ML20214G714
| ML20214G714 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 11/29/1971 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0154, CON-WNP-154 NUDOCS 8605220432 | |
| Download: ML20214G714 (20) | |
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Peter A. Morris, Director, Division of Esacter ' Licensing EANFORD NUCLEAR STATION, UNIT 2. DOCKET NO. 50-397 Adequate responses to the enclosed list of geestions prepared by the DRS Mechanical Engineering Branch are required before we een eenplete our review of the subject applicaties. These requests eencern the reactor internal structures, reactor coolant pressere boundary, esismic design criteria and pipe whip criteria submitted in Sections 3, 4, 5 and 12 and Appendix C of the PSAR.
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Edson C. Case, Director Division of Reactor Standards
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HANFORD NUC1. EAR STATION - UNIT 2 DOCKET No. 50-397 REQUEST FOR ADDITIONAL INFORMATION A.
AEC Croup A Systems (Reactor Coolant Pressure Boundary) 1.
The list of design transients for the reactor pressure vessel as specified in paragraph 4.1.5 of the PSAR appears to be incomplete.
Identify all design transients and their number of cycles, such as control system or other system malfunction, component malfunctions, transients resulting from any single operator error, etc., which are specified in the ASME Code required " Design Specifications" for all components of the reactor coolant pressure boundary. Categorize all transients or combination of transients with respect to the conditions a
identified as " normal", " upset", " emergency" or " faulted" as defined in the ASME Section III Nuclear Power Plant Components (1971) and Appendix C to the PSAR.
2.
Pumps and valves within the reactor coolant pressure boundary are classified ar either active or inactive components. Active components are required not only to serve a pressure-retaining function as in the 1/
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Active components are those whose operability is relied upon to perforn a safety function (as well as reactor shutdown function) during the transients or events considered on the respective operating condition categories.
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Inactive components are those whose operability (e.g., valve opening, or i
closure, pump operation or trip) are not relied upon to perform the system function during the transients or events considered in the respective operating l
condition categories.
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case of passive components, vessels, and piping but also to operate
,. reliably in order to perform a design safety function such as safe shutdown of the reactor or mitigation of the consequences of an hypothesized pipe break in the system. Therefore, to assure that t
active components will function as designed in the event of a pipe rupture (faulted condition) in the reactor coolant pressure boundary we consider stress limits associated with elastic action, i.e. stresses I
at or near yield stress, as appropriate in lieu of the code stress limits for the " faulted condition". Provide a list of active punps and valves as defined above and state whether it is your intention to comply with the limits indicated for active pumps and valves; justify any exceptions noted in your response.
3.
The stress and pressure limits specified in Paragraphs NB-3655 and NB-3656 of Section III, ASHE B & PV Code (1971) are considered appro-priate for emergency and faulted operating condition categories for pumps and valves within the reactor coolant pressure boundary. Indicate whether these stress limits for inactive components as defined in Ques-tion 2 above will be applied in the design of pumps and valves within the reactor coolant pressure boundary. If other stress criteria are proposed, provide the basis for their application.
4 Paragraph NB-3622.3 of the ASME Boiler and Pressure Vessel Code -
Section III requires that piping shall be supported to minimize vibragion and that the designer is responsible by observation under startup or 4
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-3 initial operating conditions to assure that vibration is within acceptable levels. Submit a discussion of your vibration oper-
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ational 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 i
valve closures, pump trips, etc. Provide a list of the transient conditions and the associated actions (pump trips, valve actu-ations, 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 dbat will result in significant vibration response of reactor coolant pressure boundary systems and com-ponents.
5.
Specify whethat the criteria to be employed in design against the effects of pipe rupture vill consider postulated pipe breaks to occur at any location within the reactor coolant pressure boundary or at limited areas within the system, (e.g., reactor coolant re-circulation line). Indicate whether these criteria take into account both longitudinal and circumferential pipe breaks and
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e provide the bases for your design approach.
6.
Provide a description of the measures that have been used to assure that the containment vessel and all essential equipment within the containment, including components of the reactor coolant prensure I
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4 boundary, engineered safety features, and equipment supporte, have beso adeguately protected against blowdown jet forces, and pipe whip. The description should includet a.
