.
(7)
Based on the surveillance data, the Calvert Cliffs Unit No. 2 vessel core beltline region materials are projected to retain sufficient toughness to meet the current requirements of 10CFR50 Appendix G through-out the design life of the unit.
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2
I II.
B%CKGROUND I
The allowable loadings on nuclear pressure vessels are determined by applying the rules in Appendix G,
" Fracture Toughness Requirements," of 10CFR50 [1].
In the case of pressure-retaining components made of fer-ritic materials, the allowable loadings depend on the reference stress in-tensity factor (Kyp) curve indexed to the reference nil ductility tempera-ture (RTNDT) presented in Appendix G,
" Protection Against Non-Ductile Failure," of Section III of the ASME Code [2].
Further, the materials in the beltline region of the reactor vessel must be mcnitored for radiation-NDT per the requirements of Appendix H,
" Reactor induced changes in RT Vessel Material Surveillance Program Requirements," of 10CFR50.
The RT is defined in paragraph NB-2331 of Section III of the ASME NDT Code as the highest of the following temperatures:
(1)
Drop-weight Nil Ductility Temperature (DW-NDT) per ASTM E 208 [3];
(2) 60 deg F below the 50 ft-lb Charpy V-notch (C )
y temperature; I
(3) 60 deg F below the 35 mil C temperature.
y The RTNDT must be established for all materials, including weld metal and heat-affected zone (HAZ) material as well as base plates and forgings, which comprise the reactor coolant pressure boundary.
It is well established that ferritic materials undergo an increase in strength and hardness and a decrease in ductility and toughness when 17 exposed to neutron fluences in excess of 10 neutrons per cm2 (E > 1 MeV)
[4].
Also, it has been established that tramp elements, particularly ccpper, and nickel affect the radiation embrittlement response of ferritic materials (5-7].
The relationship between the increase in RT and NDT 5
3
I copper content is defined in Ragulatory Guide 1.99.
Although this docu-ment is being revised by the NRC to reflect a more recent evaluation of neutron embrittlement data, estimates of shifts in RTNDT in this report are based on the current and governing Revision 1 of Regulatory Guide 1.99
[8].
In general, the only ferritic pressure boundary materials in a nuclear plant wnich are expected to receive a fluence sufficient to affect I
RTN3T are those materials which are located in the core beltline region of the reactor pressure vessel.
Therefore, material surveillance programs include specimens machined from the plate or forging material and weld-ments which are located in the core beltline region of high neutron flux density.
ASTM E 185 [9] describes the recommended practice for monitoring I
and evaluating the radiation-induced changes occurring in the mechanical properties of pressure vessel beltline materials.
Combustion Engineering has provided such a surveillance program for the Calvert Cliffs Unit No. 2 nuclear power plant.
The encapsulated C y specimens are located near the I.D. surf ace of the pressure vessel where the fast neutron flux density is between one and two times that at the ad-jacent vessel wall surface. Therefore, the increases (shifts) in tran-sition temperatures of the materials in the pressure vessel are generally less than the corresponding shifts observed in the surveillance specimens.
However, because of azimuthal variations in neutron flux density, capsule fluences may lead or lag the maximum vessel fluence in a corresponding ex-posure period.
The capsules also contain several dosimeter materials for experimentally determining the average neutron flux density at each cap-sule location during the exposure period.
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I The Calvert Cliffs Unit tio. 2 material surveillance capsules also include tensile specimens as recommended by ASTM E 185.
At the present time, irradiated tensile properties are used only to indicate that the materials tested continue to neet the requirements of the appropriate material specification.
This report describes the results obtained from testing the contents of the 263' capsule.
These data are analyzed to estimate the radiaticn-induced changes in the mechanical properties of the pressure vessel at the time of the refuelling outage as well as predicting the changes expected to occur at selected times in the future operation'of the Calvert Cliffs Unit flo. 2 power plant.
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III. DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM I
The Calvert Cliffs Unit No. 2 material surveillance program is de-scribed in detail in CENPD-48 [10), dated August 15, 1972.
Six materials surveillance capsules were placed in the reactor vessel near the vessel wall prior to startup, see Figure 1.
The vertical center of each capsule is at the midpoint of the core.
I The capsules consist of seven compartments which contain Charpy V-notch, and tensile specimens machined from the SA533 Gr B plate, weld metal, and heat-affected zone (HAZ) materials representative of those located at the core beltline.
They also contain Charpy V-notch specimens machined from a reference heat of steel utilized in the HSST program I
(Plate 01 MY).
The chemistries and heat treatments of the vessel surveil-lance materials are summarized in Table I.
All plate and HAZ test specimens were machined from the test mate-rials at the quarter-thickness (1/4T) location af ter performing a simu-lated postweld stress-relieving treatment.
The weld metal specimens were I
taken from the center 6" of the thickness.
Weld and HAZ specimens were machined from a stress-relieved weldment which joined sections of the three lower shell course plates.
The longitudinal base metal Cv specimens were oriented with their long axis parallel to the primary rolling direc-tion and with V-notches perpendicular to the major plate surf aces.
Ten-sile specimens were machined with the longitudinal axis parallel to the plate primary rolling direction.
All mechanical test specimens, shown in Figure 2, were taken at least one plate thickness from the quenched edges of the plate material.
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ARRANGEMENT OF SURVEILLANCE CAPSULES IN THE PRESSURE VESSEL I
7
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TABLE I REACTOR VESSEL SURVEILLANCE MATERIALS CALVERT CLIFFS UNIT NO. 2 l
A.
Chemical Analyses (Weight Percent)[10]
Base Plate Weld Metal Element D-8907-2 D-8907-1/D-8907-3 Si
.21
.17 5
.010
.013 P
.005
.016 Mn 1.22 1.11 I
C
.23
.13 Cr
.11
.05 Ni
.66
.04 Mo
.63
.53 V
.004
.010 Cb
<.01
<.01 g
B
.0001
.0003 Co
.011
.009 N
.006
.008 l
I Cu
.14
.20 Al
.022
<.001 W
<.01
<.01 Ti
<.01
<.01 As
.012
.013 Sn
.006
.004 Zr
.001
.001 B.
Heat Treatment All test material was prepared from the fully heat-treated shell plate.
The test plates were suosequently stress relieved for a total of 39.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at a temperature range of 1125'F - 1175'F, with heat-up and cool-down rates in accordance with Reference 10.
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FICURE 2.
VESSEL MATERIAL SURVEILLANCE SPECIMENS E
9
I The capsule contained 48 Charpy V-notch specimens (12 each longitu-dinal vessel plate material, weld metal, HAZ material, and the HSST refer-ence steel plate) and 9 tensile specimens (3 each longitudinal vessel plate, weld metal, and HAZ material).
The specimen inventury and location within the capsule is shown in Figure 3.
The capsule also contained the following dosimeters for determining the neutron flux density:
I Target Element Form Quantity Titanium Bare wire 3
Iron Bare wire 3
Copper Cd shielded 3
I Nickel Cd shielded 3
Bare wire 3
Ci shielded wire 3
I Uranium-238 Bare foil 3
Uranium-238 Cd shielded foil 3
Sulfur Powder 3
Four eutectic alloy thermal monitors had been inserted in each of the three tensile specimen compartments.
The alloy types and trelting points were as follows:
I Alloy Composition Melting Point (*F) 80 Au - 20 Sn 536 90 Pb - 5 Sn - 5 Ag 558 97.5 Pb - 2.5 Ag 580 I
97.5 Pb - 0.75 Sn - 1.75 Ag 590 I
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$P s7 436 456 45P 46A 44K 43E I
45B 437 435 44L 423 45K g
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67L 66M 667 I
E 677 676 67A E
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116 llL 14M llB 13Y 13U 145 13M 165 l
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FIGURE 3.
ARRANGEMENT OF SPECIMENS AND 00SIMETERS IN THE 263* CAPSULE I
{
11
I IV.
TESTING 0F SURVEILLANCE CAPSULE SPECIMENS The capsule shipment, capsule opening, specimen testing, and report-ing of results were carried out in accordance with the Project Plan for the Calvert Cliffs Unit No. 2 Reactor Vessel Irradiation Surveillance Program.
The SwRI Nuclear Projects Operating Procedures called out in this plan include:
(1)
XI-MS-101-1, " Determination of Specific Activity and Analysis of Radiation Detector Specimens" (2)
XI-MS-103-1, " Conducting Tension Tests on Metallic Specimens" (3)
XI-MS-104-1, "Charpy Impact Tests on Metallic Specimens" (4)
XIII-MS-103-1, " Opening Radiation Surveillance Capsules and Handling and Storing Specimens" Copies of the above documents are given in Appendix A.
A.
Shipment, Opening, and Inspection of Capsule SwRI personnel severed the capsule into two parts between the third and fourth compartments from the tcp.
They also supervised the loading of the capsule parts into the shipping cask, and transported the cask to I
San Antonio, Texas.
h The capsule ccmpartments were opened and the contents identified and stored in accordance with Procedure XIII-MS-103-1.
The specimens and spacer blocks were carefully removed and the specimens were placed in in-dexed receptacles, maintaining identity of each specimen location.
After the disassembly had been completed, each specimen was carefully checked to insure agreement with the identification and location as listed in CENPD-48 [10).
No discrepancies were found.
The thermal monitors and neutron dosimeter wires were removed from the holes in the spacers.
The thermal monitors, contained in quartz I
12 l
I vials, were examined.
The 536*F monitors had fused, and the 558*F moni-tors exhibited partial fusion.
Although it is possible that the capsule was subjected to a short excursion above 558'F, it is concluded that the major portion of the exposure period was below 558'F.
Photographs of the l
thermal monitors are included in Appendix B.
All neutron dosimeters were in the positions called out in CENPD-48 and were correctly accounted for.
However, the Cd shields had partially fused and the U-238 foils nad turned to powder.
B.
Neutron Dosimetry The gamma activities of the dosimeters were determined in accordance with Procedure XI-MS-101-1 using an IT-5400 multichannel analyzer and a Ge(Li) coaxial detector system.
The calibration of the equipment was ac-complished with 54 60 137 Mn, Co, and Cs radioactivity standards obtained from the U.S. Department of Commerce, National Bureau of Standards.
All activities were corrected to the time-of-removal (TOR) at reactor shut-down.
The dosimeter wires were weighed on a Mettler microbalance, and the I
fission monitors were weighed on a Mettler digital balance after these materials had been de-encapsulated.
The weights of the Cd-shielded U-238 dosimeters were questionable because of the possibility of contamination by the fused cadmium.
Infinitely dilute saturated activities (ASAT) were calculated for each of the dosimeters because A347 is directly related to the product of the energy-dependent microscopic activation cross section and the neutron flux density.
The relationship between ATOR and ASAT is given by:
E A TOR m=n A
- I P*(1-e m) (e -\\t )
m SAT lI 5
13
I where:
ATOR time of reactor shutdown;
=
decay constant for the activation product, day-1*
A=
I decay time after operating period m, days; t
=
m T
operating days; and
=
m P
average fraction of full power during
=
g operating period.
The Calvert Cliffs Unit No. 2 operating history up to the 1982 refuelling shutdown, which was used in the calculation of ATOR, is presented in Table II.
The primary result desired from the dosimeter analysis is the total fast neutron fluence (> 1 MeV) which the surveillance specimens received.
The average neutron flux density at full power is given by:
ASAT t4 c 9
where:
e =
ener y-dependent neutron flux density, n/cm -sec; ASAT = saturated activity, dps/mg target element; spgetrun-averagedactivationcrosssection, e =
cm ; and N
number of target atoms per mg.
=
g The total neutron fluence is then equal to the product of the average neu-tron flux density and the equivalent reactor operating time at full power.
In the 263* capsule, the tensile and Charpy specimens were located in a single specimen layer.
Since (as will be shown later) the azimuthal variation in flux density through the capsule width and the vertical vari-ation in flux density from the top to the bottom compartments are small, a single value of fluence can be assigned to the surveillance capsule.
14
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TABLE II I
l
SUMMARY
OF REACTOR OPERATIONS I
CALVERT CLIFFS UNIT NO. 2 Operating Fraction of Irradiation Decay Time I
Period Date Full Power Time After Period (M)
(Month / Year)
(P )
(T )
(t )
m m
m I
1 12/76 0.2161 31.00 2114.0 2
01/77 0.6401 31.00 2083.0 3
02/77 0.8290 28.00 2055.0 4
03/77 0.8834 31.0'0 2024.0 I
5 04/77 0.9094 30.00 1994.0 6
05/77 0.7454 31.00 1963.0 7
06/77 0.8242 30.00 1933.0 l
8 07/77 0.9094 31.00 1902.0 9
08/77 0.7179 31.00 1871.0 10 09/77 0.9179 30.00 1841.0 I
11 10/77 0.6037 31.00 1810.0 12 11/77 0.9332 30.00 1780.0 l
13 12/77 0.9025 31.00 1749.0 l
14 01/78 0.8852 31.00 1718.0 I
15 02/78 0.8938 28.00 1690.0 16 03/78 0.8821 31.00 1659.0 17 04/78 0.6319 30.00 1629.0 I
18 05/78 0.5966 31.00 1598.0 19 06/78 0.6645 30.00 1568.0 20 07/78 0.4930 31.00 1537.0 I
21 08/78 0.8496 31.00 1506.0 22 09/78 0.3987 30.00 1476.0 23 10/78 0.0000 31.00 1445.0 24 11/78 0.3333 30.00 1415.0 I
25 12/78 0.9329 31.00 1384.0 26 01/79 0.6510 31.00 1353.0 27 02/79 0.9389 28.00 1325.0 I
28 03/79 0.8283 31.00 1294.0 29 04/79 0.9856 30.00 1264.0 30 05/79 0.9221 31.00 1233.0 I
31 06/79 0.9307 30.00 1203.0 32 07/79 0.8532 31.00 1172.0 33 08/79 0.8485 31.00 1141.0 34 09/79 0.7203 30.00 1111.0 I
35 10/79 0.8588 31.00 1080.0 36 11/79 0.0000 30.00 1050.0 (Continued)
I 15
I I
TABLE II (CONTINUED)
SUMMARY
OF REACTOR OPERATIONS CALVERT CLIFFS UNIT NO. 2 Operating Fraction of Irradiation Decay Time I
Period Date Full Power Time After Period (M)
(Month / Year)
(P )
(T )
(t )
m m
m I
37 12/79 0.8627 31.00 1019.0 38 01/80 0.6756 31.00 988.0 39 02/80 0.9432 29.00 959.0 40 03/80 0.9489 31.00 928.0 I
41 04/80 0.8462 30.00 898.0 42 05/80 0.9545 31.00 867.0 43 06/80 0.9617 30.00 837.0 I
44 07/80 0.8570 31.00 806.0 45 08/80 0.9007 31.00 775.0 46 09/80 0.7872 30.00 745.0 I
47 10/80 0.8591 31.00 714.0 46 11/80 0.8751 30.00 684.0 49 12/80 0.6774 31.00 653.0 50 01/81 0.5470 31.00 622.0 5
51 02/81 0.0000 28.00 594.0 52 03/81 0.7952 31.00 563.0 53 04/81 0.7311 30.00 533.0 5
54 05/81 0.9939 31.00 502.0 55 06/81 0.9544 30.00 472.0 56 07/81 0.8668 31.00 441.0 I
57 08/81 0.8693 31.00 410.0 58 09/81 0.6881 30.00 380.0 59 10/81 0.9798 31.00 349.0 60 11/81 0.9295 30.00 319.0 I
61 12/81 0.9729 21.00 288.0 62 01/E2 0.9773 31.00 257.0 63 02/S2 0.5010 28.00 229.0 I
64 03/82 0.9832 31.00 198.0 65 04/82 0.9670 30.00 168.0 66 05/82 0.9847 31.00 137.0 I
67 06/82 0.9548 30.00 107.0 68 07/82 0.8290 31.00 76.0 69 08/82 0.7467 31.00 45.0 70 09/82 0.8847 30.00 15.0 I
71 10/82 0.8696 15.00 0.0 I
Total Irradiation Time = 1671.4 EfPD at 2700 MWt I
I 16
E I
Southwest Research Institute performed a two-dimensional discrete ordinates transport calculation with the DOT IV code, a 47-group neutron cross section library (DLC-75/8UGLE-80), a P1 expansion of the scattering matrix, and an 58 rder of angular quadrature.
