ML20330A194
| ML20330A194 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/22/1981 |
| From: | Norris E Southwest Research Institute |
| To: | Gebbie J Indiana Michigan Power Co, Office of Nuclear Reactor Regulation |
| Wall S , NRR/DORL/LPL3, 415-2855 | |
| References | |
| 02-6159 | |
| Download: ML20330A194 (105) | |
Text
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REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR DONALD C. COOK UNIT No.* 1*
ANALYSIS OF CAPSULE.X.
by E. B. Norris FINAL REPORT SwRI Project No. 02-6159 fnr Indiana & Michigan Power Company Donald C. Cook Nuclear Plant June 22, 1981 7 SOUTHWEST RESEARCH INSTITUTE SAN ANTONIO HOUSTON
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SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510f 6220 Culebra Road San Antonio, Texas 78284 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR DONALD C. COOK UNIT NO. 1 ANALYSIS OF r-
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IQI L P. 0. Box 458 Bridgman, Michigan 49106 June 22, 1981 Approved:
U.S. Lindholm, Director Department of Materials Sciences
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ABSTRACT The second vessel material surveillance capsule removed from the Donald C. Cook Unit No. 1 nuclear power plant has been tested, and the results have been evaluated.
Heatup and cooldown limit curves for normal operation have been developed for up to 12 and 32 effective full power years of operation, ii
r TABLE OF CONTENTS LIST OF FIGURES LIST OF TABLES I.
II.
III.
IV.
V.
VI.
VII.
STJMMARY OF RESULTS AND CONCLUSIONS BACKGROUND DESCRIPTION OF MATERIAL SURVEII.L.ANCE PROGRAM:
TESTING OF SPECIMENS FROM CAPSU!.E X A.
B.
- c.
Shipment, Opening, and InSl)ection of Capsule Neutron Dosimetry Mechanical Property Tests ANALYSIS OF RESULTS HEA.TUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERA-TION OF DONALD C. COOK UNIT NO. 1 REFERENCES APPENDIX A - TENSILE TEST RECORDS APPENDIX B - PROCEDURE FOR THE GENERATION OF ALLOWABLE PRESSURE-TEMPERATURE LIMIT CURVES FOR NUCLEAR POWER PUNT REACTOR VESSELS iii iv V
1 3
7 13 13 14 22 37 45 51 53 67
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Figure 1
2 3
4 5
6 7
8 9
10 11 12 13 14 LIST.OF FIGURES Arrangement of Surveillance Capsules in the Pres-sure Vessel Vessel Material Surveillance Specimens Arrangement of Specimens and Dosimeters in Cap-sule X Radiation Response of D~nald C. Cook Unit No. 1 Vessel Shell Plate (Longitudinal Orientation)
Radiation Response of Donald C. Cook Unit No. 1 Vessel Shell Plate (Transverse Orientation)
Radiation Response of Donald C. Cook Unit No. 1 Core Region Weld Metal Radiation Response of Donald C. Cook Unit No. 1 Core Region Weld HAZ Material Radiation Response of Donald C. Cook Unit No. 1 Correlation Monitor Material Effect of Neutron Fluence on RTNDT Shift, Donald C. Cook Unit No. 1 OPnPndP.nce of c__ Uooer Shelf Energy on Neutron V
Fluence, Donald C. Cook Unit No. 1 Reactor Coolant System Pressure-Temperature Limits Versus 60°F/Hour Rate Criti<;ality Limit and Hydrostatic Test Limit, 12 EFPY Reactor Coolant System Pressure-Temperature Limits Versus Cooldown Rates, 12 EFPY Reactor Coolant System Pressure-Temperature Limits Versus 60°F/Hour Rate Criticality Limit and Hydrostatic Test Limit, 32 EFPY Reactor Coolant System Pressure-Temperature Limits Versus Cooldown Rates, 32 EFPY iv 8
11 12 29 30 31 32 33 39 42 47 49 50
Table I
II III IV
.v VI
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VII VIII IX X
XI
- XII XIII LIST OF TABLES Donald C. Cook Unit No. 1 Reactor Vessel Surveil-lance Materials (12]
Summary of Reactor Operations, Donald C. Cook Unit No. l Results of Discrete Ordinates Sn Transport Analysis, Donald C~ Cook Unit No. 1, 40° Capsules Summary of Neutron Dos:imetry Results, Donald C.
Cook Unit No. l, Capsules T and X Charpy V-Notch Impact Test Results, D. C. Cook Unit No. 1, Vessel Intermediate Shell Plate B4406-3, Longitudinal Orientation Charpy V-Notch Impact Test Results, D. C. Cook Unit No. 1, Vessel Intermediate Shell Plate B4406-3, Transverse Orientation Charpy V-Notch Impac~ Test Results, D. C. Cook Unit No. 1, Vessel Core Region Weld Metal Cha11)y V-Notch Impact Test Results, D. C. Cook Unit No. 1 Vessel Core Region Weld Heat-Affected Zone Material Charpy V-Notch Impact Test Results, D. C. Cook Unit No. 1, Correlation Monitor Material Effect of Irradiation on Capsule X Surveillance Materials, Donald C. Cook Unit No. 1 Tensile Properties of Surveillance Materials, Donald C. Cook Unit No. 1 Projected Values of RTNDT for Donald C. Cook Unit No. 1 Reactor Vessel Surveillance Capsule Removal Schedule (16], Donald C. Cook Unit No. 1 V
9 16 19 21 24 25 26 27 28 34 35 40 44
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I.
SUMMARY
OF RESULTS AND CONCLUSIONS The analysis of the second material surveillance capsule removed from the Donald C. Cook Unit No. l reactor pressure vessel led to the following conclusions:
(1)
Based on a calculated neutron spectral distr1bution, Capsule X received a fast fluence of 7.4 x 1018 neutrons/cm2 (E > 1 MeV) at its radial centerline.
(2)
The surveillance specimens of the core beltline materials ex-perienced shifts in RTNDT of 110°P to 165°F as a result of exposure up to the 1980 refuelling outage.
(3)
The weld metal and heat affected zone (RAZ) materials e."dlibited the largest shifts in RTNDT" Although the intermediate shell plate mate-rials have the highest initial (unirradiated) RTNDT, the weld metal.will control the heat:up and cooldown limitations throughout: the design lifetime of the pressure vessel.
cm2 (E > 1 MeV) received by the vessel wall accrued in 3.48 effective full power years (EFPY).
Therefore, the projected maximum neutron fluence after 32 EFPY is 2.1 x 1019 neutrons/ cm2 (E > l MeV).
This estimate is based on a lead factor of 3.2 between the center of Capsule X and the point ?f maxi-mum pressure vessel flux.
- IF.,-
-.,,. I (5)
Based on the analyses of Capsules X and T, the projected values of RTNDT for the Donald C. Cook Unit l vessel core beltline region, at the l/4T and 3/4T positions after 12 EFPY of operation, are 155°F and ll5°F, respectively.
These values were used as the bases for computing heatup and cooldown limit _curves for up to 12 E:FPY of operation.
(6)
Based on the analyses of Capsules X and T, the values of RTNDT for the Donald C. Cook Unit 1 vessel core beltline region, at the l/4T and 3/ 4T positions after 32 _EFPY of operation, are projected to be 195 °F and 140°F, respectively.
These values were used as the bases for computing heatup and cooldown limit curves for up to 32 EFPY of operation.
(7)
The Donald C. Cook Unit No. l vessel plates, weld metal, and HAZ material located in the core beltline region are projected to retain suffi-cient toughness to meet the current requirements of lOCF'R.50 Appendix G through-out the design life of the unit.
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l II.
BACKGROUND
'l'he allowable loadings on nuclear pressure vessels are dateniined by applying the rules in Appendix G, "'Fracture Toughness Requirements,"
of lOCntSO [l].
In the case of pressure-retaining components ma.de of ferritic materials, the allowable loadings depend on the reference stress intensity factor (Kn) curve indexed to the reference nil duc-tility temperature (RTNDT) presented in Appendix G, "?-rotectiou Against Non-ductile Failure," of Section III of the ASME Code [2].
Furt:her, the, materials in the beltline region of the reactor vessel t11t1st be monitored for radiation-induced changes in RTNDT per the requirements of Appendix H9 "Reactor Vessel Material Surveillance PTogram Requirements," of 10CFR.50.
The RTNDT is defined in paragraph ~2331 of Section III of the ASME Code as the highest of the following temperatures:
(l)
Drop-weight Nil Duc1:ility Temperature (DW-MDT) per AST.M E 208 (3];
(2) 60 deg F below the 50 ft-lb Charpy V-no1:cil (Cv) tem:,erature:
(3) 60 deg F below the 35 mil Cv temperature.
The RTNDT must be established for all materials, including ~eld metal and heat-affected zone (BAZ) material as well as base plates and forgings, whicli comprise the reactor coolant pressure boundary.
It is well established that fer-ritic materials undergo an increase in strength and hardness and a decrease in ductility and toughness when exposed to neutron fluences in excess of 1017 neutrons ?er c:n2 (E > l MeV) [4].
AJ.so, it has been established that tramp elaments, particularly 3
copper and phosphorus, affect the radiation embrittlement response of ferrltic materials (5-6 J.
The relationship bet:Ween increase in RTNDT and copper content is defined in Regulatory Guide 1.99.
Although this docu-ment is being revised by the NRC to reflect a more recent evaluation of neutron embrittlement data by the Metal Properties CoWlcil [7], estimates of shifts in RTNDT in this report are based on the current Revision 1 of Regulatory Guide 1.99 (8].
In general, the only fen:itic pressure boundary materials in a nuclear pl.ant* which are e.--q:,ected to receive a fluence sufficient to a.ff ect RTNDT 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 place or forging material and weldments which are located in the core beltline region of high Qeutron flux density.
ASTM E 185 (91 describes the recommended practice for monitoring and evaluating the radiation-induced changes occurring in the mechanical proper~ies of pressure vessel beltline materials.
Westinghouse has provided such a surveillance program for the Donald C. Cook Unit ~o. l nuclear power plant.
Th.e encapsulated Cv specimens are located on the O.D. surface of the ther.na.l shield *mere the fast neutron flux densicy is about three times that: at the adjacent vessel,1all surface.
Therefore, the increases (shifts) in transition temperatures of the materials in the pressure vessel are generally less than the corresponding shifts observed in the surveillance specimens.
gowever, because of aziI:nlthal variations in neutron flux density, cap-sule fluences may lead or lag the ma.~imum,ressel fluence in a correspond-ing exposure period.
The capsules also contain several dosimeter materials 4
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for experimentally determining the average neutron flux density at each capsule location during the exposure. period.
The Donald C. Cook Unit No. 1 material surveillance capsules also include tensile specimens as recommended by AS111 E 185.
At the present time, irradiated tensile properties are used only to indicate that the materials tested continue to meet the requirements of the appropriate material specification.
In addition, the material surveillance capsules contain wedge opening loading (WOL) fracture mechanics specimens.
Cur-rent technology limits the testing of these specimens at temperatures well below the minimum service temperature to obtain val.ia fracture mechanics data per ASTM E 399 [10], "Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Materials."
However, recent: work reported by Mager and Witt [11] may lead to methods for evaluating high-toughness materials with small fracture mechanics specimens.
Currently, t~e NRC suggests storing these specimens until an acceptable testing pro-cedure has been defined.
This report describes the results obtained from testing the_contents of Capsule X.
These data and those obtained previously from Capsule Tare analyzed to estimate the radiation-induced changes in t~e mechanical prop-erties of the pressure vessel at the time of the refuelling outage as well as predicting the changes e..-,:pected to occur at selected times in the fu-ture operation of the Donald C. Cook Unit No. l power plant.
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III.
DESCRIPTION OF MATERIAL SURVEIU.ANCE PROGRAM The Donald C. Cook Unit No. l material surveillance program is de-scribed in detail in WCAP 8047 [l2J, dated March 1973.
Eight materials surveillance capsules were placed in the reactor vessel between the ther-mal shield and the vessel wall prior to startup, see Figure l. The verti-cal center of each capsule is opposite the vertical center of the core.
The neutron flux density at each 40° capsule location was reported to lead the maximum flux density on the vessel I.D~ by a factor of 2. 6 [121.
How-ever, in a letter to the American Electric Power Service Corpoi:a.tion (13],
Westinghouse reported that the 40° capsule lead factor had been changed to 3.7 as a result of refined calculational methods.
