ML20041E973

Integrated Reactor Vessel Matl Surveillance Program

ML20041E973
Person / Time
Site:	Midland
Issue date:	11/30/1981
From:	Aadland J, Lowe A, Moore K BABCOCK & WILCOX CO.
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ML20041E971	List: ML20041E970 ML20041E973
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BAW-1543, NUDOCS 8203160036
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( I BAW-1543 November 1981 E I. I. i INTEGRATED REACf0R VESSEL MATERIAL SURVEILLANCE PROGRAM ,I by A. L. Lowe, Jr., PE K. E. Moore J. D. Aadland 1 I i i I I I BABCOCK & WILCOX I Nuclear Power Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox I

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l1 Il s SIMIARY This report describes the integrated reactor vessel surveillance program - an innovative approach to monitoring the irradiation-induced material changes of E the steels and weldments routinely used in r2 actor vessels. Alterations to 5 the material properties of reactor vessel raterials - tensile strength, Charpy energy level, and fracture toughness - are characterized by irradiating ap-propriate test specimens in operating reactors. Federal regulations require that all operating nuclear reactors have surveil-lance programs that involve preparation, irradiation, scheduled retrieval, and subsequent testing and evaluation of irradiated specimens. The integrated re-actor vessel surveillance program not only complies with these requirements but also enhances the data acquired. The latter is accomplished by making I data-sharing possible among the nine participating power plants as we'll 3 as acquiring the fracture toughness data necessary to ensure the continued licenseability of the various reactors. Specifically, the integrated reactor vessel surveillance program, initiated in 1976, assesses data from two separate but interrelated projects: (1) the plant-specific surveillance program integrates the various plant-specific sur-veillance programs to ensure the availability of data on a timely basis and which meets the basic requirement that each reactor have a surveillance pro-gram, and (2) the power reactor program, which will provide the fracture E toughness properties of eight veld metals, which will complement the data 5 obtained from the plant-specific. capsules. The first program separates the participating power plants into two classes - those from which specimens come (guests) and those in which the specimen irradiations are performed (hosts). The nine power reactors involved (six guests, three hosts) are similar in both design and operating conditions. Specimens are enclosed in two different types of specially designed cylindrical capsules - normal plant-specific capsules and the larger research capsules. 3 I Babcock & Wilcox I

( I. I A brief description of the federal guidelines and legislation and the combina-tion of events that stimulated Babcock & Wilcox's development of the integrated reactor vessel surveillance program are discussed. The overall program is de-I- scribed and is divided into plant-specific reactor vessel surveillance programs and the power reactor research capsule program. Detailed descriptions of the types and properties of materials being investigated, the types of capsules used to contain the surveillance specimens, and the program and capsule desig-nations are given in the appendixes. II ii I,. i E I-1 I I I I I - 111 - Babcock & Wilcox

.m_ ~ I Babcock & Wilcox Nuclear Power Generation Division Lynchburg, Virginia Report BAW-1543 November 1981 Integrated Reactor Vessel Material Surveillance Program A. L. Love, Jr., K. E. Moore, J. D. Aadland Key Words: Reactor Vessel Material Surveillance, Weldments. Base Metal. Tensile Strength, Charpy Energy, Fracture Toughness, Postirradiation Examination. Plant-Specific Capsules, Research Capsules ABSTRACT An integrated reactor vessel material surveillance program was designed when the surveillance capsule holder tubes in a number of reactors were damaged and could not be repaired without a complex and expensive repair program and considerable radiation exposure to personnel. Ths integrated program is fea-sible because of the similarity of the design and operating characteristics of the affected plants. Three plants were selected fo the role of irradia-E tion sites (host reactors), and thn capsules of the 'other six plants (guest 5 reactors) were irradiated on an integrated irradiation schedule with the cap-sules of the host reactors. The program consists of two parts - the first is the plant-specific program, which is the continued irradiation of the surveil-lance capsules removed from those reactors in which the capsule holder tubes were damaged along with the capsules from the host reactors; the second is made up of a number of special research capsules designed to provide fracture I toughness data on a series of weld metals predicted to exhibit high sensitivity 5 to irradiation damage. I A I - iv - Babcock 8 Wilcox

f I. f I.- I.. CONTENTS Page 1-1 1. INTRODUCIION........................... ii 1.1. General 1-1 1-4 1.2. Objective 2-1 { 2. BACKGROUND............................ 3. INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM.......... 3-1 { 3.1. Plant-Specified Surveillance Program... 3-2 3.1.1. Structural, Hydraulic, and Thermal I Characteristics of Specimen Holder Tube and Capsule. 3-4 I 3.1.2. Dosimeters..................... 3-5 E, 3.1.3. Temperature Monitors... 3-6 B' 3.1.4. Types of Surveillance Programs and 3-6 l Capsules.. 3.2. Research Capsule Prograr. 3-8 I. 3-8 3.2.1. Introduction.................... 3.2.2. Research Capsule Design 3-9 Holder Tube.and Capsule Location.. 3-11 3.2.3. 3.2.4. Structural, Hydraulic, and Thermal Characterization of Research Capsules 3-11 3-11 1. 3.2.5. Dosimetry 3.2.6. Insertion and Withdrawal Schedules.. 3-13 I' 3.2.7. Unirradiated Control Data 3-13 3.3. Irradiation Schedule... 3-13 APPENDIXES A. Description and Properties of RVSP Materials....... A-1 B. Description and Properties of Research [ Capsule Program Materials B-1 3 C. Description of Surveillance Capsule Test Specimens - Plant-Specific and Research Capsules. C-1 D. Program and Capstile Type Designations D-1 E. References E-1 I Babcock s.Wilcox -v-I

T I, List of Tables Table Page 1-1. Comparison of Plant Parameters for Host Reactor Davis-Besse 1 and Guest Reactors Rancho Seco and Arkansas Nuclear One-1 1-5 1-2. Comparison of Plant Parameters for Host Reactor Crystal River 3 and Guest Reactors Oconee Units 1, 2, and 3.......................... 1-6 g 1-3. Comparison of Plant Parameters for Host Reactor 3 Three Mile Island 2 and Guest Reactor Three Mile Island 1......................... 1-7 2-1. Significant Differences Between Revisions of ASTM E-185.... 2-3 3-1. Typical Withdrawal Schedule for Six-Capsule Program. 3-15 3-2. Surveillance Capsule Dosimeters. 3-15 3-3. . Temperature Monitor Wires. 3-16 5 3-4. Reactor Vessel Surveillance Program -- Detailed Summary 3-17 3i 3-5. Research Capsules -- Material and Specimens per Capsule 3-18 3-6. Contents of Research Capsules. 3-19 g 3-7. Azimuthal Location of Power Reactor Research Capsules.. 3-19 E 3-8. Neutron-Detecting Elements 3-20 3-9. Matrix of Control Specimens for Welds W2, W4, W6, W8 Under Power Reactor Program. 3-20 3-10. Identification of Programs and Control Specimens for Eight Research Capsule Program Welds 3-21 3-11. Surveillance and Research Capsule Insertion and Withdrawal Scheduled for Crystal River Unit 3......... 3-22 3-12. Surveillance and Research Capsule Insertion and Withdrawal Schedule for Davis-Besse Unit 1 3-24 3-13. Surveillance and Research Capsule Insertion and g Withdrawal Schedule for Three Mile Island, Unit 2....... 3-26 g A-1. Oconee 1 Description and Properties of Reactor Vessel Surveillance Program Materials.. A-2 A-2. Oconee 2 Description and Properties of Reactor Vessel Surveillance Program Materials. A-3 A-3. Oconee 3 Description and Properties of Reactor Vessel Surveillance Program Materials.. A-4 A-4. TMI-1 Description and Properties of Reactor Vessel Surveillance Program Materials.. A-5 A-5. TMI-2 Description and Properties of Reactor Vessel Surveillance Progran Materials. A-6 A-6. ANO-1 Description and Properties of Reactor = Vessel Surveillance Program Materials.. A-7 A-7. Crystal River Unit 3 Description and Properties of g Reactor Vessel Survsillance Program Materials. A-8 g A-8. Rancho Seco Description and Properties of Reactor Vessel Surveillance Program Materials.. A-9 A-9. Davis-Besse 1 Description and Properties of Reactor l Vessel Surveillance Program Materials.. A-10 5 B-1. Chemical Composition and Unirradiated Mechanical Properties of Beltline Region Weld Metals. B-2 3' B-2. Description of Beltline Region and Surveillance g Weld Metals. B-3 - vi - Babcock s.Wilcox I

F 1 1 Tables (Cont'd) I. Table Page C-1. Dimensions of Component Fracture Specimens C-4 D-1. RVSP Capsule Types D-2 D-2. Materials and Specimens in Surveillance Capsules of Oconee Unit 1 D-4 D-3. Materials and Specimens in Surveillance Capsules of Oconee Unit 2 D-5

1 D-4.

Materials dnd Specimens in Surveillance Capsules of Oconee Unit 3 D-6 D-5. Materials and Specimens in Surveillance Capsules } of Three Mile Island Unit 1.................. D-7 I D-6. Materials and Specimens in Surveillance Capsules of Three Mile Island Unit 2.................. D-8 l D-7. Materials and Specimens in Surveillance Capsules of Crystal River 3 D-9 I D-8. Materials and Specimens in Surveillance Capsules of Arkansas Nuclear One, Unit 1................ D-10 D-9. Materials and Specimens in Surveillance Capsules I of Rancho Seco Unit 1..................... D-ll D-10. Materials and 3pecimens in Surveillance Capsulen {. of Davis Besse Unit 1..................... D-12 1 ! E' l 5 List of Figures I Figure I 1-1. Reactor Vessel Arrangement Showing Surveillance {. Capsule Holder Tube Location 1-8 1-2. Reactor Vessel Cross Section Showing Surveillance I' Capsule Locations at Crystal River Unit 3........... 1-9 3-1. Surveillance Capsule Arrangement 3-28 3-2. Surveillance Capsule Arrangement - With Compact I. Tension Specimens............ 3-29 3-3. Research Capsule, Type R-1 3-30 L. 3-4. Research Capsule, Type R-2 3-31 lg 3-5. Surveillance Capsule Holder Tube Location and l E 'd*"'i'icati " ' ' 'vi" "** ""i' 3-32 l 3-6. Surveillance Capsule Holder Tube Location and Identification for Crystal River Unit 3............ 3-33 I 3-7. Surveillance Capsule Holder Tube Location and Identification for Three Mile Island Unit 2.......... 3-34 i 3-8, 177-FA Reactor Vessel Through-Thickness Temperature Distribution at Steady-State Normal Operation.. 3-35 C-1. Standard Size Tensile Specimen - Used on Type C-5 A Capsules C-2. Miniature Size Tensile Specimens - Used on C-6 Type B Capsules. - vii Babcock SiWilcox I

I Figures (Cont'd) Figure Page C-3. Charpy V-Notch Specimen. C-7 C-4. Rectangular Compact Fracture Specimen - Standard Proportions and Modification for Measurement of Displacement at Load Line. C-8 C-5. Round Compact Fracture Specimen - Dimensions and Modification for Measurement of Displacement at Load Line C-9 C-6. Geometry of Side Grooves for 0.936 TRCT.. C-10 I I I I 1 I I I I l I-I I - viii - Babcock s.Wilcox

'i I 1. INTRODUCTION r 1.1. General l' The integrated reactor essel surveillance program (RVSP) is the result of t 4 two events: failure of the capsule holder tubes in operating plants, and the necessity to obtain fracture toughness data for irradiated weld metals to l ensure the continued licenseability 'of operating plants. l The original design of the B&W 17 /-fuel assembly (177-FA) class reactors in-cluded three reactor vessel surveillance capsule holder tubes (SCHTs) located near the reactor vessel inside wall, as shown in Figures 1-1 and 1-2. Each of the tubes was designed to hold two capsules containing reactor vessel sur-veillance specimens. In 1976, the SCHTs in a number of the 177-FA reactor vessels were found to be damaged. Subsequently, all operating 177-FA plants were shut down for inspection of the holder tubes. This inspection revealed that all of the SCHTs had been damaged to some extent. To prevent further damage and to eliminate the possibility of overall system damage if parts of the holder tube were dislodged, all the surveillance capsules and holder tubes I-that had either failed or were of the same design were removed from the vessels. Plants involved were Oconee Units 1, 2, and 3; Arkansas Nuclear One, Unit 1; Rancho Seco; and Three Mile Island Unit 1. During the same period another event occurred that led to the need to improve t. the kind of data that were being obtained from the existing reactor vessel surveillance programs of operating 177-FA plants. It was found that certain = weld metals used in the fabrication of the early-generation 177-FA resctor vessels may not meet current design requirements because of initial properties and unusual sensitivity to neutron embrittlement. This problem was further complicated by the fact that the capsules in the affected plants did not con-tain appropriate specimens, and, in some cases, the proper veld metals, to permit the required analyses of the vessels.

