ML20214M031
| ML20214M031 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/28/1987 |
| From: | Perone V, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML18150A103 | List: |
| References | |
| WCAP-11415, NUDOCS 8706010154 | |
| Download: ML20214M031 (89) | |
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{{#Wiki_filter:. WCAP 11415 WESTINGHOUSE CLASS 3 CUSTONER DESIGNATED DISTRIBUTION o' J ANALYSIS OF CAPSULE V FROM THE VIRGINIA ELECTRIC AND POWER COMPANY SURRY UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko V. A. Perone February 1987 Work performed under Shop Order No. VCKJ-106 APPROVED: jdM T.A.MeykManager Structural Waterials and Reliability Technology Prepared by Westinghouse for the Virginia Electric and Power Company Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval. 9 I~ WESTINGHOUSE ELECTRIC CORPORATION Power Systems Division o P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 nowan.- imo. 8706010154 870522 ADOCK0500gG PDR
,= , y, t ,,l) PREFACE This report has been technically reviewed and verified. ' u,, +i Reviewer Sections 1 through~5 and 7 C. C. Heinecke (( M_ Section.6' S. L. Anderson - d.T Ortdamh T O P t i f* f x i t i i ,1 i e
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TABLE OF CONTENTS Section Title Page s-1
SUMMARY
OF RESULTS 1-1 2 INTRODUCTION 2-1 3 BACKGROUND 3-l' 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE V 5-1 5-1. Overview 5-1 5-2. Charpy V-Notch Impact Test Results 5-3 5-3. Tension Test Results 5-4 5-4. Wedge Opening Loading Tests 5-4 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1. Introduction 6-1 6-2. Discrete Ordinates Analysis 6-1 6-3. Neutron Dosimetry 6-5 6-4. Transport ' Analysis Results 6-10 6-5. Dosimetry Results 6-12 6-6. Surveillance Capsule Withdrawal Schedules 6-14 6-7. Influence of an Energy Dependent Damage Model 6-14 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE 7-1 8 REFERENCES 8-1 4 220Cs/0335e-041387:10 y w
LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the Surry Unit 1 4-5 4-2 Capsule V Diagram Showing Location of Specimens, 4-6 Thermal Monitors, and Dosimeters 5-1 Irradiated Charpy V-Notch Impact Properties for 5-17 Surry Unit 1 Reactor Vessel Lower Shell Plate -C4415-1 5-2 Irradiated Charpy V-Notch Impact Properties for-5-18 Surry Unit 1 Reactor Pr. essure Vessel Weld Metal i 5-3 Irradiated Charpy V-Notch Impact Properties for 5-19 Surry Unit 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-4 Irradiated Charpy V-Notch Impact Properties for 5-20 Surry Unit 1 A533 Grade B Class 1 Correlation Monitor Material (HSST Plate 02) 5-5 Tensile Properties for Surry Unit 1 Reactor Vessel 5-21 Lower Shell Plate C4415 5-6 Tensile Properties for Surry Unit 1 Reactor Vessel 5-22 Wald Metal 6-1 Surry Unit 1 Reactor Geometry 6-39 6-2 Reactor Vessel Surveillance Capsule 6-40 l 6-3 Surry Unit 1 Maximum Fast Neutron (E > 1 MeV) Fluence 6-41 at the Beltline Weld Locations as a Function of Full Power Operating Time 6-4 Surry Unit 1 Maximum Fast Neutron (E > 1 MeV) Fluence 6-42 at the Center of the Surveillance Capsules as a Function of Full Power Operation Time 6-5 Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence. 6-43 at the Pressure Vessel Inner Radius as a Function of 4 Azimuthal Angle 6-6 Surry Unit 1 Relative Radial Distribution of Fast Neutron 6-44 (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 6-7 Surry Unit 1 Relative Axial Variation of Fast Neutron 6-45 (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 2200s/0335s-041387:10 yj
i LIST OF TABLES Table-Title Page-i 4-1 Chemical Composition of the Surry Unit 1 Reactor 4-3 Vessel Surveillance Materials 4-2 Heat Treatment of the Surry Unit 1 Reactor Vessel 4-4 Surveillance Materials 5-1 Charpy V-Notch Impact Data for the Surry Unit 1 5-5 Lower Shell Plate C4415-1 Irradiated at 550*F, Fluence 1.94 x 10" n/cm2 (E > 1 MeV) 5 Charpy V-Notch Impact Data for the'Surry Unit 1 5-6 Pressure Vessel Weld Metal Irradiated at 550*F, Fluence 1.94 x 10" n/cm2 (E > 1 MeV) ! 3 Charpy V-Notch Impact Data for the Surry Unit 1. 5-7 Pressure Vessel Weld Heat Affected Zone Metal Irradiated at 550*F, Fluence 1.94 x 10" n/cm 2 (E > 1 MeV) 5-4 Charpy V-Notch Impact Data for the Surry Unit 1 5-8 A533 Grade B Class 1 Correlation Monitor Material (HSST Plate 02) at 550*F, Fluence 1.94 x 10" n/cm 2 (E > 1 MeV) 5-5 Instrumented Charpy Impact Test Results for Surry 5-9 Unit 1 Lower Shell Plate C4415-1 .5-6 Instrumented Charpy Impacc -Test Results for Surry 5-10 Unit 1 Wald Metal 5-7 Instrumented Charpy Impact Test Results for Surry 5-11 Unit 1 Wald Heat Affected Zone Metal 5-8 Instrumented Charpy Impact Test Results for Surry Unit 1 5-12 A533 Grade B Class 1 Correlation Monitor Material (HSST Plate 02) 5-9 The Effect of 550*F Irradiation at 1.94 x 10" 5-13 i (E > 1 MeV) on the Notch Toughness Properties of The Surry Unit 1 Reactor Vessel Materials l' 5-10 Summary of Surry Unit 1 Reactor Vessel Surveillance 5-14 Capsule Charpy Impact Test Results is i 5-11 Comparison of Measured ART Versus Regulatory 5-15 Guide 1.99 Revision 2PredEedRTilDT 5-12 Tensile Properties for Surry Unit 1 Reactor Vessel 5-16 Material Irradiated to 1.94 x 10" n/cm2 y$$ noovens.-ouwe
LIST OF TABLES (Cont) Table Title Page 6-1 47 Energy Group Structure 6-16 6-2 Nuclear Parameters for Neutron Flux Monitors 6-17 6-3 Surry Unit 1 Calculated Fast Neutron (E >-1.0 MeV) 6-18 Exposure at the Pressure Vessel Inner Radius - 0* Azimuthal Angle 6-4 Surry Unit 1 Calculated Fast Neutron (E > 1.0 MeV) 6-19 Exposure at the Pressure Vessel Inner Radius - 15' Azimuthal Angle 6-5 Surry Unit 1 Calculated Fast Neutron (E > 1.0 MeV) 6-20 Exposure at the Pressure Vessel Inner Radius - 30' Azimuthal Angle 6-6 Surry Unit 1 Calculated Fast Neutron (E > 1.0 MeV) 6-21 ~ Exposure at the Pressure Vessel Inner Radius - 45' Azimuthal Angle 6-7 Surry Unit 1 Calculated Fast Neutron (E > 1.0 MeV) 6-22 Exposure at the 15' Surveillance Capsule Center 6-8 Surry Unit 1 Calculated Fast Neutron (E > 1.0 MeV) 6-23 Exposure at the 25' Surveillance Capsule Center 6-9 Surry Unit 1 Calculated Fast Neutron (E > 1.0 MeV) 6-24 Exposure at the 35* Surveillance Capsule Center 6-10 Surry Unit 1 Calculated Fast Neutron (E > 1.0 MeV) 6-25 Exposure at the 45' Surveillance Capsule Center 6-11 Irradiation History of Surry Unit 1 Reactor Vessel 6-26 Surveillance Capsule V 6-12 Measured Flux Monitor Activities from Surry Unit 1, 6-30 Capsule T 6-13 Measured Flux Monitor Activities from Surry Unit 1, 6-31 Capsule W 6-14 Measured Flux Monitor Activities from Surry Unit 1, 6-32 Capsule V 6-15 Calculated Neutron Energy Spectra at the Center of 6-33 Surry Unit 1 Surveillance Capsules 6-16 Spectrum Averaged Reaction Cross-Sections at the Center 6-35 of Surry Unit 1 Surveillance Capsules noeuem.-ocmao y$$$
~ LIST OF TABLES (Cont) Table Title _Page 6-17 Thermal Neutron Flux Data from Capsules T, W, and V 6-36 18 Comparison of Measured and Calculated Fast Neutron 6-37~ Fluence for Capsules T, W, and V 6-19 dPa/, (E > 1.0 MeV) Ratios for Surry Unit 1 6-38 b b i i. l 2200s/0335s-041347:10 jy I
SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in Capsule V, the third surveillance capsule to be removed from the Surry Unit I reactor pressure vessel, led to the following conclusions: o The capsule received an average fast neutron fluence (E > 1.0 MeV) 19 2 of 1.94 x 10 n/cm, o Irradiation of the reactor vessel lower shell plate C4415-1, to 19 1.94 x 10 n/cm, resulted in 30 and 50 ft-lb transition temperature increases of 110*F and 130*F, respectively for specimens oriented parallel ~to the major. working direction (longitudinal orientation). The upper shelf energy decreased from 125 to 116 ft-lb as a result of the irradiation. 19 2 o Weld metal irradiated to 1.94 x 10 n/cm resulted in a 30 ft-lb transition temperature increase of 240*F. The upper shelf energy decreased to ~ 50 ft-lb as a result of the irradiation. o Comparison of the 30 ft-lb transition temperature increases (aRTNDT) f r the Surry Unit 1 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2, i shows that the plate material and weld metal transition temperature increase were in relatively good agreement with predicted increases. i b a noo,mn,-oew.,o 1-1
SECTION 2 ~ INTRODUCTION This report presents the results of the examination of Capsule V, the third capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Virginia Electric and Power Company Surry Unit I reactor pressure vessel materials under actual operating conditions. The surveillance program for the Surry Unit I 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 c e presented by Yanichko.Ill The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel anc was based on ASTM E-185-73, " Recommended Practice for Surveillance Tests for Nuclear Reactors".[2] 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. j This report summarizes testing and the postirradiation data obtained from surveillance Capsule V removed from the Surry Unit I reactor vessel and discusses the analysis of the data. The data are also compared to capsule I33 I43 which was removed from the reactor in 1974 and capsule W which was T removed in 1978. It sould be noted that only dosimetry was measured for the capsule W. A new reactor vessel surveillance capsule withdrawal schedule was developed to meet the requirements of ASTM E-185-82, as proposed by Babcock and Wilcox(5), N nm.mn.* nno 2-1
SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core _and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of 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 SA533 Grade B Class 1 plate (base material of the Surry Unit i reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation. 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 III of the ASME Boiler and Pressure Vessel l. Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)* l f RT is defined as the greater of either the drop weight nil-ductility NDT transition temperature (NDTT per ASTM E-208) or tFe temperature 60*F less than the 50 ft ib (and 35-mil lateral expansion) temperature as determined from f Charpy specimens oriented normal (transverse) to the major working direction of the material. The RT f a given material is used to index that NDT material to a reference stress intensity factor curve (K curve)which IR appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to l ~ l 3-1 220cs/0335s-041347.10
the K curve, allowable stress intensity factors can be obtained for this IR 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 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 Surry Unit 1 Reactor Vessel Radiation Surveillance Program,Ill in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens 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 RT f r radiation embrittlement. This adjusted RTNDT (RTNDT initial + NDT ARTNDT) is used to index the material to the Kgg curve and, in turn, to set ope. rating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials. b"2 2200s/032ss 041387to ~
