ML20094B108
| ML20094B108 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/30/1984 |
| From: | Boggs R, Cheney C, Vanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20094B080 | List: |
| References | |
| WCAP-10509, NUDOCS 8408060231 | |
| Download: ML20094B108 (85) | |
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WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION WCAP-10509 ANALYSIS OF CAPSULE T FROM THE TENNESSEE VALLEY AUTHORITY SEQUOYAH UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM R.S.Boggs S. E. Yanichko C.A.Cheney W. T. Kaiser April 1984 Work performed under Shop Order No. UEN-6620
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APPROVED: 'l' f /$'0ffC I O' T. R. Mager, MNager Metallurgical and NDE Analysis Prepared by Westinghouse for the Tennessee Valley Authority Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.
WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 8400060231 840731 PDR ADOCK 05000328 P pm
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PREFACE This report has been technically reviewed and verified.
Reviewer Sections 1 through 5 and 7 M. K. Kunka - 7'*/'~
Section 6' Appendix A S. L. Anderson F. J. Witt 7 /,Idh[d M
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i TABLE OF CONTENTS Section Title Page 1
SUMMARY
OF RESULTS 1-1 2 INTRODUCTION 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE T 5-1 5-1. Overview 5-1 5-2. Charpy V-Notch Impact Test Results 5-3 5-3. Tension Test Results 5-3 5-4. Wedge Opening Loading Tests 5-4 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY G-1 6-1. Introduction 6-1 g
6-2. Discrete Ordinates Analysis 6-1 6-3. Neutron Docimetry 6-6 6-4. Transport Arialysis Results 6-11 6 5. Dosimetry Results 6-21 7 SURVElLLANCE CAPSULE REMOVAL SCHEDULE 7-1 8 REFERENCES 8-1 7595B.1bC32384 V
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TABLE OF CONTENTS (cont)
Section Title Page Appendix A HEATUP AND COOLDOWN LIMIT CURVES A-1 FOR NORMAL OPERATION A-1. Introduction A-1 A-2. Fracture-Toughness Properties A-1 A-3. Criteria For Allowable Pressure-Temperature A-5 Relationships A-4. Heatup and Cooldown Limit Curves A-7
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LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the 4-3 Sequoyah Unit 2 Reactor Vessel (Updated Lead Factors for Capsules Shown in Parentheses) 4-2 Capsule T Diagram Showing Location of Specimens, 4-5 Thermal Monitors, and Dosimeters 5-1 Irradiated Charpy V-Notch Impact Properties for 5-15 Sequoyah Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Axial Orientation) 5-2 Irradiated Charpy V-Notch Impact Properties for 5-16 Sequoyah Unit 2 Reactor Vessel Intermediate Shell Forging 05 (Tangential Orientation) 5-3 Irradiated Charpy V-Notch Impact Properties for 5-17 Sequoyah Unit 2 Reactor P.:=ure Vessel Weld Metal 5-4 Irradiated Charpy V-Notch Impact Properties for 5-18 Sequoyah Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-5 Charpy impact Specimen Fracture Surfaces for 5-19 Sequoyah Unit 2 Pressure Vessel Intermediate Shell Forging 05 (Axial Orientation) 5-6 Charpy impact Specimen Fracture Surfaces for 5-20 Sequoyah Unit 2 Pressure Vessel Intermediate Shell Forging 05 (Tangential Orientation) 5-7 Charpy impact Specimen Fracture Surfaces for 5-21 Sequoyah Unit 2 Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for 5-22 Sequoyah Unit 2 Weld Heat Affected Zone Metal 75958:1bO32684 Vii
LIST OF ILLUSTRATIONS (cont)
Figure Title Page 5-9 Comparison of Actual versus Predicted 30 ft Ib 5-23 Transition Temperature increases for the Sequoyah Unit 2 Reactor Vessel Material based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 5-10 Tensile Properties for Sequoyah Unit 2 Reactor 5-24 Vessel Intermediate Shell Forging 05 (Axial Orientation) 5-11 Tensile Properties for Sequoyah Unit 2 Reactor 5-25 Vessel Weld Metal 5-12 Fractured Tensile Specimens of Sequoyah Unit 2 5-26 Reactor Vessel Intermediate Shell Forging 05 (Axial Orientation) and Weld Metal 5-13 Typical Stress-Strain Curve for Tension Specimens 5-27 6-1 Sequoyah Unit 2 Reactor Geometry 3-2 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-4 6-3 Calculated Azimuthal Distribution of Maximum 6-12 Fast-Neutron Flux (E > 1.0 MeV) within the Pressure Vessel Surveillance Capsule Geometry 6-4 Calculated Radial Distribution of Maximum 6-13 Fast-Neutron Flux (E > 1.0 MeV) within the Pressure Vesse!
6-5 Relative Axial Variation of Fast-Neutron Flux 6-14 (E > 1.0 MeV) within the Pressure Vessel 6-6 Calculated Radial Distribution of Fast-Neutron 6-15 Flux (E > 1.0 MeV) within the Reactor Vessel Surveillance Capsules viii 7595B:1b 032384
[ . _ _ _ _ - _ _
l LIST OF ILLUSTRATIONS (cont)
Figure Title Page 6-7 Calculated Variation of Fast-Neutron Flux Monitor 6-16 ,
Saturated Activity within Capsules Located at 40 Degrees 6-8 Calculated Variation of Fast-Neutron Flux Monitor 6-17 Saturated Activity within Capsules Located at 4 Degrees A-1 Pre'dicted Adjustment of Reference Temperature, A-2 ARTNDT as a Function of Fluence and Copper Content A-2 Fast Neutron Fluence (E > 1.0 MeV) as a Function A-3 of Full Power Service Life A-3 Sequoyah Unit 2 Reactor Coolant System A-8 Heatup Limitations Applicable up to 9 EFPY A-4 Sequoyah Unit 2 Reactor Coolant System A-9 Cooldown Umitations Applicable up to 9 EFPY 75958:1bt)32384 IX
LIST OF TABLES Tabis Title Page 4-1 Chemical Composition and Heat Treatment of the 4-4 Sequoyah Unit 2 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the Sequoyah Unit 2 5-5
- ntermediate Shell Forging 05 (Axial Orientation)
Irradiated at 550'F, Fluence 2.20 x 10'8 n!cm2 (E > 1 MeV) 5-2 Charpy V-Notch Impact Data for the Sequoyah Unit 2 5-6 Intermediate Shell Forging 05 (Tangential Orientation)
Irradiated at 550 F, Fluence 2.20 x 10 (E > 1 MeV) 5-3 Charp. V-Notch Impact Data for the Sequoyah Unit 2 5-7 Pressure Vessel Weld Metal inadiated at 550 F, Fluence 2.20 x 10 n/cm2 (E > 1 MeV) 5-4 Charpy V-Notch Impact Data for the Sequoyah Unit 2 5-8 Pressure Vessel Weld Heat Affected Zone Metal Irradiated at 550 F, Fluence 2.20 x 10 (E > 1 MeV) 5-5 Instrumented Charpy Irnpact Test Results for 5-9 Sequoyah Unit 2 Intermediate Shell Forging 05 (Axial Orientation) 5-6 Instrumented Charpy impact Test Results for the 5-10 Sequoyah Unit 2 Intermediate Shell Forging 05 (Tangential Orientation) 5-7 Instrumented Charpy impact Test Results for 5-11 Sequoyah Unit 2 Weld Metal 5-8 Instrumented Charpy impact Test Results for 5-12 Sequoyah Unit 2 Heat Affected Zone Metal 5-9 The Effect of 550 F Irradiation at 2.20 x 10 5-13 (E > 1 MeV) on the Notch Toughness Properties of the Sequoyah Unit 2 Reactor Vessel Materials 7595D:1tr032384 xi l
LIST OF TABLES (cont.)
Table Title Page 5-10 Tensile Properties for Sequoyah Unit 2 Reactor Vessel 5-14 Material Irradiated to 2.20 x 10m n/cm2 47 Group Energy Structure 6-5 6-1 6-2 Nuclear Constants for Neutron Flux Monitors 6-7 ..
Contained in the Sequoyah Unit 2 Surveillance Capsules 6-3 Calculated Fast-Neutron Flux (E > 1.0 MeV) and 6-18 E Lead Factors for Sequoyah Unit 2 Surveillance
? Capsules 6-4 Calculated Neutron Energy Spectra at the Center 6-19
.- of Sequoyah Unit 2 Surveillance Capsules T 6-5 Spectrum-Averaged Reaction Cross Sections at the 6-20 Centar of Sequoyah Unit 2 Surveillance Capsules Irradiation History of Sequoyah Unit 2 Reactor 6-22 6-6 Vessel Surveillance Capsule T 4*i"*.
