ML19260E223
| ML19260E223 | |
| Person / Time | |
|---|---|
| Issue date: | 12/20/1979 |
| From: | Kastenberg W Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-SM-0147, ACRS-SM-147, NUDOCS 8002140110 | |
| Download: ML19260E223 (7) | |
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'h3/79 HYDROGEt1 GENERATICN AFD_ PRESSLTS SUPPRESSICN CONTAINMDITS W. E. Kastenberg, ACRS Senior Fellow 1.
INTRODrJCTION This note is intended as a follow-on to a previous note (1) which discussed the hydrogen design basis, the potential implications of NI-2 hydrogen genera-tion on design basis accidents in general, and implications for future licensing and safety analysis. Since that time, the NRC-MI-2 Les: ens Learned Task Force has issuted its final report (2), which includes recommendations on certain design features for mitigating accidents that are not provided by consideration of the set of design basis events.
In particular, the Task Force recommends that the Commission, " issue within three months a notice of intent to conduct rule making to solicit comments on the issues and facts relating to the censideration of design features to mitiga_e accidents that would result in (a) core-melt and (b) severe core damage, but not substantial melting." (2)
In particular, consideration of the employment of controlled, filtered venting for core-melt accidents and provisions to cope with hydrogen generation (beyond the current design basis) would be the focal point of the rulemaking procedure (3).
In this note, PWR ice condenser and BWR-Mark III, pressure suppression con-tainments are considered in anticipation of the rulemaking process. Some comments with respect to the use vent-filtered containments, inerting and core melt are also presented.
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II.
BACKGROUND Before commenting on the consideration of inerting and the use of vcnt-i i 1.;1tered containment it is useful to compare some pertinent character st cs Tables I and II contain some of BWR-Mark III and ice o>ndenser containments.
relevant information for typical BWR-Mark I, II and III containments aid, PWR S e evolution of BWR-containments large dry and ice condenser containments.
(ftsm Mark I to Mark III) can be characterized as moving from small volume, high In contrast, the PWR pressure design to a larger volume, low pressure design.
containment has moved from large volume, high pressure design to a snaller volume, We ice condenser and BWR-Mark III containments have con-low pressure design.
verged on comparable volumes (1.2 to 1.6 million cubic feet) and comparable design pressures (12-30 psig) te results of two design aasis accidents presented in the FSAR of the It is Grand Gulf Generating Station (a BWR Mark III), are shown in Table III.
interesting to note that peak pressure occurs within 1.3 seconds of the initiation For the ice condenser design, peak pressures of 9.0 and 11.8 psig of the event.
4 cec respectively for a (12.0 psig design pressure) occur at 50 sec and 10 double ended pump suction break, with the latter corresponding to the point where ice melting is complete.
III. INERTING CONTAINMIh7 At the present time only BWR-Mark I containments (~26), with the exception of Vermont Yankee and Hatch 2 are inerted. We NRC-TMI-2 Lessons Learned Task Force, in.ts Short Term Recommendations, recommends that all Mark I and Mark II contain-Bis was based in part because the small volumes left a small ments be inerted.
. It was felt that with margin available for acconrodating metal-water reactions.
the larger volume in PWR and Mark III containments, there is greater capability The Staff estimates to acccrxnodate the hydrogen produced in metal water reactions.
that it requires 13% clad metal-water reaction to reach the flarability limit in an ice condenser and 20 to 30% in a large, dry EWR containment, and 8% in a Mark Within the current regulations, these would be exempt from inerting.
III.
- 1. 's making, the design basis for ccrabustible gas control As a result v.
will most likely change (based on the 40% metal-water reaction calculated for If the design basis is 30% or greater, than all contaiments would have TMI-2).
to meet the new requirements for corrbustible gas control.
IV. VENT-FILTERED CONTAINMENT As stated above, a major aspect of the rule making procedure would be the In their presen-consideration of the employment of vent-filtered containment.
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tation before the ACRS subcommittee on 7MI-2 Implications, the NRC Staff consi ere combustible gas control and the use of vent-filtered containment separate issues.