I l The integrated RVSP was developed in response to the immediate problems of the capsule-holder failures and the longer-range requirement of revamping existing RVSPs to improve the quality and quantity of fracture toughness data. In this cooperative data-sharing system, surveillance capsules re-moved from vessels with damaged tubes are placed in similar reactors for ir-radia tion. The plants with the damaged SCHTs are called " guest reactors;" .l those in which the irradiations are to be done are the " host reactors." The following pairings of capsules and reactors were agreed upon by the af fected reactor owners. Guest reactors Owners Host reactors Oconee 1, 2, and 3 Duke / Florida Power Crystal River Unit 3 Arkansas Nuclear One, Unit 1 AP&L/ Toledo Edison Davis-Besse Unit 1 g E Rancho Seco SMUD Three Mile Island Unit 1 Met Ed/ Met Ed Three Mile Island Unit 2 I The rationale motivating the implementation of this unique program is dis-cussed in the ensuing paragraphs of this section. Damage to the original SCHTs precipitated the design, manufacture, and test-ing of improved tubes. SCHTs of this improved, NRC-approved design were in-stalled in Davis-Besse 1. Crystal River 3, and Three Mile Island 2 prior to their initial startup, i.e., before neutron activation of the reactor inter-nals. However, it.atalling the redesigned SCHTs in already-irradiated B&W reactors presented substantial difficulties, primarily because precision machining, alignment, and inspection had to be performed remotely and undar water. These circumstances could have caused significant radiation exposure - up to about 100 man-rem per reactor - to plant personnel. Therefore, an alternative pro-gram that did not involve reinstalling SCHTs in irradiated plants was proposed. Since Crystal River 3. Davis-Besse 1, and Three Mile Island 2 had the same reactor design as those six reactors from which damaged holder tubes were re-I moved and were scheduled for startup in an appropriate time-frame, it was as cost-effective and technically acceptable to use the three new plants as hosts for the capsule specimens of the other plants. The exchange plan presented g previously was devised and the integrated RVSP was implemented. 1-2 Babcock s.Wilcox I

I The most important aspect controlling the success of the integrated program is the consnonality of the irradiation sites and reactors involved. In the 1 l case of the B&W 177 FA plants participating in this integrated program, there are no dissimilarities among the plants except for the materials of construc-tion. This is a natural dissimilarity because heats of steel are not large enough to make more than one reactor vessel; however, the various vessels are made of the same type and grade of material, which is the primary considera-tion. The design characteristics of various host reactors and their guest units are compared in Tables 1-1, 1-2, and 1-3. 1 Other reactor parameters that are significant in evaluating the similarity of the host and guest reactors are (1) the relative neutron flux energy spectrum, I (2) the irradiation dose rate, and (3) the irradiation temperature. The rela-tive neutron energy spectrum is a function of the geometry and materials of the reactor internals components. As shown in the tables, the dimensions and ma-terials of the host and guest reactors are essentially the same. Thus, no j difference in the relative neutron energy, spectra is expected. Similarly, dif-ferences in irradiation dose rates between the guest and host reactors would be due only to variations in power levels..Since the licensed power levels are comparable, variations in the irradiation rate are attributable to plant maneu-I vering. Averaged over time, the variations in power level due to those maneu-vers will have no significant effect, as confirmed by surveillance results from a number of plants. The reactor vessel beltline region and normal surveillance specimens are ex-posed to reactor coolant inlet conditions when being irradiated in the host reactors. Two factors that could contribute to differences in the irradiation environments of the capsules are design variations among the plants and power level changes due to maneuvering. The variations due to design differences between the host and guest reactors I are insignificant, as shown in Tables 1-1, 1-2, and 1-3. Between partial-(N15%) and full-load conditions, the inlet temperature will vary by about 20F as an inverse function of power level. The duration of this variation due to maneuvering is comparable among plants over time; this is supported by the com-parability of available reactor vessel surveillance results. In any case, the inlet condition temperatures are considered too low to cause significant self-annealing. The inlet temperature will also vary about 40F between hot zero 1-3 Babcock s.Wilcox

1 power and partial load conditions. This variation is a direct function of power level (0 to 15%) and again is not significant because of the low temper-ature and the expected comparability in duration ever the long term. l 1.2. Objective The integrated RVSP is designed to provide the fracture toughness data for the materials in the 177-FA reactor pressure vessels necessary to ensure the con-tinued licenseability of the plants. As originally designed, the individual reactor surveillance programs for the plants with failed holder tubes did not provide the fracture toughness data needed to perform the currently required analytical evaluations of reactor ves-sel integrity. Although the original RVSPs were designed and fabricated in complete accordance with the existing standards and regulations, these changed with the development of fracture mechanics techniques as a method of ensuring reactor system integrity. Consequently, changes developed in the required ma-E terials data and the type of specimens necessary to obtain the data. Although 5 the integrated program eliminates the need to replace the failed specimen holder tubes on operating plants and thus eliminates the high radiation expo-sure, it does provide a systematic and redundant program to develop the new materials data needed to assess the future integrity of this grc.up of reactor plants. I l I I I I I 1-4 Babcock a.Wilcox

Table 1-1. Comparison of Plant Parameters for Host Reactor Davis-Besse 1 and Guest Reactors Rancho Seco and Arkansas Nuclear One-1 P L. uest reactors Host reactor: Plant Parametera Davis-Besse 1 Rancho Seco ANO-1 Design heat output (core), MWt 2772 2772 2568 Design overpower, % des power 112 112 112 System pressure (nom), psia 2200 2200 2200 I Coolant flow rate, 10-8 lbm/gpm 131.3/352,000 137.8/369,000 131.3/352,000 i Coolant temperatures, F Nominal inlet 555 557 554 I Avg rise in vessel 53 51 48 l Avg in vessel 582 582 579 No. of fuel assemblies 177 177 177 Type of fuel assemblies Mark B (15x15) Mark B (15x15) Mark B (15x15) Core barrel ID/0D, in. 141/145 141/145 141/145 Thermal shield ID/0D, in. 147/151 147/151 147/151 i Core structural characteristics ~ Core equiv diam, in. 128.9 128.9 128.9 Core active fuel height, in. 144 144 144 Reflector thickness, compos'n Top (water + steel), in. 12 12 12 Bottom (water + steel), in. 12 12 12 Side (water + steel), in. 18 18 18 I 9.sctor vessel design parameters Principal material SA508 C1.2 SA533 GrB C1.1 SA533 GrB C1.1 Design pressure, psig 2500 2500 2500 Design temperature F 650 650 650 Shell ID, in. 171 171 171 Overall vessel-closure head 40/8.75 40/8.75 40/8.75 height (a), ft/in. I Core barrel-thermal shield Type 304 SS Type 304 SS Type 304 SS principal material (a)0ver cladding and instrumentation nozzles. 1-5 Babcock & Wilcox

I' Table 1-2. Comparison of Plant Parameters for Host Reactor Crystal River 3 and Guest Reactors Oconee Units 1, 2. and 3 JI Host reactor: Plant parameters Crystal River 3 Oconee 1 Oconee 2 Oconee 3 Design heat output (core). MWt 2452 2568 2568 1568 Design overpower. I des power 114 112 112 112 System pressure (nominal), psia 2200 2200 2200 2200 Coolant flow rate. 10 lb/h/ gym 131.3/352.000 131.3/352.000 131.3/352.000 131.3/352.000 Coolant temperatures. F Nominal inlet 555 554 554 554 Avg rise in vessel 48 50 50 50 3 Avg in vessel 579 579 579 379 No. of fuel assemblies 177 177 177 177 Type of fuel assemblies Mark B (15=15) Mark B (15=15) Mark 7 (15=15) Mark B (15=15) Core barrel ID/0D. in. 141/145 141/145 141/1 141/145 Ther a1 shield ID/0D. in. 147/151 147/151 147/' 147/151 Core structural characteristics Core equiv diameter, in. 128.9 128.9 128.9 128.9 Core active fuel height, in. 144 144 144 144 Reflector thickness, compos'n Top (water + steel). in. 12 12 12 12 Bottom (water + steel), in. 12 12 12 12 Side (water + steel), in. 18 18 18 18 Reactor vessel design parameters SA302 CrB C1.1,) SA508 C1.2 SA508 C1.2 g Principal material SA533 GrB C1.1 Design pressure. pois 2500 2500 2500 2500 Design temperature. F 650 650 650 650 Shell OD in. 171.375 171 171 171 00 across nozzles, in. 249 249 249 249 overall vessel-closurerhead 40/8.875 40/8.75 40/8.75 40/8.75 height, ft/in.(b) l Core barrel-thermal shield Type 304 SS Type 304 SS Type 304 SS Type 304 SS l principal material (* As modified by Code Case 1339. (b)0ver cladding and instrumentation nozzles. I I I I I 1-6 Babcock s.Wilcox

I I Table 1-3. Comparison of Plant Parameters for Host Reactor Three Mile Island 2 and Guest Reactor Three Mile Island 1 i l Host reactor: Guest reactor: 1 Plant parameters TMI-2 TMI-1 Design heat output (core), MWe 2772 2568 Design overpower, % design power 112 112 System pressure (nominal), psia 2200 2200 Coolant flow rate,10-' 1bm/gpm 137.8/369,000 131.3/352,000 i Coolant temperatures, F Nominal inlet 557 554 I Avg rise in vessel 51 50 Avg in vessel 582 579 i No. of fuel assemblies 177 177 Type of fuel assemblies Mark B (15x15). Mark B (15x15) Core barrel ID/0D, in. 141/145 141/145 Thermal shield ID/0D, in. 147/151 147/151 Core structural characteristics, in. Cors equivalent diameter 128.9 128.9 Core active fuel height 144 144 Reflector thickness, composition, in. Top (water + steel) 12 12 I Bottam (water + steel) 12 12 Side (water + steel) 18 18 -Reactor vessel design parameters I, Principal material SA533, Gr B, SA302, Gr B C1.1 modified Design pressure, psig 2500 2500 Design temperature, F 650 650 I Shell ID, in. 171 171 OD across nozzles, in. 249 249 Overall vessel-closure head 40/8.75 4c/8.75 height (a), ft/in. Core barrel-thermal shield Type 304 SS Type 304 SS principal material (*}0ver cladding and instrumentation nozzles. I I l 1-7 Babcock a.Wilcox

Figure 1-1. Reactor Vessel Arrangement Showing Original Surveillance Capsule Holder Tube Location It f n=hY==$ $ ?h. l =*% fe i Holddown f l f Spring Unit-t'a s j.p g'g ,l p r 4 ~ Thermal Aging g -F gb Core Support Capsule Location 's Shield g l (Oconee 1 and TMl 1 Only) Tj { i s N / d s N 0 \\ / z ?-Reactor Vessel f e, s( l Flux Capsule {Q \\ y / ,,9 Location-s g g ( Core Midplane ( g . g Surveillance / g Thermal Shield \\ 4 Specimen & lo-/ f I.. !sV d N Holder Tube -.A Q ' m N /. pd31432 U 2 4 3 sI' wA l - c s %,i x I I l-8 Babcock a. Wilcox I

I I I Figure 1-2. Reactor Vessel Cross Section Showing Original Surveillance Capsule Locations X SCHTs - Capsules ANI-C. ANI-D f I 0 l ~~~ i i i, N N / / 0 0 e j e. e s I e v,l e e g e e e e e e o e e 1 r e e e l e e e w-L.- g ._ g g__. g .y e e e \\ g' e e e e e I/ i s e e i e e / \\ 8 I N / e e /% I / / D ~ / l SCHTs - Capsules ANI-A, ANI-B l SCHTs - Capsules j ANI-E. ANI-F I I I 1-9 Babcock 8.Wilcox

I I l 2. BACKCROUND I .It became apparent in the late 1950's that the neutrcn embrittlement sensi-tivity of steels and weldments used in reactor vessels vary significantly from steel to steel, heat to heat, and even weld to weld. Accordingly, a research and development program to address this phenomenon was initiated and in 1961, guidelines (ASTM E-185) for establishing a reactor vessel surveillance program (RVSP) were adopted " Standard Practice for Conducting Surveillance Tests for I, Light Water-Cooled Nuclear Power Reactor Vessels."* Both comercial power plants and test reactors # in several national laboratories were used in deter-mining the metallurgical parameters controlling the sensitivity of comonly used reactor vessel. steels to neutron irradiation." Selected specimens of these steels were encased in specially designed capsules. The encapsulated samples were then placed in test reactors where the neutron flux and tempera-ture experienced were comparable to those seen by the reactor vessel wall. I In this way, embrittlement (characteristics) could be assessed and predicted before it occurred in operating reactor vessels. In 1970, the copper.ad phosphorus centents in pressure vessel materials were identified by the Naval Research Laboratory in cooperation with The Babcock & Wilcox Company (B&W) as the principal parameter (s) influencing neutron embrit-tiement sensitivity. M However, despite research by these and other organiza-tions, the ef fects o' dif fering neutron spectra, radiation rates, and tempera-ture were not clearly delineated. Also, information on particular types of materials, such as submerged-arc weld metal and SA508 Class 2 forgings, was W quite limited. I This document was revised in 1966,1970,1973, and 1979 to reflect knowledge gained. These revisions are compared in Table 2-1. Specimens were irradiated at flux rates 100 to 200 times higher than those observed at the reactor vessel wall of a commercial power plant, and at nominal reactor vessel temperatures below 550F. 2-1 Babcock & Wilcox