Pipe restraint design requirements to prevent pipe whip impact, b.
The features which will be provided to shield vital equipment from pipe whip.
The measures to be takea to physically separate piping and c.
other components of redundant engineered safety features.
7.
Describe the design and installation =riteria for the m3unting of the pressure-relieving devices (safety valves a d relief valves) within the peactor coolant pressure boundary. In particular, specify the design triteria khich will be used to take f ato t
scrount full discharge loads (i.e., thrust, bending, torsion) imposed on valves and on cennected piping in the event all the valves are required to discharge. Indicate the provisions made to accommodate these loads.
L 8
To facilitate our review of the bepes for the pressure relieving capacity of the reactor coolant pressure boundary, subnit tha
" Report on Overpressure Protectiou" which has been prepared in accordance with the requirer. ente of the ASME Section III Nuclest Power Plant Components Code or, if the report is not available at this time, indicats the approxinate date for submission. In the
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event you do not intend to submit the report until either the Operating License review or late in the construction schedule for this plant, provide the bases and analytical approach (e.g. pre-liminary analyses) being utilized to establish the overpressure relieving capacity required for the reactor coolant pressure boundary.
9.
Provide the design criteria, including design loading combinations, design condition categories (normal, emergency, and faulted) and design stress limits, which will be applied to the supports and restraints of components and piping of the reactor coolant pressure bounda ry.
Identify the applicable design codes which will be used.
. 5.
Reactor Internal Structures 1.
Certain of the limits for reactor internal structures in Paragraph C.5.2.1 either exceed comparable code allowable limits or are form-ulated in accordance with tbe failure criteria of the code. These limits are considered unacceptable for application to the desien of reactor internal structures unless they are qualified as out-lined in Table C.4.4, Amendment No.10 to the Limerick Generating Station (Docket Nos. 50-352 and 50-353).
2 Provide the basis for the derivation of the forcing functions which will be used in the design for the LOCA and normal reactor operation transients. Provide a brief description of the methods and procedures which will be used to determine reactor internals dynamic response.
a Include a list of all Class I reactor internals structural components and the associated dominant responses and forcing frequencies.
3.
Paragraph 3.1.6 states that a vibration analysis of the reactor vessel internals is' included in the design to eliminate failures caused by vibration. Provide the test data and supporting analyses which form the basis for the vibration response predictions or if the validity of the methods soployed cannot be demonstrated at this time, include in your response a statement of your intent to implement a preoperational test program which includes the measures given below:
1.
A vibration analysis and test program should be developed. The test program should be submitted for review by the Commission prior
to the performance of the scheduled preoperational functional tests.
The., f i g tion testing should be conducted with the fuel elements in the core structure of the reactor internals (or with dummy elements which provide equivalent mass and flow characteristics).
Testing may also be conducted with the core structure not loaded with fuel elemente provided such conditions can be demonstrated to result in vibrational characteristic which, for the purposes of the test, will yield conservative results. The testing may also be conducted both with and without the core structure loaded with fuel.
The test program should include a.
a brief description of the vibration test program, including instrumentation types and diagrams of their location, which will be used for measurement of vibration responses and those parameters which define the input forcing functions, b.
the planned duration of the test for normal operating modes to assure that all critical components are subjected to at least 10 cycles of vibration, c.
the additional test duration for other than normal operating modes to assure that the number of cycles imposed on the critical conponents is sufficient to analyse their adequacy to withstand vibrations under these operating modes,
d.
the description of different flow modes of operation and transients to which the internals will be subjected during the test, the predominant response mode shapes and the estimated e.
range of numerical values of the response of che major
, components of the reactor internals in terms of amplitudes and, where appropriate, the anticipated values of the parameters which ray influence the input forcing function, under those flow modes of reactor operation, which are shown by the analyses to be the meat critical, f.
the test acceptance criteria and the permissible deviations from these criteria, and the bases upon which these criteria were established, g.
a description of the inspection program which will be followed af ter the completica of the vibration tests, including the areas of reactor internals subject to examination, the method of examination, the design access provisions in the reactor internals and the specialized equipment to be employed for performing such examinations.
