ML20006A865

Forwards Responses to Questions Re Portland General Electric 881024 Rept, Trojan Nuclear Plant Reactor Vessel Supports, Design Basis & Evaluation Summary, Raised During 890822 Meeting W/Nrc Re Resolution of Generic Safety Issue 15

ML20006A865
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	01/19/1990
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-015, REF-GTECI-NI, TASK-015, TASK-OR NUDOCS 9001300375
Download: ML20006A865 (22)

Text

{{#Wiki_filter:- t M tuum I WMW David W. Cockfield Vice President, Nuclear January 19, 1990 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555

Dear Sir:

Reactor Vecsel Support Information Request On August.22, 1989 a meeting was held at the Trojan Nuclear Plant between Portland General Electric Company.(PGE) and representatives of the Nuclear Regulatory Commission (NRC), which included Mr. Richard E. Johnson from the Engineering Issues Branch of the Office of Nuclear Regulatory Research. The purpose of.the meeting was to discuss the resolution of Generic Safety Issue (GSI) 15 regarding radiation embrittlement of reactor vessel supports. There n were several questions with respect to the PGE-report dated October 24, 1988, " Trojan Nuclear Plant Reactor Vessel Supports Design Basis, and Evaluation Summary" (PGE Report) for which additional information was not then availabic,- Attached are our responses to these questions. Sincerely, y Attachments c: Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission i Mr. David Stewart-Smith State of Oregon Department of Energy Mr. R. C. Barr NRC Resident Inspector Trojan Nuclear Plant 9001300375 900119 ADOCK0500g4 PDR p 121 S W Sa'non Stwt. Pvt:ard. Oteten 972CM 'It

f L l i I Trojan Nuclear Plant Document Control Desk Docket 50-344 January 19, 1990 License NPF-1 Attachment A Page 1 of 17 I l~ REACTOR VESSEL SUPPORTS INFORMATION REOUEST The following are portland General Electric Company's (PGE's) responses to questions-raised during the August 22, 1989 meeting to discuss the resolution of Generic Safety Issue (GSI) 15, which addresses radiation embrittlement of reactor vessel supports. Question 1 Are the materials of the pin and clevis carbon steel or alloy steel, and what are the consequencos with respect to them being lubricated with Molycoatt pGE Response The pin material is a low alloy steel American Society for Testing and Materials (ASTM) A"193-68 Type B-7, and the clevis material is a carbon steel ASTM A-36 (reference pGE Report, Section 3.1). Due to the unique dimensions of the pins, their material certification records could be identified in our files. Copies of these records are provided ab Attachment B. Traceability.of specific A-36 manufacturer's material certifications directly to the individual pieces of material from which specific clevises were fabricated, however, could not be achieved. The chemical composition of "Molycoat-Type B" dry flim lubricant specified for use on the pins (reference Bechtel Drawing C-368 Note 8) has still not been confirmed. The presence of a molybdenum disulfide-(MoS ) bearing lubricant in 2 contact with the pin and clovis assembly does not appear to be a concern based on the absence of the causative fr.ctors necessary for the occurrence of premature failure mechanisms auch as stress corrosion cracking. These factors include the presence of an electrolyte, hydrogen sulfide (H S) from MoS2 degradation, high 2 strength materials, and high applied stressen. The pin and clevis assembly is in a dry, 90 to 100 degrees Fahrenheit (*F) environment. Even if borated water contacted the assembly, H S generation does not occur at temperatures of 150*F or lower 2 according to testing reported in Electric Power Research Institute (EpRI) NP-3784 (Docember 1984). It was further noted in this report that support fastener failures had been confined to highly stressed low alloy steels, heat-treated to a minimum yield strength of 160 kilopounds (kips) per square inch (ksi). In the Trojan reactor vessel supports, the ASTM A-193 B7 pin (unthreaded in the region of

l l l I Trojan Nuclear Plant Document Control Desk i Docket 50-344 January 19, 1990 { License NPF-1 Attachment A i Page 2 of 17 maximum bending stress and without tensile preload) has a yield strength of 107 ksi and is subjected to a bending stress of 37.5 kel (35 percent of yield). Where failures have occurred in pressure boundary bolting heat-treated to less than 160 kai yield strength, there was routine exposure to the working fluid or to leaking borated water. Even so, according to an analysis of failure data presented in EPRI NP-5769. Volume 2 (April 1988), urdess assumptions were made about prior exposures to MoS2 lubricaats, there was no statistical evidence that Holykote C (a widely uved MoC2 bearing lubricant) had been associated with more stress co'trosion cracking than Felpro N-5000 (a widely used non-McS2 bearing lubricant). The presence of "Molycoat" in contact eith the reactor vessel support pin and clevis is not considered to represent a concern. Ouestion 2 Is the material of the lower support beam through which the grout-hole was flame cut susceptible to cracking induced by the flame cut or possible associated abusive grindingt j PGE Response The effect of flame cutting of the 4-inch diameter grout hole in the top flange of the reactor vessel support beam has been evaluated (reference PGE Report, Section 5.5.6). The Oxyfuel Gas Cutting (OFC) process produces an air-quenched heat-affected zone (HAZ) of 1/32 to 1/8-inch-deep in a 2.5-inch-thick plate of medium carbon steel, such as ASTM A-36. The slight hardening of the cut edges do not generally cause cracking. Even if initial cracks are assumed and can grow to 0.5 inches, which is unlikely in view of the results of the Robert L. Cloud and Associates (RLCA) fatigue analyses, the crack tip would be well outside the HAZ and would be governod in further growth by the base material properties'used in the analyses. Thereforo, the RLCA fracture mechanics analysis presented in the PGE Report bounds the conditions caused by the OFC process. The HAZ that could be caused by possible abusive grinding and the potential for crack initiation i is considered to be enveloped by that associated with the OFC process. Question 3 What is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code reference for using 6 ksi as the stress threshold in determining which potentially embrittled portions of the reactor vessel supports would be evaluated by fracture mechanics? di

i s LTrojan Nuclear Plant Document Control Desk . Docket 50-344 January 19, 1990 y License NPF-1 Attachment A Page 3 of 17 PGE Response The RLCA analysis screened the Trojan reactor vessel supports to - identify which potentially embrittled portions are subject to a ~ tensile stress greater than 6 ksi under seismic, accident, and normal operating conditions. Fracture mechanics evaluations were then performed for those portions (reference PGE Report, Section 5.2). 'he basis for'using 6 kai for screening the supports is found in Subsection NF of the ASME Boiler and Pressure Vessel Code, Section 3. Paragraph NF-2311(b)(7) exempts from impact (fracture toughness' 6esting "..... material for supports when the maximum stress 4 e. not exceed 6000 psi tension or is compressive". There-is little een.arn for fracture of support materials when tensile 7 stresses are' low or when stresses are compressive, and RLCA used the Code's position as the screening criteria for locating portions of the reretor vessel supports that are potential locations for subsequent fracturo mechanics analysis. A copy of NF-2311 is provided as Attachment C. Question 4 Provide a comparative discussion of the reactor vessel lower horizontal support beam bending stresses and moments determined from the Oak Ridge National Laboratory (ORNL), RLCA, and Bechtel analyses. PGE Response The predicted maximum stresces at the top surface of the reactor vessel support lower horizontal beam at the flange grout hole are quite similar for'the ORNL, RLCA, and Bechtel analyses, even though j there are differences in assumptions and modeling techniques. The differences are briefly summarized below 1 a. The ORNL analysis was based on a beam-on-elastic-foundation model in which support of the baam was modeled by a series of closely spaced vertical load carrying struts representing the distribution of support stiffneus (reference ORNL Report, Section 6.4). b. The RLCA analysis was based on a detailed finite element model (reference PGE Report, Section 5.4). The RLCA model accurately accounts for web shear deformations, and related cross-section warping, which causes the stress distribution over the beam. cross section to be highly nonlinear with a higher predicted surface stress.