In 1973, in a concerted effort to acquire the necessary information for licens-ing PWRs and to standardize the existing industry RVSPs, Appendix H of 10 CFR 50 (" Reactor Vessel Materials Surveillance Program Requirements") made the RVSP (complying with ASTM E-185-73) mandatory. Up to this point, the data gathered from the RVSPs had been diversified because the ASTM E-185 require-ments were broadly defined and gave considerable latitude to the designer. I I I I I I I 2-2 Babcock & Wilcox I

.E -E.. E._ ..E_ _E. . E E M g l ~ Table 2-le Significant Dif ferences Between Revisions of AS*1M E-185 re or--e am -iS not-tal. esea-w b, no. of me. of wiseesi s...f b.eeltee spec i-medes toep f-Co

i. enh-ese-cr asener s,-ul,e,eusenere esgeolee espeele/meternal

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7) feet egeteneet O

shell be celt-g bested y V I =t') OM

l l. l 3. INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM B. The integrated surveillance program is more than a combination of nine separate projects and the resultant sharing of irradiation sites. It addresses both the short-and long-term requirements for acquiring irradiation data and the need j to improve the quality and quantity of fracture toughness data to support the continued licenseability of the participating reactor pressure vessels. i The integrated reactor vessel surveillance program correlates data from both plant-specific and test reactor surveillance monitoring. However, since the test reactor irradiations are not performed in B&W 177-fuel assembly operating reactors, the following discussions are limited to the power reactor program, i which comprises two principal parts. The first is the continuation of the plant-specific surveillance programs which monitor the irradiation damage to selected materials, as originally planned. The capsules contain samples of plate or forging material and heat-affected-zone (HAZ) material from the ves-sel beltline as well as weld metal; thus, this part of the program will con-tinue to monitor the long-term effects of neutron irradiation on the reactor materials and will be the basis for the plant-specific fluence analysis, i The second part of the program consists of a series of specially designed cap-sules, (research capsule program) to study the effects of irradiation on a number of weld metals, which are anticipated to be highly sensitive to irradia-tion damage because of their chemistry and low initial Charpy upper shelf energies. These test capsules contain specimens primarily for obtaining frac-ture toughness properties of individual weld metals and are located in high-lead-factor irradiation holder tubes so that the needed data will be provided in a relatively short time. The data from these capsules will be compared t with data obtained on the same material be various test reactor research pro-grams. This comparison will permit evaluation of the effects of flux density and neutron energy spectrum on the irradiation damage to these materials. The integrated reactor vessel surveillance program (RVSP) complies with ASTM E185-73 and also addresses the additional requirements of Appendix H, which 3-1 Babcock a.Wilcox

affect the individual RVSPs of the plant involved. These additional changes are described below. 1. The definition of the beltline region was changed to include more of the shell course material above and below the effective height of the fuel element assemblies. 2. The transition temperature adj ust. ment (ARTg) for beltline region materials is based on the temperature shif t of the Charpy V-notch curve measured at the 50 ft-lb level or at the 35-mil lateral expan-sion. The property exhibiting the greater shift is used to define the adjustment in reference temperature. 3. The pressure-temperature operational limitations of the reactor vessel are established in accordance with Appendix G to Section III of the ASME Code. The highest adjusted RT and the lowest upper g shelf energy level of all the beltline region materials, as deter-mined by the RVSP, are used to calculate these operating limitations. 4. Since the test materials in the capsules of the plant-specific RVSP were not selected in accordance with ASTM E 185-73, the data to be generated are not necessarily app'licable to any specific plant. That is, the materials monitored by the RVSP are usually not the ma-terials judged by Appendix H to be most likely to be the controlling beltline region materials with regard to radiation embrittlement for the reactor vessel for which the RVSP was designed. Consequently, the applicability of the data to be generated by the plant-specific RVSP becomes limited; however, by combining the data from all the RVSPs, it is practical to develop a data base from which to deter-mine the probable values and predict the irradiation behavior of those welds for which there are no specific data. The two parts of the power reactor program are discussed separately in sec-tions 3.1 and 3.2. 3.1. Plant-Specified Surveillance Program l The plant-specific surveillance program includes the' irradiation of (1) the sur-veillance capsules removed from reactors without capsule holder tubes and (2) the capsules from those plants in which the irradiations are being conducted. 3-2 Babcock 3.Wilcox l w ~ = r

Each plant participating in the integrated surveillance program has a plant-specific surveillance program which wa's designed to meet the requirements of the NRC and ASTM E-185 at the time the programs were developed. The topical l reports describing the appropriate RVSPs for B&W's 177-fuel assembly (177-FA) units participating in the integrated surveillance program are as follows: Nuclear Plant

• Applicable Topical Report Oconee Unit 1 BAW-10006A, Rev. 3 Oconee Unit 2 BAW-10006A, Rev. 3 Oconee Unit 3 BAW-10006A, Rev. 3 Three Mile Island Unit 1 BAW-10006A, Rev. 3 l

Three Mile Island Unit 2 BAW-10100A Crystal River Unit 3 BAW-10100A Arkansas Nuclear One, Unit 1 BAR-10006A, Rev. 3 Rancho Seco BAW-10100A Davis Besse Unit 1 BAR-10100A I E The types and properties of the RVSP materials for each plant are described in Appendix A. Each RVSP consists of three holder tubes, each of which contains two capsules. Four of these six capsules are the prime data-collecting capsules, while the other two are considered " standby" capsules. The prime capsules are withdrawn i at designated time intervals so that the data collected correspond to irradia-tion levels ranging from a low level to that equal to the vessel inner surface at end of life (EOL). The standby capsules provide any necessary additional data late in the operating life of the plant. Table 3-1 is a typical with-drawal schedule. t Each capsule is a stainless steel cylinder approximately 2.4 feet long, 2.5 inches in outside diameter, and 2.0 inches in inside diameter. Three basic types of specimens, either singly or in combination, are placed in these cap-sules: Charpy, tensile, and compact fracture. (Appendix B describes the speci-I mens in detail.) The Charpy V-notch specimen is 0.394 inch square, 2.165 inches long, and conforms to ASTM E23-72. The tensile specimen is 4.25 inches long and conforms to ASTM E8-69T. The compact fracture specimen is 0.5 inch thick, 1.25 by 1.20 inches, and conforms to the basic requirements of ASTN E-399. 3-3 Babcock &)Milcox m e m

Specfmen identity is maintained throughout the program by die-stamping the top and bottom of each specimen with an identitifcation code (a combination of six letters and numbers). In addition to the specimens, each capsule contains dosimeters and temperature monitors. Figure 3-1 and 3-2 show typical capsules and the orientation of their specimens, dosimeters, and temperature monitors. The voids in the cap-sule are filled with aluminum spacers to minimize movement of the specimens [ inside and improve heat transfer characteristics. After the specimens.and filler blocks are positioned within the capsule, it is purged with an inert gas before the last end cap is welded in place. As stated previously, the integrated RVSP organizes and evaluates the data ac-cumulated in the individual surveillance programs designed for nine different reactors. Within this common network are three types of surveillance prograns (types A B, and C), in which six capsule types (I-VI) are irradiated. Sur-veillance program A uses capsule types I and II; program B uses capsule types III and IV; and program C uses types V and VI. The physical characteristics of the specimen holder tube and the capsule are described in section 3.1.1, while the dosimeters and temperature monitors are discussed in sections 3.1.2 and 3.1.3. The three separate programs (A, B, and C) and the types of capsules (I-VI) are described in section 3.1.4. 3.1.1. Structural. Hydraulic and Thermal Characteristics of Specimen Holder Tubc 2nd Capsule The thermal characteristics of the specimen holder tube and the czpt.J2 were I analyzed to obtain a design in which the temperature of the specimens is ap-proximately equal to that of the reactor vessel inside wall. An average heat 3 rate of 0.45 W/cm was used for the design of specimens and holders. This rate was based on a computer analysis of gamma heating from the core and from neutron capture in the internals at a core rated power of 2772 MWt. Two cases were analyzed. In the first case the holder tube was considered eolid; in the second it was considered perforated, with the area of the holes equal to 25% of the surface area of the solid tube. The perforated tube s.1-lowed enough coolant to reach the specimen capsules to cool then to within 14F of the temperature of the entering coolant water. Thereforem a perforated design was selected for the holder tube. 3-4 Babcock & )Milcox

I The capsules are locked into the holder tube by a removable closure device that subjects the capsules to a compressive load and the holder tube to an equal tensile load. This loading is designed to minimize flow-induced vibra-tion. (The tight inner packing also minimizes flow-induced vibrations within the capsule.) The perforated holder tube also causes the capsule to be sub-jected to the reactor coolant pressure. Structurally, the capsules are de-signed to withstand the compressive preload and the external pressure without failure. The capsule is designed to maintain specimens to within 125F of the reactor i vessel temperatexe at the 1/4 thickness location.* The heat transfer analysis considers the differences in thermal properties of the materials and the helitzn-filled gaps between the capsule and the internal components. Conser-vative maximum temperatures were calculated for each different cross section within the capsule. The coolant temperature serves as the lower bound and is within 125F of the vessel temperature at 1/4T. The capsules are placed in the holder tubes (two per tube), which are then i positioned so that both the time-averaged axial distribution of the axial ( peak neutron flux and the initial azimuthal distribution of fast neutron flux are maximized. (Holders are adjacent to the thermal shield in the integrated I program, whereas in other RVSPs they were adjacent to the reactor vessel in-side wall. Further, in the original surveillance programs, the self-shielding effect of the specimens was minimized by a design feature that permitted 180-degree rotation of the capsule during refueling. However, since the self-shielding effects have been found to be negligible, the new capsule holder tubes do not permit rotation.) 3.1.2. Dosimeters Dosimeters are placed in the specimen capsules to determine the actual neutron fluence levels experienced by the specimens. Each capsule contains four dosim-eter tubes, each tube accommodating six dif ferent dosimeters; the dosimeter types are defined in Table 3-2.

• The properties at the 1/4T vessel location are the basis for periodic adjust-ments of the pressure-temperature relationships for normal, upset, and test conditions throughout the vessel service life.

3-5 Babcock s Wilcox

I Dosimeter tube placement within the capsulas is shown in Figures 3-1 and 3-2. The tubes run the length of the specimen stacks so the actual fluence experi-enced by the specimens can be determined. 3.1.3. Temperature Monitors A nunber of low-melting fusible alloy temperature monitors are included in each capsule to determine the maximum temperature during the irradiation expo sure. The temperature monitors and their alloy composition and melting temperatures are given in Table 3-3. The locations of che temperature moni-tors within the capsule are shown in Figures 3-1 and 3-2. 3.1.4. Types of Surveillance Programs and Capaules As stated previously, there are three types of surveillance programs using six types of capsules in the integrated RVSP. The basic programs and capsule types are described briefly below, and more detailed information is presented in Ap-pendix D. An overview of the program and capsule types is given in Table 3-4. 3.1.4.1. Surveillance Program A Surveillance program A consists of capsule types I and II; it is described in topical report BAW-10006 Rev. 3. Types I and II are the upper and lower cap-sules in the holder tube, respectively. Capsule Type I - Capsule type I contains eight tensile specimens and 36 Charpy specimens. Tensile specimens are prepared from weld metal and base metal A in l the transverse direction. Charpy specimens are prepared from weld metal, the 1 heat-affected zone (HAZ) of base metal A in the longitudinal direction, base metal A in both longitudinal and transverse directions, and correlation moni-tor plate. l Capsule Type II - Capsule type II also contains eight tensile specimens and 36 l l Charpy specimens. Tensile specimens are prepared from the HAZ of heat B in 1 the longitudinal direction and base metal heat B in the longitudinal direction. Charpy specimens are prepared from the HAZ of heat B in the longitudinal direc-tion, base metal heat B in both the longitudinal and transverse directions, and correlation monitor plate. 3-6 Babcock a.Wilcox I

