ML17326A522
| ML17326A522 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/08/1977 |
| From: | NORRIS E B SOUTHWEST RESEARCH INSTITUTE |
| To: | |
| Shared Package | |
| ML17326A519 | List: |
| References | |
| 02-4770, 2-4770, NUDOCS 8002270331 | |
| Download: ML17326A522 (93) | |
Text
,~~$SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Culebra Road San Antonio, Texas 78284 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR DONALD C.COOK UNIT NO.1 ANALYSIS OF CAPSULE T by E.B.Norris FINAL REPORT SwRI Project 02-4770 to American Electric Power Service Corporation 2 Broadway New York, New York 10004 J December 8, 1977 Approved: '4$}%$$.$~h$i\'l~h'I~P$~i A',"}:=:"}lCP,Ii EL:.O'IHiC PG"'/II SEl'}VLCC"-
CORi.~c DATE'.S.Lindholm, Director Department of Materials Sciences 80 f'2 2V0 53 (i TABLE OF CONTENTS LIST OF TABLES LIST OF FIGURES~Pa e r iii
SUMMARY
OF RESULTS AND CONCLUSIONS BACKGROUND III.DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM IV.V.TESTING OF SPECIMENS FROM CAPSULE T ANALYSIS OF RESULTS 13 35 VI.HEATUP,AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION OF DONALD C.COOK UNIT NO.1 VII.REFERENCES j APPENDIX A-.TENSILE TEST RECORDS 47 A-1 APPENDIX B-PROCEDURE FOR THE GENERATION OF ALLOWABLE PRESSURE-TEMPERATURE LIMIT CURVES FOR NUCLEAR POWER PLANT REACTOR VESSELS B-1 I
~~lb LIST OF TABLES Table~Pa e Donald C.Cook Unit No.1 Reactor Vessel Sur-veillance Materials Summary of Reactor Operations Donald C.Cook Unit No.1 16 Summary of Neutron Dosimetry Results Donald C.Cook Unit No.1-Capsule T 17 IV Fast Neutron Spectrum and Iron Activation Cross Sections for Capsule T 19 Charpy V-Notch Impact Data The Donald C.Cook Unit No.1 Reactor Pressure Vessel Intermediate Shell Plate B4406-3 (Longitudinal Direction) 21 VI Charpy V-Notch Impact Data The Donald C.Cook Unit No.1 Reactor Pressure Vessel Intermediate Shell Plate B4406-3 (Transverse Direction) 22 VII Charpy V-Notch Impact Data The Donald C.Cook Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 23 VIII Charpy V-Notch Impact Data The Donald C.Cook Unit No.1 Reactor Pressure Vessel Core Region Weld Heat-Affected Zone Metal 24 IX Charpy V-Notch Impact Data A533 Grade B Class 1 Correlation Monitor Material 25 Notch Toughness Properties of Capsule T Specimens 31 Donald C.Cook Unit No.1 XI Tensile Properties of Surveillance Materials Capsule T 32 XII Projected Values of RTNDT for Donald C.Cook Unit No.1 for Up to 12 EFPY of Operation 40 I~~Table ZIZZ XZV LZST OF TABLES (CONT'D.)Projected Values of RTNDT for Donald C.Cook Unit No.1 for Up to 32 EFPY of Operation Proposed Reactor Vessel Surveillance Capsule Schedule Donald C.Cook Unit No.1~Pa e 41 42 LIST OF FIGURES~Ft ure~Pa e Arrangement of Surveillance Capsules in the Pressure Vessel 2 Vessel Material Surveillance Specimens 3 Arrangement of Specimens and Dosimeters in Capsule T 12 4 , Charpy V-Notch Properties of Plate B4406-3 (Long.)Donald C.Cook Unit No.1 Surveillance Program 26 Charpy V-Notch Properties of Plate B4406-3 (Trans.)Donald C.Cook Unit No.1 Surveillance Program 27 Charpy V-Notch Properties of Core Region Meld Metal Donald C.Cook Unit No.1 Surveillance Program 28 Charpy V-Notch Properties of Core Region HAZ Material Donald C.Cook Unit No.1 Surveillance Program 29 Charpy V-Notch Properties of Correlation Monitor Material Donald C.Cook Unit No.1 Surveillance Program 30 Dependence of Cv Shelf Energy on Neutron Fluence, 37 Donald C.Cook Unit No.1 10 Effect of Neutron Fluence on RTNDT Shift, Donald C.Cook Unit No.1 38 Donald C.Cook Unit No.1 Reactor Coolant Heatup Limitations Applicable for Periods Up to 12 Effective Full Power Years 45 12 Donald C.Cook Unit No.1 Reactor Coolant Cooldown 46 Limitations Applicable for Periods Up to 12 Effective Full Power Years C~~
I.
SUMMARY
OF RESULTS AND CONCLUSIONS The analysis of the first material surveillance capsule removed from the Donald C.Cook Unit No.1 reactor pressure vessel led to the following conclusions:
(1)Based on a calculated neutron spectral distribution, Capsule T received a fast fluence of 1.80 x 101 neutrons/cm2
>1 MeV.(2)The surveillance specimens of the core beltline materials ex-perienced shifts in transition temperature of 75'to 130 F as a result of the above exposure.(3)The weld metal and heat affected zone (HAZ)materials exhibited the largest shift in RTNDT.However, because the intermediate shell plate material has a high initial (unirradiated)
RTNDT, it will control the heatup and cooldown limitations at least until the next surveillance-capsule is removed.(4)The estimated maximum neutron fluence of 6.92 x 1017 neutrons/cm>1 MeV received by the vessel wall accrued in 1.27 full power years.Therefore, the projected maximum neutron fluence after 32 effective full power years (EFPY)is 1.74 x 1019 neutrons/cm
>1 MeV.This estimate is based on a lead factor of 2.6 between Capsule T and the point of maximum pressure vessel flux.(5)Based on Regulatory Guide 1.99 trend curves, the projected maxi-mum shift in ductile-brittle transition temperature of the Donald C.Cook Unit 1 vessel core beltline plates at the 1/4T and 3/4T positions after 12 EFPY of operation are 110 F and 50 F, respectively.
These values were used as the bases for computing heatup and cooldown limit curves for up to 12 EFPY of operation.
(6)The maximum shifts in the transition temperature of the Donald C.Cook unit 1 vessel core beltline plates at the 1/4T and 3/4T positions after 32 EFPY of operation are pro)ected to be 180 F and 83 F, respectively.
(7)Since the weld metal and HAZ beltline materials are more sensi-tive to radiation embrittlement than the intermediate shell plate material, the operating limf.tations may come under control of the weld metal and HAZ material late in the 32 EFPY.design life of the plant.(8)The Donald C.Cook Unit No.'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 10CFR50 Appendix G throughout the design life of the unit.
II.BACKGROUND 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 ferritic materials, the allowable loadings depend on the reference stress intensity factor (KIR)curve indexed to the reference nil ductility temperature (RTNDT)presented in Appendix G,"Protection Against Non-ductile Failure," of Section III of the ASME Code.()Further, the materials in the beltline region of the reactor vessel must be monitored for radiation-induced changes in RTNDT per the requirements of Appendix H,"Reactor Vessel Material Surveil-lance Program Requirements," of 10CFR50.The RTNDT is defined in paragraph NB-2331 of Section III of the ASME Code as the highest of the following temperatures:
(1)Drop-weight Nil Ductility Temperature (DW-NDT)per ASTM E 208;(2)60 deg F below the 50 ft-lb Charpy V-notch (Cv)temperature; (3)60 deg F below the 35 mil C temperature.
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 com-prise 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 exposed to neutron fluences in excess of 1017 neutrons per cm2 (E>1 MeV).()Also, it has been established that tramp elements, particularly copper and*Superscript numbers refer to references at the end of the text.
phosphorous, affect the radiation embrittlement response of ferritic mate-rials.()The relationship between increase in RT~T and copper content is not defined completely.
For example, Regulatory Guide 1.99, originally issued in July 1975, proposed an adjustment to RT~T proportional to the square root of the neutron fluence.westinghouse Electric Corporation, in their comments on the 1975 issue of Regulatory Guide 1.99(), believed that the proposed relationship overestimates the shift at fluences greater than 1.9 x 1019 and underestimates the shift at fluences less than 1.9 x 10 On the other hand, Combustion Engineering, in their comments on the 1975 is-sue of Regulatory Guide 1.99 , suggested that the proposed relationship is overly conservative at fluences below 1019 neutrons per cm (E>1 MeV).There is also disagreement concerning the prediction of Cv upper shelf re-sponse to exposure to neutron irradiation.()After reviewing the comments and evaluating additional surveillance program data, the NRC issued a revision to Regulatory Guide 1.99 which raised the upper limit of the transition tem-perature adjustment curve.In this report, estimates of shifts in RTNDT are based on Revision 1 of Regulatory Guide 1.99), issued in April 1977.In general, the only ferritic pressure boundary materials in a nuclear plant which are expected to receive a fluence sufficient to affect 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 plate or forging material and weldments which are located in such a region.of high neutron flux density.ASTM E 185 describes the (10)current recommended practice for monitoring and evaluating the radiation-in-duced changes occurring in the mechanical properties of pressure vessel belt-line materials.
Westinghouse has provided such a surveillance program for the Donald C., Cook Unit No.1 nuclear power plant;The encapsulated Cv specimens are located near the O.D.surface of the thermal shield at a point where the fast neutron flux density is about three times that at the adjacent vessel wall surface.Therefore, the increases (shifts)in transition temperatures of the materials in the pressure vessel are generally less than the corre-sponding shifts observed in the surveillance specimens.
However, because of azimuthal variations in neutr'on flux density, capsule fluences may lead or lag the maximum vessel fluence in a corresponding exposure period.For example, Capsule T (removed during the 1977 refuelling outage)was exposed to a neutron fluence approximately
2.6 times
that at the maximum exposure point on the vessel I.D., while Capsule X (scheduled for removal at a later date)is being exposed to a neutron flux about 60%of that at the point of maximum vessel exposure.The capsules.also contain several dosimeter mate-rials 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 in-clude tensile specimens as recommended by ASTM E 185.At the present time, irradiated tensile properties are used primarily to indicate that the mate-rials tested continue to meet the requirements of the appropriate material specification.
