ML20006F494
| ML20006F494 | |
| Person / Time | |
|---|---|
| Issue date: | 12/27/1989 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2683, NUDOCS 9002280090 | |
| Download: ML20006F494 (49) | |
Text
y ff o?bD
- 6
- c t
M/ M/kd S
55.185lEy PS' r
,.c e
P
. "3 6; j
Date Issued: December 27,1989 If,/.
0
,5 h
0 y$d;%
hl 1,
4 J. O r
0 f7 O ACRS JOINT SUBCOMMITTEE MEETING
SUMMARY
/ MINUTES I 8 f0 FOR. CONTAINMENT SYSTEMS / STRUCTURAL ENGINEERING DECEMBER 13, 1989 BETHESDA, MARYLAND PURPOSE The ACRS ' Subcommittees on Containment Systems and Structural Engineering held a joint meeting on December 13, 1989 in Bethesda, Maryland.
The purpose of this meeting was to continue the L
discussion in regard to the development of an ACRS paper on containment design criteria for future plants based on present knowledga.
A copy of the meeting agenda and selected slides from the presentations are attached.
The meeting began at 8:30-a.m.,
and adjourned at 5:30 p.m.,
and was held entirely in open session.
The principal attendees were as follows:
ATTENDEES ACRS INVITED SPEAKERS D. Ward, Co-Chairman G.
Davis, CE C.
Siess, Co-Chairman W.
Fox, Duke Power Co.
J.
Carroll, Member G.
Gyorey, GE l
W.
Kerr, Member R. Lutz, H
'C.
Michelson, Member B. McIntyre, H C.
Wylie, Member T.
Pratt, BNL y
M.
Corradini, Consultant D.
Houston, Staff NRC/RES E.
Igne, Staff B. Hardin 1
'EEVIEW DOCUMENTS 1'
There were-no formal documents -to be reviewed at this meeting.
The ACRS effort on this subject is in response to a Staff Requirements i
Memorandum dated July 28, 1988, which was written following an ACRS meeting with the Commission on July 14, 1988.
0\\
9 B
Dd 90022goe,22 U
tc c-- u
<'Ve'
y a-l o
pg
. ACTIONS. AGREEMENTS, AND COMMITMENTS E
l'. -
B. Hardin '(RES) agreed to provide a copy of the following documents as soon as feasible:
o, a.
Draf t Regulatory -' Guide, " Regulatory Guidance on Technical Issues, Form, and Content of PRAs Performed for Advanced Designs."
k b.
Draft supporting document for Regulatory
- Guide,
" Technical Basis for Functional
.k Performance Requirements for Evolutionary LWRs."
t 2.
T.
Pratt (BNL) agreed to provide a list of technical reports that have been prepared by BNL in regard to severe accident analyses and used by the NRC staff in their review of evolutionary LWRs.
Discussion f'
In his opening comments, D. Ward indicated that this was the third of a series. of meetings to gather information and opinions in p
regard to containment design criteria for future plants.
He noted that B. Spencer (ANL) had regretfully cancelled his appearance the
~
previous day but that he intended to provide some written comments.
l Mr. Ward acknowledged the presence of Dr.
T.
Rogers from the Advisory Committee on Nuclear Safety (Canada) and invited him to
(
participate informally in the meeting, as appropriate.
l B.
Hardin (RES) discussed the NRC/RES effort in preparing a f
proposed plan for implementation of the severe accident policy for
-future plants.
He indicated that SECY-88-248, dated September 6, 1988, addressed this matter.
Rulemaking hearings would have been initiated to amend 10 CFR 50.34 to require technical information
y<
s
.s.
a 2
...e-on severe accidents be included in future plant applications.
In December _1988, the Commission expressed a concern about the timing of such hearings and the potential effect that this could have on the design certification process for current evolutionary LWRs.
Thus, the staff was instructed to drop the rulemaking hearings.
The staff has pursued the review of severe accident issues in regard to specific applications - EPRI ALWR, GE ABWR, CE Systems 80+,
and W SP/90.
-In this respect, the staff has asked their consultant at BNL to assist in the design reviews and prepare the proposed regulatory guide and supporting documentation noted above (see Actions, Agreements, and Commitments).