A one-eighth segment of a plane through the vertical axis was used to model the core, core barrel, pressure vessel, surveillance capsule, and two water regions using R-e coordinates.
All boundary conditions were reflective except for the vacuum condition at the vessel 0.D.
A 29-:one radial by 29-zone azimuthal mesh structure was employed to describe the boundaries of the various structural components.
The salient features of this mesh structure, based on the as-built dimensions of Figure 4, are the core boundary outline and the surveillance capsule detail.
In addition to the physical description, the 00T IV input ircluded the average core power distribution shown for each fuel assembly over the period of operation.
The resulting radial and azimuthal dependence of the fast neutron (E > 1.0 MeV) flux density and energy spectrum within the reactor vessel and surveillance capsules were used to calculate the spectrum-averaged cross sections for the dosimeter reactions and displacements per atom (dpa) [11), and the lead factors which relate the neutron exposure of the capsule at its radial centerline to the I.D., 1/4T and 3/4T pressure ves-sel exposures.
The mesh structure provides for a direct calculation of the 1/4T and 3/4T fluxes, but the I.D. flux is derived for a volume ele-ment extending into the vessel wall.
The pertinent factors obtained from this analysis are summarized in Table III.
I I
I 17
I I
i 0
7 (263 ) Capsule-5 Azimuthal Zones (See Detail)
I 242.8 -l t
e
,f'
\\
\\
\\
\\
i 220.9 -
/
I.
l1l
/j V/
l/
192.4 -
j i
188.0 -
f
/
/
f'y 177.0 -
b
'/
/
I
/ =,,'/ /
'o 1.-.'r- --- l "l:
/'/ ',!
l l l
/
~ ' '/ -
L Ferritic Vessel
/
/,/f
' ' /
i a
- 5. S ladd ng e-r---
I f
j 1
--s..--+.---
{.
6.26 l
l l
l 6.40 5
ll l
l 6.80
- s. S. Core l
7.20 Barrel I
l i
i e
- //
I 7.60 J'
' -. L-.
,J 219.2 %
{7.74 l,/'
218.0 ~
- '-. '- [ ',.
j e' 216.4 ~
j 215.2 --
I
/
II f.jj, V'
- I l.
II I
/,{ ' /
Capsule Detail I
FIGURE 4.
CALVERT CLIFFS GEOMETRY USED IN DISCRETE ORDINATES TRANSPORT CALCULATION I
18
I I
TABLE III RESULTS OF DISCRETE ORDINATES TRANSPORT ANALYSIS 263* CAPSULE FROM CALVERT CLIFFS UNIT NO. 2 A.
Calculated Reaction Cross Sections (barns)(a)
Reaction E>1.0 MeV E>0.1 MeV E<0.1 eV 54Fe(n,p)54Mn 0.129 0.0668 N/A 58Ni(n,p)58Co 0.166 0.0859 N/A 63Cu(n,a)60Co
.0.00177 0.000918 N/A 46Ti(n,p)46Sc 0.0247 0.0128 N/A 238 (n,f)137Cs 0.453 0.233 N/A U
58Co(n,y)60Co 4/A N/A 36.4 B.
Calculated DPA Cross Sections (barns)(D)
Position 5
Surveillance Capsule 642 Vessel I.D.
499 l
Vessel 1/4T 431 Vessel 3/4T 284 C.
Calculated Capsule Lead Factors (c)
Position E>1.0 MeV E>0.1 MeV Vessel I.D.
1.48 1.01 l
Vessel 1/4T 2.44 1.28 Vessel 3/4T 16.7 4.08 I
(a) using 47-group reaction cross sections derived from 620-group I
ENDF/8-V values (b)
Using 47-group dpa cross sections derived from 640-group ASTM E 693 values.
(c)
Lead Factor = Capsule Flux Density / Vessel Position Flux Density I
I 19
I The activities of dosimeters, corrected to the time of shutdown (TOR) and the values obtained for saturated activity (ASAT) are presented in Table IV.
The resulting values computed for the neutron flux densities are summarized in Table V.
The discrepancies in the peak vessel flux values determined from the several dosimeter materials are attributed to the uncertainties in the calculated spectra, the reaction cross sections, the determination of
/N )*
disintegration rates, and the calculation of reaction rates (ASAT O
The estimated overall uncertainty in the flux determination from each dosimeter material determined in accordance with the respective ASTM Methods is as follows:
I ASTN Method and Reaction Est. Uncertainty (IS%)
E523:
Cu-63 (n.a) Co-60 10%
E263:
Fe-54 (n.p) Mn-54 t 6%
E264:
Ni-58 (n.p) Co-58
- 8%
E526:
Ti-46 (n.p) Sc-46 1 13%
E704:
! 11%
Averaging the results obtained from these neutron dosimeters, the fast neutron flux at the 263' capsule location during full pcwer operation is calculated to be 5.58 x 1010 cm-2.sec-1. E > 1 MeV.
It is of interest to note that if other averaging routines are utilized, the results are I
within 5% of the five dosimeter average.
For example, the average of the non-fission monitor flux values is 5.32 x 1010 cm-2 see-1, E > 1 MeV, and 10 the average flux value determined from the iron wires alone is 5.77 x 10 cm-2 sec-1, E > 1 MeV.
20
I TABLE IV
SUMMARY
0F NEUTRON DETECTOR MEASUREMENTS I
263* CAPSULE FROM CALVERT CLIFFS UNIT NO. 2 Dosimeter Cadmium Activation Dosimeter A
A TOR SAT I
(Location)
Shielded Reaction Weight (mg)
(dps/mg)
(dps/mg) 54Fe(n.p)S4Mn 25.6 4.000x103 4.811x103 Fe (Top)
No 3
3 54Fe(n.o)54Mn 24.6 3.761x10 4.524x10 Fe (Middle)
No 54Fe(n.p)54Mn 23.8 3.852x103 4.633x103 Fe (Bottom)
No 4
4 SSNi(n.p)58Ce 22.5 5.351x10 6.179x10 Ni (Top)
Yes 4
4 58Ni(n.p)58Co 22.4 5.097x10 5.836x10 Ni (Middle)
Yes 4
4 SSNi(n.p)58Co 22.3 5.694x10 6.575x10 Ni (Bottom)
Yes 2
2 63Cu(n,a)60Co 21.5 2.690x10 6.326x10 Cu (Top)
Yes 63Cu(n,2)60Co 18.2 2.680x102 6.303x102 Cu (Middle)
Yes 2
2 63Cu(n.o)60Co 21.0 2.868x10 6.745x10 Cu (Bottom)
Yes I
3 1.185x103 40Ti(n.p)46Sc 13.7 1.028x10 Ti (Too)
No 3
3 46Ti(n p)46Sc 13.6 1.074x10 1.238x10 Ti (Middle)
No I
46Ti(n.p)46Sc 13.1 9.164x102 1.056x103 Ti (Bottom)
No 238 (n,f)137Cs(e) 22.4(a) 6.588x102 6.653x103 g
U (Top)
No U
238 (n,f)137Cs(0) 27.8(a) 6.942x10 7.043x103 2
U (Middle)
No U
238 (n,f)137Cs(e) 33.0(a) 5.706x102 5.789x103 V (Botton)
No U
238 (n,7)137Cs(0) 28.5(a c) 1.756x10 1.781x10 2
3 U (Top)
Yes(b) g 238 (n,f)137Cs(e) 23.4(d) 3.777x102 3.833x10 3 U (Middle)
Yes(b)
U 238 (n f)137Cs(e) 47,9(a,c) 1.156x10 1.173x10 2
3 U (Bottom)
Yes(b)
U l
7 7
59Co(n,y)60Co 8.17 1.475x10 3.469x10 Co (Top)
No 7
7 59Co(n,y)60Co 8.31 1.532x10 3.603x10 Co (Middle)
No 7
7 69Co(n,y)60Co 8.20 1.083x10 2.547x10 Co (Bottom)
No 59Co(n,3)60Co 7.86 1.828x106 4.299x106 Co (Top)
Yes 6
6 59Co(n.3)60Co 8.98 1.596x10 3.754x10 Co (Middle)
Yes 6
59Co(n,y)60Co 8.53 1.766x106 4.153x10 Co (Bottom) fes I
(a) Assumed to be uranium metal.
(b)
Cd shield disintegrated at some unknown time during irradiation.
(c) May be contaminated with Cd.
(d)
Atomic Absorption determination by Combustion Engineering, Inc.
(e)
Fission Yield = 6.00%
21
I I
TABLE V
SUMMARY
OF NEUTRON D0SIMETRY RESULTS 263* CAPSULE FROM CALVERT CLIFFS UNIT NO. 2 I
2 Dosimeter Cadmium N
Neutron Flux, 0 (n/cm /sec)
Location Shielded (tarceta! cms /mg)
E>l MeV E>0.1 MeV Thermal (a)
I 17 10 11 Fe (Top) ho 6.2543x10 5.96x10 1.15x10 N/A 17 5.61x1010 1.08x1011 N/A Fe (Middle)
No 6.2543x10 17 10 11 Fe (Bottom)
No 6.2543x10 5.74x10 1.11x10 gjg 18 10 11 Ni (Top)
Yes 7.0039x10 5.31x10 1.03x10 N/A Ni (Middle)
Yes 7.0039x1018 5.06x1010 0.98x1011 N/A 18 10 11 Ni (Bottom)
Yes 7.0039x10 5.66x10 1.09x10 N/A 10 1.05x1011 18 5.45x10 N/A Cu (Top)
Yes 6.5551x10 I
18 10 11 Cu (Middle)
Yes 6.5551x10 5.43x10 1.05x10 N/A 18 5.81x1010 1.12x1011 N/A Cu (Bottom)
Yes 6.5551x10 I
11 18 4.71x1010 0.91x10 N/A Ti (Top)
No 1.0184x10 19 10 11 Ti (Middle)
No 1.0184x10 4.92x10 0.95x10 N/A I
11 10 0.81x10 N/A 18 4.20x10 Ti (Bottom)
No 1.0184x10 18 10 11 U (Top)
No 2.5298x10 6.78x10 1.32x10 N/A 10 1.40x1011 18 7.18x10 N/A U (Middle)
No 2.5298x10 18 10 11 U (Bottom)
No 2.5298x20 5.90x10 1.14x10 N/A 19 Co (Top)
No 1.0218x10 N/A N/A Co (Top)
Yes 1.0218x10 N/A N/A 8.00x1010 19 19 Co (Middle)
No 1.0218x10 N/A N/A 19 10 Co (Middle)
Yes 1.0218x10 N/A N/A 8.49x10 Co (Bottom)
No 1.0218x1019 N/A N/A 19 10 Co (Bottom)
Yes 1.0218x20 N/A N/A 5.61x10 I
(a)
Calculated per ASTM Method E 262 using 5 arn.
=
th I
22
I The calculated azimuthal variation in fast neutron flux at the pres-sure vessel I.D.,1/4T and 3/4T positions, normalized to the surveillance capsule dosimetry result, is shown in Figure 5.
Since the I.D. flux was computed for a mesh element having a midpoint 0.3 in. below the steel /
cladding interface, the results are slightly non-conservative.
At the peak azimuthal position, the steel / cladding interface flux is estimated to be 4% higher than that shown in Figure 5.
The pert'Jrbation of the fast neutron flux by the presence of the surveillance capule is also clearly evident in Figure 5.
This is mani-fested both as p taks (at the capsule and 3/4T radial positions) and de-pressions (at the I.D. and 1/4T radial positions) in the azimuthal fast flux (E > 1 MeV) distributions.
The axial flux distribution was not calculated.
Although the fast flux may peak a little below the axial centerline, the axial peaking f ac-tor is expected to be insignificant (less than 3%).
A comparison of the dosimetry results obtained from each compartment (Table V) confirms that the axial flux variation is small.
Since Calvert Cliffs Unit No. 2 operated for 1671.4 effective full power days (EFPD) up to the October 1982 refuelling, the calculated flu-ence (E > 1 MeV) for the surveillance materials in the 263* capsule is 19 2
8.06 x 1018 2
n/cm.
Other exposure values are 1.56 x 10 n/cm, E > 0.1 MeV, and 1.15 x 10-2 dpa (displacements per atom).
l C.
Mechanical Property Tests Hardness tests were run in accordance with ASTM Method E 18 [13] on each irradiated Charpy V-notch specimen.
The results are presented in Appendix B. The average hardness of each material group is as follows:
I I
23
I ll 10 I
Surveillance Capsule e:
-- z
\\
jk -
x_--
=-
6 I
_s.
y.-
_v-z: ' -:
[--~_.--h---
y
[d 3. O - Q - - = 3 W :-- -- =, _-:-- ; c :=
_r =
.r-.
=
goyy q= _ =1===='"f *=i=.- Ves sel I. D. E m
5 g.===
=.,=g,-
=3 g:
=3_= g.=___
__ _ _g
,3 6ik, bin
- ~~ ~Q:~i? **= ]
~ ~
$45~i w
2l g
.v.
[
rz
[
t
- e,..,}
'h
_a
_J-e _. e- ~
I c
r'
' mg :
g% Vessel 1/4t -"-- ##
i 3
E 1 L,10 5
I
- ;- 0.15 x Peak Vessel I.D. Flux
= --
-f y
6 _= =y =+_: = =m _ _ _
a=====c--
=
__ =
=,
. + = -
=_
_.==
_. _ - _ _ _ =
m L.
I
_..__n_
m
++-=:==r===_=
--= - - = - - - - - -
3
==-2:@t=-9=-
= = =1
= = u- =m---1 + =s= =i= :-p= -.= -
u
_zSL =_ -- p =:= = :
. _.; y 3_3.-
3 I
3
~
r paf=.
3_
2
- ~ ~ ~
~
~ ~ ' - - - - - -
rn
'z c._,
g I
.._a, 9
10
- -- i 1
1 i
r--f---
1- -- t --- -
0 5
10 15 20 25 30 35 40 45 Azimuthal Position, Degree e
FIGURE 5.