3.1.4.2. Surveillance Program B Surveillance program B consists of capsule types III and IV. The program is described in topical report BAW-10100A (referred to therein as the modified 1 program). In addition to tensile and Charpy specimens, compact tensile speci-mens 0.5 inch thick are included in capsule type IV. Types III and IV are the I upper and lower capsules in the holder tube, respectively. This program is designed for those reactor vessels with marginal material properties. Capsule Type IQ - Capsule type III contains four tensile specimens and 54 Charpy specimens. Tensile specimens are prepared for the weld metal and base metal heat A in the transverse direction. Charpy specimens are prepared for i the weld metal, HAZ heats A and B in the transverse direction, and correlation monitor plate. Capsule Type IV - Capsule type IV contains four tensile specimens, 36 Cha py specimens, and eight compact fracture specimens 0.5 inch thick. Tensile speci-j inens are prepared for the weld metal and base metal heat A in the transverse direction. Charpy specimens are prepared for the weld metal, the HAZ of heat I A in the transverse direction, and base metal heat A in the transverse direc-tion. The compact fracture specimens are prepared for the weld metal. 3.1.4.3'. Surveillance Program C Surveillance program C consists of capsule types V and VI. The program, de-scribed in topical report BAW-10100A is referred to as the basic program. Capsule types V and VI are the upper and lower capsules in the holder tube, respectively. This p agram is for those reactor vessels with controlled chem-istry and adequate initial material properties. Capsule Type V - Capsule type V contains four tensile specimens and 54 Charpy specimens. Tensile specimens are prepared for the weld metal and base metal heat A in the transverse direction. Charpy specimens are prepcred for weld metal, the dAZ of heat A in the longitudinal direction, base metal heat A in the longitudinal and transverse directions, and heat B in the transverse di-rection. Capsule Type VI - Capsule type VI contains four tensile specimens and 54 Charpy I specimens. The tensile specimens are prepared from the veld metal and base metal A in the transverse direction. Charpy specimens are prepared for the 3-7 Babcock a. Wilcox

weld metal, the HAZ of heats A and B in the longitudina. direction, base metal of heats A and B in the transverse direction, and correlation monitor plate. 3.2. Research Capsule program 3.2.1. Introduction The research capsule program is designed to evaluate eight weld metals (Wl, W2, W3, W4, W5, W6, W8, and W9) contained in six research capsules. The capsules are irradiated in the three host reactors (section 2) of the integrated RVSp. These host reactors (each containing two research capsules) are Three Mile Island Unit 2. Crystal River Unit 3, and Davis Besse Unit 1. The six research capsules are labeled TMI2-LGl. TMI2-LG2, CR3-LGl. CR3-LG2, DB1-LG1, and DB1-LG2. The first letters and the first number of these labels are the initials of the host reactor. The letters LG are an abbreviation for "large," and the last ntaber identifies the two capsules at each reactor site. The research capsules are considerably larger than the original plant-specific capsules irradiated at the same reactor sites. The original capsules have an inside diameter of 2.0 inches and an effective length of approximately 18.3 inches. The large-research capsules have an inside diameter of 2.5 inches and an effective length of 22.0 inches. Each research capsule contains Charpy Tensile, and compact fracture specimens from three welds. However, not all the capsules are alike; they have been categorized as types R-1 and R-2. The two capsule designs are shown in Figures 2-3 and 3-4. The type R-2 capsule represents an improved design since it utilizes miniature (Charpy size) tensile specimens. The min-g iature specimens allow the addition of five more tensile and three more com-5 pact fracture specimens per capsule then the' original design. In addition. ( there are small variations between types R-1 and R-2 in terms of the location of the temperature monitors and neutron dosimeters. This section describes the content of each type of capsule as well as the types of specimens. The TMI-2 capsules are type R-1, and the CR-3 and DB-1 capsules are type R2. Table 3-5 identifies the veld metals irradiated in each capsule as well as the distribution of specimens. The specimens listed as 0.394 TCT, 0.500 TCT, ant' O.936 TCT are the compact fracture specimens of 0.394, 0.500, and 0.936-g inch thickness, respectively. The 0.394 TCT and 0.500 TCT specimens are mod-E ifications of ASTM E-399 standard geometry, and the 0.936 TCT specimen is round. The number rf Charpy and tensile specimens per veld per capsule is 3-8 Babcock & Wilcox

lI il l adequate to characterize the toughness and tensile properties for each veld metal and irradiation condition. Other related research programs are expected to generate sufficient information to properly identify the methods (i.e., static versus dynamic) and test temperatures at which these research capsule compact fracture specimens should be tested. The combination of compact i fracture specimens is believed to be adequate to confirm the toughness curves. The information generated by the research capsules and the HSST program will enable accurate prediction of the toughness of the reactor vessel welds. 3.2.2. Research Capsule Design The cylindrical research capsules, like the RVSP capsules described previously, l contain Charpy, tensile, and compact fracture specimens as well as dosimeters and temperature monitors. The specimens are placed in stacks and are held in I place with alunine spacers. The cylindrical capsule is the principal charac-teristic of the B&W design. The unique advantage of the cylindrical capsule is that it allows for capsule replacement and for uniform specimen temperatures. However, the research capsules are larger than the standard capsules - in both length and diameter. The research capsules have a larger diameter because t. hey are usea to irradiate relatively thick compact fracture specimens. The original standard capsules will hold only eight 0.5-inch-thick compact fracture specimens. I When the research capsules were designed, many uncertainties were associated with the measuring capacity of the smaller compact fracture specimens as well as the required number of specimens needed to fully characterize the fracture proper-ties. The size of the capsule was also determined by the physical constraints within the reactor vessel as well as the restrictions on the specimen. metal temperature during irradiation. The end fittings are wedge-shaped and chrome-placed to minimize surface wear. The material of construction is type 304 stainless steel for both the shell and end filling. Alminum spacers hold the specimens, dosimeters, and tempera-I ture monitors in place and fill the gaps within the capsule. The remaining spaces are filled with helium gas. The capsules are locked in place in a holder tube assembly. The wedge-shaped end fittings are used to position and lock the capsule inside the holder tube. The individual capsules can be with-drawn and replaced with other capsules whenever necessary. The capsules are normally withdrawn during plant refueling. The capsule holder tube assembly is designed to permit remote removal and replacement of the capsules from the 3-9 Babcock s Wilcox

I I . fuel handling bridge without removing the plenum assembly. When all the cap-j sules containing specimens have been irradiated to the desirable neutron expo-sure, empty danmy capsules are installed to fill the available capsule loca-tions in the surveillance specimen holder tubes. The type R-1 capsules were designed before the R-2's and used the type of ten-sile specimens found in the standard capsule design of the 177-FA RVSPs. By the time the CR-3 and DB-1 research capsules were designed, it was recognized that the standard sized tensile specimens were not necessary (see Figure C-8). The use of miniature Charpy-sized tensile specimens (see Figure C-9), permitted the inclusion of three additional compact fracture specimens, which are the most important specimens included in the capsules. Also, the use of the miniature specimens permitted the addition of five more tensile specimens. The tensile specimens are also important because it is expected that at the upper shelf temperature the fracture resistance properties of the material will be depen-dent on the tensile properties. Each capsule contains specimens. from three different weld metals. The weld metals and distribution of specimens per weld are described in Table 3-5 g of this report. The size and amber of tensile specimens are sufficient to u obtain the tensile properties at several temperatures. The number of Charpy V-notch specimens is that recommended by ASTM E-185 as the minimun nmber re-quired to obtain a full Charpy V-notch data curve. The size and number of compact fracture specimens per veld is adequate for determination of the frac-ture toughness properties of the corresponding welds (based on current state-of-the-art). The tension, Charpy and compact fracture specimens are described g I in Appendix C. 3 Each capsule also contains dosimeters to measure fluence and temperature moni-tors to measure irradiation temperature. The dosimeters and temperature moni-tors are described later in this section. I The arrangements of the specimens, dosimeters, and temperature monitors with-in the capsulen are illustrated in Figures 3-3 and 3-4 for capsule types R-1 and R-2. 3-10 Babcock s.Wilcox

3.2.3. Holder Tube and Capsule Location l Surveillance capsule holder tubes are attached to the thermal shield and po-sition the capsules in the downcomer annulus near the reactor vessel wall. The holder tube is located so that the midspan elevation of the tube is at the. core midplane. The azimuthcl locations of the holder tubes are shown in Figures 3-5, 3-6, and 3-7 for the DB-1, CR-3, and TMI-2 plants, respectively. Tabla 3-7 pro-vides a list of the locations of the six capsules and their azimuthal loca-i tions. 3.2.4. Structural, Hydraulic, and Thermr.1 Characterization of ReLearch l Capsules I The capsules are locked into the holder tube by a removable closure device that subjects them to a compressive load and the holder tube to a tensile load. This preloading is designed to minimize ' flow-induced vibration. (The i tight inner packing also minimize flow-induced vibration within the capsule.) As a structural member, the capsules are designed to withstand the compressive preload and the external pressure without failure, although permanent defor-mation of the capsule cylindrical vall may occur. The capsule is designed to maintain specimens at temperatures within 25F of the reactor vessel temperature at the 1/4T vessel wall location. Figure 3-8 illustrates the calculated vessel wall temperature distribution for steady-state nomal operation. The temperature gradient is caused by gamma heating and by the fact that the reactor, vessel outside wall is insulated. The tem-perature calculated for the 1/4T vessel wall location is 576F. The maximum specimen temperature is calculated to be 595F and the minimtsn 554F, which is the temperature of the coolant at s' ady state. The capsule heat transfer 6 analysis accounts for the differences in the thermal properties of the mate-I rials and the helium-filled gaps between the capsule and internal components. The 595F maximum expected specimen metal temperature is for the specimens at the cluster of the Charpy specimen stack. 3.2.5. Do simetry Each capsule contains dosimeter tubes, which in turn contain neutron-detecting .1....t.1r.s in s. m a..t v.a.c - d, a c.. t...

r. -.gr.t.d

I flux, fast neutron spectrum, and thermal neutron spectrum. A variety of neu-tron detecting elements was chosen in accordance with ASTM Standard Recommend-ed Practice E-482. The dosimeters are distributed throughout the capsule to measure the dosage rates at various locations. m .I 3.2.5.1. Do simeters Table 3-8 lists the neutron detecting eierente. and provides the energy range and shielding required for each element. The gadolinium or alumintun shielding containing the neutron detecting elements is 15 to 50 mils thick (the lower bound) due to the necessity for eliminating thermal neutrons which cause com-peting reactions and the upper bound to prevent significant absorption of fast neutrons. The neutron detecting elements, along with their shielding, are then stacked in aluminum holder tubes, which may contain from 5 to 10 elements. 3.2.5.2. Dosimeter Locations Seven seta of dosimeters are distributed throughout the capsule in order to measure the neutron flux at various specimen stack locations. The Charpy and tensile specimen stacks are monitored by three dosimeters; the design is such that two of them are at the O' and 180* locations in the capsule with respect to the center of the reactor core. The third dosimeter is located in the center of the Charpy and tensile stacks. F'inally, two dosimeters are placed through the openings in the round compact fracture specimen stacks. Figures 3-3 and 3-4 show the exact locations of the dosimeters in the capsules. 3.2.5.3. Temperaturs Monitors ( Temperature monitors are distributed throughout the capsule to measure speci-men temperatures. Each set of temperature monitors contains five low-melting-point elements or eutectic alloys whose melting points range from 580 to 621F. By determining which monitcrs have melted, the peak temperature at various lo-cations within the capsule is determined. Melting Point Elements and Eutectic Alloys - Table 3-9 lists the five tempera-ture monitors and their respective melting temperatures. Gap dimensions within the capsule have been sized based on a heat transfer analysis to maintain the specimen temperature within 225F of the temperature at the reactor vessel 1/4T location. The range of 580 to 621F is adequate to monitor the expected speci-men temperaturas. 3-12 Babcock & Wilcox I

I Location of Temperature Monitors - Six temperature monitors are placed in the I capsule to measure the specimen temperatures. The locations of these monitors are shown in Figures 3-3 and 3-4 for the types R-1 and R-2 capsules, respec-tively. Five monitors are used for the Charpy and tensile stacks, and one monitor is placed in the round compact fracture specimen stack. 3.2.6. Insertion and Withdrawal Schedules The research capsules are incorporated in the surveillance capsule insertion and withdrawal schedules of the reactors in the program. i 3.2.7. Unirradiated Control Data The unirradiated baseline data needed to support the evaluation of the irradi-J ated capsule data from the research capsules will be obtained from two sources. The primary sources for these data are sets of specimens that have been pre-pared from the same weld metal used in the research capsules. These sets of specimens are similar to those included in the capsules but of a longer quan-tity, so that a better data base can be established. The type and number of specimens of each material are described in Table 3-9. I Some matorial in excess of the needs of the program was donated to the Heavy Section Steel Technology Program to obtain test reactor irradiation data. I Since this program would be obtaining baseline unirradiated data of the same type as needed by the Research Capsule program, it was decided not to dupli-cate the efforts of the HSST Program. The sources of the baseline data for the eight welds in the research capsule program are identified in Table 3-10. 3.3. Irradiation Schedule 1 l The capsule irradiation schedule is important to the integrated RVSP because it is not physically possible to irradiate all the capsules simultaneously. Therefore, the schedule must ensure that each participating pl at will have l capsules being irradiated and removed that lead the irradiation damage the re-actor vessel is experiencing. To ensure that the capsules lead their respec-tive plants, the lead fe. tors of the holder tubes are greater than normally permitted, but since most of the reactors have had at least one capsule with-drawn prior to initiation of the integrated RVSP, this is a basis for evalu-ating any abnormal behavior that may be attributable to the higher lead factor. I 3-13 Babcock s. Wilcox I 1

I

• Figures 3-5, 3-6, and 3-7 show the locations of the holder tubes in the vari-ous plants along with the lead factors for the holder tubes. Tables 3-11, 3-12, and 3-13 define the insertion and withdrawal schedules for each host reactor.