Mr.
Hardin also alluded to draft SERs for the EPRI ALWR (SECY-89-228) AND GE ABWR (SECY-A9-153) which were said to demonstrate the consistency of the staff's review in this area.
Dr. Kerr expressed a concern that significant severe accident policy was being established via the proposed regulatory guide and staff SERs and not spelled out in a specific-policy document or even discussed with the Commission.
He and other subcommittee members believed that this bottom-up approach to establish NRC
_ policy was inappropriate.
Such an approach would fail to provide reactor designers or the staff reviewers with much guidance on this matter.
Mr. Hardin agreed with the criticism but indicated that in the absence of rulemaking this was the only path open for the staff to proceed.
He agreed to provide copies of the proposed regulatory guide and supporting document to the ACRS as soon as feasible.
Mr. Hardin reviewed portions of 10 CFR 52 which were concerned with TMI-2 requirements (containment design pressure at 45 psig) and
- with a design specific PRA.
He indicated that the staf f review was focused only on evolutionary LWRs.
In response to Dr. Siess, he stated that the evolutionary LWRs are being reviewed to current regulatory criteria with severe accident issues superimposed on them.
i j
w g.'
y
)
1; 3
Dr.
T. Pratt (BNL) discussed the technical basis for functional performance requirements for evolutionary LWRs, an ef fort performed 4
as a consultant to the NRC staff.
The document on technical basis was intended to provide guidance on which severe accident-l vulnerabilities needed to be considered and what reasonable measures should be taken to address these issues.
The reasonable measures are intended to demonstrate compliance with the following NRC staff supplied requirements:
(a) core damage frequency below 1E-05 per reactor year, (b) frequency of a large release (early
(
failure) below 1E-06 per reactor year, and (c) assurance that the containment can effectively mitigate a range of severe accidents.
In this respect, BNL derived performance requirements for key safety systems:
reactivity control, coolant inventory control, and decay heat removal.
Dr. Pratt discussed in some detail the expected severe accident challenges to the containment as L
identified in various PRA studies and research programs.
He addressed briefly some uncertainties associated with-containment performance and then discussed the challenges to specific LWR containment types.
- Finally, he discussed the development of l
performance requirements for five identified challenges:
hydrogen combustion, high pressure core meltdown, containment bypass, core debris / containment boundary interactions, and long-term decay heat l
removal.
For each of these, he presented an aim / goal, options, key considerations, -and implications for evolutionary LWRs.
In l
summary, Dr. Pratt indicated that these performance requirements
. provide reasonable assurance that evolutionary LWR containments will have a high probability of remaining intact during a core melt accident.
He further noted that these requirements are similar to those recommended in the EPRI ALWR document.
In response to questions concerning the staff supplied requirements, B. Hardin (RES) indicated that these came from the draft implementation plan for the Safety Goal Policy.
One not in this plan but found in the GE ABWR licensing agreement is the
1 y
L. 0,-
e.-
.z i
4 conditional containment failure probability of 0.1.
Mr. Michelson noted that BNL had supplied a number of technical reports to the staff in support of SER reviews.
He requested a
- listing of those reports and Dr. Pratt said he would provide it.
In response to a question by Mr. Carroll, Dr. Pratt indicated that' they define core damage in the same sense as found in the latest r
draft of NUREG-1150.
He.also presented the reasons for supporting a metal water reaction value of 100 percent as opposed to that-I recommended by EPRI of 75 percent.
Mr.
R. Lutz (W) reviewed the current requirements for containment designs and concepts. He indicated that large dry PWR containments designed to current criteria generally have a largo margin for accommodating severe accidents.
He discussed the large dry performance in regard to short-term and long-term accident phenomena.
He stated that containment integrity could be maintained in the long term if"some path for heat removal remained functional.
Even without heat removal, the containment was predicted to remain integral for times greater than one day.
Mr.
Lutz identified a few problem areas in current regulations.that would prohibit increasing the level of safety afforded by the g
L containment, such as, water sprays to the shell exterior.
In L.
summary, he encouraged the acceptance of EPRI design guidance and indicated that a PRA must be used for the design of the containment and containment systems.
