ML19312C318
| ML19312C318 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/30/1967 |
| From: | Grimes B US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7912130847 | |
| Download: ML19312C318 (4) | |
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CNITED STATES GOVERNMENT DIemorandum To The Files or7g:
THRU: Roger S. Boyd, Chief M J Research & Power Reactor Safety Branch, DRL j
FROM Brian Grimes j
Research & Power' Reactor Safety Branch, DRL
SUBJECT:
MEETING WITH DUKE POWER COMPANY ON OCCNEE PLANTS DOCKET Nos.~501269'AND 50-270 r
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On January 18, 1966, a meeting was held in the Bethesda offices with representatives of the Duke Power Company and Babcock & Wilcox to discuss the form of future technical specifications for the two-unit station.
In response to Dr. Mann's request on December 2,1966, Duke had prepared a listing of the subjects and parameters to be incorpo-rated into the technical specifications at the operating license stage.
When asked whether Duke thought some consideration of technical speci-fications was beneficial at the construction permit stage, Mr. Lee responded that it was, in that it kent service and testing objectives before the designers as well as the operational considerations.
It was decided not to proceed further in the development of specifications at this time.
The staff held discussions with the applicant after this meeting to go over areas which we feel may present problems on the basis of our pre-liminary look at the application.
Additional subjects which we felt should be covered in the technical specifications included:
(1) pressurizer high level, (2) minimum con-tainment required for operation, (3) maximum acceptable primary system
- leakage, (4) surveillance of the primary system during the plant lifetime, and (5) surveillance of reactivity anomalies.
The following topics were indicated to the aoplicant as areas which we intended to pursue in our review. We stated that direct communication by Duke or Babcock & Wilcox engineers with Cardis Allen, DRL, on reactivity subjects, and with Don Sullivan, DRL, on instrumentation, was acceptable and desirable.
A.
Core Analysis 1.
Neutron balances over the core lifetime, particularly as related to the effects of soluble boron.
2.
Proof testing of the control rod design (delays in scram time have an effect on excursion calculations).
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l The Files M I"
Docket Nos. 50-269/270 3.
Additional information on the rod ejection, loss-of-coolant and steam-line-break accidents. Duke stated that the cooldown of the primary system after a steamline-break accident was not as much of a problem in this reactor since the inventcry in the steam generator secondary is not as great with the once-throuah design (there are no isolation valves in the system).
4.
More information on expected control rod programming.
B.
Instrument ation 1.
Separation of control and safety functions will be explored.
2.
It appears that a single failure on the final d.c. bus could prevent a scram.
3.
Isolation of the servo-function will be explored.
4.
The acceptability of the adjacent hydro plant as a source of emergency power will be explored.
5.
copy of the proposed IEEE standards and our criteria on con-tainment penetration were given to the applicant.
C.
Site 1.
The earthquake design value of 0.05g looks low for this area (my uninfor:ned guess deduced from Carolina Power and Light values).
2.
Since the dams are required to remain intact, they must withstand the design earthquake. (One possibility that was not broucht up at this time is that it may be desirable to have a method of removing normal decay heat that is independent of the dams.)
3.
The presence of residents within the one-mile exclusion radius; Duke indicated that this problem may be resolved by the time of construction permit issuance.
D.
Containment 1.
Hea will the containment be tested to justify the assumption that one-half the building leakare will be to the penetration rooms and, thus, available for filtering? Duke stated that they could presently see no way to confirm this by test and that if credit was not given for this filtering, they would change the
The Files JAN 3 a 5 Docket Nos. 50-269/270 design to eliminate the filters and go to a lower leakage con-tainment (probably about 0.3%/ day vs. the present 0.5%/ day).
This would have the effect of paying for more frequent leakage tests with money otherwise used for the filter system.
2.
The sharing of engineered safeguards between Units 1 and 2 will
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be evaluated. As vet, we have no criterion for this.
l 3.
Duke and B&W did not understand that our design basis accident postulated a coolant-line break simultaneous with loss of power i
and the safeguards had then to withstand a single failure. This would significantly affect the pumping capacity required and may I
influence their decision on whether to resort to accumulators.
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4.
The Duke failure-analysis of the safeguard systems assumed passive as well as active failures. We stated that our criterion on this was not yet fort'21sted but that it was possible that only active failures need be considered in certain situations. Although the following criterion was not communicated to the applicant, it represents my judgment on single-failure analyses at this writing:
" Safeguard systems must be operable after one active component failure when a loss-of-coolant accident is postulated concurrent with a loss of of f-site power.
Safeguard systems must be coerable with one passive component failure within the containment on a long-term basis (when off-site power will probably be available)."
5.
The following criterion was transmitted to the applicant as our current thinking on the over-ell design basis for emergency core cooling systems: "The ECCS should be designed to prevent fuel and clad damage that would interfere with adeauate emercency core cooling and to limit the clad-water reaction to less than approxi-mately 1% for all break sizes in the primary system piping up to the double-ended rupture of the largest crimary coolant pipe, for all break locations and for the applicable break time of the pipe.
An analvsis should be oerformed to show the expected marcin in the des 12n to orevent clad melting."
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6.
Duke inquired whether it was permissible to take credit for con-tainment air pressure in calculating the required NPSH of the emergency cooling pumps. They were under the impression that something of this sort had not been allowed on the Florida reactors.
I said I would find out and phoned Mr.14e on January 24, 1967, l
to inform him that accident pressure should not be taken into account in satisfying NPSH requirements and that we would look at I
the requirements for sump water height to operate the pumps. This l
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The Files
_4 jgg 3, sg7 Docket Nos. 50-269/270 requirement is based on the judgment that these pumps may be required for situations where the containment is not signifi-cantly pressurized.
Mr. Lee stated that the capacity of the borated water storage tank had been increased so that recircu-lation would not be required for about 20 minutes.
One other general consideration mentioned was that it may be necessary to start up Unit 2 with fresh, rather than used, fuel depending on 1
our analysis.
Duke said that they were willing to accept a construc-tion permit with this reservation and await our analysis at the operating stage.
Attendance at the meeting was as listed below.
In addition, Dr. Morris and Dr. Mann were present during the technical specifications discussion.
DR1 Compliance Duke B&W R. Boyd L. Kornblith W. Lee G. Kulanich D. Muller J. Hard E. Fiss B. Grimes J. Murphy D. Sullivan Distribution:
E. G. Case R. S. Boyd C. Long J. Murphy D. Sullivan B. Grimes DRL Reading R&PRSB Reading l