These sch;.fules are based on insertions and withdrawals only during refueling outages since the radiation levels for individual capsules are not defined precisely. The schedules include the capsules of both the guest and host re-E actors and the research capsules. EB I I I I I I I I l I I I 3-14 Babcock & Wilcox

I I. Table 3-1. Typical Withdrawal Schedule for I Six-Capsule Program Capsule Purpose of withdrawal First capsule When the highest predicted RTNDT shift of all the capsule materials is approximately 50F Second capsule When the capsule's accumulated neutron fluence (E > 1 Mev) corresponds to an intermediate value between those of the l 'first and third capsules Third capsule When the capsule's accumulated fluence (E > 1 Mev) cor-l responds to that at the 1/4T reactor vessel wall location I at approximately the end of vessel's design service life Fourth capsule When the capsule's accumulated neutron fluence (E > 1 Mev) corresponds to that of the reactor vessel inner wall lo-cation at approximately the end of the vessel's design service life Fifth and sixth Standby capsules I Table 3-2. Surveillance Capsule Dosimeters Eff cross section thresh. Dosimeter energy Counting technique Co-Al Thermal y - 5.3 yr Co 5"Fe 2.5 MeV y - 314 d 5"Mn 2se Cd-shielded U 1.1 MeV Appropriate fission products 237 Cd-ohielded Np 0.5 MeV Appropriate fission products Cd-shielded Co 0.5 eV y - 5.3 yr Co Cd-shielded Ni 2.3 MeV y - 71 d ss i N I. I I I 3-15 Babcock & Wilcox

l ~ I Table 3-3. Temperature Monitor Wires I l Approximate melting point, F Reference materials 580 97.5% Pb, 2.5% Ag 588 97.5% Pb, 1.5% Ag, 1.0% Sn 598 98.8% Cd, 1.2% Cu 610 100% Cd 621 100% Pb Note: The melting point of each alloy heat or l batch shall be verified by the fabricator 5 from its final form and reported to the system designer. I ~ I I I i I 3-16 Babcock s.Wilcox I +

I ~ I Table 3-4. Reactor Vessel Surveillance Program - Detailed Summary Capsule apsule Table of Date Table of Date ID, Tm mat specs tested ID Tm mat. specs tested Oconee Unit 1 Oconee Unit 2 A I B-2 Nov '81 A I B-1 B II B-2 B II B-3 l C I B-3 C I B-2 June '76 I~ D II B-3 D II B-2 E I B-3 June '76 E I B-2 l i F II B-3 July '75 F II B-2 Topical Report BAW-10006A, Rev 3 Topical Report BAW-10006A, Rev 3 Three Mile Island Unit 1 Three Mile Island Unit 2 l5 A III B-5 A I B-4 B IV B-5 B II B-4 C I B-4 C III B-5 O II B-4 D IV B-5 I E I B-4 June '76 E III B-5 F II B-4 F IV B-5 Topical Report BAW-10006A, Rev 3 Topical Report BAW-10100A Crystal River 3 Arkansas Nuclear One Unit 1 A I B-6 A III B-7 I B IV B-7 , June '81 B II B-6 C III B-7 C I B-6 D II B-6 D IV B-7 I E I B-6 Aug '76 E III B-7 F II B-6 F IV B-7 Topical Report BAW-10100A Topical Report BAW-10006A, Rev 3 Rancho Seco 1 Davis Besse Unit 1 i A III B-9 A III B-8 I B IV B-8 B IV B-9 C III B-9 C III B-8 D IV B-8 D IV B-9 E III B-8 E III B-9 I F IV B-8 F IV B-9 Topical Report BAW-10100A Topical Report BAW-10100A Oconee Unit 3 A V B-3 July '76

• The OC-3 capsules were fabri-B VI B-3 Oct '81 cated before BAW-10100A was I

C V -B-3 published; however, it was the OC-3 program that was described D VI B-3 E V B-3 in BAW-10100A. I F VI B-3 Topical Report BAW-10100A* 3-17 Babcock & Wilcox i i

~ Table 3-5. Research Capsules - Material and Specimens per Capsule pe imens Weld metal ID per capsule Tensile Charpy 0.394 TCT 0.500 TCT 0.936 TRCT Capsule TMI2-LG1 W1 3 12 2 4 3 W4 2 12 2 4 3 W5 2 12 2 4 3 Capsule TMI2-LG2 W4 2 12 2 4 3 W5 2 12 2 4 3 W8 3 12 2 4 3 Capsule CR3-LG1 W1 4 12 2 4 4 W6 4 12 2 4 4 W8 4 12 2 4 4 Capsule CR3-LG2 W1 4 12 2 4 4 W3 4 12 2 4 4 W6 4 12 2 4 4 Capsule DB1-LG1 W1 4 12 2 4 4 W2 4 12 2 4 4 W9 4 12 2 4 4 Capsule DB1-LG2 W1 4 12 2 4 4 W2 4 12 2 4 4 W9 4 12 2 4 4 I I l 3-18 Babcock t.Wilcox

I 3 !W l Table 3-6. Contents of Research Capsules l l l Type Type No. of test specimens R-1 R-2 Standard tensile 7 12 Miniature tensile I Charpy 36 36 0.394 TCT 6 6 0.50 TCT 4 4 0.936 TRCT 9 12 l Table 3-7. Azimuthal Location of Power Reactor i Research Capsules Vertical location Capsule with reference to Arimuthal i ID h Reactor core midplane location DB-LG1 R-2 Davis Besse 1 Upper 10.9* off W DB-LG2 R-2 Davis Besse 1 Lower 26.5" off Z I CR3-LG1 R-2 Crystal River 3 Upper 10.9* off W CR3-LG2 R-2 Crystal River 3 Lower 10.9* off W TM12-LG1 R-1 TMI-2 Upper 10.9* off W a TMI2-LG2 R-1 TM1-2 Lower 26.5* off Z I I I. I I I 3-19 Babcock 4,Wilcox

I Table 3-8.' Neutron-Detecting Elements Neutron-Neutron-sensitive sensitive Thickness, Max thick-energy range element Material in. Material ness, in. 0.4 eV Co coated A1 0.015-0.050 A1 0.030 with Cd 2s7 20.5 MeV Np Gd 0.015-0.050 A1 0.030 23e (*} 21.1 MeV U Gd 0.015-0.050 A1 0.030 Ni( Gd 0.015-0.050 A1 0.030 38 12.3 MeV 22.5 MeV 5"Fe Gd 0.015-0.050 A1 0.030 26.1 MeV Cu Gd 0.015-0.050 A1 0.030 All levels Co A1 0.015-0.050 A1 0.030 Data redundancy Co + Fe A1 0.015-0.050 A1 0.030 verification (a) Mini.am limit 285U 380 ppm. Maximum limit Co 0.5 ppm Table 3-9. Matrix of Cont.rol Specimens for Welds W2, WA. W6, W8 Under Power Reactor Program Power Weld reactor metal 0.394 0.500 0.936 1.000 2.000 program ident Tensile Charpy" TCT TCT TRCT TCT TCT TMI-2 W4 4 22 5 8 5 2 W8 4 22 5 8 5 2 DB-1 W2 4 22 5 8 5 2 W6 4 22 5 8 7 2 CR-3 W8 4 22 5 8 5 2 I \\ l l 3-20 Babcock & Wilcox l l t

E I I Table 3-10. Identification of Prograr,s and Control Specimens for Eight Research Capsule, Program Welds Weld ID Program W1 RSST Task 3 W2 Power Reactor Program - DB-1 W3 HSST Task 3 W4 Power Reactor Program - TMI-2 l W5 HSST Tasks 2 and 3 W6 Power Reactor Program - DB-1 I W8 Power Reactor Program - TMI-2, CR-3 W9 HSST Task 3 II I iI I I b I I I g 3 21 ..scocx wii; x

~I Table 3-11. Surveillance and Research Capsule Insertion and Withdrawal Scheduled for Crystal River Unit 3 Holder Location in tube Remove holder tube Insert Installed at Initial Fuel Load XW Top CR3-B XW Bottom CR3-D End of First Fuel Cycle WZ Top CR3-LG1 WZ Bottom CR3-LG2 ZY Top CR3-C ZY Bottom CR3-A YZ Top OCII-A YZ Bottom OCI-A YX Top OCII-E YX Bottom OCIII-D XW CR3-B Top CR3-E WX Top OCIII-B WX-Bottom CR3-F End of Second Cycle YZ OCII-A Top OCI-C WX OCIII-B Top TMI-1C End of Fourth Cycle WZ CR3-LG1 Top OCII-B End of Fifth Cycle ZY CR3-C Top OCIII-B XW CR3-D Bottom OCIII-C End of Seventh Cycle YZ OCI-A Bottom OCI-B WX TMI-1C Top Dummy #2 I 3-22 Babcock a.Wilcox I

I Table 3-11. (Cont'd) Holder Location in tube Remove holder tube Insert End of Ninth Cycle WZ OCII-B Top OCII-D YX OCII-E Top OCII-F j -YX OCIII-D Bottom OCI-D End of Tenth Cycle ( WX Dummy #2 Top None WX CR3-F Bottom None XW OCIII-C Bottom ')umm" #1 I [ End of Eleventh Cycle WX CR3-LG2 Bottom OCIII-F YZ OCI-C Top Dummy #2 { End of Fourteenth Cycle ZY CR3-A Bottom Dummy #1 XW CR3-E Top None { XW Dummy #1 Bottom None End of Sixteenth Cycle I YZ Dummy #2 Top None YZ - OCI-B Bottom None End of Eighteenth Cycle ZY OCIII-E Top None i ZY Dummy #1 Bottom None WZ OCII-D Top Dummy #1 YX OCII-F Top None YZ OCI-D Bottom None End of Twentieth Cycle WZ Dummy #1 Top None WZ OCIII-F Bottom None I 3-23 Babcock s.Wilcox

IL E Table 3-12. Surveillance and Research Capsule Insertion and Withdrawal Schedule for Davis-Besse Unit 1 Holder Capsule Location in tube ID holder tube Initial Installation WZ ANI-B Upper WZ RSI-B Lower ZY TEI-B Upper g ZY TEI-F Lower E YZ ANI-A Upper YZ ANI-C Lower YX RSI-D Upper YX TEI-C Lower XW TEI-D Upper XW RSI-C Lower WX TEI-A Upper WX RSI-F Lower At End of First Fuel Cycle Remove Insert WZ ANI-B Upper DB-LCl WZ RSI-B Lower RSI-E ZY TEI-F Lower DB-LG2 At End if Second Fuel Cycle YX RSI-D Upper RSI-A At End of Third Fuel Cycle WZ DB-LG1 Upper ANI-D ZY TEI-B Upper TEI-E YZ ANI-A Upper ANI-F At End of Fifth Fuel Cycle WX TEI-A Upper Dummy 3-24 Babcock 3. Wilcox

l Table 3-12. (Cont'd) Capsule

I Holder ID Location in l

tube (remove) holder tube Insert l At End of Seventh Fuel Cycle l YZ ANI-C Lower Dummy YX RSI-A Upper Dununy 1 At End of Ninth Fuel Cycle YX TEI-C Lower Nor' YX Dummy Upper None WX RSI-F Lower None I WX Dummy Upper None i' At End of Tenth Fuel Cycle I WZ RSI-E Lower Dummy At End of Eleventh Fuel Cycle ZY DB-LG2 Lower Dummy YZ ANI-F Upper None YZ Dummy Lower None 6 At End of Twelfth Fuel Cycle WZ ANI-D Upper None WZ Dummy Lower None XW TEI-D Upper None XW RSI-C Lower None At End of Sixteenth Fuel Cycle ZY TEI-E Upper None ZY Dummy Lower None I I 3-25 Babcock s.Wilcox

Table 3-13. Surveillance and Research Capsule Insertion and Withdrawal Schedule for Three Mile Island, Unit 2 Holder Location in tube Remove holder tube Insert At Initial Fuel Load WX Top TMI2-LG1 WX Bottom TMI-2B XW Top TMI-2A XW Bottom TMI-2D ZY Top TMI-1A ZY Bottom TMI2-LG2 End of First Cycle WX TMI2-LG1 Top TMI-1B XW TMI-2A Top TMI-2F End of Third Cycle WX TMI-2B Bottom TMI-lD ZY TMI-1A Top TMI-2C End of Sixth Cycle WX TMI-1B Top TMI-lF XW TMI-2D Bottom TMI-2E End of Tenth Cycle ZY TMI2-LG2 Bottom Dummy #1 End of Eleventh Cycle WX TMI-lD Bottom Dummy #2 End of Thirteenth Cycle XW TMI-2F Top None End of Fifteenth Cycle WX TMI-lF Top None WX Dummy #1 Bottom None 3-26 Babcock & Wilcox

I Table 3-13. (Cont'd) Holder Location in tube Remove holder tube Insert End of Sixteenth Cycle ZY TMI-2C Top Dtunmy #1 End of Nineteenth Cycle l XW Dummy #2 Top None , i l I IW TMI-2E Bottom None Pnd of Twenty-third Cycle i I ZY Dummy #1 Top None iI iI 6 I I I I 3-27 Babcock s.Wilcox