He did not believe that direct containment heating (DCH) was a significant issue in their plants due to depressurization schemes and he did not recommend vents for large drys.
Dr. G. Gyorey (GE) presented his perspective on containment design l
from the point of view of the LMFBR program.
He reviewed the current containment situation for operating plants and the safety basis for the LMFBR program.
For the
- LMFBR, the stated
,g
'O _:
5 requ'irements would most likely lead to a system that met the safety goals on the basis of just prevention considerations; mitigation.
would thus-provide margin. He discussed radiation release criteria and indicated that_10 CFR 100 plus PAG (36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />) considerations would be more restrictive than the safety Goal by two orders of
' magnitude.
He gave details of the passive systems which were intended to achieve high reliability.
These included decay heat
- removal, reactivity
- control, and containment function.
He
' indicated that GE had not been able to - identify a way that the operator in the control room could damage the core.
In summary,-
he presented five general and five specific recommendations for containment criteria for future plants.
The general recommendations were:
1..
Emphasize containment function rather than form.
2.
Allow for systems very different than current LWRs.
3.
Avoid closing out innovative approaches.
4.
Encourage passive features.
5.
Recognize interdependence among containment, decay heat removal, and reactivity control functions; and possible tradeoffs.
Mr.
G.
Davis (CE) discussed the Systems 80+ containment design bases:
the approach, codes and standards, design conditions, loading categories, acceptance criteria, and conservatisms.
The following are two conservatisms given by Mr. Davis:
(1) ultimate pressure capacity of the containment is about four times greater than design pressure limits, and (2) current source terms are conservative by orders of magnitude. He next discussed four severe
_ _ ~.-
m 6
[
accident issues and indicated the containment capability in regard to them.
These issues were:
(1) combustible gas control, (2) core debris. coolability, (3) direct containment
- heating, and (4) containment venting.
Mr.
W.
Fox (Duke Power) described the large steel spherical containment for the CE Systems 80+
plant.
It was a dual
. containment concept with a diameter of 200 feet.
The steel shell was 1.75 inches thick and had a design pressure of 49 psig.
In s
response to a question by Dr. Siess, Mr. Fox indicated that this shell thickness was selected to be exempted in the code for stress relieving.
He presented vieWF. of a number of sections of the containment which showed the plant configuration, the location of
-t water sources, the design of the core debris chamber to minimize DCH, and the post-accident ventilation paths.
Mr. Davis continued the CE discussion with a description of the safety depressurization and vent system.
He recommended the use of best estimate deterministic models (MAPP) for severe' accident analysis and the use of.EPRI recommendations.
In conclusion, he stated.that the continued use of traditional design bases is appropriate.
In closing, Mr. Ward asked the subcommittee members and consultant on the strategy to pull together their thoughts on' containment design criteria for future plants.
Dr. Siess asked if this should be restricted to LWRs.
Mr. Ward indicated that it should not if criteria could be developed general enough to apply universally.
Dr. Siess-indicated that if they were general, then they would be more in terms of performance criteria rather than design.
If design criteria, he felt they should be restricted to LWRs.
Dr.
Kerr indicated an uneasiness about any criteria without severe accident considerations.
Mr. Ward expressed a desire to have a policy on how containments should be designed.
He will attempt to synthesize the information and opinions and prepare something for
r.
-6
.u-...
r i
7 I
discussion at~a future joint subcommittee meeting.
Subsequent to this meeting, a joint subcommittee meeting was tentatively scheduled for February 28,-1990.
NOTE:
Additional meeting details can be obtained from a transcript of this meeting'available in the NRC Public Document Room, 2120 L Street, N.W.,
Washington, D.C., 20006, (202) 634-3273, or can be purchased from Ann Riley and Associates, Ltd.,
1612 K
- Street, N.W.,
Suite
- 300, Washington, D.C.,
20006, (202) 293-3950.
-5 l
1 i
>L
,3 +.. -
w.
