ML19221B042
| ML19221B042 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-03.9.2, NUREG-75-87, NUREG-75-87-3.9.2, SRP-03.09.02, SRP-3.09.02, NUDOCS 7907120297 | |
| Download: ML19221B042 (16) | |
Text
a asc NU REG-75/087 o
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. 4 U.S. NUCLEAR REGULATORY COMMISSION 1M S
O:
STANDARD REVIEW PLAN
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OFFICE OF IVUCLEAR REACTOR REGULATION SECTION 3.9.2 DYNAMIC TESTING AND ANALYSIS OF SYSTEMS, COMPONENTS, AND EQUIPMENT REVIEW RESPONSIBILITIES Primary - Mechanical Engineering Branch (MEB)
Secondary - Reactor Systems Branch (RSB)
I.
APiAS OF REVIEW MEB reviews the criteria, testing procedures, and dynamic analyses employed to assure the structural and functional integrity of piping systems, mechanical equipment, reactor internals, and their supports under vibratory loadings, including those due to fluid flow and postulated seismic events to assure conformance with General Design Criteria 1, 2, 4, 14, and 15.
The staff reviaw covers the following specific areas:
1.
Piping vibration, thermal expcnsion, and dynamic effect testing should be conducted during startup testing. The systems to be monitored should include (a) ASME Code Class I, 2, and 3 systems, (b) other high energy piping systems inside Seismic Category I Structures, (c) high energy portions of systems whose failure could reduce the functioning of any Seismic Category I plant feature to an unacceptable safety level, and (d) Seismic Category I portions of moderate-energy piping systems located outside containment. The supports and restraints necessary for operation during the life of the plant are considered to be parts o' the piping system. The purpose of these tests is to confirm that these piping systems, restraints, compo-nents, and supports have been adequately designed to withstand flow-induced dynamic loadings unde' the steady-state and operational transient conditions anticipated during service and to confirm that normal thermai motion is not restrained. The test program description should include a list of different flow modes, a list of selected locations for visual inspections and other measurements, the acceptance criteria, and possible corrective actions if excessive vibration or indications of normal thermal mot;or restraint occurs.
2.
Seismic qualification testing of safety related mechanical equipment is required to assure its ooility to function during and after a postulated seismic occurrence. At the const ruction permit (CP) stage, the staff review covers the following specific areas:
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a.
Tu criteria for seismic qualification such as the deciding factors for choos-ing test or analysis, the o msiderations defining the input motion, and the steps to demonstrate adequacy of the seismic qualification program.
b.
The methods and procedures used to assure seismic Category I mechanical equip-ment operability during and after the safe shutdown earthquake (SSE), and to assure structural and functional integrity of the equipment after several occurrences of the operating basis earthquake. Included are mechanical equip-ment such as fans, pump drives, he?t exchanger tube bundles, valve actuators, battery and instrument rat h, control consoles, cabinets, panels, and cable trays.
c.
The methods and procedures of analysis or testing for the supper.s for the seismic Category I mechanical equipment listed above, and the procedures u ad to account for the possible amplification of loads (amplituae and frequency content) unc'er seismic conditions.
At the operating licen 9 (JL) stage, the staff reviews the results of tests and analyses to assure th-proper implementation of the criteria established in the CP review, and to dem. 'trate adequate seismic qualification.
3.
Dynamic responses of structural components within the reactor vessel caused by steady-state and operational flow transient conditions should be analyzed fcr proto-type (first of a design) reactors. Generally, this analysis is not roouired for non prototypes t < cept that segments of an analysis u y be necessary if there are substantial deviations from the prototype internals design. The purpose of this analysis is to predic+. the vibration behavior of the components, so that the input forcing functicr.s and the level of response can be estimated. Before conducting the an..iyses, the specific locations for calculated responses, the considerations in defining the mather:atical models, the interpretation of analytical results, the acceptance criteria, and the methods of verifying predictions by means of tests should be determined. If the reactor internal structures are a non prototype design, reference should be made to the results of tests and analyses for the prototype reacter and a brief summary of the results should be given.