SECTION 4 DESCRIPTION OF PROGRAN Eight surveillance capsules for monitoring the effects of neutron exposure on the Surry Unit I reactor pressure vessel core region material _were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. Capsule V was removed after 8.02 effective full power years of plant operation. The capsule contained Charpy V-notch impact and tensile specimens from the lower shell' plate C4415-1 and submerged arc weld metal representative of the beltline weld seams of the reactor vessel, WOL specimens from the weld metal and Charpy V-notch specimens from weld heat-affected zone (HAZ) material (Figure 4-2). All heat-affected zone specimens were obtained from within the HAZ of plate C4415-1 of the representative weld. The chemistry and heat treatment of the surveillance material are presented in table 4-1 and table 4-2, respectively. The chemical analyses reported in. table 4-1 were obtained from unirradiated material used in the surveillance program. In addition, a ghemical analysis was performed on an irradiated Charpy specimen from the weld metal and plate C4415-1 and is reported in table 4-1. l All test specimens were machined from the 1/4 thickness location of the plate after stress relieving. Test specimens represent material taken at least one plate thickness from the quenched edges of the plate. Base metal Charpy l V-notch impact specimens were oriuted with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation). Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding direction. The WOL specimens in Capsule V were machined such that the i simulated crack in the specimen would propagate parallel to the weld direction. um.ans.-ou une 41
Capsule V contained dosimeter wires of pure copper, nickel, and aluminum-cobalt (cadmium-shielded and unshielded). In addition, cadrium-shielded dosimeters of Neptunium (Np ) and Uranium (U238) were contained in the capsule. 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) The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule V are shown in Figure 4-2. l l l l l i f l l noo.mn.-o.inr io 4-2
TABLE 4-1 CHEMICAL COMPOSITION OF THE SURRY UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Intermediate Lower Shell Plate Shell Plate Cc. relation Element C4326-1 C4415-1 Weld Metal (d) Monitor C. 0.23 0.22 0.245(b) 0.10 0.185(c) 0.22 Mn 1.35 1.33 1.46 (b) 1.49 1.47(c) 1.48 P 0.008 0.014 0.012(b) 0.011 0.011(c) 0.012 S 0.015 0.014 0.017(b) 0.010 0.017(c) 0.018 Si 0.23 0.23 0.42 (b) 0.37 0.43(c) 0.25 Ni 0.55 0.50 0.569(b) 0.68 0.643(c) 0.68 Cr 0.069 0.078 0.105(b) 0.076 0.074(c) V 0.001(a) 0.001(a) 0.004(b) 0.001 <0.002(c) Mo. 0.55 0.55 0.618(b) 0.46 0.405(c) 0.52 Co 0.014 0.015 0.006(b) 0.001 0.011(c) Cu 0.11 0.11 0.115(b) 0.25 0.243(c) 0.14 Sn 0.008 0.008 Zn 0.001(a) 0.001(a) 4 A1 0.037 0.036' 0.013 N 0.007 0.007 0.008 2 Ti 0.001(a) 0.001(a) Zr 0.002 0.002 As 0.007 0.007 B 0.003(a) 0.003(a) (a) Not detected. The number indicates the minimum limit of detection. (b) Analysis performed on irradiated Charpy plate specimen V-25. (c) Analysis performed on irradiated Charpy weld specimen W-10. [d] Surveillance weld fabricated from same heat of weld wire (299L44) and Linde 80 flux lot (8596) as used in the vessel lower shell vertical weld seam (L2). noo.mn.+inr:ia 4-3
TABLE 4-2 HEAT TREATMENT OF THE SURRY UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Material Heat 1reatment Intermediate shell 1650*-1700* - 9 hours - Water quenched (Plate C4326-1) 1210*F - 9 hours - Air-cooled 1125'F 1/2 hours - Furnace cooled to 600*F Lower shell 1650-1700*F - 9 hours - Water quenched (Plate C4415-1) 1200*F - 9 hours - Air-cooled 1125'F 1/2 hours - Furnace-cooled to 600*F Weldment 1125*F 1/2 hours - Furnace-cooled to 600*F Correlation Monitor 1675 1 25'F - 4 hours - Air-cooled 1600 1 25'F - 4 hours - Water quenched 1225 1 25'F - 4 hours - Furnace-cooled 1150125'F - 40 hours - Furnace-cooled to 600'F l. 1 noe.mn.* nr io 44
14417.1 270 T 1-CAPSULE (TYPICAL) S 1 REACTOR VESSEL i I THERMAL SNI6LO
- g. 115' ICf 1
I 9 O* 180* I V 25' (. 90* Figure 4-1 Arrangement 0f Surveillance CapsuTes in Surry Unit 1 G 4-5
l l I t l 5&F MONITOR 590*F MOMTOR SWF MOMTOR
- C TENS 8LE C,
C, C, Cv TENSOLE . C, W V V y y V V V f V W W W W W W 42 1 41_ .46 45 AA .47 10 e 12 11 16 . 18i a r cu Lco co C cwcci 4,, cwcon c. .0L .0L c. NpA U a 5 6 m 44 43 14 13 r i l b i r u E 1 E w 9 o.o 2 .o n 4 4 9 n = u g = 1 E I ; I m E s m I
- I
TO VESSEL BOTTOM VCSSEL.ALL SiOE V = PLATE C4415-1 H=HAZ W = WELD METAL R = REFERENCEMATERIAL Figure 4-2 Capsule V Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters
SECTION 5 TESTING OF SPECIMENS FROM CAPSULE V 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 consultatica by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and. H, ASTM Specification E185-82 and Westinghouse Procedure NHL 8402, Revision 0 as modified by RMF Procedures 8102 and 8103. Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-7723.(1) No discrepancies were found. t Examination of the two low-melting 304*C (579'F) and 310'C (590*F) eutectic alloys indicated no melting of either type of thermal 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 j Procedure 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 (E ). From the load-time curve, the load of general yielding (PGY),the D time to general yielding (tgy), the maximum load (P ), and the time to y ' maximum load (t ) can be determined Under some test conditions, in the y Charpy transition region, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the t 4 1 noo.mn.-ouune 5-1 1 --~c ,,_-,,.-,--,,a
,-,,,,,--,-e,,.-.r-,-,-
--,---n.- ,, ~ - -,,rx,va,.,,=,----,--n----.n.-~--.
fast fracture load (P ), and the load at which fast fracture terminated is F identified as the arrest load (P )* A The energy at maximum load (E ) was determined by comparing the energy-time M record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E ) is the difference p between the total energy to fracture (E ) and the energy at maximum load. D The yield stress (cy) is calculated from the three point bend formul'a. 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 E8-83 and E21-79, and RMF Procedure 8iO2. All pull rods, grips, and pins were made of Inconel 718 hardened to R 45. The upper pull rod was connected through a universal C 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-tube furnace with a 9-inch hot zone. All tests were conducted in air. nocam,-wuna 5.g
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 controller 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 indicated that this method is accurate to plus or minus 2'F. The yicl/ ioad, ultimate 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. CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials contained in Capsule V irradiated at 1.94 x 10 n/cm2 are presented in 19 Tables 5-1 through 5-8 and Figures 5-1 through 5-4. The transition temperature increases and upper she'If energy decreases for the Capsule V materials are shown in Table 5-9. Table 5-10 summarizes the Charpy impact test results from Capsule V with the previous Capsule T. Irradiation of vessel lower shell plate C4415-1 material (longitudinal l orientation) specimens to 1.94 x 10 m/cm2 (Figure 5-1) resulted in 30 19 i and 50 ft-lb transition temperature increases of 110'F and 130'F respectively, l* and an upper shelf energy decrease of 9 ft lbs. 1 Weld metal irradiated to 1.94 x 10 n/cm2 (Figure 5-2) resulted in a 30 19 ~' f t-lb transition temperature increase of 240'F and an upper shelf energy l decrease of ~ 20 ft-lb which resulted in an upper shelf energy of l approximately 50 ft-lb. l noo.mn.-ouuno 5-3
19 Weld HAZ metal irradiated to 1.94 x 10 n/cm2 (Figure 5-3) resulted in a 30 and 50 ft-lb transition temperature increases of 80'F and 85'F, respectively, and an upper shelf energy decrease of 8 ft-lb. However because ^ of the large data scatter these values are considered to be highly questionable. 19 Correlation monitor material (HSST Plate 02) irradiated to 1.94 x 10 n/cm2 (Figure 5-4) resulted in 30 and 50 ft-lb transition temperature increases of 145 and 150'F respectively. These increases are in good agreement with other irradiation program tests. Table 5-11 shows a comparison of the 30 ft-lb transition temperature (ARTNDT) increases for the various Surry Unit 1 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2.[6] This comparison shows that the transition temperature 19 2 increase resulting from irradiation to 0.281 and 1.94 x 10 n/cm is in relatively good agreement with the increase predicted by the Guide. 5-3. TENSION TEST RESULTS The results of tension tests performed on plate C4415-1 (longitudinal 19 2 orientation) and weld metal irradiated to 1.94 x 10 n/cm are shown in Table 5-12 and Figures 5-5 and 5-6, respectively. These resul.ts shown that irradiation produced an increase of 14 to 17 ksi in 0.2 percent yield strength for plate C4415-1 and ap' proximately a 23 to 27 ksi increase for the weld metal, 5-4. WEDGE OPENING LOADING TESTS At the request of the Virginia Elactric and Power Company, Wedge Open Loading (WOL) specimen will not be tested. The' specimens will be stored at the Hot Cell at the Westinghouse R&D Center. I i 2200s/033Ss-041347.10 5-4
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE SURRY UNIT 1 LOWER SHELL PLATE C4415-1 IRRADIATED AT 550*F, FLUENCE 1.94 x 10 n/cm2(E>1MeV) 19 Temperature Impact Energy Lateral Expansion Sample No. 'F (*C) ft-lbs (Joules) mils (mm) % Shear V50 50(10) i1.0 ( 15.0) 10.0 (0.25) 3 V52 100(38) 37.0 ( 50.0) 30.5 (0.77) V49 150(66) 50.0 ( 68.0) 43.0 (1.09) 41 V53 200 ( 93) 72.0 ( 97.5) 56.5 (1.44) 66 V54 250(121) 117.0(158.5) 79.5 (2.02) 100 V55 300(149) 116.0(157.5) 78.5(1.99) 100 V51 400(204) 115.0(156.0) 77.5 (1.97) 100 e I e noe.mn..ocune 5-5
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE SURRY UNIT 1 PRESSURE VESSEL WELD METAL. IRRADIATED AT 550*F, FLUENCE 1.94 x 10 n/cm2 (E > 1 WeV) 19 Temperature Irapact Energy Lateral Expansion Sample No. 'F ('C) ft-lbs (Joules) mils (m) % Shear ~ W12 50 ( 10) 4.0 ( 5.5) 3.5 (0.09) 0 W14 150 ( 66) 17.0 ( 23.0) 19.5 (0.50) 12 W13 200 ( 93) 22.0 ( 30.0) 17.5 (0.44) 28 W16 250(121) 39.0 ( 53.0) 28.5(0.72) 73 W10 250(121) 33.0 ( 44.5) 32.0(0.81) 52 W15 300(149) 41.0 ( 55.5) 31.0(0.79) 96 Wil 400(204) 47.0 ( 63.5) 41.0(1.04) 100 W9 450(232) 52.0 ( 70.5) 45.5(1.16) 100 noo.mni.-unerio 5-6
TABLE 5-3 CHARPY V-NOTCH IMPACT DATA FOR THE SURRY UNIT 1 ~ 4 PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL IRRADIATED AT 550*F, FLUENCE 1.94 x 10 n/cm2 (E > 1 MeV) 18 Temperature Impact Energy Lateral Expansion Sample No. 'F ('C) ft-lbs (Joules) mils (mm) % Shear H10 -25(-32) 12.0 ( 16.5) 10.0 (0.25) 12 H9 25 ( -4) 28.0(38.0) 20.5 (0.52) 31 H13 50 ( 10) 53.0.( 72.0) 31.5 (0.80) 51 H11 100 ( 38) 87.0 (118.0) 59.0 (1.50) 87 H15 150 ( 66) 22.0 ( 30.0) 20.0 (0.51) 58 H14 200 ( 93) 81.0 (110.0) 61.5 (1.56) 100 H12 300 (149) 52.0 ( 70.5) 39.0 (0.99) 100 H16 400(204) 110.0 (149.0) 70.0 (1.78) 100 l A t noo.mn.-m une 5-7
TABLE 5-4 CHARPY V-NOTCH IMPACT DATA FOR THE SURRY UNIT 1 A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL (HSST PLATE 02) 19 IRRADIATED AT 550*F, FLUENCE 1.94 x 10 n/cm2 (E > 1 MeV) Temperature Impact Energy lateral Expansion Sample No. 'F ('C) ft-lbs (Joules) mils (mm) % Shear R41 100 ( 38) 10.0 ( 13.5) 10.0 (0.25) 4 R43 150(66) 21.0 ( 28.5) 18.5 (0.47) 10 R47 200 ( 93) 33.0 ( 44.5) 27.5 (0.70) 33 R48 200 ( 93) 33.0 ( 44.5) 23.0 (0.58) 26 R46 250 (121) 73.0 ( 99.0) 45.0 (1.14) 43 R44 300(149) 92.0 (124.5) 69.0 (1.75) 92 R45 400 (204) 101.0 (137.0) 69.5(1.77) 100 R42 450 (232) 98.0 (133.0) 72.5 (1.84) 100 3200s/c33Ss 041Mt 10 5-8