ACRS JOINT SUBCOMMITTEE MEETING CONTAINMENT SYSTEMS / STRUCTURAL ENGINEERING'-
DECEMBER 13,'1989 BETHESDA, MARYLAND
- TENTATIVE AGENDA -
CONTAINMENT DESIGN CRITERIA FOR FUTURE NUCLEAR PLANTS A.
Subcommittee Chairmen Remarks D. Ward /
8:30 aan.
I C. Siesa, ACRS INVITED SPEAKERS B.
Brad Hardin, RES 8:45 a.m.
C.
Trevor Pratt, BNL 9:30 a.m.
BREAK 10:15-10:30 a.m.
D. Trevor Pratt (Continued) 10:30 a.m.
LUNCH J2:00 - 1:00 p.m.
E.
Bob Lutz,ji
-1:00 p.m..
F. Geza Gyorey, GE 2:00 p.m.
G. George Davis, CE - Bill Fox, Duke Power 3:00 p.m.
H. Subcommittee Discussion 4:00 p.m..
I.' Adjournment 5:00 p.m.
~
7 ;,7 7.-
E CONTAINMENT DESIGN CRITERIA FOR FUTURE NUCLEAR PLANTS o
I BACKGROUND INFORMATION ON
- m THE NRC 0FFICE OF RESEARCH'S i
1 ONGOING PROGRAM FOR IMPLEMENTATION OF SEVERE ACCIDENT POLICY L
i
/'
s BRAD HARDIN L
L ADVANCED REACTORS AND GENERIC' ISSUES BRANCH OFFICE OF RESEARCH l
l:
1 If PRESENTATION TO ACRS JOINT SUBCOMMITTEE MEETING CONTAINMENT SYSTEMS / STRUCTURAL ENGINEERING DECEMBER 13, 1989 l:
BETHESDA, MARYLAND L
- 1. - - e i
HISTORICAL BACKGROUND t
l
- CRITERIA FOR CONTAINMENT DESIGN FOR FUTURE PLANTS WAS PART OF THE PREVIOUS STAFF STUDIES INTO THE I
IMPLEMENTATION OF SEVERE ACCIDENT POLICY'THAT WERE PERFORMED IN -THE REGULATORY IMPROVEMENTS BRANCH OF NRR.
THIS WORK WAS DOCUMENTED IN SECY-86-76.
L
- RESPONSIBILITY FOR PREPARING A PROPOSED PLAN FOR IMPLEMENTATION OF SEVERE ACCIDENT POLICY FOR FUTURE PLANTS IS CURRENTLY ASSIGNED TO THE ADVANCED REACTORS AND GENERIC ISSUES BRANCH (ARGIB) IN THE OFFICE OF RESEARCH.
- ARGIB HAS BEEN ACTIVELY WORKING IN THIS AREA FOR THE PAST-2 YEARS WITH BNL AS 'ITS PRINCIPAL CONTRACTOR.
i PRODUCTS FROM THIS WORK INCLUDE:
- 1. SECY-88-248, " IMPLEMENTATION OF THE SEVERE ACCIDENT POLICY FOR FUTURE LIGHT WATER-REACTORS"
- 2. DRAFT REGULATORY Gu1DE, " REGULATORY GUIDANCE ON TECHNICAL ISSUES, FORM, AND CONTENT OF PRAS y
PERFORMED FOR ADVANCED DESIGNS"
- 3. DRAFT SUPPORTING DOCUMENT, " TECHNICAL BASIS FOR FUNCTIONAL PERFORMANCE REQUIREMENTS FOR EVOLUTIONARY LWRS"
4 x, ;.
(,
e I
o l-
. SAMPLE 0F PERFORMANCE-ORIENTED SEVERE ACCIDENT l
CRITERIA FOR CONTAINMENTS IN. FUTURE :(NEAR-1 TERM) LWRS q
L p
1 (TAKEN FROM-ENCLOSURE-3 0F SECY-88-248) j CONTAINMENTS SHALL BE
- DESIGNED,
. MAINTAINED AND OPERATED WITH SUFFICIENT MARGIN-TO PROVIDETREASONABLE l
-ASSURANCE THAT-IN THE EVENT.OF A SEVERE CORE DAMAGE-
. EVENT u AND THE ~ LIKELY CONSEQUENTIAL -PHENOMENA :(E.G.,
MOLTEN CORE DISPERSAL ?FROM 'THE VESSEL AND CONTAINMENT
. PRESSURIZATION), ' SUFFICIENT RETENTION. 0F FISSION PRODUCTS WOULD. BE MAINTAINED.