1 Comparison of Measured and Calculated Fast-Neutron 6-23 d7;5N'- %*
6-7 Flux Monitor Saturated Activities for Capsule T 6-8 Results of Fast Neutron Dosimetry for Capsule T 6-24 6-9 Results of Thermal-Neutron Dosimetry for Capsule T 6-25 6-10 Summary of Neutron Dosimetry Results for Capsule T 6-25 A-1 Sequoyah Unit 2 Reactor Vessel Toughness Data A-4 Xi 75958 lb O32384
SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in Capsule T, the first surveillance capsule to be removed from the Tennessee Valley Authority Sequoyah Unit 2 reactor prcoacn vessel, led to the following conclusions:
a Irradiation of the reactor vessel intermediate shell forging 05, to 2.20 x 10'8 n/cm, resulted in both 30 and 50 ft-lb transition temperature increases of 25'F for speci-mens oriented normal to the major working direction (axial orientation) and 60'F for specimens oriented in the major working direction (tangential orientation).
m Weld metal irradiated to 2.20 x 10'8 n/cm2 resulted in a 30 and 50 ft-Ib trancition temperature increase of 80 and 75'F, respectively.
e The average upper shelf energy of the limiting forging (05) decreased from 88 to 82 ft-lbs and the limiting weld metal decreased from 112 to 110 ft-lbs. Both materials exhibit a more than adequate shelf level for continued safe plant operation.
a Comparison of the 30 ft-Ib transition temperature increases for the Sequoyah Unit 2 surveillance material with predicted increases using the methods of NRC Regu-latory Guide 1.99, Revision 1, shows that the weld metal transition temperature increase was greater than predicted. Since the transition temperature increase was greater than predicted, the future operating limits for the vessel, shown in Appen-dix A, were based on a predicted trend curve which passed through the actual increase in transition temperature as determined from the irradiated weld metal specimens.
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, :As a SECTION 2 INTRODUCTION.
This report presents the results of the examination of Capsule T, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on Sequoyah Unit 2 reactor pressure vessel materials under actual . .
operating conditions.
The surveillance program for Sequoyah Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Yanichko.DI The surveillance program was planned to cover -
the 40-year design life of the reactor pressure vessel and was based on ASTM E-185-73," Recommended Practice for Surveillance Tests for Nuclear Reactors."tal Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for ..
removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes testing and the postirradiation data obtained from surveillance Capsule T removed from Sequoyah Unit 2 reactor vessel, and discusses the analysis of these data.
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SECTION 3 -
BACKGROUND $
coolant r sist fra ure co sttutes n m rtan f c o in enst.nng safety n t e n a industry. The beltline region of the reactor pressure vessel is the most critical region of y '
the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure ,
vessel steels such as A508 Class 2 (base material of the Sequoyah Unit 2 reactor pressure f vessel beltlina) are well documented in the literature. Generally, low alloy ferritic materials -
show an increase in hardness and tensile properties and a e crease in ductility and [,
toughness under certain conditions of irradiation. v A method for performing analyses to guard against fast fracture in reactor pressure vessels _-
has been presented in." Protection Against Non-ductile Failure," Appendix G to Section i Ill of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics r concepts and is based on the reference nil-ductility temperature RTNDT- 4 RT NDT is defined as the greater of either the drop weight nil-ductility transition temperature h (NDTT per ASTM E-208) or the temperature 60 F less than the 50 ft-lb (and 35-mil lateral L expansion) temperature as determined from Charpy specimens oriented normal (trans-verse) to the major working direction of the material. The RTNDT of a given material is ,
used to index that material to a reference stress intensity factor curve (KIR curve) which -
appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, t crack arrest, and static fracture toughness results obtained from several heats of pressure u vessel steel. When a given material is indexed to the KIR curve, allowable stress intensity $
factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to w account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel [
steel can be monitored by a reactor surveillance program such as the Sequoyah Unit 2 }
Reactor Vessel Radiation Surveillance Program,m in which a surveillance capsule is
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periodically removed from the operating nuclear reactor and the encapsulated specimens
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are tested. The increase in the average Charpy V-notch 30 ft-Ib temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the material to the KIR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
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Eight surveillance capsules for monitoring the effects of neutron exposure on the Sequoyah . Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel "fl ( j, prior to initial plant startup. The capsules were positioned in the reactor vessel between i'.; the thermal shield and the vessel wall at locations shown in Figure 4-1. The vertical center G5? of the capsules is opposite the vertical center of the core. <Y f: ye; Capsu'e T was removed after 1.04 effective full power years of plant operation. This capsule contained Charpy V-notch impact, tensile, and WOL specimens (Figure 4-2) from 44 lp. 4 . . the intermediate shell forging 05 and submerged arc weld metal representative of the g p{[ core region of the reactor vessel and Charpy V-notch specimens from weld heat-affected f,j. zone (HAZ) material. ? 'i ,. i The chemistry and heat treatment of the surveillance material are presented in Table = 4-1. The chemical analyses reported in Table 4-1 were obtained from unirradiated material used in the surveillance program. In addition, a chemical analysis was performed on an irradiated Charpy specimen from the weld metal and is reported in Table 4-1. All test specimens were machined from the 1/4 thickness location of the forgings. Test specimens represent material taken at least one forging thickness from the quenched end of the forging. Charpy specimens were machined from the forging in both the tangential (longitudinal axis of specimen parallel to the major working direction) and axial (longitudinal axis of the specimen perpendicular to the major working direction) orientations. Tensile specimens were machined from the forging with the longitudinal axis of the specimen perpendicular to the major working direction. Charpy V-notch and tensile specimens from the weld metal were oriented with the lon-gitudinal axis of the specimens transverse to the welding direction. Capsule T contained dosimeter wires of pure iron, copper, nickel, and aluminum-cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Nep-tunium (Np237) and Uranium (U ) were contained in the capsule and located as shown in Figure 4-2. nnamom84 4-1
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2. The two eutectic alloys and their melting points are: 2.5% Ag,97.5% Pb Melting Point 579'F (304 C) 1.75% Ag,0.75% Sn,97.5% Pb Melting Point 590 F (310*C) 42 75790.tb O31984
5396A 1 1 REACTOR VESSEL 0 270 THERMAL SHIELD (3.17) X , CAPSULE / [ - Y (3.17) (TYP) (1.02) W Z (1.02) 400 40 40 400 1800
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(1.02) V (3.17) U 900 Figure 4-1. Arrangement of Surveillance Capsules in the Sequoyah Unit 2 Reactor Vessel (Updated Lead Factors for Capsules Shown in Parentheses) nr90inost984 4-3
TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE SEQUOYAH UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS
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1 Chemical Composition (WT-%) Forging 05 Rotterdam Heat No. 288757/981057 Dockyard Weld Metal Element Westinghouse Analysistal Analysis Westinghouse Analysistal C 0.18 0.19 0.095 S 0.018 0.013 0.013 N2 0.009 - 0.012 Co 0.001 - 0.001 Cu 0.13 - 0.13 Si 0.27 0.22 0.41 Mo 0.64 0.57 0 53 Ni 0.74 0.78 0.11 Mn 0.72 0.70 1.50 Cr 0.33 0.34 0.085 V 0.022 <0.01 0.002 P 0.018 0.014 0.016 Sn 0.002 - 0.002 At 0.027 - 0.009 (a] M elements not listed are less than 0.010 weight *'. Heat Treatment Heat Treatment Material Temperature Time (hr) Coolant (*F) Intermediate 1675'F : 25'F 3 1/2 Water Quenched Shell Forging 05 1225'F z 25'F 9 Furnace cooled to 815'F Heat No. 288757/981057 1130 F 25'F 20 1/2 Furnace cooled Weldment 1130 F : 25'F 14 3/4 Furnace cooled 4-4 nne te osiss4
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s SURVEILLANCE Np 237 U 238 9 DOSIMET TENSILE CHARPY CHARPY BLOC WOL WOL WOL WOL TENSILE W NT W NT W10 60 60 58 58 NT10 W1 198 W4 W3 W2 W W9 W NT NT NT9 59 59 57 57
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TO TOP OF VESSEL SPECIMEN NUMBERING CODE .- W = WELD METAL NT - FORGING 05 ( AXI AL ORIENTATION) i NL - FORGING 05 (T ANGENTI AL ORIENT ATION) H = HEAT AFFECTED ZONE f 4 I Wa eM l 5__.___ . . .
5396A 2 i-NLE r T CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY 0 NT W NT W NT W NT H NL H NL H NL H NL H H l
- lis 56 54 54 52 52 50 50 60 40 58 38 56 3ti 54 34 52 50 I
W NT W NT W NT W NT H NL H NL H NL H NL 'H H l 56 56 53 53 51 51 49 49 59 39 57 37 56 35 53 33 51 49 , ~ s i n JL 579'F MONITOR
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Figure 4-2. Arrangement of Specimens, Thermal Monitors, and Dosimeters in Surveillance Capsule T 4-5
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SECTION 5 - TESTING OF SPECIMENS FROM CAPSULE T 7 5-1 OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accor- - dance with 10CFR50, Appendices G and H, ASTM Specification E185-82 and Westing- : house Procedure MHL 7601, Revision 3 as modified by RMF Procedures 8102 and 8103. Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully r3 moved, inspected for identification number, and checked against the master l' list in WCAP-8513.m No discrepancies were found. Examination of the two low-melting 304 C (579'F) and 310 C (590 F) eutectic alloys indicated no ndting of either type of inermal monitor. Based on this examination, the ;. maximum temperature to which the test specimens were exposed was less than 304 C (579 F). The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Pro-cedure 8103 on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED ). From the load-time curve, the load of general yielding (Pey), the time to general yielding (tGY), the maximum load (P M ), and ' the time to maximum load (tM) can be determined. Under some test conditions, a sharp t drop in load indicative of fast fracture was observed. The load at which fast fracture was 2 initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated is identified as the arrest load (PA)- l The energy at maximum load (EM) was determined by comparing the energy time record 3? and the load-time record. The energy at maximum load is approximately equiva;ent to , the energy required to initiate a crack in the specimen. Therefore, the propagetion energy for the crack (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load.