4.
Flow-induced vibration testing of reactor internals should be conducted during the preoperational and startup test program. The purpose of this test is to demonstrate that flow-induced vibrations similar to those expected during operation will not cause unanticipated flow-induced vibrations of significant magnitude or structural damage. The test program description should include a list of flow modes, a list of sensor types and locations, a description of test procedures and methods to be used to process and interpret the measured data, a description of the visual inspections to be made, and a comparison of the test results with the analytical predictions.
If the reactor internal structures are a non prototype design, reference should be
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Rev. 1 3 4 2-2
made to the results of tests and analyses for the prototype reactor ind a brief summary of the results should be given.
5.
Dynamic system analyses should be performed to confirm the structural design adequacy and ability, with no loss of function, of the reactor internals and '.nbroken loops of the reactor coolant piping to withstand the loads from a loss of coolant accident (LOCA) in combination with the SSE. The staff review covers the metiods of analysis, the considerations in defining the mathematical models, the descriptions of the forcing functions, the calculational scheme, the acceptance criteria and the inter-pretation of analytical results.
6.
A discussion should be provided which describes the methods to be used to correlate results from the reactor internais vibration test with the analytical results from dynamic analyses of the reactor internals under steady-state and operational flow transient conditions.
In addition, test results from previous plants of sic.ilar characteristics may be used to verify the mathematicai models used for the loading condition of LOCA in combination with the SSE by comparing such dynamic characteristics as the natural frequencies. The staff review covers the methods to be used for comparison of test and analytical results and for verification of the analytical models.
Computer programs used in the analyses discussed in this plan are reviewed in accordance with SRP Section 3.9.1.
The R3B verifie; on request that (1) the various flow modes to be used to conduct the vibratica test of the reactor internals are representative of the steady-state and operational transient conditions anticipated for the reactor curing its service, and (2) that an acceptable hydraulic analysis has been used to determine the loads acting on the reactor coolant system piping and the reactor internals.
II.
ACCEPTANCE CRITERIA To fulfill in part the design requirements for safety related structures, systems, and components set forth in General Design Criteria 1, 2, 4, 14, and 15, the acceptance criteria for the areas of MEB review are as follows:
1.
Vibration, thermal expansion, and dynamic effects testing should be conducted during startup functional testing for specified high-and moderate-energy piping, and their supports and restraints. The purport af these tests is to confirm that the piping, components, restraints, and supports have been designed to withstand the dynamic loadings and operational transient conditions that will be encountered during service as required by the Code and to confirm that no unacceptable restraint of normal thermal motion occurs.
146 284 3.9.2-3
An acceptable test program to confirm the adequacy of the designs should consist of the following:
a.
A list of systems that will be monitored.
b.
A listing of the different flow modes of operation and transients such as pump trips, valve closures, etc. to which the components will be subjected during the test. (For additional guidance see Reference 8.)
For example, the tran-sients associated with the reactor cooiant system heatup tests should include, but not necessarily be limited to-(1) Reactor coolant pump start.
(2) Reactor coolant pump trip.
(3) Operation of pressure-relieving valves.
(4) Closure of a +urbine stop valve.
c.
A list of selected locations in the piping system at which visual inspections and measurements (as needed) will be performed during the tests. For each of these selected locations, tha o?flection (peak-to peak) ( r other appropriate criteria, to be used to show that the stress and fatigue limits are within the desiqq levels, should be provided.
d.
A list of snubbers on systems which experience soff uient thermal movement to measure snubber travel from cold to hot positior e.
A description of the thermal motion monitoring p.
i.e.,
verification of snubber movement, adequate clearances and gaps, ir,cluding acceptance criteria and how motion will he measured.
f.
If vibration is noted beyond the acceptance levels set by the criteria of c.,
above, corrective restraints should be designed, incorporated in the piping system analysis, and installed.
