ML19290C030
| ML19290C030 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/21/1979 |
| From: | Ellison C, Lanpher L CALIFORNIA, STATE OF |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8001090056 | |
| Download: ML19290C030 (10) | |
Text
,_
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the matter of:
)
cket No. 50-312 SACRAMENTO MUNICIPAL UTILITY DISTRICT
)
(Rancho Seco Nuclear Generating Station)
)
)
)
CALIFORNIA ENERGY COMMISSION REQUESTS FOR ADMISSIONS TO THE. NUCLEAR REGULATORY COMMISSION Pursuant to 10 C.F.R.
S2.742, the California Energy Commission
(" CEC") requests the. Nuclear Regulatory Commission
( " NRC ,
to admit the truth of the following relevant matters of fact:
The following detinitions shall apply:
(a)
" Rancho Seco">means the Rancho Seco Muclear Generating Station,. including, as appropriate, the physical plan, SMUD personnel, and operating procedures.
(b)
" Forced circulation cooling" means cooling achieved by physi-cally pumpingc. coolant through the core and the primary system.
(c)
" Natural circulation cooling" means cooling achieved by the natural flow of coolant through the primary system in a liquid state as a result of temperature-related density differences within the coolant.
I730 001 8001099o57'g
/\\
. (d)
" Natural condensation cooling" means cooling achieved through the evaporation of coolant in the core and condensa-tion of coolant in the steam generator, leading to a transfer of heat to the secondary system and a flow of condensed coolant back to the core.
Requested Admissions:
1.
That Rancho Seco and the Three Mile Island Unit-2 facility ("TMI-2") are both Babcock and Wilcox reactors of the same basic design.
2.
That Rancho Seco and THI-2 are substantially identical with respect to the capacity of the feedwater side of the steam generators,despite the modificaticas made to Rancho Seco after the TMI-2 accident.
3.
That Ranch 6'ffEco and TMI-2 are substantially identical with respect to the elevation of the steam generators relative to the reactor vessel,despite the modifications made to Rancho Seco after the TMI-2 accident.
4.
That Rancho Seco and TMI-2 are substantially identical with respect to the capacity of the pressurizer,despite the modi-fications made to Rancho Seco after the TMI-2 accident.
5.
That Rancho Seco and TMI-2 are substantially identical with respect to the design of the pressure operated relief valve,
despite the modifications made to Rancho Seco after the TMI-2 accident.
1730 002 O
6.
That Rancho Seco and TMI-2 are substantially identical with respect to the design of the emergency core cooling system, despite the modifications to Rancho Seco after the TMI-2 accident.
7.
That Rancho Seco and TMI-2 are substantially identi-cal with respect to the design of the auxiliary feedwater system, despite the modifications to Rancho Seco after the TMI-2 accident.
8.
That Rancho Seco and TMI-2 are substantially identi-cal with respect to the design of the primary cooling system, despite the mofidications to Rancho Seco after the TMI-2 accident.
9.
That Rancho Seco and TMI-2 are substantially identical with respect to the design of the secondary feedwater system, despite the modifications to Rancho Seco after the TMI-2 accident.
,10. That Rancho Seco and TMI-2 are substantially identical with respect to the capacity of the reactor coolant drain tanks, despite the modifications to Rancho Seco after the TMI-2 accident.
11.
That Rancho Seco and TMI-2 are substantially identi-cal with respect to the design of the integrated control system, despite the modifications to Rancho Seco after the TMI-2 accident.
12.
That Rancho Seco and TMI-2 are substantially identical with respect to the design of the nuclear steam supply system, despite the modifications to Rancho Seco after the TMI-2 accident.
13.
That Aancho Seco and TMI-2 are substantially identi-cal with respect to the design of the containment structure,despite the modifications to Rancho Seco after the TMI-2 accident.
I730 003
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14.
That neither Rancho Seco nor TMI-2 provide direct indication of reactor coolant inventory apart from pressurizer inventory.
15.
That Rancho Seco has no method of venting gases from the primary system aside from the pressure operated relief valve located in the pressurizer.
16.
That had Rancho Seco on March 28, 1979, experienced the same system fsilures and operator responses as TMI-2 experi-enced on that date, it would have experienced an accident similar to that of TMI-2.
17.
That the failure of the pressure operated relief valve to clore.was a significant contributing factor to the
- p..m accident at TMI-2.
18.
That the small capacity of the pressurizer was a contributing factor to the accident at TMI-2.
19.
That the elevation of the steam generator relative to the reactor vessel was a significant contributing factor to the accident at TMI-2.
20.
That voiding in the primary system resulted in an inability to achieve natural circulation at TMI-2 and was a sig-nificant contributing factor to the accident at TMI-2, 21.
That the small capacity of the feedwater side of the steam generator was a significant contributing factor to the accident at TMI-2.
I730 004
. 22.
That the lack of direct indication of reactor coolant inventory apart from pressurizer inventory was a signi-ficant contributing factor to the accident at THI-2.
23.
That the ability to vent gases from high points in the primary system would have facilitated natural circulation cooling at TMI-2 and thereby may have helped mitigate the accident.
24.
That the failure of the operators at TMI-2 to main-tain operation of the high pressure injection system in the early stages of the accident was a significant contributing factor to the accident at TMI-2.
25.
That the failure of the operators at TMI-2 to main-tain operation of the primary coolant pumps in the early stages of the _ accident was a significant contributing factor to the accident at TMI-2.