In addition, the degree of radiation hardening indicated by the tensile yield strength is used to judge the credibility of the surveil-lance data.(7)Wedge opening loading (WOL)fracture mechanics specimens, machined from plate material and weld metal, are also contained in the capsules.Current technology limits the testing of these specimens at temperatures well below
~~the minimum service temperature to obtain valid fracture mechanics data per ASTM E 399~~,"Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Materials." However, recent work reported by Mager and Mitt~1~may lead to methods for evaluating high-toughness materials with small frac-ture mechanics specimens.
Currently, the NRC suggests storing these specimens until an acceptable testing procedure has been defined.This report describes the results obtained from testing the contents of Capsule T.These data are analyzed to estimate the radiation-induced changes in the mechanical properties of the pressure vessel at the time of the 1977 refuelling outage as well as predicting the changes expected to occur at selected times in the future operation of the Donald C.Cook Unit No.1 power plant.
III.DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM The Donald C.Cook Unit No.1 material surveillance program is described in detail in WCAP 8047(13), dated March 1973.Eight materials surveillance capsules were placed in the reactor vessel between the thermal shield and the vessel wall prior to startup, see Figure 1.The vertical center of each cap-sule is opposite the vertical center of the core.The neutron flux density at the Capsule T location leads the maximum flux density on the'vessel I.D.by a factor of 2.6.(The capsules each contain Charpy V-notch, tensile and WOL specimens machined from the SA533 Gr B plate, weld metal and heat affected zone (HAZ)materials located at the core beltline plus Charpy V-notch specimens machined from a reference heat of steel utilized in a num-ber of Westinghouse surveillance programs.The chemistries and heat treatments of the vessel surveillance mate-rials are summarized in Table I.All test specimens were machined from the test materials at the quarter-thickness (1/4 T)location after performing a simulated postweld stress-relieving treatment.
Weld and HAZ specimens were machined from a stress-relieved weldment which joined sections of the inter-mediate shell course.HAZ specimens were obtained from the plate B4406-3 side of the weldment.The longitudinal base metal C specimens were oriented with their long axis parallel to the primary rolling direction and with V-notches perpendicular to the major plate surfaces.The transverse base metal C specimens were oriented with their long axis perpendicular to the primary rolling direction and with V-notches perpendicular to the major plate surfaces.Tensile specimens were machined with the longitudinal axis parallel to the plate rolling direction.
The WOL specimens were machined X (220')270'(184')Y (320')Z (356)180'a S (4')V (176')T (40)u (140')90 0 Reactor Vessel Thermal Shield Core Barrel FIGURE 1~ARRANGEMENT OF SURVEILLANCE CAPSULES RT THE PRESSURE VESSEL
~~TABLE I D0NALD C.C0OK UNn No.1 REACT0R VESSEL SURVEn.LANCE MATERZALS<>>)
Heat Treatment Histor Shell Plate Material: Heated to 1600 F 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 1150 F for 40.hours, furnace cooled.Weldment: Stress relieved at 1150 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.1225 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, furnace cooled 1150 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 Si Ni Mo Cu Plate B4406-3 Weld Metal 0.24 1.40 0.009 0.015 0.25 0.49 0.46 0.14 0.26 1.33 0.023 0.014 0.18 0.74 0.44 0.27 Correlation Monitor 0.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14 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 T contained 44 Charpy V-notch specimens (10 longitudinal and 10 transverse from the plate material, plus 8 each from weld metal, HAZ 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 specimen numbering sys-tem and location within Capsule T is shown in Figure 3.Capsule T also was reported to contain the following dosimeters for de-termining the neutron flux density: Target Element Form Quantity Iron Copper Nickel Cobalt (in aluminum Cobalt (in aluminum)Uranium-238 Neptunium-237 Bare wire Bare wire Bare wire Bare wire Cd shielded wire Cd shielded oxide Cd shielded oxide 5 3 3 2.2 1 1 Two eutectic alloy thermal monitors had been inserted in holes in the steel spacers in Capsule T.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.10 46a 44'OII R.009 90~.3I.3 I4 2.I25 2.I05 l.063 l.053.3 5.393 (a)Charpy V-notch Impact Specimen.256.246 I.005.995.255.245 I6 Gage length 256.395 493 4.250 4.2 I 0.250 R I.250'.26 l.495 I.80 630.I98.I9.790.786.395.375 D.37'ECTION A-A (b)Tens ile Spec imen.375 D..380.439 499.437.04'73.0463 D.0667.0662.0667 l.45 l.4P I.I30 I.I20.765.745 I.005.995 I.005-8-~995.SOI.499 (c)Wedge Opening Loading Specimen FIGURE 2.VESSEL MATERIAL SURVEILLANCE SPECIMENS f C,COI CO.CCS CLttfC tttL~ISLICIII ILICI CLtt ff I Itl CLI ff IIL fC>COI CO CIS II~I III IIL OOL llISILC Cllltt Clutt Clllt1 CIOItl Clllt1 CllltT CClltl CllltT CILItt Cllltt Cltltt W.LI I IO I~'ll I-jl I~SI ISS lit SISS SS I I 4 IL4~I4 I.l~I SI 1 SI~SI 4 SI~SI ill ILO I~II I~I I II I ILL LI II~~'ll Y-SS I.ll I.lt~St W IL~SS~.Sl I SS I.ll I.IS 1-~I~I.IL I-~I I LS-II LI.I~.II~TOP BOTTOM ItICLLC~~IIIIII CIOC tllll ILLOI~I (IIIII'ITOIIIL
~IIICIIII)II tllIC~'LIOI.S (LIILSI(III
~IIICII4I)ISILL COIIILLIIOI Saallla I OILS OCII.IIIICLI
~IOIC IIL~IC ILL FIGURE 3.ARRANGEMENT OF SPECIMENS AND DOSIMETERS IN CAPSULE T
~~IV.TESTING OF SPECIMENS FROM CAPSULE T The capsule shipment, capsule opening, specimen testing and reporting of results were carried out in accordance with the Project Plan for Donald C.Cook Unit No.1 Reactor Vessel Irradiation Surveillance Program.The SwRI Nuclear Projects Operating Procedures called out in this plan include: (1)XI-MS-1,"Determination of Specific Activity of Neutron Radiation Detector Specimen." (2)XI-MS-3,"Conducting Tension Tests on Metallic Materials." (3)XI-MS-4,"Charpy Impact Tests on Metallic Materials." (4)XIII-MS-1,"Opening Radiation Surveillance Capsules and Handling and Storing Specimens." (5)XI-MS-5,"Conducting Wedge-Opening-Loading Tests on i Metallic Materials." (6)XI-MS-6,"Determination of Specific Activity of Neutron Radiation Fission Monitor Detector Specimens." Copies of the above documents are on file at SwRI.Southwest Research Institute utilized a procedure which had been pre-pared for the 1977 refuelling outage for the removal of Capsule T from the reactor vessel and the shipment of the capsule to the SwRI laboratories.
SwRI contracted with Todd Shipyards-Nuclear Division to supply appropriate cutting tools and a licensed shipping cask.Todd personnel severed the cap-sule from its extension tube, sectioned the extension tube into three-foot lengths, supervised the loading of the capsule and extension tube materials into the shipping cask, and transported the cask to San Antonio.13
~~~~The capsule shell had been fabricated by making two long seam welds to join two half-shells together.The long seam welds were milled off on a Bridgeport vertical milling machine set up in one hot cell.Before mill-ing off the long seam weld beads, transverse saw cuts were made to remove the two capsule ends.After the long seam welds had been milled away, the top half of the capsule shell was removed.The specimens and spacer blocks were carefully removed and placed in an indexed receptacle so that capsule location was identifiable.
After the disassembly had.been completed, the specimens were carefully checked for identification and location, as listed in WCAP 8047.(>>)Each specimen was inspected for identification number, which was checked against the master list in WCAP 8047.No discrepancies were found.The thermal monitors and dosimeter cfires were removed from the holes in the spacers.The thermal monitors, contained in quartz vials, were examined, and no evidence of melting was observed, thus indicating that the maximum temperature during exposure of Capsule T did not exceed 579 F.The specific activities of the dosimeters were determined at SwRI with an NDC 2200 multichannel analyzer and an NaI(Th)3 x 3 scintillation crystal.The calibration of the equipment was accomplished with appropri-ate standards and an interlaboratory cross check with two independent count-'ing laboratories on Co-, 54Hn-and~Co-containing dosimeter wires.All activities were corrected to the time-of-removal (TOR)at reactor shutdown.Infinitely dilute saturated activities (A8AT)were calculated for each of the dosimeters because ASAT 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 (1-e m)(e m)ATOR-XTm-Xt ASAT m~1 where: m=operating period;decay constant for the activation product, day 1;Tm equivalent operating days at 3250 MwTh for operat-ing period m;tm=decay time after operating period m, days.The Donald C.Cook Unit No.1 operating history up to the 1977 refuelling out-age is presented in Table II.The specific activity at time of removal (TOR)and the specific saturated activity calculated for each dosimeter are pre-sented in Table III.The primary result desired from the dosimeter analysis is the total fast neutron fluence (>1 MeV)which the surveillance specimens received.The average flux density at full power is given by: SAT m NOD (2)where: energy-dependent neutron flux density, n/cm-sec;ASAT saturated activity, dps/mg target element;spectrum-averaged activation cross section, cm;NO number of target atoms per mg.The total neutron fluence is then equal to the product of the average neutron flux density and the equivalent reactor operating time at full power.
