ML19316A640
| ML19316A640 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/27/1972 |
| From: | Maccary R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8001130169 | |
| Download: ML19316A640 (18) | |
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APR 2 7 m2 Richard C. DeYoung, Assistant Director for PWR's, Division of Peactor Licensing OCONEE NUCLEAR STATION, IJNITS 2 A'ID 3, DOCKL"I NOS and 50-287 Adequate responses to the enclosed request for additional infor=stion araThese required before 5,o can complete our review of the subject application.
requests, prepared by the DRS Hechanical Engineering Branch, concern the reactor coolant pressure boundary, reactor internal structures, safety related mechanical systems, seismic design criteria and pipe whip criteria submitted in Volumes I through IV of the FSAR.
The' applicant's description of seismic qualification testing for instru entation and equipment (Section 7A.2) appears to be extracted frem B & W Topical Report EAW-10003, " Qualification Testing of Protection System Instrumentation" (March 1971). The additional inforcation required to couplete our review of this report was forwarded in our request of March 13, 1972 for the Three Mile Island Station Unit No.1. Docket No. 50-289. Our review of the Oconee 2/3 application i
can therefore be expedited if the applicant vill agree to reference BAV-10003 l
and provide the requested information.
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The applicant has referenced (Section 14) B h W Topical Report BAW-10008, " Reactor Internals Stress and Deflections due to Loss-of-Coolant Accident and Maxinus 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 corjiction of this review.
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i r REQUEST FOR ADDITIO:L\\L INFORR\\ TION OCONEE NUCLEAR STATION UNITS 2 & 3 DOCKET NOS. 50-270/287 3.6 CRITERIA FOR PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH A LOCA 1. Provide a more detailed description of the ecasures that have been used to assure that the containment liner and all essential equip-ment within the containment, including components of the primary and secondary coolant syste=s, engineered safety features, and equipment supports, have been adequately protected against blow-I j down jet forces, and pipe whip resulting from a loss-of-coolant accident. The description should include: 1 a. Pipe restraint design require =ents to prevent pipe whip impact. I b. The features provided to shield vital equipment from pipe whip. c. The measures taken to physically separate piping and other cocponents of redundant engineered safety features. d. A description of the type of pipe whip restraints and the location of all restraints. 2 Describe the dynamic system analysis methods and procedures that were used to confirm the structural design adequacy of the reactor coolant system (unaf fected 'oop) and the reactor internal loadings. The 4 following information should be included: i ~ Y e m-l
-z i a, the locations of the postulated double ended pips rupture on 6 which dynamic analyses were based. b. the rupture type (s), such as circumferential and/or longitudinal break (s), for each postulated rupture location, the description of the forcing functions used for the pipe whip c. i dynamic analyses. The function should include direction, rise time, magnitude, duration and initial conditions. The forcing function should adequately represent the jet stream dynamics and the system pressure differences. d. a description of the cathematical nodel used for the dynamic
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analyses perforced to demonstrate that unrestrained motion of ruptured lines will not sever adjacent impacted piping or pierce impacted areas of containment liner. 4 e h + e 4 M. Dp 44%1 M
_ _.~ (; ~3~. i 3.7.1 SEIS'UC INPUT CRITERIA l. Provide the specific design condition (design carthquake or maximum hypothetical carthquake) which was applied to the critical damping factors for each item (except prestressed concrete structures) listed in Section SA.2.2 of the FSAR. l 1 2. Provide a list of all soil supported Category I and Category II i r structures and identify the depth of soil over bedrock for each 4 l structure listed, i i. 4 1 i i i I i 1 i { 4 ] ^1, s = s = g f,. o 8 3 '[ r {
I -4 3.7.2 SEIS:lIC SYSTEM MIALYSIS l '. Provide a sunnary of natural frequencies and response loads (in the form of critical node shapes, nodal responses, etc.) deternined 2 by the seismic system analysis. In addition, provide response i spectrum at critical plant equipnent elevations and points of support. 2. Provide the criteria used to lump masses for the seisnic system analyses (ratio of system nass and compliance to coeponent mass and compliance and the ratio of floor nass and compliance to I f supported equipment nass and compliance.) 3. Confirm the validity of a fixed base assunption in the nathenatical models for the dynteic sy; tem analyses by providing sunmary analyt-ical results that indicate that the rocking and translational responses are insignificant. A brief description should be included i of the method, mathenatical model and damping values (rocking verti-cal, translation and torsion) that have been used to consider the soil-structure interaction. l 4 Describe the measures taken to consider the effects on floor response spectra (e.g., peak width and period coordinates) of expected vari-ations of structural properties, danpings, soil properties, and soil-4 structure interactions. I 4 ~
5. Describe the method employed to consider the tornional modes of = vibration in the scismic analysis of the Category I building structures. If static factors are used to account for torsional accelerations in the seismic design of Category I structures, justify this procedure in lieu of a combined vertical, horizontal, and torsional nulti-mass system dynamic analysis. 6 The use of both the modal analysis response spectrum and time history methods provides a check on the responses at selected points in the station structure. Submit the responses obtained from both of these methods at selected points in the Category I structure to provide the basis for checking the seismic system analysis. 7. Provide the dynamic methods and procedures used to determine Category 1 structure overturning moments. Include a description of the procedures used to account for soil reactions and vertical earthquake effects. B. Provide the analysis procedure followed to account for the damping in different elements of the model of a coupled system. Include the criteria used to account for compcaite damping in a coupled system with dif ferent structural elements. y