l TABLE 5-5 INSTRLMENTED CHARPY IWACT TEST RESULTS FOR SURRY LNIIT 1 LOWER SHELL PLATE C4415-1 Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yleid Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2 No. (*F) (FT LB)
(FT-LP/in )--
(KIPS) (14Sec) (KIPS) (IISec) (KIPS) (KIPS) (KSI) (KSI) VSO 50 11.0 89 36 52 3.05 135 3.10 155 3.15 .25 102 102 VS2 100 37.0 298 210 88 3.10 125 4.40 495 4.35 .65 103 124 V49 150 50.0 403 258 145 3.15 135 4.50 585 4.45 1.55 103 126 v53 200 72.0 580 282 297 2.90 135 4.25 670 3.85 2.4 95 118 V54 250 117.0 942 283 659 2.75 135 4.10 690 91 113 4 V55 300 116.0 934 270 664 2.90 140 4.15 645 96 117 VS1 400 115.0 926 290 636 2.70 130 4.10 715 89 112 s 2200s-041387:10
.m._. 4 TABLE 5-6 INSTRLMENTED CHARPY IWACT TEST RESULTS FOR SURRY UNIT 1 WELD ETAL Normalized Energies Test Charpy Charpy mantmum Prop Yteld Time Itaximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yield Load Itaximum Load Load Stress Stress 2 No. (*F) (FT LB)
(FT-L8/in )-
(KIPS) (pSec) (KIPS) (pSec) (KIPS) (KIPS) (KSI) (KSI) W12 50 4.0 32 16 16 2.80 85 2.80 .15 W14 150 17.0 137 89 48 3.14 125 3.80 270 3.75 .50 102 113 W13 200 22.0 177 87 90 2.55 105 3.55 275 3.55 .95 85 101 W10 250 33.0 266 144 122 3.20 115 4.10 350 4.00 2.75 106 121 W16 250 39.0 314 165 s 149 3.55 130 4.25 380 4.20 3.25 117 129 T W15 300 41.0 330 140 190 3.05 125 3.80 370 101 113 5 W11 400 47.0 378 157 222 2.65 105 3.80 410 88 107 W9 450 52.0 419 184 235 3.40 140 4.20 435 113 126 l l 2200s-041387:10
.i a TABLE 5-7 ~ INSTRUl H IED CHARPY lirACT TEST RESULTS FOR SURRY UNIT 1 ELD EAT AFFECTED ZONE ETAL Normalized Energies Test Charpy Charpy maximum Prop Yleid Time Itaximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2 NO. (*F ) (FT LB)
(FT-L8/in )
gggp3). gp3,c) gggp3g gpg,c) gggp3) gggp3) gggg) gggg) e H10 -25 12.0 97 53 44 3.70 135 4.00 180 4.00 .30 122 127 ) H9 25 28.0 225 113 113 3.35 125 4.10 295 4.15 1.45 110 123 H13 50 53.0 427 225 202 .85 55 5.25 530 5.3 3.80 27 100 l H11 100 87.0 701 290 410 3.85 135 4.95 580 3.70 1.50 128 146 HIS 150 22.0 177 107 71 3.10 105 4.05 280 4.05 .80 102 118 T H14 200 81.0 652 232 420 3.75 130 4.70 485 124 140 C H12 300 52.0 419 155 264 3.15 135 3.95 400 105 118 H16 400 110.0 886 255 631 3.30 145 4.50 575 110 129 i 1 v i j i 2200s-041387:10 l ~
TABLE 5-8 INSTRtMENTED CHARPY IWACT TEST RESULTS FOR SURRY UNIT 1 AS33 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL (HSST PLATE 02) ] Normallzed Energies Test Charpy Charpy Maximum Prop Yleid Time Maximum Time to Fracture Arrest Y1 eld Flow l Sample Temp. Energy Ed/A Em/A Ep/A Load to Yleid Load Maximum Load Load Stress Stress i 2 No. (*F) (FT LD)
(FT-LB/in )-------
(KIPS) (USec) (KIPS) (USec) (KIPS) (KIPS) (KSI) (KSI) R41 100 10.0 81 50 31 3.30 110 3.60 160 3.75 .65 109 114 R43 150 21.0 169 125 44 3.20 100 3.90 325 3.90 .25 106 117 l R48 200 33.0 266 141 125 2.75 135 3.75 405 3.75 1.40 91 107 R47 200 33.0 266 164 102 2.90 135 3.90 445 3.90 1.35 95 112 R46 250 73.0 588 263 325 2.70 120 4.05 640 4.00 2.75 89 112 T R44 300 92.0 741 265 476 2.65 135 4.05 665 3.30 2.65 88 111 U R45 400 101.0 813 234 579 2.60 135 3.90 610 87 108 R42 450 98.0 789 263 526 2.80 115 4.15 625 92 115 22OOs-041387:10
s TABLE 5-9 EFFECT OF 550*F IRRADIATION AT 1.94 x 10 n/cm2 (E > MeV) I9 ON THE NOTCH TOUGHNESS PROPERTIES OF THE SURRY UNIT 1 REACTOR VESSEL MATERIALS Average Average 35 mit Average Average Energy Absorption 50 ft-Ib Temp (*F) Lateral Expansion Temp (*F) 30 f t-lb Temp (*F ) at Fu Stear (ft-lb) ' Material Untrradiated Irradiated AT untrradiated Irradiated AT Untrradiated Irradiated AT untrradiateo irradiated A(ft-le) Plate 20 150 130 10 130 120 -10 100 110 125 116 9 C4415-1 Weld 50 0 245 245 -15 225 240 70 49.5 20.5 Metal d W HAZ Metal -15 70 85 -20 70 90 -50 30 80 89 81 8 Correlatton 75 225 150 65 225 160 45 190 145 123 100 23 Mater 1al 22OOs-041387:10
TABLE 5-10 SLANARY OF SURRY UNIT NO.1 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS 68 Joule 41 Joule 50 ft-lb 30 f t-lb Decrease in Trans. Temp. Trans. Temp. Upper Shelf Fluence Increase Increase Energy 19 2 Material 10 n/cm (*F) (*F) (ft-lb) Plate C4415-1 (Long) 0.281(a) 60 50 5 1.940 (b) 130 110 9 Weld Metal 0.281 250-165 17 1.94 240 20.5 HAZ Metal 0.281 1.94 85 80 8 Correlation 0.281 80 70 18 Monitor 1.94 150 145 23 (a) Capsule T (b) Capsule V 1 e D noo.-o.intio 5-14
TABLE 5-11 COMPARISON OF MEASURED A RT ERSUS REGRATM GUIDE 1.M REVISIM 2 NDT (} PREDICTED ARTNDT A RTNDT (30 ft-lb increase) ~ Fluence RG. 1.99 Rev. 2 Measured 19 Material Capsule 10 n/cm2 (*F) (*F) Plate C4415-1 T 0.281 47 50 V 1.94 87 110 Weld Metal T 0.281 121 165 V 1.94 222 240 Correlation T 0.281 65 70 Monitor V 1.94 120 145 (a) Based on copper and nickel contents reported in WCAP 7723 [1]. l l~ i l l noe.-ou ano 5-15 L
n. TABLE 5-12 TENSILE PROPERTIES FOR SURRY UNIT 1 I8 2 REACTOR VESSEL MATERIAL IRRADIATED TO 1.94 x 10 n/cm Test 2% Yield Ultimate Fracture Fracture Fracture Ur.lform Total Reduction Sample Temp. Strength Strength Load Stress Strength Elongation Elongation in Area No. Material (*F) (ksi) (kst) (ktp) (kst) (kst) (%) (%) (%) V12 Plate C4415-1 '250 82.5 99.8 3.30 183.5 67.2 9.8 20.4 63 V11 Plate C4415-1 550 77.4 100.8 3.40 159.4 69.3 9.0 19.7 57 W4 WELD 250 90.7 102.3 3.90 180.9 79.5 38 19.7 56 W3 WELD 550 82.5 101.9 4.00 165.2 81.5 9.0 17.1 51 Y M t 2200s-041387:10 t-e e e e e
(*C) -50 O 50 100 150 200 250 I i i l i l i 100 I 80 nt 60 2 T 40 g U) 20 2 2 0 100 2.5 e 2 80 2 2.0 5 60 1.5 - X E W 40 120*r I.O - J U 2o p 20 0.5 O O 140 UNIRRADIATED 120 o 160 $ 100 S O IRRADIATED (550'F) l20 h 80 i.94 x sol 9n/cm2 ~ l 60 ,3g,g 40 - 2 ,,0 7 _e I y 40 w 20 [I I O O l -100 O I00 200 300 400 500 TEMPERATURE (*F) Figure 5-1. Irradiated Charpy V-Notch Impact Properties for Surry Unit 1 Reactor Vessel Lower Shell Plate C4415-1 ~ 012 A-20577-1 5-17 l
('C) -100 -50 0 50 100 150 200 250 I i i i i I I I 3 100 o 80 e -x ~ 60 e h 40 m O 20 o ^ O 100 2.5 m ~ 80 o 2.0 O W w 60 1.5 - 2 X O a E 1.0 - W 40 o 245 r _7e e J 20 -O e 0.5 2 d O O 80 100 70 O 60 0 80 ^ J UNIRRADIATED g 4 50 L. 60 o ~ 40 o e g 30 24o r 40 W U 20 8 1.94 x iol9n/cm2 8 IRRADIA E (550'r) 20 10 o O I I I O O -200 -100 O 100 200 300 400 500 TEMPERATURE (*F) Figure 5-2. Irradiated Charpy V-Notch Impact Properties for Surry Unit 1 Reactor Pressure Vessel Weld Metal 012-A-20577 2 5-18
(*C) -100 -50 0 50 100 150 200 250 I I l I i l i J 0 100 e 80 H ~ 60 e 40 o ko u) 20 8 O - 100 2.s e 0 2 80 0 2.0 6 e .o d
- 1. s -
2 60 X E
- r e
1.0 - W 40 2 J o e O.s 20 o p. 0 0 180 240 160 0 200 5 J 120 160 A b I00 WIMIAM 120 7 $ 80 0 I o o 5 80 6G ,m 0 IMIAE (550*F) 40 o
- F_
l.94 X IOI9n/cm2 20 o [I 9 I I I I g g -200 -100 O 100 200 300 400 500 TEMPERATURE (*F) Figure 5-3. Irradiated Charpy V Notch impact Properties for Surry Unit 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal 012-A-20577-3 5-19
(*C) -50 0 50 100 150 200 250 ~ l i l i 1 i i 100 80 60 tr5 40 b 2 20 -3 2 0 100 2.5 e n - 80 U 2.0 E ~ O a 60 1.5 g g X E W 40 ( 1.0 - iso'r 20 0.5 O O 160 200 140 O !20 160 ^ o 3 UNIRRADIATED 9 4 100 ~ i b 80 8 o 2 0 60 Ingaar47go (sso.r) - 80 l E 1.94 X loI9n/cm2 15 'r l ( w 40 g[C g as r i l i I O O -100 O !00 200 300 400 500 TEMPERATURE (*F) Figure 5-4. Irradiated Charpy V-Notch Impact Properties for Surry Unit 1 A533 Grade B Class 1 Correlation Monitor Material l (HSST Plate 02) 012-A-20577 4 5-20
(*c) O 50 100 150 200 250 300 120 I I I I I I 800 l10 = l 00 1 L 700 [ 600 Q-MATE TENSILE STREtK3TH g ^ m I @ 80 e ~ .2% YIELD STRENG 500 60 400 50 80 70 0 G _-g 60 m M REDUCTION IN AREA 50 CODE: Fg 40 OPEN POINTS - UNIRRADIATED CLOSED POINTS - IRRADIATED TO I.94 X lO I9n/cm2 30 o 7g7,L ggggg,77gg A a o 20 2 i l b n O \\O ? I ' UNIFORM ELONGATION O l O 100 200 300 400 500 600 TEMPERATURE (*F) Figure 5-5. Tensile Properties for Surry Unit 1 Reactor Vessel Lower Shell Plate C4415-1 012-A 20577-s 5-21
(*C) 0 50 100 150 200 250 300 120 i 800 IlO C 100 i i 700 ULTIMATE TENSILE STRENGTH - 90 o g 600 g @ 80 A g e a $ 70 0.2% YIELD STRENGTH ~ i 60 O oo-400 50 80 70 a n O 60 REDUCTION IN AREA M O 50 i CODE: F CLOSED POINTS - UNIRRADIATED 40 OPEN POINTS - IRRADIATED TO l.94 X IOI9n/cm2 H 30 O 6-- TOTAL ELONGATION g o 20 A G O O l 10 0 UNIFORM ELONGATION I I I O O 100 200 300 400 500 600 TEMPERATURE (*F) Figure 5-6. Tensile Properties for Surry Unit 1 Reactor Vessel Weld Metal 012.A 20577-6 5-22
~. E SECTION 6 ' RAD'IATION 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 property changes observed in materials test specimens, 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 between the environment at various positions within the reactor vessel and that experienced by the test specimens must be established. The former requirement-is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux i monitors contained in each of the surveillance ' capsules. The latter l-information, on the other hand, is derived solely from analysis. This section describes a discrete ordinates S transport analysis performed '~ n .for the Surry Unit i reactor to determine the fast neutron (E > 1.0 Mev) i flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules; and, in turn, to develop data for use in j relating neutron exposure of the pressure vessel;to that of the surveillance i capsules. Based on spectrum-averaged reaction cross sections derived from I this calculation, the analysis of the neutron dosimetry contained in Capsule V is discussed and updated evaluations of dosimetry from Capsules T and W are . presented. 6-2 OISCRETE ORDINATES ANALYSIS I l' A plan view of the Surry reactor geometry at th.e core midplane is shown in L Figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a 0*-45' i I noe.-ouwmo 6-1 I 4 - ~ - - - - - -
sector is depicted. - Eight irradiation capsules attached to the thermal sh'ield are included in the design.to constitute the reactor vessel surveillance program._ The capsules are located'at 45*, 55*, 65*, 165*, 245*, 285*, 295*, and 305* relative to the major axis at 0*.'(Refer to Figure 4-1.) I A plan view of a single surveillance capsule attached to the thermal shield is } 'shown in Figure 6-2. The stainless steel specimen container is 1-inch square and approximately 3 feet in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core. From a neutronic standpoint, the surveillance capsule structures are significant. In fact, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus between the thermal 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 Surry Unit 1 reactor geometry, two sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was utilized primarily to obtain spectrum-averaged reaction cross sections and gradient corrections for dosimetry reactions. The second set of calculations consisted of a series of adjoint analyses relating the fast neutron (E > 1.0 Mev) flux at the surveillance capsule locations and selected locations on the reactor vessel inner wall to the power distributions in the reactor core. These adjoint importance functions, when combined with cycle-specific core power distributions, yield the plant-specific fast neutron exposure at the sur-h veill'ance capsule and pressure vessel locations for each operating fuel cycle. Both the forward and adjoint calculations utilized an S6 angular l quadrature. The forward transport calculation was curied out in R,0 geometry using the l DOT two dimensional discrete ordinates code and the SAILOR cross-section ~ t 6 noo,* ma 6-2