AZIMUTHAL DEPENDENCE OF VESSEL WALL FAST NEUTRON FLUX (E > 1 MeV) INDEXED TO SURVEILLANCE CAPSULE DOSIMETRY RESULT I
I 24
I I
Material Rockwell Hardness Base Metal Specimens 95 HRB Weld Metal Specimens 94 HRB HAZ Material Scecimens 94 HRB I
Reference Material Specimens 97 HRB The irradiated Charpy V-notch specimens were tested on an instrumen-ted SATEC Model SI-1K 240 ft-lb, 16 ft/sec impact machine inspected and calibrated using specimens and procedures obtained from the Army Materials and Mechanics Research Center, Watertown, Mass., in accordance with Proce-dure XI-MS-104-1.
The test temperatures, selected to develop the ductile-brittle transition and upper shelf regions, were obtained using a liquid conditioning bath monitored with a Fluke Model 2168A digital thermometer.
The Charpy V-notch impact data are presented in Tables VI through IX.
Photographs of the fracture faces are presented in Appendix B.
The base-line Charpy data obtained on the unirradiated materials were developed by Combustion Ergineering, Inc. [14].
TheshiftsintheCharpyV-notchtrEn-sition temperatures were determined for the vessel plate, the weld metal, the HAZ and the reference materials by comparing computer-fit transition curves for the irradiated materials to those developed using the unirradi-ated test data, see Figures 6 through 9.
The curve fitting computer pro-gram utilizes a Fletcher-Reeves optimization routine to obtain a least squares fit of each material data set to the relationship:
T-T y Y = A + B tanh T
I I
I I
25
I I
TABLE VI CHARPY IMPACT PROPERTIES OF LONGITUDINAL PLATE CALVERT CLIFFS UNIT NO. 2 263* CAPSULE Fracture Lateral Fracture Specimen Temperature Energy Expansion Appearance No.
(*F)
(ft-lb)
(mils)
(% shear) 11B 25 8.5 11 nil 14M 50 7.5 10 2
I 165 75 24.0 25 10 15T 75 33.5 34 5
13Y 100 34.0 34 10 I
13U 120 46.5 43 15 166 140 49.5 49 20 145 160 52.0 52 20 13M 180 75.5 61 80 110 210 104.5 84 100 116 250 117.0 89 100 11L 300 122.5 88 100 I
TABLE VII I
CHARPY IMPACT PROPERTIES OF HAZ MATERIAL CALVERT CLIFFS UNIT N0. 2 263" CAPSULE I
Fracture Lateral Fracture I
Specimen Temperature Energy Expansion Appearance No.
('F)
(ft-lb)
(mils)
(% shear) 43E
-50 11.0 13 nil 45K
-25 18.0 17 10 423 0
23.0 23 10 1
d4L 25 28.0 27 5
I 458 50 20.5 20 10 44K 60 25.0 24 20 45P 75 36.5 34 15 456 140 50.5 48 60 46A 180 49.0 53 80 436 210 92.0 76 100 I
435 250 103.5 85 100 437 300 118.0 89 100 I
I 26
I TABLE VIII CHARPY IMPACT PROPERTIES OF WELD METAL CALVERT CLIFFS UNIT N0. 2 263* CAPSULE Fracture Lateral Fracture I
Specimen Temperature Energy Expansion Appearance No.
(*F)
(ft-lb)
(mils)
(% shear) 337
-50 5.0 6
nil I
332
-25 19.0 19 10 321 0
14.0 16 5
320 10 23.5 25 15 I
32Y 20 25.0 27 20 333 25 38.0 36 15 33P 40 57.0 51 45 I
334 50 50.0 46 20 35K 75 90.5 74 75 32L 140 98.5 83 100 34E 210 105.0 88 100 36M 250 112.5 91 100 I
TABLE IX I
CHARPY IMPACT PROPERTIES OF REFERENCE PLATE CALVERT CLIFFS UNIT N0. 2 263* CAPSULE I
Fracture Lateral Fracture Specimen Temperature Energy Expansion Appearance No.
(*F)
(ft-lb)
(mils)
(% shear)
I 67A 50 9.0 11 nil 660 75 9.0 11 5
676 100 11.0 13 5
677 120 17.5 20 10 67P 140 31.0 30 10 66T 160 31.5 30 20 66A 180 42.0 67 25 66C 210 65.0 57 25 67K 230 84.0 74 100 667 250 87.5 75 100 I
66M 300 94.0 81 100 67L 350 94.0 76 100 I
27
1 I
16.
I 4
3 3
e -
- O
.Q
. O.
0 E
o N
- * ~ " * ~ ~._
....s 820.
9, -
E O
/
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- e, /
c 100.
~
y I
8 v
/
St..
'/,
/
r o
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T 60.9
'/ '
'~
I le
- s L,,,,
s e p I
..,5
+ PLATE D-e987 UN!amclAftD s
/
8 i e...
O"-
" * - PLATE D-e9 7 ImmD!aTED e e t
t i
.M i
-1M.
.M 1M.
20.
300.
400.
TDFERATURE IDEC F3 iM.
i 4
'o I
i
~~~~~
4
~
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- O 9
4
~
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Y
- g,'
C
. 9.
- /.
g,,
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/-
b
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e e I
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+ PLATE D-e9 7-2
- LMIhthe!ATED s
- e e
_. _,1,Tr D-e9et-a - imanianD t
t
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-iM.
iM.
see.
3M.
l nmanTimt ides r FIGURE 6.
RADIATION RESPONSE OF CALVERT CLIFFS UNIT NO. 2 VESSEL SHELL PLATE D-8907-2 (LONGITUDINAL)
I I
28
I 164.
a
" -- -o-- HA TTAL F39e?-t - LMINIABIATE3
~
gg,
- - e - 6ec ETAL F8987 InitADIATO
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toe.
30s.
TEMUt4TURE EKG F3 I
see.
e i
i a
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I t
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20s.
300.
I TDFUtATE IKC F)
I FIGURE 7.
RADIATION RESPONSE OF CALVERT CLIFFS UNIT N0. 2 REACTOR VESSEL HEAT-AFFECTED ZONE MATERIAL I
I 29
I I
tae.
=
i 6
--o-- KLD ETAL D-8967-le tM!MtADIATED 146.
- -e - G CAL FSM7-te IMMIARD S-E I
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b
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300.
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TUPEMTURE (KC F) iM.
6 5
4 i
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m__
=
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a
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.00
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.00 100.
800.
300.
TDPthiJE <KG F1 i
FIGURE 8.
RADIATION RESPONSE OF CALVERT CLIFFS UNIT N0. 2 REACTOR VESSEL WELD MATERIAL I
l 30
I I
166.
I 8
C
-- O---- WSST PLATE 01 RV - LMIRRADIATED o
- g-HS$7 PLATE 01 RV - IRRADIATED E
E I
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TDFDIATUIE tDEC F) 1 E
i FIGURE 9.
RADIATION RESP 0tlSE OF CALVERT CLIFFS NO. 2 REACTOR VESSEL SURVEILLANCE CORRELATI0t1 MATERIAL (HSST PLATE 01 MY) 31
where:
C function (fracture energy or lateral expansion)
Y
=
y C test temperature, deg F T
=
y (T-T y
0 Intercept when tanh A
=
=
I Tg Slope B
=
T1= Temperature at transition midpoint, deg F T2= One-half of transition range, deg F.
The values of A, B, Ti and T2 obtained for each C curve are given in y
determined at the 30 f t-lb Table X.
A summary of the shif ts in RTNDT level as specified in Appendix G to 10 CFR 50 [1], and the reduction in Cv upper shelf energies for each material, is presented in Table XI.
The force-time and energy-time curves obtained from the Charpy machine instru-mentation are on file at SwRI.
Tensile tests were carried out in accordance with Procedure XI-MS-103-1 using a 22-kip capacity MTS Model 810 Material Test System equipped with an Instron Catalogue No. G-51-13A 2-in. strain gage extensometer and Hewlett Packard Model 7004B X-Y autographic recording equipment.
Tensile tests on the plate material and the weld metal were run at room tempera-ture (RT), 250*F and 550*F at a strain rate of 0.005 in/in/ min. through I
the 0.2% offset yield strength using servocontrol and ramp generator.
The results are presented in Table XII.
The load-strain records are included in Appendix B along with the photographs of the tested specimens.
D.
Chemical Analyses Chem' cal analyses were run on 2-3 mm thick samples cut from the i
fracture end of selected tested Charpy specimens halves.
All of the weld specimens and one plate specimen were tested for copper and nickel content at SwRI using an X-ray fleurescent technique [15] applied to the plane I
32
I TABLE X CONSTANTS FOR TANH-FIT C TRANSITION CURVES y
Material / Condition Criterion A
9 T
T i
p Long, Plate /Unirr.
Energy 74.33 72.91 62.02 75.07 Lat. Exp.
49.36 44.25 45.91 59.72 Long. Plate /Irr.
Energy 67.11 56.91 156.67 81.91 Lat. Exp.
54.47 39.22 145.88 76.73 HAZ Material /Unirr.
Energy 66.00 58.00 32.73 84.45 Lat. Exp.
45.73 36.72
- 0.60 102.19 HAZ Material /Irr.
Energy 69.05 44.92 175.55 74.90 Lat. Exp.
46.83 42.28 116.21 142.52 Weld Metal /Unirr.
Energy 68.87 64.13
-11.66 54.39 Lat. Exp.
45.19 45.15
-27.85 59.21 Weld Metal /Irr.
Energy 59.05 48.73 51.94 48.38 Lat. Exp.
46.17 44.65 38.62 59.99 I
HSST Plate /Unirr.
Energy 71.67 70.84 78.05 78.56 Lat. Exp.
45.44 39.02 61.88 64.91 HSST Plate /Irr.
Energy 50.94 40.29 184.17 53.75 I
Lat. Exp.
42.23 36.53 160.96 51.12 I
I I
I I
I I
33
m M
M W
W M
M M
M M
M m
m m
m M
M mm TABLE XI EFFECT OF IRRADIATION ON SURVEILLANCE MAIF. RIALS 263* CAPSULE - CALVERI Cliffs UNIl NO. 2 i
Weld ilAZ Long. Plate Ref. Plate Criterion (1,2)
Metal Material D-8907-2 Material Transition Temperature Shift 9 50 ft-lb 71*F 133*F 95"F 130"F 0 30 ft-lb 69"F 104"F 84*F 128"F 0 35 mil 65"F 107*F 78*F 107"F A RT 69"F 104"F 84"F 128"F HDT W
C Upper Shelf Drop 32 ft-lb 24 ft-lb 31 ft-lb 52 ft-lb y
(23%)
(19%)
(21%)
(37%)
(1) RefertoFigures5grough8 2
(2) Fluence = 8.06 x 10 n/cm, E > 1 MeV
.1 4
l l
i
W M
W W
M M
M M
M M
M M
M M
W M
M M
M TABLE XII TENSILE PROPERTIES OF SURVEILLANCE HATERIALS 263* CAPSULE - CALVERI CLIFFS UNII NO. 2 Fracture fracture Uniform Total Test Spec.
Temp.
0.2% YS UTS Load Stress Elongation Elongation R.A.
Material (a)
No.
( f )
(ksi)
(ksi)
(lb)
(ksi)
(%)
(%) _
(%)
Long. Plate IJ4 75 77.1 100.3 3250.'
189.
18.2 22.5 65.0 0-8907-2 IJ6 250 72.4 93.9 2942.
183.
17.9 24.3 67.3 IK6 550 68.6 91.5 3137.
171.
15.8 21.4 62.5 HAZ Material 4JK 75 79.4 98.4 2784.
176.
17.2 (b) 67.7 M
4JJ 250 73.6 92.7 3009.
193.
16.0 21.8 68.2 4J6 550 70.9 91.8 2761.
163.
17.4 21.0 65.4 Weld Metal 3J6 75 87.0 100.5 3182.
190.
18.0 24.4 65.9 3KI 250 79.7 94.5 2693.
178.
17.1 24.8 69.1 j
3JK 550 79.7 94.5 3347.
171.
15.5 21.5 60.0 18 2
(a) Fluence = 8.06 x 10 n/cm, E > 1 MeV (b) Failed outside gage length.
I surface parallel to and 2-3 mm from the fracture surface.
Due to inherent limitations of the X-ray flourescant technique, one base plate specimen and two of the weld specimens were then sent to Westinghouse Advanced Reactors Division Analytical Labcratory for complete analyses using gravi-metric (Si), combustion (C and S) and ICP Plasma (remainder of elements) methods of analyses.
The results are summarized in Table XIII.
The dif-ferences in the copper and nickel results obtained by the two methods can be attributed to at least three factors:
(1)
The ICP Plasma method measures the chemical content of the I
full volume of sample material while the X-ray flourescent method looks only at the surface of the sample; I
(2)
The 1 square centimeter cross sectional area available from the Charpy samples is snaller than optimum for detecting small amounts of residuals with an X-ray flourescent technique.
(3)
The gamma activity of the sample increases the difficulty in accurately measuring the flourescent peak.
For these reasons, and because of the better agreement with the preservice analyses by Combustion Engineering (see Table I), it appears that the ICP plasma method provides the more reliable results, especially for very low g
nickel contents.
I i
I I
I 36
M M
W W
W M
M M
M M
M M
M M
M M
M M
M TABLE XIII CllECK CilEHICAL ANALYSIS RESULTS Test )
Weight Percent of Element Specimen Identification (d)
Lab (U C
S P
Si Cu Ni Cr Mo V
Mn 165 (Long. Plate)
W
.186
.005
.006
.20
.12
.59
.098
.51 (c) 1.20 165 (Long. Plate)
S
.14
.52 321 (Weld Metal)
W
.091
.006
.011
.13
.24
.05
.051
.45 (c) 1.18 321 (Weld Metal)
S
.34
.27 32U (Weld Metal)
W
.123
.005
.011
.13
.24
.06
.049
.46 (c) 1.21 32U (Weld Metal)
S
.27
.20 32Y (Weld Metal)
S
.30
.21 O
33P (Weld Metal)
S
.28
.15 32L (Weld Metal)
S
.24
.25 35K (Weld Metal)
S
.22
.16 34E (Weld Metal)
S
.18
.15 l
36M (Weld Metal)
S
.22
.30 333 (Weld Metal)
S
.33
.25 334 (Weld Metal)
S
.28
.22 337 (Weld Metal)
S
.29
.21 332 (Weld Metal)
S
.30
.21 (a) Cut from tested C specimen identified by number y
(b) W = Westinghouse Atomic Power Division S = Southwest Research (c) <0.001%
I V.
ANALYSIS OF RESULTS I
The analysis of data obtained from surveillance program specimens has the following goals:
(1)
Estimate the period of time over which the properties of the vessel beltline materials will meet the fracture toughness requirements of Appendix G of 10CFR50.
This requires a projection of the measured reduc-tion in C upper shelf energy to the vessel wall using knowledge of the y
energy and spatial distribution of the neutron flux and the dependence of C upper shelf energy on the neutron fluence.
y (2)
Develop heatup and cooldown curves to describe the operational limitations for selected periods of time.
