ML20210B807
| ML20210B807 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1986 |
| From: | Bernero R NRC |
| To: | |
| Shared Package | |
| ML20209E138 | List: |
| References | |
| FOIA-87-10 NUDOCS 8702090239 | |
| Download: ML20210B807 (15) | |
Text
- ^
., [.
t.
Document Name:
BELGIUM PAPER Requestor's ID:
KRIESEL Author's Name:
RBernero Document Comments:
g g,/fff BELGIUM PAPER s..
1 i
b M
9 I
fd Z R -S 7-0 to 0702090239 870129 FOIA PDR PDR TYE87-10 i,
{0 -3/- S Q
U. S. APPRAISALS AND ACTION ON SEVERE ACCIDENT RISK Robert M. Bernero e
U. S. Nuclear Regulatory Commission Summary This paper traces the history and development of U.S. appraisals of severe accident risk.
The elements and basis of the U.S. Severe Accident Policy and Implementation Program are laid out.
Requirements of new reactor applications and existing plant assessments are described.
Regulatory changes in process or soon to come are described.
t 1.
INTRODUCTION i
A severe reactor accident is one in which substantial damage is i
done to the reactor core.
Chernobyl was indeed a severe accident.
The i
attention of the world is riveted on all nuclear power plants with the suspicion, with the fear that each one can be a Chernobyl in any I
moment.
This cycle of apprehension has occurred before, after the Three Mile Island accident in 1979. Many of us can recall those first few years after the Three Mile Island accident when so many lessons were i
being learned all over the world about severe accidents, especially how I
to prevent them. - I believe we have learned a great deal from the TMI accident - we had to because the stakes are high. The TMI accident I
destroyed a plant, at a cost of at least one billion U.S. dollars, and l
threatened worse.
Now the stakes are far higher. Chernobyl killed dozens, posed a latent cancer legacy for many and caused extensive property damage, even beyond the borders of its own country.
Now the future of nuclear power is at stake in many countries. What we in nuclear technology do now, how we do it, will have a profound influence on the outcome.
2.
BACKGROUND Appraisals of severe accidents in the U. S. go back to the earliest days of nuclear power activities. Many consider WASH-740 (1) the earliest major risk assessment for a U.S. light water reactor (LWR).
That study attempted some mechanistic analysis of core melt, fission product release, containment failure, and offsite consequences but 1acked much of the data needed to do a proper job.
In 1957 very little was known about the design of future nuclear power plants let alone about the physical phenomena of core melt and fission product transport. So little was known about design that WASH-740 was forced to use subjective estimates, or professional guesses, of system failure and release probabilities.
It was not until 1972 that the shape of the first 100 LWRs in the U.S. was clear. With that design knowledge, with at least a few reactor years of experience, and with a substantial amount of reactor safety research as a basis, work on WASH-1400 (2) was begun. This document, also called the Reactor Safety Study (RSS), was completed in 1975 and
_. _, ~. _ _ _
l l
i' stands even today as a great milestone and reference in severe accident risk.
The RSS analyzed one pressurized water reactor (PWR) and one boiling water reactor (BWR) in order to represent the spectrum of such designs used in the United States. The RSS was not widely acclaimed at first but hotly disputed.
It was only after three years of bitter dispute, in 1978, that it was essentially vindicated by the prestigious Lewis Committee Report (3), a peer review which justifiably criticized the shortcomings of the RSS but nevertheless recognized it as a great advance in reactor safety analysis and strongly recommended further use of this p.obabilistic risk assessment (PRA) in reactor safety work. The RSS then came into widespread influence in the U.S. and other countries as a basis for regulatory analysis, direction of research, and even as the basis for specific regulations.
Succeeding PRAs built upon and extended PRA methodology until now many plants in the U.S. and other countries have undergone such analysis.
i 3.
U.S. 3EVERE ACCIDENT POLICY l
In August of 1985 the U.S. Nuclear Regulatory Commission published a Severe Accident Policy Statement (4) along with the technical basis for that poli.ty.
That policy noted the wealth of severe accident analysis that had become available from the many PRAs which had been performed.
It also noted a great deal of corroborating experience derived from other ways of reviewing safety. The Policy Statement concluded that the present generation of LWRs was safe enough from severe accidents if one could be sure that no outliers, or plant specific vulnerabilities of exceptional risk, were present. The results from so many PRAs showed a core melt frequency on the order of 10 /yr and sufficient consequence mitigation by typical containments that the
-5
-6 frequency of large re' lease was on the order of 10 /yr or 10 /yr.