The capsules each contain Charpy V-notch, tensile, and WOL specimens machined from the SA533 Gr B, Cl l plate, weld metal, and heat-affected
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zone (HAZ) materials located at the core beltline plus Charpy V-notch specimens machined from a reference heat of steel utilized in a number of Westinghouse surveillance !?rograms.
The c-.hta..mi.~t:rles and heat treatment?;
of.. the vessel surveillance materials are summarized in Table I.
All test specimens were machined from the test materials at the quarter-thickness (l/4 T) location after performing a simulated postweld stress-relieving treatment.
Weld and RAZ specimens were machined from a stress-relieved weldment which joined sections of the intermediate shell course.
HAZ specimens were obtained from the plate B4406-3 side of the weldment.
The longitudinal base metal Cv specimens were oriented with their long axis1 parallel to the primary rolling direction and with V-notches perpendicular to the major plate surfaces.
The transverse base metal Cv specimens were I
.oriented with their long axis perpendicular to the primary rolling direction 7
X (220°)
W (184 °)
V (176°)
TJ (140°)
270
- I 90
- Reactor Vessel Thermal Shield Core Barrel y (320°)
z (356 °)
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. T (40°)
FIG'URE l.
AR...~.ANGEMENT OF SURVEII.U...'ICE CAPSULES I~ L':1.!:. PRESSURE VESSEL l
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TABL! I DONALD C. COOK UNIT NO. l REACTOR VESSD. SURVEII.I.ANC! MATD.IALS (12]
Heat Treatment Histor, Shell Plate Material:
Heated to 1600 r for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, water quenched.
Tempered at 1225 F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, air cooled.
Stress relieved at ll50 F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, furnace cooled.
Weldment:
Stress relieved at USO F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, furnace cooled.
Correlation Monitor:
1675 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, air cooled.
1650 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, water quenched.
12.25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, furnace cooled WO.F' 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, furnace cooled to 600 F.
Chemical Com~osition (Percent)
Material C
Mn p
s
_g_
Place B4406-3
- o. 24 l.40
- a. 2s
~eld ~.etal
- a. 26 l *.33 0.18 Correlation Monitor 0.22 1.48
- 0. 2.5 v,
9
- 0.46 *
O 44 *o t
. 0.52 -
and with V-notches perpendicular to the major plate surfaces.
Tensile spec-imens were machined with the longitudinal axis either parallel or perpendic-ular to the plate primary rolling direction.
The WOL specimens were machined with the simulated crack perpendicular to both the primary rolling direction and to the major plate surfaces.
All mechanical test specimens, see Figure 2, were taken at least one plate thickness from the quenched edges of the plate material.
Capsule X contained 44 Charpy V-notch specimens (10 longitudinal and 10 transverse from the plate material, plus 8 each from weld metal, HAZ material, and the reference steel plate); 4 tensile specimens (2 plate and-2 weld metal); and 4 WOL specimens (2 plate and 2 weld metal).
The speci-1 J~.3 men numbering system and location within Capsule Xis shown in Figure 3. }~i
/J A O ;.,£.aft-Capsule X also was reported to contain the following dosimeters for CP~Q' determining the neutron flux density:
Target Element Form Quantity Iron Bare wire 5
Copper Bare wire 3
Nickel Bare wire 3
Bare wire 2
Cd shielded wire 2
Uranium-238 Cd shielded oxide 1
Neptunium-237 Cd shielded oxide 1
Two eutectic alloy thermal monitors had been inserted in holes in the steel spacers in Capsule~-
One (located at the bottom) was 2.5% Ag and_ 97. 5% Pb with a melting point of 579°F.
The other (located at the top of the capsule) was 1.75% Ag, 0.75% Sn, and 97.5% Pb having a melting point of 590°F.
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(b)
!ansil: Spec.:.::en
- Cl!
l.4991 FIGURE 2.
VESSEL MATERIAL SURVEILLA.i.~CE SPECil1ENS 11
Spec.1.man Code:
A - Place !4406-3 (Long.)
At - Place 34406-3 (Tr.ma.)
a - w'eld l!eac A.ffac:ad Zone W -
lo'eld Maul
~ -
ASffl Canelacion Monitor R
!t 25 26 37 38 Cba.r,,y FIGURE 3.
ARR.A..'iGEMENT OF SPECIMENS AND DOSIMETERS IN CAPSULE X 12
!V.
nsTING OF SPECIMENS FROM CilSULE X The capsule shipment, capsule opening, specimen testing, and re-por~ing of results were carried out in accordance with the Pi:oject Plan for Donald C. Cook Unit ~o. 1 Reactor Vessel Irradiation Surveillance Program.
The SwRI ~uclear Projects Operating Procedures called out in this plan include:
(l)
XIII-MS-104-0, "Shipment of Westinghouse ?TNR. Vessel Materia.l Sw:ve J J ance Capsule Using SwiU Cask. and Equipment" (2)
- a-MS-101-0, "'Det:e:rm.n.at:iou of Specific Activity and Analysis of Radiation Detector Specimensrr (3) n-MS-103:..-0, "Conducting Tension Test:.s on Metallic Specimens" (4)
- O:-MS-104--0, nChU?Y !::pact Tests on Metallic Spe('i,fflens" (5)
X!II-MS-103--0, "Opem..ug Radiation Sw:veillance Cap-sules and Ha:ndJ ing and Storing Specimens" (6)
- x:I-MS-5--0, "Ccndur:tiD:g Wedge-O'peuing-Load.i:c.g Tests ou Metallic M.aterf..a.ls" Copies of the above docUlllellts are on file at SwRI.
Shiument, O'Oening, and !nsoection of Catisu.le Southwest Researc.h. I.ust~tute utilized ?Tocedure nII-MS-104-0 :or the sl:li;nnent. of Capsule ! eo ehe SwRI labora1:ortes.
SwR.! pe:rsoru::.el sev-ered c:h.e cal)sule from its eneusion eube, sec::f.oned ce a:ti:ension cube i.u1:o several len~.hs, s~ervised e.he loadi.ng oi oe ca1;)sule and e.:t-::ension rube :.iat;e.r:ia.ls into e.b.e shipping cask~ and e-:-allS-por:ad ehe cask. c:o San lnt:ou.io, Texas.
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The capsule was opened and the contents identified and stored in ac-cordance with Procedure XIII-MS-103-0.
The long seam welds were milled off using a Bridgeport vertical milling machine.
Before milling the long seam weld beads, transverse saw cuts were made to remove the capsule ends.
After the long seam welds had been milled off, the top half of the capsule shell was removed.
The specimens and spacer blocks were carefully removed and placed in indexed receptacles identifying each capsule location. After the disassembly had been completed, each specimen was carefully checked to insure agreement with the identification and location as listed in WCAP 804 7 [ 12 J.
No discrepancies were found.
The thermal monitors and neutron dosimeter iorlres were removed from the holes in the spacers.
The thermal monitors, contained in quartz vials, were examined.
No evidence of melting was observed, thus indi-cating that the maximum temperature during exposure of Capsule X did not exceed 579°'F.
All neutron dosimeters were in the positions called out in WCAP 8047 and were correctly accounted for.
B.
Neutron Dosimetry The gamma activities of the dosimeters were determined in accordance with Procedure XI-MS-101-0 using an IT-5400 multichannel analyzer and a Ge(Li) coaxial detector system.
The calibration of the equipment was ac-complished with 54Mn, 60cc, and 137cs 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*
shutdown.
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The dosimeter wires were weighed on a Mettler !!:licrobal.ance, and the fission mon.it:ors were weighed on a Metc.ler digital balance after these materials had been deencapsulated.
Infinitely dilute saturated activ-ities (AsAI) were calculated for each of the dosimeters because ASA! is directly related to the product of the energy-dependent microscopic acti-vation cross section and the a.eutrou flwc density.
The relationship be-
~een ATOR and ASAT is given by:
where:
decay constant for the activation product, day-1; equivalent operating days ae 3250 ~
for 01>-
eracillg period m; and decay t~e after O'perating period m. days.
An alternate expression which gives equivalent results is:
':Jhere:
01>era.tillg days ; and average fraction of f-ull power during operating period.
Ihe Donald C. Cook Unit No. l operating history up to tile 1980 refueling shutdowu 9 *.,1hich. r..1as used ill t:b.e ca.lculat:i.on of Aroa, is presented ill Taole The pri:::la..ry result desired from t:he dosimetar analysis is the ~otal fast neutron fluence (> 1 MeV) ~hich :he sur7eilla.nce s~eci~ens recei'7ed.
~e average flu::t de!l.Si:Y at full power is gi7en by:
1.3
~*rac1A1 Dae**
Period Scar,:
St01J l
02/02/75 02/14/75 02/15/75 02/16/75 2
02/17/75 02/17 /75 02/l8/75 02/20/75 3
02/21/75 03/18/75 03/l9/75 04/03/75 4
04/04/75 06/24/75 06/25/75 06/26/75 5
06/27/75 07/03/75 07/04/75 07/22/75 5
07/23/75 lO/ll/75 l0/12./75 l0/14/75 7
lO/lS/75 l0/31/75""
U/Ol/75 ll/14/75 a
U/1.5/75 Ol/Ol/76 01/02/76 Ol/04/76 9
Ol/05/76 04/12./76 04/13/76 05/09/76 10 05/10/76 07/0l/76 07/02/76 07/05/76 ll 07/06/76 09/10/76 09/ll/76 09/18/76
- 12.
09/19/76 ll/20/76 ll/21/76 U/21/76 13 ll/22/76 12./23/76 12./24/76 02/19/77 14 02/20/77 04/09/77 04/10/77 04/10/77 lS 04/U/77 06/17/77 06/18/77 06/19/77 16 06/20/77 07/31/77 08/0l/77 08/02/77 17 08/03/77 09/01/77 09/02/77 09/04/77 LS 09/05/77 U/lS/77 U/!.9/770 ~ U/02/77
'19 12./03/77 ~17/78 Ol/18/78
/18/78 20 Ol/19/78 04/06/78 04/07/78 06/21/78
-21 06/22/78 12/25/78 12./25/78 U/30/78 22 12/31/78 03/23/79 03/24/79 03/24/79 23 03/25/79 04/06/79 04/07/79 07/16/79 Z4 07/17/79 10/27/79 10/28/79 U/09/79 25 ll/10/79 U/24/79 12./25/79 01/16/80 26 Ol/17/80 05/30/80 TABLE II
SUMMARY
OF REACTOR OPERATIONS DONALD C. COOK UNIT NO. 1 P-r
~*racil:is SbucdOllll C:.erad.01l Dazs Davs
~MW:2 lJ 2.194 2-l 228 3
26 29,604 16 82 200.616 z
1 lS,432 l9 8l 201,.306 3
l7 40,163 l4 4a U6.552 3
99 256.178 Z7 53 143.868 4
67 205.682 8
63 196,320 l
32 92,754 38 49 107,673 l
68 l.Jl.002 z
42 85,954 2
30 75,327 3
75 lSS.989 14 46 W,.530 l
78 237,365 76 187 546,265 5
83 253,252 l
l'J 37,l.75 lOl 103 314,210
!.3 45 135,329 23 135 421.603 Toa.ls 4.130.971 (a)
Ona aqu:l.val.enc o-perac:1.ng clay
- 3250 !.twt.
(b )
?qua.ls J. 43 tn'Y.
16 *
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!qu:J. va.1.eiic Decay T:ille
~*rae1A1 Days<*>
After Period
~Tml
~ cml 0.68 l932 0.07 1929 9.U l900 6l. 73 1802 4,75 1793 62.00 1693 U.JS 1673 3S.36 1611 78.82 l509 44.27 1429 63.29 l.358 60.47 12.87 28.54 US4
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33.lJ U47 40.Jl 1078 26.4S 1034 23.lS 1002 48.92 924 38.62 864 73.04 785 168.08 522 77.92 434 U.44 420 96.68 ll6 41.64 l.58 l.29. 72 0
where:
energy-dependent neutron flux density, n/cm2:.sec; saturated activity, dps/mg target element; spectrum-averaged activation cross section, cm2; and number of target atoms per mg.
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 Capsu.le X, the weld metal and the transverse shell ?late Charpy specimens were located in the specimen layer nearest to the vessel wall and the longitudinal shell plate and the HAZ material Charpy specimens were located in the specimen layer nearest to the core.