3.7.3 SEISMIC SUBSYSTEM ANALYSIS 1. Describe the procedures used to account for the number of earth-quake cycles during one seismic event, and specify the number of maximum amplitude loading cycles for which Category I systens, components and equipment were designed. Revise Table 4-8 to reflect the number of design earthquake cycles. 2. Provide the criteria for combining modal responses (shears, noments, stresses, deflections, and/or accelerations) when modal frequencies j are closely spaced and a response spectrum nodal analysis method is used. 3. Provide the design criteria and analytical procedures applicable to piping that accounts for the relative displacements between piping support points, i.e., floors and components, at different elevations within a building and between buildings. 4. Submit the basis for the nethods used to determine the possible combined horizontal and vertical amplified response loading for the seismic design of piping and instrumentation, including the effect of the seismic response of the supports, equipment, and components. 5. Section 4B.3.2.1 of the FSAR refers to the term " root mean square" in describing a combination of modal responses. Confirn that the responses are conbined by the square-root-of-the-sum of the squares method. k 5
7-6 If static loads equivalent to the peak of the floor spectr um eurve were used for the seismic design of components and equip e t m n, the use of peak spectrum values should be justified by denonstrati ng that the contribution of all significant dynamic modes of response under seismic excitation have been included. 7. Provide the basis for simplified seismic analyses methods and procedures that were used for seismic design of structures systems and components including the criteria that was used to avoid the predominate input frequencies produced by the respon se of structures, supports and components to the earthquake input 8 Provide the criteria and procedures used to account for modal period variation in the mathematical models for Category I structures due to variations in material properties. 9. Provide the criteria employed to account for the torsional effe t A cs of valves and other eccentric masses (e.g., valve operators) in the seismic piping analyses. 10 With respect to Category I piping buried or otherwise located o t u-side of the containment structure, describe the seismic design criteria employed to assure that-allowable piping and structural stresses are not exceeded due to differential movement at support i points, at containment penetrations, and at entry points into other structures. b e i& 11. Describe the evaluation performed to determine seismic induced effects of Category II piping systens on Category I piping. 12 Provide the criteria enployed to determine the field location of seismic supports and restraints for Category I piping, piping system components, and equipment, ineluding placement of snubbers and dampers. Describe the procedures followed to assure that the field location and characteristics of these supports and restraining devices are consistent with the assumptions made in the dynamic i analyses of the system. 4 i i 4 f s
3.7.4 CRITERIA FOR SEIS?!IC INSTRUME?iTATION Provide the following information with respect to the use of the seismic instrumentation for the facility: 1 Discuss the seismic instrumentation provided and compare the proposed seismic instrumentation program with that described in AEC Safety Guide 12. " Instrumentation for Earthquakes." i Submit the basis and justification for elements of the proposed program that differ substantially from Safety Guide 12 2. Provide a description of the seismic instrumentation such as peak recording accelerographs and peak deflection recorders, that will be installed in selected Category I (Class 1 Seismic) structures and on selected Category 1 (Class 1 Seismic) components. Include the basis for selection of these structures and co=ponents the basis for location of the instrumentation, and the extent to which this instrumentation will be c= ployed to verify the seismic analyses following a seismic event. I 3 Describe the provisions that will be used to signal the control room operator the value of the peak acceleration level experienced \\; in the tendon access gallery of the reactor containment structure i to the control room operator within a few minutes after the carth-I { quake. Include the basis for establishing the predetermined values for activating the readout of the accelerograph to the control room operator. L e e 4
- 4 Provide the criteria and procedures that will be used to compare measured 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 l characteristics that may produce different dynamic responses of Category I (Class 1 Seismic) sta -tures at the site. t hk+ ' l
3.7.5 SEISMIC DESIG'l CO:iTROL MEASURES Describe the design control neasures (as specified in Appendix B - of 10 CFR Part 50 " Quality Assurance Criteria for Nuclear Power Plants") implenented to assure that appropriate seismic input data (as derived from seismic system and subsystem analyses including i any necessary feedback from such analyses) are correctly specified to the manufacturer of Category I components and equipment to constructors of other Category I structures and systems. The i responsible design groups or organizations that will verify the adequacy and validity of the analyses and tests employed by manufacturers of Category I co=ponents and equipment and con-structors of Category I structures and systems should be iden-tified. A description of the review procedures employed by each group or organization should be included. i m. 8.