library The SAILOR library is a 47 group, ENDF-BIV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with a P3 expansion of the cross sections. The energy group structure used in the analysis is listed in Table 6-1. ' The design basis core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 3-loop plants. Inherent in the development of this design basis core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuei, on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2a level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewha't conservative results. This is especially true in cases where low leakage fuel management has been employed. cross section The adjoint analyses were also carried out using the P3 approximation from the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positions along the inner diameter of the pressure vessel. Again, these calculations ~ were run in R,0 geometry to provide power distribution importance functions for the exposure parameters of interest. Having the adjoint importance functions and appropriate core power distributions, the response of interest is calculated as FFF R,0 ,RJejE (R,0,E)F(R,0,E)dERdRd8 I R where: R = Response of interest (e.g., 9 (E > 1.0 MeV)) at radius R,0 R and azimuthal angle 0. I (R,0,E) = Adjoint importance function at radius R and azimuthal angle 0 for neutron energy group E f noo.-on w.,o g.3
F (R,0,E) = Full power fission density at radius R and azimuthal angle 0 for neutron energy group E The fission density distributions used reflect the burnup-dependent inventory of fissioning actinides, including U-235, U-238, Pu-239, ar.d Pu-241. Core power distributions for use in the plant specific fluence evaluations for Surry Unit 1 are derived from measured assembly and cycle burnups for each operating cycle to date. The specific power distribution data used in the analysis is provided in Appendix A of WCAP 11015I93 The data listed in Appendix A represents cycle averaged relative assembly powers. Therefore, the adjoint results are in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence. Reactor vessel and surveillance capsule neutron fluence projections are made to several future dates. Current neutron fluences, based on past core . loadings, are defined as of the end of Cycle 8. Fluence projections are made to the expiration date of the operating license. The expiration date of the operating license for Surry Unit 1 is May 25, 2012 (forty years after the operating license was issued). In addition, projections are made to 60 calendar years beyond issuance of the operating license to illustrate the effect of a 20 year life extension. A few key assumptions are required to make the fluence projections. In particular, the cycle-averaged core power d.istribution for Cycle 8 and an 80% capacity factor are assumed to be representative of all future operation. Thus, all fluence projections reflect the low leakage fuel management strategies exemplified by the Cycle 8 core loading. Finally, it is assumed that the Surry Unit 1 core will be uprated from 2441 MWth to 2546 MWth at the beginning of Cycle 11. The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oakridge National Laboratory (ORNL) Poolside Critical Assembly (PCA) facility as well as against noo,* m;io 6-4
I the Westinghouse power reactor surveillance capsule data base [10]. The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be conservative by 7-22%. When plant specific power distributions are used with the adjoint importance functions, the benchmarking studies show that fluence predictions are within + 15% of measured values at surveillance capsule locations. 6-3 NEUTRON DOSIMETRY ' The passive neutron flux monitors included in the Surry Unit i surveillance program 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 materials properties changes. To properly account for burnout 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 are also included. The relative locations of the various monitors within the surveillance capsules are shown in Figure 4-2. The nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The iron monitors are obtained by drilling l samples from selected Charpy test specimens. In " type-II" capsules such as T, V, X and Z, cadmium-shielded neptunium and uranium fission monitors are accommodated within a dosimeter block located near the center of the capsule. The " type-I" capsules, including S, U, W and Y, do not contain the neptunium and uranium fission monitors. 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. l Rather, the activation or fission process is a measure of the integrated { effect that the time-and energy-dependent 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 J i I noo.-war io 6-5
from-the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: o The operating history of the reactor o The energy response of the monitor o The neutron energy spectrum at the monitor location o The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two procedures. First, the disintegration rate of product isotope per unit mass of monitor must 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 procedures.[11,12,13,14,15] Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium, Ge(Li), gamma spectromoter. The overall standard 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 Surry Unit 1, the overall 2o deviation in the measured data is determined to be +10 percent. The neutron energy spectra are determined analytically using the method described in Section 6-1. Having the measured activity of the monitcrs and the neutron energy spectra at the locations of interest, the calculation of the neutron flux proceeds as follows. G me.-ouw.io 6-6
The reaction product activity in the monitor is expressed as: N o(E),(E)fp*P. -At. -it N J d 2-C (1 - e )e (6-1) R=-{f$ Y 3 E j=1 where: induced product activity R = N, Avagadro's number = atomic weight of the target isotope A = weight fraction of the target isotope in the target material .f = g i number of product atoms produced per reaction Y = energy-dependent reaction cross section o(E) = time-averaged energy-dependent neutron flux at the monitor v(E) = location with the reactor at full power average core power level during irradiation period j P = 3 O noosao-osoto to g.7
e P maximum or reference core power level = max decay constant of the product isotope 1 = t = length of irradiation period j j d. decay time following irradiation period j t = 3 ((E>1.0 MeV) during irradiation period j divided by the C = average ((E>1.0 MeV) over the total irradiation period. C is calculated with the adjoint neutron transport method 3 and accounts for the change in neutron monitor response caused by core power distribution variations from cycle to cycle. P /P,,,, which accounts for the month-by-month variation 3 of power level within a cycle, is applied to the full power-based flux ratio, C. 3 Since neutron flux distributions are calculated using multigroup transport methods and, further, since 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 (6-1) is replaced by the following relation. c(E)#(E)dE = i #(E > 1.0 Mev) where: N a(E) e (E)dE { odgg G=1 g = N 1
- (E)dE
'g 1.0 Mev G= 1.0 Mov um.4o-mano 6-8
l Thus, equation (6-1) is rewritten N ["P. J), it -it. R=[N - d f Y F # (E > 1.0 Mev) C (1-e 3 g j=1 or, solving for the neutron flux, 9 (E > 1.0 Mev) = (6-2) N P. -it. -it [N d [ [ maxC3 (1-e J) e f Y3 4 j=1 The total fluence above 1.0 Mev is then given by ~ N p, t (E > 1.0 Mev) = (E > 1.0 Mev) [ [ max t (6-3) 3 j=1 where: 1 N p j total effective full power seconds of -{ I j=1 P,,x t) = o eration up to the time of capsule removal O i i noo,4a-onano 6-9
J An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered CoS9 (n,r) Co60 data by means of cadmium ratios and the use of a 37-barn 2200 m/sec cross section. Thus, D-1 (Th = (6-4) [N p P. -At. -Atd [ max(1-e J) e f Yo L g j=l where: Rbare .D is defined as RCd covered 6-4 TRANSPORT ANALYSIS RESULTS Calculated fast neutron (E > 1.0 Mev) exposure results for Surry Unit 1 are presented in Tables 6-3 through 6-10 and in Figures 6-3 through 6-7. lata is presented at several azimuthal locations on the inner radius of the pressure vessel as well as the center of each surveillance capsule. In Tables 6-3 through 6-6 cycle specific maximum neutron flux and fluence. levels at 0',15*, 30', and 45' on the pressure vessel inner radius are listed the first eight fuel cycles as well as projected to 60 years beyond issuance of the operating license. Similar data for the center of the surveillance capsules located at 15', 25*, 35', and 45' are give'n in Tables 6-7 through 6-10, respectively. Graphical presentations of the plant specific fast neutron fluence at key locations on the pressure vessel are shown in Figure 6-3 as a function of full power operating time. The pressure vessel data is presented for the O' location on the circumferential weld as well as for the 45' longitudinal ~ welds. Fast aeutron fluence at the surveillance capsule locations is shown as a function of full power operating time in Figure 6-4.
- ~"2"'"""
6-10
In-regard to Figure 6-3 and 6-4, the solid portions of the fluence curves are based directly on the cycle-specific core loadings as of the end of cycle 8. The dashed portions of these curves, however, involve a projection into the future. ' As mentioned in Section 6-3, the neutron flux average over Cycle 8 was used to project future fluence levels. It should be noted that implementation of a more severe low leakage pattern would act to reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those projections. In Figure 6-5, the azimuthal variation of maximum fast neutron (E > 1.0 MeV) fluence at the inner radius of the pressure vessel is presented as a function of azimuthal angle. Data are presented for both current and projected end-of-life conditions. In Figure 6-6, the relative radial variation of fast neutron flux and fluence within the pressure vessel wall is presented. Similar data showing the relative axial variation of fast neutron flux and fluence' over the beltline region of the pressure vessel is shown in Figure 6-7. A three-dimensional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figures *6-5 through 6-7 along with the relatica ~ $(R,0,Z) = ((0) F(R) G(Z) Fast neutron fluence at location R, 0, Z within where:
- (R,0,Z)
= the pressure vessel wall Fast neutron fluence at azimuthal location 0 on 4 (0) = the pressure vessel inner radius from Figure 6-5 L Relative fast neutron flux at depth R into the F(R) = pressure vessel from Figure 6-6 i l I j-nee,* ur in 6-11
.G (Z) Relative fast neutron flux at axial position Z = from Figure 6-7 4 Analysis has shown
- iat the radial and axial variations within the vessel wall are relatively insensitive to the implementation of low leakage fuel management schemes. Thus, the above relationship provides a vehicle for a reasonable wvaluation of fluence gradients within.the vessel wall.