Since there is a radial dependence of the fast neucron flux in the vessel, ~~e neut~on ex-posure received by the weld metal and transverse shell plate Charpy speci-mens is expected to be lower than that received by the longitudinal shell plate and a.AZ material Charpy specimens.
The dosimetry program was planned
.J,.
to provide infer.nation *on the radial dependence of the fast flu..~ si:ice the copper and nickel threshold detectors were located on t~e radial center-lines of the Charpy specimen layers nearest to and far~hesc from the core, respectively, and the iron threshold detecto-::-s were located at the radial position corTesponding to the interface be~Neen the ~#o Charpy specimen layers.
Additional dosimetry included the =ission monitors located at the radial centerline of the capsule and the ther:nal neut=on ~onitors (bare and cadmium-shielded cobalt) located at the radial centerline of the Charpy specimen layer nearest the pressure vessel wall.
As a part of the analysis of Capsule T [14], Southwest Research Institute performed a two-dimensional discrete ordinates transport cal-culation with the DOT 3.5 code, a 22-group neutron cross section library (CASK), a P1 expansion of the scattering matrix, and an s8 order of angular quadrature.
A one-eighth segment of a plane through the vertical axis was used to model the core, core barrel, thermal shield, pressure vessel, and three water regions using R-e coordinates.
The surveillance
\\_
capsules and their holders were not included in the model.
Recent analytical studies show that the perturbation of the neutron flux profile by a surveillance capsule and its holder can be significant
[15].
Both the magnitude of the flux and the spectriJm-averaged cross sections for the neutron detector reactions are affected.
When the cap-sule perturbation has little effect on the flux incident on the pressure vessel wall (as is. generally the case for Westinghouse l'WR vessels), the capsule lead factor is also directly affected by the perturbation.
The discrete ordinates Sn transport analysis for the Donald C. Cook Unit No. l reactor vessel was repeated but with the capsules and holders included in the model.to determine the a:x:i.al, radial, and azimuthal depen-dence of the fast neutron (E > 1.0 MeV) flux density and energy spectrum within the reactor vessel and surveillance.capsules.
These results were used to calculate the spectrum-averaged cross sections for the threshold detectors and the lead factors for use in relat:i:tg neutron exposure of t~e pressure vessel to that of the surveillance capsule.
The pertinent factors obtained from this transport analysis are summarized in Table III.
!he lead factors are higher than *the 2.6 average lead factor used in the analysis of Capsule T [14], as expected [15].
18
(
A.
B.
TABLE III RESULTS OF DISCRETE ORDINATES Sn TRANSPORT ANALYSIS DONALD C. COOK UNIT NO. l 40° CAPSULES Calculated Reaction Cross Sections for Analysis of Fast Neutron Monitors (E > 1.0 MeV)
Reaction u (barns) 54Fe(n,p)54Mn
.0870 58Ni(n,p)58co
.116 63cu(n,a) 60co
.00093 238u(n,f)
.387 237Np (n, f) 2.46 Calculated Caosule Lead Factors Position Ca)
Location within Capsule Lead Factor Cb) 211.7 cm Center of core-side Charpy layer 3.34 211.9 cm Center of capsule 3.19 112.1,:!!!
(' "'" f'"'.,.,,,= !'.VO ~pi:,.rimen layers 2.99 212.7 cm Center of vessel-si.de Charpy layer
- 2. 69 (a)
Distance from center of core (b)
Capsule neutron flux density, E > 1.0 MeV Maximum neutron flux density at vessel I.D., E > 1.0 MeV 19
The activities of dosimeters obtained from Capsule X and those pre-viously reported for Capsule T [14] are presented in Table IV.
A summary of the Capsule X and vessel I.D. fluxes calculated for full-power operation is as follows :
Dosimeter Measured Capsule Flux Lead Peak Vessel Flux at I.D.
Type cm-2.sec-l, E > 1 MeV Factor cm.-2.sec-l, E > l MeV Copper 5.83 X 10lO 3.34
- 1. 75 X lOlO Fission Iron 5.76 X 10lO*
2.99 1.93 X 1010 Nickel 5.60 ;,c 1010 2.69 2'.08 X 1010
- If a fission-spectrum. energy distribution is assumed at the capsule lo-cation, the cross section for the 54n(n,p)54Mn. reaction (E > 1.0 MeV) would be 98.26 mb [4], and the resulting value for fast flux at the cap-sule location would be 5.10 x 1010 Clll-2.sec-1.
This value is reported
(
for reference only and has not been used in the analysis of results.
The discrepancies in the peak vessel flux values determined from the several dosimeter materials are attributed primarily to the uncertainties,:-~,
~
in the calculated spectra and in the reaction cross sections.
Other neutronic factors contributing to the estimated+/-. 16.5% uncertainty (lcr) in a calculated flux value are the determination of disintegration rates and the calculation of reaction rates (AsAT/No).
Averaging the results obtained *from the Capsule X and T neutron dosimeters, the peak neutron flux incident of the I.D. surface of the pressure vessel to the 1980 refuelling outage is calculated to be 2.10 x 1010 cm-2.sec-l, E > 1 MeV.
The calculated full power neutron flux for the weld metal and transverse plate Charpy specimen layer is given by:
2.10 X 1010 X 2.69 = 5.6 X 10lO
( I
TABLE rv S1.JMMARY OF NEUTRON DOSIMETRY RESULTS DONALD C. COOK UNIT NO. 1, CAPSULES T AND X Oo*1-c*r l'o*iC10<i(&)
- u.1 =-
l Zll.9 ca I
1 2u.2 =-
212.7 ca I
0o*1-C*I.'
!d*nU.ticacioa Cia (To1t Mid.db) c~ (!'11d4l*>
Cia (!loccoa Middle)
Cia (To1t lUdlil&)
Cia (Middle)
Cl& (!acc.,. Midcll.a) tr-Z31 (lUddla) ll,-237 (lU.dclle) cr-231 (M:f.dGle)
!I!'"' U 1 (M1d4le)
!"* (To1t) re (To, !U.ddle) f* (Middl*i *
!'e (Joccca l!iddla)
!"* (!occca)
Te (To1t) re (T01t !Ucl.dla)
!'* (M:f.ddla)
Fe (!oc:011 !U.dlll*>
!a (3occ->
!Ii (T°' !Uddla)
- 1:1. (Middle)
!Ii (Joc:oa lUdGlal
!Ii (TOIi ~*l
!Ii (Middl.e) lli (Joccca !.U.Gdla)
Co (TO\\')
Co (lloc:-)
Co(CJ) (Tc,i)
Co(Cdl (lloci:ma)
Co (T°")
Co (lloc1:oa)
,:,i(Cd) (:QI))
C-,(Cd) (!oCtml)
(a)
Otacanca !=oo, caacer 0! cora.
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21
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- l. 93 x J.03 l.69 JC toJ l.69 ii: 103 l.&O JC lO.l J.98 X 104 4.02 X 104
- . u t: 104 3.!3 x 104-J. 77.x 104 J.95 X 104 l,.44 X 10]
- l. 4iO s lO; S.J2 s io6 S,.,3 X 106 i..a1 x ioo 5.03 X 10° l.4J X lQO 1.64 x 106 A.u:
(dpa/g) l.t.l x lO" 3.43 X 10" 3.Jl,r; lO"'
J.43 X 102 J.Jl >: 10" 4,QJ X 102
- 4. 79 X lQJ 3.39 : 104 (c)
(cl J.ll x lOJ J.ll x lOJ 3.lJ c 103 J.2, : 103 J.J8 X J.o3 J.J4 X 103 1.94 : ioJ 2.93 1: ioJ
~*uJ.* na<0>
I oa*2 *..,c*l r t > l ~17)
S.39 X lQlO 5.&3 z 1010 5.76 X lOJ.O S.o.3 1: 1010 5.11 x lolO 5.61 X tol0
!.l.6 X 10.l.Q
!.5.5 X 1010 (c)
Cc)
S.72,r; 10l0 S.90 : 10lO S.7S : 1010 o.0.5 x lolO 6.21 X l0l0
&.l4 x J.01.0 S.40 : iolO s *.1a : 1010 S.34 X lQlO
).ll x 10.J S.11 x 10.i.U A~ap
- 5.76 X t0l0 i..53 : 10" 4.51 X 10'"
- 76 1: 104
- r: 104 i..Ja i: l0'-
4.J9 :r: 10" 4.30 :r: 107 4.17 :i: 107 l.63 X 10; l.65: 107 l.~,: 107 J.36 X lC/ 7 1..i2 :r: 107 1.09 X 107 5 ** 51 : 1010 S.62 :r: 1010 S.!6 1: 1010 S.1.9 : ltllO s *.19 x 1ol0 5.:3 X 1010 (d)
I I
Similarly, the calculated full power neutron flux for the longitudinal plate, RAZ material and reference material Charpy specimens, the tensile specimens, and the WOL specimens are given by:
2.10 x 1010 x 3.34 a 7.o*x 1010 (Longitudinal Plate and HAZ Cv Specimens) 2.10 x 1010 x 2.99 a 6.3 x 1010 (Tensile and Reference Material Cv Specimens) 2.10 x 1010 x,3.19 a 6.7 x 1010 (WOL Specimens)
Since Donald C. Cook Unit No. l operated for 1271.1 effective full power days up to the May 1980 refuelling, the calculated fluences for Capsule X. and the vesse.l to that time ar.e as follows:
Weld Metal and Transverse Plate Cv Specimens - 6.2 x 1018 n/cm2 Longitudinal Plate and RAZ Cv Specimens
- 7. 7 x 1018 n/ cm2 Reference Material C.,, and Tensile Specimens
- 6. 9 x 1018 n/ cm2 WOL Specimens
- 7.4 X 1018 n/c:m2 Pressure Vessel ID Sur=ace
- 2.3 x 1018 n/cm2 C.
Mechanical Proper~v Tests Toe irradiated Charpy V-notc.h specimens were tes~ed on a SATEC im-pact machine in accordance with Procedure XI-MS-104-0.
Toe test temperatures were selected to develop the ductile-brittle transition and upper shelf regions.
The unirradiated Charpy V-notch impact data reported by Westing-house [12] and the data obtained by SwRI on the specimens contained in Capsule X are p-resented in Tables ~, through IX.
Toe Charpy V-notch trans-ition curves for the vessel plate, the weld metal, and the HAZ and reference materials are presented in Figures 4 through 8.
The radiation-induced shift in transition temperatures a.re indicated at the 50 ft-lb, 30 ft-lb, and 22 I
35 mil lateral expansion levels.
A summary of the shifts in RTNDT and Cv upper shelf energies for each material is presented in Table X.
Tensile tests were carried out in accordance with Procedure XI-MS-103-0 using a 50-kip capacity tester equipped with a strain gage extensometer, load cell, and autographic recording equipment.
Tensile tests on the plate material were run at 250°F and 550°F; those on the weld metal were rim at 160°F and 550°F.
The results, along with tensile data reported by Westinghouse on the uniITadiated materials [12}, are presented in Table XI.
The load-strain records are included in Appendi.~ A.
Testing of the WOL specimens was deferred at the request of Indiana
& Michigan Electric Company.
The specimens are in storage at the SwRI radiation laboratory.
23
TABLE V CRARPY V-NOTCR IMPACT TEST RESULTS, D.C. COOK UNIT NO. 1 VESSEL IN'IEBMEDIA'IE SHELL P¥TE B4406-3, LONGITUDINAL ORIENTATION Specimen Condition No.
Capsule X A-39 A-31 A-40 A-32 A-38 A-33 A-34 A-35 A-36 A-37 (a)
(b)
(a)
Uninadiated (12].
(b)
Not reported.
Test Temperature (deg F) 40 75 90 110 135 160 210 250 300 300
-40
-40
--40 10 10 10 40 40 40 76 76 76 110 110 110 160 160 160 210 210 210 300 300 300 24 1.1£1~ p.J.-;)...