't 75958. Ib 032684 51
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The yield stress (ay) is ca!culated from the three point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also using the three point bend formula. Percentage shear was determined from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification. Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-81 and E21-79. and RMF Procedure 8102. All pull rods, grips, and pas were made of inconel 718 hardened to Rc45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test. Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67. Elevated test temperatures were obtained with a three-zone electric resistance split-tuba furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature. Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and con-troller temperatures was developed over the range room temperature to 550 F (288 C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments indi-cated that this method is accurate to plus or minus 2 F. The yield load, ultimato load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement. 5-2 7soss n o m s4
5.2. CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials contained .. in Capsule T irradiated at 2.20 x 10'8 n!cm2 are presented in Tables 5-1 through 5-8 and Figures 5-1 through 5-4. A summary of the transition temperature increases and upper i shelf energy decreases for the Capsule T material is shown in Table 5-9. : Irradiation of vessel inurmediate shell forging 05 material specimens (axial orientation) to 2.20 x 10'8 n/cm2 (Figure 5-1) resulted in both 30 and 50 ft-lb transition temperature increases of 25'F and an upper shelf -nergy decrease of 6 ft-lb. Irradiation of vessel intermediate shell forging 05 material (tangential orientation) to 2.20 x 10'8 n'cm2 (F gure 5-2) resulted in both 30 and 50 ft-lb transition temperature increases of 60 F and an upper shelf energy decrease of 16 ft-lb. Weld metal irradiated to 2.20 x 10'8 n/cm2 (Figure 5-3) resulted in both 30 and 50 ft-lb transition temperature increases of 80 F and 75"F, respectively, and an upper shelf energy decrease of 2 ft-Ib. Weld HAZ metal irradiated to 2.20 x 10's n/cm2 (Figure 5-4) resulted in both 30 and 50 ft-lb transition temperature increases of 50 F respectively and an upper shelf energy decrease of 2 ft-lb. The fracture appearance of each irradiated Charpy specimen frem the various materials is shown in Figures 5-5 through 5-8 and show an increasing ductile or tougher appearance , with increasir,g test temperature. Figure 5-9 shows a comparison of the 30 ft-lb transition temperature increases for the various Sequoyah Unit 2 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1P1 This comparison shows that the transition temperature increases resulting from irradiation to 2.20 x 10'8 n/cm2 are less than predicted by the Guide for the axial and tangential orientations of forging 05. The weld metal transition temperature increase resulting from 2.20 x 10'8 n/cm2 is greater than predicted by the Guide. Therefore, a trend line passing through the actual increase in transition temperature as determined from the irradiated weld specimens was used to set future operating limits 3 for the vessel (Appendix A). , 5-3. TENSION TEST RESULTS The results of tension tests performed on forging 05 (axial orientation) and weld metal irradiated to 2.20 x 10's n/cm2 are shown in Table 5-10 and Figures 5-10 and 5-11, ? respectively. These results show that irradiation produced an increase in 0.2 percent yield strength of 7 to 8 ksi for forging 05 and approximately 6 ksi for the weld metal. Fractured tension specimens for each of the materials are shown in Figure 5-12. A typical stress-strain curve for the tension specimens is shown in Figure 5-13. .. 75958 IbO3Nm 5-3
i 5-4. WELD OPENING LOADING TESTS Test results for weld opening loading (WOL) fracture mechanics specimens contained in Capsule T will be reported at a later time.
I TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE SEQUOYAH UNIT 2 INTERMEDIATE SHELL FORGING 05 (AXlAL ORIENTATION) lRRADIATED AT 550'F, FLUENCE 2.20 x 10" n/cm2 (E > 1 MeV) Temperature Impact Energy Lateral Expansion No. C (F) Joules (ft-Ib) MM (mils) (%) NT54 - 32 (-25) 24.5 (18.0) 0.42 (16.5) 7 NT55 -18 ( 0) 36.5 (27.0) 0.62 (24.5) 12 NT53 -4 ( 25) 39.5 (29.0) 0.58 (23.0) 18 NT52 -4 ( 25) 27.0 (20.0) 0.51 (20.0) 16 NT59* 10 ( 50) NT50 24 ( 75) 51.5 (38.0) 0.90 (35.5) 43 NT60 24 ( 75) 47.5 (35.0) 0.85 (33.5) 31 NT56 38 ( 100) 87.0 (64.0) 1.40 (55.0) 70 NT58 66 ( 150) 101.5 (75.0) 1.45 (57.0) 94 NTS1 93 ( 200) 110.0 (81.0) 1.50 (59.0) 100 NT49 121 ( 250) 111.0 (82.0) 1.82 (71.5) 100 NT57 149 ( 300) 114.0 (84.0) 1.68 (66.0) 100
- Specimen was not centered on Anyd s458 morss4 55
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE SEQUOYAH UNIT 2 INTERMEDIATE SHELL FORGING 05 (TANGENTIAL ORIENTATION) lRRADIATED AT 550 F, FLUENCE 2.20 x 10" n/cm2 (E > 1 MeV) Temperature Impact Energy Lateral Expansion No. C (F) Joules (ft-Ib) MM (mils) (%) NL38 -46 (- 50) 12.0 ( 9.0) 0.19 ( 7.5) 3 NL36 - 32 (- 25) 12.0 ( 9.0) 0.23 ( 9.0) 5 NL33 -18 ( 0) 55.5 ( 41.0) 0.71 (28.0) 18 NL35 -4 ( 25) 69.0 ( 51.0) 1.00 (39.5) 28 NL34 38 ( 100) 88.0 ( 65.0) 1.26 (49.5) 50 NL37 66 ( 150) 143.5 (106.0) 1.96 (77.0) 85 NL39 93 ( 200) 161.5 (119.0) 2.10 (82.5) 100 NL40 149 ( 300) 158.5 (117.0) 2.26 (89.0) 100 5-6 nee inoaaa4
TABLE 5-3 CHARPY V-NOTCH IMPACT DATA FOR THE SEQUOYAH UNIT 2 PRESSURE VESSEL WELD METAL IRRADIATED AT 550 F, FLUENCE 2.20 x 10" n/cm2 (E > 1 MeV) Temperature impact Energy Lateral Expansion No. C (F) Joules (ft-Ib) MM (mils) (%) W58 - 32 (- 25) 51.5 ( 38.0) 0.85 (33.5) 48 W49 -32 (- 25) 85.5 ( 63.0) 1.21 (47.5) 25 W53 -18 ( - 0) 19.0 ( 14.0) 0.29 (11.5) 24 W55 -4 ( 25) 42.0 ( 31.0) 0.72 (28.5) 29 W50 -4 ( 25) 20.5 ( 15.0) 0.42 (16.5) 25 W52 10 ( 50) 130.0 ( 96.0) 2.17 (85.5) 96 W54 24 ( 75) 96.5 ( 71.0) 1.51 (59.5) 48 W59 38 ( 100) 42.0 ( 31.0) 0.67 (26.5) 55 W60 52 ( 125) 154.5 (114.0) 2.25 (88.5) 100 W51 66 ( 150) 153.0 (113.0) 2.13 (84.0) 100 W56 93 ( 200) 158.5 (117.0) 2.17 (85.5) 100 W57 149 ( 300) 133.0 ( 98.0) 2.16 (85.0) 100 7s9se. m o m a4 5-7
l' TABLE 5-4 CHARPY V-NOTCH IMPACT DATA FOR THE SEQUOYAH UNIT 2 PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL IRRADIATED AT 550'F, FLUENCE 2.20 x 10 n/cm2 (E > 1 MeV) Temperature Impact Energy Lateral Expansion No. C (F) Joules (ft-Ib) MM (mils) (%) H50 - 32 (- 25) 28.5 ( 21.0) 0.33 (13.0) 11 H51 -18 ( -0) 55.5 ( 41.0) 0.75 (29.5) 49 H49 -18 ( 0) 80.0 ( 59.0) 0.76 (30.0) 69 H53 -4 ( 25) 81.5 ( 60.0) 1.03 (40.5) 66 H59 10 ( 50) 96.5 ( 71.0) 1.21 (47.5) 67 H54 24 ( 75) 111.0 ( 82.0) 1.38 (54.5) 78 H60 38 ( 100) 152.0 (112.0) 1.64 (64.5) 93 H57 52 ( 125) 57.0 ( 42.0) 0.91 (36.0) 63 H52 66 ( 150) 161.5 (119.0) 1.94 (76.5) 99 H56 93 ( 200) 162.5 (120.0) 1.83 (72.0) 100 H55 149 ( 300) 167.0 (123.0) 1.93 (76.0) 100 H58 149 ( 300) 118.0 ( 87.0) 1.85 (73.0) 100 5-8 75958. ib 032684
= 6 0 I TABLE 5-5 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR SEQUOYAH UNIT 2 INTERMEDIATE SHELL FORGING 05 (AXIAL ORIENTATION) Normalized Energies Test Charpy Charpy Maximum Prop Yield T!me Maximum Time to Fracture Arrest Yield Flow Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress Gample Temp. Energy Ed/A ( Sec) (N) ( Sec) (N) (N) (MPa) (MPa) No. ('C) (Joules) (kJ/m') (kJ/m2) (kJ/m') (N) 68 14,000 90 17,600 285 17,000 0 719 811 NT54 - 32 24.5 305 237 369 88 13,900 125 18,500 450 18,000 0 714 832 NT55 -18 36.5 458 198 141 13,500 85 16,100 260 15,900 400 692 760 NT52 -4 27.0 339
