ML19257B684
| ML19257B684 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/07/1980 |
| From: | Bowers E Atomic Safety and Licensing Board Panel |
| To: | NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD), SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8001180110 | |
| Download: ML19257B684 (3) | |
Text
.
UNITED STATES OF AMERICA
@ \\,';:',j ~'*'
NUCLEAR REGULATORY COMMISSION C
c5%
<C ATOMIC SAFETY AND LICENSING BOARD C""
h s
I"
.g 7 h' gqcice,gy(,'T
~
Elizabeth S. Bowers, Chairman tecd!ffg Dr. Richard F. Cole, Member Frederick J. Shon, Member
,, 4
/
g 'p' In the Matter of SACRAMENTO MUNICIPAL UTILITY DISTRICT
)
Docket No. 50-312
)
(Rancho Seco Nuclear Generating Station)
)
ADDITIONAL BOARD QUESTIONS (January 7, 1980)
The Board will expect the Licensee and the Staff to address the following questions at the hearing.
If other parties care to address these questions, we encourage them to do so.
If any party feels that its testimony as prepared for other issues already adequately addresses them, it is only necessary to point out the pertinent sections of that testimony.
The questions are:
1.
At a meeting with owners of B&W reactors held on August 23, it was noted that, in the interim then elapsed since the TMI-2 accident, control-grade hard-wired antici-patory reactor trips (ART) had been called on to respond four times and had failed once:
a.
Is this typical of performance by control grade trips?
1763 297 8001180
. b.
What are the safety implications for opera-tion of Rancho Seco before such trips are upgraded?
2.
We note (letter D. Ross to J. J. Mattimoe, December 14, 1979) that there is still some dispute as to the fundamental logic for Reactor Cooling Pump (RCP) trip in a small-break LOCA.
a.
What current instructions to reactor opera-tors govern tripping of the pumps in small-break LOCA's and upon what theory of system behavior are those instructions based?
b.
What are the implications for safety of operating Rancho Seco until the exact behavior of the system in a small-break LOCA is well-understood?
3.
It appears from a Board Notification issued by R. H.
Vollmer on December 5, 1979, that the basic design of the once Through Steam Generator (OTSG) may so closely couple primary system behavior to secondary system disturbances that gross disturbance of the primary system is inevitable for feedwater transients.
Further, it seems there are situations in which an operator may not be able to tell exactly what is wrong or what response is appropriate (e.g. over-cooling vis-a-vis a small-break LOCA).
~
176B 298
. What changes in the system and procedures a.
have been made to ameliorate this situation?
b.
What are the implications for safety of operating Rancho Seco before any uncer-tainties are resolved?
FOR THE ATOMIC SAFETY AND LICENSING BOARD SYi Eliz/beth S. Bowers, Chairman Dated at Bethesda, Maryland, this 7th day of January, 1980.
}700