g-. 3'.9.1 DYNAMIC SYSTDi A'IALYSIS MID TESTING 1. Paragraph 1701.5.4 of the ANSI B31.7 Nuclear Power Piping Code requires that piping shall be supported to prevent excessive 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 with'n the reactor coolant pressure boundary have been designed to withstand dynamic effects due to valve closures, punp trips, etc. Provide a list of the transient conditions and the associated actions (pump trips, valve actuations, etc.) that will be used in the vibration operational test program to verify the integrity of the system. Include those transients introduced in systems other than the reactor coolant pressure boundary that will result in significant vibration response of reactor coolant pressure boundary systems and components. 2. Discuss the testing procedures used in the design of Category I mechanicat equipment 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 employed will consider the frequency spectra and amplitudes calculated to exist at the equipment supports. Where tests or analyses do not include evaluation of the equipment ~ 's ~ 4 e 9 3
13 - \\ t h t this in the operating node, describe the bases for assuring t a i ic and accident equip:nent will function when subjected to se sm loadings. l i nethods Provide a brief description of the dynamic systen ana ys s ses of reactor internals 3. i and procedures used to determine dynanic responcoolant pressure and associated Class I components of the reactor The discussion should include boundary (e.g., analyses and tests). described in Safety Guide the preoperational test program elements In the event 20, Vibration Measurements on Reactor Internals. f h require-elements of the program dif fer substantially from t e i tification for these ments of Safety Guide 20, the basis and jus differ.nces should be presented. d testing Provide a dit 2ussion of the preoperational analysis an 4. alysis results that will be used to augment the LOCA dynamic an d beam modes, guide methods and procedures, i.e., barrel ring an ffects, damping factor tube responses, water mass and compliance e selection, etc. i I. I l l .~ e M e 1 f
3.9.2 ASitE CODE CLASS 2 AND 3 C0!!PC'IENTS 1. Provide for all Quality Groups B and C sy t b design condition categorier s ems and conponents, the (normal, upset or emernency), the asso iated design loading combinations and desi c-gn stress linits which vill be applied for each loading com'iination This information may be submitted in tabular form as suggested below: System and/or Design Loading Combinations Design Condition Component Categories (Normal. Design Stress Upset, or Emergenty) Limits i 2 The FSAR states that faulted operating condition categories have i been applied to certain reactor coolant system components. Identify any other components or systems that are nota part of the reactor coolant pressure boundary for which the design stress limits associated with faulted conditions were appliedIf faulted conditions are used for such cases, then provide j ustification for applying such conditions, including the bases for the loading con dicions and combinations, and associated d esign stress limits whics. vere applied. 1 3 If any design stress limits allow inelastic d f e ormation (or are comparable to the faulted condition limits defined i n ASME Section III for Class I components) then provide the bases for the use of inelastic design limits by demonstrating that th e component will maintain its functional or structural integrityunder the specified I ~, e
~ I i f 6 Include a brief description of I I design loading cocbination.ocedures that were used in suc h cases. the methods and design pr l tion criteria applicable to the Describe the design and instal aelieving devices (safet and i 4. mounting of the pressure-r re protection of systems with relief valves) for the overpressuIn particular, spec i criteria ds (i.e., thrust, Class 2 components. account full discharge loa used to take into connected piping in lves and on bending, torsion) imposed on vaired to discharge. Indicate the i event all the valves are requ acco=modate these loads. the provisions cade to J h i i l I .p M7
4 $.2.1 DESIGN CRITERIA, METHODS AND PROCEDURES (REACTOR COOLANT PRESSURE BOUNDARY) 1. Categorize all transients or combinations of transients listed in Table 4-8 of the FSAR with respect to the conditions identified as " normal", " upset", " emergency", or " faulted" as defined in the ASME Section III Nuclear Component Code. In addition, provide the design loading combinations and the associate d stress or deformation criteria. 2. Table 4-20 of the FSAR includes faulted condition stress limits for pump casings. Describe the criteria employed to assure that active
- cocponents will function as designed in the event c f a postulated piping rupture (Faulted Condition) within the reactor coolant pressure boundary (e.g., allowable stress limits established at or near the yield stress calculated on an elastf-basis). Where empirical methods (tests) are employed, provide a sunmary description of test methods, loading techniques and reeults obtained including the bases for extrapolations to corponents larger or smaller than those tested.
3. The design criteria which was used to account for full discharge loads (i.e., thrust, bending, torsion) imposed on safety and relief valves and connected piping in the event all valves are required to discharge, including the provision esde to accommodate these loads should be specified.
- Active components of a fluid system (e.g., valves, pumps) are those whose oper-ability is relied upon to perform a safety function such as safe shutdown of the reactor or mitigation of the consequences of a postulated pipe break in the reactor coolant pressure boundary.
i I 5 2.2 OVERPPISSURIZATICN PROTECTIOT To facilit.2te review of the bases for the pressure relieving 1 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 the ASME Section III Nuclear Power Plant Components Code or, is not available, indicate the approximate date if the report for submission. In the event the report is not expected to be available until either the Operating License review or late in 1 the construction schedule for the plant, provide La the FSAR the bases and analytical approach (e.g., preliminaty analyses) being utilized to establish the overpressure relieving capacity required for the reactor coolant pressure boundary. i I
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