TABLE II
SUMMARY
OF REACTOR OPERATIONS DONALD C.COOK UNIT NO.1 Operating Period Start Dates~DS S Operating~DS s Shutdown~DS S Power Generation Equiualent Operating Days T)Decay Time After Period 10 12 2/2/75 2/15/75 2/17/75 2/18/75 2/21/75 3/19/75 4/4/75 6/25/75 6/27/75 7/4/75 7/23/75 10/12/75 10/15/75 ll/1/75 11/15/75 1/2/76 1/5/76 4/13/76 5/10/76 7/2/76 7/6/76 9/11/76 9/19/76 11/21/76 11/22/76 2/14/75 2/16/75 2/17/75 2/20/75 3/18/75 4/3/75 6/24/75 6/26/75 7/3/75 7/22/75 10/ll/75 10/14/75 10/31/75 11/14/75 1/1/76 1/4/76 4/12/76 5/9/76 7/1/76 7/5/76 9/10/76 9/18/76 11/20/76 11/21/76 12/23/76 13 26 82 81 17 48 53 67 63 32 2,194 228 16 19 14 27 29,604 200,616 15,432 201,506 40,163 116,552 256,178 143,868 205,682 196,520 92 754 Total, Cycle 1 1,501,297 0.68 0.07 9.11 61.l3 4.75 62.00 12.35 35.86 78.82 44.27 63.29 60.47 28.54 461.94 678 675 646 548 539 439 419 357 255 175 104 33 0
~~TABLE III
SUMMARY
OF NEUTRON DOSIMETRY RESULTS DONALD C.COOK UNIT NO.1--CAPSULE T Monitor Identification Activation-Reaction ATOR (d s/m ASAT d s/m Fe-Fe-Fe-Fe-Fe-Top Top Mid.Mid.Bot.Mid.Bot.54Fe(n,p)54Mn Average 193 x 103 1.69 x 103 1.69 x 103 1.69 x 103 1.80 x 103 1.76 x 103 3.34 x 103 2.94 x 103 2.93 x 103 2.93 x 103 3.11 x 103.3.05 x 103 Cu-Top Mid.Cu-Mid.Cu-Bot.Mid.Ni-Top Mid.Ni-Mid.Ni-Bot.Mid.Co-Top Co(Cd)-Top Co-Bot.Co (Cd)-Bo t.U-238 Np-237 63Cu(n,a)60Co ll tf 58Ni(n,p)58Co If II Co(n,p)Co II If II 238U(n, f)137C 237Np(n, f)137Cs 5.14 x 101 5.27 x 101 6.04 x 101 3.83 x 104 3.77 x 104 3.95 x 104 4.87 x 106 1.83 x 106 5.03 x 106 1.64 x 106 1.20 x 103 4.53 x 103 3.43 x 102 3.52 x 102 4.03 x 102 4.46 x 104 4.38 x 104 4.59 x 104 3.25 x 107 1.22 x 107 3.36 x 107 1.09 x 107 N/A N/A 17 The neutron flux density was calculated from the 4Fe(n,p)4Mn reac-tion because it has a high energy threshold and the energy response is well known.The energy spectrum for Capsule T was calculated with the DOT 3.5 two-dimensional discrete ordinates transport code with a 22-group neutron cross section library, a Pl expansion of the scattering matrix and an S8 order of angular quadrature.
The normalized spectrum for Capsule T and the group-organized cross sections for the 54Fe(n,p)54Mn reaction derived from the ENDF/B-ZV library are given in Table IV.The value of o is Fe given by: 10 MeV aF (E)g(E)dE o (>1 Mev)-1'1 10$(E)dE l.00 (3)where: VF (>1 MeV)the calculated spectrum-averaged cross Fe section for flux>1 MeV, cm2 determined for the 54Fe(n,p)54Mn reaction.The resulting value obtained for fast (>1 MeV)neutron flux density at the Capsule T location was 4.50 x 101 neutrons/cm-sec.Since Donald C.Cook Unit No.1 operated for an equivalent 461.94 full power days up to the 1977 refuelling outage, the total neutron fluence for Capsule T is equal to 1.80 x 1018 neutrons/cm (E>1 MeV)based on the calculated spectrum at the cap-2 sule location.Assuming a fission-spectrum energy distribution at the capsule location, the cross-section for the 4Fe(n,p)4Mn reaction (E>1 MeV)would be 98.26 mb.The resulting flux and fluence values would be 4.95 x 10 neu-(4)trons/cm2-sec and 1.97 x 1018 neutrons/cm2, respectively.
18 TABLE IV FAST NEUTRON SPECTRUM AND IRON ACTIVATION CROSS SECTIONS FOR CAPSULE T Energy Range (MeV)8.18-10.0 6.36-8.18 4.96-6.36 4.06-4.96 3.01-4.06 2.35-3.01 1.83-2.35 1~11-1.83 Normalized Neutron Flux 0.0098 0.0254 0.0482 0.0471 0.0855 0.1400 0.1752 0.4689 54Fe(n,p)54Mn Cross Section (barns)0.581 9.577 0.491 0.354 0.205 0.099 0.023 0.0014 VF 0.108 barns Fe 19 The irradiated Charpy V-notch specimens were tested on a SATEC impact machine.The test temperatures were selected to develop the ductile-brittle transition and upper shelf regions.The unirradiated Charpy V-notch impact data reported by Westinghouse(13) and the data obtained by SwRI on the spec-imens contained in Capsule T are presented in Tables V through IX.The Charpy V-notch transition curves for the three plate materials and the cor-relation monitor material are presented in Figures 4 through S.The radia-, tion-induced shift in transition temperatures for the vessel plates are in-dicated at 50 ft<<lb and 35 mil lateral expansion.
A summary of the shifts in RTNDT and Cv upper shelf energies for each material are presented in Table X.Tensile tests were carried out in the SwRI hot cells using a Dillon 10,000-1b capacity tester equipped with a strain gage extensometer, load cell and autographic recording equipment.
One each plate and weld metal tensile specimens was tested at room temperature (RT)and at 550 F.The results, along with tensile data reported by Westinghouse on the unirradi-ated materials(1
), are presented in Table XI.The load-strain records are included in Appendix A.Testing of the WOL specimens was deferred at the request of American Electric Power Service Corporation.
The specimens are in storage at the SwRI radiation laboratory.
The Charpy V-notch results indicate that the HAZ is more sensitive to radiation embrittlement than the as-rolled and heat-created plate and about equal to that of the weld metal.This is surprising because the copper con-tent of HAZ is reported to be'uch lower than that of the weld metal.(3)20 TABLE V CHARPY V-NOTCH IMPACT DATA THE DONALD C.COOK UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B4406-3 (LONGITUDINAL DIRECTION)
Condition Baseline Capsule T Spec.No.(a)A-44 A-45 A-49 A-50 A-41 A-47 A-42 A-48 A-43 A-46 Test Temp.(p)-40-40-40 10 10 10 40 40 40 76 76 76 110 110 110 160 160 160 210 210 210 300 300 300 10 40 82 110 135 160 185 210 250 300 Impact Energy (ft-1b)10 ll 11.5 24.5 33 31.5 57 42 65 82 70 78 93.5 100 88 110 131.5 115.5 120 144 125 131.5 126 132 10.5 29 38 46.5 62.5 84 99 105 110 105.5 Shear (x)9 11 13 23 25 29 45 37 37 52 59 52 95 100 95 100 100 100 100 100 100 1 5 20 35 25 55 95 95 100 100 Lateral Expansion~Mls 13 10 11 24 29 28 49 40 54 67 60 61 72 77 72 84 95 83 89 98 95 90 92 93 10 24 31 38 53 58 80 83 89 89 (a)Not reported.21 TABLE VI CHARPY V-NOTCH IMPACT DATA THE DONALD C.COOK UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B4406-3 (TRANSVERSE DIRECTION)
Condition Baseline Capsule T Spec.Na.(a)AT-44 AT-45 AT-49 AT-50 AT-41 AT-47 AT-42 AT-48 AT-43 AT-46 Test Tempt~P)-40-40-40 10 10 10 40 40 40 76 76 76 76 110 110 110 160 160 210 210 210 300 300 300 1O 40 82 110 135 160 185 210 250 300 Impact Energy~ft-1b)11 11.5 14 28 23 30 40 41 37 83 43 50 50 84 54 68 97 77 90 95 97 100 94 101 6 25 35 37 49.5 57 73.5 87 87 89 Shear~7.)14 9 9 18 23 18 27 27 32 27 48 37 41 90 90 100 100 100 100 100 100 5 5 20 30 25 40 100 100 100 lOO Lateral Expansion~milt 12 15 15 28 22 26 36 35 34 56 44 46 44 71 51 57 80 71 75 79 79 83 75 85 8 23 30 35 44 47 63 73 71 83 (a)Not reported.22 TABLE VII CHARPY V-NOTCH IMPACT DATA THE DONALD C.COOK UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL Condition Baseline Spec.No.(a)Test Temps ('p)-140-140-140-100-100-100 Impact Energy~ft-1b1 ll 21 19 23.5 29 20 Shear (X)18 20 11 Lateral Expansion~m11s 10 19 18 22 26 18 Capsule T W-33'-35 W-34 W-39 W-40 W-37 W-38 W-36-70-70-70-40-40-40 10 10 10 76 76 76 210 210 210>>40 10 75 82 110 160 210 300 45.5 51 54 63 59 69 83 84 92 114 107 107 110 112 111 24.5 50 75.5 44 85 75 98 68.5 24 42 32 47 34 47 73 71 75 99 100 100 100 100 100 5 20 70 20 95 100 100 100 39 47 49 52 53 60 69 72 75 88 87 88 90 87 93 19 41 67 34 69 66 66 66 (a)Not reported.23 TABLE VIII CHARPY V-NOTCH IMPACT DATA THE DONALD CD COOK UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION MELD HEAT-AFFECTED ZONE METAL Condition Baseline Spec.No.(a)Test Tempo~7)-175-175-175-140-140 Impact Energy (ft-lb)5.5 7 7 16 22 Shear~(/Lateral Expansion~mals 12 18-100-100-100 30 33 45 13 14 20 25 28 40 Capsule T H-33 H-35 H-34 H-39 H-40 H-37 H-38 H-36-70-70-70-70-40-40-40 10 10 10 76 76 76 210 210 210-40 10 45 82 110 160 210 300 52 47 27 30 54 71 47 97 89 82 112'40 131 129 104 105 10 40.5 30.5 52.5 62.5 84 111.5 83 21 25 14 20 55 50 43 90 43 69 100 100 100 100 100 100 5 15 25 25 40 100 100 100 39 35 21 24 53 50 45 83 67 64 86 84 82 85 94 87 9 30 27 41 46 65 78 54 (a)Not reported.