~7 Individual health risks of early fatality of 10 /yr or less were being calculated even for people very close to the plant.
These figures being calculated were essentially the same ones that were being put up as candidate safety goals. The conclusion was therefore a logical one; if one can eliminate the prospect of a special vulnerability, the typical plant is safe enough. The legacy of this policy was the obligation to do a systematic search through every reactor for these special vulnerabilities or outliers. The U.S. safety goal was published in 1986 (5).
The U.S. reactor owners banded together in 1980 to form the Industry Degraded Core Group (IDCOR) to perform independent studies of severe accidents in U.S. LWRs.
Their summary report (6) estimated U.S.
reactors to be even safer than was estimated by many of the PRAs, principally because many PRAs still used older WASH-1400 source term methodology. The IDCOR group recognized the need for a systematic search for outliers and pledged to develop effective methods to conduct such searches.
4.
U.S. SEVERE ACCIDENT POLICY IMPLEMENTATION The NRC staff developed a plan to implement the Severe Accident Policy and set this plan out in a Commission paper (7) in 1986. This plan covered the three major elements of (a) future applications, (b) existing plant assessments, and (c) the revision of regulatory practices to reflect new source term information.
e*
4.1 Future Applications The USNRC encourages the development of standard reactor designs.
The U.S. industry has suffered a great dissipation of its resources in supporting the design analysis and surveillance of such a bewildering array of different plants, over 100 in number, that we have in the U.S.
Our laws do not permit us to force standard designs but licensing conditions are now set to encourage their pursuit alone. New designs in the U.S. are acceptable if they:
a.
comply with all current regulations; b.
resolve all Unresolved Safety Issues as well as Medium-and High-Priority Generic Safety Issues; i
c.
have completed a PRA and used it in design analysis; and d.
NRC staff review of the design has been completed.
Procedures are also in place to encourage standard plant license applicants to go a step beyond NRC staff approval and to obtain Commission certification of the design.
Such certification would eliminate reactor design contention and litigation from individual licensing cases.
4
[
4.2 Existing Plant Assessments The IDCOR group has submitted proposed methods for Individual Plant Evaluations (IPE) to the NRC staff for review.
That review is going on at this time.
In addition, hRC is preparing guidelines and procedural criteria for the conduct of these reviews. By August of 1987 the approach should be clear and the work of IPEs begun.
In at least one area, generic requirements have been identified to supplant IPE and individuti plant decisions.
It has long been recognized that BWR pressure-suppression containments are especially vulnerable to failure in severe accidents. A set of generic i
requirements has been' proposed to U.S. BWR owners to enhance containment integrity and thereby close the severe accident issue with respect to these containments.
The basic requirements for the 24 Mark I containments are:
a.
Hydrogen:
Ensure strict control of inert atmosphere (N )
2 during plant operation;
~
b.
Containment Spray:
Provide a primary and two backup sources of drywell spray (one for blackout);
c.
Pressure Control:
Provide a reliable path to vent the wetwell to the stack if overpressure failure of containment is threatened; d.
Core Debris:
Understand the possible flow of core dv.bris and ensure that wetwell water will be retained in a contnlled area in the event of rupture; and e.
Procedures and Training:
Provide these for severe accident management.
Formal licensing requirements of this type are expected to be published in April 1987.
Similar requirements for BWR Mark II and Mark III containments are being formulated.
4.3 Revision of Regulatory Practices The NRC is proceeding on many fronts to revise regulatory practices in light of new severe accident information, ten areas have been identified for such revision:
O 4.3.1.)
Revised treatment of severe accidents in NRC risk estimates.
Following the TMI accident, the Commission directed the staff in June 1980 to provide a discussion of accident risks, including those for severe accidents, in the Environmental Impact Statement (EIS) prepared for each plant at the time of the application of the operating license. About 30 such EIS's have been issued since then.
There are no major technical issues for this area, and since the Commission policy statement directs the staff to discuss health and safety risk "in a sanner that i
fairly reflects the current state of knowledge," it is considered that this regulatory change has been effected, and future risk estimates will use the best available source terms.
4.3.2.)
Revision of requirements for PWR s) ray additive systems.
