ML19319D664
| ML19319D664 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/04/1974 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19319D663 | List: |
| References | |
| NUDOCS 8003180699 | |
| Download: ML19319D664 (10) | |
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-s MECHANICAL ENGINEERING BRANCH DIRECTORATE OF LICENSI'?G CRYSTAL RIVER UNIT 3 DOCKET No. 50-302 3.6 Protection Acainst Effects Associated with the Postulated Ruoture of Piping The design criteria used for detersining the break locations and break orientations for the reactor coolant pressure boundary are consistent with the AEC staff position.
Both longitudinal and circumferential pipe breaks were assu=ed to occur at any location.
Generally physical separation is e= ployed 'to assure that a single incident will cot da= age both portions of redundant safety-related piping or equipment.
Where physical separation cannot be achieved, shields and restraints are employed.
The proposed desi;n of piping restraints as applied to the reactor coolant pressure boundary provides adequate protection of the contain=ent structure, the unaffected reactor coolant system components, and these systems important to safety which are either interconnected with the reactor coolant system, or in close proximity to the reactor coolant pressure bo'undary in which postulated pipe failures are assumed to occur as a design basis loss-of-coolant accident.
These provisions for protection against the dyna =ic effects associated with pipe rutpures and the resulting discharging coolant provide adequate assurance that, in :ne event of the occurrence of 8003180[ N I
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T the combined londings.1= posed by an earthquake of the =agnitude specified for the Safe Shutdewn Earthquake and a concurren; single pipe break of the largest pipe at one of the design basis break locations, the following conditions and safety functiens will be accc=:cdated and assured:
(1) the =2gnitude of the design basis loss-of-coolant accident can not be aggravated by potentially cultiple failures of ii P P ng, (2) the reactor e=ergency core cooling syste=s can be expected to perform their intended function.
The criteria used for the identification, design, and analysis of piping syste=s where postulated breaks =ay occur constitute an acceptable design basis in meeting the applicable requirements of AEC General Design Criteria 1, 2, 4, 14, 15, 31 and 32 and are consistent wit'. the staff position.
High energy piping breaks outside containment are presented in the report, " Effects of High Energy Piping System Breaks Outside Reactor Building," October 1973, revised November 1973.
The criteria employed by the cpplicant is consistent with the staff position transmitted to the applicant by letter dated December 22, 1972.
The protection provided against the dynamic effects of postulated pipe breat;- and discharging fluids in piping systems containing high energy fluids and located outside the containment is adequate
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4 to prevent damage to structures, systems and components to the
---l extent considered necessary to assure the maintenance of their structural integrity.
Such protection provides reasonable assurance that the safe shutdown of the reactor can be accon-1 plished and maintained, as needed.
The criteria used for the. identification, design and analysis of high energy fluid lines where postulated breaks may occur constitutes an acceptable design basis for satisfying the j
applicable require =ents of AEC General Design Criterion 4, and I
are consistent.with those specified in the staff position.
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4 3.9 Mechanical Systems and Components 3.9.1 Dynamic-Systen Analysis and Testing The applicant has designated Oconee 1 as the prototype plant from which precperational vibration test results are applicable in evaluatin; the design adequacy of the reactor internal structures of the Crystal River plant.
Thus, only confirmatory tests in accordance with Regu..atory Guide 20 will be conducted. We find this progran of preuperational vibration testing of reactor internals is acceptable.
The reactor internals of Crystal River were designed to withstand the dynanic effects of the postulated accident, a simultaneous occurrence of loss-of-coolant due to coolant pipe rupture near the nozcle and safe shutdown earthquake.
The cpplicant has referenced Topical Reports (1) 3AW-lC003-1-Rev.1, " Reactor Internals Strass and Deflection due to Loss-of-Coolca: Accident and Maximum Hypothetical Earthquakes," (2) BAW-10035, " Fuel Assembly Stress and Deflection Analysis for Loss-of-Coolant Accident and Seis=ic Excitation." The final evaluations of these topicals have been conpleted.
We find.these topicals are acceptable for Crystal River.
A series of preoperational functional tests will be perfor=ed on piping systems both inside and outside the reactor coolant pressure boundary, in accordance with Paragraph I-701.5.4 of ANSI B31.7 Nuclear Power Piping Code. This code requires that piping shall
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~acceptable levels. T>ese tests will verify that the piping and piping restraints have been designed to withstand dyna =ic effects due to valve closures, pump trips, and operating modes associated with the design operational transients.
The applicant has submitted a test acceptance criteria.
Also, during the testing of the main steam turbine stop valve closure and relief valve opening, instru-mentation will be installed to provide tent data that will be compared with calculations to insure that displacements are within allowable 1:.mits.
If the allowable limits are e:cceeded, steps will be-taken to reduce the displacement to within accept-able limirs. We find this program to be acceptable.
Th'e applicant has stated that he has or will conduct either tests or analysis for each item of Category I =echanical equip-ment to assure functional capability of that equipment during a seismic event.
n'e find this criteria to be acceptable.