TABLE IX CHARPY V-NOTCH EPACT DATA A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL Condition Baseline Capsule T Spec.No.(a)R-33 R-37 R-38 R-39 R-40 R-34 R-35 R-36 Test Tempr-50-50-50-20-20-20 10 10 10 40 40 40 85 85 85 110 110 110 160 160 160 210 210 210 300 300 300 40 82 110 160 210 300 350 400 Impact Energy~fe-1b)6.5 9 6 12 14.5 13.5 22 36 35 58.5 41.5 52 82.5 85.5 63.5 108.5 81 109 117 115 121 125 117.5 127 13.5 18.5 35 55.5 86.5 100 111 96.5 Shear~X 9 13 13 23 23 23 33 29 29 43 41 42 58 67 55 84 85 87 98 98 100 100 100 100 5 10 20 40 95 100 100 100 Lateral Expansion~mals)6 10 9 15 14 14 23 32 32 51 42 45 60 71 54 72 69 79 84 88 87 87 83 84 13 18 32 45 66 57 84 84 (a)Not reported.25 e~~160~'I I~I I 1 t t I~e i I i t I~I I I l I t 1 I I e I I 1 Qt I~I~I I I t I I~t!I t'I I I~I I e I I I 120 I I I~I~e t I 1 I~I~I i I I!I j I~e I l~i~I~I I I e e+lg!I I I e I~I I i t~I~I I i I~'I f.e I I I f~~~e I I I I I 1 e~I C I~J I I I I~'00 SI u fz1)C3 80 40 e I e I e~I t I, I I~e I~e I~I I I t 1 T~I~I I I e I I'I t I I I I I~I I I!I I i i~I I I I I I I'1 I I I I i~e~I I~~I.'I I I~I~lt~C~I~t ,I I t I j: t~I I t I I!I I I'~I'-I I I..L'L.l.e~~'I I I'I I~~1 t e I~I I I~'I I e I~I 1 J~~I~e t e I i,t 1 I I I I I I~I';!I I I i I I I I I'I~" e e~-Baseline
~~I~e I't I I I~I j C~I e I e I t', 1~I~I I, i I'-Irradiated Capsule T e~~e~~'I C I"~0!*I<e I I'I I I-200-100 0 100 200 Temperature, deg F 300 400 100~I~I'I'l~I~~~I 75 I e I e e~~I~I~I C 0 50 e I I~I e)C c7 25 1 e I e 1~I I I I i I I I I I I I I I~I~I I~'~e I'I I I I I~I~j I I I~~I!I I'.I~j e I'~t i e e;i I I I I I e....I,.e I~~e I~Unirradiated Baseline~~ee e~,~\-Xrradiated Capsule T 0 I e I'I~I~.I I I I 1 I I'I~~'~~e e I~e L I-200-100 0 100 200 Temperature, deg F 300 400 FIGURE 4~CHARPY V-NOTCH PROPERTIES OF PLATE B4406-3 (LONG-)DONALD C.COOK UNIT NO.1 SURVEILLANCE PROGRAM 26
~c 160 a~a e 120~I~I I I 1,!a, I a 1 a ja a I~I I 7 a a a~I a I'I a ,'I ae 80 I I~I.'I 7~','I I~I C 1 I Ia I I I I I I I I I 9~I 1~aa I~1 I I t I I~a~I I I a~!I I I.I 1 I I~9 I't" I~~'I~'~I I i I 1~a I!40~e I\';~I I~l I I I 1 I I I~.'I I~1~~e I 1~a a a I I'-'I I~I I I I:~t I e~~I a, 1~~-I=~a"~-200-100 0 100 200 Temperature, deg F 300 400100 co 75 8~'*f e'I 1~1 1 I I'~I.'dMted&ad~Lat-dd'-Capj I: I I'I I*a 1 t I'I~I I I I I!I I I I~t 1, I'I~~I a e 1~>>~'1.L Ot~J'1 a~~~'I" 1 0~50 1~ee I~'I I a I I!~I~.6=-~.I P 25 I~~at, I I'I Jaai~t:~!I I I I I I+I I LJ'1 a'.~~I: I't 1 I!J I I I a~I."'a-H I I I a I I I~at I I I I I I I I I I I I I I I I~I I I I'I iI I a I I~I a I I I'i t i I~.~L j I I~I e~I 1."..~t.I a.I l T I;..L I I/I'.a J f)I f[I~'I I J I I I I I I, I, 9-200-100 0 100 200 Temperature, deg F 300 400 FIGURE 5.CHARPY V-NOTCH PROPERTIES OF PLATE B4406-3 (TRANS.)DONALD C.COOK UNIT NO.1 SURVEILLANCE PROGRAM 27 160 4~>I~~~'4 i i>!~I I I i j l-r+'!-'.H.I~'I I I~~I~I I I I~I I f: ';~I I 120~!I~I I I~;I I I I I I.I I I i I!>I>1~~~I I I I I I I~1 4>1 I~~80 i).~j t~~~I I I t I>I I 4 i 4'I~>>I i 1~I I I~I~I I I I>I~>I~I I>~~>I I 4>I I': i'I~~I>,'I~f 1 I<<I~~,C I I I I I l 4 I>i I~~4~~1 t I I>j~.I I~~~i i I->I I>>t i i i I i I~~40 0 i C t I i t~t I I>I~I I~~~I I~4 I I I~I I~I I I I I I>~>>~U~diatad-'Baseline j i I~~I I.I I i I I I I I j i I I I.'t 1 I~~I~I~I I I I'I',w j-.'-+-.-&@Bated-Capsule T-200-100 0 100 Temperature, deg F 200 , 300 400 100 I I 4 I~.'1 i I I I>I I:.'4'~t'C;4~I~1 4 I I~>I~1,'>~m 75 8 1~>~I I I~~C A j Q 50 X c5 25~~I~1'>'4>I'~~r I 4~>~f~f I l.4~'i I~~I I~I I~I.i.I: I I I,'1 j I>I I I, 1~4 I'I;I!~'~f i 4 t I I~'-~Coda-.4 I I,.I I~~~,','~.'-,.":,;&--.UnMrediated Baseline I I~adiated Capsule T I~I;'4 i I I: I I I>i't I I..*.~~I>I I I'.~>4'I~'I-200-100 0 100 200 Temperature, deg F 300 400 FIGURE 6.CHARPY V-NOTCH PROPERTIES OF CORE REGION WELD METAL DONALD C.COOK UNIT NO.1 SURVEILLANCE PROGRAM 28 160!!~~!I I"!~I l!D I W 120~I I='I I I I I~~I I~i I I~I I'!l I~I i I i I I: '!I i I I'+l~~.I I I!j I I t"~~!80)CD~~I!!I I~!!~I t I I!l!." I!~I I I I I I I'"I i!~~I l~I.!+:'I I I!(~l I P 40 0 I I I I I I I I~I I!I I I I'L t f I I l'I I I I I I l:~!I~$J~l (iT i~i(!!,.'I~A~I~!I~ll't~Base1ine.~I!!,'Xrmdia5edl.
Capsule I,~I I I'(-200-100 0'00 200 Temperature, deg F 300 400 100!I~I i.I I i I I'~'~J.'l I'I I l~'!I I~!0 75 l~I~I', i I I (J~l I I I I I 1 I:~j I I I!I~I I I I I I.'~LA I I I~I~~I'I I 50 I I, I I I I I l I I'j I I;I 4~t I!I~'I'I~l!j j!~f I I (I~w I J u 25 0~!!I I I I l I I~~!(Q (I I I I;I~" o-'-I I I'I.l I I I I!!I~-~l I I!I I r I I 0 I~I'i'I I I I~I~1 l j t~I I t I'~', Unirradiathd Baseline 4 ,'-Irradiated(Capsu1e T I I~!!'!-200-100 0 100 200 Temperature, deg F 300 400 FIGURE 7.CHARPY V-NOTCH PROPERTIES OF CORE REGION HAZ MATERIAL DONALD C.COOK UNIT NO.1 SURVEILLANCE PROGRAM 29 160 I~'.~I f I I, I f~f I i I I I I~f~I~', I f I'~I I j I I I~~I'~I I!f I~I I I~~I~I I I l~~~I I I~I j~I I I~~I 120~, I I I rratHQtedj iz1+~.I~~I~~r, f I I I I I I~I.t!I I I~I~I I~~~l I~~'I!~~~'"'0"f I I~~I 80 C)o~I I~~I I~~~r~I 1 l"'~I I I'~~~I I~I!I I!I i l 1 I I I I~ii~I j I I I I~I;;~i!II I I~j I!I I~I~I'I!I I'~I I f~~~'I I I~I 1 I~'I I L!I~!!I I I~I~~!J~~I I I l!'~'I i~I f I I~~I~t!l;I~~~I I 40 I I I~I~I!~I I I f~jt I f\~j I'~;~I I I~>'.I f~f I!~I~I I I I I I I~~I~I~I I l 1 I~i i I I I f I I I I'I T~I I I I I I I I I I I I I I I~~I I j j!\I~1,'I 0-200 t!-100 0 100 200 Temperature, deg F 300 400 100 j.I.2008~i~75 8~"-~Zrred kited-'Gape
~~j~I I~~I I I'~j I 0 50 LI o 25 I!I I~~~!~I~~t!1!I I f I~~-I!, I I I I.I~'~~~!~.I~~~I'.I'~l I i I I I j I~I I I I~I~I'I~I I!!I I~I!'I 1~I!I~~I j~!!1!I~~~: j i I i I I I~'1 i r i~~r I!tt, I~!e*0 I!I!~I 1~~~I i~I i!~1 I~I r I-200-100 0 100 200 Temperature de@F 300 400 FIGURE 8.CHARPY V-NOTCH PROPERTIES OF CORRELATION MONITOR MATERIAL DONALD C.COOK UNIT NO.1 SURVEILLANCE PROGRAM 30 TABLE X NOTCH TOUGHNESS PROPERTIES OF CAPSULE T SPECIMENS DONALD C.COOK UNIT NO.1 50 ft-lb C Tem.(de F)Plate B4406-3 Weld Weld~(Lan.)(Trans.)Metal HAZ Correlation Monitor Irradiated Unirradiated AT 35 mil C Tem.(de F)Irradiated Unirradiated AT C U er Shelf Ener ft-lb)150(a)140 75(a)65 75()135 (b)110 60(b)40 75(b)60 70-70-60 130 130 50 55-80-75 130 130 145 75 70 125 60 65 Unirradiated Irradiated hE, ft-lbs AE, 130 108 22 16.9 94 84 10 10.6 110 120 80 93 30 27 27.3 22.5 120 102 18 15 (a)Energy transition at 77 ft-lb.(b)Lateral expansion transition at 54 mil.31 TABLE XI TENSII E PROPERTIES OP SURVEILLANCE MATERIALS CAPSULE T Condition Specimen Ident.Test Temp.('P)0.2X Yield Tensile Total Reduction Strength Strength Elongation in Area~si~sf~I (%)Baseline Capsule T Baseline B4406-3 (Long.)A-1 A-2 B4406-3 (Trans.)Room Room 300 300 600 600 Room 550 Room Room 300 300 600 600 68,650 68,250 61,350 61,200 58,000 58,550 72,700 66,700 68,700 67,600 61,000 60,900 58,300 55,900 90,650 90,250 82,650 82,300'87,000 87,400 99,800 93,000 90,300 89,450 82,800 81,900 86,000 86,600 27.7 27.4 23.4 22.6 26.0 25.4 24.3 20.2 26.6 25.6 23.0 23.3 24.8 24.7 70.4 69.6 69.4 69.7 65.1 67.0 65.7 64.3 65.8 65.0 65.0 64.6 58.8 58.6 Capsule T W-9 M-10 Baseline Veld Metal Room Room 300 300 600 600 Room 550 66,900 67,350'9,700 59,800 57,200 56,300 86,100 75,800 81,500 82,250 74,600 74,500 79,400 78,500 103,400 95,300 28.7 25.0 24.0 23.3 23.4 23.6 23.6 19.3 73.2 65.3 72.9 71.8 65.2 63.4 65.0 60.8 32