Chemical additives such as sodium lydroxide (NaOH) are usually added to U.S. PWR spray systems to enhance removal of elemental iodine. This requirement stems from the assumptions of 1962 source terms, TID-14844 (8). The spray additive system adds complexity, and that inadvertent operation may expose equipment to a potentially corrosive, damaging environment.
The NRC is investigating eliminating mandatory requirement of spray additives but retaining the requirement for post-accident pH control in the sump to prevent evolution of iodine i
from the sump solution.
The latest research information on iodine chemical form is being followed closely in this area.
l 4.3.3.)
Credit for fission product scrubbina in BWR suppression pools.
Present Standard Review Plans and regulatory guidance do not allow suppression pools to be considered as fission 2
product cleanup systems.
The source tern code package (STOP) developed under NRC research, however, leads to the conclusion that suppression pools are very effective at retaining fission products during severe accidents.
l The NRC is studying revisions to Regulatory Guide 1.3 and Standard Review Plan 6.5.3 to incorporate credit for i
suppression pools as fission product cleanup systems.
4.3.4.)
Containment Leaktichtness and Integrity The NRC is studying means of replacing the present emphasis on integrated leak rate with greater assurance against failure of containment isolation. Possible changes include a slight relaxation of the allowable leak rate, coupled with a more restrictive policy aimed at assuring that inadvertent breaches of containment integrity do not occur.
Factors to be considered include the possibility of continuous proof of containment integrity by maintaining a differential pressure vs. the
4 use of administrative procedures alone, and consideration of pressure relief to protect containments against overpressure.
4.3.5.)
Emergency Planning Current NRC regulations require emergency planning for two concentric areas around a nuclear power plant.
These are a ten mile radius plume exposure emergency planning zone (EPZ), where actions might be taken to protect individuals exposed to a radioactive plume, and a fifty mile radius ingestion pathway EPZ, where actions might be taken with regard to contaminated food or water.
These requirements stem largely from the accident source terms taken from WASH-1400.
The NRC plans to reassess emergency planning requirsents in light of revised accident source term information and the Chernobyl experience.
4.3.6.)
Control Room Habitability and Air Filtration Systems.
Current design and surveillance requirements for control rooms are dominated by provisions to filter the air they contain through charcoal (to remove radioactive iodine) and to attenuate direct gamma ray transmission from the containment, assuming an accident as described in TID-14844.
These provisions may not be optimal for the protection of plant personnel in the event of a severe accident.
The NRC is studying possible revisions to control room habitability requirements and filtration systems in light of these considerations.
4.3.7.)
Environment Qualification of Safety-Related Eculpment.
Safety-related equipment is presently either cesigned to survive in or be protected from the radiation environment defined by the TID-14844 assumptions and the temperature and pressure environment consistent with a l
of the radiation environments during selected severe large pipe break accident.
The NRC has begun estimations l
l accident scenarios, and will reassess the radiation environment in which safety-related equipment should be qualified to operate.
4.3.8.)
Safety Issue Prioritization.
Prioritization of safety issues is made using WASH-1400 i
accident source terms.
The staff anticipates revising the source terms used in prioritization to incorporate the insights gained from source term research.
Because the relative importance of the prioritized issues is not expected to change, no reprioritization is needed at this i
time.
Revision of the methodology is expected to and to be completed by November 1986.
4.3.9.)
Accident Monitoring and Management.
Equipment and procedures are required to be available in i
=-
s the event of a severe accident to be used in making decisions concerning the likely future course of the accident and the minimization of its impact.
Future capabilities of accident monitors and generic guidance on management of the course of accidents will be based upon revised source term information rather than WASH-1400.
Reassessment of instrumentation to monitor accident conditions (Regulatory Guide 1.97) could commence at the end of 1986.
4.3.10.)
Siting.
10 CF.t Part 100 requires that site and plant design, in combination, be such to yield doses no greater than specified guidelines values, based upon an accidental fission product release as defined in TID-14844.
The guideline values, in keeping with TID-14844 assumptions, are given in terms of the hazards of iodine and noble gas fission products.
The NRC will consider revising the guidelines to more accurately represent off-site hazards of severe accidents. This is expected to require a moderate effort over the next one to two years.
5.
CHERNOBYL INPLICATIONS j
By now many people using LWRs have said about Chernoby1; "It can't happen here!" We should not take false comfort from that partial truth.