Also considered are (The final report on Lessons Learned lumps them together).
degraded cores (as a result of events involving degraded core cooling, but events arrested short of core-melt) where the previous design basis is adequate save for hydrogen control, and core melt considerations (beyond the design basis) wher use of vent-filtered containment would be utilized.
It was stated by the Staff that if one had filtered-vented containment at Three Mile Island, it wuld not have been used. Cnly if the core had melted, ard ld the contaiment pressure has risen to some predetermined threshold level, wou it have been used.
_4-IV.
SOME CONEIDERATIONS Based >n the above discussion, the following points should be considered:
'l)
Although there are no BWR-Mark III contaiments on presently operating reactors (Grand Gulf, the lead plant is scheduled for AGS OL Review in November 1980), they should be considered as candidates for inerting along with the Mark I and II containments. 'Ihis consideration should be made in spite of the fact that Grand Gulf will be equipped with two hydrogen reembiners to cope with post LOCA radiolysis.
(2) Consideration should be given to the integration of the hydrogen control systam and the potential use of vent-filtering for those contain-ments with low design pressures (Mark III's and ice condensers). For degraded cores (beyond the current design basis), a large anount of fission product activity (noble gases and volitiles) might be released to the For these. ow pressure contairraents, the principle modes of l
containment.
failure include overcressurization due to non-condensible gases as wil as hydrogen burning. It is not at all clear that one can be treated separately than the other.
(3)
In the consideration of core melt, both in the final report of the NRC Staff task force and in their presentation at the subcommittee meeting, the only mitigating system mentioned is vent-filtered contain-No consideration has been given to molten core retention such as me t.
the core ladle considered for floating nuclear power plants.
For the low pressure contaiments (e.g. an ice condenser) the primary mode of containment failure following a Wss of Coolant Accident (LOCA), falling beyond the design basis, is overpressurization due to noncondensible gases (3).
Hence containment failure can potentially be averted in all types degraded
{
l cores using a vent-filter approach. For high pressure contairinents (e.g.
l Mark I and large dry IHR), the principle failure modes include steam explo-sions. While a vent-filtered approach might mitigate against the short tenn threat to containment due to a steam explosion (in the vessel), the question of long term effects due to vessel and base mat penetration need to be answered.
In view of this the Staff should consider different accider.t scenarios for different reactor containment designs and consider whether or not retention should be part of mitigation. Consideration of "beyond design basis accidents" should include the investigation of degraded core coolability and molten core retention following such an event.
(4) The use of any mitigating device or system to be employed to cope with events beyond the design basis should be predicated on some measure of risk reduction.
REFERENCES:
1.
W.E. Kastenberg, "mI-2 Hydrogen Generation: Licensing Implications and Design Basis Accidents" Memo to D. Okrent, Oct. 3,1979.
"mI-2 Lessons Learned Task Force, Final Report" NUREG-0585, October 1979.
2.
" Liquid Pathway Generic Sttx3y" NUREG-0440, February 1978.
3.
WASH-1400 Appendix VIII, Sections 2.2.7 and 3.2.7 (See also Appendix D 4.
to Appendix VIII).
TABLE I Centainment Parameters for BWRs Long Term
- sis" "r*SS"r*
WetWellF5*)*
DrywellFrge Design Pressure Volume (ft (psig) h Volume (ft )
(psig)
Mark I 159,00.0 52-62 117,245 62 (Dresden)
Mark II 180,000 45 95,350 45 (Zinner)
Mark III 270,000 30 1,400,000 15 (Grand Gulf)
TABLE II Containment Parameters for PhRs 3
g Free Volume ft Design Pressure (psig) 6 Large Dry 2.6x10 75 (?)
(Diable Canyon) 6 Ice Condenser 1.13x10 12 (Sequoyah)
TABLE III SHORT-TERM ACCIDENT RESULTS FOR GRAND GULF (MARK III)
Recirculation Steam Line Line Break Break 1.
Peak drywell pressure 19.4 (30)~*/
22.0 (30)~*/
13ig 2.
Time of peak pressure 1.3 1.3 Sec 3.
Peak drywell temperature 240 330 OF 4.
Peak wetwell pressure N.A.
9.0 (15)-*/
psig
- /
Design Pressure