If, during the test, pip:ng system restraints are determined to be inadequate or are damaged, corrective restraints should be installed and another test should be performed to determine that the viorations have been reduced to an acceptable level. If no snubber piston travel is measured at those stations indicated in d., above, a description should be provided of the corrective action to be taken to cssure that the snubber is operable.
2.
A test program is required to confirn +he ability of all seismic Category I mechan-ical equipment to function as needed during and after an earthquake of magnitude up to and including the SSE.
1/6 285 k
Rev. 1 3.9.2-4 1
a.
Analysis without testing is acceptable if structural integrity alone can assure the intended function. When a complete seismic test is impracticat,le, a combina-tion of test and arialysis is acceptable.
b.
Equipment should be tasted in the operational condition. Loadings simulating those of plant normal operation, such as thermal and flow-ir loadings, if any, should be concurrently superimposed upon the seismic lo r Operability should be verified during and after the test.
c.
The characteristics of the seismic input motion should be specified by one of the following:
(1) Response spectrum.
(2) Power spectral density function.
(3) Time history.
Such characteristics, as deriveu from the structure or system seismic analysis, should be representative of the seismic input motion at the equipment mounting locations.
d.
The test input motion should be characterized in the same manner as the seismic input motion, and the conservatism in amplitude anc frequency content should be demonstrated.
e.
Seismic excitations generally have a broad frequency content. Multi-Frequency l
input motion should be used in the testing. However, single frequency input, such as sine " beats," may be applicable provided one of the following condi-tions are met:
(1) The characteristics af the seismic input motion indicate that the motion is dominated by one frequency (e.g., by structural filtering effects).
(2) The anticipated response of the equipment is adequately represented by one mode.
(3) The test input motion has sufficient intensity and duration to excite all modes to the required amplitudes, such that the testing response spectra will envelope the corresponding response spectra of the individual modes.
The test input motion should be applied to one vertical axis and one principal horizontal axis (or two orthogonal horizontal axes) simultaneously unless it 9
can be femonstrated that the equipment response in the vertical direction is not '.isitive to the vibratory motion in the horizontal directicn, and vice 146 286 3.9.2-5 Rev. 1
versa. The time phasing of the inputs in the vertical and horizontal direc-tions must be such that a purely rectilinear resultant input is avoided. An acceptable alternative is to have vertical and horizontal inputs in phase, and then repeated with inputs 180 degrees out-of phase. In addition, the test must a" repeated with the equipment rotated 90 degrees horizontally.
g.
Dynamic coupling between the equipment and related sytems, if any, such as cor ected piping and other mechanical components, should be considered.
h.
The fixture design should meet the following requirements:
(1) Simulate the actual service mountir.g.
(2) Cause no extraneous dynamic coupling to the test item.
i.
The in situ application of vibratory devices to superimpose the seismic vibra-tory loadings on a complex active device for operability testing is acceptable if it is shown that a meaningful test can be made in this way.
i.
The test program may be based upon selectively testing a representative number of mechanical components according t-
- ype, load level, size, etc., on a proto-type basis.
k.
Analyses or tests should be performed for all supports of rechanical equipment to assure their structural capability to withstand seismic excitation. The analytical results must include the following:
(1) The required input motions to the mounted equinment should be obtained and characterized in the manner as stated in subsection 2.c, above.
(2) The combined stresses of the support structures should be within the limits of the ASME Ccde, Subsection NF, " Component Support Structures."
1.
Supports should be tested with equipment installed or with an equivalent mass that simulates the equipment dynamic coupling to the support. If the equipment is installed in a nonoperating conditin for the support test, the response at the equipment mounting location should be characterized in the manner as stated in subsection 2.c, above. In such a case, the equipment should be tested separately for operability and the actual input to the equipment should be more conservative in amplitude and frequency content than the.aonitored response.
m.
The requirements of subsections 2.c, 2.d, 2.e, 2.f, and 2.h, above, are applic-able when tt i conducted on equipment supports.