~~The tensile properties of the weld metal appeared to be the most af-fected by the radiation exposure in Capsule T as expected from.the reported copper contents.33
'~V.ANALYSIS OF RESULTS 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 reduction in C upper shelf energy to the vessel wall using knowledge of the energy and spatial distribution of the neutron flux and the dependence of Cv upper shelf energy on the neutron fluence.(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 RTNDT to the vessel wall using knowledge of the dependence of the shift in RTNDT on the neutron fluence and the energy and spatial dis-tribution of the neutron flux.The energy and spatial distribution of the neutron flux for Donald C.Cook Unit No.1 was calculated for Capsule T with the DOT 3.5 discrete ordi-nates transport code.The lead factor for Capsule T reported by Westinghouse is 2.6 for the vessel I.D.surface.()This was supported by the SwRI DOT 3.5 analysis.The DOT 3.5 analysis also predicted that the fast flux at the 1/4T and 3/4T positions in the 8-5/8-in.pressure vessel wall would be 49%and 7.8%, respectively, of that at the vessel I.D.These figures are in good agreement with fluence attenuation determinations of 46%and 10%for an 8-in.steel plate by the Naval Research Laboratory.()However, currently the NRC pre-fers to use more conservative figures of 60%and 15%, respectively, for the attenuation of fast neutron flux at the 1/4T and 3/4T positions in an 8-in.
vessel wall.(16)This conservatism allows for the increased fraction of neutrons which might accrue in the 0.1 to 1.0 MeV range in deep penetra-tion situations.
For the 8-5/8-in.wall thickness of the D.C.Cook Unit No.1 vessel, the attenuations become 57%and 12.5%for the 1/4T and 3/4T positions, respectively.
A method for estimating the reduction in Cv upper shelf energy as a function of neutron fluence is given in Regulatory Guide 1.99, Revision 1.()The results from Capsule T are compared to a portion of Figure 2 of.(7)Regulatory Guide'.99, Revision 1, in Figure 9.The embrittlement response of the weld metal, reported to contain 0.27%Cu(), is in good agreement with the prediction of Regulatory Guide 1.99, Revision 1.However, the plate is less sensitive and the HAZ is more sensitive than predicted for the 0.14%copper content.The behavior of the HAZ specimens may reflect some copper pickup in the HAZ from the weld deposit or the placement of the notch unusually close to the fusion line.Using the dashed curve drawn through the data point for the weld metal, it is predicted that the weld metal Cv shelf energy will reach 50 ft-lbs at a fluence of about 2.1 x 10 (E>1 MeV).This corresponds to approximately 38 effective full power years (EFPY)of operation at the vessel I.D., in excess of the 32 EFPY design life of the plant.The plate and HAZ materials are projected to require even larger fluences to reach the 50 ft-lb shelf level.These projections will be reex-amined after the next surveillance capsule has been removed.A similar approach can be taken to estimate the increase in RTHDT as a function of reactor power generation.
Figure 10 compares the Donald'.Cook Unit No.1 surveillance data on the three surveillance materials to selected portions of Figure 1 of Regulatory Guide 1.99, Revision 1.The results 60 40 l s I I~I'.]j.!~I lit!~~I Is s!il,!!!'ii ss RR..OI.'!I I.jj sr s'l:lli I!!!its jls I!.I i ss s I't:.s I it ili I)~i lt;}.sl s s s~I'I~I:I~: list~~Isl l I s'I I I~I l s j s s I~Is sl 4l e 20 W 10 sl I~I fl I I!III j I,-I-I'0 I I s, s fili~ls tel.j!j.l!j Igloo ll if)):III jig!rI js\~s Ill~~tjj.l.!~I.fjl jjj ll!~~s l!j i!i st I I jjjj s>>I I i ii v)6 0 0 Qt A I I I!I~~l~I s I ss ls)l:;l',I,", s sl i!l tlji jl!i,:.l I~I s s s ss st I js I s!Is i.;"'I ssl:Isl iL" ls lrss: I>>>>I, I;I,'~s!jl I~!'i ill;I sj!I I~s sill sss I ls>>I I>>!'!I'.Ill j~I s's~l+II I j";l s Is!'..'Ill I~ls',~I!'ii!~hl.:m lil llj'-I:j l I t" r ,.)j, jl t s sl ll I~II li I'.,.'jii>>s'.~l Ill'is's'I sSI It Is s 1 l ill~~s~~ls Illj ss il!I lsll I!@pe.-I.s.i"I','s" t.s~'I I I'l j.~s-j~~~2 x 1017 4 6 8 1018 2 4 6 8 1019 2 4 Neutron Fluence, n/cm (E>1 MeV)FIGURE 9.DEPENDENCE OF Cv SHELF ENERGY ON NEUTRON FLUENCE, DONALD C.COOK UNIT NO.1 600 400 200 100 80 60 40 20 I~ll i.i'~I'Ij I I.: 1~I I~>>I-~I'.lji:.Il'I I/I I ,II!'ill"I 11~', lait I!I)1 jl ilj I I I tj I~I I I I lf jl Igt il;,'" 1 it!i., I)K I C4!.I..i.I j I Iij l I I 1 I I f.I)I>>.j-I II, lll II!l Ill~ll irl: '11 II;!I': jl, 1l I Il I)I Ij tjt~~I!!Ii I,1;r!,i l ,!,!j'll'l T~).I,I'11~I'I I~II I'l t.l j,j tll.!I lF~I)1~I I':i]j t!I il I:,I i~'I I I>>I iii I.~I I j-I I.I~I':I~'I I f I 77'~I I I~~I I 1'l I!ij!I!1 j>>i i, I)::;hii ili:.II-': I~I tl 2 x 1017'4 6 1018 8 1019 2 4 Neutron Fluence, n/cm (E>1 HeU)FIGURE 1 EFFECT OF NEUTRON FLUENCE ON RTNDT SIIIFT>>DONALD C COOk UNIT NO' indicate that the measured shift in RTNDT of the weld metal is in agreement with that predicted by Regulatory Guide 1.99, Revision 1, but that the mea-sured shifts in RTNDT for the plate and HAZ materials are underpredicted by the guide.The predicted shifts in RTNDT for the Donald C.Cook Unit No.1 reac-tor pressure vessel obtained from Figure 10 are summarized in Tables XII and XIII.The values predicted at the 1/4T and 3/4T after 12 EFPY (Table XII)are used to develop heatup and cooldown limit curves to meet the require-ments of Appendix G to Section III of the ASME Code, as described in Section VI of this report.These projections for Cv shelf energy reductions and RTNDT shifts, and the resulting heatup and cooldown limit curves, are based on extrapolations from one data point representing the most sensitive material.After a second capsule has been removed and tested, one will be able to inter-polate between two data points.The Donald C.Cook Unit No.1 reactor vessel surveillance program sched-ule proposed by Westinghouse~
~is summarized in Table XIV.It has been or-ganized to satisfy Appendix H of lOCFR50 as closely as possible.There are seven additional capsules in the vessel, all of which contain base plate, weld metal and HAZ specimens.
There is no reason to consider changing the proposed capsule removal schedule at this time.39 TABLE XII PROJECTED VALUES OF RTNDT FOR DONALD C.COOK UNIT NO.1 FOR UP TO 12 EFPY OF OPERATION Location Material Calculated Fluence (n/cd E>1 MeV)Initial RT (de F))Shift 12 EFPY(a Vessel I.D.~~Vessel 1/4T Vessel 3/4T Inter.Shell Plate Weld Metal HAZ Inter.Shell Plate Weld Metal WZ Inter.Shell Plate Weld Metal MZ 6.55 x 1018 3.73 x 1018 45(b)-52(b)-60(c)45(b)52(b)-60(c)45(b)52(b)-60(c): 145 245 245 110 185 185 50 87 87 190 193 185 155 133 125 95 35 27 (a)1 EFPY 1,186,250 M&t.(b)Reference 18.(c)References 13 and 18.