-4 39.5 491 388 103 15,100 90 18,500 420 18,500 100 T!6 865 NT53 NT59 10 384 209 14,400 100 18,600 430 18,200 4200 74u 848 NT60 24 47.5 593 376 268 14,100 90 18,000 425 18,000 4100 728 826 NT50 24 51.5 644 470 614 14,600 90 19,000 500 17,200 8300 751 865 NT56 38 87.0 1085 460 811 14,200 90 18,600 500 13,800 8900 729 843 NT58 66 101.5 1271 430 943 12,800 95 17,500 510 11,700 8400 656 779 NTS1 93 110 0 1373 411 978 10.600 90 16,700 520 555 707 NT49 121 111.0 1390 436 987 11,200 70 17,700 505 575 743 NT57 149 114.0 1424
L I I b l g ,; g
TABLE 54 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR SEQUOYAH UNIT 2 INTERMEDIATE SHELL FORGING 05 (TANGENTIAL ORIENTATION) Normelland Eswrgpos Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed'A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress No. ('C) (Joutes) (kJ/m2) (kJ W ) (kJ/m2) (N) ( Sec) (N) (pSec) (N) (N) (MPa) (MPa) 9 8 70 15,500 90 16,700 125 16,900 0 795 827 NL38 - 46 12.0 153 82 NL36 - 32 12.0 153 82 70 14,700 85 15,700 125 15,600 0 754 780
- NL33 -18 55.5 695 602 93 16,400 90 20,800 580 20,700 0 844 957 NL35 -4 69.0 864 539 325 14,100 100 19.000 585 18,300 0 726 853 NL34 38 88.0 1102 613 489 14,200 90 19,200 645 19,000 5600 731 859 NL37 66 143.5 1796 601 1195 13,200 85 18,500 655 13,800 8900 678 816 NL39 93 161.5 2017 577 1440 13,100 90 18.200 640 672 805 NL40 149 158.5 1983 729 1254 11,100 120 16,700 700 572 715 s
3 n f TABLE 5-7 INSTRURAENTED CHARPY IRAPACT TEST RESULTS FOR , SEQUOYAH UNIT 2 WELD RAETAL Normakaed Energies Test Charpy Charpy beanimum Prop Yield Time leanimum Time to Fracture A. M Yield Flow Sample Temp. Energy Ed'A Em'A Ep/A Load to Yield Load b8animum Load Load Stress Stress No. [C) (Joules) (kJ w ) (kJ'm*) (kJw) (N) (pSec) (N) (pSec) (N) (N) (ISPs) (RAPa) y W49 - 32 85.5 1068 577 491 15,400 95 19,100 600 18,100 200 791 887 W58 - 32 51.5 644 515 129 13,700 80 17,700 570 17,600 0 706 808 W53 -18 19.0 237 197 41 14,900 85 17,100 235 17,100 100 765 823 W50 -4 20.5 254 192 62 13.400 85 15,800 250 15,800 1200 688 751 W55 -4 42.0 525 377 118 13,900 115 17,800 445 17,100 0 713 815 W52 ,1G 130.0 10.27 593 1034 13,300 100 17,800 675 10,700 0 683 800 W54 24 96.5 1203 595 E03 13,300 '125 17,800 700 14,800 2600 685 80t W59 38 42.0 525 383 143 13,400 85 17,000 445 16,900 4300 688 731 W60 52 154.5 1932 622 1310 11,800 90 16,100 765 605 718 W51 66 153.0 1915 574 1341 11.900 80 16,800 675 613 738 W56 93 158.5 1983 624 1359 12,300 90 17,700 705 635 ,71 W57 149 133.0 1661 480 1181 10.000 85 14,600 655 516 634
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a E TABLE 5-8 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR SEQUOYAH UNIT 2 WELD HEAT AFFECTED ZONE METAL Normalized Energies Arrest Yield Flow Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress Sample Temp. Energy Ed/A ( Sec) (N) ( Sec) (N) (N) (MPa) (MPa) ( C) (Joules) (kJ/m') (kJ/m') (kJ/m') (N) No. 337 19 16,900 90 20,600 335 20,600 0 869 964 H50 - 32 28.5 356 m i(,,300 21,100 665 20,700 5900 840 962 307 90 y H49 -18 80.0 1000 693 298 15,700 105 19,500 430 19,400 5000 809 906 HS1 -18 55.5 695 397 527 490 14,000 85 19,000 560 17,000 200 718 848 H53 -4 81.5 1017 482 721 15,300 90 19,400 500 18,100 9100 786 893 H59 10 96.5 1203 555 834 14,800 90 19,500 575 12,000 5500 763 882 H54 24 111.0 1390 028 1270 15,200 90 20,200 630 13,700 6200 782 910 H60 38 152.0 1898 472 240 14,500 95 19,200 505 18,700 4900 748 869 H57 52 57.0 712 1384 13,800 90 19,500 665 712 858 H52 66 161.5 2017 632 739 1295 13,600 85 19,900 755 702 864 H56 93 162.5 2034 481 994 10,500 85 16,700 585 540 700 H58 149 118.0 1474 643 1441 13,100 95 18,200 715 674 805 H55 149 167.0 2085 a N E i
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G h 6 f TABLE 5-9 EFFECT OF 550 F 1RRADIATION AT 2.20 x 10 (E> 1 MeV) ON THE NOTCH TOUGHNESS PROPERTIES OF THE SEQUOYAH UNIT 2 REACTOR VESSEL MATERIALS Average Average 35 mit Average Average Energy Absorption 30 ft-lb Temp (*F) Lateral Expansion Temp (*F) 50 ft-lb Temp ( F) at Full Shear (ft-ib) Material Unirradiated irradiated AT Unitradiated irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated A (ft-lb)
$ Forging 05 0 25 25 20 50 30 60 85 25 88 82 6 cAxial)
Forging 05 - 70 -10 60 -45 20 65 - 25 35 60 134 118 16 (Tangential) Weld Metal - 75 5 80 - 50 15 65 - 40 35 75 112 110 2 HAZ Metal - 60 -10 50 - 25 15 40 - 30 20 50 122 120 2
TABLE 5-10 TE:NSILE PROPERTIES FOR SEQUOYAH UNIT 2 REACTOR VESSEL MATERIAL 1RRADIATED TO 2.20 x 10a n/cm2 Test .2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp Strength Strength Load Stress Strength Elongation Elongation in Area No. Material (*F) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) (%) Forging 05 200 66.6 84.5 2.95 1E8.4 60.1 9.7 20.2 62 i NT9 (Axial) s Forging 05 550 61.5 86.0 3.30 147.1 67.2 9.5 18.4 54 NT10 (Axial) Weld 150 68.2 82.5 2.72 221.6 55.4 9.0 21.0 75 W10 Metal Weld 63.2 80.3 3.10 170.8 63.2 7.8 16.0 63 W9 Metal 550 a 6 f
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Figure 5-3. Irradiated Charpy V-Notch Impact Properties for Sequoyah Unit 2 Reactor Pressure Vessel Weld Metal 75958: tbOM684 5-17
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SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1. INTRODUCTION Knowledge of the neutron environment within the pressure vessel / surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced properties, changes observed in materials test specimens and the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship must be established between the environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and mea-surements obtained with passive neutron flux monitors contained in each of the surveil-lance capsules. The latter information is derived solely from analysis. This section describes a discrete ordinates Sn transport analysis performed for the Sequoyah Unit 2 reactor to determine the fast neutron (E > 1.0 MeV) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules. The analytical data were then used to develop lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule T is discussed and comparisons with analytical predictions are presented. 6-2. DISCRETE ORDINATES ANALYSIS A plan view of the Sequoyah Unit 2 reactor geometry at the core midplane is shown in Figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a zero- to 45-degree sector is depicted. Eight irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program. Four capsules are located symmetrically at 4 and 40 degrees from the cardinal axes as shown in Figure 6-1. nnamoa2284 6-1
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A plan view of a single surveillance capsule attached to the thermal shieid is shown in Figure 6-2. The stainless steel specimen container is 1-inch square and approximately 38 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet nf the 12-foot-high reactor core. From a neutronic standpoint, the surveillance capsule structures are significant. In fact, J as is shown later, they have a marked effect on the distributions of neutron flux and energy spectra in the water annulus between the th 3rmal shield and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. Use of at least a two- ,
- dimensional computation is therefore mandatory.
