ML20006F576
| ML20006F576 | |
| Person / Time | |
|---|---|
| Issue date: | 10/25/1989 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2673, NUDOCS 9002280195 | |
| Download: ML20006F576 (58) | |
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ACRS 00lNT SUBCONilITTEES MEETING
SUMMARY
/ MINUTES FOR CONTAINMENT SYSTEMS / STRUCTURAL ENGINEERING OCTOBER 17, 1989 ROSEMONT, ILLIN0!$
PURPOSE The ACRS Subcommittees on Containment Systems and Structural Engineering held a joint meeting on October 17,19P.9 in Rosernont, Illinois. The purpose of this meeting was to continue the discussion in regard to the development of an ACRS paper on containment design criteria for future plants bcsed on present knowledge. A copy of the meeting agenda and selected slides from the presentetions are attached.
The meeting began et 8:30 a.m. and adjourned at 4:00 p.m., and was held entirely in open session. The principal attendees were as follows:
MIENDEES ACRS INVITED SPEAKERS D. Wcro, Chairmar R. Henry FAI J. Cerro11, Member L. Minnick, Private Consultant C. Wylie, Member P. North, EG60-Idaho M. Bender, Consultant W. Snyder, SNL D. Houston, Staff W. von Riesemann, SNL A. Walser, Sargent and Lundy-N,.RC B. Hardin, RES G. Bagchi, NRR REVIEW DOCUMENTS There were no formal documents to be reviewed at this meeting.
The ACRS effort on this subject is in response to a Staff Requirements Memorandum-dated July 28, 1908, which was written following an ACRS meeting with the Commission on July 14, 1988.
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Containatnt Systems / Structural Engineering Meeting Minutes October 17, 1989 f
ACTIONS, AGREEMENTS AhD CO MITMENTS j
None i
i DISCUSSION i
In his opening couents. D. L'cro expressed regrets that C. Siess, Chairman of the Structural Engineering Subcomittee, could not attend due to 111ntss. He indicated that the purpose of the meeting was to discuss contai ment design criteria for future plants.
He stated that over the last five to ten years, there has been a considerable growth of 4
scientific information and a general understanding of the nature of severe accidents. However, this has not jelled into new guidance for designers to use when considering cer.tainments or containment systems.
He indicated that this was an information gathering meeting to aid in the development of new guidance or design criteria.
W. Synder (SNL) expressed his opinion that it is very timely to develop l
a modern set of containment system design criteria, but he also feels that it night be too late for some of the advanced designs already on the drawing board. Hc believes the concept of multiple barriers should I
be retained.
He indicated that the NSSS design is bottom-up engineering i
whilethebalanceofplant(BOP)istop-dcen.
He stated that 70 to 80 percent of outages at the plants originate in the BOP. He recommended that the total plant be designed on the bottom-up approach to achieve balanced reliability performance across the whole plant.
Further, he recomendeo that all systems be classified as safety systems, dropping the notations of safety-related and non-safety.
This approach is being taken in France. He indicated there is a reluctance in the industry to l
embrace these ideas because of the legacy embedded in the regulatory process. He stated that he has been close to the PRA studies-over the 3
past 15 years and that he is uneasy about the conclusions one draws from
-PRAs. He feels that better conclusions can be drawn from conventional reliability analysis.
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Cer,teinment Systems / Structural I
Engineering keeting Minutes October 17, 1989 1
P. North (EG&G) discussed the philosophical foundation for the growth of
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f nuclear energy both in the United States and the World. He indicated that the largest growth would be expected in areas with low current per l
capita energy consumption (~1/8 of USA).
Fossil fuel plants have become a greater concern (acid rain, C0, greenhouse effect). He l
2 indicated that a large number of people must support the use of nuclear energy if it it to inake an appropriate contribution. He discussed a foundation to provide the basis for this support based on:
(1)contain-l ment criteria lir.ked to clear protective objectives, (2) criteria that I
allow progressive design innovation, and (3) an approach based on rising j
standards of adequacy. He felt that a judgement by the Consnission at this tirne that a traditional containment structure is necessary would be disappointing en a technical basis. He reconenended a sound engineering 5
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approach base's on best estimate analyses with explicit factors of safety aoded. He indicated that new systens must demonstrate a i
robustness in achieving the containment function and that.there should
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be a balance between prevention and mitigation. He noted that longer plant lifetimes might be possible (80 to 100 years). He discussed the approech related to protective objectives:
near term similar to EPRI ALWR requirements and long term as eliminating the need for offsite emergency planning.
He also recommended testing of a full scale prototype to demonstrate analytical validation and fault tolerance 4 With these assurances, one could allow progressive design flexibility-and strong societal support, i
R. Henry (FAI) addressed the question of whether design criteria for containments should be altered.
He indicated that there are only two types of containment to be considered:
(1) large drys and (2) pressure suppression. These can be designed to:
(1)containfissionproducts, (2)passivelycontainstoredenergy,and(3)provideforheatremoval i
over the long term. He provided some calculational results to support various designs to contain stored energy.
He concluded that current criteria are bounding and well conceived, thus should be retained.- For i
fission product retention, he indicated that containments must have an-i
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i Centainment Systems / Structural Engineering Peeting tlinutes October 17, 1909 j
I integral steel liner. He reviewed the observations made at Titl-2 and Chernobyl end concluded that current criteria are well conceived. He indicated that future designs sheuld focus on:
(1)addingwatertothe' l
core or cavity to cool debris and protect the liner and (2) imbedding the liner in concrete to minimize thermal loads. He discussed severe l
accident issues and indicated how these could be addressed to enhance the capabilities in the following areas:
(1)hydrogencombustion,.(2) l liner protection. (3) tunnel configuration to restrict debris dispersal, i
and (4) containment floor design to achieve maximum cooling and minimum debris accumulation.