TABLE XIII PROJECTED VALUES OF RTNDT FOR DONALD C.COOK UNIT NO.1 FOR UP TO 32 EFPY OF OPERATION Location Material Calculated Fluence (n/cm2 E>1 MeV)Initial R DT (de F))32 EFPY(a Shift Vessel 1/4T Vessel 3/4T Inter.Shell Plate Meld Metal HAZ Inter.Shell Plate lfeld Metal HAZ Inter.Shell Plate Meld Metal HAZ'1.0 x 1019 2.2 x 1018 45(b)-52(b)-60(c)45(b)-52(b)-60(c)45(b)-52(b)60(c)240 320 320 180 285 285 83 142 142 285 268 260 225 233 225 128 90 82 (a)1 EFPY=1,186,250 MMDt.(b)Reference 18.(c)References 13 and 18.
TABLE XIV PROPOSED REACTOR VESSEL SURVEILLANCE CAPSULE SCHEDULE DONALD C.COOK UNIT NO.1 Capsule Identification Lead Factor Removal Time 2.6 2.6 0.6 Removed and tested at end of first core cycle 10 Years (postirradiation test)10 Years (reinsert in Capsule T location)0.6 10 Years (reinsert in Capsule X location)2.6 20 Years (postirradiation test)0.6 2.6 0.6 20 Years (reinsert in Capsule U location)30 Years (postirradiation test)30 Years (reinsert in Capsule Y location)
~~VI.HEATUP AND COOLDOMN LIMIT CURVES FOR NORMAL OPERATION OF DONALD C.COOK UNIT NO.1 Donald C.Cook Unit No.1 is a 3250 Mwt pressurized water reactor oper-ated by American Electric Power Service Corporation.
The unit has been pro-vided with a reactor vessel material surveillance program as required by 10CFR50, Appendix H.The first surveillance capsule (Capsule T)was removed during the 1977 refuelling outage.This capsule was tested by Southwest Research Institute, the results being described in the earlier sections of this report.In sum-mary, these results indicate that: (1)The RTNDT of the surveillance materials in Capsule T increased a maximum of 130 F as a result of exposure to a neutron fluence of 1.80 x 10 neutrons/cm2 (E>1 MeV).(2)Based on a ratio of 2.6 between the fast neutron flux at the Capsule T location and the maximum incident on the vessel wall, the vessel wall fluence at the I.D.was 6.92 x 1017 neutrons/cm2 (E>1 MeV)at the time of removal of Capsule T.(3)The maximum shift in RTNDT after 12 effective full power years (EFPY)of operation was predicted to be 185 F at the 1/4T and 87 F at the 3/4T vessel wall locations, as controlled by the weld metal and HAZ materials.
(4)The intermediate shell plate material, although less sensitive to radiation embrittlement than the weld and HAZ materials, is projected to control the limiting RTNDT for a considerable length of time because of a much higher initial (unirradiated)
RTNDT of 45 F.(43
~~The Unit No.1 heatup and cooldown limit curves for 12 EFPY have been computed on the basis of (4)above because it is anticipated that the RTNDT of the primary pressure boundary materials will be highest for the plate ma-terial at least through that time period (see Table XII).The procedures employed by SwRI are described in Appendix B.The following pressure vessel constants were employed as input data in this analysis: Vessel Inner Radius, ri Vessel Outer Radius, ro Operating Pressure, Po Initial Temperature, To Final Temperature, Tf 86.50 in., including cladding 95.34 in.2235 psig 70 F 550'F Effective Coolant Flow Rate, Q~135.6 x 10 ibm/hr Effective Flow Area, A 26.72 ft2 Effective Hydraulic Diameter, D~15.05 in.Heatup curves were computed for a heatup rate of 60 F/hr.Since lower rates tend to raise the curve in the central region (see Appendix B), these curves apply to all heating rates up to 60 F/hr.Cooldown curves were com-puted 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 cooldown rates up to 20 F/hr;the 40 F/hr curve would apply to rates from 20 F to 40 F/hr;the 60 F/hr curve would apply to rates from 40 F to 60 F/hr;the 100 F/hr curve would apply to rates from 60 F/hr to 100 F.hr.The Unit No.1 heatup and cooldown curves for up to 12 EFPY are given in Figures ll and 12.44 2600 2400 2200~I~~I lt I I~2000 1800 1600 1400 u 1200 1000 800 I~~1 t]t...I 1~t I~lf~~i I~'ff if}l,l 600 400 200 l g~:it}r'i y~~II~~~I I~~, l I:1 t 1~<<f~f I.,f~f1~I I"I~~lf:1~~1~r~jt it'r i H f)L1 1,~I f.[~]~f l f 60 100 150 200 250 300 350 400 Indicated Temperature, deg F FIGURE ll.DONALD C.COOK UNIT NO.1 REACTOR COOLANT HEATUP LIMITATIONS APPLICABLE FOR PERIODS UP TO 12 EFFECTIVE FULL POfKR YEARS 2600 2400 2200 2000 800 600:.,;I)l)~'l e~1~~~~I 1~~er~I:.:[::II 1 I.;I'i>I:;I,-: I I}i I'l e.l,)IL)"!I t e 400 200 6 0 1~~1)~I,.1 1800 Aj 1600 P4 1400~Q Ai 1200 1000~~I.".s~'-s e-I s~ll 1:~1)I!-.1;I: I!~si j!I'.I I)~s~sell I 1~I.',: I l'all I~-i): ls~~~It l~1-: I~~I i'~~".I)'e~I'!1)~r:li~~ee)e i::I?.I g!ji'.l)!~.I I~~~I~~1t1'e ii!l.f~ie)tl~ij~~i~.~)~I~e=-~'1 e I~j'I 100 e fje::ff!."Ie s e>}s)1 Is!',I i a e+I I;.I ,~ll~e)i}'e s 1~ls.I lr', ll)~I eI'.>I;.:I~1 e I~~I~~1~i I~I 1 e~g I I'..I 1~~t~~150 1~I~~el>a~I sell~e e a I~~'I Rl I~~e~1~tt I\I Is~la f I>1~1)I~t I ,.I))~>I:lj 1 1'i" Fig+:I.~'I~1 ae ,.)I ll~I I~~I.';ll Sel 1~1~1 itis ,';)I eisa Ill 11~~i:I I'I}ll~~g~!'.I!!ilI)1;I)s~~s s e i;.1'~~'.L c'l,!)le~iIi;f})Jjl I:,I 1~I 1-"~e..I:}i.~f:I'I:i!~l)iife'Il 'll;~~~l 200 s)~Sl'1 e~~~1 I'~1~~ii~i>le l f:I!.p.)1~le'I I)!Ij)s e1~I~)I~I~ll!e~~.~1 1~~'~~~,iii j~I s}esj)rjl li I!1 I Ill si),.I's If!a~~~s~~I~~~I'.I)ffjf i li I'I s s g I 1 1 il>!I I!e!~)'-ii'sl:::II l I!I s~~~tp;if g I A)~I:e)r)')1 lel)ls I~~I I ej 1~~~e e~1 I~250 s~1 s;-1~~~):I s s I'l~1 f~~I stj t~.I~~1 I eae::il s'Ii, I sl s~I 1 f e".I-'ll:;.L'::.:.-:'I.':.i I~*~el')t si e~I I II): I'll~1 s g~1~~')tI-!I'I~)I~-1!(}I 1>~~~~I~~e I'.1t!I!I t)}?4>I l I r I~~~'.~e~~1 I;r f:Itj I~as~300~~>l~I?)j:-j-'~t's 1 lj:sn i l'l'll.'g l t ,1 I~~.1 e al I~~I I'1>f 11>~~II l j~:, Ii S1~~I.:;.I 1 j)II~,J e)~f I.:::)I:!i j I 1~e I"~~4~e el g e e~4 ll::I I~~)t j~el Ta e~e~~s 350 1 g~~rma~i~~~~:i>i st';le)~~I~1 I,,~~f II~rg l alj)IIj al: s.A~e'1~~{g II t I't'~~I}L>I l}I li rf I'r I 400 Indicated Temperature, deg F FIGURE 12.DONALD C.COOK UNIT NO.1 REACTOR COOLANT COOLDOWN LIHITATIONS APPLICABLE FOR PERIODS UP TO 12 EFFECTIVE FULL POWER YEARS VII.REFERENCES 1.Title 10, Code of Federal Regulations, Part 50,"Licensing of Produc-tion and Utilization Facilities." 2.ASME Boiler and Pressure Vessel Code, Section III,"Nuclear Power Plant Components," 1974 Edition.3.ASTM E 208-69,"Standard Method for Conducting Drop-Weight Test to De-termine Nil-Ductility Transition Temperature of Ferritic Steels," 1975 Annual Book of ASTM Standards. 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 Embrittlement of Reactor Pressure Vessel Steels," International Atomic Energy Agency, Technical Reports Series No.163, 1975.6.ASME Boiler and Pressure Vessel Code, Section XI,"Rules for Inservice Inspection of Nuclear Power Plant Components," 1974 Edition.7.Regulatory Guide 1.99, Revision 1, Office of Standards Development, U.S.Nuclear Regulatory Commission, April 1977.8.Comments on Regulatory Guide 1.99, Westinghouse Electric Corporation,'btained from NRC Public Document Room, Washington, D.C.9.Position on Regulatory Guide 1.99, Combustion Engineering Power Sys-tems, Obtained from NRC Public Document Room, Washington, D.C.10.ASTM E 185-73,"Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," 1975 Annual Book of ASTM Standards. 11.ASTM E 399-74,"Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Materials," 1975 Annual Book of ASTM Standards. 12.Witt, F.J., and Mager, T.R.,"A Procedure for Determining Bounding Values of Fracture Toughness KIc at Any Temperature," ORNL-TM-3894, October 1972.13."American Electric Power Service Corporation Donald C.Cook Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-8047, March 1973.14.ENDF/B-IV, Dosimetry Tape 412, Mat No.6417 (26-Fe-54), July 1974.15.Loss, F.J., Hawthorne, J.R., Serpan, C.Z., Jr., and Puzak, P.P.,"Analysis of Radiation-Induced Embrittlement Gradients on Fracture Characteristics of Thick-Walled Pressure Vessel Steels," NRL Report 7209, March 1, 1971.47 16.Telecon, E.B.Norris to Ken Hogue (NRC Staff)January 19, 1977.17.Hazleton, W.S., Anderson, S.L., and Yanichko, S.E.,"Basis for Heatup and Cooldown Limit Curves," WCAP-7924, July 1972.18.Donald C.Cook Unit No.1 Technical Specifications, as of November 30, 1977.48 APP END IX A TENSILE TEST RECORDS Southwest Research Institute Department of Materials Sciences TENSILE TEST DATA SHEET Test No.T-..l Spec.No.-1 Est.U.T.S.Initial G.L.PS1 r 41Z1~Machine No.Temperature I 4'F ts J Strain Rate,<2 tzpi>Initial Dia..I in.Inisial Thickness in.Date Initial Area 77 Initial Width in.Top Temperature Bottom Temperature Final Gage Length Final Diameter Final Area'F p 4T ine/~~I in.ine 2 0.2'%ffset Load 88 9 D lb 0.02%Offset Load Upper Yield Point lb Maximum Load 4 0 lb r Maximum Load Initial Area P 2 Init1al Ar ea psi cjoy 2-~g psi 0 02/Y S 0.02%Offset Load Initial Area PS1 Y S Upper Yield Point UPPer..I tial Area PS1 Final G.L.