In the analysis of the neutron environment within the Sequoyah Unit 2 reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOTW two-dimensional discrete ordinates code. The radial and azimuthal distributions were obtained from an R,0 computation wherein the geometry shown in Figures 6-1 and 6-2 was described in the analytical model. In addition to the R,0 computation, a second calculation in R,Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R,Z analysis, the reactor core was treated as an equivalent-volume cylinder and, of course, the surveillance capsules were not included in the model. t t Both the R,0 and R,Z analyses employed 47 neutron energy groups and a P3 expansion of the scattering cross sections. The cross sections used in the analyses were obtained from the SAILOR cross section library (51 which was developed specifically for light water + reactor applications. The neutron energy group structure used in the analysis is listed in Table 6-1. A key input parameter in the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For this analysis, power distributions representative of time-averaged conditions derived from statistical studies of long-term operation of Westinghouse 4-loop plants were employed. These input distributions include rod by-rod , spatial variations for all peripheral fuel assemblies. i e ; i nne.Some4 6-3 {
1 5396A-17 (40OR 400) (30OR 390) r CHARPY SPECIMEN
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TABLE 6-1 47 GROUP ENERGY STRUCTURE Lower Energy Lower Energy Group (MeV) Group (MeV) 1 14.19tal 25 0.183 2 12.21 26 0.111 3 10.00 27 0.0674 4 8.61 28 0.0409 5 7.41 29 0.0318 6 6.07 30 0.0261 7 4.97 31 0.0242 8 3.68 32 0.0219 9 3.01 33 0.0150 10 2.73 34 7.10x10 -3 11 2.47 35 3.36x10-3 12 2.37 36 1.59x10-3 13 2.35 37 4.54x10 -* 14 2.23 38 2.14x10 -* 15 1.92 39 1.01 x10 -* 16 1.65 40 3.73x10 -5 17 1.35 41 1.07x10-5 18 1.00 42 5.04x10-e 19 0.821 -- 43 1.66x104 20 0.743 44 8.76x10-7 21 0.608 45 4.14x10-7 22 0.498 46 1.00x10 -7 23 0.369 47 0.00 24 0.298 (a] The upper energy of group 1 is 17.33 MeV. nne: mea 2284 6-5
}
d it should be noted that this generic design basis power distribution is intended to provide a vehicle for long-term (end-of-life) projection of vessel exposure. Since plant-specific power distributions reflect only past operation, their use for projection into the future may not be justified; the use of generic data which reflects long-term operation of similar reactor cores may provide a more suitable approach. Benchmark testing of theso generic power distributions and the SAILOR cross sections against surveillance capsule data obtained from 2-loop and 4-loop Westinghouse plants indicate that this analytical approach yields conservative results, with calculations exceed-ing measurements from 10 to 25 percent.I'l One further point of interest regarding these analyses is that the design basis assumes an out-in fuel loading pattem (fresh fuel on the periphery). Future commitment to low-leakage loading pattems could significantly reduce the calculated neutron flux levels presented in Section 6-4. In addition, capsule lead factors could be changed, thereby influencing the withdrawal schedule of the remaining surveillance capsules. Having the results of the R,0 and R,Z calculations, three-dimensional variations of neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor. 4(R,Z,0Eg ) = 6(R,0Eg) x F(Z,Eg) (6-1) where 4(R,Z,0Eg) = neutron flux at point R,Z,0 within energy group g 4(R,0Eg) = neutron flux at point R,0 within energy group g obtained from the R,0 calculation F(Z,Eg) = relative axial distribution of neutron flux within energy group g obtained from the R,Z calculation j i I 6-3. NEUTRON DOSIMETRY The passive neutron flux monitors included in Capsule T of Sequoyah Unit 2 are listed in Table 6-2. The first five reactions in Table 6-2 are used as fast neutron monitors to relate neutron fluence (E > 1.0 MeV) to measured material property changes. To properly account for bumout of the product isotope generated by fast neutron reactions, it is necessary to also determine the magnitude of the thermal neutron flux at the monitor location. Therefore, bare and cadmium-covered cobalt-aluminum monitors were also included. 6-6 nnamon284
TABLE 6-2 NUCLEAR CONSTANTS FOR NEUTRON FLUX MONITORS CONTAINED IN
,THE SEQUOYAH UNIT 2 SURVEILLANCE CAPSULES Target Fission Weight Product Yield Monitor Material Reaction of Interest Fraction Half-life (%)
Copper CuS3 (n,a) Co" 0.6917 5.27 years Iron Fe" (n,p) Mn5' O.0585 314 days Nickel Nise (n,p) Co" 0.6777 71.4 days Uranium-238W U23' (n,f) Cs'37 1.0 30.2 years 6.3 Neptunium-237W Np237 (n,f) Cs'37 1.0 30.2 years 6.5 Cobalt-aluminumW Coa (n,y) Co* 0.0015 5.27 years Cobalt-aluminum Coa (n,y) Co" 0.0015 5.27 years [a] Denotes that monitor is cadmium-shielded k l ! 75798.1bO32284 6-7
l l l The relative locations of the various monitors within the surveillance capsule are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter bloc :!ccated near the center of the capsule. The use of passive monitors such as those listed in Table 6-2 does not yield a direct measure of the energy dependent flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent 1 neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest. m The operating history of the reactor a The energy response of the monitor a The neutron energy spectrum at the monitor location a The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two operations. First, the disintegration rate of product isotope per unit mass of monitor trust be determined. Second, in order to define a suitable spectrum-averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated. The specific activity of each of the monitors is determined using established ASTM pro-cedures?8Ael Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The overall stan-dard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from Sequoyah Unit 2, the overa!! 2e deviation in the measured data is determined to be plus or minus 10 percent. The neutron energy spectra are determined analytically using the method described in paragraph 6-1. 6-8 7sn e w o22284
Having the measured activity of the monitors and the neutron energy spectra at the locations of interest, the calculation of the neutron flux proceeds as follows. The reaction product activity in the monitor is expressed as n R= A fy i E ir@)S(EW k p (be9)e 9 (6-2) e max i1 where R = induced product activity No = Avogadro's number A = atomic weight of the target isotope fj = weight fraction of the target isotope in the target material Y = number of product atoms produced per reaction cr(E) = energy dependent reaction cross section 6(E) = energy dependent neutron flux at the monitor location with the reactor at full power Pj = average core power level during irradiation period j Pmax = maximum or reference core power level A = decay constant of the product isotope tj = length of irradiation period j td = decay time following irradiation period j Because neutron flux distributions are calculated using multigroup transport methods and, further, because the prime interest is in the fast neutron flux above 1.0 MeV, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the following relation. E <r(E)S(E)dE = J 6 (E > 1.0 MeV) where N ex. { <r9*9
,U ir(E)6(E)dE gg Tr = =
*6(E)dE
,wv h 09 9 = 9 s o wv 7579fi ib OM84 6-9
- - - - - - ~ - - - - - - , ., _ , _ _ _ , _ _ , _
~
Thuc, equation (6-2) is rewritten R= fi y d 4 (E > 1.0 MeV) I A Pmax(1-e-^'j) e 4-^'d j=1 or, solving for the neutron flux, R 4(E > 1.0 MeV) = n N P- (6-3) fi y d [ I -- (1-e ^'j) e 'A'd max the total fluence above 1.0 MeV is then given by n P-1 (6-4)
<b(E> 1.0 MeV) = 4(E> 1.0 MeV) { gmaxtj 3.,
y where Pj total effective full power seconds of reactor operation up to the time of capsule removal
]- . i.i pmax An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered Co58 (n,y) Co* data by means of cadmium ratios and the use of a 37-barn,2,200 m/sec cross section. Thus, Rbare I b }
D
&Th = n (6-5)
N
-0 L A fj y a T P-4 Pmax (1-e-^'j) e -^'d l- t l
where D is defined as Rbare/ rcd covered-6-10 75798 lb'032284 I
6-4. TRANSPORT ANALYSIS RESULTS Results of the Sn transport calculations for the Sequoyah Unit 2 reactor are summarized in Figures 6-3 through 6-8 and in Tables 6-3 through 6-5. In Figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius,1/4 thickness location, and 3/4 thickness location are presented as a function of
. azimuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident. In Figure 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 MeV) through the thickness of the reactor pressure vessel is shown. The
~
relative axial variation of neutron flux within the vessel is given in Figure 6-5. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in Figure 6-3 or 6-4 by the appropriate values from Figure 6-5. In Figure 6-6, the radial variations of fast neutron flux within surveillance capsules at 4 and 40 degrees are presented. These data, in conjunction with the maximum vessel flux, are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E > 1.0 MeV) at the dosimeter block location (capsule center) to the maximum fast neutron flux at the pressure vessel inner radius. Updated lead factors for all of the Sequoyah Unit 2 surveillance capsules are listed in Table 6-3. Since the neutron flux monitors contained within the surveillance capsules are not all located at the same radial location, the measured disintegration rates are analytically adjusted for the gradients that exist within the capsules so that flux and fluence levels may be deriveo en a common basis at a common location. This point of comparison was chosen to be the capsule center. Analytically determined reaction rate gradients for use
~
in the adjustment procedures are shown in Figures 6-7 and 6-8 for capsules at 4 and 40 degrees. All of the applicable fast neutron reactions are included. In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center of each of the Sequoyah surveillance capsules l is listed in Table 6-4. The associated spectrum-averaged cross sections for each of the l fast neutron reactions are given in Table 6-5. - l l f l l 7snenomu 6-11
$396A-18 1011 SURVEILLANCE 8 - CAPSULES R = 211.41 cm 6 -
4 - PRESSURE VESSEL IR 2 - 1/4T LOCATION "4 X 10 3 10 u. 8 - g N 6 - S Z 4 - 3/4T LOCATION 2 - l 108 0 10 20 30 40 50 60 AZIMUTHAL ANGLE (dog) l Figure 6-3. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux I (E > 1.0 MeV) Within the Pressure Vessel- Surveillance Capsule Geometry i 1 6-12 759ss:ino32ss4 I
5336A-19 \ 10 11 _ 8 - 6 - 4 - 219.71 l
- l 8 IR 225.19 cf 2 -
l
!e I g 1/4T 3 10 10 _
d 8 z O 6 - B - 236.14 g 4 _ l 241.62 2 - l l l l l l l l l l l l l O5 10 9 214 216 218 220 222 224 226 228 230 232 234 236 238 240 242 RADIUS (cm) Figure 6-4. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 75958: 1b032584 6-13
5396A-20 10 0 _ 8 ~_ 6 - 4 - 2 - y 10~l -- J g _
- u. _
z 6 - e - g 4 - y -
$ 2 -
P 5 E 10~2 -- 8 __ 6 - 4 - CORE MIDPLANE 2 - TO VESSEL CLOSURE HEAD l l l ! ! 10-3
-300 -200 -100 0 100 200 300 400 DISTANCE FORM CORE MIDPLANE (cm) l l
l l Figure 6-5. Relative Axial Variation of Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 6-14 75958: 1bO32684
5396A-21 1012 _ 8 - 6 - 4 - 2 - 211 41 400CAPSULES 1011 - o 8 - I - N 6 - E g - y 4 - 40CAPSULES 3 - a L z o 2 -
% CAPSULE 3 CENTER Z
1010 _ 8 - 6 - ( _ 4 - l 2 - THERMAL! , fO W! SHIELD $W9
; :! TEST SPECIMEN ;z;9j 3 j ;<;3:
, h; PN 10 9 !
