ML19308E108
| ML19308E108 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/18/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Mallay J BABCOCK & WILCOX CO. |
| Shared Package | |
| ML19308E107 | List: |
| References | |
| NUDOCS 8003200830 | |
| Download: ML19308E108 (5) | |
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UNITED STATES e,,
ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 19848
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.i OCT 181971 Mr. James F. Mallay Manager, Licensing Nuclear Power Generation P. O. Box 1260 Lynchburg, Virginia 24505
Dear Mr. Mallay:
We have completed our initial review of your topical reports, listed below, and find that we need additional information to complete our evaluation.
BAW-10037, Revision 1, " Reactor Model Flow Testing"
. BAW-1003B, " Prototype Vibration Measurement Program for Reactor Internals" BAW-10050, " Evaluation of Oconee Reactor Component Failure" BAW-10051, " Design of Reactor Internals and Incore Instrument Nozzles for Flow-Induced Vibration" The specific information required is listed in the enclosures.
In order to maintain our licensing review schedules for facilities referencing these topical reports we will need a prompt and completely adequate response.
Picase inform us within seven (7) days after receipt of this letter of your schedule for submitting the complete 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 co'be extended.
Picase contact us if you desire any discussion or clarification of the material required.
Sincerely, l * ***
R. C. DeYp6ng, Ass tant Director j
for Pres'surized Nater Reactors l
Directorate of Licensing
Enclosure:
Requests for Additional Information l
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REr7EST FOR ADDITIONAL INF010fATION
" REACTOR MODEL FLOW TESTING" B&W REPORT BAW-10037, REVISION 1, SEPTDtBER 1972 1.
Verify the possible omierl.an of the flow area term in the squation l
(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.
1 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 shMding frequency since this model is valid only for a simple flow condition.
4 REQUEST FOR ADDITIONAL INFORMATION
" PROTOTYPE VIBRATION MEASUREMENT PROGRAM FOR REACTOR INTERNALS" B&W REPORT BAW-10038, SEPTEMBER 1972 1.
Since flow-induced forcing functions have not been identified or postulated provide a description of the method that was employed to determine the predicted responses.
e 2.
Supplement Table 6-1 by providing predicted readings or estimated stress levels at all sensor locations.
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REQUEST FOR ADDITIONAL INFORMATION
" EVALUATION OF OCONEE REACTOR COMPONENT FAILURE" B&W REPORT BAW-10050, SEPTEMBER 1972 1.
As stated in page 4-12, the first mode frequency of the instrumentation guide tube is 250 Hz while the vortex shedding frequency is approximately 385 Hz, therefore, the first mode response may be excluded as a failure mode. However, higher modes may be in the range of the vortex shedding frequency or other forcing frequencies.
(a)
Provide a comparison of the high*.t mode guide tube frequencies with the shedding frequency.
(b)
Provide the criteria that was used for redesign of the instrumentetion guide tubes.
(c)
Provide a discussion of other possible causes of failure, such as the mentioned radnom excitation of turbulence and the reactor coolant pump excitation. Include the effect of the pump shaft frequency of 20 Hz (page 4-9).
2.
Provide a d*scussion on the following possible failure mode on the incore instrument nozzles: The core structure vibratory motion and the cross flow loading may produce a rotational vibration model in the guide tubes and associated lateral deformation of the lower tips.
The lateral motion may produce vibratory contact with the inserted tip of the incore instrument nozzle and result in cyclic bending stresses at the bottom of the nozzle to failure.
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REQUEST FOR ADDITIONAL INFORMATION
" DESIGN OF REACTOR INTERNALS AND INCORE INSTRUMENT N0ZZLES FOR FLOW-INDUCED VIBRATION" B&W REPORT BAW-10051, SEPTEMBER 1972 1.
Describe the loading combinatio6s 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 functional testing.
2.
As shown in Table 3-3 (page 3-26) the cantilever part of the guide tube and the flow distributor assembly (vertical) have approximately the same,first node frequencies. The configuration shown in Figure 3-3 indicates that the vertical action of the flow distributor may produce rotation and therefore lateral motion at the lower tip of the guide tube. Provide a sunsaary of the dynamic analyses used to account for possible dynamic coupling of the guide tube and the flow distributor assembly. 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 8 value of the drag force acting on the incore instrument nozzles was actually based upon a two (2) inch diameter (page 3-5) of the lower portion. Since the upper portion is 1 1/8 inch diameter (8-1), provide a summary of the analysis to show that excessive response amplitudes of the instrument nozzles will not occur.
Provide the basis for assuming that the lowest mode deflect?on of_the _
thermal shield is 0.06 inches.
5.
Provide the basis for asswaing that the amplitude of other predominate modes of the thermal shield are a fanction of the ratio of the frequencies squared to the first mode (page 3-14).
6.
Provide the basis for neglecting the combined modal contribution effects in predicting the maximum radial deflection of the thermal shield under the hot functional testing and normal operational loadings (Table 3-5).
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