ML20040F684
| ML20040F684 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Limerick |
| Issue date: | 05/17/1971 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20040F238 | List: |
| References | |
| FOIA-81-385 NUDOCS 8202100112 | |
| Download: ML20040F684 (5) | |
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Adequate responses to the enclosed request for additional information
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tion. These requests, prepared by the DRS 14achanical Engiamering 3 ranch.,--
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concera dyasmic effects induced ta the reactor coolant pressure homedary '-
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piping, design of valves, and r=e1===ificatise of sei==1c=11y designed systems. The enclosed supplements our requests for information of j
July 20,1970, March 4,1971, and April 9,1971.
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R. Boyd, DEL R. DeToong, DEL R. R. Maccary, DRS D. Kauth, DRL D. Lange, DRS i l C. L==r, DRL
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REQUEST FOR ADDITIONAL INFORMATION LIMERICK CENERATINC STATION UNITS 1 AND-2
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DOCKET NOS. 50-352 AND 50-353 1.
Paragraph I-701.5.4 of the ANSI B31.7 Nuclear Power Piping Code requires that piping shall be supported to rninimize vibration and that the designer is responsible by observation under startup or r-initial operating conditions to assure that vibration is within acceptable icvels.
Submit a discussion of your vibration operational test program which vill be used to verify that the piping and piping restraints within the reacter coolant pressure boundary have been designed to withstand dynamic effects due to valve closures, pump trips, etc.
Provide a list of the transient conditions and the associated actions (pu=p trips, valve actuations, 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 that will result in significant vibration response of reactor coolant pressure boundary systems and co=ponents.
2.
Appendix A of the PSAR specifies the ASME Code for Pumps and Valves I
for Nuclear Power (NP&V Code) as applicable to the design of Classifica-tion Group Nuclear I (Class I NP&V Code) valves. However, this Code, in
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conjunction with two recent code case interpretations, allow the option of selecting any of the following design procedures for Group A valves:
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2-Paragraph 452.la of the NP&V Code, Standard Pressure Rated a.
Valve 1
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Paragraph 452.lb of the NP&V Code, Non-Standard Pressure Rated Valve c.
MSS SP-66, Pressure-Temperature Ratings for Steel Butt-Welding
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End Valves,1964 edition (as referenced in the March 1970 Addenda to the NP&V Code) d.
ASME Code Case 1466 Indicate which of the above design standards will be used in the
'l design of Group A valves within the reactor coolant pressure boundary.
3.
The ASME Code for Pumps and Valves for Nuclear Power (NP&V Code) which is specified for the design of pumps and valves within the reactor coolant pressure boundary for this plant, stipulates the use of ANSI B31.7 Nuclear Power Piping Code for design under earth-quake loadings (paragraph 424).
For the combination of loadings which include those due to earthquake, emergency and faulted operating c
condition categories may apply in conjunction with the associated stress limits as given in Case 70 of Interpretations of Code for 4
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Pressure Piping.
Indicate whether the stress limite of Code Case 70 will be applied in the design of pu=ps and valves within the reactor coolant pressure boundary for the emergency and faulted operating condition categories'.
If other stress criteria are proposed, pro-vide the bases for their application.
4.
Distinguich between the stress limits proposed for active and pumps and valves, e.g., certcin pumps and valves (classified inactive as active co=ponents) within the recctor coolant pressure boundary are required not only to serve as a pressure-retair.ing component (as in the case of passive conponents, vessels, and piping) but also to operate reliably in order to perform a safety function; such as safe shutdown of the reactor, or, in the event of a pipe break in the system, to nitigate the consequences of the accident under the loading combinations considered in d'esign.
Therefore, to assure that an active component will function (e.g. closure of containment isola-tion valvec) in the event of a pipe rupture in the reactor coolant pressure boundary (faulted condition), we consider the stress limits for the " emergency condition" as appropriate in lieu of.the code stress limits for the " faulted conditions".
State whether it is your intention to co= ply with the limits indicated for active pumps and valves as c
defined, or justify any exceptions in your response.
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NOTES:
1.
Inactive congonents are those whose operability (e.g., valve opening, j'
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or closure, pump operation or trip) are not relied upon to, perform the system function during the transients or events considered in the respective cperating condition categories.
2.
Active components are those whose operability is relied upon to perfore a safety function (as well as reactor shutdown function) e during die transients or events considered in the respective operating 4
condition categories.
5.
The list of structures and systems that have been desipaated as seismic Class I in Appendix A of the PSAR does not include that
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portion of the main steam system extending from the outermost containment isolation valve up to the turbine casing and connected
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piping inclusive of the first valve (either normally closed or capable of automatic closure), nor does it include the liquid radioactive vaste storag'e treatment, handling, and disposal systems.
Indicate your intention either to reclassify these systems or portions thereof as seismic Class I, or submit your justification for the pro-posed Class II classification as presently specified.