ML20037A422
| ML20037A422 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/31/1972 |
| From: | Maccary R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8001140822 | |
| Download: ML20037A422 (15) | |
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MAY 311972 Richard C. DeTount Assistant Director for Pk'R's, Directorate of Licensirg OCONEE NUCLEAR STATION, U!IITS 2 AND 3, DOCKET NOS. 50-270 AND 50-287 Adoquate responses to the encioned request for additional inforr.ation are required before we can complete our review of the subject application. These requests, prepared by the L Mechanical Engineering Branch, concern the reactor coolant pressura boundary, reactor internal structures, safety related nachanical systecs, seismic design criteria and pipe whip criteria submitted in Volumes I through IV of the FSAR. This request supersedes our previous request dated 4-27-72 and reflects the results of the reevaluation discussed
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in our letter dated 5-27-72 The applicant's description of ceismic qualification testing for instrumentation and equipment (Section 7A.2) appears to be extracted from B & W Topical Report EAW-10003, " Qualification Testing of Protection. System Instrunentation" (March 1971). The additional information required to complete our review of this report was forwarded in our request of March 13, 1972 for the Three Mile Island Station Unit Nc.1. Docket No. 50-289. Our review of the Oconee 2/3 application can therefore be expedited if the applicant will agree to reference BAW-10003 and provide the requested information.
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The applicant has referenced (Section 14) 3 & W Topical Report BAW-10008, " Reactor Internals Scress and Deflections due to Loss-of-Coolant Accident and Maximue Hypothetical Earthquake" (June 1970). The sections of the report applicable to Oconce are currently under review by the MEB and additional information may be required prior to completion of this review.
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REQUEST FOR ADDITIONAL INFOR".ATION OCONEE NUCLEAR STATION UNITS 2 & 3 DOCKET NOS. 50-270/287 3.6 CRITERIA FOR PROTECTION AGAINST DYNAMIC EF IATED k'ITH A LOCA 1
Provide a core detailed description of the raeasur es that have been used to assure that the containment liner and all ess ential equip-ment within the containment, including conponents of the primary and secondary coolant systems, engineered safety feat ures, and equipment supports, have been adequately protected a gainst blev-dova jet forces, and pipe whip resulting from a l oss-of-coolant accident.
The description should include:
Pipe restraint design requirements to prevent pi a.
pe whip impact, b.
The features provided to shield vital equipment f rom pipe whip.
The measures taken to physically separate piping a d c.
n other conponents of redundant engineered safety features i description of the type of pipe whip r d.
estraints and the location of all restraints.
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Describe the dynamic system analysis methods and procedures that vere used to confirm the structural design adequacy of the r j-eactor coolant systen (unaffected loop) and the reactor internal loadi ngs. The following information should be included:
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the locations of the postulated double ended pipe rupture on which dynamic analyses were based.
b.
the rupture type (s), such as circumferential and/or longitudinal break (s), for each postulated rupture location.
c.
the description of the forcing functions used for the pipe whip dynamic analyses. The function'should include direction, rise time, magnitude, duration and initial conditions. The forcing function should adequately represent the jet stream dynamics
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and the system pressure differences.
a description of the mathe=atical model used for the dynamic d.
analysis.
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analyses performed to,dceonstrate that unrestrained motion of ruptured lines will not sever adjacent impacted piping or pierce impacted areas of containment liner.
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3.7.2 SEISMIC SYSTDt A'IALYSIS 1
Confirm the validity of a fixed base assumption in the mathematical models for the dynamic system analyses by providing summary analyt-ical results that indicate that the rocking and translational responses are insignificant. A brief description should be included of the method, mathcestical codel and damping values (rocking verti-cal, translation and torsion) that have been used to consider the soil-structure interaction.
2.
Describe the method e= ployed to consider the torsional modes of vibration in the seismic analysis of the Category I building structu.ts.
If static factors are used to account for torsional acceleratices in the seismic design of Category I structures, just:fy this procedure in lieu of a combined vertical, horizcutal, and torsional =ulti=acs systen dynemic analysis.
3.
The use of both the moda! enalysis response spectrum and time history eethods provides a check on the respenses at selected points in the station structure.
Submit the responses o,btained from both of these.eethods at selected points in the Category I structure to provide the basis for checking the seismic system analysis.
4 Provide the dynamic methods and procedures used to determine Category I structure overturning moments.
Include a description
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of the procedurcs used to account for soil reactions and vertical i
earthquake effects.