Fracture Lateral Fracture Energy Expansion Appearance (ft-lb)
(mils)
(% shear) 15 14 5
20.5 18 5
31 28 10 39.5 36 15 49.5 45 15 51 48 20 78.5 68 75 99.5 83 100 106 87 100 102 83 100 10 13 6
11 10 3
11.5 11 2
24.5 24 9
33 29 11 31.5 28 13 57 49 23 42 40 25 65 54 29 82 67 45 70 60 37 78 61 37 93.5 72 52 100 77 59 88 72 52 110 84 95 131.5 95 100 115.5 83 95 120 89 100 144 98 100 125 95 100 131.5 90 100 126 92 100 132 93 100 f
(
( /
(
(
TABLE VI CRARPY V-NOTCH IMPACT TEST RESULTS, D.C. COOK UNIT NO. 1 VESSEL INTERMEDIATE SHELL PLATE B4406-3, TRANSVERSE ORIENTATION
&,.,:;;_ ~ /8 fj..,1,1.,
Test Fracture Lateral Fracture Specimen Condition No.
Capsule X AT-40 AT-31 AT-32 AT-39 AT-33 AT-38 AT-34 AT-35 AT-36 AT-37 (a)
(b)
(a)
Unirradiated [12].
(b)
Not reported.
Temperature (deg F) 40 75 110 135 160 185 210 250 300 300
-40
-40
-40 10 10 10 40 40 40 76 76 76 76 llO 110 110 160 160 210 210 210 300 300 300 25 Energy Expansion Appearance (ft-lb)
(mils)
(% shear) 9 9
5 13.5 16 10 27.5 30 J,..)
31.5 32 15 46 39 20 49 48 50 71 65 90 75 83 100 81.5 77 100 80 68 100 11 12 5
11.5 15 5
14 1.5 5
28 28 14 23 22 9
30 26 9
40 36 18 41 35 23 37 34 18 83 56 27 43 44 27 50 46 32 50 44 27 84 71 48 54 51 37 68 57 41 97 80 90 77 71 90 90 75 100 95 79 100 97 79 100 100 83 100 94 75 100 101 85 100
TABLE VII CHAR.PY V-NOTCR Il1PACT TEST RESULTS, D.C. COOK UNIT NO. 1 VESSEL CORE REGION WELD METAL Specimen Condition No.
Capsule X W-31 W-30 W-25 W-32
- W-26 W-27 W-28 W-29 (a)
(b)
(a)
Unirradiated [12J.
(b)
Not reported.
Test Temperature (deg F) 0 40 75 90 110 160 210 250
-140
-140
-140
-100
-100
-100
-70
-70
-70
-40
-40
-40 10 10 10 76 76 76 210 210 210 26
{p.;/£18p.J.J--
Fracture Lateral Fracture Energy Expansion Appearance (ft-lb)
(mils)
(% shear) 18 14 nil 23.5 21 10 51 45 60 45.5 42 55 40 32 15 71.5 65 100 66 61 100 82.5 78 100 11 10 0
21 i7YP 19 3
<£).J-1.r-18 3
22 18 29 ----
26 20 20 18 11 45.5 39 24 51 47 42 54 49 32 63 52 47 -
59 53 34 69 60 47 83 69 73 84 72 71 92 75 75 114 88 99 107 87 100 107 88 100 110 90 100 112 87 100 111 93 100
(
(
(
C
(
TABLE VIII CBARPY V-NOTCR nrPACT TEST RESULTS, D.C. COOK UNIT NO. 1 VESSEL CORE REGION WELD HEAT-AFFECTED ZONE MATERIAL
- 1. 1 I & fJ, Ad.
Test Fracture Lateral Fracture Specimen Temperature Energy Expansion Appearance Condition No.
Capsule X H-32 H-31 H-.25 H-26 R-30 R-27 H-28 H-29 (a)
(b)
(a)
Unirradiated [12].
(b)
Not reported.
(deg F) 0 40 75 110 135 160 210 250
-175
-175
-175
-140
-140
-100
-100
-100
-70
-70
-70
-70
-40
-40
-40 10 10 10 76 76 76 210 210 210 27 (ft-lb)
(mils)
(% shear) 24 16 5
31.5 24 10 31.5 26 15 49.5 45 45 54 47 40 89 69 100 80.5 66 100 75 64 100 5.5 3
0 7
5 0
7 5
0 16 12 3
22 18 5
30 25 13 33 28 14 45 40 20 52 39 21 47 35 25 27 21 14 30 24 20 54 53 55 71 50 50 47 45 43 97 83 90 89 67 43 82 64 69 112 86 100 140 84 100 131 82 100 129 85 100 104 94 100 105 87 100
TABLE IX CHARPY V-NOTCR IMPACT TEST RESULTS, D.C. COOK UNIT NO. l CORRELATION MONITOR MATERIAL Specimen Condition No.
Capsule X R-25 R-26 R-32 R-27 R-28 R-29 R-30 R-31 (a)
(b)
(a)
Unirradiated [12].
(b)
Not reported.
Test Temperature (deg F) 75 110 135 160 210 250 300 300
-so
-so
-so
-20
-20
-20 10 10 10 40 40 40 85 85 85 110 110 110 160 160 160 210 210 210 300 300 300 28
/..p,,1 1'9 p.J.'J-Fracture Lateral Energy Expansion (ft-lb)
(mils) 7 6
14.5 15 23 23 40.5 35 57.5 53 79 70 86 73.
78, 70 5
3 5
5 3
4 6.5 6
9 10 6
9 12 15 14.5 14 13.5 14
- 22.
23 36 32 35 32 58.5 51 41.5 42 52 45 82.5 60 85.5 71 63.5 54 108.5 72 81 69 109 79 117 84 115 88 121 87 125 87 117.5 83 127 84 Fracture Appearance
(% shear) nil 10 15 15 50 100 100 100 9
9 9
9 13 13 23 23 23 33
-'29 29 43 41 42 58 67
'55 84 85 87 98 98 100 100 100 100
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TABLE X EFFECT OF IRRADIATION ON CAPSULE X SURVEILLANCE MATERIALS DONALD C. COOK UNIT NO. 1 Crite.rion (1)
Transition Temperature Shift
@ 50 ft-lb
@ 30 ft-lb
@ 35 mil llRTNDT(5)
Cv Upper Shelf Drop CF Refer to Figures 4-8.
Weld Meta1(2)
(,. :*
165°F 165°F 160°1?
165°F 37 ft-1b (34%)
\\ \\ \\
D,; i o,74
']()(,.~
HAZ Material (3)
- 1, 7 170°F 160°F 160°F 160°F 38 ft-lb (32%)
I ") /')
.,.., \\._,
Trans. Plate B4406-3(2}
(
)
'*t'} !
H0°F H0°F 100°F 110°F 17 ft-lb (18%)
cl *,
(1)
(2)
(3)
(4)
(5)
(6)
Fluence -- 6.2 x 1018 n/cm2, E > l MeV.
Fluence - 7.7 x 1018 n/cm2, E > l MeV.
Fluence - 6.9 x 1018 n/cm2, E > 1 MeV.
Maximum transition temperature shift at Transition temperatures at 77 ft-lb, 46 30 ft-lb or 35 mil.
ft-lb, and 54 mils (17).
Long. PtatJ;_
84406-3 3,b) 120°F H0°F 100°F ll0°F 28 ft-lb (22%)
\\ 30 c\\ I*\\
o,c\\'\\
,1*1 **.i Ref. Plate Materia1(4)
<1, I H5°F 100°F 100°F 100°F 42 ft-lb (34%)
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w l.Jl
-~"
TABLE XI TENs*u.E PROPER'l'IES OF SURVEILLANCE MATERIALS DON~LP C. COOK UNIT NO. l Fracture Test
- Spec, Temp.
O.~% YS UTS Load Condition Material No.
(OF)
~
s i)
Jksi)
(lb)
Capsule X Plate 84406-3 AT-7 250 7J.7 97.6 3560 J
(Transverse)
AT-8
'550 71.0 95.6 3590 Weld Metal W-8 160 BC. 8 98.6 3960 W-7 550 8C.8 99.l 3680 (b)
Plate Bl1406-3 Room 68.7 90.3 (c)
(Transve ~se)
Room 6i.6 89.4 I
300 61.0 82.8 300 60.9 81.9 600 58.3 86.0 600 55.9 86.6 Weld Metal Room 66.9 81.5 j
Room 6 7. 4 82.2 300 59.7 74.6 JOO 59.8 74.5 600 57.2 79.4 600 56.3 78.5 (a)
Result obtained invalid (larger than t )tal elongation).
(b)
Unlrradiated [12).
(c)
Data not reported (12].
Fracture Stress (ksi) 172.8 166.2 234.J 172.o (c)
- t. 9'£. 18 F.:J &~
- Uniform Total Elongation Elongation R.A.
(%)
(%}
_m_
15.4 19.7 57.7 15.1 18.4 55.6 (a) 24.3 65.3 14.9 19.6 56.4 15.4 26.6 65.8 15.7 25.6 65.0 13.o 23.0 65.0 12.6 23.3 64.6 15.8 24.8 58.8 16.3 24.7 58.6 16.6 28.7 73.2 14.8 25.0 65.3 12.5 24.0 72.9 12.4 23.3 71.8 13.5 23.4 65.2 14.0 23.6
- 63. 2
V.
ANALYSIS OF RESULTS The analysis of data. obeained from sun-eillance p-rcgram specimens has the following goals:
(l)
Estimate t...~e period of time over wirl.c.h t:he propel:"t:ies of the vessel belt.line materi.al.s will meet the fracture toughness requirements of Appendix G of l0CTit.5O.
This requires a p-rojection oi t:he ~ea.sured reduc-tion in Cv upper shelf energy to the vessel wall using knowledge of the energy and spatial distribution of the neutron flux and t..~e dependence of
~ upper shel.f energy on the neu::-ron fJ.uence.
(2)
Develop heatup and cooldo,;.m, curves to descr...be the operational l.im:i.tations for selected pertocis of ti.lie.
This requires a projection of the measured shi.ft in RTMI)T to the vessel ~all using knowledge of the dependence of the shi.£1: in Rl'NDT on the neut:ron fluenc:e and the energy and spatial dis-tn.butiou of the rieutron flu:..
The energy and spatial distribution of the neutron flu:t for Donald C.
Cook Unit No. l was calculated for Capsule X with a discrete ordinates transport code.
This analysis, also applicable to Capsule T, predicted that the lead factor (ratio of fast flux at the capsule location to the maximum p-ressure vessel flu:) was 3.19 at the capsule centerline, 3.34 for the core-side Cha.rpy layer, and 2.69 for the vessel-side Char?Y layer (see Table III).
This analysis also predicted that the fast flux at the l/4! and 3/4! positions in the 8.5-in. pressure vessel -.all would be 54%
and 10%, respectively, of that a.t the,,essel !.D.
liowe"7er, in this report the projection of Capsule T results to the pressure vessel wall utilizes the more conser1ative attenuation figures of 60% and 15: :or :he 1/4!
37
and 3/4! positions to allow for the increased fraction of neutrons which might accrue in the 0.1 to 1.0 MeV range in deep penetration situations.
A method for estimating the increase in RTNDT as a function of neu-tron fluence and chemistry is given in Regulatory Guide 1.99, Revision 1 (7].
However, the Guide also permits interpolation between credible sur-veillance data and extrapolation by constructing response curves through the data points and parallel to the Guide trend curves.
The data from Caps~es T and X are deemed to be credible because (1) the surveillance materials are judged to be controlling W'ith regard to radiation damage, (2) the scatter in the Charpy data is small, (3) the changes in yield strength are consistent with the Charpy curve shifts, and (4) the response of the con-elation monitor material is W'ithin the expected response range.
The slopes of the response curves are less than the square root of fluence utilized in Regulatory Guide 1.99, and recent work [7] indicates that the square root of fluence dependence may be too high.
However, the pro-jected responses of the Donald C. Cook Unit No. l vessel bel~line materials are based on the trend curves of Figure 9 ~h~ch ~ere constructed in accor-dance rith ReguJ.atory Guide 1.99 procedures.
The Donald C. Cook Unit ~a. 1 weld metal and HAZ surveillance ma-terials are more sensitive than the vessel plate surveillance material to irTadiatio~ embrittlement.
llthough the unirTadiated values of RTNDT for the intermediate shell plates are higher than those of the weld and HAZ materials [16], the beltline region weld metal is projected co control the adjusted value of RT~T through the 32 EFPY design life of Donald C.
Cook Unit No. 1.
A sUII!lilary of the ?rejected values of RT:IDT for 12 and 32 ~z?Y of operation of Donald C. Cook Unit ~o. l is presented in Table XII.
38
~
00 (l)
'd QJ
~
µ Ill i~
QJ !