M. Pender (Querytech) discussed the system concept to define contain-t ment. This definition included a boundary closure, a heat sink and a
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radionuclide trapping or stabilizing capability.
He then discussed some characteristics of current containments and reviewed reactor accident cxperience.
He indicated that neither PWR nor BWR containments would I
contair en ATWS. He noted that no accidents with radionuclide releases have been experienced at high power.
He stated that one should have a design basis accident concept but that one should censider realistic times for accident progression and recovery activities.. He'recomended effective accident sensing devices and systems for controlled contain-rientfailure(ventingshculdbeconsidered). He indicated that not enough attention has been ghen to make things better if the containment failed.
He expressed a concern that pressure vessel / concrete codes are not well integrated with regulations and that too few engineers really understand the codes.
He discussed a number of aspects that should be revisited to determine the right basis for evaluating containments.
L. Minnick (Private Consultant) reviewed the historical development of containments for Yankee-Rowe and Connecticut Yankee. He indicated that there was a reluctance to install a pressure relief system on these early designs. He recomended that a passive means for cooling core debris and for relievirg containment overpressure be considered for
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j Centainrent Systems / Structural Engineering Peeting Minutes October 17, 1989 future reactors.
He further indicated that these devices should have a
-I minimal effect on the basic design of the plant and that these devices must provide a substantial improvement in safety assurance. There should be a careful analysis of any detrimental effects from these devices. He then discussed a self-actucted pressure-relief device for reactor ccntainments. This device was conceived by L. Minnick and investigated for EpRI by Sol Levy, Inc. A copy of the EPRI report was prov Wed. This system is comprised of multiple standpipes with water box seals and was reported to cost about $15H. Mr. Minnick discussed the operational features of the system to relieve pressure, to scrub fission products, and to provioe water inside containment.
i A. Walser (Sargent & Lunty) discussed containment design criteria from the standpoint of a structural engineer.
As a designer, one needs to knew the applied loads with some time dependency and probability of occurrence. Given that information and the space requirtments for the i
plant, one can then design and build a suitable containment. He re-viewed the current requirements for containment design: LOCA loads from the NRC regulations and combined LOCA plus 1/2 SSE from the ASME code.
He discussed safety factors and the effect of discontinuities (pene-trations, hatches,stiffners). He indicated that the effect of discon-l tinuities can not be codified. He concluded that current design crite-ria are adequate and should not be changed in the near future.
If they are changed, he reconrnended that an industrial task force with input 5
from research, universities and NRC be formed to address the matter.
Another recommendation was that ASME codes be revised from deterministic to probabilistic in terms of load factors and.allowables, and with an emphasis on ductility. He estimated that the efforts in his recorriendations would require about 20 years to complete.
W. von Riesemann (SNL) presented his personal thoughts on the subject of containment design criteria, mostly for LWRs. He discussed the primary and secondary purposes for containnents. One secondary purpose is to
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1 Conteinment Systems / Structural Engineering Meeting Minutes October 17, 1989 protect against external threats - missiles, tornadoes and sabotage.
He indicated that the contaircent is a system, not an isolated component.
Its performance depends on the response of the parts and any possible interactions.
He discussed the current approach to designs and the lessons learned from scale tests performed mostly at SNL. He noted that a decade of knowledge en containment behavior and severe accidents has rot been factored into the ASME code and in agreement with A. Walser, recomended that a cemittee be formed to revise the code considering the containment as a system. He discussed goals and some potential difficult points for new requirements.
He noted that a probabilistic
,;4 design approach is beyond the current state-of-the-art.
In the wrap-up session, W. Snyder emphasized a need for better commu-nichtions between the severe accident analysts and the civil engineers.
He felt that civil cr.gineers would have to change their philosophies when designing systems that may go beyond the elastic limit. He also stated that in his discussions with designers, he believes that they are already a helf step beyond current reciuirements for the next generation of plants.
R. Henry encouraged designers to think more in tems of thermal loads than pressure loads.
He also endorsed a more realistic-approach to integrated leak test (proposed by W. von Riesemann) and a l
rnore realistic source term analysis.
M. Bender emphasized the' system approach for containment design and the load conditions as a function of time.
W. von Riesemann proposed an ACRS workshop with all interested parties to discuss the conclusions drawn from the joint Subcommittee meetings,aproposalalsoendorsedbyG.Bagchi(NRC/NRR).
B. Hardin (NRC/RES) discussed the status of staff activities for evolutionary plants and indicated that efforts for improving the source tem are being reactivated. He also discussed the disagreement between the NRC l
and industry in respect to the metal water reaction for hydrogen calcu-lations [100% MWR (NRC) vs 75% MWR (Ind)].
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i Containment Systems / Structural l
Engineering Meeting Minutes October 17, 1989 NOTE:
Additionti meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, NW, Washington, DC 20006.(202)634-3273,or-can be purchased from Ann Riley and Associates, Ltd., 1612 K Street, hW, Suite 300, Washington, DC 20006 (202) 293-3950.
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ACRS JOINT SUBC0fEITTEE MEETING CCNTAINMENT SYSTEMS / STRUCTURAL ENGINEERING i
CCTOBER 17, 1969 ROSEMONT, ILLINDIS
- TENTATIVE AGENDA -
1 CONTAlkEENT DESIGN CLITERIA FOP. FUTURE NUCLEAR PLANTS A.
Subconsnittre Chaimen Renarks D. Ward /
8:30 a.m.
