ML20037A614
| ML20037A614 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/18/1972 |
| From: | Maccary R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8003250618 | |
| Download: ML20037A614 (10) | |
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,c Richard C. DaT5ung, Assistant Director for Pressurizad Wster Reactors Directorate of Licensing
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CRTSTAL RIVER - UNIT 3r DOCKET NO. 50-302 g ;
Adequate responses to the~ anclosed list' of questions prepared by the -
Mechanical Engineering Braw hg L, are required before we can complate our review of the subject application. These questions concern the reactor coolant pressure boundary, reactor internals, aclamic system analysis and seizaic instrumentation submittesi in Amendment IS of the FSAR.
The opplicant has rafarenced, in Section 7 of tho'l'SAR, B & W Topical Report BMT-20GO3 "Oualification Testing of Protection System Inntru-sentation," dated !hrch, 1971. The additional infor=ation requirod, to coe lete our review of thic report was requested in cur letters of " arch 13, 1972 ann May 23, 162 concerning Thrca Mila Island Station, Unit No. 1, Docket Eo. 50-289. Our review cf the subjec: application can therc ore i.e cnpedited.han this inforntic,n la raccivad.
a The applicant has also referanced in Section 3 of the FSAR, 2 L W Topical leport 3EJ-10'"P3, Part 1, "hetor Internais St,ross and
aflections Dus to Loss-of-Coolsat t.ccid?nt and M w N2113 pethetical Earthquake," dated June. 1970. The sections of the report applicablo to the subject application are currently under raview by the Machanical Engineering Branch and additional information may be required prior to coupletion of this review.
This request for additionel ir'orittion supplemen* : rhe provious request dated December 9, 1971.
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d '. Responsible Branch and Project Leader PWRf4/H. Faulkner a.
Roquested Complation Datos. 8/1/72 f.l Applicants response data necessary for.complation of next action ~
.p anned on projects. 9/20/72-f_
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Description oflresponset..Resposee'to~ questions h.
Review Status: Awaiting'informatten.
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w-v CRYSTALRIb1-UNIT 3'-
DOCKET NO. 50-302 REQUEST FOR~ ADDITIONAL INFORMATION-A.
Reactor Coolant Pressure Bcundarv 1.
Supplement the response to Question 4.19,'A=endment No. 18 of
- the FSAR, by describing the acceptance program that will be implemented to determine the acceptable amplitudes 3f the vibration for confirming the structural integrity of the piping and pipe restraints.
2.
Provide the dynamic testing procedures used in the design of Category I =echanical equipment (such as fans, pump drires, valve operators, heat exchanger tube bundles) to withstand seismic, accident and operational vibratory loading conditions, including the methods and procedures employed which consider the frequency spectra and amplitudes calculated to arist at the equipment supports. Where tests or analyses do not include evaluation of the equip =ent in the operating code, describe the bases for assuring that this equipment will function hen subjected to seisnic accident loadings.and vibratory loadings.
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.-..3, Clarify.the responselto Ouestions 4.20 and 4121 by-providing a
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more detailed description of the dynamic analyses performed for Class 1 piping and associated supports which determine the resulting loadings as a result cf a postulated pipe break, s
including:
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breaks on which the dvnanic analyses'are based.
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the postulated rupture orientation, such as a circumferential and/or longitudinal break (s), for each postulated design basis break location.
- c. _ description of the forcing' functions'to be used.for the pipe.
whip _ dynamic analyses, including direction, rise time, magnitude, duration and initial conditions'.that adequarely represent.the jet strea dynanics and the system pressure differences.
d.
typical diagrams of the mathecatical codels used for the' dynamic analysis.
e.-.'a su==ary of the analyses performed to 'emonstrate that motion-d of ruptured lines where unrestrained vill not sever adjacent impacted piping er pierce impacted areas cf containment steel
-liner.
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4.
Describe the analytical methods used t'o evaluate stresses,(e.g...
elastic or inelastic) including a discussion of their compatibility with the. type of dynamic systen analysis for the combination of:
loadings; normal plus design basis. earthquake plus LOCA.
If in-
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elastic component stress analyses and inelastic d.tsign stress
-limits are used in conjunction with an elastic dynamic systen t
analysis provide juscification for using such design procedures.
