ML19319E202
| ML19319E202 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/27/1972 |
| From: | Schwencer A US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8003310715 | |
| Download: ML19319E202 (4) | |
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Docket File R. W. Klecker I
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S. Hanauer Licensing Assistan R. S. Boyd R. C. DeYoung Docket No. 50-312 OCT 2 7 gra D. Skovholt F. Schroeder R. R. Maccary D. Knuth R. Tedesco i
Mr. R. K. Davis, General Counsel Sacramento Municipal Utility District 6201 S. street, P. O. Box 15830 Sacramento, California 95813
Dear Mr. Davis:
On the basis of our continuing review of the Final Safety Analysis Report for the Raneho Seco Nuclear Generating Station, we find that we need additional information c5 complete our evaluation. The specific information required is listed in the acciosure.
In order to maintain our licensing review schedule, we will need a completely adequate response by November 6, 1972. Please inform us within seven (7) days af ter receipt of this letter of your confirmation of the schedule or the date you will be able to meet. If you cannot meet our specified date or if your reply is not fully responsive to our requests, it is highly likely that the overall schedule for completing the licensing review for this project will have to be extended. Since reassignment of the staff's
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efforts will require sempletion of the new assignment prior to returning
_ y' to this project, the estaat of extension will most likely be greater than
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the extent of delay in your response.
Sincerely, Original Signed by h
mert Schwencar
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A. Schweneer, Chief ls Pressurised Water Raastors Braneh No. 4 i
Directorate of Licensing
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Enslosure:
l Request for Additional Information es w/ encl:
David 8. Raplan, Soaretary and Attorney
.6201 s. street, P. o. Bom 15830 i
Seermeento, California 95813 l
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RECUEST FOR ADDITIONAL INFORMATION SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR CENERATING STATION DOCKET NO. 50-312 9.0' AUXILIARY AND EMERGENCY SYSTEMS 9.35 Complete your response to Request 9.35 by providing the results of an evaluation for cases 9.35.1,.2, and
.3, using conservative conditions, for the following:
1.
The magnitude of impact level, its effect on the reactor vessel's points of support and the possible consequences of disrupting the flow of coolant to and frca the reactor vessel and refueling canal and the consequences of such a failure; and 2.
The possibility for failure of the water seal between the reactor vessel and the refueling canal structure and the con-sequences of such a failure.
9.46 Your present design of the pump room ventilation system equipment is seismic Class III. ECCS equipment located in the pump room may be required to operate for long periods of time following an accident. Leakages may occur in this ECCS equipment, and there-fore there is the potential that radioactive fluid may escape containment when the ECCS equipment is operating in a recirculation mode.
1.
If the assumption is made that in the event of a loss-of-coolant accident the equipment is leaking at a maximum possible leakage rate (i.e., postulate a damaged seal, or packing, or some other leakage path in which leakage would be at a maximum but not great enough to cause the pump or equipment to be inoperable) following a design basis seismic event, calculate the radioactive release to the environment over the 30-day period of operation. State all assumptions that were used and verify that they are conservative.
2.
Provide an estimate of the total amount of leakage that could
. occur prior to isolation of failed equipment.
'3.
Verify that the design of your ventilation and cleanup system, whose design function is to control the envirotunent and exhaust fron 'the ECCS pump rooms, meets our present requirement to minimize potential radioactive releases to the environment and
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. 9.47 Indicate'the backup system to the CO2 fire protection system.
1.
the effect that the backup avstem is a portable system, describe the means by which ingress t, ;- entially smoke-filled rooms will
-be assured..
9.48
-The two Nuclear Service Spray Ponds comprise the Rancho Seco Class I (seismic) ultimate heat sink. Verify that the failure of al) non-seismic Class'I piping and equipment physically connected to these ponds will not cause the drainage of these ponds following a design basis seismic event, thereby reducing the capability of the ponds to provide 30 days of post-accident cooling.
10.0
. STEAM AND POWER CONVERSION SYSTEM 10.8
-Your response to Request 10.8 was not complete in that the relation of the detectable flaw size to the critical crack size for the turbine generator highly stressed parts and rotating members was omitted. Provide the required information for the above so that
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we can complete' our evaluation.
-14.0 SAFETY ANALYSIS 14.11 Your response to Request 14.11 states that the 5.0 f t.2 break with continuous steam release, which is analyzed in paragraph 14.4.1, is a cold leg break. However, your response to Request 14.12 states that the 5.0 ft.2 break with continuous steam re-lease is a hot leg break. Clarify this inconsistency.
14.12 It is not apparent from your response to Request 14.12 that the internal vent valves will preclude steam venting through the steam generators for a cold leg, pump suction break at a point close to a steam generator. Therefore, provide the reactor building pres-sure transient analyses, for a spectrum of cold leg breaks. The analyses should be extended through the initial blowdown, reflood, and post-reflood phases of the postulated accidents. All assump-tions used in the analyses should be explained. Assumptions should be conservative with respect to the calculation of reactor building pressures..
14.13
'For the spectrum of hot' leg breaks analyzed, the 5.0 ft.2 break results-in the high'est reactor building pressure. Discuss why
-an intermediate size break should result in the highest reactor 1
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building pressure. Would a pipe. rupture having a break size 4
between 3 f t.2 and 5 ft.2 result in a higher peak reactor building pressure?
14.14 Discuss how the reactor coolant system blowdown analyses for loss-
.of-coolant accidents were modified to assure a conservative cal-culation of the mass and energy releases.co. the reactor building for' the reactor building pressure response analyses.
14.15 For the cold leg break resulting in the highest calculated reactor building pregsure, tables of mass release (1b/sec) and the enthalpy of the mass released (Btu /lb) should be provided throughout the blowdown and reflood phases,of the accident. A graph showing core flooding velocity lui a function of time should also be provided for the reflood phase of the accident.
14C.
HYDROGEN PURGING
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14C.3 With respect to the post-accident hydrogen generation analysis based on Safety Guide 7, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident," provide the following information:
1.
Specify the surface area of Ethe aluminum that is not assumed
-to -react instantaneously with the spray solution.
2.
Graphically show.the integrated hydrogen production (ft.3) within the containment as a function of time (hr.) following the_ design b dis accident for each of the potential sources I
of hydroge
,.1.e.,
radiolytic decomposition of water, metal (zirconiump-water reaction, evolution of entrained hydrogen
. in primary coolant and aluminum corrosion.
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