Figure 3-1. Surveillance Capsule Arrangement TENSILE SPEClufNS CHARPT SPECluENS () @ 00$1 METER C O ~ @ TEMPERATURE MONITOR I THIS SECil0N ROTATED e 45* TOR CLARiff +l -o-1 x vxx l, "T~ x x x7 x s t xx x x x x vrxxxxxxxx xxx xx x x x x vv s 's s s s (\\ ///r / 3s / s,s i 1 y ~~ bl/ // / / / / / / / / / / ~ ~ ~ ~ //// \\ \\ \\ x x t' 1x (((((((( i ( (lix x x x x x x x x x xxxxx xx x

• I

+T UPPER CAPSULE CAPSULE IDENilFICAil0N SHOIN HERE BT VIBR A100 LING CODE USLNG 1/2" HIGH g3 L E TTE R ING. TTP BOTH CAPSULES e 8 x 9' bw I f i

M M t M cN tue e ts sNu M c e c-A M M ,Qs\\N, T. / m e s T n / _m e \\ / j 1 m S N \\ A/;. i M \\ c l \\ / e C W lt \\ ( x p P \\ / ^ S s \\ / o o i n \\ / ~ U o N \\~ s ' / i [ n \\ _./Qu 1 s o O M ~ e L T \\ / ~ ^ \\ t S N c \\^ ~ E E a l \\ ~ m_ p c E m P \\ ~ S o C \\ E \\^ R a___ . g u l h t \\ t AR ^ ~ ___Y W i ( \\ W ~ I C \\~ a A ~ Pe \\^ \\ 0 t C E ~ \\ n 0 M 1 e m \\- x eg \\ n \\/ a E 8 \\/ r f M C r A \\/ i e \\/ l u - u \\/il', g a _ _ t{ s \\/ { t M. s p _ ___~ s a \\/ a u C n \\/ a ._ T @e _8 e S N c \\/ \\ E n I \\/il', M W a C \\ s E \\/ l P ^ w l \\/ i I P e R \\/ n A v H \\/ r C M u f \\/ S \\ \\/Il), \\/ 2 \\/ m a 3 W \\/ m e \\/ r r u T NN\\ g i \\ F \\ M \\ /j / m a M a ._ _D M g yK8x t o M

I I l I m I' V///////4 i d l' j.l O. 4 l l 7 i l l m g .c a' E E N 's ( f J l C 3, I I i

m,m f

Y li 1 e hi V h [ ae j p_3 i l }(,_ e..m N! i e n i h I g Il l l 5 l O @O i O f ( 9 5 W 1; l I 3-30 Babcock & Wilcox I w-,

I I I' I R@[W m W ! ij I. I 1 j ( l I \\(: l g s 4 N 0 t b W g. l cw G: a / I i x m V f [ _ J _11 l_ [ _l__@L I r: i x u f/__ L #1- _L_ f - ~ \\ l f \\ _J JLL- \\ i i g \\ i I s i e !j , ld-. I c L_j._il_i_ Jew j = E I clcS c!OIR, r ;J.j !OiO'@lOlOi ' g. s I lGlOICici (?' 't gg s l 3-31 Babcock & *ifilcox

I Figure 3-5. Surveillance Capsule Holder Tube Location and Identification for Davis Besse Unit 1 I l-HOLDER -TUBE XW LEAD FACTOR, T/4 = 7.0 X l HOLDER TUBE-WX HOLDER TUBE-YX LEAD FACTOR, T/4 = 9.7 a LEAD FACTOR, T/4 = 9.7 s s e o I / e. O s, O e e I e c e c,; S e O e / L W l c 9,___ V_,y I /, s-c e e e e i / I l e e O e e x e e. \\ e t I e e / \\ e i g E e f ~~_ f HOLDER TUBE-YZ LEAD FACTOR, T/4 = 9,7 HOLDER TUBE-WZ LEAD FACTQR, T/4 = 9.7 Z l L. HOLDER TUBE-ZY LEAD FACTOR, T/4 = 7.0 I 3-32 Babcock & Wilcox I

> I I Identification fo'r Crystal River Unit 3 Figure 3-6. Surveillance Capsule Holder Tube Location and .- HOLDER TUBE-IW I LEAD FACTOR T/4 = 7.0 HOLDER TUBE WX HOLDER TUBE YI LEAD FACTOR. T/4 = 9.7 a LEAD FACTOR, I T/4 = C.7 I I, \\ / e O \\ / e-e e e g [ e e O_ \\ I 9 e e e e t o e e g-J e s e e e e t g O Y 0 0 0 i 0 W VJ r 8 e O O O e e e e I \\ l e e e i s e e I o e e l t i ~~ { HOLDER TUBE YZ I HOLDER TUBE-WZ LEAD FACTOR, LEAD FACTOR, T/4 = T/4 = 9.7 g,7 l Z L HOLDER TUBE ZY LEAD FACTOR. T/4 = 7.0 I I 3-33 Babcock s.Wilcox

i Il Figure 3-7. Surveillance Capsule Holder Tube Location and Identification for Three Mile Island Unit 2 1-HOLDER TUBE XW l LEAD FACTOR, T/4 = 7.0 X HOLDER TUBE-WX LEAD FACTOR, T/4 = 9.7 g .~~ s N e e E / e 8 g / G o l e e s E y e e e \\ e e e e e e e e e e f g y e-W y 3 e e e e e e s_ e e o e 0 \\ ~~ e e e e \\ e e e / ,I s e-e e e I s e o x s ~~- t L HOLDER TUBE-ZY J LEAD FACTOR T/4 = 7.0 I Z I I 3-34 Babcock & Wilcox

i Figure 3-8. 177-FA Reactor Vessel Through-Thickness Temperature Distribution at Steady-State Normal Operation l 590 I I 580 I e c5 E a a l G 570 -5 l I. l,it, l 1 l I 3 560 .l = EI ai ?;; I 550 i l l 540 I I I 0 2 4 6 8 10 l. Distance from vessel inner surface (in) I 2-35 Babcock s.Wilcox

i a l l i i i i 1 APPENDIX A l Description and Propertias of RVSP Materials { i I 1I l l I I l A-1 Babcock a.Wilcox 1 6- _. - _ _ _,,

Table A-1. Oconee 1 Description and Properties of Reactor Vessel Surveillance Program Materials Impact properties Chemical composition. % T RT C -USE. hterial g. ET. y MS. TS. ID C Mn P S Si Ni g Mo Cu F F ft-lb kai kai Base metal A 0.21 1.42 0.015 0.015 0.23 0.50 0.49 0.10 0 20 109 87.0 66.0 Base metal B 0.20 1.40 0.012 0.017 0.20 0.63 0.50 0.11 20 20 119 88.5 69.0 Weld metal 0.075 1.50 0.024 0.006 0.60 0.58 0.51 0.22 -50 0 65 83.0 66.0 Material Heat Spec ID No. No. Supplier Austenitizina Tempering Stress relief Base metal A C3265-1 SA 302 CR.B Lukens 1600-1650F for 9.75 1200-1220F for 9.5 h. 1100-1150F for 40 h. Y

h. brine quench brine quench furnace-cooled to 600F Base metal B C2800-2 SA 392 CR.B Lukens 1600-1650F for 9.5 1200-1225F for 9.5 h.

1100-1150F for 40 h.

h. brine quench brine quench furnace-cooled to 600F Weld metal W-112 NA NA NA NA

!!00-1150F for 31.0 h. furnace-cooled 8-8x to a-m m m m m e e e e e

,E E E.E. ._ W.. E _. . W _W. _. E E E Table A-2. Oconee 2 Description and Properties of Reactor Vessel Surveillance Program Materials Impact properties em ca position, I T -USE,

UTS, YS, Material FDT' NDT' v

ID C Mn P S Si Ni Cr Mo Cu F F ft-lb kai kai Base metal A 0.24 0.63 0.006 0.012 0.25 0.75 0.36 0.62 0.04 20 20 133 93.2 72.1 Base metal B 0.21 0.62 0.010 0.010 0.23 0.80 0.39 0.58 0.02 -10 -10 138 71.3 90.6 Weld Metal 0.067 1.58 0.020 0.005 0.56 0.48 0.12 0.33 0.30 -20 10 69 87.5 72.5 Material Heat Spec ID No. No. Supplier Austenitizina Temperina Stress relief Base metal A 3P-2359 SA 508 CL.2 LADISH 1640F 1 20F held at 1260F 1 20F held at Il25F 1 25F held at T color 4h, cold water color 10 h. cold color 40 h, furnace quenched at 1590F 1 water quenched. cooled to below 600F 20F held at color 4 h. ccid water quenched. Base metal B 4P-1885 SA 508 CL.2 LADISH Same as above Same as above Same as above Weld metal WF-209-1 NA NA NA NA 1100-Il50F for 33 h, furnace-cooled 8' Rx 9' W l N l l l

i Table A-3. Oconee 3 Description and Properties of Reactor Vessel Surveillance Program Materials Impact properties

• • IC' "I *

-USE. UTS. YS, Material NDT' NDT' v ID C Mn P S Si Ni Cr Ho Cu F F ft-lb ksi ksi l Base metal A 0.24 0.72 0.014 0.012 0.21 0.76 0.34 0.62 0.02 40 86.5 61.3 Base metal B 0.21 0.58 0.011 0.015 0.24 0.73 0.30 0.60 0.01 40 84.8 60.8 Weld metal 0.067 1.58 0.020 0.005 0.56 0.46 0.12 0.33 0.30 60 66 87.5 72.5 Material Heat Spec ID No. No. Supplier Austenitizing Tempering Stress relief Base metal A 522194 SA 508 CL.2 LADISH 1640F 1.20F held at 1250F 2 20F held at ll25F 1 20F, held at Y color 4 h. cold color for 10 h, cold color 40 he furnace t; l vater quenched.1590F wat.er quenched. cooled to below 600F. 1 20F held at color for 4 h cold water quenched. s Base metal B 522314K SA 508 CL.2 LADISH Same as above 2ame as above Same as above I Weld metal WF-209-1 NA NA NA NA 1100-1250F for 30 h, furnace-cooled en Rw 9' N w is E M E E m w m m M M M e_ A

M,M M .M M M._M M __ M _ . M _m... ._M M M M. m l Table A-4. THI-l Description and Properties of Reactor Vessel Surveillance Program Materials Impact properties Chemical Composition, % T RT C -USE. -UTS. YS. hterial ET * - NDT. ID _ C_ Hn P S Si Ni Cr Ho Cu F F ft-lb ksi kai 0.51 0.09 10 30 98 92.0 67.0' Base metal A 0.24 1.36 0.010 0.017 0.23 0.59 0.47 0.12 -10 20 112 86.0 64.25 Base metal B 0.21 1.24 0.010 0.016 0.27 0.55 Weld metal 0.088 1.50 0.015 0.013 0.45 0.71 0.11 0.33 0.29 -20 -14 81 80.75 66.5 I l "**E E*E"*"" Haterial Heat Spec ID No. No. Supplier Austenitizing Tempering Stress relief l Base metal A C2789-2 SA 302 GR.B LUKENS 1600-1650F for 9,5 h brine quench 1100-1150F for 40 n. 1200-1225F for 9.5 h. brine quench furnace cooled. 1600-1650F for 9.5 h. brine quench 1600-1650F for 9.5 h. brine quench 1510-1535F for 5 h brine quench l 1200-1225F for 5 h brinu quench l Base metal B C3307-1 SA 302 GR.B LUKENS 1600-1650F for 9.5 h. brine quench Same as above 1200-1225F for 9.5 h. brine quench 1225-1250F for 9.5 h brine quench Weld metal WF-25 NA NA NA NA Il00-Il50F for 27.5 h. furnace cooled 8' Rw 9' w

l Table A-5. THI-2 Description and Properties of Reactor Vessel Surveillance Program Materials Impact properties Chemical Composition, % T RT C -USE. UTS. YS. g NDT. ID C Hn P S Si Ni Cr Ho Cu F F ft-lb ksi ksi Base metal A 0.22 1.35 0.010 0.015 0.23 0.50 0.47 0.12 0 40 100 87.0 63.5 Base metal B 0.22 1.42 0.010 0.015 0.19 0.50 0.47 0.14 -10 16 101 91.75 68.5 0.46 0.21 -30 19 84 81.0 NA i Weld metal 0.09 1.74 0.13 0.017 0.45 0.61 i i E"*E"*"E Haterial Heat Spec ID No. No. Supplier Austenitizing Tempering Stress relief Base metal A C1946-2 SA 533 LUKENS 1600-1650F for 9.5 h 1150-1200F for 4.5 h. 1100-Il50F for 40 h. CR.B. CL I brine quench. brine quench. furnace cooled. Base metal B C1937-2 SA 533 LUKENS Same as above ll80-1220F for 9.5 h, Same as above CR.B. CL. I brine quench. 1200-1225F for 9.5 h. brine quench. Weld metal WF-182-1 hA NA NA MA !!00-Il50F for 48 h. furnace-cooled 5 K 8x P -n mas sus amm aus uma sus ums num aus man ums