Trojan' Nuclear Plant Document control Desk Docket 50-344 January 19, 1990 License NPF-1 Attachment ~A page 4 of 17 c. The Bechtel analysis was based on a beam-on-clastic-foundation model in which the support beam steel components and concrete prisms considered to be effective were analytically modeled as equivalent clastic springs. Deep beam effects were also considered (reference PGE Report, Section 4.4.3). The stress calculated by Bechtel includes a factor of.1.20 to reflect deep beam effects. The max 3 mum beam stresses predicted by ORNL at the grout hole are -based on the case where the concrete in front of the pedestal is assumed to be ineffective due to crushing. The RLCA and Bechtel analyses consider that this concrete remains intact for all load I cases. In comparing these analyses, it should also be noted that slightly different loads were used for the controlling load case; namely l .828 kips was used by RLCA and Bechtel and 779 kips was used by ORNL for the controlling load on a single beam. In the following stress comparison, the ORNL results have been retlood up to a load of 828 kips. 1 Stresses are nominal at the grout hole center line, with consideration of the hole, but without stress intensification. i Concrete Crushed Concrete Intact j ORNL 26.0 kai 21.6 ksi 25.1'kai 4 RLCA' .Bechtel 28.8 kai -{ As can be seen from the above comparison of stresses resulting from the ORNL. RLCA,~and Bechtel analyses, which used different analytical 1 models, calculated nominal flexural stresses at the centerline of the upper flange grout hole are in reasonable agreement. I l A comparison of the lower horizontal beam bending moments predicted from the ORNL, RLCA, and Bechtel analyses using the independent structural models is shown in Figure 1 (ORNL Report, Figure 6.7). [ These bending moment diagrams are also in reasonable agreement. ) In the fracture mechanics evaluations contained in the PGE report, beam stresses from the RLCA analysis were used because they aro based on what is considered to be the most representative analytical modeling of the beam.

m 1? Trojan Nuclear plant Document Control Desk Docket 50-344 January 19, 1990 License NPF-1 Attachment A Page 5 of 17 Ouestion 5 Determine if the liner piste embedment steel angle is welded to the reactor vessel lower horizontal supports. PGE Response The liner plate embedment steel angle was not designed to be and is b not welded to the reactor vessel lower horizontal supports. Question 6. Determine the feasibility of performing nondestructive examination (NDE) or physical sampling of the reactor vessel lower horizontal support beam. PCE Response General: I PGE-has made=a preliminary investigation of the feasibility of both an ultrasonic volumetric examination of the support beam upper flange and removal of a sample of the upper flange for subsequent laboratory fracture toughness testing. -Indirect access to and confined space at the location of the support beam upper flange and potential radiation exposures significantly diminish the practicality of either approach i as discussed further herein, j Access to the Lower llorizontal Support Beam: i Access to the lower horizontal support beam will be described with reference to Figures 2 through 5. Figure 2 shows a typical section at the location of a reactor vessel support. Access from below the support is essentially precluded by the lack of adequate clearance between the continuous horizontal tube steel section (Ts 8x4x3/8) at elevation 42 feet 10 inches and the reactor vessel insulation. The available clearance is only about 5-1/16 inches. Circuitous access may be available from above the support. Neutron shielding at elevation 67 feet 1-1/2 inches over the reactor cavity accident pressure relief ports, which is in place during power operations, is removed during refueling. As shown in the plan in Figure 3, a space about 1 foot 3 inches in width and 2 feet or more in length then exists between the edges of the pressure relief port and the reactor hot or cold leg piping insulation. This space could allow limited access from elevation 67 feet 1-1/2 inches to a ledge at elevation 57 feet 6 inches shown in the plan in Figure 4. The estimated l l

y 1 ~i f Trojan Nuc1' ear Plant Document Control-Desk -Docket 50-344 January 19, 1990 t License'NPF-1 -Attachment A' Page;6 of 17 4 radiation field at this ledge location is 250 to 300 millirem per hour. Below this ledge, the clearance between the reactor primary shield wall and the reactor vessel insulation is about 1 foot 10-1/16. inches.The top of the lower horizontal support beam is 9-feet 2 inches below the ledge. The minimum clearance between the } end of the support beam and reactor vessel insulation is about i 4-1/16 inches as shown in the plan in Figure 5. While access to the reactor vessel lower support beam may be possible as described above, staging to provide for personnel safety (ladders and scaffolding, illumination, communications, dosimetry, cooling, etc.) and test or sampling equipment support would be very difficult in the confined spaces available. Staging, testing or sampling, and disassembly would also involve substantial radiation exposures to personnel and could require an extended refueling outage for performance of the work. Remote controllable test or sampling. equipment staged-from the ledge at elevation 57 feet 6 inches and controlled from the slab at elevation 67 feet 1-1/2 inches would reduce the concerns for personnel safety and unnecessary radiation-exposure but would result in considerable overall program complexity l gi and cost. Ultrasonic Examination: The objective of ultrasonic volumetric examinations would be to determine if a preexistent flaw at a critical orientation exists in the region of the upper flange flame cut hole. Based upon -discussions with specialists in ultrasonic examination techniques, the test equipment transducer (2.25 megahertz straight beam) should be coupled to the end of a representative beam upper flange and moved in increments across-the width and thickness of the flange to provide a complete scan of the flame cut hole region (reference Figure 2). The examination should be capable of detecting a flaw size on the order of 1/8-inch-deep by.1/2-inch-long at right angles to the beam. The condition (roughness) of the end of the flange is not known and adequate coupling of the transducer may introduce the complication of required surface preparation or the addition of a coupling medium. While ultrasonic volumetric examination of the upper flange flame cut hole region may be valid in principle, mockup testing would be required to validate the process. Physical Sampling and Testing: Regarding the concept of obtaining a physical sample from a representative support beam upper flange, PGE has retained a consultant to advise us on the size of sample needed to perform meaningful metallurgical tests and the types of tests recommended.