H Q)
(J p
Q)
H Q)
'H (I)
~
w 4-i
- 4) 0
µ r::
Q) 13
.j..J U}
~
'17
~
'tj (I)
.j..J u
- rl
'O (l)
H P-t 600.,,.,.,........ ~..,..,........,.,..,.,...,.,.,.,.,,...,....,,.,.....,...,..,.,....,.,.,.,.,.,..,...,.,.,.._"_._.......,.,"Tii-ni'TT+nrrr:m,TITIITTTlrrrmmmTITTrmiTITlmTimTIITTITTrr-1r-T~~~t'.'nrT.f':TTT'ITl'"MTll'TlTnfflTTTITITlM'T'ITI 4 00*++-1-+.. HI-H 200 100 -
6(]
.27% Cu 40 Reg. Guide 1. 99
_,_._,_,..,_,H+I H+l+H-1-l1/2IIIHHHi+H+iHff+HI-Hl+IHHIHt- ~
Reg. Guide 1.99 Upper Limit Code:
0
- b.
-k- - -
Plate B4lt06-3, Trans.
Plate 84406-3, Long.
Weld Metal IIAZ Material Reference Mat erial 20~-~-1,4+~14+1,1~um,.U-1.,1i.u,UUJW,,1,UW.WW,1,1...-'-J........... ~-'-'-t-M..i..w~*~i...uJW,!.1.1,U,1.1.'Wi-U-l-141.1.-U-LU~I.IWJIUL...-l-----..,_----------'
2 X 1017 1018 1019 N1!utron Fluence, n/cm2 (E > 1 MeV)
FIGURE 9.
EFFEC'f OF NEUTRON FLUENCE ON RTNDT SHIFT, DONALD C. COOK, UNIT NO. l
1L ~-?l/
?ABLE XII PROJECTED VALUES OF RTNDT FOR DONALD C. COOK UNIT NO. 1
(
EFPY P.V. Mat:erial Locat:ion 12 Place B4406-3 I.D.
i l/4T 3/4T HAZ Material I.D.
t l/4T 3/4T Weld Metal I.D.
i l/4T 3/4T 32 Plate B4406-3 I.D.
I 1/4T 3/4T HAZ Material I.D.
I l/4T 3/4T Weld Metal I.D.
t l/4T 3/4T (a)
Neutrons/cm2, E > l,MeV.
(b)
Reference 17.
Initial RTNDT 45 °1(b) 45°F 45°F
-60°F(b)
-60°F
-60°F oop(c) 0°F 0°F 45°F(b) 45°F 45°F
-60°F(b)
-60°F
-60°F.
OoF(c) 0°F 0°F (c)
Estimated per Reference 17.
- 40.
FluenceCa)
~RTNDT Adj* RTNDT 7.9 X 1018 125 170 (w.~. I /I p *911 P~.oo9 P*
(
4.8 X 1018 100 145 1.2 X 1018 60 105 7.9 X 1013 175 115 4.8 x 1018 145 85 L2 x 1018 110 50 7.9 X 1018 185 185 Cl.LL::, "J...°Y f*S 4.8 X 1018 155 155 p = 'o~.3 f-1 1.2 X 1018 115 115 2.1 X 1019 205 250
(
1.3 X 1019 160 205 3.2 X 1018 85 130 2.J. X 1019 265 205
- l. 3 x 1019 210
. 150 3.2 X 1018 135 75 2.1 X 1019 305 305 1.3 X 1019 240 240 3.2 X 1018 140 140
A method for estimating the reduction in Cv upper shelf energy as a function of neutron fluence is also given in Regulatory Guide 1.99, Revi-sion 1 [8].
The results from Capsules T and X are compared to a p*ortion of Figure 2 of the Regulatory Guide 1.99, Revision 1, in Figure 10.
The shell energy responses of the pressure vessel surveillance materials from
. Capsule T and all the materials but the reference steel from Capsule X are in good agreement with the prediction of Regulatory Guide 1.99, Revision 1.
Refen-ing
- to the conservative NRC design curves from Regulatory Guide 1.99 shown in Figure 10, the projected Cv shelf energies of the vessel materials are as follows:
Plate B4406-3 (Unirradiated £i Shelf
- 96 ft-lb) 32 EFPY at I.D. -
69 ft-lb 32 EFPY at l.4T-72 ft-lb 32 EFPY at 3/4T -
78 ft-lb Weld Metal (Unirradiated C.,,. Shelf
- 110 ft-lb) 32 EFPY at I.D.
59 ft-lb 32 EFPY at l/4T 64 ft-lb 32 EFPY at 3/4T 77 ft-lb HAZ Material (Unin-adiated Cv Shelf* 120 ft-lb) 32 EFPY at I.D.
32 EFPY at l/4T 64 ft-lb 70 ft-lb 32 EFPY at 3/4T -
84 ft-lb These projections indicate that the core beltline materials in the Donald C.
Cook Unit No. l pressure vessel will retain adequate shelf toughness through-out the 32 EFPY design lifetime.
41
60 rl'1"TTTMTt1TnfflTlrltmttm1t1'1'itTITmtl'T1Tl'ITl'ffl1rm-~-rrr-,..,..,l"T1"TTTTTTTTTlrmiTrmmmmmTtTMTT'.rffnfflTTmm~~'T.=r"T:.'!:='rT'l=t"'l~'l"m"lffTTTfflTffffl_.......
- -- 1--- ~ *-
~ 1- -
40 N
~
H 20..._._.-1-+-1-,........... -H-H++I+ Reg. Guide 1. 99 (lJ Ji
'H
.-I (lJ
~
Ul A
- rl
~
10 IH-Hd-H-Hft-l-ti~H~IH!-Hmlltl+l+H-H-HilHHHHlil~ll-::-1-:+,.,li-+--1+.hHf.+.F.~+H
.po
1 MeV)
FIGURE 10.
DEPENDENCE OF Cv UPl>ER SHELF ENERGY ON NEUTRON FLU ENCE* DONALD C. COOK, UNIT NO. 1
(
The cur-rent Donald C. Cook Unit No. 1 reactor vessel surveillance program removal schedule, revised to conform to ASTM E 185-79 [18], is summarized in Table XIII.
There are six capsules remaining in the ves-sel, of which three are standbys.
43
TABLE XIII REACTOR VESSEL SURVEILLANCE C.A.PSUlE REMOVAL SCHEDULE [16]
DONALD C. COOK UNIT NO. l WOL Removal Equivalent Vessel Capsule Material Time Fluence T
Weld Metal (a) 4 EFPY at I.D..
X Trans. Plate (b) 11 EFPY at I.D.
y Weld Metal 5 EFPY E.O.L. at l/4 T u
Weld Metal 9 EFPY E.O.L. at I.D.
s Trans. Plate 32 EFPY E.O.L. at I.D.
V Trans. Plate Standby w
Trans. Plate Standby z
Weld Metal Standby (a)
Removed at 1.13 EFPY of operation (3.4 EFPY equivalent vessel fluence).
(b)
Removed at 3.48 EFPY of operation (10.4 EFPY equivalent vessel fluence).
44
VI.
HEATUP AND COOLDOWN LD1IT CURVES FOR NORMAL OPERATION OF DONALD C. COOK UNIT NO. 1 Donald C. Cook Unit No. 1 is a 3250 Mwt pressurized water reactor operated by American Electric Power Service Corporation.
The unit has been provided with a reactor vessel material surveillance program as re-quired by 10CFR.50, Appendix H.
The second surveillance capsule (Capsule X) was removed during the 1980 refueling outage.
This capsule was tested as described in earlier sections of this report.
In summary, these test results indicate that:
/
/(1)
The RTNDT of the surveillance materials in Capsule X in-creased a maximum of 165°F as a result of exposure to a neutron fluence of 6.2 x 1018 neutrons/cm2 (E > 1 MeV).
v(2)
Ba.sed on an analysis of the dosimeters in Capsule X and those of Capsule T (removed in 19i7) the vessel wall fluence at the I.D. was 2.3 x 1018 neutrons/cm2 (E > 1 MeV) at the time of removal of Capsule X.
of operation was predicted to be 155°F at the l/4T and 115°F at the 3/4T vessel wall locations, as controlled by the core beltline material weld metal.
(4)
The maximum RTNDT after 32 E:FPY of operation was predicted to be 240°F at the l/4T and 140°F at the 3/4T vessel wall locations, as controlled by the core beltline weld metal.
The Unit No. 1 heatup and cooldown limit curves for 12 E:FPY and 32 EFPY have been computed on the bases of (3) and (4) above.
The procedures employed by SwRI are described in Appendix B.
45
'l'he following pressure vessel c:oustants were employed as input data in this analysis :
Vessel Inner Radius, r1 86.50 in., including cladding Vessel Ou1:er Radius, r 0 95.34 in.
Operating PTessure, P0 2235 psig Initial Tem;,erature, ~
70°'F Final Tempera-cure, Tf 550°!
Effective Coolant now Ra.ta, Q 135.6 x 106 l~/hr Effective Flow Area, A.
- 26. 72 f-t2 Effective Rydraul.ic: Diamecar, O
- l5. OS in.
~eacu;, curves were c:cmputad for a heatup rate of 60°'F/hr.
Si:lce lower rates tend to raise r:he c:urve in the c:enttal region (see Appendix B), t!lese c:urves ap-ply to all heating r:a.tes u;, to 60 °F /hr.
Coolci0W'C1 curves wen com-puted for c:ooldova. rates of 0°:/hr (steady state), 20°:/br, 40°:/hr, 60°:/hr, f,
and 100°:/hr. !he 20°F/hr curve would ap-ply to cooldOW"ll races up to 20°F/hr; the 40 °'F /b.r curve "JOuld ap-ply r:o rates u;, to 40 °'F /hr; oe '60."! /hr curvt: w-ould apply to rates up to 60°:/b.r; the 100°7/hr curve ~ou.ld apply to rates u;, to 100°7/hr.
!he unit No. 1 hea~ and c:ooldowu curves for u;, to 12 En"! are given in Figures ll and l.2; those for up to 32 E:FPY are given in Figuras'lJ and 14.
46
(.!)
(/)
0..
~
(/)
(/)
w
~
0..
ffi U)
(/)
.i:,--
~
§ f5 E
~
~
2600 llitU:H+/-HltttWl Ulllt.U llitl lltiJlttl Util ttiUHi.l REACTOR COOLANT SYSTEM HEHUP LIMITATIOOS AP-2400 PU CABLE FOR Fl RST 12 EFFl:CT I VE FULL PCl'l'ER YEARS.
(MARGINS OF 60 PSIG AND 10°F ARE IM-2200 2000 1800 1600 1400 1200 1000 CLUll::D FOR POSS IBLE I NSTRLMENT ERROR. )
ll~llll~lllllllllll~llllmlllllllllllllmlm~m1 111111111111111111111 IHllltl ti LEAK TEST LIMIT llffl MATERIAL PR(fERTY BASIS WELD METAL CU= 0.27%
INITIAL RTNDT = 0°F 12 EFPY RTNl)l (l/4T) = 155°F:
l.JJACCEPTABLE ACCEPTABLE *
(3/4T) = 115°f OPERATION OPERATIOO
- 1mmi~m111m1 llmlllml m
l PRESSURE-TEMPfRATURE CRITICALllY
. Ii LIMIT FOR HEATIJP RATES UP TO 60°F/HR LIMIT 800 I!
600 400 t
i 200 60 100 150 200 250 300 350 400 AVERAGE REA<:TOR COOLANT SYSTEM TEWERATURE ( 0 F)
F'IGURE 11.
REACTOR COOLANT SYST*~M PRESSURE--TEMPERATURE LIMITS VERSUS 60"F/H0UR RATE CRITICALITY LIMIT AN) HYDROSTATIC TEST LIMIT, 12 EFPY 450
2600 111111111111 m 11111111 rm 11111111111111111111111111111111111111111111111111111 n Im 111111111111 REACTOO COOU\\NT SYSTEM COOL.l)(MN LIMITATIOOS 2400 APPLICABLE FOO FIRST 12 EFFECTIVE FULL PMR
- YEARS, (MARGINS OF 60 PSIG AND 10°F ARE IN-2200 CLUDED FOO POSSIBLE INSTRLM:NT ERROR,)
G -
(/)
2000 e::.
