ML19220A569
| ML19220A569 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/18/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Mallay J BABCOCK & WILCOX CO. |
| References | |
| NUDOCS 7904230075 | |
| Download: ML19220A569 (5) | |
Text
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UNITED STATES
/fg/fj y, ATOMIC ENERGY COMMISSION WASHINGTON Q.C. 20545
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- n o OCT 1 5 E7?.
Mr. James F. Mallay Manager, Licensing Nuclear Power Generation P. O. Box 1260 Lynchburg, Virginia 24505
Dear Mr. Ma!. lay:
We have completed our initial review of your topical reports, listed below, and find that we need additional information to complete our evaluation.
3AW-10037, Revision 1, " Reactor Model Flow Testing" 3AW-10038, " Prototype Vibration Measurement Program for Reactor Internals" 3AW-10050, " Evaluation of Oconee Reactor Component Failure" 3AW-10051, " Design of Reactor Internals and Incore Instru=ent Nozzle.4 for Flow-Induced Vibration" The specific information required is listed in the enclosures.
In order to =aintain our licensing review schedules for facilities referencing these topical reports we will need a prompt and completely adequate response.
Please infor= us within seven (7) days after receipt of this letter of your schedule for submitting the co=plete response.
If your reply is not prompt or fully responsive to our requests it is highly likely that the overall schedule for completing the licensing reviews for these facilities will have to be extended.
Please contact us if you desire any discussion or clarification of the
=aterial required.
Sincerely, s -Wj R. C. DeYp6:ig, Assistant Director for Pres'surized Water Reactors Directorate of Licensing
Enclosure:
Requests for Additional Information 7.90423607f 2
~
66'001
REQUEST FOR ADDITIONAL INFORMATION
" DESIGN OF REACTOR INTER.'ULLS AND INCORE INSTRUMENT N0ZZLES FOR FLOW-INDUCED VI3 RATION" 3&W REPORT BAW-10051, SEPTEMBER 1972 1.
Describe the loading combinations and the analytical methods used to confirm the structural integrity of the instrumentation guide tubes. Provide the basis for the criteria that redesign is not necessary if two guide tubes fail during hot functienal testing.
2.
As shown in Table 3-3 (page 3-26) the cantilever part of the guide tube and the flow distributor assembly (vertieni) have approx 1=ately the same first mode frequencies. The configuration shown in Figure 3-3 indicates that the vertical motion of the flew distributor may produce rotation and therefore lateral motion at the lower tip of the guide tube. Provide a su= mary of the dynamic analyses used to account for possible dynamic coupling of the guide tube and the flow distributor asse=bly. Include the efforts of cross flow on the cantilever portion of the guide tube. The associated cyclic bending stresses at the incore instrument nozzle should also be provided.
3.
The shedding frequency used for computing the S value of the drag force acting on the incere instrument nozzles was actually based upe-a two (2) inch dia=eter (page 3-5) of the lower portion.
Since the upper portion is 1 1/8 inch diameter (!=1), provide a su==ary of the analysis to show that excessive response amplitudes of the instrument nozzles will not occur.
4.
Provide the basis for assuming that the lowest mode deflection of the thermal shield is 0.06 inches.
5.
1.uvide the basis for assuming that the a=plitude of other predominate modes of the thermal shield are a function of the ratio of the frequencies squared to the first = ode (page 3-14).
6.
Provide the basis for neglecting the combined modal contribution effects in predicting the max 1=um radial deflection of the tner=al shield under the hot functional testing and nor=al operational loadings (Table 3-5).
GG oog,
O REOUEST FOR ADDITIONAL INFORMATION "EVAI,UATION OF OCONEE REACTOR CCMPONENT FAILURE" 3&W REPORT BAW-10050, SEPTEMBER 1972 1.
As stated in page 4-12, the first = ode frequency of the instrumentation guide tube is 250 R: while the vortex shedding frequency is approxi=ately 385 Hz, therefore, the first = ode response =ay be excluded as a failure = ode. Hottver, higher = odes =ay be in the range of the vortex shedding frequency or ether forcing frequencies.
(a)
Provide a ce=parison of the higher = ode guide tube frequencies with the sheddt4g frequency.
(b)
Provide the er.teria that was used for redesign of the instru=entatien guide tubes.
(c)
Provide a d1>cussion of other possible causes of failure, such as the nentioned radno= excitation of turbulence and the reactor coolant pu=p excitation.
Include the effect of the pu=p s'taf t frequency of 20 Hz (page 4-9).
2.
Provide a discuasion on the following possible failure mode on the incore instrurent nozzles: The core structure vibratory motion and the cross flos loading =ay produce a rotational vibration =odel in the guide tubes and associated lateral defor=ation of the lower tips.
The lateral r.ntion =ay produce vibratory contact with the inserted tip of the incore instrument nozzle and result in cyclic bending stresses at the botto= of the nozzle to failure.
EG 003
REOUEST FOR ADDITIONAL INFORMATION
" PROTOTYPE VI3 RATION MEASU' 3*r FROGRAM FOR REACTOR IMERNALS" B&W REPORT 3xn-10038, SEPTEMBER 1972 1.
Since flow-induced forcing functions have not been identified or postulated provide a description of the method that was e= ployed to detetuine the predicted responses.
2.
Supplement Table 6-1 by providing predicted readings or estimated strces levels at all sensor locations.
EG-~004
~.
REOUEST FOR ADDITICNAL INFORMATION
" REACTOR MODEL FLOW TESTDiG" 35W REPORT BAW-10037, PR ISION 1, SEPTEMBER 1972 1.
Varify the possible omiselon of the flow area term in the equation (1 -1).
2.
The flow frequency content and the related energy distribution was not determined by the measurements during the 1./6 scale model testing.
Identify the contribution of this model testing to the postulation of forcing functions for response prediction analysis. Provide the basis for the use of the simple equation setforth on Page 3-4 of BAW-10051 to compute the shedding frequency since this =odel is valid only for a si=ple flow condition.
gG-005