26.
That the inability of the reactor coolant drain tanks to contain all the coolant vented through the pressure operated relief valve exacerbated the accident at TMI-2 and led to offsite releases of radioactivity.
27.
That the failure of operators at TMI-2 to recognize or appreciate the significance of indications that coolant was being discharged to the reactor coolant drain tanks was a signi-ficant contributing factor to the accident at TMI-2.
28.
That the inability of the operators at TMI-2 to achieve core cooling by using the decay heat removal system exacerbated the TMI-2 accident.
1730 005
. 29.
That the '
of sufficient radiation monitoring equipment at TMI-2 and knowledge of plant systems and areas likely to contain radioactivity limited the ability of operators to respond to the accident.
- 30. That the lack of a hydrogen recombiner at TMI-2 may have resulted in a significant build up of hydrogen leading to an explosion sufficient to exceed the containment design.
- 31. That a second pressure operated relief valve on the pressurizer may have aided operators at TMI-2 in responding to the accident.
32.
That the failure of the operators at TMI-2 to recognize and unde'rstand the degraded condition of the reactor and the primary system was a significant-oontributing factor to the accident at TMI-2.
- 33. That procedures instituted at Rancho Seco since TMI-2 direct operators to rely upon natural circulation cooling for at least 20 minutes following a feedwater transient which activates the HPI.
- 34. That natural circulation cooling is unreliable once significant voiding occurs in the primary system.
- 35. That forced circulation cooli 2g is unreliable once significant voiding or loss of pressure o: curs in the primary system.
- 36. That forced circulation cooling is ueavailable at Rancho Secc where there is.a_ loss of off-site power together with a turbine trip.
1730 006
_7_
37.
That a significant number of operators at Rancho Seco exhibited an insufficient understanding of natural circulation cooling when first tested by the NRC in relation to SMUD's compliance with the NRC's May 7 shutdown order.
38.
That where both forced and natural circulation cooling are unavailable because of voiding in the primary system, there is no means of reasonably assuring adequate core cooling for extended time periods at Rancho Seco.
39.
That where both forced and natural circulation cooling are una'7ailable, there is no means to place the reactor core in a cold shut down without using the decay heat removal system.
40.
That where both forced and natural circulation cooling are unavailable because of voiding in the primary system, the only means of core cooling for extended time periods is to rely upon boiling and venting from the reactor core.
41.
That boiling and venting in the primary system may lead to core damage if used for extended time periods.
42.
That boiling and venting cooling has never been relied upon to assure adequate core cooling in pressurized water reactors.
43.
That boiling and venting cooling was not considered as a reliable method of assuring adequate core cooling during the licensing of Rancho Seco.
44.
That boiling and venting cooling in pressurized water reactors is an untested theory that has not been scientifically demonstrated.
45.
That boiling and venting cooling has never been successfully achieved in a pressurized water reactor like Rancho Seco.
1730 007 46.
That natural condensation cooling may be unreliable where significant amounts of non-condensible gases are present in the primary system.
47.
That natural condensation cooling has never been relied upon to assure adequate core cooling in pressurized water reactors.
48.
That natural condensation cooling was not considered during the licensing of Rancho Seco as a method of core cooling.
49.
That natural condensation cooling in pressurized water reactors is an untested theory that has not yet been scienti-fically demonstrated.
50.
That natural condensation cooling has never been successfully achieved in a pressurized water reactor like Rancho Seco.
51.
Th'at the h.igh, pressure injection system can not maintain reactor coolant inventory if certain size small breaks occur and the reactor coolant pumps are operating.
52.
That procedures now in effect at Rancho Seco direct operators to manually raise inventory on the feedwater side of the steam generator in the event of a feedwater transient.
5?.
That without the action described in requested admission number 54,above, natural condensation cooling can not be achieved at Rancho Seco.
"G.
That procedures now in effect at Rancho Seco direct operators to maintain operation of the high pressure injection system in the case of a small break LOCA.
1730 008 55.
That procedures now in effect at Rancho Seco allow operators to cease operation of the high pressure injection system after twenty minutes in the event of a feedwater transient accompanied by a 50 degree F. margin above saturation.
56.
That the Rancho Seco high pressure injection system will automatically be activated during many feedwater transients not accompanied by a small break LOCA.
57.
That if operators at Rancho Seco mistakenly shut off the high pressure injection system for any significant period of time -during a small break LOCA, voiding would occur in the primary system.
58.
That as a result of procedures now in effect at Rancho Seco, the high pressure injection system will operate more often than was anticipated when Rancho Seco was licensed.
59.
That as a result of modifications to Rancho Seco resulting from the NRC's May 7 order, the facility will experience a significantly greater number of reactor trips than was anticipated when Rancho Seco was licensed.
60.
That operators at Rancho Seco have no direct indication of all possible small break LOCAs and must deduce some of tnem from indications of resulting events such as loss of pressure or inventory in the primary system.
1730 009 61.
That Rancho Seco has no overpressurization protection system (such as controlled, filtered venting from containment) that would mitigate the consequences of containment failure.
62.
That the design of the Rancho Seco control room is such that operators must sometimes control the facility based upon indications not located within a functionally useful distance of the related controls without the assistance of a second operator.
Respectfully submitted, CALIFORNIA ENERGY COMMISSION z,-
Christopher Ellison l
Lawrence Coe Lanpher Attorneys for the California Energy Commission Date:
December 21, 1979 1730 010
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