C. Siess ACRS I
INVITED SPEAKERS B.
Bill Snyder, SNL 6:45 a.m.
C.
Ptul North, EGl.G 9:30 a.m.
- BREAK ****
10:15 - 10:30 a.m.
D.
Bob Henry, FAI 10:30 a.m.
E.
Mike Eender, Qutrytech 11:15 a.m.
l
- LUNCH ****
12:00 - 1:00 p.m.
F.
Larry Minnick, Private Consultant 1:00 p.m.
G.
Adolph Falser, Sargent & Lundy 1:45 p.m.
- BREAK ****
?:30 - 2:45 p.m.
H.
Welt Von Riesenann, SNL 2:45 p.m.
I.
Subcommittees Discussion 3:30 p.m.
J.
Adjournment 5:00 p.m.
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Presentation l
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ACRS Joint Subcommittees' Meeting on Containment Systems and Structural Engineering
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A. Wm. Snyder l
Sandia National Laboratories i
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October 17,1989
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i Containment System Design Criteria
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Future Generations of U. S. Nuclear Power Plants t
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A difficult challenge, given:
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the bias of the legacy being limited to the LWR experience i
the investment in making a success of the concepts and designs of current plants i
current institutionalization of the U. S. nuclear power enterprise
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the sharply focussed attention being given to I
i the understanding of severe accidents the predictions of threats to and the response of l
contemporary containments l
ACRS Joint Subcommmeet Meeting AWS:10/15/89 L
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"'.j The Concept and Design Legacy Vis-a-Vis An Alternative Future Design Agenda
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The Concept and Design Legacy:
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design approach j
NSSS;~predominantly " bottom up" j
Balance-of-Plant; predominantly " top down"-
e safety systems; mostly additions / auxiliaries to the base plant a
the containment. building, last barrier of the multiple defenses-in-depth, designed to withstand a surrogate (DBA) for all plausible accidents the multiple sequential barriers of the defenses-in-depth susceptible e
to common cause and interdependent failures-4 1
ACRS Joint Subcommittees' Meeting AWS:10/15/89 d
.1 (2 of 3)-
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TheLConcept.and Design Legacy Vis-a-Vis An Alternative Future Design Agenda 1
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'* 1 An Alternative Future Design Approach
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NSSS & BOP; both designs mostly " bottom up
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.no distinctions between safety, safety-related, and non-safety l
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i ACRS Joint Subcommittees' Meeting.
AWS: 10/15/89' I-y i
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The Concept and Design Legacy Vis-a-Vis An Alternative Future Design Agenda
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An Alternative Future Design Approach (continuing) r e
Total Performance Management (TPM) i Total. complete plant system; over the full projected plant life L
optimization of the performance of the complete plant -
system to all vital ~ performance success indices (safety, i
economics, etc.).
include in design, full objective consideration of both i
deterministic and-probabilistic events and their costs-L l
- excellence keyed to plant system reliabilities as metrics of -
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quality attained in design, operations, maintenance, and i
Lmanagement.-
ACRS Joint Subcommittees'. Meeting AWS:10/15/89 -
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Translation to the objectives of safety, the " language" of containment and containment systems, and the definition of design and performance criteria w/r/t internal events
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l Retain the cardinal concept of multiple barriers to attain safety-in-1 s
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Define multiple reliability criteria as indices of successful
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performance for each of the multiple barriers to attain safety-thru-1 quality, e.g.,
i the reliability of a barrier to withstand successfully credible i
threats from credible internal initiators I
the reliability of the collective internal systems that credibly, thru l
failure and malfunction, could initiate a threat to the barrier l
e Define a total plant system reliability critenon as an mdex of j
successful performance of the composite containment function i
ACRS Joint Subcommittees' Meeting
' AWS:10/15/89
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D C0191ENTS ON CONTAINMENT DESIGN CRITERIA FOR FUTURE NUCLEAR PLANTS
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Nattendt Gitgineering.
Labetetyry
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F WORLD ENERGY PICTURE i
0 LARGE ENERGY! CONSUMPTION GROWTH PROJECTED BY WORLD i
ENERGY STUDIES i
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S LARGEST GROWTH IN AREAS WITH LOWER CURRENT PER. CAPITA l
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ENERGY CONSUMPTION THAN-IN THE UNITED STATES L
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GLOBAL ENVIRONMENTAL CONCERNS IF THIS GROWTH IS
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PROVIDED1BY FOSSIL-FUEL BURNING S
INDICATIONS THAT UNITED STATES, EUROPEAN AND JAPANESE NUCLEAR INDUSTRIES WILL SEEK. TO SERVE THIS GLOBAL l
ENERGY MARKET.
l CONCLUSION - WE MUST ADDRESS THE POSSIBILITY OF MUCH WIDER USE OF NUCLEAR ENERGY THAN IS EVIDENTLTODAYlAND IN A MUCH I
BROADER GEOGRAPHIC AND SOCIETAL SETTING s
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A RESULTANT FOUNDATION
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CONTAIIBIENT CRITERIA LINKED TO CLEAR PROTECTIVE I
-(REGULATORY) OBJECTIVES FORMULATED ON THE BASIS i
0F WIDE APPLICATION OF. NUCLEAR ENERGY WITHIN.THE
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UNITED-STATES AND IN THE WORLD AT LARGE i
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CONTAINMENT CRITERIA THAT ALLOW PROGRESSIVE DESIGN INNOVATION IN MEETING THE PROTECTIVE OBJECTIVES l
9 AN APPROACH BASED ON RISING STANDARDS OF ADEQUACY
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FROM DESIGN GENERATION TO DESIGN GENERATION l
e AN APPROACH AND RELATED METHODS THAT PROVIDE THE BASIS.FOR STRONG. SUPPORT OF. NUCLEAR ENERGY BY
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.LARGE NUMBERS OF PEOPLE l
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DEFINING THE APPROACH e
RELATED CONDITIONS
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SOUND ENGINEERING APPROACH BEST ESTIMATE, MECHANISTIC. ANALYSES i
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SUPPORTED 18Y ADEQUATE PHYSICAL UNDERSTANDING-i 1
" FACTORS-0F SAFETY" ADDED EXPLICITLY.