-Initial x 100=Initial Area-Final Area 1 p@~7 Initial Area Signature: A-2 -0;0 rZ i9ahJ A-3 Southwest Research Institute Department of Materials Sciences TENSILE TEST DATA SHEET Test No.T-.Z Est.U.T.S.psi Spec.No.Initial G.L..O in.Tem per afore~P'F rr/Strain Rate.C'~/W Initial Dia..g C'n.Initial Thickness in.Initial Area.+H/Initial VTidth in.Tap Temperature Bottom Temperature I'F Maximum Load S~7S lb 0 2%%uo Offset Load 5 2.=.~~lb Final Gage Length Final Diameter.l+J ln~0.02%%utf Offset Load Upper Yield Point lb lb Final Area.o'72 2 r Initial Ar ea 0.2%Offset Load Initial Area 0 02%%u Y S~02%%u'ff et Load Initial Area psl U er Yield Point PPer.-I tlal Area Final G.L.-Initial%%utl Elongation x 100'=~~'%%uo Initial Area-Final Area 100 Initial Area tt )~~a'0'0 g~<A-5 Southwest Research Institute Department of Materials Sciences TENSILE TEST DATA SHEET Test No.T-Spec.No.Est.U.T.S.Initial G.L.psi dd in.Machine No.)>/J~~Temperature >+'F Initial Dia.Initial T hie kne s s in.Date Initial Ar e a'~8 7 Initial Width in.Top Temperature oF Maximum Load 5.>G lb 0.2%Offset Load~~n,~>lb~s Final Gage Length Final Diamete" Final Area 111~in.sP/74+m.2 0.02%Offset Load Upper Yield Point lb lb Maximum Load 0.2'ls Offset Load g~gg.Initial Area 0 02$Y S=2/o Offset Load Initial Area'er Yield Point pp,~tel Area ps1 p81%u Fin G.L.-Initial G.L.%Elongation Initial G, L.%R A Initial Area-Final Area Initial Area Signature: A-6 'A-7 ~~1 Southwest Research Institute Department of Materials Sciences TENSILE TEST DATA SHEET Test No.T-Spec.No.Temperature 5 ft<'F Est.U.T.S.Initial G.L.Initial Dia.psi Project No.Machine No.Date 6<-a>>n-of"/Strain Rate Initial Thickness Initial Width 1ne Initial Area.OHg'7 Top Temperature 5~l~'F Maximum Load+C~~'0 lb Bottom Tempe ratur e o 8 4 0.2%Offset Load~?~.5 ib 0.02%Offset Load lb in.Upper Yield Point Final Area Maximum Load Initial Ar ea 0.2%%uo Offset Load Initial Area p 0 2%%u Y S 0.0 2%0 f f s e t L o a d Initxal Area ps1 U er Yield Point Initial Ar ea ps'inal G.L.-Initial G L lpp 0 E OIlgation-~.+l G L x-//7'Initial Area-Final Area Initial Area Signature: A-8 b,t, A-9 ~1)1 APPENDIX B PROCEDURE FOR THE GENERATION OF ALLOWABLE PRESSURE-TEMPERATURE LIMIT CURVES FOR NUCLEAR POWER PLANT REACTOR VESSELS PROCEDURE FOR THE GENERATION OF ALLOWABLE PRESSURE-TEMPERATURE LIMIT CURVES FOR NUCLEAR.POWER PLANT REACTOR VESSELS 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 pressure vessels~The safety margins employed in these procedures equal or exceed those recommended in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G,"Protection Against Nonductile Failure." B.Back round The basic parameter used to determine safe vessel operational conditions is the stress intensity factor, KZ, which is a function of the stress state and flaw configuration. The KI corresponding to membrane tension is given by KIŽm'm where Mm is the membrane stress correction factor for the postulated flaw and o.m the membrane stress.Likewise, KI corresponding to bend-ing is given by KIbŽb 0'b (2)where Mb is the bending stress correction factor and o.b is the bending stress.For vessel section thickness of 4 to 12 inches, the maximum B-2 postulated surface flaw, which is assumed to be normal to the direction of maximum stress, has a depth of 0.25 of the section thickness and a length of l.50 times the section thickness. Curves for Mm versus the square root of the vessel wall thickness for the postulated flaw are given in Figure 1 as taken from the Pressure Vessel Code (ref.Figure G-2114.1).These curves are a function of the stress ratio parameter r/r, where o.(Py is the material yield strength which is, taken to be 50,000 psi.The bending correction factor is defined as 2l3 Mm and is therefore determined from 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.The Code specifies the minimum KI that can cause failure as a func-tion of material temperature, T, and its reference nil ductility temperature, RTNDT.This minimum KI is defined as the reference stress intensity fac-tor, KIR, and is given by KIR=26777.+1223.exp 0.014493(T-RT+160)NDT (3)where all temperatures are in degrees Fahrenheit. A plot of this expression. is given in Figure 2 taken from the Code (ref.Figure G-2010.1).C.Pressure-Tem erature Relationshi s 1.Inservice Leak and H drostatic Test During performance of inservice leak and hydrostatic tests, the reference stress intensity factor, KIR, must always be greater than B-3 3.8 3.2 MEh<8RAHQ I(m M Im m~ra Mb x Mb<2/3hlm, 1.0 0.7 0.5 O.I 3.0 E 2.~i 2.2 2.0 1.6 1.2 1.0 1.0 1.2 I A 1.6 1,0 2.0 2.2 2.~i 2.6 2.8 3.0 3.2 3A 3.6 3.8 4.0 FIGURE 1.STRESS CORRECTION FACTOR I70 l30 I20 I I 0 Lg tco SO 70 60 50 40 I'R 26 777)V'IIERE RTHPT'EFEAFHCE STRESS INTENSITY FACTOR TEhIPERATURE AT VIHICH I'IR IS PERhIITTED,'F 'EFERFHCE HIL-DUCTILITY TEMPERATURE IO 0-240-200-IGO-I20-eO-40 0 40 80.I20 IGO 200 240 TEIAPERATUAE RELATIYE TO ATHP,(T-ATHPT), FAHREIIHEI D GREES FIGURE 2.REFERENCE STRESS INTENSITY FACTOR B-5 l.5 times the KZ caused.by pressure, thus l.5 Kl'p KZR (4)or'5 Mm<m~K1R (5)For a cylinder with inner radius ri and outer radius ro, the stress distribution due to internal pressure is given by With 1/4T flaws possible at both inner and outer radial locations, i.e., at rl/4=ri 4 1/4(ro-ri)and r3/<r j+3/4(ro-ri), the maximum stress will occur at the inner flaw location, thus I r j r+(1/4ro+3/4ri)4.2 o.=P max o ro2-ri2 (1/4ro p 3/4r i)2 With the operation pressure known, i.e., Po, we deter-mine the minimum coolant temperature that will satisfy Equation (4)by e valuating KlR='5Mm<max and determine the corresponding coolant temperature, T, from Equa-tion (3)for the given RT~~DT at the 1/4T location.For this calculation, Equation (3)takes the form I-*I-6..6.I[-666-'].S-6 The inservice curves are generated for an operating pres-sure range of~96 Po to l.14 Po, where Po is the design operating pres sur e.2.Heatu and C ooldown 0 e rations At all times during heatup and cooldown operations, the ref-erence stress intensity factor, K1R, must always be greater than the sum of 2 times the Klp caused by pressure and the Klt caused by thermal gra-dients, thus 2.0 Klp+l.0 Klt<KZR (10)or 2 0 Mm 0 max-K1R-KZt where o max is the maximum allowable stress due to internal pressure, and KZt is the equivalent linear stress intensity factor produced by the thermal gradients. To obtain the equivalent linear stress intensity fac-tor due to thermal gradients requires a detailed thermal stress analysis.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.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 this point in the vessel wall and, therefore, the steady state pressure stress would represent the maximum stress condition at the 1/4T location.At OUTER RAD IUS 3/4T Z/4T INNER RAD IUS Pressure stress distribution Thermal stress distribution (a)Heatup OUTER RAD IUS 3/4T 1/4T INNER RADIUS Pressure stress distribution Thermal stress distribution (b)Cooldown Figure 3.Heatup and Cooldown Stress Distribution B-8 the 3/4T flaw location, the pressure stress and thermal stress add and, therefore, the combination for a given heatup rate represents the maxi-mum stress at the 3/4T location.The maximum overall stress between the 1/4T and 3/4T location then determines the maximum allowable reac-tor 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 max(K1R r j ro+(1/4ro 0 3/4r;)2Mm roZ-rj (1/4ro+3/4rj)2 (12)where Mm is determined from the curves in Figure 1 and K1R is obtained from Equation (3)using the coolant temperature and RTNDT at the 1/4T location.Here we may note that Mm must be iterated for since it is a function of the final stress ratio to yield strength (0./ay).At the 3/4T location, the maximum pressure is determined from Equation (ll)as P (3/4T)-KZR-Ku r j r oZ+(1/4r j+3/41 o)2M roZ r.Z (1/4ri+3/4ro)2 (13)where K1R is obtained from Equation (2)using the material temperature and RTNDT at the 3/4T location and Klt is determined from the analysis procedure outlined in Section D.Mm is determined from Figure 1, B-9 The minimum of these maximum allowable pressures at the given coolant temperature determines the maximum operation pressure.Each heatup rate of interest must be analyzed on an individ-ual bas is.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 location 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 Equation (12).For each coo)down rate, the maximum pressure is evalu-ated at the 1/4T location from max(ri ro~+(3/4ri 0 1/4r o)2M r-r~(3/4ri+1/4r)(14)where KIR is obtained from Equation (3)using the material temperature and RTNDT at'the 1/4T location.KIt is determined from the thermal analysis 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 gradients 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 pro-duced by steady state analyses and transient thermal analysis to determine the minimum of the maximum allowable pressures. 