6-5 00SIMETRY RESULTS The irradiation history of the Surry Unit I reactor is given in Table 6-11. The data were obtained from several sources including Nucleonics Week (16),, Surrysemi-annualoperatingreport{17), NUREG C020(18] and Virginia PowerI193. ~ Measured saturated activities of the flux monitors contained in Capsules T, W, and V are listed in Tables 6-12 through 6-14, respectively. The measured results for Capsules T and W were ~ derived from Battelle reports (3,4], whereas those for Capsule V were obtained by Westinghouse. The data are presented as measured at the actual monitor radial locations as well as adjusted to the capsule radial (191.15 cm) and azimuthal center where possible.- Adjustment factors for each monitor were obtained 'from reaction rate gradients through each capsule as calculated from the forward transport - calculation described in Section 6-2. In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cr'oss sections are required. The neutron energy spectrum at the radial and azimuthal center of each surveillance capsule, shown-in Table 6-15, was taken from the forward calculation. The resulting spectrum-averaged cross sections for each of the five fast neutron reactions are given in Table 6-16. The fast neutron (E > 1.0 Mev) flux levels derived for Capsules T, W, and V are presented in Tables 6-12 through 6-14, respectively. The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in Table 6-17. Due to the relatively low thermal neutron flux at the capsule locations, no um-ouwm 6-12
~ burnout correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than one percent 'for the'NiS8(n,p)CoS8 reaction and even less.significant for all of the oliher fast reactions. A comparison of the measured and calculated fast neutron fluence for each flux monitor of Capsules T, W, and V is shown in Table 6-18. Examination of the 4 data ir. Table 6-18 shows that. neutron fluences corresponding to the average of the monitors at each location agree within 5% of the calculated fluences based on-the plant-specific power distributions. It should be mentioned that, in the case of Capsule V, the excellent agreement between the measured fluences derived from the individual flux monitors and the calculated fluence is due largely to the use of the flux ratio, C, 3 mentioned in Section 6-3. Recall that C accounts for the impact of power 3 distribution changes on neutron monitor response. For example, low leakage core power distributions in cycles 6 and 8 caused the fast neutron (E>1.0 i MeV) flux et the 15' surveillance capsule location to be 15-16 percent lower S8 than the lifetime-average flux at that position. The Co product from the i S8 Ni n p reaction has a half-life of only 71 days, which implies that the nickel monitor is probably not sensitive to the irradiation history dating back more than one fuel cycle. Without the power distribution correction factor, the derived flux from the nickel monitor would have been low compared to the other monitors having significantly longer-lived reaction products. !~ Under these circumstances, the results.from the nickel monitor would have been ignored. I A similar, but not quite as severe, under prediction would have occurred if the flux was derived from-the iron monitor without use of the C factor 3 54 (Mn half-life is 312 days). Thus, one can see that meaningful results can j be obtained from the entire set of neutron dosimetry when the power distribution correction is made. This correction may ba even more important or necessary in future dosimetry analyses where Type I capsules, lacking the uranium and neptunium monitors, are examined. ux,-ou uno 6-13 =. -... ~...
6-6 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES As discussed in Section 6-4, plant specific fluence evaluations for the center of surveillance capsules located at 15*, 25*, 35*, and 45' were presented in Figure 6-4 for Surry Unit 1. The data presented on those curves represent the best available information upon which to base the future withdrawal schedules for capsules remaining in the Surry Unit i reactor. In the past, withdrawal schedules have been based on the assumption of a constant exposure rate at the surveillance capsule center and a constant lead factor relating capsule exposure to maximum vessel exposure. With the widespread implementation of low leakage fuel management neither of these assumptions can be assumed to be universally valid. It becomes prudent, therefore, to utilize the actual anticipated capsule exposure in conjunction, with appropriate materials properties data to establish capsule withdrawal dates that will provide expsrimental information that is of most benefit. In evaluating future withdrawal schedules, it must be remembered that the fluence projections shown in Figure 6-4 assume continued operation with the low leakage fuel management scheme currently in place.. The validity of this assumption should be verified as each new fuel cycle evolves and if significant changes occur withdrawal schedules should be adjusted accordingly. 6-7 INFLUENCE OF AN ENERGY DEPENDENT DAMAGE MODEL The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials property changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to a reduction in the uncertainties associated with damage trend ~ curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall. 220Cs-04 t sgy_ p g 6-14
Because of this potential shift away from a threshold fluence toward an energy . dependent damage function for data correlation, ASTM Standard Fractice E853 " Analysis and Interpretation of Light Water Reactor Surveillance Results", recommends reporting calculated displacements per iron atom (dPa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dPa function to be used for this evaluation is specified in ASTM Standard Practice E693 " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dPa)." For the Surry Unit 1 pressure vessel, iron atom displacement rates at each surveillance capsule location and at positions within the vessel wall have been calculated. The analysis has indicated that for a given location the . ratio of dPa/9(E > 1.0 MeV) is insensitive to changing core power distributions. That is, while implementation of low leakage loading patterns significantly impacts the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum at a given location are of second order. The dPa/v(E > 1.0 MeV) ratios calculated for key locations in the Surry reactor geometry are given in Table 6-19. The data in Table 6-19 may be used in conjunction with the fast neutron fluence data provided in Section 6-4 to develop distributions of dPa within the surveillance capsules and the reactor pressure ves'sel. nu,*nna 6-15
4 ^ TABLE 6-1 47 GROUP ENERGY STRUCTURE Lower Energy Lower Energy Group ~ (Mev) Grgug (Wev) 1 14.19* 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 -3 10 2.73 34 7.10 x 10 -3 11 2.47 35 3.36 x 10 ~3 12 2.37 36 1.59 x 10 -4 . 13 2.35 '37 4.54 x 10 ~ ~4 14 2.23 38 2.14 x 10 -4 15 '1.92 39 1.01 x 10 16-1.65 40 3.73lx 10-5 -5 17 1.35 41 1.07 x 10 -6 18 1.00 42 5.04 x 10 -6 . 19 0.821 43 1.86 x 10 -7 20 0.743 44 8.76 x 10 -7 21 0.608 45 4.14 x 10 -7 22 0.498 46 1.00 x 10 - 23 0.369 47 0.00 24 0.298
- The upper energy of group 1 is 17.33 Mev.
noo.-m mio 6-16
l } il 1 TABLE 6-2 j Fh . NUCLEAR.PAlAk kTERS FOR NEUTRON FLUX MONITORS , y 7 ( ;, Target ' Fission 4 .-Monitor
- R,eaction Weight
Response
Product Yield q) of/ Interest Fraction Range Half-Life (%) Material t ,c -q q Copper Cu63(li,a)Co60 0.6917 E>4.7 Mev 5.272 years YIr$n Fe54(n.p)Mn54 0.058 E>1.0 Mev 312.2 days Nickel NiS8(n,p)CoS8 0.6827 E>1.0 Mev 70.91 days e Uranium-238* U238(n,f)Cs137 1,'O E>0.4 Mev 30.17 ye'ars 6.0 r Neptunium-237* Np237(n,f)Cs137 1.0 E>0.08 Mev ' 30.17 years 6.5 Cobalt-Alumir.um* CoS9(n,r)Co60 0.0015 0'.4eV< E<0.015 Mev 5.272 years-Cobalt-Aluminum CoS9(n,r)Co60 t 0.0015 E<0.015 Mev 5.272 years
- Denotes that monitor is cadmium shielded.
I t 9 j, e
- j_
i e i
- rs
\\ 6 noo,wano 6-17 s s . l c L.--,, ..),, --n---,,-- --n ..n ...,n.----
TABLE 6-3 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE (*) l l Elapsed Irradiation Irradiation Avg.2 Flux Beltline Region Interval Time (EFPY) (n/cm sec) CumulativeFluence(n/cm$ 10 18 CY-1 1.1 5.03 x 10 1.70 x 10 10 18 CY-2 1.6 5.73 x 10 2.70 x 10 10 18 CY-3 ' 2.3 5.22 x 10 3.87 x 10 10 18 CY-4 3.4 4.86 x 10 5.49 x 10 10 18 CY-5 4.6 4.40 x 10 7.10 x 10 10 18 CY-6 5.9 3.96 x 10 8.75 x 10 10 19 CY-7 6.8 5.91 x 10 1.05 x 10 CY-8(b) 8.0 4.05 x 10 1.20 x 10 10 19 CY-9 + CY-10(c) 10.3 4.05 x 10 1.49 x 10 10 19 CY-11
- 6/25/2008(d) 25.6 4.22 x 10 3.54 x 10 10 19 6/25/2008 + 5/25/2012 *)
28.8 4.22 x 10 3.96 x 10 I 10 19 5/25/2012
- 5/25/2032(f) 44.8 4.22 x 10 6.09 x 10 10 19 (a) Applicable to the peak locations (0*, 90*, 180*, 270*) on the intermediate and lower shell plates and the intermediate to lower shell circumferential weld.
(b) Current neutron fluences are defined at the end~ of CY-8. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. B(yond the end of CY-8 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the original license expiration date. (e) 5/25/2012 corresponds to 40 calendar years beyond issuance of the operating license and is the license expiration date. (f) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating license, illustrating the effect of a 20 year life extension, noo.-o iur in 6-18