This requires a projection of the measured shift in RTilDT to the vessel wall using knowledge of the de-pendence of the shift in RTt4DT on the neutron fluence and the energy and spatial distribution of the neutron flux.
The energy and spatial distribution of the neutron flux in the pres-sure vessel was calculated for Calvert Cliffs Unit 2 with a discrete ordinates transport code.
This analysis predicted that the lead f actor (ratio of fast flux, E > 1 MeV, at the capsule location to the peak pres-sure vessel flux) was 1.48 at the vessel I.D. and 2.44 at the 1/4T loca-tion.
Although the transport calculation predicted that the fast flux, E > 1 MeV, at the 3/4T depth in the pressure vessel wall was 9% of that at the I.D.
location, the conservative fast flux attenuation figure of 15%
was utilized in this analysis to allow for the increased fraction of neu-trans which accrue in the 0.1 to 1.0 MeV range in deep penetration situa-tions and which might contribute to the damage.
At the peak I.D. flux I
I 38
I location, for example, 60% of the neutron population having energies greater than 0.1 MeV were calculated to lie in the 0.1 to 1.0 MeV range.
At the 3/4T depth, however, the calculated fraction of neutrons in the 0.1 to 1.0 MeV range increased to 87% of t.he total having energies greater than 0.1 MeV.
The following neutron flux densities were used in this analysis:
2 Location Neutron Flux Density, n/cm /sec, E > 1 MeV 10 1.D.
3.77 x 10 1/4T 2.29 x 1010 9
3/4T 5.66 x 10 A mathod for estimating the increase in RTNDT as a function of neu-tron fluence arr' chemistry is given in Regulatory Guide 1.99, Revision 1
[8].
However, the Guide also permits extrapolation of credible surveil-lance data by constructing response curves through the data and parallel to the Guide trend curves. The Calvert Cliffs Unit 2 response curves con-structed in Figure 10 were used to estimate the value of RTNDT for the three beltline materials (plate, HAZ material and weld metal) after 12, 16 and 32 Effective Full Power Years (EFPY) of operation. These projections, summarized in Table XIV, indicate that the HAZ material will control the vessel beltline RTNDT throughout the lifetime of the pressure vessel.
A method for estimating the reduction in C upper shelf energy as a y
function of neutron fluence is also given in Regulatory Guide 1.99, Revi-sion 1 [8].
The results from the 263* capsule are compared to a portion of Figure 2 of the Regulatory Guide 1.99, Revision 1, in Figure 11.
The shelf energy responses of the pressure vessel surveillance materials from the 263* capsules fall on or below the predictive trend curves of Regula-tory Guide 1.99, Revision 1.
39
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I TABLE XIV FOR CALVERT CLIFFS UNIT NO. 2 PROJECTED VALUES OF RTNDT Initi Adj.
EFPY P.V. Material Location RTNDT )
Fluence (b)
ART RT NDT NDT 19 12 Plate D-8907-2 1.0.
+10*F 1.43 x 10 112 122 18 1/4T
+10"F 8.66 x 10 g7 97 18 3/4T
+10*F 2.14 x 10 43 53 19 HAZ Material I.D.
+10*F 1.43 x 10 139 149 19 1/4T
+10*F 8.66 x 10 138 118 18 3/4T
+10*F 2.14 x 10 54 64 19 Weld Metal I.D.
-60*F 1.43 x 10 92 32 19 1/4T
-60*F 8.66 x 10 72 12 I
18 3/4T
-60*F 2.14 x 10 36
-24 19 16 Plate D-8907-2 I.D.
+10*F 1.90 x 10 129 139 5
19 1/4T
+10*F 1.15 x 10 100 110 19 3/4T
+10*F 2.86 x 10 50 60 I
19 HAZ Material I.D.
+10*F 1.90 x 10 160 170 19 1/4T
+10*F 1.15 x 10 124 134 18 3/4T
+10*F 2.86 x 10 62 72 19 Weld Metal I.D.
-60 F 1.90 x 10 106 46 19 1/4T
-60 F 1.15 x 10 82 22 18 3/4T
-60*F 2.86 x 10 41
_13 19 I
32 Plate D-8907-2 I.D.
+10*F 3.8 x 10 180 190 19 1/4T
+10*F 2.3 x 10 140 150 18 3/4T
+10*F 5.7 x 10 70 80 19 HAZ Material I.D.
+10*F 3.8 x 10 230 240 19 1/4T
+10*F 2.3 x 10 180 190 18 3/4T
+10*F 5.7 x 10 90 100 I
19 Weld Material I.D.
-60 F 3.8 x 10 150 90 19 1/4T
-60*F 2.3 x 10 120 60 18 3/4T
-60*F 5.7 x 10 60 0
(a) Reference 1.
2 (b) Neutrons /cm, E > 1 MeV.
I I
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M M
M M
M M
M M
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M M
M 60 l.
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fleutron Fluence, E di fleV FIGURE 11.
EFFECT OF FAST flEUTRON FLUEllCE Oil C UPPER V
SilELF EllERGY, CALVERT CLIFFS UllIT 2 1
E Referring to the trend curves constructed in Figure 11 for the Calvert Cliffs Unit 2 vessel beltline materials, the projected C shelf y
energies of these materials are as follows:
o Plate D-8907-2 (Unirradiated Cv Shelf = 95 ft-lb*)
32 EFPY at I.O. -- 67 ft-lb (30% reduction) 32 EFPY at 1/4T -- 70 ft-lb (26% reduction) 32 EFPY at 3/4T -- 77 ft-lb (19% reduction) o Weld Metal (Unirradiated Cv Shelf = 137 ft-lb) 32 EFPY at I.D. -- 93 ft-lb (32% reduction) 32 EFPY at 1/4T -- 98 ft-lb (29% reduction) 32 EFPY at 3/4T -- 109 ft-lb (20% reduction) o HAZ Material (Unirradiated Cv Shelf = 83 ft-lb*)
32 EFPY at I.D. -- 61 ft-lb (27% reduction) 32 EFPY at 1/4T -- 63 ft-lb (24% reduction)
I i
32 EFPY at 3/4T -- 69 ft-lb (17% reduction)
This conservative projection indicates that the Calvert Cliffs Unit No. 2 vessel beltline weld metal will remain above the 50 ft-lb limit called out in Appendix G to 10CFR50 [1] to at least the end of the 32-EFPY design life.
I 5
I E
- Longitudinal values decreased by 35% to estimate transverse values [16].
I 43
VI. HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION OF CALVERT CLIFFS UNIT NO. 2 l
Calvert Cliffs Unit No. 2 is a 2700 Mwt pressurized water reactor operated by Baltimore Gas & Electric Company.
The unit has been provided with a reactor vessel material surveillance program as required by 10CFR50, Appendix H.
1 The first surveillance capsule (263* location) was removed during the 1982 refuelling outage.
This capsule was tested as described in earlier sections of this report.
In summary, these test results indicate that:
(1)
The maximum RTNDT after 12 effective full power years (EFPY) of operation was predicted to be 118'F at the 1/4T and 64*F at the 3/4T vessel wall locations, as controlled by the core beltline HAZ material.
(2)
The maximum RTNDT after 16 EFPY of operation was predicted to be 134*F at the 1/4T and 72*F at the 3/4T vessel wall locations, as con-trolled by the core beltline HAZ material.
The Unit No. 2 heatup and cooldown limit curves for 12 and 16 EFPY have been computed on the bases of the above using the generic procedure described in Appendix B.
The following pressure vessel constants were em-l plcyed as input data in the Calvert Cliffs Unit No. 2 analysis:
Vessel Inner Radius, rj
= 86.97 in., including cladding Vessel Outer Radius, r
= 95.59 in.
o Operating Pressure, P
= 2235 psig o
Initial Temperature, T
= 50*F o
Final Temperature, T
= 550*F f
Effective Coolant Flow Rate, Q = 128.8 x 106 lb /hr m
Effective Flow Area, A
= 39.83 ft2 Effective Hydraulic Diameter, D = 22.44 in.
I I
44
E Heatup curves were computed for a heatup rate of 60*F/hr.
Since lower r'ates tend to raise the curve in the central region, these curves I
apply to all heating rates up to 60*F/hr.
Cooldown curves were computed for cooldown rates of 0 F/hr (steady state), 20 F/hr, 40*F/hr, 60*F/hr, and 100*F/hr.
The 20*F/hr curve would apply to rates up to 20*F/hr; the 40*F, 60*F/hr, and 100 F/hr curves would apply to rates up to 40*F/hr, 60*F/hr, and 100*F/hr respectively.
I The unit No. 2 heatup and cooldown curves for up to 12 EFPY are given in Figures 12 and 13; those for 16 EFPY are given in Figures 14 and 15.
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6Lsd 'aanssaJd PG U M puI 49
I VII. REFERENCES 1.
Titic' 10, Code of Federal Regulations, Part 50, " Licensing of Pro-duction and Utilization Facilities," 1983 Revision.
I 2.
ASME Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components," 1980 Edition through the Summer 1982 Addenda.
3.
ASTM E 208-81, " Standard Method for Conducting Drop-Weight Test to I
Determine Ni-Ductility Transiticn Temperature of Ferritic Steels,"
1982 Annual Book of ASTM Standards.
t Steele, L.
E., and Serpan, C.
Z., Jr., " Analysis of Reactor Vessel Radiation Effects Surveillance Programs," ASTM STP 481, December 1970.
5.
Steele, L.
E.,
" Neutron Irradiation Emorittlement of Reactor Pres-sure Vessel Steels," Internat anal Atomic Energy Agency, Technical Reports Series No. lti3,1975.
6.
ASME Boiler and Pressure Vessel Code,Section XI, " Rules for In-service Inspection of Nuclear Power Plant Components," 1974 Edition.
7.
Perrin, J.
F.,, Wullaert, R.
A.,
Odette, G. R., and Lombrazo, P.
M.,
" Physically Based Regression Correlations of Embrittlement I
Data from Reactor Pressure Vessel Surveillance Programs," EPRI NP-3319, January 1984.
8.
Regulatory Guide 1.99, Revision 1, Office of Standards Development, I
U.S. Nuclear Regulatory Commission, April 1977.
~
9.
ASTM E 185-79, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," 1981 Annual Book of ASTM Standards.
10.
" Summary Report on Manufacture of Test Specimens and Assembly of I
Capsules for Irradiation Surveillance of Calvert Cliffs - Unit 2 Reactor Vessel Materials." CENPD 49, August 15, 1972.
11.
ASTM E 693. " Practice for CFaracteriz1ng Neutron Exposures in Fer-I ritic Steeis in Terms of Displacements Per Atom (DPA),1984 Annual Book of ASTM Standards.
l 12.
Perrin, J.
S.,
et al., Calvert Cliffs Unit No. 2 Nuclear Plant Reactor Pressure Vessel Surveillance Program:
Capsule 263," Final Report,_ Battelle CoNatus Laboratories, December 15, 1980.
13.
ASTM E 18, " Standard Test Methods for Rockwell Hardness and Rockwell Superficial Hardness of Metallic Materials," 1982 Annual Book of ASTM Standards.
I I
g 50
I 14.
Byrne, S.
T., Biemiller, E. C. and Rag 1, A., " Testing and Evaluation of Calvert Cliffs, Units 1 and 2 Reactor Vessel Materials Irradi-ation Surveillance Program Baseline Samples," TR-ESS-001, Combustion Engineering, Inc., Jcnuary 31, 1975.
15.
ASTM E 322, " Standard Method for Spectrochemical Analysis of Low I
Alloy Steels and Cast Irons Using an X-ray Fluorescence Spectro-meter," 1982 Annual book of ASTM Standards.
16.
US NRC Standard Review Plan, NUREG-0800, Section 5.3.2, Pressure-I Temperature Limits, Revision 1, July 1981.
I I
I I
I I
I I
I I
51
I I
I I
I I
APPENDIX A SOUTifdEST RESEARCH INSTITUTE NUCLEAR PROJECTS OPERATING PROCEDURES I
I I
I I
I
.4 Cl-23-lM -1 SOUTHWEST RESEARCH INSTITUTE 7bg
-y NUCLEAR PROJECTS xover. der 19s2 i /-e OPERATING PROCEDURE
~
Page 1 of 8 I
Title I
DETER'1I5ATION OF SPECIFIC ACTIVITY AND A';ALYSIS OF NEUTRON RADIATION DETECTOR SPECIMENS EFFECTIVITY AND APPROVAL 12/0/02 Revision 1
of nis crececure cecame effective on Tnis crocecure consists of the cages and changes listed below.
Page No Change Date Effect:ve I
I
' I I
l I I
l I
I I
cr A c-.,,
'I c
T_, e e.,e.
c.,e
.,,e.ey
<k 6. haa.el5v442.
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~
SOUTHWEST RESEARCH INSTITUTE I
Ntvember 1982 NUCLEAR PROJECTS OPERATING PROCEDURE Page 2 of 8
=
DETERMINATION OF SPECIFIC ACTIVITY AND ANALYSIS OF NEUTRON RADIATION DETECTOR SPECIMENS XI-MS-101-1 1.0 PURPOSE The purpose of this procedure is to describe the methods by which I
radiation material surveillance program neutron radiation detector speci-mens (also referred to as neutron flux monitors) are tested to determine their specific activity end analyzed to compute the neutron fluence to which they were exposed.
2.0 SCOPE AND APPLICATION (1)
This procedure describes the techniques and equipment to be utilized, and the responsibilities of the operator.
(2)
The requirements of this procedure shall be satisfied when measuring the specific activity of a gan:ma-emitting neutron flux monitor removed from a nuclear power reactor material surveillance program capsule.
2.1 Applicable Documents The following documents form a part of this procedure, as (1) 10CFR50, Appendix G, " Fracture Toughness Requirements."
(2) 10CFR20, " Standards for Protection Against Radiation."
(3)
SwRI Nuclear Quality Assurance Program Manual.
(4)
SwRI Radiological Health and Safety Manual.
I 3.0 RESPONSIBILITY (1)
The Director of the Department of Materials Sciences, Engineering and Materials Sciences Division, shall be responsible for the application of this procedure.
I I
gs_
E XI-MS-101-1 SOUTHWEST RESEARCH INSTITUTE I
November 1982 k
NUCLEAR PROJECTS OPER ATING PROCEDURE Page 3 of 8 ne I
(2)
The Manager of the Nuclear Materials Program, the Department of Materials Sciences, shall be responsible for the implementation and control of this procedure in accordance with the SwRI Nuclear Quality Assurance Program Manual in effect on the effectivity date I
of this procedure.
(3)
The operator shall be responsible for implementing the requirements of this procedure.
(4)
The Manager of the Nuclear Materials Program, the Department of I
Materials Sciences, shall be responsible f or the storage of records generated in accordance with this procedure.
4.0 PROCEDURE REOUIREMENTS The requirements listed below shall apply when testing and analyzing neutron radiation detector specimens in accordance with this procedure:
(1)
Personnel (operators) perf orming these tests shall be qualified by training and experience to the satisf action of the Manager of the I
Nuclear Materials Program of the Department of Materials Sciences.
Documentation of the approval of operators shall be on file in the Depart ment of Materials Sciences.