Rev. 1 3.9.2-b
3.
The following guidelines, in addition to Regulatory Guide 1.20 (Reference 7), apply to the analytical solutions to predict vibrations of reactor internals for prototype plants. Generally, this analysis is required only for prototype designs.
a.
The results of vibration calculations for a prototype reactor should consist of the following:
(1) Dynamic responses to operating transients at critical locations of the internel structures should be determined and, in particular, at the loca-tions where vibration sensors will be mounted on the reactor internals.
For each location, the maximum response, the modal contribution to the total response, and the response causir] the maximum stress amplituae should be calculatEi (2) The dynamic propertie.,.. internal structures, including the natural frequencies, the dominant mode shapes, and the damping factors should be characterized. If analyses are performed on a component structural element basis, the existence of dynamic coupling among comoonent structure elements should be investigated.
(3) The response characteristics, such as the dependence on hydrodynamic excitation forces, the flow path configuration, coolant recirculation pump frequencies, and the natural frequencies of the internal structures, should be identified.
(4) Acceptance criteria for allowable responses should be established, as should criteria for the location of vibration sensors. Such criteria should be related to the Code allowable stresses, strains, and limits of deflection that are established to preclude loss of function with respect to the reactor core structures and fuel assemblies.
o.
The forcing functions should account for the effects of transient flow condi-tions and the frequency cc it.
Acceptable methods for formulating forcir.g functions for vibraticn pred.ction include the following:
(1) Analytical method: based on standard hydrodynamic theory, the governing differen'ial equations for vibratory motions should te developed and solutions obtained with appropriate boundary conditicns and parameters.
This method is accejtabla ',h?re the geometry along the fluid flow paths is mathematically tractable.
(2) Test-analysis combination method based on data obtained ' rom plant tests or scaled model tests (e.g.
velocity or pressure distribution data), forcing functions should be formu.ated which will include the effects of complex flow path configurations and wide variations of pressure distributions.
146 288 3.9.2-7 Rev.
(3) Response-deduction method: based on a derivation of response character-istics from plant or scaled model test data, forcing functions should be formulated. However, since such functions may not be unique, the computa-tional procedures and the basis for the selection of the representative forcing functions should be described.
Acceptable methods of obtaining dynamic responses for vibratiori predictions are c.
as follows:
(1) Force-response computations are acceptable if the characteristics of the forcing functions are predetermined on a conservative basis and the mathe-matical model of the reactor interna!s is appropriately representative of the design.
(2) If the foning functions are not predetermined, either a special analysis of the -
anse signals measured from reactor internals of simiiar design
,aay ' r *ormed to predict amplitude and modal contributions, or param-eter stuu.es useful for extrapolating the results from tests of internals or components of similar designs based t.n composite statistics may be used.
d.
Vibration predictions should be verified by test results. If the test results differ substantially from the predicted response behavior, the vibration analysis should be appropriately modified to improve the agreement with test results and to validate the analytical method as appropriate for predicting responses of the prototype unit, as well as of other units where confirmatory tests are to be conducted.
4.
The preoperational vibration test program for the internals of a prototype (first of a design) reactor should conform to the requirements for a prototype test, as speci-fied in Regulatory Guide 1.20, including vibration prediction, vibration monitoring, l data reduction, and surface inspection. The test program should include, but not necessarily be limited to the following:
a.
The vibration testing should be conducted with the fuel elements in the core or with dummy elements which provide equivalent dynamic effects and flow character-istics. Testing without fuel elements in the core may be acceptable if it can be demonstrated that testing in this mode is conservative.
b.
A brief description of the vibration monitoring instrumentation should be provided, including instrument types and diagrams of locations, which should include the locations having the riost severe vibratory motions or having the most effect on safety functions.
lev. I 3.9.2-8 146 289
c.