2.
A vibration test program should be implemented during the pre-operational functional testing program to measure the responsdL W
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Frequency and amplitudes of vibration, in terms of velocities, accelerations and displacements or strains.
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of the reactor internals and, where appropriate, the values of those parameters which will define the input forcing functions for the more critical modes of reactor operation. The data obtained by these measurements on reactor Luternals should be sufficient to verify that the cyclic stresses in the components, as determined by analyses of these data, are within the acceptable design stress limits set forth in the design specifications and applicable code requirements and that the results meet the acceptance criteria of the vibration test program.
3.
The extent of the measurements should be determined, for cach individual case, on the basis of the design and configuration of those structural elements of the reactor internals important to safety and their predicted behavior as determined from the vibration analyses used in their design. The type of vibration test instrumentation used, the number of measurements taken, and the distribution of measuring devices within the reactor should be adequate to detect the presence of lateral, vertical, and torsional amplitudes of vibration (e.g., beam, column, and shell modes of vibrations, as applicable to the geometry of the internals) and at sufficient locations to determine the points of predominant maximum vibratcry oscillations.
4 Af ter the reactor internals have been subjected to the significant flow modss expected during service lifetime under normal reactor 4
. operation, and other modes of reactor operation, visual and nondestructive surface examinations of reactor internals should be conducted to detect any evidence of the effects of vibrations.
These examinations should be conducted preferably following re-moval of the internals from the reactor vessel.
Where removal is not feasible, the examinations should be performed by means of examination equipment appropriate for in situ examination.
The areas examined should include all major load-bearing elements of the reactor internals which are relied upon to retain the core structure in place, the lateral, vertical, and torsional restraints provided within the reactor vessel, those locking and bolting devices whose failure could adversely affect the structural intee,rity a
of the internals, and those critical locations on reactor internal components as identified from the vibration analyses.
5.
In the event either the inspections of reactor internals reveal unacceptable defects or the results of the vibration test program fail to meet the specified acceptance criteria, a report should be prepared and submitted to the Commission for review, which includes an evaluation and a description of the corrective actions planned in order to justify the adequacy of the reactor internals design to withstand the vibrations expected in service.
6 If the test and examination program is acceptable, a summary of the results obtained from the vibration tests and inspection should be submitted to the Co= mission after completion of the tests.
The summary should includes a description of any differences from the specified vibration a.
test progran, instrumentation reading anomalies and instru=ent
- failures, b.
a comparison between the measured values of vibration responses
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2 from which input forcing functions are determined and the predicted valuee from the analysis.
This comparison should be cade for those components of the reactor internals for which the acceptance criteria under
- 3. f. have been established with respect to the different modes of vibration, an evaluation of measurements that exceeded acceptable limits c.
or of observations that were unanticipated, and the disposition of such deviations.
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Where measurements to determine forcing functions cannot be obtained practically in all areas by ceans of pressure transducers (or other instru=ents), such values may be estimated from measured responses and from analytical and empirical results.
C.
AEC Groups B & C Systens (Other Safety Related Fluid Systees and Comeonents) 1.
Provide for all Quality Croups B and C systems and components, the design condition categories (normal, upset or emergency), the assoc-1sted design loading combinations and design stress limits which will be applied for each loading combination. This information may be submitted in tabular form as suggested below System Design Loading Design Condition Design and/or Conbinations Categories (Normal, Stress Component Upset, or Emergency.)
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D.
Seismic System Desien Criteria and Analysis 1.
With regard to the development of system and equipment seismic design criteria by the time history methodsprovide plots that show a comparison of the smoothed site response spectra and the spectra derived from the earthquake records for all damping values which were used in the time history system analyses. Identify the system period intervals at which the response spectra acceleration values were calculated and demonstrate that the period interval used is sufficient to produce accurate spectra that do not deviate below the smooth response spectra for the site.
2 Because various assumptions are made regarding structure material a
properties and soil structure interaction, calculated periods of vibration are not exact. Describe the measures that will be taken to assure that the calculated response of Class I (seismic) structures by the normal mode response spectrum method will conservatively reflect the expected variations in the periods of vibration of the structures.