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Provide the analysis procedure followed to account for the ' damping -
in different elecents of the rodel of a coupled system. Include the criteria use( to account for. co=posite damping in a coupled system with different structural elements.
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' 3.7.3 SEISMIC SUBSYSTCt ANALYSIS 1.
Provide the criteria for combining modal responses (shears, moments, 4
stresses, deflection's. and/or accelerations) when modal frequencies are closely spaced and a response spectrum nodal analysis method is-used.
2.
With respect to Category I piping buried or otherwise located out-side of the containment structure, describe the scismic design criteri2 employed to assure that allowable piping and structural stresses are not exceeded due ia differential movement at support points, at containnen. penetrations, _ and at entry points into other structures.
3.
Describe the evaluation perfor=ed to deternine seismic induced.
ef fects of Category II piping (syste:-s on Categr,ry 1 piping.
4 Provide.the criteria employed to determine the field location of seismic supports and restraints for Category I piping, piping system components, and equipeent, including placement of snubbers and da=pers. Describe the procedures followed to assure
- that' the field locatica and characteristics of these supports and' restraining devices are consistent with the assueptions made in the dynamic analyses of the system.
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3.7.4 CRITERIA FOR SEISMIC INSTRTESTATION
' Provide ~ the following information udth respect -to the use of the seismic instru=entation for the facility:
1 Discuss the seismic instrumentation provided and compar e the proposed seismic instrumentation program with that described in AEC Safety Guide 12. " Instrumentation for Earthquakes "
Submit the basis and justification for elements of the propo sed program that differ substantially from Safety Guide 12 4
2.
Provide a description of the seismic instrumentation such a n= peak e
recording accelercgraphs and peak deflection recorders, thativill be installed in selected Category I (Class l' Seismic) struct I
ures and on selected Category I (Class 1 Seismic) conpenents.
I Include the basis for selection of these structures and component s, the basis for locatien of-the instrumentation, and the extent to which this instrumentation will be employed to verify the seis im c analyses following a seismic event.
3.
Describe the provisions that will be used to signal the cont I
rol room operator the value of the peak acceleration level experienced-in the tendon access gallery of the reactor, containeent structure -
to the control room operator within a few minutes af ter th e earth'-
quake.
Include the basis for establishing the predetermined values for activating the readout of the accelerograph to the control roo m
operator.
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Provide the criteria and procedures-that vill be 'used to compare neasured responses of Category I (Class 1 Seismic) structures and the selected components in the event of an earthquake with the results of the system dynamic analyses.
Include consideration of different underlying soil conditions or unique structural dynamic characteristics that c:ay produce different dynamic responses of Category 1 (Class 1 Seismic) structures at the site.
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3.7.5 SEISMIC DESIGN cc:; TROL MEASURES Describe the design control measures (as specified in Appendix B -
" Quality Assurance Criteria for Nuclear Power Plants") implemented to assure that appropriate seismic input data (as derived from. seismic system and subsystem analyses including any necessary feedback from such analyses) cre correctly specified to the r.anuf acturer of Category I components and equipment to constructors of other Category-I structures and systems.
The responsible design groups or organizations that will verify the adequacy and validity of the analyses and tests employed by manu-facturers of Category I components end equipment and constructors of-Category I structures aiid systems should be identified.. A description.
of the review proce[ures enployed by each group or or ;anization should be included.
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3.9.1 DYNAMIC SYSTDI /JiALYSIS /J;D TESTING 1
Paragraph 1701.5.4 of the ANSI B31.7 Nuclear Power Piping Code requires that piping shall'be supported to prevent excessive
'I vibration under startup and initial operating conditions.
Submit.
a discussion of your vibration operational-test program which will be used to verify that the piping and piping restraints within the reactor coolant pressure boundary have been designed to withstand dynamic,,, effects due to valve closures, pump trips, etcProvide a list of the transient conditions and the associated actions'(pump trips, valve actuations, etc.) that will be used in the vib i
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1 operational test program to verify the integrity of the system Include those transients introduced in systees other than the reactor coolant pressure boundary that will result in sienificant vibration response of reactor coolant pressure boundary systems 1
and components.