8 THIS.. APPROACH CAN BE UNDERSTANDABLE AND CONVINCING TO PEOPLE NOT: INVOLVED IN THE WORK: AND IS THEREFORE CONDUCIVE TO THE-GENERATION! 0F; SUPPORT-j-
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i DEFINING THE APPROACH
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4 RELATED CONDITI'ONS T
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THE NEW SYSTEMS SHOULD DEMONSTRATE ROBUSTNESS IN ACHIEVING THE CONTAINMENT. FUNCTION USE OF. BASIC PHYSICAL CHARACTERISTICS i
j-FAULT TOLERANCE
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j CAREFUL IMPLEMENTATION OF DEFENSE IN DEPTH WITN IISEPENDENT MULTIPLE LAYERS, EFFECTIVE FOR THE ENTIRE ACCIDENT SPECTRUM i
ABSENCE l0F TNE POSSIBILITY 0F BYPASS L
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. BALANCE BETWEEN PREVENTION AND MITIGATION (THERE-
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WILL ALWAYS BE : RESIDUAL UNCERTAINTY IN PREVENTION) i j
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DEFINING THE APPROACH
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RELATED CONDITIONS i
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- 0 POSSIBILITY 0F LONGER PLANT LIFETINES (80.TO 160' YEARS) i j.
ORIGINALLY REMOTE LOCATIONS MAY BECOME MORE POPULATED
~ i IT WILL NOT BE A SERVICE TO SOCIETY TO LIMIT LAND DEVELOPMENT j.
POSSIBILITIES 1
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WITH INCREASING " NUCLEAR FLEET" APPROACHES THAT ALLOW-i EVEN:THE REMOTE POSSIBILITY.OF FARMLAND WITHDRAWAL AND L
CLOSUREc0F NEIGHBORHOODS (CHERNOBYL) WILL BE. INCREASINGLY
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-UNACCEPTABLE.TO SOCIETY 4
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9 BOTX10F.THESE: FACTORS. MILITATE FOR AN APPROACH THAT CONCENTRATES 0NuTHE1 CHARACTERISTICS 0F THE: PLANT I
ITSELF ANDLDOESLNOT RELYiON EXTERNAL RESPONSES BY-1 THE: REST 0FlSOCIETY l
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FOUNDATION ELEMENT - RISING STANDARDS OF ADEQUACY 8
CONSISTENT WITH THE ADVANCED REACTOR POLICY STATEMENT l
8 LEVELS OF " ADVANCED DESIGNS" SHOULD BE RECOGNIZED AND APPROACHES DEFINED:ACCORDINGLY l'
DESIGNS THAT.ARE.A LOGICAL EVOLUTIONARY STEP FROM OPERATING LWRS -< BUILD FROM EXISTING RULES; DEMONSTRATE COWLIANCE WITH
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SEVERE l ACCIDENT POLICY; DEMONSTRATE-IWROVED-FISSION PRODUCT l-RETENTION;JCOUPLE WITN: FEATURES SUCN AS LONG TRANSIENT TIME; DESIGN TO TIGHTER PROTECTIVE OBJECTIVES-DESIGNS THAT REPRESENT A GREATER DEVELOPMENT STEP AND ARE n
AIMED AT LATER DEPLOYMENT - USE MORE PERFORMANCE RELATED i
- CRITERIA-TO ALLOW DESIGN INNOVATION; ESTABLISH EVEN TIGHTER l-PROTECTIVE OBJECTIVES 3
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I FOUNDATION ELEMENT - CONTAINMENT CRITERIA RELATED.
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NEAR TERM ADVANCED LIGHT WATER REACTORS q
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'l CORE DAMAGE FREQUENCY s'l X 10-5 PER-YEAR SITE BOUNDARY i-WHOLE BODY DOSE LESS THAN'25 REM FOR-ACCIDENTS WH0SE CUMULATIVE
- FREQUENCYEXCEEDS1X10.6LPERYEER s.-
8 LONGER TERM OBJECTIVES p
4 GO:BEYOND CONSIDERATION:0F N0'0FFSITE EMERGENCY PLANNING
- REQUIREMENT AND MAKE:THIS CONDITION:A SPECIFIC DESIGN OBJECTIVE 1
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i SHOULD THEJCONTAINMENT-DESIGN CRITERIA BE ALTERED?.
l RosERT E.. HENRY FAUSKE &-ASSOCIATES, INC.
L 16WO70 WEST 83no STREET' l-BURR RIDGE, ILLINOIS..
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CHICAGO, ILLINOIS l-OCrosER 17, 1989 i
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i BASIC CRITERIA FOR ACCIDENT CONDITIONS 1.
CONTAIN FISSION PRODUCTS RELEASED FROM-THE FUEL AND THE PRIMARY 4
SYSTEM (FIRST:AND-SECOND BARRIERS)..
2.