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g 3.9.2 Structural Intecrity of Pressure Retainin Components Pressure retaining components in fluid systems designated Seismic Category I which are within the boundaries of AEC System Quality Group Classifications A, 3 or C were designed to meet 4
the requircments of the codes and standards specified in 10 CFR 50.55a or Regulatory Guide #26 as appropriate. All components are designed to sustain nor=al operating loads, anticipated operational occurrences and the operational basis earthquake (1/2 SSE) within the stress limits of the code specified.
In addition, Quality Group A conpenents are designed for a limiting pri=ary stress of two-thirds of ulti= ate strength for the combin-ation of design 1 cads plus Safe Shutdown Earthquake and pipe rupture loadins.
Quality Group 3 and C components are desi;nad to sustain che Safe Shutdown Earthquake wittin stress limits comparable to those associated with the en.rgency operating condition of current component codea.
The specified design basis combination of loading as applied to the design of the safety-related ASME Code Class 1, 2 and 3 pressure-retaining components in systems classified as Seismic Category I provide reasonable assurance that in the event (a) of earthquake should occur at the site, or (b) an upset, emergency, or faulted plant transient should occur during normal plant operation, the resulting combined stresses imposed on the system components may be-expected not to exceed the allowable design stress and strain limits for the materials of construction.
Limiting the stresses under such
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loading combinations provides a conservative basis for the design of the systen cecponents to withstand the most adverse combination of loading events without gross loss of structural integrity.
The design load ceabinations and associated stress and defor:stion limits specified for ASME Code Class.1, 2 and 3 components constitute an acceptable basis for design in satisfying the General Design Criteria l
1, 2 and 4.
l The criteria used in developing the design and nounting of the safety and relief valves of ASME Class 1 and 2 systens provides adequate assurance ths.t. under discharging conditions, the resulting stresses i
are expected not to axceed the allowable design stress and strain li=its for the materials of construction.
Limiting the stresses under the loading combinatiens associated with the actuation of these pressure relief devices provides a conservative basis for the design of the system co=ponents to withstand these loads without loss of structural integrity and i=pairment of the overpressure protection function.
I The criteria used for the design and installation of overpressure relief devices in ASME Class 1 and 2 systems constitute an accept-able design basis in meeting the applicable requirements of AEC General Design Criteria 1 & 2, 4, 14, 15.
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8-3.9.3 Co=ponents Not Covered bv ASME Code The design criteria applied and the tests perfor=ed on the fuel and control rod asse=blies and control rod drives of Crystal River Nuclear Power Stations are co= parable to those of prior designs which were found' acceptable for Arkansas Nuclear Power Station.
The use of these criteria provide reasonable assurance that the 4
fuel and control rod assemblies and control rod drives =ay be expected to withstand the i= posed loads associated with normal reactor operation, a'nticipated operational transients postulated accidents, ani seismic events without gross loss of their structural integrity or i= pair =ent of function.
Co=pliance with these design criteria fulfills the requirements of AEC General Design Criteria 2 and 14 as these criteria relate to fuel and control rod asse=blies, and control rod drives.
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3.10 Seismic Qualification of Category I Instrumentation and Electrical Equionant The reactor protection system, engineering safety feature circuits and the energency power syste= vere designed to =eet Category I (seismic) design criteria.
A seismic qualification progran "as implemented for confirming that all Category I instrumentation and electrical equip =ent will function properly during the safe shutdova earthquake and the post-accident operation, and all their support structures are adequately designed to withstand the seismic disturbance.
The operability of the instrumentation and electrical equip =ent were assured by testing.
The design adequacy of their supports were assursd by either analysis or testing.
The applicant has referenced Topi:21 Report 3AW-lC003, Qualification Testing of protection Syste= Instrumentation." The final evaluation of this report has been completed.
We find the referenced topical is acceptable provided (1) the responses o' cabinet assemblies at various instrument or device mounting locations due to seisnic disturbances at the plant site are deter =ined in order to define qualification input levels for the instruments and devices to be mounted in the cabinets, and (2) the maximum response determined in (1) does not exceed lg.
The applicant has agreed to meet these stipulations.
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- s 4.0 Reactor 4.2 Mechanical Desi2n of Reactor Vessel Internals For nor=al design loads of mechanical, hydraulic and thermal origin, including anticipated plant transients and the operational basis earthquake, the reactor internals vare designed to the stress limit i
-criteria of Article 4 of the ASME Boiler and Pressure Vessel Ccde Section III.
For the loads calculated to result fro: the loss-of-coolant accident (LOCA), the design basis earthquake (SSE) and the cembination of these postulated events the reactor internal components were designed to the criteria submitted in 3 & W Topical Report 3AW-10u08, " Reactor Internals Stress and Deflection Due to a LOCA and Maxic.r hypothetical Earthquake" which was referenced in the FSAR.
These criteria are consistent with ce= parable cede e=ergency and faulted operating condition category linics and the criteria which have been accepted for all recc: tly licensed plants. We find these critet acceptable.
The dynamic analyses of the Arkansas Nuclear Plant Unit 1 reactor internals are discussed in Section 3.9.1, " Dynamic System Analysis and Testing."
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