Trojan Nuclear plant-Document Control' Desk Docket 50-344 January' 19,'1990 License NPF-1 Attachment A page 7 of 17 Specialists who have the technology to cut such samples without introducing undesirable metallurgical effects have also been interviewed. e ( The largest sample of a representative support beam upper flange that could reasonably be obtained without affecting the support beam structure would be from the end of the beam nearest the reactor vessel and would have dimensions of 16 inches in' width, 2-1/2 inches in thickness.. and about 1-1/2 inches in length along the flange. The present activity of such a sample would be about 2.3 rem per hour on contact end about 35 millirem per hour at 18 inches from the sample. The sample configuration and test specimen blanks are shown in Figure 6. The recommended test matrix is shown in Figure 7. Test program objectives would be to: 1. Determine the fracture toughness at the operating temperature of the reactor vessel support material. 2. Determine other properties of the reactor vessel support

material which are of importance in determination of the potential for failure of the. vessel-supports and for indirect determination of the material's fracture toughness.

These include: a. Tensile properties (yleid strength, tensile strength, fracture strength,- ductility properties,-and strain hardening characteristics) over a range of temperatures corresponding with the lower-shelf, transition region, and upper-shelf temperatures. b. Charpy V-notch impact properties (absorbed energy and fracture modes) versus temperature. This should include the determination of the lower-shelf, transition region, -and upper-shelf properties; the nil-ductility transition temperature; and the fracture appearance transition temperature. c. Microstructure including general microstructure, grain size, and microstructural abnormalities. d. Chemical composition. As is apparent, the recommended testing would involve a substantial program.

p V~ Trojan Nuclear plant Document Control Desk - Docket 50-344: January 19, 1990 License NPF-1 Attachment A page 8 of 17 Removal of the representative sample of the support-beam upper-flange would require a vertical cut downward through the 2-1/2-inch-thick and 16-inch-wide flange and horizontal cuts (parallel to the beam axis)~through the two 1-3/4-inch-thick beam webs and along the 1/4-inch-thick end plate. There are several cutting techniques available which may, in principle, be suitable for removal of the -sample without significantly affecting its metallurgical characteristics. One such method would be use of an ultrahigh pressure abrasive water jet cutter. In this process, fine abrcsive j particles are rootered into an ultrahigh pressure water jet through a j special' nozzle assembly having a small exit orifice. This high 1 velocity abrasive water jet removes material from the workpiece in a controlled fashion and produces the cutting. The ebrasive used is usually garnet and the water jet pressure is on the order of 35,000 psig. For 2-1/2-inch-thick carbon steel (thickness of the upper-flange) the nozzle orifice would be about 0.022 inches in diameter J the water jet flow rate would be about 2.8 gallons per . minute, and the maximum cutting rate would be on the order of 2 inches per minute. The total cutting time estimated to be required is approximately 10 minutes with approximately 30 gallons of water used in:the process. Although the cutter nozzle assembly is j reasonably compact, the water and abrasive feed hoses create additional clearance requirements. For either a manual or remote [ control application, a track system to accurately guide-the cutter in the; required positions would be necessary. In a more conventional cutting application, the abrasive water jet would be trapped and _ -dissipated in a catcher-as it exits the material being cut. In the i case of the beam flange sample, for the vertical cut the jet will ~ exit against the webs or concrete infill in the middle portion of the beam, and for the horizontal cuts the jet will exit against the concrete infill. The reactor vessel insulation, the pin and clovises would, of course, have to be protected from the abrasive water jet j splashback. Splashback.would also affect visual control of the cutter for either a direct or remote-controlled operation. Other cutting methods that may be feasible would be the electrical discharge machining (EDM) or metal disintegration machining (MDM) processes. A specially designed cutting head would be delivered to the top flange of the beam with a remote-controlled device. pneumatic or hydraulic clamps would be used to fix the head to the beam. These clamps would be integral with the cutting head. Both the EDM and MDM processes require a dielectric cutting modium which in this case would be either deionized or demineralized water. A flow rate of 5 gallons per minute would be required and the cutting time would be on the order of 2 hours. The water would, of course, j have to be collected by some means. l

u L Trojan Nuclear plant Document Control Desk l Docket 50-344 January 19 1990 License NpF-1 Attachment A Page 9 of 17 3,; Summary: This response was intended to provide an initial evaluation with l respect to the concepts of performing ultrasonic examinations or physical sampling of a representative reactor vessel support lower horizontal beam top flange for subsequent metallurgical testing. It is apparent to us that any investigative effort requiring access to the lower reactor vessel support bee.m would require very careful i planning (including the use of mockups). considerable effort and expense, significant radiation exposure to personnol, and possible n extension of a refueling outage. These concepts must, therefore, be carefully evaluated against alternativo means of resolution of Generic Safety Issue 15. l The 'RLCA fracture mechanics analysis of the Trojan reactor-supports was presented in PGE Report, Section 5. We.believe that this I analysis demonstrated acceptable factors of safety against support-structure fracture enhanced by neutron embrittlement. There are j efforts underway which should result in significantly increasing l these factors of-safety. pCE presently has a leak-before-break (LBB) - analysis of the primary reactor coolant loop piping which has been approved by the Nuclear Regulatory Commission (NRC). The associated reductions-of effects on reactor vessel support-loads are reflected-in the present RLCA analysis, pCE is also pursuing LBB analysis on reactor coolant loop auxiliary piping. If pGE's LBB analysis on the auxiliary piping (residual heat removal,-accumulator, and surge lines) is approved, the present loss-of-coolant accident (LOCA)' j vertical load on an individual support (two-horizontal beams) of 740 ' kips (reference pCE Report, Table 4-1) would be reduced to an I Lf estimated value of 220 kips. The total load on a reactor vessel support based on the governing load combination of dead weight + thermal + SRSS [ safe shutdown earthquake (SSE)+LOCA) (reference pGE Report, Section 4.4.2) would become 1296 kips in lieu of the present value of 1656 kips. A 30 percent reduction in load and, thus, a 30 percent increase in factors of safety would result. In the RLCA analysis, the neutron flux and fluence exposure of the reactor vessel supports at 32 effective full power years of operation was conservatively estimated based on available data which was not Trojan specific. A dosimetry experiment is presently planned by the National Institute of Standards and Technology (NIST) for the NRC, in cooperation with PGE, to measure the broad-band energy level neutron fluence exposure of the Trojan reactor vessel supports over one operating cycle. Data obtained from the NIST experiment should be evaluated with respect to the neutron exposure that was assumed in the RLCA and ORNL fracture mechanics analyses. Results from the NIST

'_p, L _ Trojan Nuclear plant Document Control Desk -Docket 50-344 January 19,-1990 License NPF-1 Attachment-A Page 10 of 17 dosimetry experiment are expected to be available in late 1990 if the. experiment is carried out as presently planned. Recently questions have been brought to bear.on the interpretation of the ORNL High Flux Isotope Reactor fHFIR) surveillance results. An overview of the HFIR data and test results from thu Shippingport reactor shield tank are summarized in Westinghouse publication WCAP-12345, Revision 1, which is provided as Attachment D. In this l WCAPJit is stated that the HFIR results should not be. considered-typical of those for a power reactor. The reactor. support = integrity issue at low fluences should be questioned and the concern may not be valid. 4 In view of the considerations summarized above, we do not feel that. the encumbrances' associated with nandestructive or destructive examination of the reactor vessel support lower horizontal beam are warranted at this time. We will continue to stay abreast of developments associated with Generic Safety Issue 15 and actively seek resolution. ~LGD/bsh 3807W.0109. l 1 l s