3.Core 0 eration At all times that the reactor core is critical, the temperature must be higher than that required for inservice hydrostatic testing, 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 constructed directly from the inservice leak and.hydrostatic test and heatup analysis results.D.Thermal Stress Anal sis 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 PcT<-K[(1/r)T+T.1=o (15)subject to initial condition T(r,0)=T and boundary conditions-KTr(ri, t)=hLTc(t)-T(ri t)I (17) and Tr(roit)=0 (18)whe re Tc=To+Rt.(19)p is the material density, c the material specific heat, K the heat conduc-tivity of the material, h the heat transfer coefficient between the water coolant and vessel material, R the heating rate, To the initial coolant temperature, T(r, t)the temperature distribution in the vessel, r the spatial coordinate, and t the temporal coordinate. 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 6r apart, across the vessel are: for 1<n<N htK T=Ll-2(2-)JT QtK~gr+(g)Z L (1+-)Tn+1.+Tn-1J (2o)(21)B-12 andfor n=N t+()t N[pc(()r)Z J N pr())r)2 N-1 (22)For stability in the finite difference operation, we must choose ht for a given hr such that both 2(2+-)c 1 ()t K Zr pc(kr)2 r1 (23)and ht K (Ih,r~(1+)+C 1 pc(hr)rl pc(hr)(24)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/pc), 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 (E/pc)called the thermal diffusivity and E, the thermal conductivity, can be determined from the ASME Boiler and Pressure Ves-sel Code, Section III, Appendix I-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) (25)B-13 and k(T)"-(K/pc)=0.6942-0.000432~T (FT/HR)(26)where T is in degrees Fahrenheit. The heat transfer coefficient is calculated based on forced con-vection 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 conductivity of the coolant.For water coolant, allowance for the variations in physical properties with temperature may be made by writing~h(T)=170(1+10~T-10~T)v/D (27)where v is in ft/sec, D in inches, the temperature is in'F, and h is in Btu/hr-ft-'F.The values for the heat-transfer coefficient given by this 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 Q (Lbm/hr)and effective flow area A (ft).Given the relationship p(T)=62.93-0.48 x 10 2<'-T-0.46 x 10 4" T2 (28)for the density of water as a function of temperature, the mean velocity of the coolant is obtained from v=O/(3600>p (T)~A)(29)Glasstone, S., Princi les of Nuclear Reactor Engineerin, D.Van Nostrand Co., Inc., New Jersey, pp.667-668, 1960. The thermal stress distribution is calculated from r2+ri2 C ro aT(r,t)=t[3 j T(r,t)rdr-T(r,t)+ 3 (3 3)j T(r,t)rdrj (30)ri 0 1 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 Theor of Elasticit by Timoshenko and Goodier, pp.408-409, when im-posing 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 n and E are evaluated as a function of the average temperature across the vessel T=~(3 j T(r)rdr ri (31)The dependence of the coefficient of thermal expansion on temperature is taken to be a(T)=5.76 x 10-6+4.4 x 10-9 4 T (32)and the dependence of Young's modulus on temperature is taken to be E(T)=27.9142+2.5782 x 10~" T-6.5723 x 10 6 4 T (33)as 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 from the resulting moment.The moment produced by the nonlinear B-15 r~~stress distribution is given by ro M(t)=b f a T (r, t)rdr (34)where b is*unit depth of the vessel.Here we note that the moment is a function of time, i.e., coolant temperature via Tc=To+Rt.For a lin-ear stress distribution we have that P Mc~max=I (35')where 0 ax is the maximum outer fiber stress, c the distance from the neutral axis, taken to be (ro-ri)/2, and I the section area moment of inertia which is given by bh b(ro-r;)3 12 12 (36)Combining these expressions results in the equivalent linear stress due to thermal gradients ro rrttax rbt TJ't'T (r')r~(r.-r)J 1i (37)The thermal stress intensity factor KIt is then defined as KIt=Mb 0 bt (38)where Mb is determined from the curves given in Figure 1 wherein Mb=2/3 Mm.It is of interest to note that a sign change occurs in the stress calculations during a cooldown analysis since the thermal gradients produce compressive stresses at the vessel outer radius.This sign change must then be reflected in the Klt calculation for the cooldown analys is.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, by (T-Tavg)max (39)The thermal stress and equivalent linearized stress, as calculated by Equations (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 temperature and thermal stress distributions are developed during cool-down.The trends are nearly identical as those shown in Figure 4 when the inner and outer vessel locations are reversed with the I/4T location becoming the critical point.E.Exam le Calculations The following example is based on a reactor vessel with the follow-ing characteristics: Inner Radius Outer Radius Operating Pressure 82.00 in.(r)90 00 in.(r)2250 psig (Po) OUTER WALL 1.0 0.8 0.6 0.4 0.2//////-1.0 1.0-1.0 INNER WALL 1.0 Norma lized temperature distribution (4T/h,Tma) Normalized stress distribution (o/omax)Figure 4.Typical Normalized Temperature and Stress Distribution During Heatup Initial Temperature Final Temperature Effective Coolant Flow Rate 70'F (To)550'F 100 x 10 Lbm/hr (Q)Effective Flow Area 20.00 ft2 (A)Effective Hydraulic Diameter=10.00 in.(D)RTNDT (1/4T)RTNDT (3/4T)200OF 140'F 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 thermal analysis for the 100'F/hr heatup rate.The results of the computation are shown in Figures 5 through 9.Figure 5 gives the reference stress intensity factor, KIR, as a function of temperature indexed to RTNDT (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-perature at the 1/4T and 3/4T and thermal stress intensity factors Kzt are required to compute allowable pressure via Equations (13)and (14).The material temperatures versus coolant temperature during the 100'F/hr heatup and cooldown analyses are given in Figure 6.These temperatures allow computation of the corresponding reference stress intensity factors, KIR (3/4T)and KIR (1/4T).Figure 7 gives the corresponding thermal stress intensity factor at the 3/4T and 1/4T locations as a function of coolant tempe rature. 200 160 RTNDT(1i4T) -200 F~-120 hC I tV o 80 40 50 150 200 250 TEMPERATURE (F)300 350 400 Figure 5.Reference Stress Intensity Factor as a Function of Temperature Indexed to RTNDT(1/4T ) 400-100'F/HR HEATUP i 3/4T Location i--100'F/HR COOLDOWN (1/4T Location)300 200 100 50 100 150 200 250 COOLANT TEMPERATURE ('F)300 350 Figure 6.Vessel Temperature at 1/4T and 3/4T Locations as a Function of Coolant Temperature 10 6 cu hC-100'F/HR HEATUP (3/4T Location i--100'F/HR COOLDOWN (1/4 Location)50 10Q 150 200 250 COOLANT TEMPERATURE ('F)3QQ 350 Figure 7.Thermal Stress Intensity Factor at 3/4T and 1/4T Locations as a Function of Coolant Temperature
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 for possible instrumentation error.Figure 8 also shows 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 cor-responds to 1.1 times the operating pressure.The criticality 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 be greater than that required by the leak test limit.B-23 2400 LEAK TEST LIIIIIIT 2000 COMPOS ITE CURVE-100'F/HR HEATUP (Margins of+10 F and-60 psig for instrument error)1600 I 1200 STEADY STATE CR I TI CALI TY LIMIT 800 HEATUP 400 50 100 150 200 250 INDICATED TEMPERATURE (F)300 350 400 Figure 8.Pressure-Temperature Curves for 100 F/Hr Heatup 2400 2000 1600 COMPOSITE CURVE-100 F/HR COOLDOWN (Margins of+10 F and-60 psig for instrument error)CXI PJ 1200 CD Ch 800 COO LDOWN STEADY STATE 400 50 100 150 200 250 INDICATED TEMPERATURE ('F)300 350 Figure 9.Pressure-Temperature Curves for 100'F/Hr Cooldown
ADDENDUM TO FINAL REPORT ON"REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR DONALD C.COOK UNIT NO.1, ANALYSIS OF CAPSULE T" Plate B4406-3 Held Held ,30 ft-1b C Tem.'(de T)~(lan.)(Ttana.)Metal Mtt Correlation Monitor Irradiated Unirradiated. AT 65 5 60 90~.-10'0 20-90-100 70 80 120 105 45 60 Monitor Identification Fe-Top Fe-Top Mid.Fe-Mid.Fe-Bot.Mid.Fe-Bot.Cu*-Top Mid.Cu-Mid.Cu-Bot.Mid.Ni-Top Mid.Ni-Mid.Ni-Bot.Mid.Co-Top Co(Cd)-Top Co--Bot.Co(Cd)-Bot.U-238 NP-237 Height~(m)18.2 15.3 17.2 16.6 16.4 64.9 62.9 70.9 22.9 25.5 24.5 9.3 8.7 9.5 7.7 12.0(a)20.0(a)(a)As reported in WCAP-8047. i ADDENDUM NO.2 TO FINAL REPORT ON"REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR DONALD C.COOK UNIT NO.1, ANALYSIS OF CAPSULE T" Additional Tensile Test Data Specimen No.Fracture Load si 64,700 63,250~Fracture Stress 188,600 177,000 Uniform Elongation< >%%u4 5.00 2.45 W9 87,600 757800 250,000 193,700 4.56 2.87 (a)Using method of change in cross-sectional area of unnecked portion of specimen per ASTM E 184-62.}}