PASSIVELY CONTAIN (ACCOMMODATE)-THE i
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. ENERGY STORED.IN THE PRIMARY SYSTEM-COOLANT AND FUEL AT NORMAL OPERAT--
ING-CONDITIONS.
(LARGE LOCA ISLA WAY OF CONCEPTUALIZING THIS: RE-QUIREMENT.)
3.
REMOVE DECAY HEAT OVER THE LONG TERM.
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CRITERION:
CONTAIN THE ENERGY' L
STORED IN THE PRIMARY SYSTEM THE PREVIOUS CALCULATIONS-ARE ONLY APPROXIMATE TO-ILLUSTRATE THE SIZES NECESSARY TO SATISFY THEeCRITERION.
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OTHER ASPECTS NEED:TO-BE CONSIDERED, l
PARTICULARLY THOSE ASSOCIATED WITH l
NORMAL OPERATION.
s CONCLUSION - THIS-CRITERION FOR-CURRENT PLANTS:
- IS ENVELOPING.
- IS WELL CONCEIVED..
- SHOULD BE1RETAINEDLFOR. FUTURE-l L
PLANTS.
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CRITERION:
CONTAIN FISSION 4
PRODUCTS RELEASED FROM THE 4
FUEL AND PRIMARY SYSTEM L
FOR THE TWO CONCEPTS-USED IN THE-I U.S.,
THE CONTAINMENT COULD.POTEN-l TIALLY PRESSURIZE FOR-TENS OF
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MINUTES OR LONGER DURING A SEVERE j
ACCIDENT.
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j TO SATISFY THE CRITERION, THE CON TAINMENT MUST HAVE AN INTEGRAL STEEL-l LINER.
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l CRITERION:
CONTAIN FISSION PRODUCTS RELEASED FROM THE FUEL ONLTHE PRIMARY SYSTEM CONCLUSION:
WHILEETHE-VALUES'SHOWN-IN THE PREVIOUS SLIDE ARE AP-PROXIMATE, IT IS-CLEAR THAT THE CRITERION FOR CURRENT PLANTS IS:-
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- WELL CONCEIVED, AND L
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.SHOULD BE RETAINED FOR FUTURE' PLANTS.
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OTHER LESSONS FROM REACTOR ACCIDENTS THE TMI ACCIDENT WAS CAUSED:BY A LACK OF WATER.
THE TMI ACCIDENT WAS TERMINATED-BY ADDING WATER.
THE DAMAGED CHERNOBYL REACTOR-WAS.
STABILIZED FOR SEVERAL HOURS BY l
WATER ADDITION-(FIRE FIGHTERS)
BUT l
WAS HAULTED BECAUSE THE WATERcWAS h
SPILLING INTO,AND CONTAMINATING UNITS 3, 2 AND 1.
L CONCLUSION:
WATER'WOULD BE VERY EFFECTIVE IN-RECOVERING FROM AN-ACCIDENT STATE AND FUTURE DESIGNS, LIKE THE CURRENT PLANTS ~,
SHOULD FOCUS ON WAYS T0-SUPPLY WATER-TO THE CONTAINMENT AND TO REMOVE'THE DECAY.
HEAT.
1 J
i CONTAINMENT LINER
. INTEGRITY IS IMPORTANT Furuns DESIGNS SHouLD Focus ON-i 4
l COOLING THE DEBRIS TO PROTECT:THE LINER.
l-l IMBEDDING THE-' LINER IN'-CONCRETE TO I
MINIMIZE THERMAL LOADS FROM DEBRIS.
i OR BOTH.
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t FUTURE DESIGNS CAN-ADDRESS SEVERE-ACCIDENT ISSUES l
LIKE THE CURRENT PLANTS, FUTURE l
DESIGNS SHOULD PROTECT AGAINST
'i OVERPRESSURE DUE TO HYDROGEN COMBUS-l TION.
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--VOLUME AND ULTIMATE PRESSURE i
CAPABILITY TO ACCOMMODATE A COM-i PLETE BURN OF HYDROGEN-GENERATED BY THE OXIDATION OF 75% OF THE ACTIVE CLADDING.
i INERT THE CONTAINMENT.
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l INTENTIONAL IGNITION (IGNITERS).
PROVIDING PROTECTION FOR THEHLINER.
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- WATER.
l IMBEDDED.
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.s-1 FUTURE DESIGNS CAN ADDRESS SEVERE ACCIDENT ISSUES BY (CONTINUED)
USE A REACTOR CAVITY / INSTRUMENT l
TUNNEL CONFIGURATION WHICH DRASTI-CALLY-REDUCES OR ELIMINATES THE P'OTENTIAL FOR DEBRIS-DISPERSAL GIVEN' A.HIGH PRESSURE MELT EJECTION CONDI-TION.
MAXIMIZE THE' CAPABILITY-OF PUTTING WATER ON:THE CONTAINMENT FLOOR.
.. - MAXIMIZE THE POTENTIAL FOR ACCIDENT RECOVERY BY MAXIMIZING THE-FLOOR AREA-FOR DEBRIS ACCUMULATION.
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1 CONCLUSIONS i
THE GENERAL CRITERIA USED FOR DESIGNING THE-CURRENT PLANTS ARE WELL CONCEIVED.
THE PRUDENCE'0F THE CRITERIA USED IN THE U.S.-IS DEMONSTRATED BY THE.-
EXPERIENCE FROM REACTOR ACCIDENTS.
. THE GENERAL CRITERIA USED FOR CUR-RENT PLANTS ARE APPLICABLE TO FOTURE DESIGNS.
i
. THE IMPLEMENTATION OF THE CRITERIA CAN BE STREAMLINED.