Certainly an accident scenario such as occurred in Chernobyl can't happen in an LWR but for years we have estimated some LWR accident scenarios with releases greater than those at Chernobyl, see, for example, the releases in WASH-1400 (2) designated'as PWR-1, PWR-2, BWR-1 and BWR-2.
New severe accident or source tem information lowers their probability more than their quantity. The NRC is vigorously working on a Chernoby1 Implication Report which we expect to finish in December of this year. We are searching our knowledge of the Chernobyl events not for exclusion arguments but for possible implications of value in better control of LWR design and operation.
1 6.
SUPMARY AND CONCLUSIONS The U.S. was well along on a massive program of severe accident assessment when Chernobyl occurred. That accident has not suspended the l
U.S. effort but, if anything, adds impetus to it. The U.S. Severe l
Accident Program continues and major results are coming as it proceeds.
REFERENCES l
(1) USAEC, " Theoretical Possibilities and Consequences of Major l
Accidents on Large Nuclear Power Plants," WASH-740, March 1957.
i l
(2) USNRC, " Reactor Safety Study: An Assessment of Accident Risks in
(
U.S. Commercial Nuclear Power Plants" (WASH-1400), (NUREG-75/014),
October 1975.
(3) USNRC, " Risk Assessment Review Group to the U.S. Nuclear Regulatory Commission," H. W. Lewis, et al., NUREG/CR-0400, September 1978.
l
~ ~ - - - -
i r
a
~
p v
(4) USNRC, "NRC Policy on Future Reactor Designs, "NUREG-1070, July 1985.
(5) USNRC, Federal Register, " Safety Goals for the Operations of Nuclear Power Plants; Policy Statement," 10 CFR Part 50, August 4, 1986.
(6)
IDCOR, Technology for Energy Corp., " Technical Summary Report:
Nuclear Power Plant Response to Severe Accidents," November 1984.
(7) USNRC, " Implementation Plan for the Severe Accident Policy Statement and the Regulatory Use of New Source-Term Information,"
SECY-86-76, February 28, 1986.
(8) USAEC, " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, March 23, 1962.
e
'I v
p
,l.
g SEVERE ACCIDENT SAFETY IN BOILING WATER REACTORS WITH MARK I CONTAINMENT As the name indicates, a boiling water reactor (BWR) is a reactor in which the water fed to the reactor core boils right there in the reactor vessel and then passes as steam directly out to the turbine generator where its energy is converted to electricity.
The exhausted steam, after condensation, is returned to the reactor as feedwater.
Figure I shows a simple schematic of a BWR plant.
The reactor is enclosed in a special containment structure. The feedwater enters and the steam leaves this containment structure through multiple, large diameter pipes equipped with redundant valves which can be closed in an emergency.
In the pressure suppression containment which is used i
in all large U.S. BWRs, a very large quantity of water, up to one million gallons, is stored in a special compartment of the containment called the suppression pool.
Many auxiliary and emergency cooling systems are provided to pump cooling water into the reactor and to cool the containment atmosphere and its suppression pool.
If a pipe breaks by accident, the containment closes to isolate the reactor in the containment and many cooling systems are called into play to cool the reactor and the suppression pool, removing i
l the stored energy and heat generated by radioactive decay.
i Thus, the BWR is an open system removing large quantities of energy to nearby equipment which, in emergencies, converts to a closed system, basically l
relying on external cooling of the containment to remove the bottled-up r
energy.
Themost'commonlyps'ofpressureEppressi6ncontkinmentintheU.S.
is the Mark I type shown in Figure 2, which is used in the 24 U.S. BWRs listed in Table 1."The reactor is contained in the drywell portion of the containment, shaped like an electric light bulb standing upside down. The i
suppression pool partially fills a toroidal shell around the base of the
" bulb" and a series of ducts is installed to guide steam and other releases into the suppression pool which quenches the steam and also absorbs much of the radioactive material (except gases).
l
" Severe accidents" is the term most commonly used to describe accidents in which the reactor core is severely damaged.
As happened at Three Mile Island, i
prolonged loss of core cooling can allow the heat of radioactive decay in the p /}-$7-d/6 l
O[/.