I ! 5 I ! 207 208 209 210 211 212 213 214 RADIUS (cm) Figure 6-6. Calculated Radial Distribution of Maximum Fast Neutron Flux
- (E > 1.0 MeV) Within the Survelilance Capsule i
I-75958: 1b032684 6-15
5396A 22 1 l 1 109 8 - - 6 - s 4 - I I l 2 - 1 NiS8 (n, p) CoS8 8_ 6 Np237 (n, f) Cs137 [ l p 4 - l
.e . _
$ 2 -
g 211.41 P N 107 -- 8
@ Z U238 (n, f) Cs137 Q 6 -
p,54 (n, p) MnM e -
$ 4 -
l 5 - 2 - CAPSULE CENTER 100 - 8 - 6 2 Cu63 (n, a) Co60 4 - 2 THERMAL $ lj$l TEST SPECIMENS SHIELD j 5j
, j \ lC d' h; I I l$$@!'
{ 5 ' 207 208 209 210 211 212 213 214 R ADIUS (cm) Figure 6-7. Calculated Variation of Fast Neutron Flux Monitor Saturated l Activity Within Capsules Located at 40 Degrees l 6-16 7s% s. m 2ss4 1
5396A-23 108 _ 8 ^ 6 -
~
NiS3 (n, p) CoS8 2 - Np 237 (n, f) Cs137 10 7 .-- 8 - 6 - 211.41 e 4 -
- Fe54 (n,p) MnN f 2 -
U238 (n, f) Cs137 3 _1% 5 l
@ 10 6 _
o 8 - CAPSULE CENTER
# 6 -
g 4 - g Cu63 (n, d) Co60 10 5 _,,, 8 - i 6 - 4 - TH "M^' TEST SPECIMENS ; HiELD i $ l h l l '
' 5 104 207 208 209 210 211 212 213 214 RADIUS (cm)
Figure 6-8. Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules Located at 4 Degrees l 7595B: 1bC32684 6-17
l
)
TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AND LEAD FACTORS FOR SEQUOYAH UNIT 2 SURVEILLANCE CAPSULES Capsule Azimuthal 4(E > 1.0 MeV) Lead identification Location (dog) (n/cm2-sec) Factor S 4 3.04 x 10'o 1.02 V 4 3.04 x 10'o 1.02 W 4 3.04 x 10'o 1.02 Z 4 3.04 x 10'o 1.02 T 40 9.44 x 10'o 3.17 U 40 9.44 x 10'o 3.17 X 40 9.44 x 10'o 3.17 Y 40 9.44 x 10'o 3.17 6-18 7579e:ino32284
4 d TABLE 6-4 CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF 4 THE SEQUOYAH UNIT 2 SURVEILLANCE CAPSULES 4 ' 4 (n/cm2-sec) N 4 (n/cm2-sec) Group Group _- No. 4 Capsules 40' Capsules No. 4 Capsules 40 Capsules 2.08x107 25 3.59x10' 3.42x10'o a 1 1.35x107 2 4.85x107 7.65x 107 26 8.10x10' 3.29x10'o d 3 1.56x108 2.67x108 27 6.50x10$ 2.67x10'o 9 4 2.74x108 4.89x108 28 4.80x10' 1.99x10'o I 5 4.32x108 8.20x108 29 1.68x10' 6.92x10' % 6 9.33x108 1.85x10' 30 1.04x10' 4.27x10' ; 7.15x10' 7 1.18x10$ 2.57x10' 31 1.71x10' 8 2.07x10' 5.17x10' 32 1.65x10' 4.41 x10' 3 9 1.62x10' 4.54x10' 33 2.52x10$ 1.05x10'o , 4.21x10$ 1.75x10'o 2 10 1.27x10' 3.71x10' 34 11 1.46x10' 4.38x10' 35 5.66x10' 2.35x10'o N 12 7.19x108 2.19x10' 36 5.16x10' 2.17x10'o f5 13 2.12x108 6.52x108 37 7.79x10' 3.31x10'o 3 14 1.04x10' 3.22x105 38 4.42x10' 1.88x10'o f 15 2.67x10' 8.37x10' 39 4.68x10' 2.01 x10'o 3 16 3.21x10' 1.05x10'o 40 6.27x10' 2.71x10'o i 17 4.67x10' 1.57x10'o 41 7.59x10' 3.31x10'o j 18 8.45x10' 2.98x10' 42 4.33x10' 1.90x10'o - 19 5.73x10' 2.09x10'o 43 5.24x10' 2.31x10'o i 20 2.83x10$ 1.04x10'o 44 3.46x10' 1.53x10'o ] 21 8.14x10' 3.14x10'o 45 2.93x10$ 1.29x10'o "f 22 6.21 x10' 2.45x10'o 46 47 5.59x10' 1.41x10'o 2.41x10'o 5.66x10'o
}
23 7.46x10' 2.93x10'o j 24 6.51 x10' 2.59x10'o l lll 2 75790 tbO32284 6-19 a l
l TABLE 6-5 SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF CEQUOYAH UNIT 2 SURVEILLANCE CAPSULES d (barns)tal Reaction Capsules at 4 Capsules at 40' Fe" (n,p) Mn" 0.0980 0.0735 CuS3 (n,a) Co" 0.00112 0.000659 Niu (n,p) Co* 0.127 0.0993 2.62 Np237 (n,f) Cs'37 2.83 U23 (n,f) Cs'37 0.385 0.385 fxo a(E)6(E)dE [*4(E)dE 1MeV
=
(*- 1 l 6-20 7579B in032284 j l l
.--.---.\
6-5. DOSIMETRY RESULTS The irradiation history of the Sequoyah Unit 2 reactor up to the time of removal of Cap-sule T is listed in Table 6-6. Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsule T based on the irradiation history shown in Table 6-6 are given in Table 6-7. The data are prasented as measured at the actual monitor locations as well as adjusted to the capsule center. All gradient adjustments to the capsule center were based on the data presented in Figure 6-7. The fast neutron (E > 1.0 MeV) flux and fluence levels derived for Capsule T are presented in Table 6-8. The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in Table 6-9. Due to the relatively low thermal neutron flux at the capsulo location, no bumup correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than 1 percent for the Nise (n,p)CoS8 reaction and even less significant for all of the other fast neutron reactions. An examination of Tcble 6-8 shows that the fast neutron flux (E > 1.0 MeV) derived from the five th.eshold reactions rangas from 6.69 x 105 to 8.46 x 10m n/cm2.sec, a total span of less than 10 percent. It may also be noted that the calculated flux value of 9.44 x 105 n/cm2-sec exceeds all of the measured va:ues, with calculation to experimental ratios ranging from 1.12 to 1.41. Comparisons of meesured and calculated current fast neutron exposures for Capsule T 4 as well as for the inner radius of the pressure vessel are presented in Table 6-10. Measured values are given based on the Fe5d(n.p) Mn54 reaction alone as well as for the average of all five threshold reactions. Based on the data given in Table 6-10, the best estimate , exposure of Capsule T is
'DT = 2.20 x 10's n/cm2 (E > 1 MeV)
Since the calculated fluence levels were based on conservative representations of core power distributions derived for long term operation while the Capsule T data are repre- 6 ' sentative only of cycle 1 operation, it is recommended that projections of vessel toughness into the future be based on design basis calculated fluence levels. Withdrawal of future 1 surveillance capsules should further substantiate the adequacy of this approach. m es m o m s4 6-21