The planned duration of the test for the normal operation modes to assure that 6
all critical components are subjected to at least 10 cycles of vibration l
should be provided. For instance, if the lowest response frequency of the core internal structures is 10 Hz, a total test duration of 1.2 days or more will be l acceptable.
d.
Testing should include all of the different flow modes of normal operation and upset transients. The proposed set of flow modes are acceptable if they provide a conservative basis for dr termining the dynamic response of the reactor inter-nals and are reviewed by RSB on request.
e.
The methods and procedures to be used to process the test data to obtain a meaningful interpretation of the core structure vibration behavior should bq provided. Vibration interpretation should include the amplituda, frequency content, stress state, and the possible effects on safety functions.
f.
Vibration predictions, test acceptance criteria and bases, and permissible deviations from the criteria should be provided before the test.
g.
Visual and nondestructive surface inspections should be performed after the completion of the vibration tests. The inspection program description should include the areas subject to inspection, the methods of inspection, the design access provisions to the reactor internals, and the equipment to be used for performing such inspections. These inspections should be conducted preferably following the removal of the internals from the reactor vessel. Where removal is not feasible, the inspections should be performed by means of equipment appropriate for in situ inspection. The areas inspected should include all load-bearing interfaces, core restraint devices, high stress locations, and locations critical to safety functions.
For internals of subsequent reactors that have the same design, size, configuration, and operating conditions as the prototype reactor internals, the vibration test program should conform to the requirements of the appropriate non prototype program as specified in Regulatory Guide 1.20.
5.
Dynamic system analyses should be performed to confirm the structural design ade-quacy of the reactor internals and the reactor coolant piping (unbroken loops) to withstand the dynamic loadings of the most severe LOCA in combination with the SSE.
Where a substantial separation between the forcing frequencies of the LOCA (or SSE) loading and the natural frequencies of the internal structures can be demonstrated, the analysis may treat the loadings statically.
p l
The most severe dynamic effects from LOCA loadings are generally fcund to result from a postulated double-ended rupture of a primary coolant loop near a reactor vessel inlet or outlet nozzle with the reactor in the most critical normal operating 146 290 3.9.2-9 Rev. 1
mode. However, all other postulated break locations should be evaluated and the location producing the controlling effects should be identified.
Mathematical models use = for dynamic system analysis for LOCA in combination with the SSE effects should include the following:
Modeling should include reactor internals and dynamically related piping, pipe a.
supports, components, and fluid-structure interaction effects when applicable.
Typical diagrams and the basis of modeling should be developed and described.
b.
Mathematical models should be representative of system structural characteris-l tics, such as the flexibility, mass inertia effect, geometric configuration, and damping (including possible coexistence of viscous and Coulomb damping).
Any system structural partitioning and directional decoupling employed in the l
c.
dynamic system modeling should be justified.
d.
The effects of flow upon the mass and flexibility properties of the system should be discussed.
Typical diagrams and the basis for postulating the LOCA-induced forcing function should be provided, including a description of the governing hydrodynamic equations and the assumptions used for mathematically tractable flow path geometries, tests for determining flow coefficients, and any semiempirical formulations and scaled model flow testing for determining pressure differentials cr velocity distributions.
The acceptabilty of the hydraulic analysis, as reviewed by RSB on request, is based on established engineering practice and generic topical reviews performed by the staff.
The methods and procedures used for aynamic system analyses should be described, including the governing equations of motion and the computational scheme used to derive results. Time comain forced-response computation is acceptable for both LOCA and SSE analyses. The response spectrum modal analysis method may be used for SSE analysis.
The stability of elements in compressica, such as the core barrel and the control rod guide tubes under outlet pipe rupture loadings should be investigated.
Either response spectro or time histories may be used for specifying seismic input motions of the SSE at the reactor core supports.
The criteria for acceptance of the analytical results are as provided in SRP Sections 3.9.3 and 3.9.5.
\\kb Rev. 1 3.9.2-10
6.
Regarding the correlation to be made of tests and analyses of reactor internals, a discussion covering the following items should be provided:
a.