~
1800
~
~
1600 a..
rn 1400
(/)
~
1200 "-
(X)
§ 1000 lllllllllllllttlllWillllllllllllltllllllllWlllillllllllllll
~II j
MATERIAL PROPERTY BASIS WELD METAL CU= 0.27%
- UNACCEPTABLE INITIAL RTNDT = 0°F EFPY RTNDT (1/4T) = 155°f OPERATIOO (3/4'1') = 11s°F llil1Malill111111 ACCEPT ABLE.
OPERATIOO.
PRESSU<E-TEMPE~Tl.RE LIMITS g
800 u
u) 600 0::
coo~
I RATE °F HR 0
20 400 40 60 I
100 200
- II 111111 60 100 150 200 250 300 350 400 450 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (°F)
FIGURE 12.
REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS VERSUS COOLDOWN RATES, 12 EFPY
r-l!)
U) 0... -
w er
(/)
(/)
~
0...
in l-(/)
. (/)
~
+".
\\D 0 8 ts i 2600 2400 2200 2000 1800 1600 1400 1200 1000 800 600 400 200 60 1111rJ11111111111111111111111111111111111111111111111111111111111111111111111 II 11m HIii ii 11 11111111 REACTOR COOLANT SYSTEM HEATUP LIMITATIOOS AP-
- PLICABLE FOR FIRST 32 EFFECTIVE FULL PCMER
- YEARS, (MARGINS OF 60 P!ilG AND 10°F ARE IN-CLUDED FOR POSSIBLE INSTRLMENT ERROR.)
LEAK TEST LIMIT 1111111111111111 l lltl I I I I I I IIIII II II I II Ill I II I I II Ill l11111111111 MATERIAL PROPERTY BASIS WELD t-ETAL CU= 0.27%
INITIAL RTNDT = 0°F EFPY RTNDT (1/4T) = 240°F (3/4T) = 140°F 111111 UNACCEPTABLE OPERATION PRESSUF!E-TEMPERAllJRE LIMIT FOR HEATUP RATES UP TO 60°f/HR I
I 100 150 200 250 300 350 AVERA(,iE REACTOR COOLANT SYSTEM TEMPERATURE ( 0 F)
I I
ACCEPT ABLE I Cf'ERA Tl 00 :
ITtn*_..........,,.,. *.,.,..-*.,....-..,-,
CRITICALITY LIMIT
' i i
400 450 FIGURE 13.
REACTOR COOLANT SYS'l'EM PRESSURE-TEMPERATURE LIMITS VERSUS 60°F/HOUR RATE CRITICALITY LIMIT ANO IIYUROS'l'ATIC TEST LIMIT, 32 EFPY
l!)
(/)
a_
~
- J
(/)
(/)
~
a_
~
(/)
§ VI 0
§
~ ;
2600 2400 2200 2000 1800 1600 1400 1200 1000 800 600 400 200 60 1*IIIIIIIIIIIIIHlllllllll,,, ****,,,,,,,,,,,,,,,,,,1111nllllltlllllltltlltl,,1,1111n11.
REACTOR COOLANT SYSTEM COOlDO'-IN LIMITATIOOS APPLICABLE FOR FIRST 32 EFFECTIVE FULL Pa,.JER YEARS.
(MARGINS OF 60 PSIG AND l0°F ARE IN-CLUDEO FOR POSSIBLE INSTRUMENT ERROR,)
II II II I II I II I Ill I I II Ill I I I I Ill llltl I Ill I I I I I I I I II Ill I I I Ill II I I II MATERIAL PROPERTY BASIS WELD METAL CU= 0.27%
INITIAL RTNOT = 0°F EFPY RTmr (1/4T)' = 240°F UNACCEPTABLE (3/4T) = 140°f OPERATIOO 11!1 PRESSURE-TEWERATURE ACCEPTABLE LIMITS OPERATIOO COOL.l)GIN RATE °F/HR.
0 20 40 60 100 100 150 200 250 300 350 400 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE ( 0 F)
FIGURE 14.
REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS VERSUS COOLDOUN RATES, 32 EFPY 450
l VII.
REFERENCES L
Title 10, Code of Federal Regulations, Part SO, "Licensing of Production and Utilization Facilities."
- 2.
ASME Boiler and P-ressure Vessel Code, Section !II, "Nuclear Power Plant Components," 1974 Edition.
- 3.
AS'l'M E 208-69, "Standard Method for Conducting Drop-Weight Test to Deter.nine Nil-Ductility Transition Temperature of Ferritic Steels," 1975 Annual Book of ASTM Standards.
- 4.
Steele, L. E., and Serpan, C. z., Jr., "Analysis of Reactor Ves-sel Radiation Effects Surveillance Programs," ASTM STP 481, December 1970.
S.
Steele, L. E., "Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels,lt International Atomic Energy Agency, Technical Reports Series No. 163, 1975. *
- 6.
ASME Boiler and P-ressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1974 Edition.
- 7.
"Prediction of Shift in the Ductile-Brittle Transition Tempera-ture of tWR Pressure Vessel Materials," prepared by the Metal Properties Cotm.cil, Subcommittee 6 on Nuclear !iaterials, July 1, 1980.
- 8.
Regulatory Guide 1.99, Revision 1, Office of Standards Develop-ment, U. S. Nuclear Regu.Latory comm.iss1.on, apri:: l':Ji7.
- 9.
ASTM E 185-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," 1975 Annual Book of ASTM Standards.
- 10.
ASTM E 399-74, "Standard Method of Test for Plane-Strain Frac-ture Toughness of Metallic Materials," 1975 Annual Book of ASTM Standards.
- 11.
Witt:, F. J., and Mager, T. R., "A Procedure for Determining Bound-ing Values of Fracture Toughness Krc at Any Temperature," ORNL-TM-3894, October 1972.
- 12.
"American Electric Power Service Corporation Donald C. Cook Unit No. l Reactor Vessel Radiation Surveillance Program," WCAP-804 7, March 1973.
- 51
- 13.
Letter. F. Noon of ~estinghouse to J. R. Jensen of the American Electric Power Service Corporation, Document AEP-80-528, March 19, 1980.
- 14.
Norris, E. B., "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. l; Analysis of Capsule T," SwRI Report 02-4770, December 8, 1977.
- 15.
McElroy, W. N., et al, "Analysis of Interpretation of Nuclear Surveillance Results," Irradiation Embrittlement, Thermal Anneal-ing and Surveillance of Reactor Pressure Vessels, International Atomic Energy Agency Report: IWG-RRPC-79/3, Vienna, Austria, Feb-ruary 26 to March l, 1979, pp. 144-173.
- 16.
Donald C. Cook Unit No. 1 Technical Specifications.
- 17.
US NRC Standard Review Plan, NUREG-75/087, Section 5.3.2, Pressure-Temperature Limits, November 24, 1975.
- 18.
ASTM E 185-79, "Standard P-ractice for Conducting Surveillance Tests for Light-W~ter Cooled Nuclear Power Reactor Vessels," 1979 Annual Book of ASTM Standards.
52
AHl'ENDIX A T.ENSII.E TEST RECORDS 53
Test No. T------
Sou1:hwest Research Institute
)
Department o£ Material~ Sciences TENSII..E TEST DATA SHE.ET Est. U. T. S. _____ psi Initial G. L. _....,f,_O __ o __ o ___ in.
Machine *No.
2.2 r<, 2 Spec. No. 1J - 8 Temperature 1 G:.o °F Initial Dia.
- .249
in.
Date
/ 2 -;, - 9o Strain Rate
- OOS°JN/,,4..,,;_!IJ.itial Thickness in.
!IJ.itial Area.
1 0 !+ 81') i1 t Initial Width _____ in.
-Top Temperature OF Maximum. Load 4Eoo Bottom Temperature OF
- 0. 2% Of!set Load Final Ga.ge Length 1,243 in *
- 0. 02% Of£set Load Final Diameter
- 1 Y-1 in.
Upper Yield Point Fmal Ai-ea U. T. S. =
0.2o/o*Y.S.
O. 02% Y. S.
Upper Y. S.
.QI 1.29 2
- m.
Ma.xim:~ Load Initial Area
=
0. 2% Offset Load
Initial Area
=
- 0. 02% Offset Load Initial Area
= UE-oer Yield Point Initial Area
=
=
psi
_____ psi
% Elongation = Final G. L. - Initial G. L. x 100 =
'2, 4. 3>
Initia,l G. L.
% R.A. = Initial Area - Final Area x 1 00 =
~ 5 3 Initial Area lb lb lb lb
~
I}
~
I
~
- . = :_--: -
r -
- ':)-*-*
I* 'r'
-=:::..i.::
(
j
Southwest *Research. Institut~
Department of Material.a Sciences TENSU.E TEST DATA SHEET rl
- {o. T-C Spec. No.
AT q.
Est. U. T. S.
psi Initial G. L.
I, ooo in.
Project No. Q~ - C:, ( ~~
Machine No.
- 2. ~ K, ~
!'emperatu:re 2.S-O °F Initial Dia. --*
_2._~_q __ in.
Initial Area.
, 0 LJ. g '1, *N '?o Initial Width.
in.
Top Temperatu.re OF Ma.ximum Load 411~~
lb Bottom Temperature OF O. 2% Ofiset Load
'31j 64 lb Final Gage Length Lr qry in.
- 0. 02% Offset Load lb Final Diameter
- I l:,Z in.
Upper Yield Point lb Final Area.
oz.oc.,
2
- m.
-t Ma.xim:.im Load U. T. S.
1 Inltia Area
=
O. 2% Y. S. = 0. 2% Offset Load =
- Initial Area O. 02% Y. S.
- 0. OZ% Off.set L~ad =
=
Initial Area psi Upper Y. S.
Up-oer Yield Point
=
=
Initial Area psi Final G. L. - Initial G. L. x l OO __
/ q 7 a1.0
% Elongation =
___. ___ -1(
Initial G; L.
Initial Area - Final Area x 100 =
% R. A. =
J.nitial Area
t c:--
1-
<(_
j....
~
I
~
I i
!~
12 I
I I i""' -.
("'I C
~
I
-*-~
I *."..
C
-*-- "=:t--*~
~f;*o.*~~0,,L,)~:(l)l
(
l r
I
=:1131119*~
QM:Mt,lt,, ll lJI
- t,1:::i,,.t-. OJ,.Ot :( Ol r -
Southwest Research wt:itute Department oi Materials Sciences TENSILE TEST DATA SHEET Test No. T-3 Spec. No.
W-'7 Est. U. T. S. _____ psi Initial a. L. ___ / __, __ o __ -e __ ~_ in.
Project No. Q--Z.. -/;:,/ S'f Machine No. 2:2 K,;>
-:s...
Temperature ;;,
0 °F Initial Dia. __
, 2_~~-- in.
Date
/ 2... 4 :,
St:rain Rate,Oo~;"i,,,,,/,.,.,,._ Initial Thickness ___ in.
Initial Area 1 0 49 ' tN~
Initial Width _____ in.
Top Temperature OF Maximum Load -1+~~4 Bottom Temperature OF O. 2% Offset Load Final Gage Length
,\\,q{g Final Diameter
,\\ {ps' Final Area U. T. S. =
I QZ I 'i:
Ma.xim.:.un Load Initial Area in.
O. OZ% Ofiaet Load in.
Upper Yield Point 2
- m.
=
O. 2% Y. S. = O. 2% Ofiset Load = 2-c '~ \\
Initial Area.
- 0. 02% Y. S. = O. 02~ Ofi'set Load =
Initial Ar ea.
psi Upper Y. S.
~ UE;Eer Yield Point =
Initial Area.
_____ psi a,_
l Final Q. L.
- Initial G. L. x l OO = \\ q_ \\o a,_
1o E ongation =
Initial Q. L.
, __,o
% R. A.
= Initial Area
- Final Area x 100 =
Initial Are a Signature\\,~~
lb lb lb lb
(
.. ~
(
I l
I,
---=
~="
ti C
11111 IE*
=
(-
'),:'
- r. -
~
- Southwest Reselich Instimte Department of Materials Sciences TENSILE TEST DAT.A. SHEET 1(. No. T-i Spec. No. A,... S Temperature sS'7J °F Est. U. T. S. _____ psi Initial G. L.
I( 0 oo in.
Project No. OZ- '1 I ~=f Machine No. 22 k1P Initial Dia.
, 2 4-'f
________ llle Strain Rate,oo~--,,..(,,\\1/;,_ Initial Thickness in.
Initial Area
, 0% 'l ir1' Initial Width _____ in.
