ML19261F019
| ML19261F019 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/24/1972 |
| From: | Maccary R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910180628 | |
| Download: ML19261F019 (11) | |
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AtJG 2 41972 Richard C. Defoung, Assistant Director for Pressurized Water Reactors Directorate of Licensing THREE MILE ISLAND NUnm STATION UNIT NO.1. DOCKET NO. 50-289 Plant Naza Three Mile Island Nuclear Station Unit No. 1 Licensing Stage: OL Docket Number 30-289 Responsible Branch and Project Leader FWR-4, H. Faulkner Requested Completion Date 8/18/72 Applicants response date necessary for coupletion of next action planned on project: N/A Description of Responses N/A Review Status: Complete The finsi evaluation for the subject plant which was prepared by the Mechante=1 Engineering Branch, Directorate of Licensing, dated Mt y 11, 1971, has been revised to reflect significant changes submitted in Amentimants through No. 27. Our updated evaluation as requested in your mano of June 22, 1972 is enclosed.
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R. R. Maccary, Assistant Director for Engineering Directorate of Licensing Enclosures Final Evaluation - Mechanical for ihres Mile Island ec w/ enc 1:
S. H. Hanauer, DRTA J. M. Hendrie, L 1/86 0'87 A. Schwencer, L
- o. 1. tan e, t H. J. Faulkner, L J. P. Knight, L N. H. Davison, L Docket File (50-2891 L, Reading File cc w/o encli L:MEB File A. Giambusso, L
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3.6 Criteria for Protection Against Dynamic Effects Associated with a Loss-of-Ccolant Accident In the analysis of the reactor coolant loop, the applicant has applied an equivalent static approach to the criteria for protection against a pipe rupture considering both impact and jet loadings on the restraint systems designed to prohibit damage to containment or other safety related systems. Circumferential and longitudinal breaks were postulated at locations where they would impose the most severe loading conditions on piping components and supports.
Factors greater than 2 were applied to the normal thrust load at postulated pipe breaks as a result of conservative assumptions of maximum theoretical momentum change and ideal flow at the postu-lated break locations.
Restraint systems were designed to withstand the combined loads calculated to. result from postulated pipe rupture
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and from the maximum hypothetical earthquake; resulting stresses were limited to the yield stress of the restraint materials. Additional protection for vital syste::.2 is provided by the secondary shield walls surrounding each steam generator and its pair of reactor coolant pumps and from the routing of safety related pipirg to attain separation of systems within the shield area.
Restraint systems and protection criteria for main steam and feedwater piping were designed and developed on similar bases. Under the combined loads calculated to result from impact and jet thrust some plastic action will result but this action is well within the energy absorbing 1486 088
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capabilities of the structural steel members comprising these systems.
We find this approach for protection against pipe rupture to be acceptable.
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e 3.6.1 Seismic Inout The seismic design response spectra submitted produce a magnification factor greater than 3.5 in the period range appropriate for the response of structures, systems, and components.
Proposed structure and equipment damping factors are in accordance with those recommended by N. Newmark. The response spectra are derived from the most critical combination of the normalized Golden Gate and El Centro (1940) earthquake records. These records were also used as input to confirm the structural integrity of structures, systems, and components.
We conclude that the seismic input criteria proposed by the applicant provide an acceptable basis for seismic design.
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3.6.2 Seismic System and Subsystem Dvnamic Analyses Modal response spectrum multi-degree-of-freedom and normal mode-time history methods are used for the analysis of all Category I structures, systems, and components. Governing response parameters have been combined by the square root of the sum of the squares to obtain the modal maximums when the modal response spectrum method is used.
The absolute sum of responses is used for clcsely spaced frequencies. Floor spectra inputs used for design and test verification of
. structures, systems and ccaponents were generated by semi-empirical methods and confirmed by the normal mode-time history method. A vertical seismic-system dynamic analysis was employed to account for significant vertical amplifications for the seismic design of structures, systems, and components.
Constant vertical load factors were employed only where analysis showed sufficient vertical rigidity to preclude significant vertical amplifications in the seismic system being analy::ed.
We and our seismic consultants conclude that the seismic-system dynamic methods and procedures proposed by the applicant provide an acceptable basis for the seismic design.