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//// / / / / / ///) O O O O O O 15" f \\ O O O O O O xmxmxxxxxxxxxxxxxxxxxxxw/o CUT BLANKS FOR CHARPY V-NOTCH SPECIMENS 7 BLANKS FOR TENSILE SPECIMENS e D1 mansions and Orientation of Test Block and Layout of Blanks for Test Specimens. Note that specimen blanks are oriented such that fractures in test specimens machined from those blanks vill be the same as the probable orientation of the fracture in the Vessel Support if it failed. 4 Figure 6 -. ~ ~ - - " ~ ~ ' ~ ~ ~ ~ ~ " ' ' ' ~ ~

I Trojtn Nuclear Plent Documsnt Control Dask [ Dockst 50-344' . January 19, 1990 'Liesnst NPF-1 Attachmant A Page 17 of 17 l The tentative test antr'ix'is shown in Table 1. The test temperatures are shown only for purposes of illustration, and they are likely to - l-change somewhat aince they vin' he chosen as work progresses.: Similarly, the actual choice of strain rates any differ from those shown. i e Table 1 y Tentative Test Matrir* Test Charov Tensile at t J Tests at t Blunt Fatigue 7 Temo.. 'T V-notch.001 A &.001 & 0.10 j!2Eib. Crack Growth -250 I II -200 I I I -150 II I I I -100 II I I I II - 50 II I II I I 0 II I I .I I 50 I .I i I 75 I I I I l 100' x I ~

• Blanks for chemical analysis and metallographic examinations are not included in this table but will be taken from broken portions of other test specimens.

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= ~ ~ " = ' " ~ ~ ~ \\ Trojan Nuclear Plant Document Control Desk i Dockst 50-344 January 19, 1990 Licanse NPF-1 Attachment B Page 3 of 3 .i 4 NIETEATION OF HEAT-TREATMENT j SCHMITT STEEL INC. 2765 N. W. NICOLAl. PORTLAND. OREGON 97210 i CUSTOMER 78 Md. M"/[ N DATE CUSTOMER'S ORDER NO. /daga3f JOB. NO. ND PART NO. I 8/ /.2k Mu d* CUANTITY M MATERIAL SPEC. /40 M/ 4 "/ HEAT-TREATED PER: Aff /ff[ 4 6! jA;de 9tN5 0D A&Ml1 //2.soG ~ Nud[?2.) I HARDNESS TEST- %/n I REQUIRED PERCENT INSPECTED RESULT * / 27)P l WE CERTIFY THAT THE HEAT-TREATMENT ABOVE IS TRUE AN g., r, AND THAT THE TEMPERATURES AND TEST RESULTS WERE OBTAIN STANDARD APPROVED METHODS. r E.6# ce.TI, led Bv ,u ( FORM 130 MC AT TREAT gypg AVI SOR e O N "J ~ .0 3* %p D. ~ g }' ~-

L. l Trojan' Nuclear Plant Document Control Desk [ Dockat 50-344-January 19, 1990 Licensa.NPF-1 Attachment C, 1 Page 1 of 1 I 1986 Edities NF.2000 - MATERIAL NF.2226.4-NF.2321 specimens shall be at least h in. from any heat treated (4) material for 8ttings with all pipe connections surface and the midlength of the specimens shall be at of % in, nominal wall thickness and less-least 1% in from any second heat treated surface. The (J) austenitic stainless steels; F CertiAcate Holder shall specify the surfaces of the (6) nonferrous materials; i Amshed product subjected to high tensile stresses in (7) material for supports when the maximum service. stress does not exceed 6000 psi tension or is comptes. sive; (8) rolled structural shapes, when the thickness NF 2227 ' Rolled Shapes of a Sange is */ in, or less; For rolled shapes, the coupons shall be taken so that (9) materials for Class 1, 2, or MC supports, q specimens shall have their longitudinal axes on a line Heted in Table W.231MI, for hans 24 in. representing the center of the thickest element of the and less when se lowest ne tempuaW ts at shape and with the midlength of the specimen at least 307 abow the tabulaW tempwatum. nis ifrom a heat treated and.- 888mpdon from impact testing does not apply to either the weld metal (NF.2430) or the weld proce-dure qualiacation (NF 4335). NF.2300 FRACTURE TOUGHNESS (10) materials for Class 3 supports, listed in Table REQUIREMENTS FOR W 231 $ 1, for sh 24 in. and M when g the lowest service temperature is equal to or greater than the tabulated temperature. This exemption from NF.2310 MATERIAL TO BE IMPACT impact testing does not apply to either the weld metal TESTED (NF 2430) or the weld procedure quah8 cation (NF. p 2311 seppen for meh Impact Tudog of 4335)jl) matwials for Class 2 supierts for w L Material Is Required 8 g lowest service temperature exceeds 150'F; (a) Support matenals shall be impact tested in (12)matenals for Class 3 suppons for which the accordance with the requirements listed below. lowest service temperature exceeds 100'F. A87 (1) Integral attachments to the component or (c) The Design S =riwon (NCA 3250) shall -l t piping (NF 1222) shall meet the requirements for state the lowest service temperature (LST) for the impact testing stipulated for such components or piping in the applicable Subsection. component support and the designated impact test temperature, when required. (2) Class I component supports shall mest the requirements of NF.2300. t A8e (J) For Class 1 piping supports, Class I standard NF.2320 IMPACT TEST PROCEDURES supports, and all other types and Classes of supporta, the Desip Speci$ cation (NCA 3250) shall state NF 2321 Charpy V Notch Tests whether or not impact testing is required for the j material of which the support is constructed. When The Charpy V. notch test (C,), when required, shall impact testing is required, the tests shall become a be performed in accordance with SA 370. Specunens 'I requirement of this Subsection. shall be in accordance with SA 370, Fig.11 Type A. (b) The requirements for supports shall be as A test shall consist of a set of three full size 10 mm X speciAed in NF-2300, except that the matenals de-10 nun pens. He IM exp and absorbed _. scribed in (1) through (12) below are not to be unpact energy, as applicable, and the test temperature, as well tested as a requirement of this Subsection: as the orientation and location of all tesu performed to (1) material with a nominal section thickness of moet the requirements of NF.2330 shall be reported in ,e m. and less; the Certaaed MaterialTest Report. (2) botting, including studs, nuts, and bolts, with a nommalsize of1 in and less; (J) bars with a nominal cross > sectional area of 1 D' 8'" ""'F" GE " d" *""""" '""P"" sq in. and less; wisch will be maistaaned inside the contanasset venel duttes the pies: opersoon (for suppens within containmensk or aisarnaavely. 'When import tasons is required, the methods of Appedas G unny the *=lala'a4 or measured maassnum metal tasaperamre of the be used as an attemative desasa procedure for assurms pronmetaan empport espected durins norinal operstaan when-the pamoure withm the component escends 20% of the syssent hydressatac test asainst nooductde fracture. presure. 21

lC }' ~__ } Trajcn Nucloer Plant _ Document Control Desk. Dscket 50-344 January 19, 1990 + Licsnes NPF-1l_ Attachmant D l 31 pages ^ i WCAP 12345 REVISION 1 _t WESTINGHOUSE OWNERS GROUP IRRADIATION-EFFECTS ON REACTOR VESSEL SUPPORTS E!. W. H. Bamford E. R. Johnson R 4_ October, 1989 1 9 Verified by: I kem' Verified by: I. ~S. A. SFamy' H. L. Ott StructuM Materials ' Civil & Structural Engineering Engineering Approved by:

1. NW Approved by:

C fa S.S/Palusamy, Manager S. A. Palm, Manager = Structural Materials Civil & Structural Engineering Engineering Work Performed Under Shop Order MUHP-4001 WESTINGHOUSE ELECTRIC CORPORATION I Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 L[hk bh(

-s ' Y', 1 TABLE OF CONTENTS .Section Title Page 1,0 INTRODUCTION 11 2.0 . IRRADIATION DAMAGE ASSESSMENT 2-1 3.0. FRACTURE ANALYSIS CRITIQUE 3-1 l 4.0 REVIEW 0F SUPPORTS CONFIGURATION FOR WESTINGHOUSE PLANTS 4 '1 5.0

SUMMARY

AND CONCLUSIONS. 5 ; - 6.0 ' REFERENCES 6-1 i APPENDIX A REVIEW 0F REACTOR VESSEL SUPPORTS SURVEY A-1 4 j l 1 k 1 l- ' jl ' 11 D A

'I =t 9

1.0 INTRODUCTION

1 s The Westinghouse Owner's Group has received technical information concerning the potential for damage to reactor' vessel supports by low q i . flux, low fluence irradiation. This summary of our view of the issue has !I been prepared from the proceedings of a meeting held with the USNRC on June 15, 1989, e This report contains an independent assessment of the reactor vessel support irradiation damage issue, and a critique of the study performed I earlier by Oak Ridge National Labs Ill. A review of the range of support configurations which exist in Westinghouse plants is provided, based on a survey conducted among all the Westinghouse plants. Also included is a summary.of the key conservatisms in the integrity studies performed to date. 1 l i 'j. i 1 1 1 'l I 1-1 .e P l .' y

.~ 2.0 IRRADIATION DAMAGE ASSESSMENT ] l . This issue of irradiation damage at low temperature and low fluence'has been considered inconsequential since the middle 1960's, when research was redirected to higher fluence and higher temperature conditions. This issue arose again recently with the publication of the ORNL HFIR reactor surveillance resultsIII. In this section a summary of all available data will be'given, along with a fresh look at the HFIR results. The best summary of available data is that compiled by Porter [2]. f U.S. o ~ Steel, and his results are summarized in Figure 2-1. He surveyed the literature, and did a statistical study, constructing tolerance limits on the available data. Note that the slashed-points are the only data irradiated at higher than 250'F. Porter's data included both carbon and -l low alloy steels, and associated welds, and a partial list of materials is shown in table 2-1. Recent. test results from the Shippingport shield tank material provide.a good assessment of 'the level of irradiation damage which might be expected to occur in a typical rector vessel support system located immediately 1 opposite the vessel core region. These resultsI33 are summarized in table 2-2, and show an irradiation-induced shift of a maximum of 52F, with a dpa dosage of 0.00167, and fluence = 6.lE17. Plotting these results on Porters data, we find that they fit directly on the mean curve, as shown on ' Figure 2-2. Note that both 15 and 30 ft.lb. shifts were used for completeness, since the 15 ft.lb shift is often used for carbon steels. -1 The surveillance results from the HFIR reactor are shown in Figure 2-3. Note that the data all still fit within the 75-90 tolerance bounds developed by Porter. Looking at the data by themselves shows a different .y slope than Porters, but there appears to be a good reason for such a disparity. The HFIR reactor contains a thick beryllium reflector which I prevents a large percentage of the high energy neutrons from reaching the l' tank wall. Therefore the dpa is a much more accurate measure of total 2-1 L =

irradiation exposure. The Shippingport and HFIR results compare better p based on dpa, as seen in Figure 2-4, but the comparison is marred by the ORNL calculation, which considered only er.ergies greater than 0.1 MeV. This calculation s'ould be redone to consider all energies, as the n Shippingport calculations have done. An example of the energies missed in the ORNL calculation is given in table 2-3. Figure 2-5 shows the HFIR results in terms of both fluence (> 1 Mev), and dpa-(>0.1 Nev) and illustrates the different slopes which result. l The'HFIR results are not typical of those for a power reactor, because the j e'nergy spectrum has such a low proportion of low energy neutrons. It is our belief that use of these results is misleading for the vessel support issue. The more appropriate data are from Shippingport, and they match Porter's data very well. Porters data base was used in earlier years to conclude that no integrity issue exists fur low fluences, and the same conclusion appears to hold today. The fluence calculations need to be redone for the HFIR reactor, and the whole issue reassessed at that time. I 4 2-2

g y .t= o TARE 24 MATERIALS INCLUDED IN PORTERS DATABASE ASTM A106 ASTM A201 1 ASTM A212B (HOT ROLLED; NORMALIZED) J E7016 WELD i AISI C 1019 ASTM A203 ASTM A285 (HOT ROLLED; NORMALIZED) ASTM A293 ' ASTM A3018 (NORMALIZED, ANNEALED) ASTM A3028, ASSOCIATED WELDS ASTM.A336 ASTM A353 HY 65 HY-80 E 10016 WELD USS."T-1" STEEL E 12016 WELD DUCOL W30, ASSOCIATED WELDS 2.25 CR - 1 MO Ni--Mo-Cu-V (AUSTEMPERED; QUENCHED) 1 2-3 .......~.. ..-- 1 J

s. TABLE 2-2 SHIPPIN6 PORT RESULTS USED 15 FT-LB A m 30 FT-LB SHIFTS l ? j 30 FT-LB15 FT-LB ] l t0cAn0n etHEncE Op. snirT snIrv 3,9,2,8 WR6.10 x 10II 0.00167 52F-45F INER HALF, IMER WALL (3/89 PROG RPT.) ~ ^ 3,9,2,8 WR2.03 x 10I7 0.00056 40F 38F OUTER. HALF, IMER WALL (3/89. PROG RPT.) 3,9,2,8 TR6.10 x 1017 0.00167 20 30 INNER HALF, INNER WALL (4/89 PROG. RPT.) l 17 3,9,2,8 TR2.03 x 10 0.00056 30 30 OUTER HALF, IMER WALL (4/89 PROG. RPT.) l m