Top Temperature ______ °F
~
Ma.ximwn Load ts: 9{,sc, lb Bottom Temperature _____ °F
- 0. 2% Offset Load lb Final Gage Length. ___ l __,-_\\ _9_4-__ in.
0.02% Of:fset Load°'
lb Final Diameter
, l t,t, in.
Upper Yield Point ____ lb Final Area __
...... Q_-Z._\\_C, _____ in. 2 U. T. S.
Inltia 1.Area
=
- 0. 2% Y. S.
O. Z% Offset Load
'"10,C\\ri
=
=
Initial Ar ea
~Ks; O. OZ% Y. S. =
- 0. 02% Of:fset Load Initial Area
_____ psi Upper Y. S. = UP;Eer Yield Point
.Initial Area psi a7. Elonga-:...... -
Final G. L. - Initial G. L. - l 00 =
I~. 4 a,_o 10
.Initial G. L.
7<
% R.A. = Initial Area - Final Area x 1 00 =
..; {. L, S %
I:c.:i:tial Area
i I !
-1 I i
i I
~
\\
II
\\.
-I ~
~
i! ~
I
.:*-:*,****F*-*
~
APPENDIX B PROCEDURE FOR THE GENERATION OF ALLO~ABU:
PRESSURE-TEMPElimRE LIMIT CURVES FOR NUCUAR POWER PI.ANT REACTOR VESSELS 67
PROCEDURE FOR THE CENEllATION OF AL.I.OW.A.BU:
PRESSURE -TEMP Ell.A TUltE I..IMIT CURVES FOR NUCLEAR. POWER PLANT REACTOR V'l:SSELS A.
Introduction The following i.s a. description of the ba.1i..s far the generation oi pr-essure-tem-pe.rature limit c~s £or in.service leak a.nd !lydrostatic te st:s? b.ea.m:p a.nd cooldowu operations, and c or-e operation o£ re a.ct or.
pressure 7'!ssels. The sa..fety margins em:ployed in these procedures equal or e.xc:eed those recommended i:n the.Aasz.4.E Boiler and Pressure Vessel Code,Section III. Appendix C, "Protection Against Nonduc:tile Fa.ilure.
- a.
The basic: pa..~eter tUed to determine sa..fe 'Tes sel ope.rational c0ndit:i0n.s is the stress intensity fa.ctor. Kr, wilic:l:f is a. function oi the stress.state a.lld £la.w con:figura.:t:ion. The KI c:or-respondinJ to membrane tens ion is given by we.ere ~
is t:ie mem.bra.ne sn-ess cor'?'eetion fa.ctor for the pcstw.atad flaw a..nd o-m the membrane sn-es.s.
Likewise, KI c:orTespondmg to bend-ing is given by (l) where Mb is the bending suess c:orreetion factor and a-~ is the bending
(
sttess. For 'Tessel section thickness 0£ 4* to 12 inches, the ma."'imum
postulated surface flaw, which ia assumed to be normal to the direction of maximum stress, ha..s a. depth oi 0. 25 of the section thickness a.nd a.
length of l. 50 times the section thickness. Curves for Mm versus the square root 0£ the vessel wall thickness for the postulated flaw a.re given in Figure l a.s taken from the Pressure Vessel Code (re£. Figure G-2114. 1 )..
These curves a.re a. function of the stress ratio pa:ameter a-/ <f'v-, where a-
'I is the material yield strength which is ta.ken to be 50,000 psi. Th.e bending correction factor is de£ined a.s 2/3 Mm and is thereiore determined from Figure 1 a.s well.
Tb.e basis !or these cu:ves is given in ASME Boiler and Pressure Vessel Code,Section XI. "Rules for In.service Inspection of Nu*
clear Power Plant Components, Article A-3000.
The Code specifies the minimum Kr that can cause failure a.s a. func*
tion oi material temperature, T, a..nd its reference nil ductility temperature, RTNDT* This mini.mum KI is defined a.s the reference stress intensity fac~
tor, Km., and is given by Km, = 26777. + 122.3. exp [ O. 014493(T - RTNDT + 160)]
(3) where a.ll temperatures a.re in degrees Fahrenheit. A plot o:£ this e.:cpression is given in Figure 2 taken from t.b.e Code (re:f. Figure G-2 010. 1 ).
- c.
Pressure-Temoerature Relationsb.i:,s In.service Leak a.nd Hydrostatic Test During performance oi inser.,.ice leak and hydrostatic tests, the reference st:ress intensity factor, KI.R, must always be greater than
FIGURE l. STRESS CORRECTION FACT OR
170 IGO 150 I~
130
!20 110
!~ !CO
'! 90
~. so
- ic: 70 60 so 40 30 20 10 0
I L
.,_..-~--.,
(l": 1R-2S.777) 21.223e0.01493(T-HtTNOT-160l)
\\":HE:liE
!(!R
- f<c:FtRENCE STRESS tNTI:::NSITI FACTOR T
- TI:MPERATURE AT \\'/HICH l<IR I
I IS PERMITTED, *F I
I RTf.:DT
- REFERENCE NIL-OUCilllTY TEMPERATURE.
I I
/
/
/
I - l---"'"
I I
I I
i I
I i
I I
I ii I
I I l I I
I/
Ii I
I i
l I
I I
-240 -200 -!GO
-120
-eo
-40 0
40 80
.120 IG0 200 240 FIGURE z. REFERENCE STRESS IN'TENSlTY FACTOR
- l. S times the KI c:a.u.1ed by pressure, thua 0%'
For a. c:yl.inder with inner ra.ciiiu ri a.nci ou.ter ra.diu.s r o*
the stress dist:-ibu.tion due tc internal pres sure i.s given by With l/4T flaws possible at both inner and. ou.ter radial lcca.ticn.s, i.e.,
(4)
(5)
(6) at r 114 = ri + l/4(r0 - ri} and. r3/4 = ri + 3/4(r0 - ri.), the maximum stress (
will occur a.t the im:.er flaw location. th.us With the Ol'er-aticn pressare k::c.own, i.e.* P 0
- we_deter-rniz::i.e the minimum coolant b!mpera.ture that will sati.siy Equation (4) by ev-aluating Km = l. 5 Mm_ rrtna.x and determine the c:orresponclio.g c:oolaJ:it temperature, T, f:rom Equa-tion (3) for the given RTNDT a.t the l / 4T location. For this calc~latio:c.,
Zqua.tion ( 3 ) takes the form.
(7)
(8)
I g
[Krg - 26 7i7. 1 T = RTNDT{l 4T) - 160. +08. 998 1n 1223 _
j.
(9)
C
(
The inservice curves are generated !or an operating pres*
sure raJ1ge of. 96 P 0 to 1. 14 P 0, where P 0 ia the design operating pressure.
- z.
Heatup and Cooldown Operations At all times during heatup and c:ooldown opera.tiona, the ref-erence stress intensity factor., Km,, mast always be greater than the sum.
of Z times the K1p caused by pressure and the K1t caused by thermal g*ra.-
dien.ts. thus
- z. 0 Krp + 1. 0 K1t < Km or wh.ere <rma.x is the m.a.ximwn allowable stl'ess due to internal pressure, and Ku is the equivalent linear stress intensity factor produced by the thermal gradients. To obtai:l the equivalent linear stress intensity fa.c-( l 0)
{11) f tor due to therm.al gradients requires a detailed thermal sttess an.alys is.
The details of the required analysis are given in Section o.
During b.eatup the radial stress distributions due to internal pressure an.d therm.al gradients are shOW'1'1 schematically in Figure 3a.
Assuming a possible flaw at the 1/4T location., we see from Figure 3a.
that the thermal stress tends to alleviate the pressure stress at th.is point in the vessel wall and, therefore, the steady state pres sure stress would represent the max:imum stress condition at the 1 /4T location. At
OUTER RADIUS.
3/4T
+
114T INNER RADIUS Pressure stress distribution n, ernia I stress distribution
( a ) Heatup OUTER RAD l US 3/4T l/4T INNER RAD I US Pressure stress distribution Th ermai stress distribution
( b ) Cooldown Figure 3. Heatup and Cooldown Stress Distribution
the 3/4T fla.w location, the pressure stress. and thermal stress add and, therefore, the combination for a given heatup rate represent., the maxi-mum stress at the 3 / 4T location. The maximum overall stress between the l/4T and 3/4T location then determines the maximum allowable reac-tor pressure at the given coola.nt temperature.
The b.eatup pressure-temperature curves are thus generated by calculating the maximum steady state pressure based on a possible flaw a.t the
- 1 / 4T location £:rom
(
where Mm is determined from the ctll"v*es i.t1 Figure l and Km. is obtained from Equation (3) using the coola.nt temperature and RTNDT a.t the l/4T
!..:,.;,;.4,~luu.
Hc.n: we may riote that Mm, must be iterated !or sillc:e*it *hr *a.
function of the final stress ratio to yield strength (rrlir-1 ).
At the 3/4T location, the maximum pressure is determined
!:rom Eq~tion { 11) as where Km is obtained from. Equation (2) using the material temperature a.ad RTNDT at the 3/4T location and Krt is determined from the analy-sis procedure outlined i.e. Section D. Mm is determined from Figu.:-e l.
(l:?)
(13)
Th.e minimum of these maximum allowable pressu..res a.t the give.n coola.nt temperature determines the maximum operation pressure. Ea.c:.h b.ea.tu-p rate o:f interest mu.st be a..na.ly-z;ed on a.n indirid*
u.a..l ba..sis.
Tb.e cooldowt1 a.na.lysi.s pr1::iceeds ill a. *imilar fashion a..s that described for heatup with the following e.xceptio:n.s: We note fr-o~ Fi~re 3b that durmg cooldown the l / 4T location always c-ont:=1::11..s the maximum saess sinc:e the thermal gradient prQduces tensile ste:sse:s a.t the l/4T location.
T!lu.s the steady state pressure is the same a..s that given in Equation (lZ). For each cooldo"Wt1 rate, the maximum pressu:re i.s eV'alu*
a.ted a.t the l / 4T location from where Km, i.s obtained from Equa.tio.n (3) 1.J..S:ing the material temperature and RTNDT a.t the l / 4T J.ocaticn.
Kn: i.s determined from the thermal a..nalysis described in Section 0.
It Ls of interest to.!lO"t!! th.at duri:c.g c ooldown the mate:ial temperature will lag the coola.nt temperature-a.nd, therefore, t!le steady state pressure, which is evaluated a.t the coola.nt temperature, will ini-tially yield the lower ma.xi.mum allowable pres su:re.
Wb.ell the tb.e rmal g:radie.nt.s* inc.rease, the st:'esses do likewise, and, finally-, t+/-::.e t:-ansient
( 14}
a.na.ly3i.s goveni.s t.b.e rna.xim.urn allowable ~:res sure. Ee.nc e a. ?OiJ:l.t-by- ~oi.nt
comparison must be made between the maximum allowable pres su.res proc duced by steady state analyses and transient thermal a.nalysia to determine the minim.um of the maximum allowable pressures.
- 3.
Core Ot>e:ration At a.ll times that the reactor core is critical, the temperature must be higher than that reqtlll'ed for in.service hydrostatic testing" and in addition, the pressure-temperature relationship shall provide a.t least a.
40 °F margin over that required for b.eatup a.nd cooldown operations. Thus the pressure-temperature limit cu~s for core ope.ration may be constructed.
directly from the mservice leak and b.y-drostatic test a.:c.d h.ea.tup a.nalys is 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 m the vessel wall is governed by the partial differential equation
( l S) subject to initial condition T(r, 0) = T O,
(16) and boundary-conditions
( 1 7)
a.nci where p i.1 the ma.teri&l density, c the ma.taria.1 speeific heat, K the heat condu.c:-
tivity of. the material, h the heat tra.n.sier c:oei:ficient betw~en the water coolant a.nci vessel material. R the heating rate, T O the irlitia.l c:oola.nt temperature, T(:r, t) the temperature distribution in the ~ssel9 r the s-pat:i.al c:oordina.te, anci t the temporal. coorci:m.ate.