1486 091 3.6.3 criteria for seismic Instrumentation Program The type, number, icoation and utilization of strong motion accelerographs to record seismic events and to provide data on the frecuency, amplitude and phase relationship of the seismic response of the containment structure corresponds to the recommendations of Safety Guide 12.
Supportiag instrumentation will be installed on Category I structufes, systems, and components in order to provide data for the verification of the seismic responses determined analytically for such Category I items.
A plaa for the utilization of the acquired seismic data will be developed before start-up.
We conclude that the Seismic-Instrumentation Progran proposed by the applicant is acceptable.
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1486092 4.2.3.1 Reactor Internals - Design For normal design loads including the operational basis earthquake and anticipated transients, the reactor interna'.s have been designed to operate within the acceptable allowable stress intensity limits of Article 4,Section III of the AS'1E Boiler and Eressure Vessel Code.
All internals conponents have been designed to withstand the loads _calcualted to result from the Design Basis Earthquake, the Design Basis Accident and the combinatien of these postu-lated events.
Strain limits for the internals under these combined loads will correspond to an elastically calculated stress limit of not greater than 2/3 of the ultimate tensile strength. Allowable deflection limits are generally within 50% of loss-of-function defor=ation limits. We consider these design limits to be acceptable.
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4.2.3.2 Dynamic System Scismic and LOCA Analysis Topical Repor?. EAW-10008, Parts 1 and 2, is referenced in the FSAR as outlining the =e: hods of analysis employed for the internals and fuel assemblies under loss-of-coolant and design basis earthquake loadings fo'. skirt supported reactor vessels. We have, with the aid of our consultant, reviewed the =cthods of analyses presented in this report and find them acceptable.
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4.2.3.3 Reactor Internal Structures - Vibration Centrol Verification of the calculated vibration responses will be accomplished by comparing vibration response measurements made during the Three Mile Island preoperational testing with similar measurements =ade at the designated prototype plant for the Babcock & Wilcox Cc=pany product line, Oconce I.
A portion of the Oconee I instrumentation will be duplicated ir. design and location at Three Mile Island to allow direct comparison of data.
We find the proposed preoperational test program acceptable pro-vided that the Oconee I tests are successfully completed and that comparative data demonstrating the validity of the methods utilized to predict vibration responses for thu Babcock & Wilcox product line are available-prior to the coupletion of the Three Mile Island test program in accordance with AEC Safety Guide 20.
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4 5.2.1 Reactor Coolant System The reactor coolant system has been designed to withstand normal design loads including anticipated plant transients and the Operational Basis Earthquake within the acceptable stress limits of the appropriate codes given below.
The steam generator, pressurizer, and reactor coolant pump casings have been designed to Class A requirements of Section III of the ASME Boiler and Pressure Vessel Code, 1965 edition, including the Summer 1967 Addenda.
Safety and relief valves are in accordance with the requirements of Article C of the above edition and addenda of Section III.
The design, fabrication, inepection and testing of the reactor coolant piping including the pressurizer surge line and spray line is in accordance with the USAS B31.7, Code ror Pressure Piping, Nuclear Power Piping, dated February, 1968, including the June 1968 Errata.
me Nondestructive examination requirements for reactor coolant system pumps and valves are given in Table 4-12 of the FSAR.
These exa=-
inations include radiography of castings, ultrasonic testing of forgings, dve penetrant examination of pump and valve body surfaces, and radiography of circumferential welds. This program upgrades the nondestructive examination of pumps and valves within the reactor coolant pressure boundary to essentially that required by the ASME C-de for Pumps and Valves for Nuclear Power.
6 096 The design, fabrication and inspectica criteria discussed above are censistent with those accepted for all recently reviewed plants of this type and we find them acceptable.
Components of the reactor coolant system (RCS) have also been designed to withstand the loads calcualr d to result from the Design Basis Earthquake, the Design Basis Accident, and the combination of these postulated events.
Strain li=its for the RCS components under these combined leads correspond to an elastically calculated stress limit of not greater than 2/3
'of the ultimate tensile strength. We consider these design limits to be acceptable.
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