(2)
This procedure shall be used for the testing and analysis of the following detector specimens :
I
<a (a)
~ ~ Co ( n,Y ) y" Co (b)
" Ni (n,p)
Co (c)
Fe (n,p)
Mn (d)
~'Cu (n,a) ;-"Co However, these procedures are generally applicable to testing any gamma-emitting detector specimens. Therefore, other detector I
specimens may be tested and analyzed with this procedure with the approval of the Manager of the Nuclear Materials Program of the Department of Materials Sciences.
(3)
Personnel performing these tests chall observe the radiation safety procedures required by the SwRI Radiation Safety Officer or his designated alternate.
I SwA! Form QA 3 2
I XI-MS-101-1 I
SOUTHWEST RESEARCH INSTITUTE November 1982 NUCLEAR PROJECTS OPERATING PROCEDURE Page 4 of 8 I
e 5.0 EOUIPMENT The items of equipment needed to follow this procedure are listed below:
(1)
NaI(Tl) or Ge(Li) detector (2)
Multi-channel pulse height analyzer (3)
" Co,
Mn, and Cs gamma standards, with disintegration l#
rates traceable to the U.S. Bureau of Standards (4)
Laboratory balance, with sensitivity of 0.1 mg.
6.0 PROCEDURE 6.1 Determination of Number of Tarcet Atoms I
The nu=ber of target atoms per mg in each detector specimen is calculated as follows :
I t;0 "
where:
0 = number of target atoms per mg N = Avogadro's number (6.02 x 100 ) nuclei per gm atom A = atomic weight of detector element, gms c = weight fraction of detector isotope in detector specimen 6.2 Countine Apparatus I
(1)
The detector should be set up in a controlled access laboratory with the associated power supplies and instrumentation needed to measure and record gamma ray spectra.
(2)
Each gamma source to be measured shall be supported in a i
position along the projected long axis of the detector I
at any convenient distance which meets the following requirements:
SwRI Form QA 3 2
I SOUTHWEST RESEARCH INSTITUTE xt-MS-101-1 l
November 1982 k
NUCLEAR PROJECTS OPER ATING PROCEDURE I
n i y y Page 5 of 8 I
(a)
The multichannel analyzer dead time meter reads ten percent or less.
(b)
There are no extraneous materials in the path between the I
source and the counter which cause significant scattering or absorption of the gamma rays of interest.
6.3 Determination of Specific Activity of a Detector Specimen (1)
Detector specimens contained in material surveillance I
capsules are most commonly in the form of small diameter wires. Other common forms include powders (such as UO2 fission monitor specimens) or slices of metal removed from tested mechanical property specimens. A portion or all of each specimen may be counted, depending on the total activity of each specimen.
The specimen should be cut or shaped to lie in a plane area having no dimension greater than 1/4 in.
and having a thickness no greater than 1/16 in.
(2)
Make a preliminary scan of all specimens representing each I
group of unknown detector materials at the source-to-detector distance (s) previously established.
(a)
If the instrumentation shows that the dead-time exceeds 10% increase the source-to-detector distance.
The determination of standard specimen source efficiencies (Paragraph (3) below) and specific activities of I
unknown specimens (Paragraph (4) below] should be accomplished at the same source-to-detector distances established for each group of detector materials.
(b)
If the instrumentation shows that the dead-time is less than 10% but the counts in the peak channel are less than 0.1 of the counts in the peak channel obtained I
with the appropriate standard source, the source-to-detectcr distance may be decreased, but in no case closer than 10 cm f rom the detector.
If the appro-I priate standard source cannot be counted at this reduced source-to-detector distance because of the dead-time 11 aitat ions, intrinsic efficiency factors are computed for the ntandard source (T(E)s] and the I
unknown source [T(E)u] geometries per Appeadix IV of Adams and Dams, [Ref. (3)], and a geometry correction factor of T(E)s,T(E)u is determined.
SwRI Form QA 3 2
I
"~""~
~
I SOUTHWEST RESEARCH INSTITUTE November 19F.2 NUCLEAR PROJECTS OPERATING PROCEDURE Page 6 of 8 d
I (3)
Determine the experimental ef ficiency for the appropriate standard source ence each day for each group of unknown detector materials being tested that day.
A convenient counting period is 2000 sec, but other counting periods may be employed. At least 20,000 counts should be accumulated in I
for Co and Cu detector specimens is 60Co, for Fe specimens is the photo peak of interest. The appropriate standard source 54Mn, and for the fission-monitor specimens is 137Cs. When counting other detector specimens, determine the experimental efficiencies for all three standard sources and calculate the required efficiencies using a total efficiency / log gamma ray energy linear interpolation or extrapolation.
Sum the counts in the appropriate photo peak and compute the experimental efficiency for a given photo peak as in the following example:
Eff(E)t = Total counts under peak of Energy E less "backcround" A (T) t where:
" background" is the area beneath the photo peak as I
determined by a straight line or curvlinear inter-polation f rom the background on either side of the photo peak.
and:
)
i counting time, see I
~
=
A disintegration rate at time of counting, sec-1
=
g For a atanciard source, A = Ao(exp-At) t where:
Ao disintegration rate of standard at time of cali-
=
bration, sec-1 elapsed time between calibration date and count-t
=
ing date, days 0.69315/T /2, day-1 A
=
l T1/2 half-life of isotope, days
=
SwRl Form OA 3-2
E
~'
SOUTHWEST RESEARCH INSTITUTE November 1982 d
NUCLEAR PROJECTS OPERATING PROCEDURE Page 7 of 8 bb~
I (4)
Determine the specific activity for the unknown detector specimen.
Sum the counts in the appropriate photo peak and compute the specific activity of the unknown I
detector as in the following example:
A(TOR) = Total counts under peak of energy E less " background" x T(E)s Eff(E)t (T) (w) (exp-Ati)
T(E)u where:
A(TOR) specific activity of selected isotope in unknown
=
I specimen, disintegrations per second (dps) per mg of target element at time of removal (TOR) weight of target element, og w
=
t1 elapsed time between TOR and counting date,
=
days 6.4 Determination of the Saturated Activity of a Detector Specimen Calculate the saturated activity at TOR after n operating periods as follows:
I A(TOR) m=n P (1-exp-AT ) (exp-At )
I s
m=n where:
n' I
A saturated activity, dps/cg target element
=
s P
fraction of full power generated in period m
=
m Tm operating time at selected power level for the
=
mth operating period, days m
elapsed time from end of the m h operating t
t
=
period to TOR, days 6.5 Calculation of the Neutron Flux Density Calculate the neutron flux density as follows:
I As NC O
I y
s__
I XI-MS-101-1 SOUTHWEST RESEARCH INSTITUTE November 1982 NUCLEAR PROJECTS OPERATING PROCEDURE Page 8 of 8 I
, [F
.4 where:
2 p
= energy-dependent neutron flux density, n/cm sec 2
= spectrum averaged activation cross-section, em Note: A IN is also known as the reaction rate.
s o i.0 RECORDS (1)
Printouts of the pulse-height analyzer data shall be filed in the project folders in the files of the Department of Materials Sciences.
(2)
These documents shall be stored by the Manager of the Nuclear Materials Program of the Department of Materials Sciences f or the period specified by the contractual agreement with the customer. These records shall be indexed, filed, and main-tained in facilities that provide suitable environment to minimize deterioration or damage and to prevent loss.
8.0 REFERENCES
I The following references were utilized as bases for the procedure described herein:
(1)
ASTM Method E181, " Analysis of Radioisotopes."
(2)
ASTM Method E261, " Measuring Neutron Flux, Fluence and Spectra by Radioactivation Techniques."
I (3)
Adams and Dams, Applied Gamma Ray Soectrometry, Perga mon Press, 2nd Edition, 1970.
I I
I s*m Form OA 3 2
l
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S ag SOUTHWEST RESEARCH INSTITUTE xt-MS-103-1 NUCLEAR PROJECTS x,ye 3,7 1932 OPERATlNG PROCEDURE 4
~
Page 1 of 8 Titte CONDUCTING TENSIOS TESTS ON METALLIC SPECIMENS EFFECTIVITY AND APPROVAL 12/8/82 This orocecure censms of tne pages and Revision 1
of tnis crececure became effective on changes hsteo beiow.
Page No Change Date Effective I
I I
I I
I I
SA Accrovais Written By Date Tecnnical Review Date f,
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SOUTHWEST RESEARCH INSTITUTE November 1982 NUCLEAR PROJECTS OPERATING PROCEDURE Page 2 of 8 CONDUCTING TENSION TESTS ON METALLIC SPECIMENS XI-MS-103-0 I
1.0 PURPOSE The purpose of this procedure is to describe the methods by which tension tests are performed on metallic specimens at SwRI.
2.0 SCOPE AND APPLICATION (1)
This procedure describes the techniques and the equipment to be I
utilized, and the responsibilities of the operator.
(2)
The requirements of this procedure shall be satisfied when I
conducting tension tests on specimens utilized in nuclear power reactor material surveillance programs.
(3)
This procedure may be applied for nonnuclear applications when I
measuring the tensile properties of structural materials for conformance to applicable codes.
2.1 Applicable Documents The following documents form a part of this procedure, as (1) 10CFR50, Appendix G, " Fracture Toughness Requirements."
(2)
ASME Boiler and Pressure Vessel Code,Section III, 1980 Edition, and addenda through the Su=mer,1982 Addenda.
(3)
SwRI Nuclear Quality Assurance Program Manual.
3.0 RESPONSIBILITY (1)
The Director of the Department of Materials Sciences, Engineering Sciences Division, shall be responsible for the application of this I
procedure.
I SwRI Form OA 3 2
I
- '~**~' '~*
SOUTHWEST RES EARCH INSTITUTE November 1982 NUCLEAR PROJECTS OPERATING PROCEDURE I
Page 3 of 8
_gv (2)
The Manager of the Mechanical Properties Section, Department of Materials Sciences, shall be responsible for the implementation and control of this procedure in accordance with the SwRI Nuclear Quality Assurance Program Manual in effect on the date this I
procedure is approved.
(3)
The operator shall be responsible for implementing the requirements of this procedure.
(4)
The Manager of the Mechanical Properties Section, Department of I
Materials Sciences, shall be responsible for the storage of records generated in accordance with this procedure.
4.0 PROCEDURE REOUIREMENTS The requirecents listed below shall apply when perf orming tension tests in accordance with this procedure:
(1)
Persennel (operators) perf orming tension tests shall be qualified by training and experience to the satisfaction of the Manager of I
the Mechanical Properties Section of the Department of Materials Sciences.
Documentation of the approved operators shall be on file in the Department of Materials Sciences.
(2)
The test temperatures shall be approved by the Manager of the Mechanical Properties Section, Department of Materials Sciences.
(3)
Personnel performing these tests shall observe the radiation safety procedures required by the SwRI Radiation Safety Officer or his designated alternate.
5.0 EOUIPMENT 5.1 Testing Machine The tension testing machine shall be of a type meeting the requirements of ASIM Method E4, " Standard Methods of Verification of Testing I
Machines."
5.2 Extensometer I
(1)
For tests at room temperature, a Class B-2 extensometer shall be attached to the gage section of the tension specimen I
either directly or indirectly with the use of averaging-type extension arms.
SwRI Form OA 3 2
XI-MS-103-1 SOUTHWEST RESEARCH INSTITUTE November 1982 k
NUCLEAR PROJECTS OPERATING PROCEDURE I
)g g Page *, o f 8 (2)
For tests at temperatures prohibiting the use of direct attachment of the extensometer to the gage section of the test specimen, averaging-type extension arms shall be used.
5.3 Temperature Measuring Equipment Each thermocouple shall be f abricated f rom pairs of thermocouple I
materials appropriate for the test specimen temperature to be measured. The accuracy of the thermoelectric millivoltage output of each pair of thermocouple materials shall be within the limits given in ASTM E 320.
5.4 Auxiliary Ecuipment (1)
An X-Y recorder (or equivalent) shall be used to record the I
load-strain history for each tension test.
(2)
A commercially-available 0.001-in. micrometer shall be used to measure the initial diameter of the test specimen gage section.
l (3)
A " point-type" 0.001-in. micrometer or a 0.001-in. dial caliper shall be used to measure the diameter of the test specimen at the location of fracture after the test is completed.
6.0 PROCEDURE The following procedure is based on the references listed in Section 8.0 of this procedure.
6.1 Checking and Calibration of Tension Test Equipment (1)
Each day that the tension testing machine is used, the accuracy of the strain gage load cell shall be checked either I
by using standard weights as described in ASE Method E4 or by employing a shunt calibration technique which utilizes the load equivalent of a standard resistor.
(2)
If the tension testing machine load cell has been loaded beyond its nominal capacity, the accuracy of the tension I
testing machine load cell shall be checked using an elastic proving ring traceable to the U.S. Bureau of Standards. This calibration shall be perf ormed in accordance with ASTM Method E4.
I I
SwRI Form QA 3 2
I Xf-MS-103-1 SOUTHWEST RESEARCH INSTITUTE November 1982 NS NUCLEAR PROJECTS OPERATING PROCEDURE re Page 5 of 8 (3)
The accuracy of the temperature meas tring device used to I
measure specimen temperatures shal' be checked at six-month intervals with Standards traceable to the U.S. Bureau of Standards.
The instrument shall be tagged to indicate the I
date of the last calibration and the date of the next calibration.
I (4)
The accuracy of the micrometers used to measure specimen gage section diameter shall have been checked with gage blocks no more than two months prior to their use.
(5)
The extensometer-autographic recorder system shall be checked with a 0.0001-in. micrometer at the beginning of each day that the equipment is to be used for conducting tension tests in accordance with this procedure.
~
6.2 Tension Testing Procedure (1)
Tension test specimens to be tested at room temperature
- shall be brought to the testing laboratory at least one hour prior to conducting the tests.
(2)
A 1-in., 1.4-in., or 2-in. gage length (as appropriate to the size of the tension test specimen) shall be lightly marked on I
the gage section.
The diameter of the gage section shall be measured with 0.001-in. micrometers.
(3)
When tension specimens are to be tested at other than room temperature, the specimens shal.1 be heated in air with a laboratory furnace or cooled in a cryogenic chamber.
(4)
Tension test specimens to be tested at other than room temperature shall be instrumented by attaching thermocouples to the reduced (gage) section.
For tension test specimens I
with 1-in. or less gage section, two thermocouples shall be used:
one wired to the top of the gage section and the other wired to the bottom of the gage section.
For tension test I
specimens with a gage section of 1.4-in. or 2-in. le ng th,
three thermocouples shall be employed (at the top, center, and bottom).
I
- Room temperature is defined as 80F t10F.
I Swm Form QA 3 2
XI ""S-103-1 SOUTHWEST RESEARCH INSTITUTE November 1982 lh.
/R d4 NUCLEAR PROJECTS OPERATING PROCEDURE Page 6 of 8 I
4 l /-,.,/
(5)
The specimen is installed in the upper or lower grip of the I
tension testing machine.
(6)
The extensometer is attached to the specimen gage section.
(7)
For tests conducted at other than room temperature, a furnace or cryogenic chamber which has a length of at least three times the length of the gage section should be used.
(8)
Room temperature tests may be run as outlined in Paragraph (10) as soon as the requirement of Paragraph (1) has been I
met.