FUTURE DESIGNS COULD ADDRESS SEVERE-F I
ACCIDENT ISSUES TO REDUCE THE IN--
FLUENCE OF UNCERTAINTIES.
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1 i
i CONTAINMENT DISCUSSlQN Presented to the NRC ACRS Subcommittees on Coritainment and Structures--Chicago, Illinois, October 17,1989 Prepared by M. Bender, Querytech Associates, Inc..
O DEFINITION OF CONTAINMENT,- A SYSTEMS CONCEPT 0
REFERENCE EXPERIENCE O
CURRENT UNDERSTANDING FROM NRC! AND INDUSTRY SPONSORED RESEARCH 0
DEVELOPING A DESIGN BASIS e
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j CONTAINMENT DEFINED:
-_A SYSTEM INTENDED TO PREVENT THE SPREAD OF-RADIONU-CLIDES, RELEASED IN BULK FROM THE REACTOR CORE, BEYOND-SPECIFIED SITE LIMITS IN THE EVENT OF A NUCLEAR ACCIDENT.
ESSENTIAL SYSTEM PROPERTIES:
1 Boundary closure sufficient to limit: dispersal: of radionuclides postulated.to be present during Land-subsequent to an accident, 2. An effective heat sink to absorb nuclide decay l energy' and stored energy in coolants and surrounding structure for the purpose of controlling temperature conditions to limit subsequent chemical, physical state, or fluid perturbations that would aggravate radionuclide dispersal j
conditions, i
3 Radionuclide trapping or stabilizing capability to prevent
~
further dispersal of all but the noble gases during and subsequent to.an. accident including those caused by transient effects. (Holdup _ to permit noble gas (xenon)-
decay-can be a valuable capability, but the trapping-mechanisms must.be of high reliability; the physical flow path may be the most effective device for this purpose.)
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REFERENCE REACTOR ACCIDENT EXPERIENCE L
i-1.
NO RADIONUCLIDE RELEASES AT HIGH POWER-2.
OPERATOR ALERTNESS HAS PREVENTED ' FUEL FAILURE 'AT POWER (E.G. BROWNS FERRY ATWS, DAVIS BESSE FEEDWATER-TRANSIENT)
I 3.
PREVIOUS PRACTICE HAS EXCLUDED SEVERE EVENTS FROM CONTAINMENT REQUIREMENTS (BWR ATWS, CORE COOL. ANT BLOCKAGE) j 4.
EARLY ACCIDENTS IN SMALLER INSTALLATIONS HAVE. GUIDED SAFETY REQUIREMENTS (SL-1, NR-X, WINDSCALE)'
l 5.
TARAPUR AND CHERNOBYL SHOWN POTENTIAL RISK (NOT AS l
EXTENSIVE AS " DOOMSDAY" PREDICTIONS BUT EXTENSIVE AND SERIOUS) j 6.
TMI-2 SHOWED THAT CORE MELTING DOES NOT NECESSARILY.
}
VIOLATE CONTAINMENT.
.WITH MINIMAL COOUNG UNDER j
SHUTDOWN ' CONDITIONS LOW CONTAINMENT PRESSURES l
EASILY MAINTAINED. LOW LEAKAGE WASNT HARMFUL i.
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. 1 WHAT ARE THE LESSONS FROM AOCIDENT RESEARCH?
d 1.
ACCIDENT PROGRESSION 1.1
- Murphy's Law" logic does not give effective design guidance.
1.2. Unencumbered accident progression will inevitably lead to
~
imponderable accident conclusion.
1.3 Time is available for control accident interdiction.
1.4 The operator is an impor1 ant. part of accident control and operator interdictive provisions should not involve co'mplex logic based on accident progression analysis.
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LESSONS FROM SEVERE ACCIDENT RESEARCH l
2.
CONTAINMENT STRUCTURAL RESPONSE 2.1 Containment structural behavior is predictable and reliable uo to elastic resoonse limits.
Reinforced concrete appears to-provide non-catastrophic failure.caoability beyond elastic resoonse limits.
2.2 Liner reliability contingent - on assuring' controlled; structural _
^
movement under accident loadings--discontinuities still the major uncertainty in liner response..
2.3 Closures sealed with elastomers -are the main; source of leakage vulnerability. Experimental testing suggests that up to the point of significant leakage (observable flow) gasket materials in current use are. functionally effective over the t
anticioated times of active accident orooression if orotected from overtemoerature and intense radiation.
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11; DESIGN BASIS i
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" DESIGN BASIS ACCIDENT" DEFINITION 1.-
ACCIDENT INITIATORS NEED. TO BE POSTULATED-LOCA'S,'
LOPA'S, STEAM GENERATOR RUPTURES, ETC.
2.
SEVERITY OF THE CONDITION NEEDS BETTER RATIONALE I.E.
WORST CONDITION LOCA'S DISTORT BEHAVIORAL -
CHARACTERISTICS AND MISUSE SAFETY RESOURCESiEXAMINE SYSTEM PROPERTIES FOR~A REALISTIC ACCIDENT BASIS..
i 3.
ATWS TYPE EVENTS NEED TO BE INCLUDED IN SOME FORM.
ENOUGH EXAMPLES EXIST TO DEFEAT ANY PROBABILISTIC-ARGUMENT THAT THEY ARE~ OUT OF THE REALM OF PROBABluTY.
4.
RADIONUCUDE RELEASES SHOULD BE BASED ON REAL TIME EVENTS--ARBITRARY RELEASES
~DO NOT.
PROPERLY ~
CHARACTERIZE THE ACCIDENTS : AND DO NOT' EFFECTIVELY.
l COMBINE RELATED CIRCUMSTANCES.