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With respect to the responsa 4.22,. provide'the stress or.
deformation limits and the design standards or' codes appli-,
cable to the pressurizer support.
6.
Section 4.1.3.4 states that the-pressurizer code safety valves comply with Article 9,Section III, of the ASME Boiler and Pressure Vessel Code.
However, Article.9 does not address itself_to the structural integrity of valves.
Provide the stress or deformation li=its which will be'used:for safety valves and' justify the basis for their application.
7.
Section 4.2.2.8 states that loads imposed on the safety and relief valves within the reactor coolant pressure boundary are-limited to acceptable values.
Define the term " acceptable" in terms of stress or-deformation limits and justify the basis for their application.
3.
The response to Question 4.27, Amendment 18 of the FSAP. states that a plastic analysis was used to design and analyze the pipe whip restraints installed to prevent a ruptured main steen or feedwater pipe fron damaging the component of reactor coolant pressure boundary. Provdie a su= mary of the =ethods of the plastic analysis used including justification of the validity of such nethods.
In addition provide the stress or strain licits used to compare with the results of the plastic analysis and justify the basis for their application.
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Reactor Internals l'.
Describe the' dynamic:syste::r analysis methods and procedures which will be used.to deter =ine dynamic responses of reactor internals:
and associated Class 1 components.of the reactor' coolant pressure
' boundary which'have effect on the responses (e.g., analyses and-tests).
Describe your preoperational test program'as related-
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to Safety Guide 20, Vibration Measurements on Reactor Internals.
If elements of the test program differ substantially from the requirements of Safety Guide 20,, submit the basis and justifi-cation for these differences.
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C. ~ Seismic System Analysis l.
With regard to response 5.14, provide sketches of che cathematical models employed including models of soil-structure interaction effects.
2.
Response 5.16 is inadequate. Confirm the validity ofLthe fixed base assumption by providing summary analytical results that'
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indicate that the rocking and translational response is insig-nificant.' Include a brief description of the mathematical' model and damping values (rocking,. vertical, translation, and torsion) that have been used to consider the soil structure interaction.
3.
Response 5.18 is inadequate. Descr_ibe the measures taken to consider the effects on floor response spectra (e.g. peak vidth and period coordinctes) of expected variations of structural properties, damping, soil properties.and soil structure. inter-action.
4.
The response to Question 5.26.1 is not acceptable.
It has not been de=onstrated theoretically that the response spectrum method gives conservative results compared to the time history nethod.
Demonstrate, for all floor response spectra developed, that the response spectrun method used gives conservative results conpared to the time history =ethod unless the use of the tine-history (or
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equivalent)'=ethod'is planned to develop the floor responsez spectra.
. 5.
Clarify response 5.26.6 by stating whether the internal shears,
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moments and internal reactions were determined for each mede separately an'd'then co=bined using the-square root of the sus-af the squares or by describing any-other method that was used-to obtain the combined total response.
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Seismic Instrumentation
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Appendix A to 10 CFR Part 100 requires that suitable f.nstrumentation be provided to determine, promptly, the. seismic responsa of. nuclear power plant features important to safety.and to permit comparison of
-such responses with that used as the design basis.
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A single strong motion accelerograph installed on the top of the containment structure is not acceptable to meet the intent of Safety Guide 12, " Instrumentation for T.arthquakes.*' In addition, seismic instru=entation is also needed for other Category I structures and components.
(1) Provide an appropriate?v modifiel seismic instrumentation program which includes the selection of the type, runber, location and utili::ation of strong ration accelerographs to record seismic events and data on the frequency, amplitude and phase relation-ship c: the seismic response of the containment structure.
(2)
In addition, describe the seismic instrumentation system (such as peak recording accelerographs and peak deflection recorders) that.
you propose to install in selected Category I structures (other than containment structure) and on selected Category I compenents.
Specify the basis for selection of these structures and components, and for the location of instrumentation, as well as the extent to which this instrumentation vill serve, following a seismic event to verify the predicted respcase for these structures and components as derived from saisnic analyses.
Include the criteria and procedures O
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(3) Describe the provisions that will be utilized to signal'to the.
control room operator the value of the peak acceleration level experienced at the facility founcations within a few minutes
. after the earthquake.
Include the basis for establishing pre-deternined values for correlating the readout of the seismic.
instruments with the ground motion specified for this site.
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