A finite difference solution procedure bs employed to solve for the radial tempera.tu.re distribution a.t 'l'a.riows time steps a.l0ng the b.eatup or coclciow-::t :ycle. The finit!! differ~n-:e eq_t.~tion~ fo?'_ N :-adia.l poic.t3, a.t di.stance Ar a.pa.rt, a.cross the ~ssel a.re:
£or 1 < n < N for n = 1 6.t: K r,\\-
+
{,\\.. )2
( 1 + ::.:.. )
~c.....
r1
( 18)
( 19)
(ZO)
(Z 1)
and for n = N For stability in the finite difie:rence operation. we must choose At !or a. given Ar such. that both and a.re satisfied. These conditions assure u.s that heat will not !low in the direction of increasing temperature, which. of course, would violate the second law of thermodynamies~
Since a. large variation.~ -~oo~ temperamre is considered9 the
\\,.Ji'\\,,\\',-,*~,~--'----
dependence of. (K/pc),) K. and hon temperature is included in the analysis by treating these a.s constants only during every 5 °F increment in coolant
{ZZ)
(23)
(24}
temperature a.nd then updating their values for the next 5 °F increment.
The dependence of (K/ pc) called the the rm.a..l dii!usivity a.nd K. the thermal conductivity, can be determined from the ASME Boiler and Pressure Ves-sel Code, Section m. Appendix 1 e Stress Tables. A linear regression*
analysis of the tabular values resulted in the following expressions:
K(T) = 38. 211 - 0. 01673
- T (BTU/HR-FT- °F)
(2S)
and k(T) = (K/~c) = 0.6942
- O. 000432.
- T (FTZ/Hlt) where T i.1 in degrees Fa.b.renheit.
The b.eat t:-ans:fer coei:£icient ia calculated ba.1ed cm foreed con-vection tmder mzbulent flaw conditions. The variables m.vo.lved a.re the mean velocity of the fluid cool.a.at, the equiva..lent (b.yd:ra.ulic) diameter of (26) the coolant channel, a.a.d the density., b.ea.t* ca.pa.c:i:y, V'iscoaity, and thermal c:ondw:tivity of tne cool.am:. For water c:00.La.ut, allowance for the v-a.riationa in physical properties with tempera.tu.re may be ma.de by WT'iting*
where -, is :in ft/ sec, 0 in inches, the temperature is in *F, and b. is in rela.tia.a.s.b:ip a.re in good a.p-e4ment with those obta.med f:rom tile Oitm.1-Boelter. equation for temperamr1ts u.p to 600 *F. The mean velocity of the c:oola.:t, v, is generally. given in terms of the effective coolant flow ra.te Q
( Lbm/ b.:r) a.nci effective flow area. A (ft2 ).
Given the relations b.i-p for the density of water a.s a. function of tempera.tu:re, the mean V1!locity o:f the coolant i.s o bta.ined from v- = Q/(3600
- o (T)
- A) *
- Gla..ssto.c.e, S., Principles o:f Nuclear Reactor Eng+/-c.ee.ring1 O. Va.n Nosn-and Co., Inc., New Jersey, pp. 667-668, 1960.
(28)
The thermal stress distribution is calc:u.la.ted from.
wl:iere a. is the coefficient 0£ thermal e.xpanaion (in/in *r). E is Young1 s modulua. and v is Poisson's ratio. This expression can be obtained from Theorv oi E.la.stic:ity by Timosh.enko and Goodier, pp. 408-409, wb.en im-
,(30) posing a zero radial stress condition at the cylinder inner and outer radius.
Pois~on' s ratio is ta.ken to be constant at a value of O. 3 wb.ile a. and E a.re evaluated as a func:t:ion of the average temperature across the vessel (31)
The dependence of the coefiic:ient of thermal expansion on temperature is a.(T) = S. 76.x: 10-6 + 4. 4.x: 10-9
- T (32) and the dependence of Young's modulus on temperature is ta.ken to be E(T> = z1. 9142 + z. s1sz x 10-4
- T - 6. s1z3 x 10-6
- T 2 (33) as obtained from regression analysis of tabular values given in Section Ill.
Appendix I of the AS:ME Boiler and Pl'essure Vessel Code.
The resulting stress distribution given by Equation (30) is not linear; however, an equivalent linear stress distribution is determined l ':'Om the :resulting moment. The moment produced by the nonlinear
sues s distribution is given by r o M(t) = bf a-T (r, t) rdr
?'i where b i.s a. writ depth oi the ~ssel. Here we note that the moment i.s a.
functicn of. tim.e, i.e., coo.J.a.nt temperature ri.a. Tc = To + Rt. For a. lin-ear sues s di.st:-ibu:ticn we have that Mc a-max= T wb.ere a-max i.s the ma.ximwn outer fiber stressy c the dista.nc:e from the
- i.eut:'al a;.x.u, taken to be (r 0 - ri)/Z, and I the section a.rea moment 0£ inertia. wbic:.b. i.s given. by Combinittg these expressions resulb in the equivalent linear st::'ess due to thermal gra.dient.s The thermal st:-ess intensity factor Krt is then deiined a..s wb.e.re Mb i.s determined from. the curves given in Figure 1 wherein M:, = 2/3 ~-
It is oi interest to :iote that a sign change occur!! in ehe
.3t::'ess calculations du.ring a cooldown a.nalysi.s since the the.rrr..a.l gradients
{35)
(36)
(38)
C produce 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* 1:ieatup a.re given in Figure 4. The radial temperature is shown normalized with respect to the average temperature, T avg* by T
- Tavg T=------------
(T - Ta.vg>max The thermal sa-ess and equivalent linearized sa-ess, a.s calculated by Equationa (30) a.nd (37), are normalized with respect to the maximum thermal stress. Here we note that the acmal therm.al stress at the 3/4T location is considerably less tha.n tb.e maximum equivalent linear stress
-wb.ic:h rielda additional s.uttty margins during the heatup cycle. Similar temperature and thermal stress dbtributions are developed during cool-dowu. The trends a.re nearly identical a.a those sh.own in Figure 4 when the inner a.cid outer vessel locations a.re reversed with the l / 4T location becoming the critical point.
E.
Ex.a.mole Calculations
{39)
The following e.xample is based on a. reactor vessel with the follow-ing characteri:stics:
Inner Radius Outer Radius Operating Pressure
=
=
=
- 82. 00 in.
90.00 in.
2250 psig
OUTER WALL 1.0 0.8 0.6 0.4 0.2 0
-LO 0
-LO INNER WALL No rm a I ized tern peratu re distribution { ~TI~ T max )
0 LO Normalized stress distribution ( cr / a-max }
Ftgure 4. iypicaJ Normalized Temperature and Stress Distribution Du ring Heatup
Initial Tempera.tu.re
=
70*F (To)
Final Temperature
=
sso*F (Tf)
Effective Coolant Flew Rate
=
100 x 106 Lbm/hr (Q)
Effective Flow Area
=
zo. 00 ft2 (A)
Effective Hydraulic Diameter =
10.00 in.
(D)
RTNDT (l/4T)
=
zoo*F R';' NDT (3 / 4T)
=
l40*F In the thermal stress a.naly~is 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 (tempera.tu.re via T = To + Rt) steps w*Te required in the thermal aJ2alysi.:s for the lOO*F/hr heatup rate. The
(
results 0£ the computation a.re shown in Figures 5 through 9.
Figure 5 gives the reference st:resa intensity factor, K.m,. a.s a function of temperature indexed to RTNDT (l/4T.). For the steady state analysis, Km, is converted directly. to allowable pressure via Equation. 12.
During the heatup and cooldown thermal analyses the material tem-perature a.t the l/4T a.n.d 3/4T and thermal stress intensity factors Ku a.re required to compute a.llowable pressure via Equations ( 13) and ( 14). The material temperatures versus coolant temperature during the 100 °F /hr heatup a.nd cooldown analyses a.re given in Figure 6. These temperatures a.llow computation of the corresponding reference stress intensity factors, l
(3/4T) and Km (1/4T). Figure 7 gives the corresponding therm.al stress intensity factor at the 3/4T a.nd l/4T locations as a function of coolant temperat'.ire.
160 -
0:::
80
~
40 O~---..a-------'-------L-------'-----.L-----..1.-----__.
50 100 150 200 250 300 350 400 TEMPERATURE f *r)
Figure 5. Reference Stress Intensity Factor as a function of Temperature Indexed to RTNDT ( 1141)
l.L 4 -
u.J
~
=>
~
400 300 -
ffi 200 a..
- cf
~
_J w
U'l Vl w >
100
-- I00°f /HR HEATUP ( 3/41 location)
~ 100 °F / HR COOLDOWN ( l/4 T location )
o __________________
.a.,._ ___
--L-___
50 100 150 200 250 300 350 COOLANT TEMPERA TUR[ ( °F )
'figure 6. Vessellemperature at l/4T and<f/4T locations as a Function of Coolant Temperr~~ure
~
~*
~
1-i
~
l2r---..,........,.-.,,-....---------.,_....--,,------,---------------
10 -
8 -
6 4 -
2 -
~- 100 °F /HR HEATU P ( 3/4 T Location )
- -- 100 °f /HR COOLDOWN { 114 Location )
0--------------------,,.-&-----...__ __________ __
50 100 150.
200 250 300 350 COOLANT TEMPERATURE ' 0f) fi9u, u 7. Thermal Stress Intensity Factor at 3141,o,d H4l tocations as a function oi Cooiant 1f~ *** ~err,1aur1 J
l Figures 8 a.nd 9 demonstrate the constl'uction oi the allowable com-posite pressure and temperature curves for the lOO*F/hr heatup and cool-down rates. The composite curves represent the lower bound of the the,ma+
and steady state curves with the addition of margin.a of +l O *F and -60 psig
!or possible inst:rumenta.tion error. Figure 8 also shows the leak test limit, corrected for instrument error, a.s obtained from Equation {9 ).
The limit points a.re a.t the operating pressure Z250 psig a.nd a.t 2475 psig which cor*
responds to 1. l times the operatmg pressure. The criticality limit is also shown in Figure 8 a.nd is constructed by providing for a. 40 °F :margin over.,..
that required for heatup and cooldowu and by requiring that,the minim.um temperature be greater than th.at required by the leak test limit.
2400.
2000 _s
- 0)
V)
Cl. --
w 1600
~
=>
COMPOSITE CURVE -l00°FIHR ttEATUP I
Margins of +1o*r.and -60 pslg for Instrument error I\\
l/)
l/)
w
~
- i 0...
0
- 1200* -
~
STEADY STATE CRITICALITY:
LIMIT
<(
u a z -
800 HEATUP
- 400 50 100 150 200 250 300 350 400.
INDl'CATED TEMPERATURE C °F, flgura 8. Pressure-Tt)!ll()erature Curves for 100 *F / Hr Heatup
2400 2000 0->>
COMFUS ITE CURVE -100 °FI HR COOLDOWN V) 0..
( Margins of +I0°F and
-60 psig tor instrument error)
LLJ a:.:
1600
=>
t
(/)
Vl w
a:.:
Q_
0 1200 w
I-
<t::
u COOLDOWN-
~
0
- z f(
800 STEADY STATE\\.
_1-------
400 -
50 100 150 200 250 300 INDICATED TEMPERATURE 1*r)
(
X ~Hern _ f.> e
- a,..;_ R,=_{>R
- RM-C-Pl!.
X HAz-PR
- X k Q.a., T-..PR.
- x 1-\\ Q.C.t.:r _. t, ST
- Re.~c.- PR.
- X
~.e.a..c _ £.. s 7
- I
{<.e,f - l-ST.
V
-R.e-P... t ; el *
$HFi-PR *
(
-retJ_{)R X
lu£'-'l>..- PR..
6 P fl{!,~ l57
- Thwe.;:
1 p_ -,-
p_ T_Reac v ar/ WoL LM.d
~-ID)\\;).
- P<!.l(tc\\
- We..~ lC>l
- \\-\\C.l<.lOl
- S \\-\\.S~Cdi.
1-U!,\\(.. \\ D 1
-fro rn e 44ll4> -3 p G.l<. { 0 I R, J..\\-L\\-Olt, - 3 WCK.t0l *
~-~Or,: F ctlr bl dabb
,J/A ddtl -IFu!JjtriM.aJ add />>ta~. ~
~ fiAUMLld. (Vu~
"1}A Al/ A
/JlA add {If Ul /q /{p -,_ l) I;._~
Ai/A I
add fa U!J add~ IL5 add 1:1) UD Odd ~ad. ~- llhvMad.cw. ~
JJ/Pt add b U!>
f>* Jli,j ~.3~ 5 o ~Wt.
. JJ/-Pt CV
\\QIJ u)CL
- l b L -r l* I D TL
- ~ -r
~ -r1-8
- ~.. *l B
8 L -r