a TABE 2-3 I DAMAGE CROSS SECTIONS BY GROUP i i LOWER DAMAGE CROSS SECTION (B) ) ENERGY GROUP LIMIT q n n, av l 1 7.79 MeV 1374 6895 456.2 1826 .2 6.07 1365 5743 537.9 1726 1 3 4.72 1261 4633 598.7 1682 4 3.68 1847 4248 641.5 1551 5 2.86 1799 3583 709.6 1340 i 6 2.23 1531 2696 744.8 1188-7 1.74 1308 2054 594.1 916.5 8 1.35 1114 1573 469.8 684.8 9 1.05 826 1095 347.0 471.0 10 0.821 683 789 239.8 289.3 11 0.639 843 843 285.2 290.2 12 0.498 535 535 163.7 167.3 13 0.387 729 729 272.1 274.6 14 0.302 402 402 109.2 111.0 -15 0.235 327 327 118.7 119.9 16 0.183 268 268 148.3 149.2 17 0.143 264 264 82.57 83.32 18 0.111 176 176 32.77 33.42 19 86.5. kev 192 192 91.52 92.08 20 67.4 138 138 44.35 44.83 21-40.9 115 115 55.59 55.99 22 24.8 350 350 64.65 04.90 23 15.0 21.8 21.8 5.851 5.890 90T mm 24 9.12 31.6 31.6 13.95 13.95 TO DATE 25 5.53 28.1 28.1 19.45 19.45 FOR HFIR 26 3.35 11.2 11.2 7.757 7.757 27 2.03 6.7 6.7 5.714 5.714 28 1.23 4.2 4.2 4.691 4.691 29 0.749 1.7 1.7 2-5 ~

/ L FIGURE 2-1 400 Solid Points were Not Used in ,/ - the Statisticci Anolysis a g Substre Contilever-Type Impoet Specirrens / j l g a Chorpy V Noten impact Specimens f f

• Chorpy V-Notch Slow Bend Specimera. /.

f, j . [ 300 Moon / - i 75 per cent Toleennee Limits *f

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i. FIGURE 2-2 400 7 I Solid Points were Not Used in ,/ I the Statistical Anolysis ( g . Subsite Contilever Type impoet Scocir ens # f g a Chorpy V Notch impact Specimens f f . Chorpy V*No,ch Slow Send Specimera. /,

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• o SHIPPINGPORT SHIELD TANK RESULTS NOTE:

LOWER POINT IS 15 FT-LB SHIFT, UPPER IS 30 FT-LB SHIFT i EFFECT OF NEUTRON RADIATION ON THE NOTCH TOUGHNESS OF i CARBON AND ALLOY STEELS IRRADIATED BELOW 500 F: PORTER (SLASHED POINTS IRRADIATED ABOVE 250 F) 2-7 iR.-

i i 4 I I FIGURE 2-3 l I l sco. /- 180 ,4 MPIR SURVtlLLANCE; / $ A3129 (e e 2.4 I 104 vom!.ti e A19811 (o e 8 7 I 108) 290 " 9 ASSO LF3 to e i.61.41100) j ,ce-( l -( CART U I 9 MPIR Atitt (e s 9.4 X 1988) { .. mumni ..m e 1. a WTR 1 93.e,2:... > ge t

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FIGURE 2-4 400 t Solid Points were Not Used in ,/ the Stolistical Analysis t g . Subsite Contilever Type impoet SpeciEns,# j g, a Chorpy V Notch impoet Specimens a f f Chorpy V-Notch $ low Benc Specimeria. I I 300 Moon .,/,4 ;p, 75 per cent Tolerance Limits f- / oi is per cent Cont eene. g,,/. s i h

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l p FIGURE 2-5 I I l effe DATA tothas Cuev8 see&wtesosses) a le 4E e i teeWI e i X tela num. 4 letA SURVinLLA46C4 ise 6 e allt 8 ($. I 4 8 48 3 360

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....it i e e i....i e 10** i t*3 10 3 ese (8 m 8.1 beeV) COMPARISON OF HFIR SURVEILLANCE DATA: FLUENCE > 1 Mev vs dpa 2-10 ~ A_

3,0 FRACTURE ANALYSIS CRITIQUE The fracture analysis performed by theverton et.al.[1] appears to be technically sound, and there are only a few items worthy of discussion. The strain rate effect on toughness is a real effect, but the use of strain rates equivalent to impact loads seems unrealistic for a seismic event. 9 The strain rate of 0.1 inch / inch per second is a realistic upper bound. I The loadings used are upper bound design loadings, and are therefore very conservative. Realistic loadings have not been calculated for the Trojan j supports. Elimination of the need to corsider large break loca loads for j Turkey Point by utilizing leak before break could significantly increase the critical crack size for this plant to the point that a concern would not exist. As stated in the ORNL report, low cycle fatigue is not viable mechanism for creation of flaws on the order of the critical crack size calculated, thus such flaws would have to exist at the time of fabrication. ORNL further stated that at the two locations considered for the Trojan supports other than the grout hole, the critical crack size is the full width of the flange (16 inches) and that a flaw of this size would be readily detected during fabrication. The credibility of the existance of a flaw of critical crack size magnitude relies on the validity of the assumption that such a 6 flaw exists at the flame cut grout hole. We believe that the existance of a 0.4 inch flaw at the grout hole to be unlikely especially, as was mentioned in the ORNL report, since the flame cut hole was dressed. The j grinding operation employed during the dressing operation would lower the potential for any pre-existing flaws of critical crack size magnitude. The i net effect is that the likelihood of cracks anywhere in the support i configuration is very low, either from fabrication or service. A best estimate of the NDT shift in the supports is SOF, but much higher values were used in the analysisIll. A more realistic assessment, including use of leak before break to reduce the postulated loads, and use of best estimate instead of bounding loads, would lead to no integrity concerns, eveh at the governing plants. 3-1 l

I , ( 4 4.0 REVIEW 0F SUPPORTS CONFIGURATIONS FOR WESTINGHOUSE PLANTS 3 A review was made of the support configurations for all Westinghouse f plants, and a few discrepancies were found. These are listed in table i 4-1. The dimensions of the various support configurations are listed in { table 4 2, which refers to the geometry shown in Figures 41, 4 2 and 4-3. The stresses for all support configurations were reviewed, and results j showed that vertical tension loads existed on only three plants: Trojan and Turkey Point Units 3 and 4. All other plants have supports loaded entirely in compression during normal operation. Therefore, the key plants identified in reference [1] are correct. It should be emphasized that the i loads available for all support configurations are upper bound faulted loads used for design type calculations, and these loads were not intended to be used in integrity calculations such as those in reference [1]. ) l a, l' t F P 4-1 .-.-d. -- u

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TABLE 4 -1 REACTOR VESSEL SUPPORT l CONFIGURATIONS COMPARISON l j t NUREG DATA M DATA

• l San Onofre Short column?