3.
With regard to the seismic analysis in Paragraph 12.3 of the PSAR, provide applicable stress or deformation criteria and descriptions (sketches) of the mathematical models used.
4 Describe the method which is employed to consider the torsional modes of vibration in the seismic analysis of the Class I building structures.
5.
Describe the procedures which will be used to account for the number of earthquake cycles during one seismic event, and specify the number
of loading cycles for which Class I systems and components will be designed for this event as determined from the expected duration of the seismic motions or the number of major motion peaks.
6 Submit the criteria to be employed for determination of the following seismic loadings (a) The possible combined horizontal and vertical amplified response loading for the seismic design of equipment and components, including the effect of the seismic response of the building and floors.
(b) The possible combined horizontal and vertical amplified response loading for the seismic design of piping and instrumentation, a
including the effect of the seismic response of the building, floors, supports, equipment, component, etc.
7 With reference to the seismic analysis of Class I items by the response spectrum method using floor response spectra, the shape of these floor response spectra derived from the time history method is dependent on the assumptions made for the structural properties, dampings, and soil structure interactions. Describe the measures which will be taken to consider the effcets on floor response spectra of expected variations in structure response.
8.
The use of both the modal analysis response spectrum and time history will provide a check on the response at selected points in the station structure. List the responses obtained f rom both methods at selected i
points in the Class I structure to provide the basis for checking the seismic system analysis.
9.
Provide the criteria which will be employed to account for the torsional effects of valves and other eccentric masses (i.e., valve operators) in the seismic piping analyses.
- 10. Provide the criteria which will be used with respect to basis for design to preclude failure of Class I piping systems by interaction of Class II piping systems.
11.
Indicate all Class I (Seismic Design) systems and components located within Class II (Scismic Design) structures and the measures which will be provided to ensure their continued safety function under seismic loading.
12 With respect to Class I (Seismic Design) piping buried (e.g., emergency cooling water piping and conduit) or otherwise located outside of the containment structure, describe the seismic design criteria which assures that allowable piping and structural stresses are not exceeded due to dif ferential movement at support points, at contain-ment penetrations, and at entry points into other structures, 13.
Indicate the provisions which will be applied to assure that the crane located in the reactor buildings will be held on rails to preclude its dislodgement during seismic excitation.
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14 The list of structures and systems that have been designated as seismic Class I in Appendix C of the PSAR does not include that portion of the main steam system extending from the outermest containment isolation valve up to the turbine casing and connected piping inclusive of the first valve (either normally closed or capable of automatic closure), nor does it include the radioactive waste storage treatment, handling, and disposal systems. The proposed Class Il classification as presently specified for these systems is not acceptable. Indicate the extent to which these systems or portions thereof will be reclassified as sesimic Class I.
15.
Describe the seismic design criteria which will assure the adequacy of Class I mechanical componcnts such as pumps, heat exchanc,ers, and electrical equipment such as cable trays, battery racks, instrument racks and control consoles. Describe the measures to be taken for seismic restraint to meet these criteria, the analytical or testing methods to verify the adequacy of these restraints and the methods which will determine the seismic input to these components.
16.
The list of damping coefficients in Table 12.3-1 does not include values for vital piping. Revise Table 12.3-1 to include the darping coefficients to be used in the seismic analyses of piping.
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E.
Seismic Instrumentation With respect to seismic instrumentation, submit a statement of your intent to implement a program such as described in AEC Safety Guide 12, Instrumentation for Earthquakes (March 10, 1971). Submit the basis and justification for elements of the proposed program which differ substantially from Safety Guide 12 o
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F.
Seismic Quality Assurance Describe the design control measures which will be instituted to assure that adequate seismic input, including any necessary feedback from structural and system dynamic analyses is specified to vendors of purchased Category I components and equipment. Identify the responsible design groups or organizations who will assure the adequacy and validity of the analyses and tests employed by vendors of Category I components and equipment. Provide a description of the review procedures to be utilized by each group or organization, n
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