l TABLE 6-6 IRRADIATION HISTORY OF SEQUOYAH UNIT 2 SURVEILLANCE CAPSULE T l PJ Pmax Pj/Pmax Irradiation Time Decay Timetal Month Year (MW) (MW) (Days) (Days) 12 1981 27 3565 .007 9 679 1 1982 264 3565 .074 31 648 2 1982 763 3565 .214 28 620 3 1982 1223 3565 .343 31 589 . 4 1982 2251 3565 .632 30 559 5 1982 1282 3565 .360 31 528 6 1982 2706 3565 .759 30 498 7 1982 3389 3565 .951 31 467 8 1982 3287 3565 .922 31 436 9 1982 2883 3565 .809 30 406 10 1982 3123 3565 .876 31 375 11 1982 1366 3565 .383 30 345 12 1982 4 3565 .001 31 314 1 1983 M54 3565 .857 31 283 2 1983 3495 3565 .980 28 255 3 1983 3403 3565 .955 31 224 4 1983 3472 3565 .974 30 194 5 1983 3485 3565 .978 31 163 6 1983 3160 3565 .886 30 133 7 1983 3179 3565 .892 19 114 EFPS = 3.28E+07 SEC g
= 1.04 EFPY
[a] Decay time is referenced to 11/11,83. 6-22 75798:1b O32284
TABLE 6-7 COMPARISON OF MEASURED AND CALCULATD FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE T Saturated Activity Adjusted Saturated Activity Reaction Radial (dis g
/s) (dis /s) 9 and Location Axial Position (cm) Capsule T Calculated Capsule T Calculated Fe54 (n,p) Mn54 Top 211.68 3.01x108 3.17x108 Top-Middle 211.68 3.03x108 3.20x108 Middle 211.68 3.01x108 3.17x108 Bottom-Middle 211.68 3.08x108 3.25x108 Bottom 211.68 3.11x108 3.28x108 Average 3.05x108 4.30x108 3.21x105 4.53x108 t Cua 3 (n,a) Co" Top-Middle 211.18 3.22x105 3.06x105 Middle 211.18 3.24x105 3.08x105 l
Bottom Middle 211.18 3.36x105 3.20x105 Average 3.27x105 4.32x105 3.11x105 4.11x105 ! NiS (n.p) CoS8 Top-Middle 212.18 4.06x107 4.66x107 Middle 212.18 3.99x107 4.58x107 Bottom-Middle 212.18 4.16x107 4.78x107 Average 4.07x107 5.75x107 4 67x107 6.59x107 NP237 (n,f) Cs'37 Middle 211.41 3.95x107 4.41x107 3.95x107 4.41x107 U238 (n,f) Cs'37 Middle 211.41 5.02x108 5.31x108 5.02x108 5.31x10' 75790.1tkC32284 6-23
TABLE 6-8 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE T dis /s 4 (E > 1.0 Mev) 4 (E > 1.0 Mov) g (n/cm2-sec) (n/cm2)
- Reaction Measured Calculated Measured Calculated Mamanrod Calculated g
a Fe"(n.p)Mn" 3.21x10' 4.53x10' 6.70x10* 9.44x10$ 2.20x10* 3.10x10*
< Cu"(n,u)Co" 3.11x105 4.11x105 7.14x10* 9.44x10'* 2.34x10* 3.10x10*
Ni"(n.p)Co" 4 67x10' 6.59x10' 6.69x10* 9.44x10* 2.19x10* 3.10x10* Np22'(n,f)Cs"' 3.95x10' 4.41x10' 8.46x10* 9.44x10* 2.77x105 3.10x10* 4.42x105 5.31x10' 7.19x10* 9.44x10* 2.36x10* 3.10x10* Ua(n.f)Cs'* 141 u2= amusted saturated aciuny has been mumphed by a se io correct for 350 ppm u2mvnpurny 3 5 b i
TABLE 6-9 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE T Saturated Activity g (dis /s} Axlal 6Th Location Bare Cd-Covered (n/cm2-sec) Top 7.01 x 107 2.79 x 107 7.48 x 10'o Bottom 6.94 x 107 2.73 x 107 7.38 x 10'o Average 6.98 x 10' 2.76 x 107 7.43 x 10'o TABLE 6-10 SUMMAPY OF NEUTRON DOSIMETRY RESULTS FOR CAPSULE T Current 4 (E > 1.0 mov) EOL 6 (E > 1.0 mov) (n/cm') (n/cm8) Location Measured Calculated Measured Calculated Capsule T 2.20 x 10 3.10 x 10'8 Vessel IR 6.94 x 10 9.78 x 10" 2.13 x 105 3.01 x 10 Vessel 1/4T 3.85 x 10 5.42 x 10" 1.18 x 10 1.67 x 10 Vessel 3/4T 7.93 x 10 1.12 x 10 2.44 x 10's 3.43 x 10 Note: EOL fluences are based on operation at 3565 MWt for effective full-power years. 75796.1t>032284 6-25
SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule per ASTM E185-82 is recommended for futyre capsules. to be removed from the Sequoyah Unit 2 reactor vessel: Vessel Estimated Capsule Location Leaa Removal Fluence (deg) Factor Timetal (n/cm2) T 40 3.17 1.04 (f emoved) 2.20 x 10'8 U 140 3.17 3 8.39 x 10's X 220 3.17 6 1.68 x 10'*i Y 320 3.17 11 3.06 x 10'*1 S 4 1.02 34 3.06 x 10'S V 176 1.02 standby W 184 1.02 standby Z 356 1.02 standby [a Effective full power years from plant startup [b Approximate fluence at 1/4 thickness vessel wall at end of hfe [c] Approximata fluence at vesselinner wall at end of hfe 7sesaino32384 7-1
SECTION 8 REFERENCES
- 1. Yanichko, S. E., Davidson, J. A., and Phillips, J. H., " Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8513, November,1975.
- 2. ASTM Standard E185-73," Recommended Practice for Surveillance Tests for Nuclear Reactors" in ASTM Standards, Part 45 (1973), American Society for Testing and Materials, Philadelphia, Pa.,1973.
- 3. Regulatory Guide 1.99, Revision 1," Effects of Residual Elements on Predicted Radia-tion Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, April 1977.
- 4. Soltesz, R. G., Disney, R. K., Jedruch, J., and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation. Vol. 5 - Two-Dimen-sional Discrete Ordinates Transport Technique," WANL PR(LL)034, Vol. 5, August 1970.
- 5. SAILOR RSIC Data Library Collection DLC 76, Coupled Self Shielded,47 Neutron, 20 Gamma Ray, P3, Cross Section Library for Light Water Reactors.
- 6. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology (To be published).
- 7. ASTM Designation 261-77," Standard Practice for Measuring Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards (1981), Part 45, Nuclear Standards, pp 915-926, American Society for Testing and Materials, Phila-delphia, Pa.,1981.
- 8. ASTM Designation 262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards (1981), Part 45, Nuclear Stan-dards, pp 927-935, American Society for Testing and Materials, Philadelphia, Pa.,
1981. nne. mosm < 81
1
- 9. ASTM Designation 263-77, " Standard Method for Measuring Fast-Neutron Flux by l Radioactivation of Iron," in ASTM Standards (1981), Part 45, Nuclear Standards,
. pp 936-941, American Society for Testing and Materials, Philadelphia, Pa.,1981.
- 10. ASTM Designation 481-78," Standard Method of Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1981), Part 45, Nuclear Standards, pp 1063-1070, American Society for Testing and Materials, Philade'ephia, Pa.,1981.
- 11. ASTM Designation 264-77, " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel," in ASTM Standards (1981), Part 45, Nuclear Standards, pp 942-945, American Society for Testing and Materials, Philadephia, Pa.,1981.