L
u.
i i
the direct cycle
~
~
l boiling water reactor system
\\
\\
i Turbine Generator P
t l
Reactor l
Extraction Lines Condenser t
1 Separators & Dryers kru c= m3 + drc
- ~
Core
(.,.
h I
l l
Jet Jet (h
N i
Pump Pump
]
g f
Demineralizer Feed y
2, l Pump g
i n
L I-1 d
l I
ll d
I 4
1 li.. m _ _, __ E E
Ei _.
I i
.i,-
I
(
en Heater i
l I
l Recirculation Pumps Demineralizer i
l FIGURE 1.
I
,i s-
~
~ ^ '
I l
3
- igg Building Sh' If l
- 2. Refueling Floor j
),
- 3. Drywerr
- 4. Reactor Vessei
- 5. Pedestal I
s 7 00 c
er
~
~
- Concrete Shielding I ~..
w Figure 2.
Boiling Water Reactor Mark l Containment
F 4
TABLE 1 BOILING WATER REACTORS WITH MARK I CONTAINMENTS LICENSED OPERATING PLAN 4, POWER LICENSE NAME LEVEL DATE COUNTY STATE UTILITY BROWNS FERRY 1 3293 12/20/73 LIMESTONE COUNTY AL TVA BROWNS FERRY 2 3293 08/02/74 LIMESTONE COUNTY AL TVA BROWNS FERRY 3 3293 08/18/76 LIMESTONE COUNTY AL TVA BRUNSWICK 1 2436 11/12/76 BRUNSWICK COUNTY NC CAROLINA POWER & LIGHT BRUNSWICK 2 2436 12/27/74 BRUNSWICK COUNTY NC CAROLINA POWER & LIGHT COOPER 2381 01/18/74 NEMEHA COUNTY NE NEBRASKA PUBLIC POWER DISTRICT DRESDEN 2 2527 12/22/69 GRUNDY COUNTY IL COMMONWEALTH EDISON DRESDEN 3 2527 03/02/71 GRUNDY COUNTY IL COMMONWEALTH EDISON DUANE ARNOLD 1658 02/22/74 LINN COUNTY IA IOWA ELECTRIC POWER & LIGHT FERMI 2 3292 07/15/85 MONROE COUNTY MI DETROIT EDISON FITZPATRICK 2436 10/17/74 OSWEGO COUNTY NY POWER AUTHORITY OF STATE OF NY HATCH 1 2436 10/13/74 APPLING COUNTY GA GEORGIA POWER HATCH 2 2436 06/13/78 APPLING COUNTY GA GEORGIA POWER HOPE CREEK 1 3293 04/11/86 SALEM COUNTY NJ PUBLIC SERVICE ELECTRIC & GAS MILLSTONE 1 2011 10/16/70 NEW LONDON CT NORTHEAST NUCLEAR ENERGY MONTICELLO 1670 01/19/71 W!!IGHT COUNTY MN NORTHERN STATES POWER NINE MILE POINT 1 1850 08/22/69 OSWEGO COUNTY NY NIAGARA M0 HAWK POWER OYSTER CREEK 1 1930 08/01/69 OCEAN COUNTY NJ GPU NUCLEAR CORP PEACH BOTTOM 2 3293
-12/14/73 YORK COUNTY PA PHILADELPHIA ELECTRIC PEACH BOTTOM 3 3293 07/02/74 YORK COUNTY PA PHILADELPHIA ELECTRIC PILGRIM 1998 06/08/72 PLYMOUTH COUNTY MA BOSTON EDISON QUAD CITIES 1 2511 12/14/72 ROCK ISLAND COUNTY IL COMMONWEALTH EDISON QUAD CITIES 2 2511
~12/14/72 ROCK ISLAND COUNTY IL COMMONWEALTH EDISON VERMONT YANKEE 1593 02/02/73 WINDHAM COUNTY VT VERMONT YANKEE NUCLEAR POWER
- t-t 1
coretobuild.uftothepointthatthefuelbeginstodisintegrate,the zirconium metal cladding melts or reacts with residual steam to form a
combustible hydrogen, and even the ceramic uranium oxide fuel pellets can mel t.
'k' great deal cf attention is being given to understanding the behavior of reactors and their containments in severe accidents, especially since the Three Mile Island accident. The objectives are to ensure that the likelihood N
of core melt accidents-is very low and that, should one occur, there is substantial assurance that the containment will mitigate its consequences.