M.M .M M._ __ M _ M.. M _ M. ._ m._ M m M Table A-6. ANO-1 Description and Properties of Reactor Vessel Surveillance Program Materials Impact propertie.. -USE,

UTS, YS, Material NDT' NDT' v

ID C Hn P S St Ni Cr Ho Cu F F ft-lb ksi kai l Base metal A 0.21 1.32 0.010 0.016 0.20 0.52 0.57 0.15 10 30 93 92.5 69.0 Base metal B 0.21 1.32 0.010 0.016 0.20 0.52 0.57 0.15 20 10 107 29.0 55.7 Weld metal 0.065 1.50 0.016 0.008 0.42 0.59 0.36 0.19 -20 30 73 82.0 9 Material Heat Spec ID No. No. Supplier Austenitizing Tempering Stress relief Base metal A C5114-1 SA 533 CR.B LUKENS 1650-1700F, held 1 1200F, held I h/in./ 1100-Il50F, held 60 h, a CL. I h/in./ min and water min and air cooled. and furnace cooled p quenched to 400F. within a rate of 35F/h to below 600F. Base metal B C5114-2 SA 533 CR.B LUKENS Same as above Same as above Same as above CL. 1 Weld metal WF-193 NA NA NA NA !!OO-il50F for 27.5 h, furnace-cooled X Rx P n

Table A-7. Crystal River Unit 3 Description and Properties of Reactor 7essel Surveillance Program Materials Impact properties emical Composition, % T -USE,

UTS, YS, Material NDT*

NDT' v ID C Hn P S St Ni Cr Ho Cu F F ft-lb ksi ksi 0.55 0.20 -10 20 88 86.7 62.8 Base metal A 0.23 1.30 0.008 0.016 0.22 0.54 Base metal B 0.23 1.30 0.008 0.016 0.22 0.54 0.55 0.20 -10 20 88 90.0 66.4 Weld metal 0.067 1.58 0.020 0.005 0.56 0.48 0.12 0.33 0.30 -50 43 63 87.5 72.5 Heat treatment Material liest Spec ID No. No. . Supplier Austenitizing Tempering Stress relief l Base metal A C4344-1 SA 533 CR.B LUKENS 1650-1700F, held 1 1180F, held 0.5 1100-Il50F held 60 T CL. I h/in./ min water h/ia./ min air cooled h, furnace cooled to below 600F. ll quenched to 400F. O' Il Base metal B C4344-2 SA 533 GR.B LUKENS Same as above. Same as above. Same as above CL. 1 Weld metal WF-209-1 NA NA NA NA 1100-1150F for 27 h, furnace-cooled m Rx to bw M M M M M M M M M M M M

M - M M M M- .M- ...M M M.. M M l Table A-8. Rancho Seco Description and Properties of Reactor Vessel Surveillance Program Materials Impset properties T -USE. UTS. YS. '"I** "E ' *I "* NDT. NDT* v Material ID C Mn P S St Ni Cr Ho Cu F F ft-lb kai kai 0.52 0.10 -20 0 92 86.25 65.75 Base metal A 0.20 1.33 0.010 0.015 0.19 0.58 0.35 0.12 -10 4 90 85.0 64.0 Base metal B 0.20 1.26 0.013 0.017 0.15 0.60 0.36 0.19 -100 15 66 82.0 65.0 Weld metal 0.065 1.50 0.016 0.008 0.42 0.59 Material Heat Spec ID No. No. Supplier Austenitizing Tempering Stress relief Base metal A C5070-1 SA 533 CR.B LUKENS 1650-1700F, held I h/ 1200F. held 0.5 h/ 60 h at !!00-1150F CL. 1 in./ min and water in./ min and air cooled and furnace cooled below 600F quenched to 400F 4 Base metal B C5062-1 SA 533 CR.B LUKENS Same as above Same as above Sam'e as above CL. 1 Weld metal WF-193 NA NA NA NA Il00-1150F for 27.75 h. furnace-cooled N W R 3t* P I I E W

Table A-9. Davis-Besse 1 Description and Properties of Reactor Vessel Surveillance Program Materials Impact properties Chemical Composition, % T RT C -USE. UTS. YS, Material g. ID C Hn P S Si Ni Cr Ho Cu F F ft-lb ksi ksi Base metal A 0.22 0.63 0.011 0.011 0.27 0.81 0.32 0.63 0.02 50 50 !!8 91.6 71.4 Base metal 8 0.26 0.68 0.004 0.006 0.30 0.77 0.38 0.64 0.04 20 20 144 89.8 71.5 Weld metal 0.09 1.74 0.013 0.017 0.45 0.61 0.46 0.21 -20 2 81 81.0 NA Heat treatment Material Heat Spec ID No. _ No. Supplier Austenitizing Tempering Stress relief Base metal A SP4056 SA 508 CL.2 LADISH 1640F i 10F held at 1240F 1 10F held at 1125F 1 25F held at T color 4 h cold water color 6 h air cooled color 40 h. furnace 5 quenched. Reausten-cooled below 600F itized 1570F i 10F held at color 4 h cold water quenched Base metal B 123x244 SA 508 CL.2 LADISH Same as above 1240F 1 10F held at Same as above color 5 h then air cooled Weld metal WF-182-1 NA NA NA NA I100-1150F for.15 h. furnace-cooled K Rx P w m m m m m e uma e e mas

l l I. 1 i i I i ) I 1 I APPENDIX B Description and Properties of Research Capsule Program Materials i l I I l I I 3-1 Babcock & Wilcox l

Table B-1. Chemical Composition and Unirradiated Mechanical Properties of Beltline Region Weld Metals Impact properties Chemical composition. Z T RT C -USE. g. ET. MS YS. Ident No. C Mn P S Si Ni Cr Mo Cu F F ft-lb kai kai Weld metal-W1 0.09 1.63 0.018 0.009 0.54 0.59 0.11 0.40 0.42 85.5 69.0 Weld metal-W2 0.075 .l.50 0.024 0.006 0.60 0.58 0.51 0.22 -50 0 65 83.0 66.0 Weld metal-W3 0.08 1.45 0.016 0.016 0.51 0.59 0.09 0.38 0.21 31.0 NA Weld metal-W4 0.09 1.53 0.013 0.017 0.53 0.70 0.08 0.42 0.37 88.0 NA Weld metal-W5 0.09 1.58 0.015 0.016 0.54 0.67 0.09 0.42 0.35 Weld metal-W6 0.08 1.53 0.021 0.016 0.58 0.60 0.10 0.40 0.22 81.5 64.0 Weld metal-W8 0.09 1.55 0.014 0.015 0.55 0.70 0.08 0.41 0.35 Weld metal-W9 0.08 1.43 0.011 0.013 0.49 0.59 0.08 0.38 0.27 81.5 67.0 K Rx P n-e mas amm num

Table B-2. Description of Beltline Region 2nd Surveillance Weld Metals Filler metal Flux Welding Test qualification Ident No. type type procesa post-weld heat treatment Weld retal-W1 Mn,Mo,Ni Linde 80 Sub. are 48 h at 1100-1150F Weld metal-W2 Mn,Mo,Ni Linde 80 Sub, are 48 h at 1100-1150F Weld metal-W3 Mn,Mo,Ni Linde 80 Sub, are 80 h at 1100-1150F Weld metal-W4 Mn,Mo,Ni Linde 80 Sub. are 48 h at 1100-1150F Weld metal-W5 Mn,Mo,Ni Linde 80 Sub, arc Weld metal-W6 Mn Mo Ni Linde 80 Sub. are 48 h at 1100-1150F Weld metal-W8 Mn,Mo,Ni Linde 80 Sub. arc Weld metal-W9 Mn,Mo,Ni Linde 80 Sub. arc Eight 6-h cycles at 1100-1150F I I B-3 Babcock & Wilcox

l l 1 I. I l I { l 1, \\I APPENDIX C l 3 Description of Surveillance Capsule Test Specimens - Plant-Specific and Research Capsules I I I I I I c-1 Babcock & Wilcox

This appendix describes the tensile, Charpy V-notch, and compact fracture specimens included in the research capsules. 1. Tensile Specimens Two different sizes of tensile specisnens are used in the research capsules; both conform to the requirements of ASTM E8-69T. The Type A research capsules contain the standard size specimens with a gage length of 1.428 inches. The tensile specimens in the other four capsules (Type B) are smaller and fit in a Charpy specimen envelope. The gage length for the miniature tensile specimen is 0.840 inch. Figures C-1 and C-2 illustrate the standard and miniature size tensile specimens, respectively. 2. Charpy V-Notch Specimens The Charpy V-notch specimens conform to the requirements of ASTM E23-72 and are 0.394 inch square and 2.165 inches long. Figure C-3 describes the Charpy specimen.used. 3. Compact Fracture Specimens There are two configurations of compact fracture specimens: rectangular and round geometry. Two rectangular specimen sizes and one round specimen size were used. The configurations and sizes of specimens are described in the following sections. 3.1. Rectangular Compact Fracture Specimens The rectangular compact' fracture specimens were prepared in accordance with ASTM E 399-74. The specimen geometry is illustrated in Figure C-4. As illus-trated in the figure, the specimens were modified for measurement of load ver-sus load line displacement. Two sizes of this type of specimen are included. The specimen sizes (in terms of thickness) are 0.394 and 0.50 inch. The dimen-sions of these specimens are listed in Table C-1. 3.2. Round Compact Fracture Specimen When tha research capsules were designed, it was recognized that the round compact fracture specimen would make the most efficient use of the capsule volume. Figure C-5 illustrates the round compact fracture specimen with its corresponding dimensions. C-2 Babcock s.Wilcox

3.3. Side-Grooved Specimens As indicated in section 3.4, the thick specimens for the two research capsules at Crystal River 3 have been side-grooved. These are the 0.936-inch TRCT spec-imens. The geometry of the side grooves is shown in Figure C-6. The depth of the grooves is 10% of specimen thickness, with a total reduction of 20%. The angle and radius of the grooves are the same as for the notch of the Charpy specimens. The decision to side-groove the specimen was made based on the information generated by Shih, et al.' In general, the side grooves kept the crack front of the stable crack relatively straight. A large degree of crack tunneling was cbserved in the testing of non-side-grooved specimen. Shih found that the 25% total side-grooving (12.5% on each side) was sufficient for the tough ma-terials used for his davelopment. Shih tested 12.5, 25 and 50% total side-grooved specimens. The 12.5% side-grooved specimens did not completely sup-press the shear tip formation, and the 50% showed higher stable crack growth extension near the tip of the side-grooves than at the center of the specimen. For the materials of this program the 20% side-grooving was selected because it was believed to be adequate and also minimized' the reduction of the net ~ section thickness of the specimen (reducing J measuring capacity). The irra-diated welds are not expected to be as ductile as the material used by Shih, et al., in their studies. Side-grooving is also expected to affect the slope of the J versus Aa R-curves because of the straightening of the crack front, which affects the determination of Aa. The J-Aa curves determined with side-groove specimens are believed to be more representative of the extension of a crack on a thick-walled component. The side-grooves affect neither the deter-mination of J nor the slope of the J-Aa curve at very small Aa. Ic C-3 Babcock & Wilcox

a) Table C-1. Dimensions of Component Fracture Specimens Dimensions, in. Specimen Lead line to Thickness. Length.