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12 DESIGN BASIS
- DESIGN BASIS ACCIDENT
- DEFINITION 5.
ACCIDENTS SHOULD NOT BE ASSUMED TO PROCEED TO THEIR NATURAL ENDPOINT UNLESS THE INTERDICTIVE OPPORTUNITIES ARE BEYOND ACCESS.
AN ATWS MIGHT NOT BE CONTROLLABLE; A SMALL LOCA HEAT SINK BYPASS COULD BE-CORRECTED IF KNOWN TO EXIST. ACCIDENT SENSING NEEDS TO BE BUILT IN TO THE DBA ASSESSMENT, 3
6.
DESIGN CONTAINMENT ENCLOSURE FOR CONTROLLED FAILURE:
ALLOW CONDITIONS NEAR TO STRUCTURAL YlELDING AND PROVIDE RUPTURE RELIEF THROUGH A KNOWN TRAPPING PATH BEFORE BURSTING.
7.
PROVIDE FOR EFFICIENT TRAPPING MEDIA SUCH AS CAUSTIC SPRAYS, CHEMICALLY ACTIVE TRAPPING PONDS, RUGGED AND ACCIDENT INSENSITIVE TRAPPING DEVICES LIKE " SAND FILTERS",
)
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i SELF-ACTUATED PRESSURE RELIEF DEVICE FOR l
REACTOR CONTAINMENTS (CONCEIVED BY L. MINNICK; j
INVESTIGATED FOR EPRI BY S. LEVY, INC.)
PATENT APPLIED FOR BY EPRI I
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FUNDAMENTAL PURPOSE l
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5 TO PREVENT OVER-PRESSURIZATION OF REACTOR l
CONTAINMENT DURING ANY POSTULATED ACCIDENT i
OTHER THAN INSTANTANEOUS RELEASE OF ENERGY l
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1 SELF ACTUATED PRESSURE RELIEF DEVICE FOR REACTOR CONTAINMENTS ADDITIONAL FUNCTIONS PERFORMED i
SCRUBS RELEASED GASES OF PARTICULATES AND ANY MATERIAL j
HAVING AN AFFINITY FOR WATER.
l PROVIDES DILUTED, ELEVATED AND HEATED RELEASE OF NOBLE l
GASES.
i CONDENSES ESSENTIALLY ALL STEAM AND RETURNS THE WATER FORMED TO THE CONTAINMENT, l
REESTABLISHES CONTAINMENT' INTEGRITY WHENEVER-CONTAINMENT OVER PRESSURE IS TERMINATED.
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PROVIDES RELIEF OF POTENTIAL CONTAINMENT VACUUM i
FOLLOWING INCIDENT,
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i SELF ACTUATED PRESSURE RELIEF DEVICE 1
FOR 4 -
REACTOR CONTAINMENTS INHERENT CHARACTERISTICS i
TOTALLY PASSIVE ACTUATION, OPERATION AND RESET:
1 i
- NO ACTIVE DEVICE OR MECHANISM,
- NO OPERATOR ACTION, f
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- NO POWER REQUIREMENT, 3
- NO INSTRUMENTATION OR CONTROL, AND
- NO MAKEUP WATER ARE REQUIRED THROUGHOUT THE COURSE OF THE TRANSIENT, REGARDLESS OF DURATION.
SHIELDS ALL RADIOACTIVE MATERIAL COLLECTED AND, ULTIMATELY, CONTAINS WHATEVER HAS NOT BEEN RETURNED TO
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THE CONTAINMENT IN A SINGLE UNDERGROUND TANK.'
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Review of Present Structural Containment Design
. The LOCA load is well defined. The NSS supplier provides this load. It is coupled to the reactor's l
thermal capability.
i
- The ASME Containment Codes are complete.
i They are:
l Section 111 - Division 1 - Subsection MC, Section lli - Division 2 - Subsection CC, l
and have been developed and are maintained by the Industry, Research, and Universities with participation i
by the NRC. These codes are based on LOCA loads.
4 4
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' C1923.005 10-16-89 I
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Review of Present Structural Containment Design i
. The containment capability of existing containments i
for an upper bound pressure load have been determined and safety margins compared to LOCA loads have been computed. The acceptance criteria in all these capability evaluations were beyond code allowables.
- Based on these studies, containments designed to current codes show considerable margins.
}
- Some of these results used in PRA have shown acceptable risk to public within current understanding of acceptable risk.
l
- Testing by Sandia of scaled containment models in steel and reinforced concrete have shown that in most cases, the scaled containments behave in a ductile manner.
C1923.006 10-16-89 SasCENTS LENBY
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Review of Present Structural Containment Design l-l
- The work required to determine the containment capabilities was sponsored by:
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The Industry Degraded Core Rulemaking Program, l
Utilities commissioning plant unique probabilistic I
l risk assessment studies, i
i Sandia-NRC sponsored workshops, Sandia effort on NUREG 1150.
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. The Advanced Ught Water Reactor Study utilizes a l
containment designed for LOCA loads and using the ASME Code. System and layout provisions are made in consideration of-severe accidents.
l C1923.007 10-16-89
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Review of Present Structural Containment Design i
l
- Lessons learned from the Containment Capability Studies have highlighted that the containments must be ductile 1
l and must not have a weak link anywhere. Designs and care of details is of utmost importance and can be provided l
within current design basis.
Conclusion
. The present Structural Containment Design Criteria is i
adequate and should not be changed in the near future.