~
a The severe accident behavior of a BWR with a Mark I containment, the Peach Bottom plant, was assessed in'the Reactor Safety Study (WASH-1400 or NUREG-75/014) which was published in 1975.
That study indicated a relatively low overall risk for the BWR, principally due to its ability to prevent core melt.
The containment was estimated to provide very little mitigation of core melt consequences because ths buildup of pressure under accident conditions would be a direct cause of containment failure unless adequate cooling was preserved.
Consistent with operating procedures in place in 1975, the Study assumed little effort by.the reactor operators which might effectively preserve' the# containment's integrity.
~
The situation, more than ten years later, is different and still changing for the better.
It is recognized today that molten core material melting into the groundthrougithethickcontainmentbaseisnottheprincipalthreat;rather, it is an atmospheric release of radioactive material which is the principal threat. The principal factors which can cause containment failure with atmospheric release are hydrogen ignition, gas overpressure buildup to rupture, and direct attack cf the 'drywell by core melt debris. The general situation for each of these is summarized as follows:
Hydrogen Ionition Recog6izing Ubat combustible hydrogen can be generated and released in severe accidents 7 all Mark I containments now are purged and filled with inert nitrogen gas during operation so that even if hydrogen gas is formed it has insufficient.
I oxygen available to support combustion.
Remaining questions in this area relate to how long the containment may be without this inert atmosphere in order to s
... 1 permit inspections, 'and how air might leak in or hydrogen leak out to nearby rooms under accident conditions.
t Overpressure Failure Careful analysis indicates that a typical Mark I containment can withstand pressures of more than twice the design pressure without rupture.
Nevertheless, severe accidents in the extreme can generate such pressures and cause containment rupture. Overpressure damage control procedures have been developed for pressure suppression containments and are already in place for operator use.
With these procedures the containment remains closed for most accident conditions; but, if overpressure failure threatens, large vent valves above the suppression pool chamber are opened so that the excess pressure is released gradually by bubbling the releases through the pool, forming a filtered vent containment system. With this path assured, virtually,nothing but the noble gases are released. The radioactive noble gases pose a modest exposure threat offsite only in the area very close to the plant. A number of questions are being pursued in this area. All the plants have suitably large vent valves and ducts but.they vary one to another in the ability to open
~
these valves under accident conditions.
The valves are designed for highly reliable closure, not opening.
Consideration is being given to modifying valve controls.
In addition, the vent ductwork downstream of the valves may warrant modification.
In most plants it is fairly light ga,uge ductwork and T
might be breached in accident venting.
If so, consideratio'n is being given to the effects of secondary release of radioactive gas, hydrogen, and perhaps steam into the reactor building.
l l
l Direct Attack The core melt debris, since it has melted through the reactor vessel into the drywell may, by direct radiation of heat, cause failure of connections in the drywell shell; or the debris, if sufficiently fluid, may flow out to the wall
~
and melt through the steel.
The Mark I containments have one or more spray systems in the drywell which are able to spray water along the walls and onto
~
the floor of the drywell inhibiting direct attack.
Concerns in this area are in three general areas:
core debris modeling, shell and concrete attack modeling, and spray reliability.
In the first area, it is recognized that a molten reactor core, to melt through the bottom of a BWR, must dissolve a very
s 7
_4 large amount of inert metal in the lower reactor vessel, probably diluting the core melt.
The key question is whether the melt would come out moving sluggishly like Hawaiian volcano lava or as a hot free flowing liquid.
The latter is the more threatening condition.
If core melt debris reaches the concrete floor and steel shell of the wall, it is important to understand that the path to the outside that might be opened bypasses the beneficial scrubbing of radioactive material passing through the pool.
i As noted earlier all these plants have drywell spray systems, but they are designed as a secondary mode of operation for a reactor safety system.
Strong consideration is being given to enabling hookup of these systems to fire protection systems so that spray capability is almost always available.
Substantially different emergency operating procedures and training were put in place at all reactors after the Three Mile Island accident; further improvements in these procedures are still being made.
For the Mark I containments both industry and NRC studies are being used to identify the best combined strategy for procedures and perhaps some changes in equipment such'as alternate vent paths, or improved valve operability.
The Mark I studies are being given highest priority by the NRC staff and the industry. The expectation is that, with modest improvements of this type, one can achieve substantial assurance of core melt consequences mitigation by a Mark I containment.
i O
_