TABLE 6-4 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15' AZIMUTHAL ANGLE Elapsed Irradiation Irradiation Avg.2 Flux Beltline Region 2 Interval Time (EFPY) (n/cm -sec) Cumulative Fluence (n/cm ) 10 17 CY-1 1.1 2.40 x 10 8.12 x 10 10 18 CY-2 1.6 2.72 x 10 1.29 x 10 10 18 CY-3 2.3 2.49 x 10 1.84 x 10 10 18 CY-4 3.4 2.34 x 10 2.62 x 10 10 18 CY-5 4.6 2.06 x 10 3.38 x 10 10 18 CY-6 5.9 1.88 x 10 4.16 x 10 10 18 CY-7 6.8 2.50 x 10 4.92 x 10 10 18 CY-8(a) 8.0 1.88 x 10 5.62 x 10 10 18 CY-9 + CY-10(b) 10.3 1.88 x 10 6.96 x 10 10 19 CY-11 4 6/25/2008(c) 25.6 1.97 x 10 1.65 x 10 10 19 6/25/2008
- 5/25/2012(d) 28.8 1.97 x 10 1.84 x 10 10 19 5/25/2012 4 5/25/2032(')
44.8 1.97 x 10 2.84 x 10 (a) Current neutron fTuences are defined at the end of CY-8. (b) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond the end of CY-8 a 80% capacity factor is assumed. i (c) Exposure period from the onset of the uprating to the original license expiration date. (d) 5/25/2012 corresponds to 40 calendar years beycod issuance of the operating license and is the license expiration date. (e) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating license, illustrating the effect of a 20 year life extension. noo.-o.in7;io 6-19
TABLE 6-5 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE I VESSEL INNER RADIUS - 30* AZINUTHAL ANGLE Elapsed Irradiation Irradiation Avg.2 Flux Beltline Region 2 Interval Time (EFPY) (n/ cia -sec) Cumulative Fluence (n/cm ) 10 17 l CY-1 1.1 1.30 x 10 4.40 x 10 I 10 17' 'CY-2 1.6 1.54 x 10 7.09 x.10 10 18 CY-3 2.3 1.34 x 10 1.01 x 10 10 18 CY-4 3.4 1.30 x 10 1.44 x 10 10 18 CY-5 4.6 1.09 x 10 1.84 x 10 10 18 CY 5.9 1.02 x 10 2.27 x 10 9 18 CY-7 6.8 9.80 x 10 2.56 x~10 I 9 18 CY-8 ") 8.0 9.86 x 10 2.93 x 10 CY-9 + CY-10(b) 10.3 9.86 x 10 3.63 x 10 9 18 CY-11 4 6/25/2008(c) 25.6 1.03 x 10 8.62 x 10 10 18 4 6/25/2008 + 5/25/2012(d) 28.8 1.03 x 10 9.63 x 10 10 18 5/25/2012 - 5/25/2032(*) 44.8 1.03 x 10 1.48 x 10 10 19 (a) Current' neutron fluences are defined at the end of CY-8. (b) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond the'end of CY-8 a 80% capacity factor is assumed. -(c) Exposure period from the onset of the uprating to the original license expiration date. _ (d) 5/25/2012 corresponds to 40 calendar. years beyond issuance of the opere d ng license and is the license expiration date. (e) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating ~ license, illustrating the effect of a 20 year life extension. k noo.-o.im;io 6-20
r TABLE 6-6 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 NeV) EXPOSURE AT THE PRESSURE VESSEL.. INNER RADIUS - 45 AZIMUTHAL ANGLE (*) Elapsed ' Irradiation Irradiation Avg.2 Flux Beltline Region 2 Interval Time (EFPY) (n/cm -sec) Cumulative Fluence (n/cm ) 9 17 CY-1 1.1-8.59 x 10 2.91 x 10 10 17 CY-2 1.6 1.05 x 10 4.75 x 10 9 17 CY-3 2.3 9.08 x 10 6.78 x 10 9 17 CY-4' 3.4 8.71 x 10 9.68 x 10 9 18 CY-5 4.6 7.11 x 10 1.23 x 10 9 18. CY-6 5.9 6.86 x 10 1.51 x 10 9 18 CY-7 6.8 6.14 x 10 1.70 x 10 9 18 CY-8(b) 8.0 6.54 x 10 1.94 x 10 'CY-9 + CY-10(c) 10.3 6.54 x 10 2.41 x 10 9 18 9 18 - CY-11
- 6/25/2008(d) 25.6 6.82 x 10 5.71 x 10 9
18 6/25/2008 4 5/25/2012(') 28.8 6.~82 x 10 6.39 x 10 5/25/2012 4 5/25/2032(f) 44.8 6.82 x 10 9.83 x 10 9 18 (a) Applicable to the longitudinal welds at 45*., 135', 225', 315' in the peak axial flux. (b) Current neutron fluences are defined at the end of CY-8. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond the end of CY-8 a 80% capacity factor is assumed. (d) -Exposure period from the caset of the uprating to the original license expiration date. -(e) 5/25/2012 corresponds to 40 calendar years beyond issuance of the operating license and is the license expiration date. (f) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating license, illustrating the effect of a 20 year life extension. noo.-mune 6-21
TABLE 6-7 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15* SURVEILLANCE CAPSULE CENTER Elapsed Irradiation Irradiation Avg.2 Flux Beltline Region 2 Interval Time (EFPY) (n/cm -sec) Cumulative Fluence (n/cm ) 10 18 CY-1 1.1 8.31 x 10 2.81 x 10 10 18 CY-2 -1.6 9.42 x 10 4.46 x 10 10 18 CY-3 2.3 8.61 x 10 6.39 x 10 10 18 CY-4 3.4 8.11 x 10 9.08 x 10 10 19 CY-5 4.6 7.08 x 10 1.17 x 10 10 19 CY-6 5.9 6.47 x 10 1.44 x 10 10 19 CY-7 6.8 8.76 x 10 1.70 x 10 CY-8(a) 8.0 6.46 x 10 1.94 x 10 10 19 CY-9 + CY-10(b) 10.3 6.46 x 10 2.40 x 10 10 19 CY-11 4 6/25/2008(c) 25.6 6.74 x 10 5.67 x 10 10 19 6/25/2008 - 5/25/2012(d) 28.8 6.74 x 10 6.34 x 10 10 19 5/25/2012 4 5/25/2032 ') 44.8 6.74 x 10 9.74 x 10 I 10 19 (a) Current neutron fluences are defined at the-end of CY-8. (b) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond the end of CY-8 a 80% capacity factor is assumed. (c) Exposure period from the onset of the uprating to the original license expiration date. (d) 5/25/2012 corresponds to 40 calendar years beyond issuance of the perating license and is the license expiration date. (e) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating license, illustrating the effect of a 20 year life extension. n oo.-o.i m :ia 6-22
TABLE 6-8 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 25* SURVEILLANCE CAPSULE CENTER Elapsed Irradiation Irradiation Avg.2 Flux Beltline Region 2 Interval Time (EFPY) (n/cm -sec) Cumulative Fluence (n/cm ) 10 18 CY-1 1.1 5.26 x 10 1.78 x 10 10 18 CY-2 1.6 6.14 x 10 2.85 x 10 10' 18 CY-3 2.3 5.40 x 10 4.06 x 10 10 18 CY-4 3.4 5.24 x 10 5.81 x 10 10 18 CY-5 4.6 4.48 x 10 7.44 x 10 10 18 CY-6 5.9 4.15 x 10 9.17 x 10 10 19 CY-7 6.8 4.17 x 10 1.04 x 10 10 19 CY-8(8) 8.0 4.02 x 10 1.19 x 10 10 19 CY-9 + CY-10(b) 10.3 4.02 x 10 1.48 x 10 10 19 CY-11 4 6/25/2008(C) 25.6 4.20 x 10 3.52 x 10 6/25/2008 4 5/25/2012(d) 28.8 4.20 x 10 3.93 x 10 10 19 5/25/2012 4 5/25/2032(*) 44.8 4.20 x 10 6.05 x 10 10 19 (a) Current neutron fluences are defined at the end of CY-8. (b) At the beginning of CY-11, the core thermal power will be unrated to 2546 MWth. Beyond the end of CY-8 a 80% capacity factor is assumed. (c) Exposure period from the onset of the uprating to the original license expiration date. (d) 5/25/2012 corresponds to 40 calendar years beyond issuance of the operating license and is the license expiration date. (e) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating license, illustrating the effect of a 20 year life extension. me.enna 6-23
TABLE 6-9 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 35' SURVEILLANCE CAPSULE CENTER Elapsed Irradiation Irradiation Avg.2 Flux Beltline Region 2 Interval Time (EFPY) (n/cm -sec) Cumulative Fluence (n/cm ) = 10 18 CY-1 1.1 3.56 x 10 1.21 x 10 10 18 CY-2 1.6 4.28 x 10 1.95 x 10 4 10 18 CY-3 2.3 3.71 x 10 2.78 x 10 10 18 CY-4 3.4 3.58 x 10 3.97 x 10 10 18 CY-5 4.6 2.93 x 10 5.05 x 10 10 18 CY-6 5.9 2.78 x 10 6.21 x 10 10 18 CY-7 6.8 2.58 x 10 6.99 x 10 CY-8(a) 8.0 2.68 x 10 7.99 x 10 10 18 CY-9+CY-10(b)' 10.3 2.68 x 10 9.90 x 10 10 18 CY-11 4 6/25/2008(c) 25.6 2.80 x 10 2.35 x 10 5 10 19 6/25/2008 + 5/25/2012(.d) 28.8 2.80 x 10 2.62 x 10 10 19 '5/25/2012 + 5/25/2032(*) 44.8 2.80 x 10 4.04 x 10 10 19 (a) Current neutron flut.nces are defined at the end of CY-8. .(b) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond the end of CY-8 a 80% capacity factor is assumed. (c) Exposure period from the onset of the uprating to the original license - expiration date. (d)-5/25/2012 corresponds to 40 calendar years beyond issuance of the operating license and is the license expiration date. 1 i (e) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating license, illustrating the effect of a 20 year life extension. 4 n + c -: oo -war ia 6-24 y g'* w--.ip
- --w-.r v.
y-e+--.---re. .e +F---*--=- 9-*'"'* e-t-iev* e-mv*e-ew* mew-+e--w=r-*-wm-----rw--
"r-
c-TABLE 6-10 SURRY UNIT 1 CALCULATED FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45* SURVEILLANCE CAPSULE CENTER Elapsed Irradiation Irradiation Avg.2 Flux Beltline Region 2 Interval Time (EFPY) (n/cm -sec) Cumulative Fluence (n/cm ) 10 17 CY-1 1.1 2.79 x 10 9.45 x 10 10 18 CY-2 1.6 3.43 x 10 1.55 x 10 10 18 CY-3 2.3 2.95 x 10 2.21 x 10 10 18 CY-4 3.4 2.83 x 10 3.15 x 10 10 18 CY-5 4.6 2.29 x 10 3.98 x 10 10 18 CY-6 5.9 2.21 x 10 4.91 x 10 10 18 CY-7 6.8 1.97 x 10 5.50 x 10 10 18 CY-8(a) 8.0 2.10 x 10 6.28 x 10 10 18 CY-9 + CY-10(b) 10.3 2.10 x 10 7.78 x 10 10 19 CY-11
- 6/25/2008(c) 25.6 2.19 x 10 1.84 x 10 10 19 6/25/2008
- 5/25/2012(d) 28.8 2.19 x 10 2.06 x 10 10 19 5/25/2012
- 5/25/2032(*)
44.8 2.19 x 10 3.16 x 10 (a) Current neutron fluences are defined at the end cf CY-8. (b) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond the end of CY-8 a 80% capacity factor is assumed. (c) Exposure period from the onset of the uprating to the original license expiration date. (d) 5/25/2012 corresponds to 40 calendar _ years beyond issuance of the operating license and is the license expiration date. ~ (e) 5/25/2032 corresponds to 60 calendar years beyond issuance of the operating license, illustrating the effect of a 20 year life extension. noe.-oem.io 6-25
TABLE 6-11 IRRADIATION HISTORY OF SURRY UNIT 1 . REACTOR VESSEL SURVEILLANCE CAPSULE V P P J MAX IRRADIATION DECAY P /P ' MONTH YEAR' (WW) (MW) J RAX TIME (DAY) TIME (DAY) 7 1972 83 2441 0.034 31 5202 8' -1972 83 2441 0.034 31 5171 9 1972 436~ 2441 0.178 30 5141 10 1972 0 2441 0.000-31 5110 11 1 1972 1 2441 0.000 30 5080 12 1972 1129 2441 0.463 31 5049 1 1973 285 2441 0.117 31 5018 2 1973 1649 .2441 0.676 28 4990 3 1973 1416 2441 0.580 31 4959 4 1973 1371 2441 0.562 30 4929 5 1973 1361 2441 0.557 31 4898 6 1973 1181 2441 0.484 30 4858 7 1973 1838 2441 0.753 31 4837 8 1973 1777 2441 0.728 31 4806 9 1973 1250 2441 0.512 30 4776. 10 1973 674 2441 0.276 31 4745 11 1973 2073 2441 0.849 30 4715 12 1973 0 2441 0.000 31 4684 1 1974 0 2441 0.000 31 4653 2 1974 0 2441 0.000 28 4625 3 1974' 1047 2441 0.429 31 4594 4 1974 2192 2441 0.898 30 4564 5 1974 2073 2441 0.849 31 4533 6 1974 .2224 2441 0.911 30 4503 7 1974 1400 2441 0.574 31 4472 8 1974 2384 2441 0.977 31 4441 9 1974 2095 2441 0.858 30 4411 10 1974 1129 2441 0.462 31 4380 Capsule T Removed 11 1974 0 2441 0.000 30 4350 12 1974 0 2441 0.000 31 4319 1 1975 1 2441 0.000 31 4288 2 1975 1979 2441 0.811 28 4260 3 1975 1873 2441 0.751 31 4229 4 1975 20s3 2441 0.839 30 4199 5 1975 2302 2441 0.943 31 4168 6 1975 2176 2441 0.891 30 4138 7 1975 1799 2441 0.737 31 4107 8 1975 2180 2441 0.893 31 4076 9 1975 1971 2441 0.807 30 4046 10 1975 0 2441 0.000 31 4015 11 1975 0 2441 0.000 30 3985 12 1975 1209 2441 0.495 31 3954 noo.-ou ar.io 6-26
