ML19322B997
| ML19322B997 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/29/1970 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7912200712 | |
| Download: ML19322B997 (7) | |
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UU?I E & i' liq Tcter A. Morrin, Director Division of Rer.ctor Licensing 50-269, 270 OCO; LEE liUCLEAR STATIO:1 U'i1TS 1, 2, Al!D 3, DOCKET NOS.
N;D 287 The enclosed evaluation of the recctor pressure vessel, reactor internal structures, reactor coolant systens, and Class I acchanical equipnent of the Oconce nuclear Station was prepared by the DRS Structural Engineering Branch.
Original s!cr.ed by C. G. Case Edson G. Cese, Director Division of Reactor Standards
Enclosure:
Octance Revice cc v/ enc 1:
R. DcYoung, DEL R. Maccary, DPS C. Long, DPJ, A. Drocerick, DRS A. Schucacer, DEL K. Uichman, DRS i
bec: E. G. Case Dis tribution:
Suppl. Docket File:
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OCO';EE NUCLEAR STATIO:: U:;ITS !;0. 1, 2, AND 3 DOCKET !!OS. 50-269, 270, AND 2S7 t
henctor Coolant System The reactor coolant system has been designed as a Class I (ceiscic) system to withstand the normal loads of ecchanical, hydraulic, and thermal origin including anticipated plant transients and the operational basic carthquake within the stress linits o'f the appropriate codes given below.
The steam generator, pressurizer, and reactor coolant pump casings have been designed to Class A requireccats of Section III of the AS:fE Boiler and Pressure vessel Code,1965 edition, including the Susser 1967 Addenda.
Safety and relief valves are in accordance with the require-ments of Article 9 of the above edition and addenda of Section III.
Piping which is part of the reactor coolant syste= has been designed to the ANSI E31.7 Code for ::ucicar Power Piping, dated February 1960, includinr, the June 1968 Errata.
Rondestructive exacination requirements for reactor coolant systen purps and valves are given in Table 4-12 of the FSAR. These examinations include radiography of castings, ultrasonic testing of forgings, dye penetrant inspection of pu=p and valve body surfaces, and radiography of i
circumferentini veldncnts. This program upgrades the nondestructive testing of pu=ps and valves within the reactor coolant pressure boundary to essentially that of the AS:iE Code for Pumps and Valves for Nuclear Power.
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. The applicant states that carthquate loads for the OBE and DtE have been determined by dynanic analysca. We are currently exanining the analyt-ical techniques employed in these analyses in conjunction with our consultants.
We expect to have this natter resolved prior to the ACRS nceting for this plant.
Reactor Vessel The reactor vessels have been designed and fabricated in accordance with Class A requirencnts of Section III of the ASME Eoiler and Pressure Vessel Code, 1965 edition, including the sunner 1967 Addenda. Applicabic Code Cases are 1332, 1335, and 1336.
The vessels are essentially identical to those intended for the Russelville, Crystal River 3 and 4. Rancho Seco 1, Midland 1 and 2, and Oyster Creek 2 plants, and have been designed to permit complete renoval of the veasci internals. Fabrication natcrials are low alloy steel plates Type SA-533, Grade E, class 1, and forgina steel Type SA-50S-64, class 2.
The vessel interior is clad with Type 304 austenic stainless steel applied by veld over1cy techniquc. The applicant han inforr.ed us.that furnace sensitization of stainicss steel vessel raterial has been limited to the non-pressure-bearing interior cladding. The requirencnts for nondestructive examinations have been linited to those required by Section III, except that 1
head and shcll plate naterial and flange forgings have been given a 100%
volunetric examination using both longitudinal and shear wave UT techniques.
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The Oconce reactors are the prototype vessels of the Babcock &
Wilcox supplied 850 MVe Clacs of reactor vessels. However, no unusual t
design or f abrication prob 1c=s either before or during manufacture have bec'n identified. We conclude that the reactor vessels as designed and fabricated are neceptable.
Renctor internals For normal design loads of nechanical, hydraulic, and thermal origin, including the operational basis earthquake and anticipated transients, the reactor internals have been designed to operate within die allowable r
stress intensity linits of Section III of the ASME Boiler and Pressure Vessel Code.