Room temperature shall be recorded.
(9)
For tests at other than room temperature, the test specimen shall be brought to the nominal test temperature *1.7C (*3F) as indicated by the top thermocouple attached to the specimen gage section. The maximum variation along the specimen shall I
not exceed i3C ( 5F). The specimen shall be held within the above specified limits for at least 15 minutes prior to conducting the test.
(10) The sequence of operations for conducting a tension test should be as follows:
(a)
Attach the remaining tension test machine grip to the test specimen.
I (b)
Select tne appropriate strain rate setting on the tension test machine.
(c)
Start the machine and observe the autographic I
load-strain record to assure that load and strain indicators are functioning properly.
(It may be of help to draw a light line on the recorder paper I
representing the expected slope of the clastic portion of the curve.)
I (d)
After the specimen has exceeded the 0.27. offset yield load, the extensometer may be removed from the specimen (or from the extension arms in the case of elevated or I
cryogenic temperature tests) to prevent damage to the extensometer when the specimen f ractures.
Continue to record load for calculation of ultimate tensile strength.
(e)
After the specimen has fractured, stop the test machine and remove the broken pieces.
I SwQlForm OA 3 2
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- '-"S-SOUTHWEST RESEARCH INSTITUTE Uovember 1962 rq M Mr,
NUCLEAR PROJECTS OPER ATING PROCEDURE Page 7 of 8 I
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(11) The tension specimen number and thermocouple readings shall be recorded on a data sheet. The 0.2% of f set (yield) and maximum (ultimate) loads from the autographic recording shall be recorded and entered on a data sheet. After fitting the I
two halves of the specimen together, the dimensions of the minimum cross-section and the final gage length shall be measured and recorded on the data sheet. The autographic load-strain record shall be identified by project number and specimen number.
I (12) The ultimate tensile strength (UTS), 0.2% yield strength (0.2% YS), percent elongation (% El), and percent reduction in area (% RA) are computed as follows :
(a)
UTS =
Maxieum Load (lbs)
Initial Cross-Section Area (in.4)
(b) 0.2% YS =
0.2% offset Load (lbs)
Initial Cross-Section Area (in.')
1 (c)
% El = Increase in Cage Length x 100 Initial Gage length I
(d)
% RA = (Initial Area) - (Final Area) x 100 Initial Area The person performing these calculations shall sign (or I
initial) the data sheet.
(13)
If unif orm elongation, f racture strength, and f racture stress are required to be determined, record the load vs. cross-head movement to specimen fracture.
Calculate these values as described in ASIM E184.
(14) The tested tension specimens shall be placed in a marked container (or an indexad storage receptacle in the case of irradiated specimens) and returned to storage.
(15) The calculations described in Paragraphs (12) and (13) shall be checked by a person other than the one perf orming the I
original calculations.
The person checking the calculations shall sign (or initial) the data sheet.
I SwAl Form QA 3 2
I
'XI ~us-103-1 SOUTHWEST RESEARCH INSTITUTE November 1982 g
NUCLE AR PROJECTS OPER ATING PROCEDURE Page 8 of 8 7.0 RECORDS 7.1 Test Data Records E
(1)
Each data sheet shall be filed in the project f olde rs in the files of the Department of Materials Sciences.
(2)
These documents shall be stored by the Manager of the Mechanical Properties Section, Department of Materials I
Sciences for the period specified by the contractual agreement with the customer.
These records shall be indexed, filed, and maintained in facilities that provide suitable environment to minimize deterioration or damage and to prevent loss.
7.2 Calibration Records The test equipment calibration records shall be kept on file in the Department of Materials Sciences.
8.0 REFERENCES
The following references were utilized as bases for the tension testing I
procedure described herein:
(1)
ASTM Method E4, " Verification of Testing Machines."
I (2)
ASTM Method E8, " Tension Testing of Metallic Materials."
I (3)
ASTM Practice E184, " Effects of High-Energy Neutron Radiation on the Mechanical Properties of Metallic Materials."
(4)
ASTM Method E21, " Elevated Temperature Tension Tests of Metallic I
Materials."
(5)
ASTM Method E83, " Verification and Classification of I
Extensometers."
(6)
ASTM Method E185, " Surveillance Tests for Nuclear Reactor Vessels."
(7)
ASTM Standard E320, " Temperature-Electromotive Force (EMF) Tables for Thermocouples."
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<ie, OPERATING PROCEDURE Page 1 of 7
.e CHARPY I!: PACT TESTS ON !!ETALLIC SPECI:! ENS E.:FECTIVITY AND APPRCVAL Revision I of this procecure became effecove on 12/8/82 This procecure c=nsists of :ne cages anc cnanges listed below.
Page No Change Date Effect:ve i
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SOUTHWEST RESEARCH INSTITUTE I
November 19 82 h g..m NUCLEAR PROJECTS OPERATING PROCEDURE Pa ge 2 of 7 I
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CHARPY IMPACT TESTS ON METALLIC SPECIMENS XI-MS-104-1 I
1.0 PURPOSE I
The purpose of this procedure is to describe the methods by which Charpy impact tests are performed on metallic specimens at SwRI.
2.0 SCOPE AND APPLICATION (1)
This procedure describes the techniques and the equipment to be utilized, and the responsibilities of the operator.
(2)
The requirements of this procedure shall be satisfied when con-ducting Charpy impact tests on specimens utilized in nuclear power reactor material surveillance programs.
(3)
This procedure may be applied when measuring the Charpy V-notch I
impact properties of structural materials for conformance to applicable codes.
2.1 Applicable Documents The following documents form a part of this procedure, as (1) 10CFR50, Appendix G, " Fracture Toughness Requirements."
I (2)
ASME Boiler and Pressure Vessel Code,Section III, 1980 Edition, and addenda through the Summer,1982 Addenda.
(3)
SwRI Nuclear Quality Assurance Program Manual.
3.0 RESPONSIBILITY I
(1)
The Director of the Department of Materials Sciences, Engineer-ing Sciences Division, shall be responsible for the application of this procedure.
(2)
The Manager of the Nuclear Materials Program, Department of Materials Sciences, shall be responsible for the implementation I
and control of this procedure in accordance with the SwRI Nuclear Quality Assurance Program Manual in effect on the date this procedure is approved.
I
$wAIForm OA J 2
I X1-MS-10J.-1 SOUTHWEST RESEARCH INSTITUTE November 1982 Mh Page 3 of 7 l
QS NUCLEAR PROJECTS OPERATING PROCEDURE 5
(3)
The operator shall be responsible for implementing the require-ments of this procedure.
(4)
The Manager of the Mechanical Properties Section, Department of I
Materials Sciences, shall be responsible for the storage of records generated in accordance with this procedure.
4.0 PROCEDURE REOUIREMENTS The requirements listed below shall apply when perf orming Charpy impact tests in accordance with this procedure:
(1)
Personnel (operators) perf orming these tests shall be qualified by training and experience to the satisfaction of the Manager of the Nuclear Materials Program of the Department of Materials Sciences.
Documentation of the approval of operators is on file in the Department of Materials Sciences.
(2)
The test temperatures shall be approved by the Manager of the Nuclear Materials Program, Department of Materials Sciences.
(3)
Personnel perf orming these tests shall observe the radiation safety procedures required by the SwRI Radiation Safety Officer l
or his designated alternate.
l 5.0 EotIPMENT 1
5.1 Testine Machine The Charpy impact machine shall be of a type meeting the requirements of ASD! Method E23, " Standard Methods for Notched Bar Impact I
Testing of Metallic Materials."
5.2 Tecmerature Measuring Equipment I
l Each thermocouple shall be fabricated from pairs of thermo-couple materials appropriate for the test specimen temperature to be j
I measured. The accuracy of the thermoelectric millivoltage output of each pair of thermocouple materials shall be within the limits given in r
l ASn! E320.
5.3 Micrometer A commercially-available 0.001-in. micrometer shall be used to measure the initial and f' nal specimen widths for determining lateral i
l expansion.
l I S
I
- RI Form OA 3 2 i
l
I SOUTHWEST RESEARCH INSTITUTE
- *S-November 1982 i
NUCLEAR PROJECTS OPERATING PROCEDURE Page 4 cf 7 6.0 PROCEDURE 6.1 Checking and Calibration of Charpv Test Ecuipment (1)
Each day that the Charpy machine is used, the free swing of the pendulum, with the appropriate hammer installed, shall be checked prior to conducting any tests. With I
the indicator set at the factory-recommended position, a free swing of the pendulum shall indicate zero energy.
Should a zero energy reading not be obtained, the scale I
compensator shall be adjusted until three consecutive swings read zero energy.
I (2)
The accuracy of the machine shall be checked at inter-vals not to exceed one year with a set of specimens obtained from the Army Materials and Mechanics Research Center (MDRC), Watertown, Massachusetts.
Should the I
machine not meet the requirements of the set of calibration specimens, the cause shall be investigated and remedied, and a new sct of specimens shall be I
tested. During any period in which the machine has been shown to be out of calibration, it shall not be used to test nuclear reactor material surveillance specimens.
(3)
The accuracy of the temperature measuring device used to measure specimen conditioning bath temperatures shall be checked at 6-month intervals with standards traceable to the U.S. Bureau of Standards.
(4)
The accuracy of the nl:rometers used to measure the I
specimen dimensions shall have been checked with gage blocks traceable to the U.S. Bureau of Standards no more than two months prior to their use.
6.2 General Charpv Imract Testine Procedure (1)
Charpy specimens to be teste3 at room temperature
- shall I
be brought to the laboratory n least one hour prior to conducting the tes ts.
I (2)
When Charpy specimens are to be tested at other than room temperature, the specimen conditioning bath shall be filled with an appropriate liquid. Acceptable I
- Room temperature is defined as 80F t 10F.
5*m Form QA 3 2
I XI-MS-104-1 SOUTHWEST RESEARCH INSTITUTE November 1982 NUCLEAR PROJECTS OPERATING PROCEDURE Page 5 of 7
)
I liquids include, but are not necessarily limited to, the following:
(a)
Methanol (-100F to room temperature).
(b) k'ater [9C (32F) to 100C (212F)].
I (c)
High Temperature Oil (room temperature to 300C (572F)).
(3)
Specimens to be tested at a given temperature shall be I
removed f rom storage and placed in the specimen condi-tioning bath.
(4)
Room temperature Charpy tests may be run as outlined in Paragraph (7) as soon as the requirement of Para-graph (1) has been met.
Room temperature should be recorded.
(5)
For Charpy tests below room temperature, the condition-ing bath should be cooled with an appropriate medium I
such as dry ice. The bath shall be stirred conti-nuously. The bath temperature shall be checked with a calibrated temperature measuring device. The specimens I
shall be held in the bath at the desired test te mpe r-ature within +0 and -1.5C (+0 and -3F) f or at least five minutes before testing.
(6)
For Charpy tests above room temperature, the condition-ing bath should be heated with the conditioning bath electric heating elements. The bath shall be stirred I
continuously. The bath temperature shall be checked with a calibrated teuperature measuring device. The specimens shall be held in the bath at the destred test I
temperature within +0 and +1.5C (+0 and +3F) for at least ten minutes before testing.
I (7)
The sequence of operations to conduct a Charpy test shall be as f ollows:
(a)
Cock the hammer.
(b)
Set the energy indicator at the factory-recommended set point.
(c)
Place a specimen on the machine anvil with a centering device and release the pendulum within I
five seconds. After at least one full swing of the pendulum, apply the brake.
I SwRIForm OA 3 2
I XI-MS-104-1 SOUTHWEST RESEARCH INSTITUTE November 1982 NUCLEAR PROJECTS OPER ATING PROCEDURE Page 6 of 7 w
(d)
Record the Charpy specimen number, test tempera-ture, absorbed energy, and date of test.
Deter-mine the fracture appearance and record on the da ta sheet. Measure the lateral expaasion and I
record on the data sheet.
(e)
Replace the tested Charpy specimen halves in a marked envelope or an indexed storage receptacle and return to storage.
6.3 Instrumented Charpy Testine Procedure ( Applicable When specified (1)
All requirements of Section 6.2, General Charpy Testing I
Procedure, are applicable to conducting instrumented Charpy tests.
(2)
The strain gaged cup shall be calibrated statically by applying a series of loads at the location of impact using a testing machine and following the applicable portions of AST". Method E4.
This calibration shall be I
subsequently accomplished by use of a shunt calibration technique each day that the instrumentation is used.
(3)
Instrumentation shall be set up to provide an oscillo-I graphic recording of both load-time and integrated energy-time during the period of specimen fracture.
I (4)
A permanent record of the load-time and integrated energy-time curves shall be made.
i 7.0 RECORDS 7.1 Test Data Records (1)
Ea :h data sheet shall be filed in the project folders in j
thi files of the Department of Materials Sciences.
I (2)
Trese documents shall be stored by the Manager of the Nuclear Materials Program, Department of Materials Sciences for the period specified by the contractual I
agreement with the customer. These records shall be indexed, filed, and maintained in facilities that provide suitable environment to minimize deterioration or dam.ga and to prevent loss.
I I
$wRI Form OA 3 2 I
I XI-MS-104-1 SOUTHWEST RESEARCH INSTITUTE November 1982 NUCLEAR PROJECTS OPER ATING PROCEDURE Page 7 of 7 g
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'w 7.2 Calibration Records The AMMRC Charpy machine calibration records shall be kept on file in the Department of Materials Sciences.
8.0 REFERENCES
I The following references were utilized as bases for the impact testing procedure described herein:
I (1)
ASTM Method A370, " Standard Methods and Definitions for
'techanical Testing of Steel Products."
(2)
ASTM Method E23, " Standard Methods for Notched Bar Impact Test-I ing of Metallie Materials."
(3)
ASTM Method E185, " Standard Recommended Practice for Surveil-lance Tests for Nuclear Reactor Vessels."
(4)
ASTM Method E4, " Standard Methods of Verification of Testing Machines."
(5)
ASTM Method 320, "re=perature-Electromotive Force (EMF) Tables for Thermocouples."
I I
I I
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I I
SwRIForm OA J 2
E NUSLEE PRECTS RLE SOUTHWEST RESEARCH INSTITUTE I
XIII->:S-103-1 NUCLEAR PROJECTS OPERATING PROCEDURE Nove= der 1982 Pace 1 of 5 OPCiING RADIATION SURVEILLANCE CAPSULES AND HANDLING AND STORING SPECIliENS EFFECTIVITY AND APPROVAL 12/8/82 Revision N of mis crocecure became effecitve on This procedure consists of :ne cages and cnanges listed below.
Page No Change Date Effective I
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NUCLEAR PROJECTS OPERATING PROCEDURE
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I OPENING RADIATION SURVEILLANCE CAPSULES AND HANDLING AhT) STORING SPECIMENS XIII-MS-103-1 I
1.
PURPOSE I
The purpose of this procedure is to describe the methods by which radiation material surveillance program capsules are handled, opened, and stored at SwRI.
2.
SCOPE AND APPLICATION (1)
This procedure describes the techniques and equipment to be utilized, and the responsibilities of the operator.
(2)
The requirements of this procedure shall apply to the handling, opening, and storage of nuclear reactor radiation material surveillance capsules received at SwRI for testing and evaluation.