Width, Load line Notch ID back face W B - W/2 1.25 W 1.2 W opening, D opening, N 0.394 TCT 0.788 0.394 0.985 0.945 0.100 0.064 0.50 TCT 1.000 0.50 1.25 1.20 0.150 0.064 1.00 TCT 2.000 1.00 2.50

~2.40 2.00 TCT 4.00 2.00 5.00 4.80 0.150 0.127 ("}The round compact fracture specimens are illustrated in Figure C-5. .? s e !r 8x 9* Ef O em men

M M -- ~ ~ ~ m Figure C-1. Standard Size Tensile Specimen - Used on Type A Capsules 4 11 1 = 1 4 64

f I[c (t h,

{ t h ~ l.750 .o o = - f.428t.005 - GAGE LENeTH, REF SPEctMEN IDENTIPCCATION APPEN To SOTH ENDS d_o RM __\\_ m ls ..,s o s o -l (-{- + t l- .szaf:80sa p ~ tn = .3571.oozW .437f.002W 080 R mat 2 PLACES k 9* lei w

.4 oo.( i 1 o i z. = s (4 o. s e e lusp o a M. C B [ e p } y T

ll r

no 'f d E m e m, m o o s U oH go

c. T oo.

o, c. oo I !d o ct tM s e +4 s c E. . c. 5 o. o. n e 4t +- m 8.N 9 5 a ic s a-e C i. l. p S R e l is .ll neT s ,o= e.o q e z o. u i S m + - m wa er 3 u g s. lot t

c. o a

+ i = n fH i M 9-6a = 1 2 i MS C oo e t w A D rug C 3 I M i FE F I T H MT xf a_ Ec N(*- OS S M0 rMb _ 7 [ Nf p, C P P SA XOOWP I=OOM na@ l

I ~ I ^ I 1.---, '/ l I I I 0 l i1i 2 ?- l L.. J [ I e if q ( a i I i E I- .c l t s 8 2 +' h 'I iI ) 1 ls e 3 .? t l I v a I I I u l Babcock & Wilcox c-7 I

I I Figure C-4. Rectangular Compact Fracture Specimen - Standard l Proportions and Modification for Measurement of 5 Displacement at Load Line 1.t v -* = I e-I I I ..v I = 55 w .at w cxA. \\ / / g (+ I + u I I a. Ltso.cc E l V U g i I 4 p I cu C-8 I

M, -- M.M M .. M MM Figure C-5. Round Compact Fracture Specimen - Dimensions and Modification for Hessurement of Displacement at Load Line e E.0902.084._ % ^ ~.9362.OeO t.010 " 395_.o00

2. 2* 3o' t O* 30' TYP

= !.585 l .atfpt.OeO M TYP yyp 4.005 f-.444 .00 0 Tusto F,. HOLES SEE DerAsL A +, p/ l .fso!.os s 357 ser I] smwew soewTeriam e APPLV TO SOTed SsDES 8&mm$~.i. c' w

t., _,,

bb z l TYP R 94 ELE S j n b 2.49 2 f.00 7. ~ l 8_s sj -- RE W TV9" E PLAct s -+ SEE DETAIL e' D - st7t.cos !:EE . e sO 19' l l 4 i I Z- / I .303 THsqu up W MNIFE E06E .005s4 Mag .005 M MAIL. LPLACES .g 5 DETAlt.- A DE. Tall - S DETAll-C i

I I Figure C-6. Geometry of Side Grooves for 0.936 TRCT ~ l i 0.108 I i 1 45' 45' Y I o.o / I 1 - _ _ __j l I w. I I I I I I Babcock a. Wilcox C-10 l

I E

I.

\\ 1 I t l APPENDIX D Program and Capsule Type Designations I i I I; I' I I D-1 Babcock & Wilcox I

i Table D-1. RVSP Capsule Types No. of specimens Material description Tensile Charpy Capsule Type I Weld metal 4 '8 HAZ, heat A, longitudinci 0 8 Baseline material plate Heat A, longitudinal 4 8 transverse 0 4 Correlation material 0 8 Total per capsule 8 36 I. Capsule Type II HAZ, heat B, longitudinal 4 10 Baseline material plate Heat B, longitudinal 4 10 g transverse 0 8 g Correlation material 0 8 Total per capsule 8 36 Capsule Type III Weld metal 2 12 ( Weld-HAZ l Heat A, transverse 0 12 m Heat B, transverse 0 6 i Base metal forgings Heat A, transverse 2 12 Heat B, transverse 0 6 Correlation material 0, j Total per capsulo 4 54 l D-2 Babcock & Wilcox

I I Table D-1. (Cont'd) No. of specimens Material description Tensile Charpy 0.5 TCT Capsule Type IV Weld metal 2 12 8 Weld-HAZ, heat A, transverse 0 12 0 I' Base metal, heat A, transverse 2 12 0 Total per capsule 4 36 8 Capsule Type V J Weld metal 2 12 HAZ, heat A, longitudinal. 0 12 Baseline material Heat A, longitudinal 0 9 transverse 2 12 Heat B, transverse 0 9 Total per capsule 4 54 Capsule Type VI Weld metal, longitudinal 2 12 Weld-HAZ Heat A, longitudinal 0 12 I Hoat B, longitudinal 0 6 y Baseline material Heat A, longitudinal 0 0 I transverse 2 12 Heat B, longitudinal 0 0 transverse 0 6 I Correlation material O_ _6, Total per capsule 4 54 I I I D-3 Babcock & Wilcox

I I Table D-2. Materials and Specimens in Surveillance Capsules of Oconee Unit 1 No. of specimens Material description Tensile Charpy Capsules OCI-A, -C, -E Weld metal, WF 112 4 8 HAZ Heat C3265-1, longitudinal 0 8 Baseline material plate Heat C3265-1, longitudinal 4 8 transverse 0 4 Correlation, HSST plate 02 0 8 Total per capsule 8 36 Capsules OCI-B, -E, -F HAZ Heat C2800-2, longitudinal 4 10 Baseline material plate Heat C2800-2, longitudinal 4 10 transverse 0 8 Correlation, HSST plate 02 0 8 Total per capsule 8 36 I I I I I I D-4 Babcock s. Wilcox

I I Table D-3. Materials and Specimens in Surveillance I Capsules of Oconee Unit 2 No. of specimens I Material description Tensile Charpy l Capsules OCII-A, -C, -E Weld metal, WF 209 4 8 HAZ I-Heat AAW163, longitudinal 0 8 i Baseline material plate Heat AAW163, longitudinal 4 8 I transverse 0 4 Correlation, HSST plate 0 8 Total per capsule 8 36 Capsules OCII-B, -D, -F HAZ 0 Heat AWG164, longitudinal Baaeline material plate B, Heat AWC164, longitudinal 4 10 transverse 0 8 Correlation HSST plate 02 0 8 Total per capsule 8 36 I I I I. I. I' e D-5 Babcock & Wilcox i 1

I-Table D-4. Materials and Specimens in Surveillance Capsules of Oconee Unit 3 No. of specimens Material description Tensile Charpy Capsules OCIII-A, -C, -E Weld metal, WF 209 2 12 HAZ Heat A ANK191, longitudinal 0 12 Baseline material Heat A ANK191, longitudinal 0 9 transverse 2 12 Heat B AWG192, transverse 0 9 Total per capsule 4 54 Capsules OCIII-B, -D, -F Weld metal WF 209 Longitudinal 2 12 Weld - HAZ Heat A ANK191, longitudinal 0 12 Heat B AWG192, longitudinal 0 6 Baseline material Heat A ANK191, longitudinal 0 0 transverse 2 12 Ileat 3 AUG192, longitudinal 0 0 transverse 0 6 Correlation HSST plate 02 0 6 Total per capsule 4 54 l I I D-6 Babcock & WilCOX

I I Table D-5. Materials and Specimens in Surveillance Capsules of Three Mile Island Unit 1 No. of specimens Material description Tensile Charpy Capsules TMI-1A, C, E Weld metal, WF 25 4 8 I HAZ Heat C-2789-2, longitudinal 0 8 1 Baseline material, plate 1 Heat C-2789-2, longitudinal 4 8 transverse O' 4 Correlation, HSST plate 02 0 8 Total per capsule 8 36 1 Capsules TMI-1B, D, F M r Heat C-3307-1, longitudinal 4 10 Baseline material, plate Heat C-3307-1, longitudinal 4 10 transverse 0 8 Correlation. HSST plate 02 0 8 Total per capsule 8 36 I I I I Ie D-7 Babcock & WilCCX I.

I I Table D-6. Materials and Specimens in Surveillance Capoules of Three Mile Island Unit 2 No. of specimens Material description Tensile Charpy 0.5 TCT Capsules TMI-2A, C, E Weld metal, WF 182-1 2 12 HAZ Heat C-1946-2, transverse 0 12 Heat C-1937-2, transverse 0 6 Base metal forging l Heat C-1946-2, transverse 2 12 Heat C-1937-2, transverse 0 6 m Correlation HSST plate 02 0 6 Total per capsule 4 54 Capsules TMI-2B, D, F Weld metal, WF 182-1 2 12 8 HAZ g Heat C-1946-2, transverse 0 12 0 E Base metal forging Heat C-1946-2, transverse 2 12 0 Total per capsule 4 36 8 I I I I D-8 Babcock & Wilcox I

'I I I Table D-7. Materials and Specimens in Survaillance Capsules of drystal River 3 No. of specimens Material description Tensile Charpy 0.5 TCT Capsules CR3-A, -C, -E Weld metal..WF 209 2 12 I' Weld-HAZ Heat C4344-1, transverse 0 12 l Heat C4344-2, transverse 0 6 Base metal forgings Heat C4344-1, transverse 2 12 Heat C4344-2, transverse 0 6 Correlation material 0 6 ( Total per capsule 4 54 Capsules CR3-B, -D, -F Weld metal WF 209 2 12 8 Weld-HAZ e j Heat C4344-1, transverse 0 12 0 Base metal I Heat C4344-1, transverse 2 12 0 Total per capsule 4 36 8 I-I- I: I 6 I' I I

Table D-8. Materials and Specimens in Surveillance Capsules of Arkansas Nuclear One, Unit 1 No. of specimens Material description Tensile Charpy Capsules ANI-A, -C, -E Weld metal, WF 193' 4 8 HAZ Heat C5114-1, longitudinal 0 8 Baseline material plate Heat C5114-1, longitudinal 4 8 transverse 0 4 Correlation, HSST plate 02 0 8 Total per capsule 8 36 Capsules ANI-B, -D, -F Heat C5114-2, longitudinal 4 10 Baseline material plate Heat C5114-2, longitudinal 4 10 transverse 0 8 Correlation, HSST plate 02 0 8 Total per capsule 8 36 I I I I I I D-10 Babcock & Wilcox

I I I Table D-9. Materials and Specimens in Surveillance Capsules of Rancho Seco Unit 1 No. of specimens I Material description Tensile Charpy 0.5 TCr Capsules RSI-A, -C, -E j Weld metal. WF 193 2 12 Weld-HAZ Heat C5062-1, transverse 0 12 heat C5070-1, transverse 0 6 j Base metal plate Heat C5062-1, transverse 2 12 Heat C5070-1, transverse 0 6 Correlation, HSST plate 02 0 6 Total per capsule 4 54 Capsules RSI-B, -D, -F Weld metal, WF 193 2 12 8 I Weld-HAZ Heat C5062-1, transverse 0 12 0 Base metal plate Heat C5062-1, transverse 2 12 0 1 Total per capsule 4 36 8 I, I I i. I I I D-11 Babcock & Wilc0X

I Table D-10. Materials and S;, -imens in Surveillance Capsules of Davis Besse Unit 1 No. of specimens Materials description Tensile Charpy 0.5 TCr Capsulen TEl-A, -C, -E Weld r.etal, WF 182-1 2 12 Weld-HAZ Heat SP4086, transverse 0 12 Heat 123x244, transverse 0 6 Base metal forgings Heat SP4086, transverse 2 12 3' Heat 123x244, transverse 0 6 E Correlation material 0 6 Total per capsule 4 54 Capsules TEl-B, -D, -F Weld metal, WF 182-1 2 12 8 Weld-HAZ Heat SP4086, traasverse 0 12 0 Base metal Heat SP4086, transverse 2_ E O Total per capsule 4 36 8 I I I I I I D-12 Babcock & Wilcox

I I l 'I I I I E I i E APPENDIX E References I 1 I i I I I I E-1 Babcock & WilCOX

1 L. E. Steele, G. W. Knighton, and U. Pocapovs Radiation Embrittlement of Pressure Vessels and Procedures for Limiting This Effect in Power Reactors, Nuclear Applications 4 4. April 1958, p 230. 2 L. E. Steele and J. R. Hawthorne, New Information on Neutron Embrittlement and Embrittlement Relf: ' - " Reactor Pressure Vessel Steels, NRLRep. 6160, Naval Research Laboratory (October 6, 1964), ASTM STP-380, Am. Nuc. Soc. Test Mater., Philadelphia, (1965) p 283. 3 L. E. Steele, Radiation Embrittlement of Reactor Pressure Vessels, Nucl. Eng. Des. 3, (1966) p 287. " J. R. Hawthorne, Jr. and L. E. Steele, Initial Evaluation of Metallurgical Variables as Possible Factors Controlling the Radiation Sensitivity of Structural Steels, NRL Rep. 6420, Naval Research Laboratory (September 29, 1966), ASTM STP-426, Am. Nuc. Soc. Test Mater., Philadelphia, (1967) p 534. s J. R. Hawthorne. E. Fortner, and S. P. Grant, Radiation Resistant Experi-mental Weld Metals for Advanced Reactor Vessel Steels, Weld. J. Research Supplement H 10, October 1970, p 453s.

• U. Potapova and J. R. Hawthorne. The Effect of Residual Elements on 550*F Irradiation Response of Selected Pressure Vessel Steels and Weldments NRL.

Rep. 6803, Naval Research Laboratory (November 22, 1968), Nuclear Applica-tions 6_ 1, January 1969, p 27. 7 L. E. Steele, The Influence of Composition on the Fracture Toughness of Commercial Nuclear Vessel Welds, Proc. 2nd Interamerican Conf. Materails Technology, Mexico City, Mexico August 1970.

• C. Z. Serpan, Jr., H. E. Watson, and J. R. Hawthorne Interaction of Neutron and Thermal Environmental Factors in the Embrittlement of Selected Struc-tural Alloys for Advanced Reactor Applications, Nucl. Eng. Des. H. April 1970, p 368.
• C. F. Shih, et al., " Methodology for Plastic Fracture," Final Report to EPRI, Conimet No. RP601-2, General Electric Company, Schenectady, New York.

August 1980. I E-2 Babcock & Wilcox}}