All interncis components are designated as Class I sotamic items, and vill be decigned to withstand loads resulting from a conbined design basis i
carthquake and locc-of-coolant accident. Strain limits for the internals under this combined load vill be held to Icss than 20*.' of the unifore utlimate strain for this naterini (334S.S.) corresponding to an clastically calculated stress limit of not greater than 2/3 of the ultinate tensile strength. Allouable deficction limits will generally be within 50*4 of loss-of-function deformation limits. We consider these design limits to be acceptable.
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! Topical Report LAW-10008 Parts 1 and 2 is referenced in the FSAR as outlining the methods of analysis to be employed for the internals and fuel ascenblics under loss-of-coolent endibaign basis carthquake loadings for skirt supported reactor vessels. We are presently, with the aid of our consultant, reviewing the analyses presented in the topical report; however, coapiction of this review awaits the subnittal of the applicant's responses to our concerns. We c=pect to receive these responses early in July.
3 Other Class 7 (Scisnic) Mechanical Equipnent Quality control standards for engineered safety features are su=uarized belou:
All velding procedures and operators conectned with the fabrication of j
pumps and valves have been qualified to Section IX of the ASME Loiler and i
Pressure Vesocl Code, liydrostatic tests o. valve bodies cnd valve cents ucre conducted in i
accordance with ANSI D16.5 and Mss Sr-61.
runpc have been hydrostatically tested to the requirenents of UG-99 of Section VIII - Division 1 of the
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i A5M Code.
The quality control standards for pu=ps and valves require incpection of rav natcrial and review of natorial certification in conformance to ANSI B31.7 requirencnts..In addition, radiography and lionid penetrant tests of valve bodies, valve bonnets, and pump casinge are perforned to nect ANSI B31.7 acceptance standards.
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5-which These requirenents result in a fabrication and inspection pronrcm d Valves for contain the essential ciceents of the AS'!E Code for Pumps an Uc find these requireecnts neceptable.
!!uclear Fouer.
h been The codes and ctandards applicabic to other Clocs I syste=s have reviewed and are considered adequate from a safety standpoint.
i All Class I equip =cnt has been designed to withstand the desip,n Ve are presently, with the aid of earthquake without loss of functien, d to enIculate the our consultant, reviewing the analytical procedures use seinnic loadings on Class I equipment.
f d of In conjunction vith this effort we are also revicuing the netho l
I equipment, cpecifying the seistic design requirements for purchased C ass
' rethods of certifi-the adequacy of the appifcent's check on the vendors l
i nd their cation, the design orteni ntions involved in scionic des gn a h intcrchan;;e recponsibilitics, and docu:sented procedures to provide for t e I
Uc pinn to report of design infornation between the involved crecni ntiens.
l l t on these untters prior to the ACRS meetint, for this p an.
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_ Vibration Control l
Flow induced vibration analyses have been nade for reactor interna illance tube such as the thereal shicid, fuel assechlics, fuel rods, surve bly, and piping and specimen holder assembly, control rod guide tube asse=
The thereal chield analysis for vibration for the in-core r-onitors.
i cting on the problems showed that the f1cu induced pressure fluctuat ons a
~6-nurface of the shield renulted in r,odal amplitudes less than 0.002 inch.
These analyncs considered inlet ficni itpingecent and turbulent flow, as vc11 as natural frequency cniculations, to establish that a factor of at 1 cast two exista betvoen conditions of possibic resonance end excitatica frequencies.
It has also been detcrained that the flow induced pressure fluctuations actin,, on the dinc of the vent valve are such that for normal i
operation there i= alwayn a positive net closing force acting on the disc We understand that the applicant vill present a program to the staff f
which outlines his plans for confirmatory vibration conitoring during picoperational testing of the Oconce Plant.
Ve expect to have this infor-nation prior to the ACRS necting for this plant.
The fensibility of inservice nonitoring for vibration and the detection of loose parts in being cyplored by the nuc1 car steam systcu supplier, D&W k
They,have invcetigated the application of such sensors cs accclerometers strain ga;:cc and load cells to r,onitor vibration of internals, and of inertially loaded-force pickups to nonitor for loose parts.
i EtU plans addi-
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tional discussion with consultants and instrucentation vendors in ord er to detert:ine the fcanibility and practicality of such systemn in operating PWr cystcon.
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