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l c19n006 1416-89 S
GENTSLW57 n
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Recommendations for Future Development
~
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. It is recommended that an industry effort, in participation j
with research, universities and the NRC, should be a
4 i.
undertaken to develop loads and design criteria for 1
containment based on severe accidents.
. The goals of this effort should be: Define severe accident loads in terms and ways that can be utilized in structural l
design without ambiguity.
1
. A consensus has to be reached regarding.the events involved in a. severe accident.. Loads, in terms of time-L dependent pressures and temperatures and their probability of occurence have to be established.
l
. A consensus-has to be reached regarding an acceptable i
L probability of risk to the public in case of a severe accident.
f C1923.003 10-16-89 sansseraunst 1
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Recommendations for Future Development
~
- Future structural designs will be based on probabilistic l
assessement of loads and resistance to achieve a safe j
structure. When this can be done appropriately, it is then the proper time to change the containment design basis.
l
. Revise present ASME design codes from deterministic to probabilistic in terms of load factors and allowables, and emphasize ductility.
. Based on the present work of the' Advanced Light Water l
Reactor Industry Group, future containments may have only l
one of two configurations: the large dry containment for PWRs l
and a modified Mark 11. containment for the BWRs.
Limiting.
i consideration to these possibilities will facilitate the above l
tasks considerably.
l l
. It is anticipated that such efforts will require a considerable amount of time.
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THOUGHTS AND REFLECTIONS ON L
L CONTAINMENT DESIGN CRITERIA l
l I
W. A. von RIESEMANN CONTAINMENT TECHNOLOGY DIVISION l
SANDIA NATIONAL LABORATORIES i
l PRESENTATIONTO ACRS JOINT SUBCOMMITTEE MEETING l
CONTAINMENT SYSTEMS / STRUCTURAL ENGINEERING c
I OCTOBER 17,1989
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SUMMARY
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A DECADE OF KNOWLEDGE ON CONTAINMENT BEHAVIOR AND l-L SEVERE ACCIDENTS HAS NOT BEEN FACTORED INTO THE ASME CODE c
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RECOMMEND THAT A COMMITTEE (INDUSTRY, RESEARCHERS,
)
(
L REGULATORS) BE FORMED TO REWRITE THE CODE (DESIGN, FABRICATION, INSPECTION INCLUDING LEAK RATE MEASUREMENTS, SEVERE ACCIDENTS) CONSIDERING THE i
CONTAINMENT AS A SYSTEM i
FIRST STEP WOULD BE TO DETERMINE THE PHILOSOPHY j
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-l CONTAINMENT (cont'd)
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CONTAINMENT IS A SYSTEM--NOT ~AN ISOLATED COMPONENT (SHELL)-
l 1.E. SYSTEM CONSISTS OF
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Structure (She:i)
Penetrations (Operable and Fixed) l Bellows l-Drywell Head.(BWR)
L FuelTransferTubes
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isolation Valves Basemat l
Instrumentation (Status of System)
THE PERFORMANCE (BEHAVIOR) DEPENDS ON i-THE RESPONSE OF ALL OF THE PARTS AND ANY POSSIBLE INTERACTIONS; e.g., REACTOR VESSEL j
j
-SUPPORT FAILURE WHICH THEN WILL LOAD j.
CONTAINMENT THROUGH THE STEAM LINES.
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i LESSONS LEARNED i
q Current Design Personnel Airlocks and E!ectrical Penetration i
Assemblies (Except for Electrical Peformance) Behaved l
Well(Leakage and Strength) i l
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Equipment Hatches j
Sleeve Ovalizes-Leakage May Occur Pressure Unseating-Not Desirable r
i Seals and Gaskets - Performed Well Up to About 5000F 1
i Inflatable Seals - Leakage will Occur at Overpressurization Basemats - Data from a Recent Test Result has to be Interpreted; i
Additional Work may have to be Performed.
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LESSONS LEARNED (cont'd)
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l Stiffening Around Penetrations and ' Area Replacement' Rule i
Causes Strain Risers and May Lead to Early Failure In Particular, for Liners With Studs and j
(on Ring Stiffened) Steel Cylinders Basemat - Cylinder intersection in Reinforced Concrete Containments is Overdesigned Tori-spherical Heads do Buckle but do not Fall (i.e. Leak) till the Pressure is Several Times the Buckling Pressure F
Consequences of a Core / Concrete Interaction Depend on the Chemical Composition of Concrete 4
I.
i
e LESSONS LEARNED (cont'd)
Substantial Corrosion of the Steel (Where it Enters the Concrete) May Occur Aerosol Retention in Concrete has not been Quantified Retention in Secondary Buildings has not been Quantified Containments have had Isolation Valves Left Open for Extended Periods il
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i GOALS FOR THE NEW REQUIREMENTS i.
Benign failure modes i
Long Life i
Simple Inwiicn, including On-Line Monitoring i
Construction Ease
~
Designers must become Familiar with Severe i-Accidents and Loads Beyond the Design Basis and the i
Fact that some Loads are not well Defined; i.e., Mind j
Set must Change i
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L GOALS FOR THE NEW REQUIREMENTS (cont'd) i Internal Structure (Compartments, Rooms) should be Designed to Minimize Effects of Fire, Flooding and Hydrogen Combustion.
l Realistic Leakage Requirements Realization that Buckling, per se, is not Necessarily j
Failure L
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POTENTIAL DIFFICULT POINTS ~
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1 Definition of L.oads 1
Design Criteria vs. Performance Requirements s
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Overpressure Protection
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Leak Rate Testing I
Current Licensing is done on a Prescriptive L
Basis-Difficult to Accommodate Guidelines 1
Probabilistic Design Beyond the Current j
State-of-the-Art
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