2.1 Applicable Documents I
The following documents form a part of this procedure, as applicable:
(1) 10CFR20, " Standards f or Protection Against Radiation."
I (2)
SwRI Nuclear Quality Assurance Program Manual.
(3)
SwRI Radiological Health and Safety Manual.
3.
RESPONSIBILITY (1)
The Director of the Department of Materials Sciences, Engineering Sciences Division, shall be responsible for the application of this procedure.
I (2)
The Manager of the Nuclear Materials Program, the Department of Materials Sciences, shall be responsible for the implementation and I
control of this procedure in accordance with the SwRI Nuclear Quality Assurance Program Manual in ef fect on the date this procedure is approved.
(3)
The operator shall be responsible for implementing the requirements of this procedure.
5 SOUTHWEST RESEARCH INSTITUTE xt u-Ms-103-1 O[R % g '
y, November 1982
~
NUCLEAR PROJECTS OPER ATING PROCliDURE g
i [ [%y Page 3 of 5 g.
8 (4)
The Manager of the Nuclear Materials Program, the Department of Materials Sciences, shall be responsible for the storage of records generated in accordance with this procedure.
4.
EQi1IPMENT (1)
Milling machine.
(2)
Band saw.
(3)
Hot cell facilities.
5.
PROCEDURE 5.1 Preparati_on for Opening Radiation Surveillance Capsules (1)
Radiation surveillance capsules are of many designs, but they I
generally incorporate stainless steel pipe or shells made of stainless steel sheet. Radiation surveillance capsules made of stainless steel pipe may generally be opened with a band saw.
Radiation surveillance capsules made of stainless steel I
sheet may require a milling machine as well as a saw.
(2)
An engineering drawing of the capsule (s) or instructions for opening the'eapsule(s) shall be obtained from the customer.
(3)
A list of the types, quantities, and identifications of all I
spec'aens and monitors contained in the radiation surveillance capsule (s) shall be obtained from the customer.
(4) ludexed receptacles for the test specimens, a code sheet for I.
identifying the specimen locations in the indexed receptacles, and markeri containers for the thermal and flux monitors shall be prepared in advance of opening the capsules (s).
5.2 Opening of 4 dic. tion Surveillance Capsules I
(1)
All personnel shall obs.rve radiation safety requirements of the Radiation Laboratory, as well as the capsule opening procedure.
(2)
Perform the cutting operations on the capsule (s) per instructions issued by the Project Manager.
I A
I SwAI Form OA 3-2 v
I SOUTHWEST RESEARCH INSTITUTE xu r-nS_103-1 g
k November 1982
/RjeQ NUCLEAR PROJECTS OPERATING PROCEDURE I
lgu Page 4 of 5 7
(3)
An example of the steps required to open capsules of the design utilized by Westinghouse is as follows:
B (a)
Clamp capsule in saw and cut the bottom end plug off, locating the cut no more than 1/8 inch above the bottom end plug-to-capsule enclosure weld.
(b)
Clamp capsule in saw and cut the top end plug of f, locating the cut no more than 3/8 inch below the top end plug-to capsule enclosure weld. Note:
Step 2 may be I
carried out before Step 1.
(c)
Clamp the capsule in a milling fixture and remove the nuggets of the two longitudinal welds which join the capsule enclosure halves. Note: To minimize handling, milling of one longitudinal weld should be completed before starting on the second weld.
(d)
With the capsule clamped in a fixture, remove one capsule enclosure half to expose the contents.
5.3 Identification and Storage of Specimens i
The contents of the capsule (s) shall be carefully removed and stored in appropriate containers as outlined below. These steps should be carried out by two persons to provide added assurance that indentifications are properly made.
j (1)
Examine each test specimen, with the aid of a magnifier if necessary, to determine its identification number. Place each I
specimen in an indexed receptacle.
(2)
Record the specimen number on a code sheet which references a schematic drawing of the indexed receptacle.
(3)
Examine and identify all thermal and flux monitors and place i
8 in marked containers which identify the monitor and the location in the capsule from which it was re moved.
(4)
Check the specimen identification numbers against the list supplied by the customer. Report to the customer any inconsistencies prior to conducting any destructive tests.
I (5)
Store specimens and radioactive scrap in the high level 1
storage area.
(6)
Prepare all equipment for storage in hot cell per instructions I
of radiation laboratory supervisor.
SwRI Form QA 3-2 r
I SOUTHWEST RESEARCH INSTITUTE xill-Ms-10-1 November 1982 NUCLEAR PROJECTS OPER ATING PROCEDURE I
l_
G Page 5 of 5
.s 5.4 Disposal of Scrap Materials (1)
Scrap capsule containment material should be disposed of* as soon as convenient, but shall not be transferred until all contents of the capsule (s) have been accounted for and I
properly stored.
(2)
Broken specimens and thermal and flux monitors shall be held I
in storage in the hot cell for a period of at least six months af ter issuance of the Test Report unless the reactor owner releases them prior to that time or requests that they be retained for a longer period.
(3)
Broken specimens and thermal and flux monitors shall be disposed of* only at the direction of the Project Manager.
6.
RECORDS (1)
Specimen identification records shall be filed in project folders in the files of the Department of Materials Sciences.
I (2)
These documents shall be stored by the Manager of the Nuclear Materials Program, Department of Materials Sciences, for the pariod specified in the contractual agreement with the customer. These I
records shall be indexed, filed, and maintained in facilities that provide suitable environment to minimize deterioration or damage and to prevent loss.
I I
8 I
I I
... the.s1 mess Manu.1 for disposa1 of rae1oactive.aste.
I
I I
I I
I APPENDIX B HARDNESS TEST DATA, I
TENSILE TEST RECORDS AND i
TESTED SPECIMEN PHOTOGRAPHS I
I I
I I
I
I TABLE B-1 ROCKWELL "B" HARDNESS READINGS TAKEN ON CHARPY V-NOTCH SPECIMENS Specimen Specimen Material No.
Hardness (HRB)
Material No.
Hardness (HRB)
Plate llB 95.0 95.5 94.5 Weld 237 97.0 95.0 95.5 Plate 14M 95.0 96.5 95.5 Weld 332 96.5 97.0 99.5 Plate 165 95.5 94.0 95.0 Weld 321 96.5 97.5 98.0 I
Plate 15T 93.5 94.0 96.0 Weld 32U 97.0 96.0 95.0 Plate 13Y 96.0 96.0 97.0 Weld 32Y 97.0 96.0 96.0 Plate 13U 95.5 95.5 96.0 Weld 333 97.0 97.5 94.0 Plate 166 94.0 93.5 95.0 Weld 33P 97.5 96.5 93.0 I
Plate 145 95.0 95.5 96.0 Weld 334 94.0 96.5 95.5 Plate 13M 95.5 94.5 98.0 Weld 35K 96.5 96.0 96.0 Plate llD 94.5 93.5 95.5 Weld 32L 95.0 95.5 97.0 I
Plate 116 95.5 95.5 95.0 Weld 34E 97.0 97.5 98.0 Plate llL 95.0 95.5 94.0 Weld 36M 96.5 99.0 93.5 I
3 HAZ 43E 88.0 87.0 88.0 Ref.
67A 97.5 97.0 98.5 R
HAZ 45K 94.5 94.5 95.0 Ref.
66D 97.0 96.0 97.5 HAZ 423 96.5 95.0 96.5 Ref.
676 96.5 97.5 97.5 HAZ 44L 95.0 94.5 93.5 Ref.
677 98.5 96.5 99.5 HAZ 458 95.0 94.0 95.0 Ref.
67P 96.5 96.0 95.5 HAZ 44K 94.0 95.5 95.5 Ref.
66T 97.0 97.5 98.0 HAZ 45P 97.0 95.5 95.0 Ref.
66A 97.0 96.5 97.0 HAZ 456 94.0 94.5 94.5 Ref.
66C 97.0 98.0 96.5 HAZ 46A 96.0 96.0 95.0 Ref.
67K 96.5 96.5 96.5 HAZ 436 90.0 85.0 85.5 Ref.
667 97.5 96.5 99.5 HAZ 435 94.0 94.0 93.0 Ref.
66M 97.5 96.5 96.5 I
HAZ 437 89.5 87.0 87.0 Ref.
67L 97.0 97.0 97.0 I
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I I I I I I I I l 1 APPENDIX C PROCEDURE FOR THE GENERATION OF ALLOWABLE PRESSURE-TEMPERATURE LIMIT CURVES FOR NUCLEAR POWER PLANT REACTOR VESSELS I I I I I .I I I I
I PROCEDURE FOR THE GENERATION OF ALLOWABLE PRESSURE-TEMPERATURE LIMIT CURVES FOR NUCLEAR POWER PLANT REACTOR VESSELS I A. Introduction The following is a description of the basis for the generation of pressure-temperature limit curves for inservice leak and hydrostatic tests, heatup and cooldown operations, and core operation of reactor pres-sure vessels. The safety margins employed in these procedures equal or exceed those recommended in the ASME Boiler and Pressure Vessel Code, I Section III, Appendix G, " Protection Against Nonductile Failure." B.
Background
The basic parameter used to determine safe vessel operational con-ditions is the stress intensity factor, K;, which is a function of the stress state and flaw configuration. The Ky corresponding to membrane tension is given by (1) K M .e = Im I where M is the membrane stress correction factor for the postulated flaw g I and the membrane stress. Likewise, K corresponding to bending is m y given by (2) Ib "b * #b K I where M is the bending stress correction factor and is the bending b b stress. For vessel section thickness of 4 to 12 inches, the maximum pos-tulated surf ace flaw, which is assumed to be normal to the direction of j I I
I I maximum stress, has a depth of 0.25 of the section thickness and a length of 1.50 times the section thickness. Curves for M versus the square root m of the vessel wall thickness for the postulated flaw are given in Figure 1 as taken from the Pressure Vessel Code (ref. Figurc G-2114.1). These curves are a function of the stress ratio parameter c/c, where e is the y y material yield strength which is taken to be 50,000 psi. The bending cor-rection factor is defined as 2/3 M and is therefore determined from m Figure 1 as well. The basis for these curves is given in ASME Boiler and Pressure Vessel Code, Section XI, " Rules for Inservice Inspection of Nu-clear Power Plant Components," Article A-3000. that can cause failure as a func-The Code specifies the minimum Kg tion of material temperature, T, and its reference nil ductility tempera-ture, RTNDT. y is defined as the reference stress intensity This minimum K factor, KIR, and is given by 26777. + 1223. exp [ 0.014493(T-RTNOT + 160) ] (3) K = yp where all temperatures are in degrees Fahrenheit. A plot of this expres-sion is given in Figure 2 taken from the Code (ref. Figure G-2010.1). C. Pressure-Temperature Relationships 1. Inservice Leak and Hydrostatic Test During performance of inservice leak and hydrostatic tests, the reference stress intensity f actor, KIR, must always be greater than 1.5 times the Kg caused by pressure, thus 1.5 Kyp < KIR (4) i
I I I 3.0 l I I l l I /y 3.G f/.EMBRANE Kim " Mm * "m f 1.0 BENDING Kib ' Mb * "b / g ,/ / O'. M = 2 / 3 f.i b m g ,n 2.0 I 2.0 d +gh 2 l.G I l.4 1.2 1.01.0 1.2 1.4 1.0 1.8 2.0 2.2 2.4 2.G 2.8 3.0 3.2 3.4 3.6 3.8 4.0 YTHICKNESS (IN.) I I l Flaure 1. Stress Correction Foctor I I I
I I 17 0 ICO 15 0 -- (Kig - 2G.777) -1.2 2 3 e0 0M e.h -(RT Hor ie v.HERC 13 0 K = REFERDJCE STRESS INTENSITY FACTOR ig 32p ___J_. T = TEMPERATURE AT v/HICH Kin l IS PERMITTED,'F liO RT = REFEREf!CE I!!L-DUCTILITY l isDT TEl. FERATURE l:.100 5._ S O I / f CD E x 70 00 I'
- ,0 40 30 20 10 O
l. - 24 0 -200 -100 -12 0 -80 -40 0 40 eo .i20 IGO 200 240 TEMPERATURE RELATIVE TO RTg g,(T-RT,.g), FAHRENHEIT DEGREES I I I I Figure 2. Reference Stress Intensity Factor I I I
or 1.5 M
- K (5) m m IR For a cylinder with inner radius rj and outer radius r, the g
stress distribution due to internal pressure is given by r,2 \\r 2+r\\ 2 (6) a(r) = I 2 2 2 I -r r o 4 With 1/4T flaws possible at both inner and outer radial locations, i.e., $ + 1/4(r -r$) and r3/r = r$ + 3/4(r -r$), the maximum stress at rif4 = R g g will occur at the inner flaw location, thus [r 3 2+(1/4r +3/4r 2-2 r j g g 3 - r$ j (1/4r + 3/4r$)2 max o 2 2 r g g With the operation pressure known, i.e., P, we determine the o minimum coolant temperature that will satisfy Equation (4) by evaluating 1.5 M (8) K = IR m max I and determine the corresponding coolant temperature, T, from Equation (3) for the given RTNDT at the 1/4T location. For this calculation, Equation (3) takes the form KIR - 26777. I) T = RTNDT(1/4T) - 160. + 68.9988 in 1223. I I
The inservice curves are generated for an operating pressure range of.96 P to 1.14 P, where P is the design operating pressure. a g o 2. Heatup and Cooldown Operations At all times during heatup and cooldown operations, the ref-erence stress intensity factor, KIR, must always be greater than the sum of 2 times the K caused by pressure and the Kgg caused by thermal gradi-ip ents, thus I 2.0 Kyp + 1.0 kit
- KIR (10) or i
2.0 M =KIR - kit (II) m max where e is the maximum allcwable stress due to internal pressure, and max K is the equivalent linear stress intensity factor produced by the It thermal gradients. To obtain the equivalent linear stress intensity fac-tor due to thermal gradients requires a detailed thermal stress analysis. I The details of the required analysis are given in Section D. During heatup the radial stress distributions due to internal pressure and thermal gradients are shown schematically in Figure 3a. As-suming a possible flaw at the 1/4T location, we see from Figure 3a that the thermal stress tends to alleviate the pressure stress at this point in the vessel wall and, therefore, the steady state pressure stress would re-present the maximum stress condition at the 1/4T location. At the 3/4T flaw location, the pressure stress and thermal stress add and, therefore, the combination for a given heatup rate represents the maximum stress at I I
I l OUTER RADIUS = l 3/4T 3 I 3 Il4T g 1 l INNER RADIUS Pressure stress distribution Thermal stress distribution 'I (a) Heatup I OUTER RADIUS = } 3/4T l 33 l ll4T 1 I INNER RADIUS Pressure stress distribution Thermal stress distribution (b) Cooldown I 1 Figure 3. Heatup and Cooldown Stress Distribution I I
I the 3/4T location. The maximum overall stress between the 1/4T and 3/4T location then determines the maximum allowable reactor pressure at the given coolant temperature. The heatup pressure-temperature curves are thus generated by calculating the maximum steady state pressure based on a possible flaw at the 1/4T location from I I P (1/4T) = 2 (12) [r )[r +(1/4r +3/4r,)2\\ 9 g g (1/4r +3/4r )2 2-2 r$ j r g 9 g is obtained where M is determined from the curves in Figure 1 and KIR m from Equation (3) using the coolant temperature and RTNOT at the 1/4T location. Here we may note that M must be iterated for since it is a m function of the final stress ratio to yield strength (c/o ). At the 3/4T location, the maximum pressure is determined from Equation (11) as Kyp - kit P""*(3/4T) = (13) 2 (1/4r,+3/4r )2 \\ 2 )[r / r + 9 g g - r$ j (1/4r +3/4r )2 2 2 r 9 g g IR is obtained from Equation (2) using the material temperature and where K is determined from the analysis proce-RTNDT at the 3/4T location and kit dure outlined in Section D. M is determined from Figure 1. m The minimum of these maximum allowable pres:Ures at the given coolant temperature determines the maximum operation pressure.. Each heat-up rate of interest must be analyzed on an individual basis. I I