Long column Unit 1 Point Beach Short column Ring girder / Long column Units 1&2 Seabrook Ring girder Short column Units 1&2 on concrete S. Harris Ring girder Short column Units Ifi2 on concrete Prairie IslandLong column Short column Units 1&2 -ALL OTHER PLANT DATA ARE IN AGREEMENT-OPreliminary data, to be confirmed by each utility. L l l 4-2

---

Y-'

- -.-.~- TABLE 4-2 i DIMENSIONAL COMPARISON OF SUPPORT CONFIGURATION DESIGN VERTICAL (1) HORIZONTAL (2) DIMENSION DIMENSION (INCHES) (INCHES) TYPICAL SHORT COLUMNS (3) +26 7 TO 9 WITH BOX TYPICAL BOX (N0 COLUMNS) (4). -4 TO 2 7 TROJAN (NUREG-CR-5320) +94 26 (COLUMNSONCANTILEVERBEAMS) l TURKEY POINT (NUREG-CR-5320) +28 26 lji I i i (1) FROM TOP OF CORE DOWNWARD TO BOTTOM OF SUPPORT ,iij (2) FROM VESSEL OUTSIDE DIAMETER RADIALLY OUTWARD T d (3) SURVEY OF EIGHT PLANTS WITH COMMON CONFIGURATION m f (4) SURVEY OF TEN PLANTS WITH COMMON CONFIGURATION i 1 43

L l 9 FIGURE 4-1 REACTOR VESSEL SUPPORTS / FUEL CONFIGURATION l s g 1r N 1 REL SUPPORT \\ s 'l c H e 4-4 ,t i t

FIGURE 4-2 .j tSYM +Z ^ TO REACTOR +X +Y OUT OF PAPER 1 9 9 +++ VIEW b w a wt ww ma cuumv sNN NW 4 s l

c= =m p

i s YM LTllllE $+Y s f: 9 +Z +X INTO PAPER - i l / k sectiew ~ l AIR C001.E0 Rf. ACTOR SUPPORTS 4-5 l

1 1 1 O FIGURE 4-3 \\ V u a m / ./ 9 / e e a m g g a +Y +)( i la I l +2 INTO l ,PAPED 9 4 (W AIR C001.ED RfACTOR SUPPOR% f e 4-6 - Aseen --w.-- .,.v--.--, -,-.-.,---.----,~,w,,-.-s -,--.-..---.-..-.~.,-..--.,,--_--__---.m_ __-.-.~,--..--%.---

) 5.0

SUMMARY

AND CONCLUSIONS l c@ In reviewing the ORNL report on integrity of reactor vessel support systemsW a number of conservatisms have been identified. It now is evident that the HFIR irradiation spectrum should not be used to model the l behavior of power reactor supports. A more realistic model would be i Porter's model, which has been confirmed by the Shippingport results. Other key conservatisms identified are the relatively large postulated flaws at the grout hole, and the use of worst case design loads instead of best estimate loads. A realistic assessment of the reactor vessel supports leads to the conclusion that the issue here is a long range one, if it exists at all. The governing support configuration is a short beam, cantilevered from the vessel cavity wall. Only three plants have this arrangement, so this issue is not an Owners Group issue. Based on the available information the priority should be lowered to " low" for. generic issue 15. l l-i 5-1 a

i 600 REFERENCES ] 1. Cheverton, R.D. et.al. ' Impact of Radiation Embrittlement on Integrity of Pressure Vessel Supports for Two PWR Plants". Oak Ridge National Laboratories Report NUREG CR-5320, ORNL/TM/10966. January

1989, 2.

Porter.L.F., " Radiation Effects in Steel' published in ASTM STP 276, 1959, pp 147. 3. Chopra, 0.K., Monthly Reports on ' Aging Studies on Material from the Shippingport Reactor *, March through May 1989, transmitted by letter to E. Woolridge, USNRC. l L~ L l l l i i 6-1 i i I (

4 APPENDIX A REVIEW OF REACTOR VESSEL SUPPORTS SURVEY i During the summer of 1989, a survey was carried out of all members of the Westinghouse Owners Group. The survey was designed to obtain as much information as possible from all Westinghouse plants regarding the configuration and operating conditions for the reactor vessel supports. ] The survey was conducted in response to a concern which arose over the j I potential for damage to the supports by low flux, low fluence irradiation. A meeting was held on this subject between the Westinghouse Owners Group and the Nuclear Regulatory Commission on June 15, 1989, at which time an independent l assessment of the issue by Westinghouse was presented. This presentation is documented in the main body of this report. A study had been completed earlier for the NRC'by Oak Ridge National Laboratory (1), and contained in this reference is a summary of all support configurations for operating plants. One of the key goals of the survey was to verify the support configurations listed in that document. The survey results are summarized in table A-1, which reveals that responses were obtained from 53 of 56 operating plants. The survey showed that several plants have support configurations which are different from those reported in (1). Seven plants reported differences, and some provided detailed drawings, which have been included here as figures A-1 through A-3. Excore dosimetry measurements are available on six plants to date, and others have measurements in progress. The values shown in table 2 have been calculated from actual measurements taken in the reactor cavity, over one fuel 17 2 cycle. Results show fluence values ranging from 1 to 4 x 10 n/cm at the 17 2 top of the core, and from 6 to 16 x 10 n/cm at the core mid plane. Only one measurement is available (thus far) at the nozzle bottom, where most vessels are supported, and this measurement shows a very low value of 16 2 5.1 x 10 n/cm, 11 Cavity temperatures were reported for 18 units and revealed a range of I" 80-150*F. This result is consistent with the maximum value of 150'F reported ,j, mwi im io A-1

L

1 in(1). Materials of construction were requested, and the responses revealed a variety of structural materials used. Sore responses included bolting materials as well, but these ware not included on the summary in table A-1. The survey revealed that very few utilities had records of preservice inspections performed on their supports. Very few unusual features were reported, and those reported did not appear to be of concern relative to the structural integrity of the supports, as shown in table A-1. Virtually all of the vessels supports received some form of preservice inspection during the construction period. Most of the more recent plants have conducted volumetric examinations on the reactor vessel supports, and some of the early plants have done so as well. The variety of support configurations and range of construction dates make further generalizations difficult. In summary, the survey has shown that in general the configurations reported in reference [1] are correct, and where they are incorrect, the corrected configurations have been identified. Reacter cavity temperatures were found to be bounded by the maximum value of 150*F reported in reference (1). A wide range of materials were identified, but all carbon and low alloy steels fall in the same general category relative to susceptibility to low fluence irradiation, as shown by Porter (2). The survey also provides a variety of other useful information, including cavity dosimetry results showing expected dosages for 32 effective full power years. The fluence in the nozzle support region was found to be extremely low, for the one plant which reported available data (5 x 1016 3fe,2), Since the' majority of plants have their key support configurations located in this area, there should be no cause for concern here. The results of the survey support the conclusions of (1) as to the governing plants for possible i susceptibility,and these two geometries have already been analyzed in detail (1). t swainoiuo 'o A-2 l