r TABLE 6-11 (Continued) IRRADIATION HISTORY OF SURRY UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE V P P g MAX IRRADIATION DECAY P /P ' MONTH YEAR (MW) (MW) J MAX TIME (DAY) TIME (DAY) 1 1976 2430 2441 0.995 31 3923 2 1976 2399 2441 0.983 29 3894 3 1976 1634 2441 0.669 31 3863 4 1976 -1894 2441 0.776 30 3833 5 1976 2022 2441 0.828 31 3802 6 1976 2423 2441 0.992 30 3772 7 1976 1777 2441 0.728 31 3741 8 1976 1834 2441 0.751 31 3710 9 1976 1914 2441 0.784 30 3680 10 1976 1258 2441 .0.516 31 3649 11 1976 0 2441 0.000 30 3619 12 1976 0 2441 0.000 31 3588 1 1977 600 2441 0.246 31 3557 2 1977 2143 2441 0.878 28 3529 3 1977 2414 2441 0.989 31 3498 4 1977 861 2441 0.353 30 3468 5 1977 1319 2441 0.540 31 3437 6 1977 2441 2441 1.000 30 3407 7 1977 2389 2441 0.979 31 3376 8 1977 2213 2441 0.906 31 3345 9 1977 2402 2441 0.984 30 3315 10 1977 2441 2441 1.000 31 3284 11 1977 1623 2441 0.665 30 3254 12 1977 1906 2441 0.781 31 3223 1 1978 2436 2441 0.998 31 3192 2 1978 2434 2441 0.997 28 3164 3 1978 2437 2441 0.998 31 3133 4 1978 1703 2441 0.698 30 3103 Capsule W Removed 5 1978 0 2441 0.000 31 3072 6 1978 0 2441 0.000 30 3042 7 1978 1708 2441 0.700 31 3011 8 1978 2383 2441 ~0.976 31 2980 9 1978 2148 2441 0.880 30 2950 10 1978 2430 2441 0.996 31 2919 11 1978 2425 2441 0.993 30 2889 12 1978 785 2441 0.322 31 2858 1 1979 2287 2441 0.937 31 2827 2 1979 2443 2441 1.001 28 2799 3 1979 1097 2441 0.450 31 2768 4 1979 0 2441 0.000 30 2738 5 1979 0 2441 0.000 31 2707 6 1979 0 2441 0.000 30 2677 noo.-o w no 6-27
TABLE 6-11'(Continued) ~ IRRADIATION HISTORY OF SURRY UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE V P P J MAX IRRADIATION DECAY P /P MONTH YEAR (MW) (MW) J MAX TIME (DAY) TIME (DAY) 7 1979 0 2441 0.000 31 2646 8 1979 0 2441 0.000 31 2615 9 1979 0 2441 0.000 30 2585 10 1979 488 2441 0.200-31 2554 11 1979 2428 2441 0.995 30 2524 12 1979 1463 2441 0.599 31 2493 1 1980 1798 2441 0.736 31 2462 2 1980 1582 2441 0.648 29 2433 3 1980 0' 2441 0.000 31 2402 14 1980 0 2441 0.000 30 2372 5 1980 1520 2441 0.623 31 2341 6 1980 2361 2441 0.967 30 2311 7 1980 2341 2441 0.959 31 2280 8 1980 1405 2441 0.576 31 2249 .9 1980 859 2441 0.352 30 2219 10 1980 0 2441 0.000 31 2188 11 1980 0 2441 0.000 30 2158 12 1980 0 2441 0.000 31 2127 1 '1981 0 2441 0.000 31 2096 2 1981 0 2441 0.000 28 2068 3 1981 0 2441 0.000' 31 2037 4 1981 0 2441 0.000 30 2007 3 1981 0 2441 0.000 31 1976 6 1981 0 2441 0.000 30 1946 7 1981 1427 2441 0.584 31 1915 8 1981 2340 2441 0.959 31 1884 9 1981 1391 2441 0.570 30 1854 10 1981 1981 2441 0.812 31 1823-11 1981 2214 2441 0.907 30 1793 12 1981 1441 2441 0.590 31 1762 1 1982 2141 2441 0.877 31 1731 2 1982 1120 2441 0.459 28 1703 3 1982 2332 2441 0.955 31 1672 4 1982 2021 2441 0.828 30 1642 5 1982 2441 2441 1.000 31-1611 6 1982 2437 2441 0.998 30 1581 7 1982 2354 2441 0.964 31 1550 i 8 1982 2311 2441 0.947 31 1519 9 1982 2426 2441 0.994 30 1489 10 1982 997 2441 0.408 31 1458 11 1982 2104 2441 0.862 30 1428 12 1982 2374 2441 0.973 31 1397 mm.-m une 6-28
n TABLE 6-11 (Continued) IRRADIATION HISTORY OF SURRY UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE V P P J MAX IRRADIATION DECAY P /P MONTH YEAR (MW) (MW) g MAX TIME (DAY) TIME (DAY) 1 1983 2373 2441 0.972 31 1366 2 1983 510 2441 0.209 28 1338 3 1983 0 2441 0.000 31 1307 4 1983 0 2441 0.000 30 1277 5 1983 0 2441 0.000 31 1246 6 1983-554 2441 0.227 30 1216 7 1983 2370 2441 0.971 31 1185 8 1983 2430 2441 0.996 31 1154 9 1983 1574 2441 0.645 30 1124 10 1983 1950 2441 0.709 31 1093 11 1983 2330 2441' O.5;5 30 1063 12 1983 2561 2441 1.049 31 1032 -1 1984 2264 2441 0.927 31 1001 2 1984 1928 2441 0.790 29 972 3 1984 1595 2441 0.654 31 941 4 1984 1047 2441 0.429 30 911 5 1984 1956 2441 0.801 31 880 6 1984 1335 2441 0.547 30 850 7 1984 1387 2441 0.568 31 819 8 1984 1910 2441 0.783 31 788 9 1984 1672 2441 0.685 30 758 10 1984 0 2441 0.000 31 727 11 1984 0 2441 0.000 30 697 12 1984 62 2441 0.025 31 666 1 1985 1737 2441 0.711 31 635 2 1985 2426 2441 0.994 28 607 3 1985 2440 2441 1.000 31 576 4 1985 2169 2441 0.888 30 546 5 1985 1248 2441 0.511 31 515 6 1985 2441 2441 1.000 30 485 7 1985 2325 2441 0.952 31 454 8 1985 860 2441 0.352 31 423 9 1985 2017 2441 0.826 30 393 10 1985 2232 2441 0.914 31 362 11 1985 2321 2441 0.951 30 332 12 1985 2325 2441 0.952 31 301 1 1986 1579 2441 0.647 31 270 2 1986 1704 2441 0.698 28 242 '3 1986 2439 2441 0.999 31 211 4 1986 2394 2441 0.981 30 181 5 1986 1947 2441 0.798 10 171 Decay time is referenced to 10/28/86. noo.-om e.ie 6-29 ~.
TABLE 6-12 ~ MEASURED FLUX MONITOR ACTIVITIES FROM SURRY UNIT 1, CAPSULE T* Adjusted Radial Saturated Saturated Measured Reaction and Location Activity Activity ( (E>.0 Mev) (n/cm{-sec) Axial Location (cm) (DPS/gm) (DPS/gm) Fe54(n,p)Mn54 - 0 10 Top 4.44 x 10 9.63 x 10 6 10 Middle 3.71 x 10 8.01 x 10 6 10 Bottom 3.45 x 10 7.45 x 10 10 Average 8.36 x 10 NiS8(n,p)Co58 7 7 10 Middle 190.92 6.11 x 10 5.84 x 10 8.24 x 10 Cu63(n,a)Co60 4.15x10f 3.98x10! 8.84x10h Top 190.92 Bottom 190.92 4.24 x 10 4.06 x 10 9.03 x 10 10 Average 8.94 x 10 Np 37(n,f)Cs137 7 10 Middle 191.15 3.23 x 10 7.01 x 10 U238(n,f)Cs137, 6 10 Middle 191.15 4.40 x 10 8.19 x 10
- U238(n,f)Cs137 activities corrected by a factor of 0.89 for 300 ppm 'J-235
. impurity and Pu-239 buildup. +Neither measured activities nor radial locations were reported in Reference 3. Saturated activities were determined from the reported nuclear constants and resultant flux levels. Adjusted saturated activities are not given for the iron monitors because the radial and azimuthal position of the charpy chips is uncertain. noo. mune 6-30
TABLE 6-13 ~ MEASURED FLUX MONITOR ACTIVITIES FROM SURRY UNIT 1, CAPSULE W ") I Adjusted Radial Saturated Saturated Measured Reaction and Location Activity Activity e(E>}.0Mev) Axial Location (cm) (DPS/gm) (DPS/gm) (n/cm -sec) Fe54(n,p)Mn54 Mid Top 1.99x10f 3.69x10$ Mid Bottom 1.94 x 10 3.60 x 10 10 Average 3.64 x 10 NiS8(n.p)CoS8 7 7 10 Middle 190.92 3.47 x 10 3.31 x 10 4.10 x 10 Cu63(n,a)Co60 2.38x10! 2.28x10l 3.92x10$ Top 190.92 Bottom 190.92 2.42 x 10 2.32 x 10 3.98 x 10 10 Average 3.95 x 10
- )Neither measured activities nor radial locations were reported in Reference 4.
Saturated activities were determined from the reported nuclear constants and resultant flux levels. Adjusted saturated activities are not given for the iron monitors because the radial and azimuthal position of the tensile specimen chips is uncertain. O noo.e nna 6-31 1
TABLE 6-14 ~ MEASURED FLUX MONITOR ACTIVITIES FROM SURRY UNIT 1, CAPSULE V Adjusted Radial Saturated Saturated Measured Reaction and Location Activity Activity 9(E>{.0Nev) Axial Location (cm) (DPS/gm) (DPS/gm) (n/cm -sec) Fe54(np)Mn54 6 6 10. R-41 191.92 3.13 x.10 3.54 x 10 7.49 x 10 6 6 10 H-10 191.92 3.16 x 10 3.74 x 10 7.89 x 10 6 6 10 V-50 190.92 3.86 x 10 3.77 x 10 7.95 x 10 10 Average 7.78 x 10 NiS8(n.p)CoS8 7 7 10 Middle 190.92 5.45 x 10 5.21 x 10 7.30 x 10 CuS3(n a)Co60 5 5 10 Top 190.92 3.84 x 10 3.68 x 10 8.17 x 10 Np237(n,f)Cs137 7 7 10 Middle 191.15 3.22 x 10 3.22 x 10 7.02 x 10 U238(n f)Cs137, 6 6 10 Middle 191.15 4.13 x 10 4.13 x 10 7.69 x 10
- U238(n,f)Cs137 activities corrected by a factor of 0.84 for 300 ppm U-235 impurity and Pu-239 buildup.
e noa..m une 6-32
e. TABLE 6-15 CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF SURRY UNIT 1 SURVEILLANCE CAPSULES 2 Neutron Flux (n/cm -sec) Group No. 15' Capsules T & V 35' Capsule W 7 7 1 2.64 x 10 1.62 x 10 7 7 2 9.64 x 10 5.85 x 10 8 9 3 3.34 x 10 1.95 x 10 9 8 4 6.08 x 10 3.48 x 10 9 8 5 1.01 x 10 5.62 x 10 9 9 6 2.26 x 10 1.23 x 10 9 9 7 3.11 x 10 1.63 x 10 9 9 8 6.17 x 10 3.01 x 10 9 9 9 5.40 x 10 2.48 x 10 9 9 10 6.97 x 10 1.99 x 10 9 9 11 5.16 x 10 2.30 x 10 9 9 12 2.57 x 10 1.14 x 10 8 8 13 7.81 x 10 3.45 x 10 9 9 -14 3.82 x 10 1.68 x 10 9 9 15 9.81 x 10 4.29 x 10 10 9 16 1.25 x 10 5.30 x 10 10 9 17 1.18 x 10 7.73 x 10 10 10 18 3.52 x 10 1.43 x 10 10 9 19 2.52 x 10 9.97 x 10 10 9 20 1.22 x 10 4.82 x 10 10 10 l 21 3.78 x 10 1.44 x 10 Note: These spectra were obtained from the forward 00T calculation using a design basis core power distribution with a core thermal power rating of 2900 MWt. noe.-ou nna 6-33
TABLE 6-15(Continued) CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF SURRY UNIT 1 SURVEILLANCE CAPSULES 2 Neutron Flux (n/cm -sec) Group No. 15'. Capsules T & V 35' Capsule W 10 10 22 2.90 x.10 1.09 x 10 10 10 23 3.40 x 10 1.29 x 10 10 10 24 3.09 x 10 1.16 x 10 10 10 25 3.99 x 10 1.50 x 10 10 10 26 3.87 x 10 1.44 x 10 10 10 27 3.07 x 10 1.14 x 10 10 9 28 2.27 x 10 8.37 x 10 9 9 29 7.49 x 10 2.77 x 10 9 9 30 4.19 x 10 1.55 x 10 9 9 31 9.18 x 10 3.36 x 10 9 9 32 5.52 x 10 2.01 x 10 10 9 33 1.29 x 10 4.74 x 10 10 9 34 1.97 x 10 7.24 x 10 10 10 35 3.03 x 10 1.04 x 10 10 9 36 2.52 x 10 9.21 x 10 10 10 37 3.81 x 10 1.39 x 10 10 9 38 2.18 x 10 7.94 x 10 10 9 39 2.35 x 10 8.52 x 10 10 40 3.17 x 10 1.15 x '1010 10 10 41 3.87 x 10 1.39 x 10 10 9 42 2.22 x 10 7.95 x 10 10 9 43 2.69 x 10 9.63 x 10 10 9 44 1.78 x 10 6.37 x 10 10 9 ~ 45 1.50 x 10 5.37 x 10 10 10 46 2.88 x 10 1.03 x 10 10 10 47 5.30 x 10 1.91 x 10 4 am.-mun e 6-34 ~. -. -.