e 9 9 82 7s798 in o326s4
APPENDIX A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RTNDT of the materialin the core region of the reactor vesselis determined by using the preservice reactor vessel material properties and estimating the radiation-induced ARTNDT.RT NDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the tem-perature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60 F. RTNDT increases as the material is exposed to fast-neutron radiation. Thus, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. The extent of the shift in RT NDT si enhanced by certain chemical elements (such as copper and phosphorus) present in reactor vessel steels. Design curves which show the effect of fluence and copper and phosphorus contents on ARTNDT for reactor vessel steels are shown in Figure A-1. Given the copper content of the most limiting material, the radiation-induced ARTNDT can be estimated from Figure A-1. Fast neutron fluence (E > 1 Mev) at the vessel inner surface, the 1/4 T (wall thickness), and 3/4 T (wall thickness) vessel locations are given as a function of full power service life in Figure A-2. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limitinc with respect to RTNDT-A 2. FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the Sequoyah Unit 2 reactor vessel materials are presented in Table A-1. The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review PlanPI The postirradiation fracture-toughness prop-erties of the reactor vessel beltline material were obtained directly from the Sequoyah Unit 2 Vessel Material Surveillance Program. 7sess mome4 A1
C o-2 LEGEND- UPPER LIMIT
$ 400 -
.1 R TNDT = [40 + 1000 (% Cu - 0.08) + 5000 (% P - 0.008)] [f/1019] 1/2 300 A WELD METAL y - G SHELL FORGING 05 y
- us /
x / y 200 - UPPER LIMIT / i ua / -
- E /
, u. p 4 O /
> /
d 100 - / 2 / w g/ s
? / > c /
l 6 4 / ' O 50 - #
"8 0.35 0.30 0.25 0.2(ry. Cu 0.15% Cu 0.10% Cu
$ LOWER LWIT
%P = 0'012 5 % Cu = 0.08
. y % P = 0.008 c. 1' s e t t I l 1 I I I l 1 l w E 2X10 17 4 6 8 10 18 2 4 6 8 10 19 2 4 6 FLU ENCE, n/cm2 (E > 1MeV)
) Figure A-1. Predicted Adjustment of Reference Temperature, ARTNDT, as a Function of Fluence and Copper Content. For Copper and Phos-j phorus Contents Other Than Those Plotted, Use the Expression I for ARTNDT Given on the Figure. y g
I
a 8 1020 8 8 = E 6 3 INNER SURFACE 4 2 - 10 19 _ 8 : 1/4 T N' 6 - E
, ae 4 -
g 2 - z
$ 10 18 3/4 T
_a 8 _-
' 6 > z 6 o 4 x
3 2 - z 1017 8 : 6 i 4 - 2 - 10 16 l l l l l l l l l l l l l l l l 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 SERVICE LIFE (EFFECTIVE FULL POWER YEARS) Figure A-2. Fast Neutron Fluence (E > 1 MeV) as a Function of Full Power Service Life $ M e'
TABLE A-1 {
$ SEQUOYAH UNIT 2 REACTOR VESSEL TOUGHNESS DATA E
Minimum 50 ft-lb/35 Average Upper Shelf Heat Material CU P NDTT } RTNOT - Component No. Grade (%) (%) (*F) PMWDai NMWDW ( F) PMWDM NMWDM _ CL Hd. Dome 52899-1 A533BCL1 - -
- 13 28 48m -12 7Sa CL Hd. Ring -
A508CL2 - - 5 34 5 424 5 1255* Hd Flange 4890 A508CL2 - -
-13 < - 67 < - 67k' - 13 141M Vessel Flange 4832 A508CL2 - -
- 22 - 47 - 27kt - 22 155.5m inlet Nozzle 4868 A508CL2 - -
- 22 41 61m 1 79:a inlet Nozzle 4872 A506CL2 - -
- 22 12 32m - 22 108*
Inlet Nozzle 4877 A508CL2 - -
- 31 1 21m - 31 113*
y intet Nozzle 4886 A508CL2 - -
- 31 - 52 -32" - 31 138t*
A Outlet Nozzle 4867 A508CL2 - -
- 31 19 39+4 - 21 854ai Outlet Nozzle 4873 A508CL2 - -
- 22 21 41m -19 76*
Outlet Nozzle 4878 A508CL2 - -
- 40 -6 14:4 - 40 105W Outlet Nozzle 4887 A508CL2 - -
- 22 - 11 9ki - 22 143.5m Upper Shell 4885 A508CL2 - -
5 25 454 5 1044 Inter Shell 4853 A508CL2 0.13 .014 - 22 19 70 10 138 93 Lower Shell 4994 A508CL2 0.14 .012 - 40 8 38 - 22 140.5 100 Trans. Ring 4879 A508CL2 - - 5 27 47m 5 985 Bot. Hd. Ring 52835-18 A533BCL1 - -
-4 48 68:4 8 81m Bot. Hd. Ring 52835-2 A533BCL1 - -
- 22 25 45m -15 81m Bot. Hd. Ring 52899-2 A533BCL1 - -
-13 39 59ki -1 62ial Bot. Hd. 52979-1 A533BCL1 - -
- 31 14 34'o - 26 99.5m Weld -
Weld 0.13 .016 -4 - 14 -4 - 101 HAZ - - HAZ - -
-13 -
17 -13 - 120 [a] Paranel to ma or workeg erecton (b) Normal to ma-x workog erecton Icl Estimate base 3 on the NRC Regulatory Standarri Review Plan, Secton 5 3 2 MTEB 5 2 [d] Percentage shear not reported, therefore value may not be on the upper sh#
A-3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. K IR isobtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code.t2i The KIR curve is given by the equation: KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)] (A-1) where K IR is the reference stress intensity factor as a function of the metal temperature
, T and the metal reference nil-ductility temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Codetal as follows:
C K;y + Kit 5KIR (A-2) where K IM is the stress intensity factor caused by membrane (pressure) stress Kitis the stress intensity factor caused by the thermal gradients , K IR is a function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, KIR si determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From Equation (A 2), the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. For the calculation of the allowable pressure versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because A-5 nea tooms4
n the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Aliowable pressure-temperature relations are generated for both steady-state ard finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in KIR exceeds Kit, the calculated allowable pressure during cooldown will be greater than the steady state value. The above proceduies are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup ra,tes. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The thermal gradients
- during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip
. lags the coolant temperature; therefore, the KIR for the 1/4 T crack during heatup is lower than the K IR orf the 1/4 T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower KIR's do not offset each other, and the pressure temperature curve based on steady state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady state and finite heatup rates is obtained.
A6 n958 4CM84
The second portion of the heatup analysis concerns the calculation of pressure-temper-ature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, tho thermal gradients established at the outside surface during heatup produce stresses which are tensits in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible ' for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most entical criterion. Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temper-ature sensing instruments by the values indicated on the respective curves. A-4. HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in Section A-3. The derivation of the limit curves is presented in the NRC Regulatory Standard Review PlanP Transition temperature shifts occurring in the pressure vessel materiale due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance pro-gram. Charpy test specimens from Capsule T indicate that the core region weld metal and limiting core region shell forging 05 (Heat no. 4853) exhibited shifts in RTNDTOf 80*F and 25 F, respectively. These shifts at a fluence of 2.20x10's n/cm2 are plotted in Figure A-1. A modified trend curve was developed for the weld metal by drawing a line parallel to the present trend curves through the surveillance data point. Figure A-1 shows the modified trend curve as a dashed line. The modified trend curve is used to obtain the i ARTNDT n i the weld metal, and the heatup and cooldown curves are based on this ARTNDT. The resultant heatup and cooldown limit curves for normal operation of the reactor vessel are presented in Figures A 3 and A 4 and represent an operational time period of 9 effective full power years. I nun mtmsa4 A7
L 8311E 3 MATERI AL PROPERTY BASIS: CONTROLLING MATERIAL : WELD METAL COPPER CONTENT : 0.13 WT% PHOSPHORUS CONTENT : 0.016 WT% 0 RTNDT INITIAL : -4 F
- 1/4T,1100 F RTNDT AFTER 9 EFPY 0
- 3/4T, 47 F CURVE APPLICABLE FOR HEATUP RATES UP TO 600F/HR FOR THE SERVICE PERIOD UP 0
TO 9 EFPY AND CONTAINS MARGINS OF 10 F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 2600
' ? ?
LEAK TEST - 1 LIMIT M 2000 f 6 I J J k HEATUP y j g CURVE , 7 j 3 J J h / / r .r _e I i a 1000 9 / O r c CRITICALITY LIMIT E / I BASED ON INSERVICE
#/
HYDROSTATIC TEST r TEMPER ATURE (2550F) - FOR THE SERVICE PERIOD UP TO 9 EFPY 0 0 100 200 300 400 500 INDICATED TEMPERATURE (OF) i I Figure A-3. Sequoyah Unit 2 Reactor Coolant System Heatup Limitations Applicable up to 9 EFPY l A-8 7s958 m a sa4 a -
8311 E-4 MATERIAL PROPERTY BASIS : CONTROLLING MATERIAL WELD METAL COPPER CONTENT 0.13 WT% PHOSPHORUS CONTENT : 0.016 WT% RTNDT INITIAL : -40F RTNDT AFTER 9 EFPY : 1/4T,1100 F
- 3/4T, 470 F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 1000F/HR FOR THE SERVICE PERIOD UP TO 9 EFPY AND CONTAINS MARGINS OF 100F AND 60 PSIG FOR POSSIBLE INSTRUMENT g
ERRORS 2600 ,,,i llIl J lill f llll i llll 1 5 2000 llll j g llll 1 llll 1 E ll11 ! D ll11 1
$ lll1 /
m Illi /
$ lill J g Illi /
0 1000
!!!! J 5 'COOLDOWN g 7 z RATES g, (OF/HR)
I 2
\ ~ ' dE l flil 40
-N:sf7 '
1III 00 V - lill
~#
lill
'_100 llll . III IIII O
'I 300 400 500 O 100 200 INDICATED TEMPERATURE ( F) Figure A-4. Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations Applicable up to 9 EFPY 759sainoa2sa4 A-9
c Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure A-3. This is in addition to other criteria which must be met before the reactor is made critical. The leak test limit curve shown in Figure A-3 represents minimum temperature require-ments at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of References 2 and 3. Figures A-3 and A-4 define limits for insuring prevention of nonductile failure. a A-10 7s958 in032es4 c_
_ = _ - -. ._ APPENDIX A REFERENCES
- 1. " Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981.
- 2. ASME Boiler and Pressure Vessel Code, Section 111, Division 1 - Appendices, " Rules for Construction of Nuclear Vessels," Appendix G, " Protection Against Nonductile Failure," pp. 559-564,1983 Edition, American Society of Mechanical Engineers, New ,
York,1983.
- 3. " Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
75968.10032004 A 11
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