TABLE A-1: StFPORTS SURVEY SUsummRY EX-CORE

RESPONSE

NUREG DOstmETRY7 CAVITY UNUSUAL PLANT RECEIVED CDNF IRMED7 (See Table 2) TEMP. m4TERIAL FEATURE 57 A X Y(4C) A588 None D X Y(4C) A588 None C X Y(3) 100-150F A516 Gr 60 None D 'X Y(3) 100-150F A516 Gr 70 Mone E X Y(4E) F X Y(4E) G X Y(4E) n x Y(4E) I None J X Y(4G) 123F A516 Gr 70. A3G K X Y(4G) 123F A516 Gr 70. A3G L X Y(4C) ff0F m X Y(4C) 80F N X N(30) I11 A36 y O X N(4E) {GI A588 Gr 8 ( e P X N(4E) (G) A588 Gr 8 w 0 X Y(4F) Yes R X Y(4F) Yes 5 X Y(4C) (2) 143F T-1 None T X Y(4C) Yes A441. A36 None U X Y(4F) None V None W X Y(4C) 104F A572. A516 Gr 70 None X X Y(4C) 104F A572. A516 Gr 70 None Y X Y(4C) 90F A588 None Z X Y(4G) 123F A516 Gr 70 8 - 4* diam. vent AA X Y(4G) 123F A516 Ge 70-holes in 3* plate NOTE: Brackets refer to notes on last page of Table A-1. 3983s/101789: 10

.y TABLE A-1: SUPPORTS SURVEY SusanARY (cont.) EX-CORE

RESPONSE

NUREG-D051ssETRY7 CAVITY UNUSUAL PLANT RECEIVED CONFIRMED 7 (See Table 2) TEMP. st4TERIAL FEATURES 7 BB X Y(3) A516 Gr 70. A537 CC X Y(3) A516 Gr 60 None 00 X Y(3) A516 Gr 60 None EE X N(2F) Yes 80-90F T-1. A53 None FF X N(2F) Yes 80-90F T-1. A53 None GG X Y(4c) (3) 133F A588. Gr A Mone ful X Y(4c) (3) 133F A588. Gr A Mone II X Y(4F) 110F A441 Or t lled and niactilned JJ X V(4F) llOF A441 shear pin hole KK X N(28) (41 11IF A3028. A36 Mone LL X Y(4F) (5) hun X Y(4C) 120-130F A572 None me X Y(4C) 120-130F A572 Mone 00 X N(4C) None PP X M(4C) None 00 X Y(4C) RR X Y(4C) SS X Y(3) A516 Gr 60 Mone TT X Y(3) A516 Gr 60 None UU X Y(4A) VV X Y(44) Yes WW X Y(4A) XX X Y(4C) A572. A3028, A588 3.5* diam. holes in e m s i YY X Y(4C) 120-13OF A572 Mone ZZ X Y(4C) 120-130F A572 Mone ZA None NOTE: Brackets refer to notes on last page of Table A-1. 4 4 l 3983s/101789:10

m e e.w< V-TABLE A-1: SUPPORTS SURVEY SUNIARY (cont.) EX-CORE

RESPONSE

NUREG 005tIRETRY7 CAVITY UNUSUAt. PLANT RECEIVED CONFIRGIED7 (See Table 2) TEluP. IIATERIAL FEATURES 7 ZB X Y(48) 122F A208 Gr B None ZC X Y(4E) Yes ZD X Y(4E) Yes [1] 38 - four block supports on steel cylindrical shell {2) 4C is the closest geometry, but the actual ts slightly different see figure A-1 (3) Sitghtty different, see figure A-2 [4] 28 - long column attached to vesset support brackets between nozzles, see flgure A-3 (5) 4F* This girder'Is not continuous (6) Weldment pedestal is embedded in concrete: support confIguratton looks 1Ike 4C. but no figure provfoed Y tn 3983s/101789:10

TABLE A 2 EXCORE DOSIMETRY Plant Core Nicplane Core Top Nozzle Bottom 2 2 2 (n/cm) (n/cm ) (n/co) 17 17 2C 6.5 x 10 1.7 x 10 17 17 20 5.8 x 10 4.3 x 10 17 17 0 8.1 x 10 2.4 x 10 17 17 R 8.2 x 10 2.1 x 10 18 17 W 1.4 x 10 3.8 x 10 18 17 16 T 1.6 x 10 3.4 x 10 5.1 x 10 Notes: 1. Core top is near bottom of box beam supports. 2. These values are for 32 EFPY, estimated from measurements made over one fuel cycle. 3. All values are taken at energies greater than 1 Her 1 l 3 .ien ie A-6 1..

RV Nosale A ( yy Sliding foot & shoe 4' Top Flange 8' Web 4' 1 ,,,i v,,s v,,i v,i, w s w,,,% v,,% v,i w,,,i v,s<i, w,,% v,,a w,,,, w i L w,,,;q,,;v,sy,w;v,,;v,,;q,,;,w,,,,;q;,;q,,;q,aq,cq,, 6 L 4 se-i c l 15' l-t,%,y;y;y,y;%,?2 Stainless Steel I .i i 1T-1 Steel l l i l t l 9; 9 da'i9;9;949;9;';9;9;9;9;9;9;';9;a; a,.aEMau<aaamaM

9<;<;%%;9;%;%;9;9;9;9;9;9;9;949;9; i49;' 4 N44%4%%%4%4%4 9<

9d4;9;9;9;9;9;9;9;9'Q;9;9;9;9;9;' %M6M4MM6%$%$%4%%4M Apprar, 39' View A-A u I Figure A-1. Reactor Vessel Support Configuration for Plant S A-7 l f

l \\ FMEDDED CONTIMUS l STEEL BNC Rim i ? l 8 m rp A o LOX 1RDER DeEDDED -TEE l TIE]BIActu (AIR SLPPLY Ape WWED To ;WD 11m %g DisCH4mEDUCTSYSTEM " **) W Mu h 1DtActs Gusset \\ \\ \\ / 1 \\ ) ) l EreEDDED a h /(CONCRETE IW SCNN) STEEL COWPNS DeEDDED h w b Q BASE PLATE he ANOCR BOLTS b 4 4 -, i Figure A-2. Reactor Vessel Support Configuration - Plants GG and HH A-8 s 4 -_._ _- -__. : _ _ -.-- -._,..L!

) ) es l-q / l j l i \\A G / \\ / m / / l' g i i 0 s / / \\ ~ 'T N I i 4, w REACTOR SPHERE S l C - TURBINE GEfERATOR e SECTion A REACTOR SUPPORT ^~ TYP. 3 PLACES g .\\ l 1 g \\ \\- \\ \\ \\ \\ i \\ \\ t g PLAN \\g ,l k S r EcTion f A e f f i i s O .%~ g. N I ,6 l ,6 i ifc i_ n w =* A W Figure A-3. Reactor Vessel Support Configuration - Plant KK 1 A-9 D W h -m =---m,-w=w r

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