Comparison of the measured response frequencies with the analytically obtained natural frequencies of the reactor internals for possible verification of the mathematical model used in the analysis.
b.
Comparison of the analytically obtained mode shapes with the shape of measured motion for possible identification of the modal combination or verification of a specific mode.
c.
Comparison of the response amplitude timo variation and the frequency content obtained from test and analysis for possible verification of the postulated forcing function.
d.
Comparison of the maximum responses obtained from test and analysis for pos-sible verification of stress levels.
Comparison of the mathematical model used for dynamic system analysis under e.
operational flow transients and under the combined LOCA and SSE loadings, to note similarities.
III. REVIEW PROCEDURES The reviewer will select and emphasize material from the procedures described below, as may + 3ppropriate for a particular case.
Genera Design Criteria 1, 2, 4, 14, and 15 state that all structures, system and compo-nents i1portant to safety should be designed and tested to assure that safety functions can be ;erformed in the event of operational transients, earthquakes, and LOCA loadings.
Under trese guidelines, the staff reviews the treatment of dynamic responses of safety-relatec piping systems and reactor internal structures by the following procedures:
1.
D. ring the CP stage, the PSAR is reviewed to assure that the applicant has provided a commitment to conduct a piping steady-state vibration, thermal expansion and ope ational transient test program. The applicants program description should be suffi:iently comprehensive to contain all the elements of an acceptable program as described in subsection II.l.
During the OL stage, the FSAR is reviewed to assure that the applicants PSAR commit-ment is fulfilled and the program is developed in sufficient detail. The reviewer should be assured that the applicants program is sufficiently developed to:
146 292 Rev. 1 3.9.2-11
(a) Establish the rationale and bases for the acceptance criteria and selection of locations to monitor pipe motions.
(b) Provide the displacement or other appropriate limits at locations to be monitored.
(c) Describe the techniques and instruments (as needed) for monitoring or measuring pipe motions.
(d) Assure that the NRC will be provided documentation of any corrective action resultirg from the test and conformation by additional testing that substan-tiates effectiveness of the corrective action.
2.
At the CP stage, the staff reviews the program which the applicant has described in the preliminary safety analysis report (PSAR) for the seismic qualification of all seismic Category I mechanical equipment. The program is measured against the require-ments listed in the acceptance criteria subsection of this SRP section. Of particular interest ar. the proper use of test and analytical procedures. Equipment which is too complex for reliable mathematical modeling should be tested unless the analytical l
procedures and corresponding design are demonstrated to be conservative. Both the Mr assurance that all important modes of test and the analysis methods are review e response have been excited is tests or cons.uered in analyses. Proper application of test input motions so as to er.velop the required input, whether in terms of response spectra, power spectral density, or time history, and in all necessary directions, is verified. The use or treatment of supports is also reviewed.
At the OL stage, the staff reviews the program a vin as described by the applicant in the Final Safety Analysis Report (FSAR). In addition, the FSAR is reviewed for documentation of successful implementatian of the seismic qualification program, includir,g test and analysis results. Also, the acceleration levels used in the tests and in the analyses are reviewed for assurance that they equal or exceed the acceleration at the equipment mounting locations derived from structural response studies of the plant structure as built or as designed.
3.
At the CP stage, the applicant should commit to performing an analysis of the vibra-tion of the reactor internal structures if they are designated as a prototype design.
A brief description of the methods and precedures to be used for the analysis should be provided.
At the OL stage, a detailed dynamic analysis should be provided for a prototype design, to be used for vibration prediction prior to the performance of preopera-tional vibration tests. Acceptance of the analysis is based on the technical sound-ness of the analytical method and procedures used and the degree of conformance to the acceptance criteria listed above.
In addition, the analysis is verified by correlation with the test results when these are available.
Rev. 1 3.9.2-12 146 293
For both CP and OL stages, if the reactor internal structures are a non prototype design, then reference should be made to the reactor which is prototypical of the reactor being reviewed. A brief summary of test and analysis results for the proto-type should be given, Alternatively, the information rray be contained in another applicable document, such as a tcpical report, to which reference should be made.