TABLE 6-16 SPECTRUM AVERAGED REACTION CROSS-SECTIONS AT THE CENTER OF SURRY UNIT 1 SURVEILLANCE CAPSULES o (barns) Capsules T & V Capsule W Reaction (15*) (35') CuS3(n a)Co60 0.00068 0.00088 Fe54(n.p)Mn54' O.074 0.086 Ni58(n,p)CoS8 0.100 0.114 U238(n,f) 0.354 Np237(n,f) 2.79 a(E)e(E)dE 0 g = v(E)dE 1.0 Mev F e um..io-emer io 6-35
TABLE 6-17 THERNAL NEUTRON FLUX DATA FRGt CAPSULES T, W, AND V Axial SaturatedActivity(DPS/gm) eth Adjusted 4th 2 2 Capsule Location Bare Cd-Covered (n/cm-sec) (n/cm -sec) 7 7 10 10 T Top 7.32 x 10 2.76 x 10 8.04 x 10 6.47 x 10 i 7 7 10 10 Middle 6.34 x 10 2.27 x 10 7.18 x 10 5.77 x 10 7 7 10 10 Bottom 6.87 x 10 2.57 x 10 7.59 x 10 6.10 x 10 7 7 10 10 W Top 2.35 x 10 1.05 x 10 2.29 x 10 1.83 x 10 7 6 10 10 Middle 2.35 x 10 7.56 x 10 2.81 x 10 2.24 x 10 7 6 10 10 Bottom 2.63 x 10 9.60 x 10 2.95 x 10 2.34 x 10 7 7 10 10 V Top 5.30 x 10 2.42 x 10 5.08 x 10 4.08 x 10 7 7 10 10 Middle 4.33 x 10 1.66 x 10 4.71 x 10 3.78 x 10 7 7 10 10 Bottom 4.47 x 10 1.70 x 10 4.89 x 10 3.93 x 10 e noo.e nne 6-36
TABLE 6-18 ~ CONPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUENCES FOR CAPSULES T, W, AND V* Irradiation 6 (E>1.0 MeV) 2 Time (n/cm ) Capsule Reaction (EFPS) _ Measured Calculated T Fe54(n,p)Mn54 7 18 3.39 x 10 2.83 x 10 NiS8(n.p)CoS8 18 2.79 x 10 Cu63(n,a)Co60 18 3.03 x 10 Np237(n,f)Cs137 18 2.38 x 10 U238(n,f)Cs137 19 2.78 x 10 18 18 Average 2.76 x 10 2.81 x 10 W Fe54(n,p)Mn54 8 18 1.07 x 10 3.89 x 10 NiS8(n,p)CoS8 18 4.39 x 10 Cu63(n,a)Co60 18 4.23 x 10 18 18 Average 4.17 x 10 3.97 x 10 Y Fe54(n.p)Mn54 8 19 2.53 x 10 1.97 x 10 NiS8(n.p)CoS8 19 1.85 x 10 Cu63(n,a)Co60 19 2.07 x 10 Np237(n,f)Cs137 19 1.78 x 10 U238(n,f)Cs137 19 1.94 x 10 19 19 Average 1.92 x 10 1.94 x 10
- Measured data have been adjusted to the radial and azimuthal center of the 238 capsule where possible.
In addition, corrections were made to the U 235 239 monitor activites to account for the U impurity and build-in of Pu O me,*nr io 6-37
TABLE 6-19 dPa/p(E > 1.0 MeV) RATIOS FOR SURRY UNIT 1 Location dPa/v(E > 1.0 MeV) -21 15' CAPSULE 1.69 x 10 -21 25' CAPSULE -1.65 x 10 -21 35* CAPSULE 1.63 x 10 -21 45' CAPSULE 1.62 x 10 -21 VESSEL INNER RADIUS (0*) 1.62 x 10 -21 VESSEL 1/4 THICKNESS (0') 1.87 x 10 -21 VESSEL 3/4 THICKNESS (0*) 2.95 x 10 -21 VESSEL INNER RADIUS (45') 1.66 x 10 -21 VESSEL 1/4 THICKNESS (45') 1.92 x 10 -21 VESSEL 3/4 THICKNESS (45*) 3.08 x 10 NOTE: RATIOS ARE IN UNITS OF 2 [ DISPLACEMENTS PER ATOM]/[n/cm ) i 9 i nea.enna 6-38
O' (MAJOR AXIS). I5* (CAPSULES V,T) e 25- (CAPSULES S,Z,X) y / 35' (CAPSULES Y,W) / g 45' (CAPSULE U) I '//// / um], PRESSURE VESSEL gr } y; I / \\ / l l @\\ f f / THERMAL SHIELD / I / I I i [l / / CORE BARREL ~l#/ I BAFFLE ~il/ l 1 / REACTOR CORE llll I/ Figure 6-1 Surry Unit 1 Reactor Geometry 6-39
l T 1 CHARPY [ SPECIMEN / / / / ///// ///////// THERMAL SHIELD i j Figure 6 2-Reactor Vessel Surveillance Capsule 6-40 007-A-1969014
4 10003-2a 1020 l 7 /* d ,/p CIRCUMFERENTIAL WELD, O' 3 s',e / ~ / / / / N 10 19 /p, sc 7 /', LONGITUDINAL WELDS, 45* 5 /s'e Y / 3 / / s M 10 18 ACTUAL 7 .--_. PROJECTED 5 3 LICENSE EXPIRATION END OF l CYCLE 8 60CY 4I I l I l I igi7 t o 10 20 30 40 50 60 70 l-OPERATING TIME (EFPY) Figure 6-3 Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Beltline Weld Locations as e Function of Full Power ~ Operating Time 6-41 -.....~..
i l imso j 1 2 /0 o . Is s o 7 i s 4, N s Og, . g, ,a J s \\ /
- p,,,
s s / ,s',# s ,,s* W ,O g s' # I N /s',s / s s y s 5 4cI(/4 ECpg ~%s%. c J 7 LICENSE ^ END OF CYCLE 8 60CY 10 18 I I I F I I O 10 20 30 40 50 60 70 OPERATING TIME (EFPY) 4 1 Figure 6-4 Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Center of the Surveillance Capsules as e Function of Full Power Operating Time i 6-42
teoos-32 3o20 7 ACTUAL ~'%
PROJECTED 5
ss 's ^ N E o 3 s N s A c \\ \\ \\ \\ w g \\ 14 g y \\g 's' 10 19
- %'* 's's'*-
Jh- -. 60CY s s N. 7 g 3 E 5 LICENSE EXPIRATION g h-3 END OF CYCLE 8 10 18 I l l I I I o 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEG.) Figure 6-5 Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle r l 6-43
10003-34 10 7 5 3 199.39 1.0 204.90 d 7 g etAO IR 5 d t l/4T 3 215.13 P 220.24 r-! O.I g 7 3/4T 45' 5 O' 3 REACTOR VESSEL OR I I I I O.01 195 199 203 207 211 215 219 223 RADIUS (cm) Figure 6-6 Surry Unit 1 Relative Radial Distribution of Fast Neutron (E > 1.0 MeV) Flux and Fluence within the Pressure Vessel Wall 6-44
18003-35 1.0 7 5. 3 g W Wy 0.1 N x 7 D l 5 3 -D 8 ~ w I p O.01 1-J 7 1 -l 5 3 CORE MIDPLANE O.001 -300 -200 -100 O 100 200 300 DISTANCE FROM CORE MIDPLANE (cm) Figure 6-7 Surry Unit 1 Relative Axial Variation of Fast Neutron l* (E > 1.0 MeV) Flux and Fluence within the Pressure l Vessel Wall l ~ 6-45 i
SECTION 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE The following withdrawal schedule per ASTN E185-82 is recommended for future capsules to be removed from the Surry Unit I reactor vessel. Vessel Estimated Location Withdrawal Fluence I 2 Capsule (deg) Time 'l (n/cm ) 18(b) T 285* 1.07 (removed) 2.81 x 10 18(b) W 55' 3.39 (removed) 3.97 x 10 19(b] V 165* 8.02 (removed) 1.94 x 10 19 X 65* 21.2 2.78 x 10 19ECl Z 245' 28.8 3.78 x 10 S 295' standby (d) Y 305' standby (d) U 45' standby (d) a. Effective full power years from plant startup b. Actual Fluence c. Approximate maximum neutron fluence on vessel inner wall at-32 EFPY. d During 20 year inservice inspection capsules shou'd possibly be transferred to higher flux capsule location. The surveillance weld represents the Surry Unit i reactor vessel lower shell vertical weld seam (L2) which is located at 45' azimuthal position where the maximum neutron fluence after long operating times (30 to 40 EFPY) will be 18 2 approximately 6 to 8 x 10 n/cm. The data obtained for capsule V there-fore represents a fluence for seam (L2) which will be well beyond any fluence experienced by seam (L2). Since the surveillance weld is typical of other welds in the beltline region because it was fabricated with Linde 80 flux, data from this weld to be obtained at 21.2 and 28.8 EFPY will be useful in further assessing this type of weld for normal operating life and plant life extension. noo *nne 71 4
i It should be noted that the Point Beach Unit 1 vessel surveillance program contains a surveillance weld which was fabricated with the same heat of weld wire (72445) and type of flux (Linde 80) as the Surry Unit 1 intermediate to lower shell girth weld. Data from this weld [20] shows a ART f 165 ND 19 to IS0*F at a neutron fluence of approximately 2.1 x 10 n/cm and an upper shelf energy of approximately 55 ft-lbs. It is recommended that future weld test results from the Point Beach Unit 1 surveillance program be considered when evaluating the Surry Unit 1 reactor vessel at later times in
- 1ife, i
e i ) 4 i 2200s-04 tM F 10
SECTION 8 REFERENCES C 1. Yanichko, S. E., " Virginia Electric and Power Company Surry Unit No.1 Reactor Vessel Radiation Surveillance Program" WCAP-7723, July 1971. 2. ASTM E185-73 " Practice for Surveillance Tasts for Nuclear Reactor Vessels" in ASTM Standards, Part 10 (1973), American Society for Testing and Materials, Philadelphia, Pa., 1973. 3. Perrin, J. S., etal, "Surry Unit No.1 Pressure Vessel Irradiation Capsule Program: Examination and Analysis of Capsule T," Battelle Columbus Laboratories, June 24, 1975. 4. Perrin, J. S., etal, "Surry Unit No. 1 Nuclear Plant Reactor Pressure Vessel Surveillance Program: Examination and Analysis of Capsule W," Battelle Columbus Laboratories Report BCL-585-8R, March 30, 1979, a 5. Lowe, A. L. " Reactor Pressure Vessel and Surveillance Program Materials Licensing Information for Surry Units 1 and 2," Babcock and Wilcox Report BAW-1909, March, 1986. 6. Proposed Revision 2 to Regulatory Guide 1.99 " Radiation Damage to Reactor Vessel Materials,", U.S. Nuclear Regulatory Commission, February, 1986. 7. Soltesz, R. G., Disney, R. K., Jedruch, J., and Zeigler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and input Data l Preparation. Vol. 5 - Two-Dimensional Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970. 8. SAILOR RSIC Data Library Collection "0LC-76, Coupled, Self-shielded, 47 Neutron, 20 Gamma-ray, P3, Cross Section Library for Light Water Reactors." noo.* nn o 8-1
9. Furchi, E. L., Perone, V. A., Weaver, M., and Wrights G. N., "Surry Units 1 and 2 Reactor Vessel Fluence and RT Evaluations," WCAP-11015, PTS December 1985.
- 10. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -
to be published.
- 11. ASTM Designation E261-77, Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 76-87, American Society for Testing and Materials, Philadelphia, Pa., 1983.
- 12. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 88-96, American Society for Testing and Materials, Philadelphia, Pa., 1983,
- 13. ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of. Iron," in ASTM Standards (1983),
Section 12, Nuclear Standards, pp. 97-102, American Society for Testing - and Materials, Philadelphia, Pa.,1983.
- 14. ASTM Designation 481-78, " Standard Method of Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 228-235, American Society for Testing and Materials, Philadelphia, Pa., 1983.
- 15. ASTM Designation E264-82, " Standard Method for Determining fast-Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards (1983),
Section 12, Nuclear Standards, pp. 103-107, American Society for Testing and Materials, Philadelphia, Pa.,1983. 4
- 16. Nucleonics Week, October 1972 and November 1972 issues.
1
- 17. Docket 50280-160, "Surry Power Station Units 1 and 2 Semi-Annual Operating Report, SOR-2, for the period January 1,1973 - June 30,1973" (Virginia Electric and Power Company, Richmond).
noe,* u n e 8-2
- 18. " Licensed Operating Reactors Status Summary Report," NUREG 0020, May 1974 through June 1986.
O
- 19. Private communication from Vivian Jones of Virginia Electric and Power Company to V. A. Perone of Westinghouse (December 18,1986).
- 20. Yanichko, S. E., etal, " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP-10736, Decembe-1984.
s O a O o noo.+inr io 8-3 _.}}