I I The cooldown analysis proceeds in a similar fashion as that described for heatup with the following exceptions: We note from Figure 3b that during cooldown the 1/4T locations always controls the maximum stress since the thermal gradient produces tensile stresses at the 1/4T location. Thus the steady state pressure is the same as that given in I Equation (12). For each cooldown rate, the maximum pressure is evaluated at the 1/4T location from KTR - kit MD" 2 (14) [ r$ )[r 2+(3/4r$+1/4r )2\\ max g g 2M* 2 2 - r$ (3/4r$+1/4r ) j r g g where K is cbtained from Equation (3) using the material temperature and IR RTNDT at the 1/4T location. K is determined from the thermal analysis It described in Section D. It is of interest to note that during cooldown the material temperature will lag the coolant temperature and, therefore, the steady state pressure, which is evaluated at the coolant temperature, will ini-tially yield the lower maximum allowable pressure. When the thermal gra-dients increase, the stresses do likewise, and, finally, the transient analysis governs the maximum allowable pressure. Hence, a point-by-point comparison must be made between the maximum allowable pressures produced by steady state analyses and transient thermal analysis to determine the minimum of the maximum allowable pressures. 3. Core Operation At all times that the reactor core is critical, the tempera-ture must be higher than that required for inservice hydrostatic testing, I E
I and in addition, the pressure-temperature relationship shall provide at least a 40*F margin over that required for heatup and cooldown operations. Thus the pressure-temperature limit curves for core operation may be con-structed directly from the inservice leak and hydrostatic test and heatup analysis results. D. Thermal Stress Analysis The equivalent linear stress due to thermal gradients is obtained from a detailed thermal analysis of the vessel. The temperature distribu-tion in the vessel wall is governed by the partial differential equation ecTT - K [(1/r)T +Trr] = 0 (15) r I subject to initial condition T(r,0) = T (16) o, and boundary conditions -KT ("i,t) = h [T (t) - T(rj,t)], (17) r o and E T (r,t) = 0 (18) r o where T =To + Rt. (19) c I o is the material density, c the material specific neat, K the heat con. ductivity of the material, h the heat transfer coefficient between the water coolant and vessel material, R the heating rate, T the initial o coolant temperature, T(r,t) the temperature distribution in the vessel, r the spatial coordinate, and t the temporal coordinate. I E
I A finite difference solution procedure is employed to solve for the radial temperature distribution at various time steps along the heatup or cooldown cycle. The finite difference equations for N radial points, at distance ar apart, across the vessel are: 1. for 1 < n < N I ~ atJ ^# t T t+at, 1, (2 + ) T oc(ar)2 r n. atK ) Tf + 1 + T t (1 + -1 (20) + oc(ar)2 n for n = 1 ath t T +at, ~ atK t y, [ g, g), T 1 oc(ar)2 ry oc(ar). 1 t" atK (1 + )Tf+ath T (21) + oc(ar)2 1 I and for n = N I atK t Kat t t T +at = 1-I + T (22) oc(ar)2 - N-1 oc(ar)2 - N N For stability in the finite difference operation, we must choose at for a given or such that both I ^ oc(ar)2 ry ) s 1 (23) (2 + i lI w ,&m-e---y w y-e %gwwm--,, - .----w--- y---w---------,------7
w----m*s-=-%-e-agr-w--,yy.-
g
I l and Ath atK (1 g). c(ar) < 1 (24) oc(ar)2 1 ~ are satisfied. These conditions assure us that heat will not flow in the direction of increasing temperature, which, of course, would violate the second law of thermodynamics. Since a large variation in coolant temperature is considered, the dependence of (K/oc), K, and h on temperature is included in the analysis by treating these as constants only during every 5'F increment in coolant temperature and then updating their values for the next 5'F increment. The dependence of (K/oc) called the thermal diffusivity and K, the thermal conductivity, can be determined from the ASME Boiler and Pressure Vessel Code, Section III, Appendix I - Stress Tables. A linear regression anal-ysis of the tabular values resulted in the following expressions: K(T) = 38.211 - 0.01673
- T (BTU /HR-FT *F)
(25) E and I 2 k(T) = (K/oc) = 0.6942 - 0.000432
- T (FT /HR)
(26) E where T is in degrees Fahrenbeit. The heat transfer coefficient is calculated based on forced convec-tion under turbulent flow conditions. The variables involved are the mean velocity of the fluid coolant, the equivalent (hydraulic) diameter of the coolant channel, and the density, heat capacity, viscosity. and thermal I E
I conductivity of the coolant. For water coolant, allowance for the varia-tions in physical properties with temperature may be made by writing
- I
/0 2 (27) 0 h(T) = 170(1+10-2
- T 5
- T ) y0.8 2
where v is in f t/sec, D in inches, the temperature is in 'F, and h is in 2 8tu/hr-ft 'F. The values for the heat-transfer coefficient given by this I relationship are in good agreement with those obtained from the Dittus-Boelter equation for temperatures up to 600*F. The mean velocity of the coolant, v, is generally given in terms of the effective coolant flow rate 2 Q (Lbm/hr) and effective flow area A (ft ). Given the relationship I o (T) = 62.93 - 0.48 x 10-2
- T - 0.46 x 10-4
- T2 (28)
I for the density of water as a function of temperature, the mean velocity of the coolant is cbtained from I v = Q/(3600
- o(T)
- A)
(29) i The thermal stress distribution is calculated from I I I r+r r c(r,t)=yf -h [I T(r,t)rdr-T(r,t)+h(r T 2 2) T(r,t)rdr (30) r .r i r -r i g 4 I Glasstone, S., Principles of Nuclear Reactor Engineering, D. Van Nostrand Co., Inc., New Jersey, pp. 667-668, 1960. I I
I where a is the coefficient of thermal expansion (in/in 'F), E is Young's modulus, and v is Poisson's ratio. This expression can be obtained from Theory of Elasticity by Timoshenko and Goodier, pp. 408-409, when imposing a zero radial stress condition at the cylinder inner and outer radius. Poisson's ratio is taken to be constant at a value of 0.3 while a and E are evaluated as a function of the average temperature across the vessel r T(r)rdr. (31) T 2 2 r avg (r - r$ ) i g The dependence of the coefficient of thermal expansion on temperature is taken to be
- T ) x 10-6 (32) 2 a(T) = (at+a2 3
- T+a Material at a2 a3 A302-B 6.776 0.003636
-0.1381 x 10-5 A533-B 6.776 0.003636 -0.1381 x 10-5 A508-2 6.125 0.004131 -0.6735 x 10-6 The dependence of Young's modulus on temperature is taken to be I E(T) = (29.87 - 0.005363
- T + 0.1918 x 10-6
- T ) x 10.
(33) 2 6 I Equation 32 and 33 were obtained from regression analysis of tabular values given in Section III, Appendix I of the ASME Boiler and Pressure Vessel Code. The resulting stress distribution given by Equation (30) is not linear; however, an equivalent linear stress distribution is determined I E
I I from the resulting moment. The mcment produced be the nonlinear stress distribution is given by r M(t) = b [ U T (r,t)rde (34) c I where b is a unit depth of the vessel. Here we note that the moment is a function of time, i.e., coolant temperature via T = To + Rt. For a lin-c ear stress distribution we have that I ( ) e max where e is the maximum outer fiber stress, c the distance from the neu-max tral axis, taken to be (r - r$)/2, and I the section area moment of iner-o tia which is given by 3 b(r - r,)3 I = bh g (36) 12 12 E Combining these expressions results in the equivalent linear stress due to thermal gradients "T(r,t)rdr (37) max " "bt " (r - r )2 i e r g g I is then defined as The thermal stress intensity factor kit I It " "b 'bt (38) .I E
I I where M is determined from the curves given in Figure 1 wherein Mb = 2/3 b M. It is of interest to note that a sign change occurs in the stress m calculations during a cooldown analysis since the thermal gradients pro-duce compressive stresses at the vessel outer radius. This sign change must then be reflected in the kit calculation for the cooldown analysis. Normalized temperature and thermal stress distributions during a typical reactor heatup are given in Figure 4. The radial temperature is shown normalized with respect to the average temperature, Tavg' DY I T-T T (39) = T-T avg max The thermal stress equivalent linearized stress, as calculated by Equa-tions (30) and (37), are normalized with respect to the maximum thermal stress. Here we note that the actual thermal stress at the 3/4T location is considerably less than the maximum equivalent linear stress which yields additional safety margins during the heatup cycle. Similar temper-ature and thermal stress distributions are developed during cooldown. Tne trends are nearly identical as those shown in Figure 4 when the inner and outer vessel locations are reversed with the 1/4T location becoming the critical point. l I l l E. Example Calculations I The following example is based on a reactor vessel with the follow-ing characteristics: Inner Radius 82.00 in. (rj) = Outer Radius 90.00 in. (r ) = g I E L
I I I I OUTER WALL I 1.0 rl \\ l 0.8 / l 0.6 -/ t g 0.4 / g 0.2 p /' 0 ll - 1. 0 0 1.0 - 1. 0 0 1.0 INNER WALL Normalized temperature Normalized stress distribution ( ATl ATmax ) distribution ( al a I max g Figure 4. Typical Normalized Temperature and Stress j Distribution During Heatup
- I I
I I .E
I I 2250 psig (P ) Operating Pressure = o 70*F (T ) Initial Temperature = g 550*F (T ) Final Temperature = f 100 x 106 lbm/hr (Q) Effective Coolant Flow Rate = 2 20.00 ft (4) Effective Flow Rate = 10.00 in. (D) Effective Hydraulic Diameter = 200*F RTtIDT (1/4T) = 140*F RTt1DT (3/4T) = In the thermal stress analysis 21 radial points were used in the finite difference scheme. Going from 70*F to the final temperature of 550*F, approximately 12,000 time (temperature via T = To + Rt) steps were required in the tnermal analysis for the 100*F/hr heatup rate. The re-sults of the example computation are shown in Figures 5 through 9. Figure 5 gives the reference stress intensity factor, Kgg, as a function of temperature indexed to RTilDT (1/4T). For the steady state analysis, KIR is converted directly to allowable pressure via Equation 12. During the heatup and cooldown thermal analyses the material tem-I perature at the 1/4T and 3/4T and thermal stress intensity factors kit j are required to compute allowable pressure via Equations (13) and (14). The material temperatures versus coolant temperature during the 100*F/hr heat-up and cooldown analyses for an 8-in. A302B wall are given in Figure 6. Figure 7 gives the corresponding thermal stress intensity factor at the 3/4T and 1/4T locations as a function of coolant temperature. For an 8-l 3/4-in. wall thickness commonly encountered in the larger pressurized water reactors, the 50'F dif ference between the coolant ar.d 3/4T vessel wall temperatures shown in Figure 6 would increase to 60-65'F. Also, the I E
ma m um uma m Wu m gun aa um aus seu e um um en m 200 i i i i i i 160 RTg ( 1/4 T ) = 200 *F .E 120 i ? 3 5 80 i x 40 0 i 50 100 150 200 250 300 350 400 TEMPERATURE (*F ) i Figure 5. Reference Stress Intensity Factor as a Function of Temperature indexed to RTNDT( ll4T ) P
I 400 l 100 *FIHR HEATU P ( 3/4T Location ) / I -- 100*FlHR C00LDOWN (1/4T Location ) / / / E / / 300 / / ~ / I / 2 / / ~ g 200 7 I / / 5 / / ~ l h / O / / EXAMPLE VESSEL CONDITIONS: g 100 8-in. A302B Wall RTNDT (1/4T) = 200 F l RTNDT (3/4T) = 140 F I 0 50 100 200 300 g l COOLANT TEMPERATURE (*F ) Figure 6. Vessel Temperature as a Function of Coolant Temperature I
E I g 16 - I l 14 g i 12 - I I ^ .5 10 - f l [ 8-I g 6 100 *F /HR HEATU P ( 3/4T Location ) -- 100 *FIHR C00LDOWN (1/4 Location ) 4-I EXAMPLE VESSEL CONDITIONS: l 8-in, A302B Wall 2-RTNDT (1/4T) = 200 F RTNDT (3/4T) = 140 F l 0' 50 100 200 300 l COOLANT TEMPERATURE (*F ) l Figure 7. Thermal Stress Intensity Factor as a Function of Coolant Temperature
values for kit would be of the order of 20 ksi 'in, considerably above that shown for an 8-in, wall in Figure 7. Figures 8 and 9 demonstrate the construction of the allowable com-posite pressure and temperature curves for the 100*F/hr heatup and cool-down rates. The composite curves represent the lower bound of the thermal and steady state curves with the addition of margins of +10*F and -60 psig i for possible instrumentation error. Figure 8 also snows the leak test limit, corrected for instrument error, as obtained from Equation (9). The limit points are at the operating pressure 2250 psig and at 2475 psig which corresponds to 1.1 times the operating pressure. The critically limit is also shown in Figure 8 and is constructed by providing for a 40*F margin over that required for heatup and cooldown and by requiring that the minimum temperature ba greater than that required by the leak test limit. I I I I I I I I
W W W m mae M M M emem M M M ee I I I l I I I LEAK lEST LIMIT 3 / 2400 l' 1 2000 COMPOSITE CURVE ~100*FlHR llEATUP ( Margins of +10 F and -60 psig for instrument error )] / i \\ /I \\ / 1600 / / STEADY STATE \\ CRITICALITY _ W / LIMIT 8 / z / IlEATUP 800 N 400 1 I I I I I I 50 100 150 200 250 300 350 400 INDICATED TEMPERATURE (*F ) Figure 8. Pressure-Temperature Curves for 100*FIHr lleatup
l !g* - N R N N 'I N N c N 3: I N '52 o N 8 8 o l3 - s m i= N y '\\ 2 l \\ eE i R *u. S I z i E a 1 o s S \\ w E g-e e o \\ 8 \\ E E \\ 5 g s E'3 t E U \\ E a F = \\ I E o Sz5 o o W H tSE \\ Q z I >.m 2 3 i w o g o \\ e5 E S,6 9 m m z g o+m ~ g w3s 8 o b.5 5 N i 8 gma @N9 ( E r m ( E o-I -^ 't l i R 8 b n ~ I ( B sd ) 3BBSS3Bd 03.1.V31QN1 i I .}}