4.
At the CP stage, the staff review of the program for preoperational vibration test-ing of reactor internals for flow-induced vibrations includes the following matters:
The applicant should clarify his intention to perform either a prototype test a.
or a non prototype test.
l b.
If the plant is designated as a prototype, a brief description of the preopera-tional vibration test program should be provided. The staff review will be based on the conformance of this program to the requirements as listed in subsection II.4, above.
If the piant is a non prototype, the applicant should ioentify the e>isting l
c.
plant of similar design that is the prototype plant. The staff reviews the validity of the designated prototype, including any design difference of reactor internal structures from the prototype plant to verify that any design modifica-tions do not substantially alter the behavior of the flow transients and the response of the reactor internals. Additional detailed analysis, scaled model tests, or installation of some instrumentation.during the confirmatory test may be required in order to complete tha review. In addition, the applicant should commit to performing the prototype test if adequate test results are not obtained on a timely basis for the designated prototype.
At the OL stage, the staff review includes the following procedures:
A detailed preoperational vibration test program and the tentative schedule to a.
perform the test are reviewed. If elements of the program differ substantially from the guidelines specified in Regulatory Guide 1.20, discussion of the need and justification for the differences rhould be given. On request, R$8 verifies that the flow modes to be used are :ceptable.
b.
Fcr a prototype plant, the review covers the acceptability of vibration predic-tion, the visual surface inspec6 ion procedures, the details of instrumentation for vibration monitoring, the methods and [ ocedures to process the test results, and possible supplementary tests, auch as component vibration tests, flow tests, and scaled model tests.
For a non pretotype plant, the staf f verifies the applicability of tr.e desig-c.
nated prototype, including the design similarity of the reactor internal struc-tures to the prototype. Additional detailed analysis, scaled model tes s, o f
3.9.2-13 Rev. 1
vibration monitoring in the confirmatory tests may be needed in order to complete the review.
5.
In the CP stage review of the dynamic analysis of the reactor internals and unbroken loops of the rector coolant piping under faulted condition loadings, the applicant commits to perform this analysis or identifies the applicable document, generally in form of a topical report, containing the required information. A brief description of the scope and methods of analysis should be provided.
In the OL review, the staff reviews the detailed information to confirm that an adequate analysis has been made of the capability of reactor internal structures and unbroken loops to withstand dynamic loads from the most severe LOCA in combination with the safe shutdown earthquake. The staff review covers the analytical methods and procedures, the basis of the forcing functions, the mathematical models to represent the dynamic system, and the stability investigations for the core barrel and essential compressive elements. Acceptance of the analysis is based on (1) the technical sov.dness of the analytical methods used, (2) the degree of conformance to the acceptance criteria listed above, and (3) verification that stresses under the combined loads are within allowable limits of the applicable code and deformations are within the limit set to assure the ability of reactor internal structures and piping to perform netded safety functions. On request, RSB verifies that an acct,.
able hydraulic analysis has been used.
6.
MEB reviews the program which the applicant has committed to implement as part of the preoperational test procedure, principally to correlate the test measurements with the analytically predicted flow-induced dynamic response of the reactor inter-nals. ME8 reviews the applicant's statements ir, this areas to assure that there is a commitment to submit a report on a timely basis. The report should summarize the analyses and test results so that MEB can review the compatibility of the results from tests and analyses, the consistency between mathematical models used for dif-ferent loadings, and the validity of the interpretation of the test and analysis results.
IV.
