ML19296C039
| ML19296C039 | |
| Person / Time | |
|---|---|
| Site: | Black Fox |
| Issue date: | 03/29/1977 |
| From: | Dalton A Citizens Association for Sound Energy, DALTON, A.T. |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML19296C015 | List: |
| References | |
| NUDOCS 8002250106 | |
| Download: ML19296C039 (6) | |
Text
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.*. }.y EXHIBIT D
. SQ UNT?ED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD IN THE MATTER OF
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DOCKET NOS. STN 50-556 PUBLIC SERVICE COMPANY OF
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STN 50-557 OELABOMA, ASSOCIATED ELEC-
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TRIC COOPERATIVE, INC. AND
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WESTERN FARMERS ELECTRIC
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COOPERATIVE, INC.
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(Black Fox Stations Units 1 )
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INTERVENORS ', C. A.S.E., ILENE YOUNCHEIN, LAWRENCE BURdELL AND CLARK GLYMORE, RESPONSE TO BOARD'S REQUEST FOR CREDIBLE MECHANISMS The Atonio Safety and Licensing Board is presently considering admitting into controversy the issue of whether class nine accidents--that is, all of those accidents which are more severs than the design basia accidents--should be anilyzed and evaluated for their likelihood and potential hamful consequences.
The
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Board has ruled that this matter is "nnMaE dvisema'at pending "M the production of a credible mechanism" for such' accidents to I '
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occur, other than the possibility of a spontaneous pressure
.4 (reactor) vessel failure or a failure of the emargency core cooling system (ECCS) to operate in the event of a design basis, loss of coolant accident.
This present pleading supplies
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avamples of credible mechanisms.
These are as follows:
The Board's order was received March 4, 1977.
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- 1. _ Control rod drop out without. scram.
The control rod disengages or breaks clear of its driva shaft, due to design.
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or manufacturing defects, or flow-induced vibration effects,
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or both, which no one can predict but which no one can show to be impossible (e.g., manufacturing defects could conceivably lead to flow induced vibration and metal fatigue or fretting corronicas so such failure mechanisms < annat be ruled out as impossible), and the rod falls out of the core by gravity, and the scram system fails to function.
A severe power excursica could result.
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2.
A control rod drop out occurs involving a high worth of reactivity, due to a failure of the systems designed to control the reactivity worths of the control rods.
A severs power excursion could result.
3.
Control rod ejection.
A faulty control rod drive mer hanism housing ruptures under pressure and breah off the f.
reactor vessel bottom, causing the associated control rod to be ejected, provided the control rod drive housing support structure is missing or ineffective, due to faulty assembly, faulty manufacture, or left uninstalled due to human error or carelessness.
(The Vermont Yankee criticality incident underscores the possibility for human error and violation of safety procedures.)
A power excursion accident could result.
Conceivably, a faulty support structure might prevent thermal expansion of the control rod drive mechanism housings, damaging both the mechanism housings and the support structure.
4.
A multiple control rod ejection accident.
Assuming i
conditions of No. 3 above, the steam pressure shock wave amanating from the housing rupture, or a pressure pulse from i
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Na the core due to an initial power excursion, could conceivably e
impose sufficient stress loading on adjacent control rod housings, which could be similarly weak (due to a common defect), to cause them to rupture, as in a cascade, thereby amplifying the power excursion accident, by the ejaction of additional control rods.
5.
All manner of steam valve closure transients without scram ands (a) without the tripping of both recirculation pumps; or (b) without the successful functioning of the entire sequences of associated safety actions, consisting of the tripping of both recirculs. tion pumps and the feedwater pumps, high pressure ecolant injection, proper opening and closing of the relief valves, and boron injection.
Such incidents could be caused, for example, by an inadvertant f.
trip of the main steam isolation valve (s), or by an uncontrolled startup of a coolant recirculation pump without scram, which would eventually lead to the rapid closure of the steam isolation valve (s).
6.
A continuous or intermittent control r:.td withdrawal without scram.
7.
A severe power excursion accident, triggered by one of the above causes, coincident with defective (weak) reactor vessel head closure bolta.
The pressure surge or shock wave caused by the power excursion could provide tha extra loading needed to break the bolts and allow the closure head to be blown off or the reactor vessel to noro easily be ruptured by a power excursion.
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A main coolant pipe rupture without scram.
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The suptsgo of the = min steam pipe without scram.
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A pressure surge, due to any transient of the class
" anticipated transients", causes a coolant pipe rupture, due to defective piping on the verge of rupture.
11.
Blockage o' coolant flow to one or more fuel assembly ducts with and without scram, due to a rag or other foreign object.
12.
xisposicianing of control rods at full power, causing excessive power peaking; and gross fuel lo d ing errors.
13.
All of the above at various power levels and zenon states.
14.
Massive fuel crumbling in a local region of the core due to fuel manufacturing defects, followed by a failure to shut-down the reactor.
15 Same as 14 above, except that the scram functions, 16.
All of the above with the added assumption of faulty f
fuel rod spacer grids.
17.
All of the above with the added assumption of a faulty system of coolant flow orifices in the inlets to the fuel assemblies, due to manufacturing and assembly errors.
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18.
All possible criticality accidents in the reactor during ref r ling, such as a rod withdrawal (control rod) or a control rod drop-out without scram.
This would occur when the contm4 = nt is open.
The reactor vessel may be open as well.
19.
A control red housing rupture assuming the housing support structure is in place.
Analyse the possibility of a chain reaction failure of adjacent mechanisms, which could impc de excessive loading
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on the support structure.
20.
Analyse the possibilities for reactor vessel movements (such as due to a power excursion accident, loss-of-coolant accident,
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,or an earthquake or saboteur's bomb) and associated reactivity effects due to coolant motion caused by the vessel movements.
s 21.
All of the above plus all design basis accidents with the added assumption of one, and more than one, control rods stuck in place.
22.
Analyze for the possibility of a slug of water thrown upwards from the core in a power excursion, rebounding off the reactor vessel head and internals, and reentering the core rapidly to trigger a severe secondary power excursion by an increase in neutron moderation.
23.
A power excursion accident (any one of the above listed ones) coincident with heavily corroded or weak fuel rods, due to manufacturing defects or inadequate strength, due to a loss of fuel rod strength against power excursions caused by neutron
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irradiation, which might develop because of a lack of adequate fuel experimental programs to establish fuel rod performance capabilities.
24.
A power excursion accident with plutonium recycled fuel in which the plutonium oxide is not well mixed with the uranium oxide during manufacture.
This could not only impair fuel rod strength but might possibly reduce the negative doppler reactivity feedback as well during power excursions.
25.
Other accident possibilities.
The NPC should provide the parties with a list of all accident possibilities of which they have knowledge which are more severe than the design basis accidents and which are not mentioned above.
7 26.
Vessel rupture due to cold. emergency cooling water.
This s
refers to the thermal shock of interjecting cold cooling water.
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'see Marshall Report, Great Britain, 1976.
7 27.
Zipper (sequential failure from sequential or incremental stress) of head bolts.
ANDREW T. DALTON, JR.
Attorney for Intervenors 2536 East 51 Street Tulsa, OK 74105 m
Datad this 29th day of March, 1977.
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