1 2
Discuss the testing procedures used in the design of Category I mechanical equip =ent such as fans, pumps, drives, valve operators and heat exchanger tube bundles to withstand seismic, accident and operational vibratory loading conditions, including the manner in which the methods and procedures erployed will consider the frequ 1
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spectra and amplitudes calculated to exist at the equipment supports i
Where tests or aaalyses do not include evaluation of the equipment in the operating mode, describe the bases for assuring that this i
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1 equipnent will function when subjected to seismic and accident loadings.-
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Provide a brief description of-the dynamic system analysis methods and procedures used to determine dynamic responses of reactor internals-and associated blass I components of the reactor coolant pressure boundary (e.g., analyses and tests).
The discussion should include the preoperationel' test program eierents described in Safety Guide 20, Vibration Measure =ents on Reactor Internals.
In the event
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elements of the prograc differ substantially from the require-cents of Safety Guide 20, the basis and justification for these differences should be presented.
4 Provide a discussion of the preoperaticnal analysis and testing results that will be used to augment the LOCA dynamic analysis methods and procedures, i.e., barrel ring and bean modes, guide-tube responses, water mass and compliance effects, damping factor selection, etc. ~~
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- s 3.9.2 ASME CODE CLASS 2 AND 3 COMPONENTS 1.
The FSAR' states that faulted operating condition categories have
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been applied to certain reactor = coolant systesa components.
Identify any other components or systems that are not a part of the reactor coolant pressure boundary for which the design stress limits associated with faulted conditions were applied.
If faulted conditions are used for such cases, then provide justification for applying such conditions, including the bases for the loading con '
'l ditions and' combinations, and associated design stress limits which were applied.
In addition, for all co=ponents and systems comparable to ASME Code Class 2 and 3, provide the design condition categories (normal, upset' or emergency), the associated design loading combinations and the-design-stress limits which will be applied for each loading combination.-
This information cay be submitted in tabular form as suggested below:
System Desien Loadi g Design Condition Design and/or Combinations Categories (Norr.al, Stress Component Upset, or E=crgency)
Limits l
2 If any design stress limits allow inelastic deformation (or are comparable tc the faulted condition limits defined in ASME Section l
111 for Class I components) then provide the bases for the use of l
ineisstic design limits by demonstrating that the co:ponent will j
maintain its functional or structural integrity under the specified i
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d design loading combination.
Include a brief descr!ption of the methods and design procedures that were used in such cases.
- 3. -Describe the design and installation criteria applicable to the counting of the pressure-relieving devices (safety valves and relief valves) for the overpressure protection of systems with Class 2 components.
In particular 'specify the design criteria used to take into account full discharge loads (i,+., - thrust, bending, torsion) imposed on valves.and on connected piping' in f
the event a? 1 the valves are required to' discharge.
Indicate the
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DESICI CRITERIA. METHODS AND PROCEDURES (REACTOR 5.2.1 Categorize all transients or combinations of transients listed in 1.
Table 4-8 of the FSAR with respect to the conditions identified as "na?:21% " upset", "energency", or " faulted" as defined in the Its addition, provide the ASNI Lection III Nucicar Conponent Ccde.
design loading co=binations and the associated stress or deformation criteria.
Table 4-20 of the FuAR includes inulted cond.'91on stress limits fo 2.
Describe the criteria employed to assure'that active
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co=ponents vill function as designed in the event of a postulated piping rupture (Faulted Conditica) within the reactor coolant pressure allevabic stress li=its established at or near the
'oundary (e.g.,
L'here c:pirical methods yield stress calculated on an clastic basis).
(tests) are c= ployed, provide a su==ary description of test methods, loading techniques and results obtained including the bases for extrapolations to ce=ponents larger or smaller than those tasted.
The design criteria which was used to account for full dischargs 3.
cending, torsion) inposed on safety and relief loads (i.e., thrust, valves and connected piping in the event all valves are requiref to discharge, including the provisien =ade to acco==odate these loads should be spec $fied.
- Active cosponents of a fluid sys!.em (e.g., valves, pumps) are those where oper ability is relied upon to perfor= a safety f unctica suc coolant pressure boundary.
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5.2.2 OVERPRESSURIZATION PEOTECTION 1
To-facilitate review of the bases for the pressure relieving capacity of the reactor coolant pressure boundary, submit (as-an appendix to the FSAR) the " Report on Overpressure Protection" that has been prepared in accordance with the requirements of th2 ASME Section III Nuclear Power Plant Components Code or, if the report is not available,- indicate the approximate date for submission.- In the event the report is not expected to lat available until either the Operating License review or late in the construction schedule for the plant, provide in the FSAR i
the bases and analytical approach (e.g., preliminary analyses) being utilized to establish the overpressure' relieving capacity required for the reactor coolant pressure boundary.
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