EVALUATION FINDINGS The reviewer verifie. that sufficient information has been provided and that the review supports conclusions of the following type, to be included in the staff's safety evalua-tion report:
"The vibration, thermal expansion, and dynamic effects test progrem which will be conducted during startup and initial operation on specified high-and moderate-energy piping, and all associated systems, restraints and supports is an acceptable program. The tests provide adequate assurance that the piping an )iping restraints of the system have been designed to withstand vibrational dynam-effects due to valve closures, pump trips, and other ope,ating modes associated with the design Rev. 1
,,g,p_14 146 29c3
basis flow conditions In addition, the tests provide assurance that adequate clearances and free mu<ement of snubvers exist for unrestrained thermal movement of piping and supports during normal system heatup and cooldown operations. The planned 9
tests will develop loads similar to those experienced during reactor operation.
Compliance with this test program constitutes an acceptable basis for fulfilling, in part, the requirements of General Design Criteria 14 and 15.
"The capability of safety-related mechanical equipment to perform necessary protec-tive actions in the event of a safe shutdown earthquake (SSE) is essential for plant safety. The qualification testing program which will be implemented for seismic Category I mechanical equipment provides adequate assurance that such equipment will function properly under the loads from vibratory forces imposed by the safe shutdown earthquake and under the conditions of post-earthquake operation. This program constitutes an acceptable basis for satisfying, in part, the requirements of General Design Criterion 2.
"The preoperational vibration prngram planned for the reactor internals provides an acctptable basis for verifying the design adequacy of these internals under test loading conditions comparable to those that will be experienced during operation.
The combination of tests, predictive analysis, and post-test inspection provide adequate assurance that the reactor internals will, during their service lifetime, withstand the flow-induced vibrations of reactor operation without loss of struc-tural integrity. The integrity of the reactor internals in service is essential to assure tha proper positioning of reactor fuel assemblies and unimpaired operation of the control rod assemblies to permit safe reactor operation and shutdown. The conduct of the preoperational vibration tests is in conformance with the provisions of Regulatory Guide 1.20 and constitutes an acceptable basis for demonstrating design adequacy of the reactor internals, and satisfies the applicable requirements of General Design Criteria 1 and 4 "The dynamic system analysis to be performed provides an acceptable basis for con-firming the structural design adequacy of the reactor internals and unbroken piping loops to withstand the combined dynamic loads of postulated loss of coolant acci-dents (LOCA) and the safe shutdown earthquake (SSE) and the combined loads of a postulated main steam line rupture and SSE (for a BWR). The analysis provides adequate assurance that the conbined stresses and strains in the components of th<
reactor coolant system and reactor internals will not exceed the allowable design stress and strain limits for the materials of construction, and that the resulting deflections or displacements at any structsral elements of the reactor internals will not distort the reactor internals geometry to the Extent that core cooling may be impaired. The methods tsed for compcnent analysis ha 7 been found to be compat-ible with those used for the systems analysis. The proposed combinations of compo-nent and system analyses are, therefore, acceptable. The assurance of structural integrity of the reactor internals under LOCA conditions for the most adverse postu-lated loading event provides added confidence that the desiga will withstand a 146 296 3.9.2-15 Rev. I
spectrum of lesser pipe breaks and seismic loading events. Accomplishment of the dynarcic system analysis constitutes an acceptable basis for satisfying the applic-able r3quirements of General Design Criteria 2 and 4."
For the FSAR, ;he review should provide justificatior, for a finding similar to that stated above with the phrase "will be implemented" modified to read "has been irrplemented."
V.
REFERENCES General Design Criterion 1, " Quality Standards and 1.
10 CFR Part 50, appendi-n.
Records."
2.
li CFR Part 50, Aptendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
3.
10 CFR Part 50, Appendix A, General Design ~rr erion 4, " Environmental and Missile Design Bases."
4.
10 CFR Part 50, Appendix A, General Design Criterion 14, " Reactor Coolant Pressure Boundary."
5.
10 CFR Part 50, Appendix A, General Design Criterion 15, " Reactor Coolant System Design."
6.
ASME Boiler and eressure vessel Code,Section III, " Nuclear Power Plant Components,"
American Society of Mechanical Engineers.
7.
Regulatory Guide 1.20, " Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing."
8.
Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors."
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