THIS REQUIREMENT IS-m
' APPLICABLE ONLY TO. THOSE. SEVERE ACCIDENT EVENTS ~-
CONSIDERED TO BE SIGNIFICANT POTENTIAL CONTRIBUTORS.TO RISK ~TO-THE PUBLIC HEALTH AND SAFETY.
' t I
=
[ ; 'l :
p 2
9
-IN SPITE OF COMPLEXITY OF SUBJECT 0F CONTAINMENT
-DESIGN CRITERIA'FOR FUTURE PLANTS AND THE 1
L
. CORRESPONDING DIVERSITY OF VIEWS, SOME THEMES J
APPEAR T0.BE COMMON AND UNCHANGING:
- 1. THERE IS AN EXPRESSED NEED FOR GUIDANCE IN THIS AREA FOR USE BY BOTH THE INDUSTRY AND-THE NRC ~
REVIEWERS.
THIS NEED HAS BEEN EXPRESSED BY BOTH-PARTIES.
EXISTING
- GUIDANCE, INCLUDING THE SEVERE L
ACCIDENT POLICY STATEMENT, IS NOT SUFFICIENT.-
q L
- 2. THERE IS A DESIRE TO MAINTAIN DEFENSE-IN-DEPTH THROUGH THE USE OF MITIGATIVE DESIGN FEATURES AND PROCEDURAL STRATEGIES (IN SPITE OF POSSIBLE CLAIMS L
' OF' EXTREMELY LOW CDFS).
L 1
- 3. Av01D EMPHASIS ON " BOTTOM-LINE" PRA NUMBERS.
USE INDIVIDUAL ACCIDENT SEQUENCE CONTRIBUTIONS TO CDF AND INDIVIDUAL SYSTEM FAILURE CONTRIBUTIONS TO CDF TO IDENTIFY RELATIVE IMPORTANCE AND T0 PRIORITIZE'MORE DETAILED FOLLOW-UP EVALUATION.
n
- 4. TREAT PROPOSED SAFETY GOALS AS JUST THAT-GOALS AND NOT ACCEPTANCE CRITERIA.
l l
1'
---m,-,,
n
.w-,,
-m v
- -, -,. ~v
,-e--
l..,
s t
k
' POTENTIAL CONSTRAINTS TO CHANGING REGULATORY l
CRITERIA FOR CERTAIN FUTURE REACTOR CONTAINMENT DESIGNS:
- 1. EVOLUTIONARY. REACTOR CONTAINMENT DESIGNS HAVE BEEN RELATIVELY' COMPLETE. FOR SOME TIME NOW.
RESISTANCE TO CHANGING DESIGNS TO ACCOMODATE NEW
. REGULATORY CRITERIA WITHIN BOTH THE INDUSTRY AND.THE NRC.
2.
SAME COMMENT (BUT TO A
MUCH LESSER' DEGREE, HOPEFULLY) FOR THE " PASSIVE" DESIGN,S NOW STARTING TO RECEIVE ATTENTION AT THE NRC.
9 0
9
v y
.. [
TECHNICAL BASIS FOR FUNCTIONAL PERFORMANCE REQUIREMENTS FOR
]
EVOLUTIONARY LWRs W. T. Pratt-Department of Nuclear Energy i
Brookhaven National Laboratory Upton, New York 11973 Presented to ACRS Joint Subcommittee Meeting Containment Systems / Structural Engmeermg December 13, 1989 bni l
aui
= ', '
f OUTLINE-Objectives, scope, and approach Severe accident containment challenges Containment performance requirements l
l H2 combustion High pressure meltdown l
a Containment bypass j
a i
1 Core. debris / containment boundary interactions Long-term decay heat removal 3
bnl
~ aui:
\\
_.<g.y.
.g.
.w-p
,4,.w,
-y*w
..wy t
e
,e s nl yv-u y
2__,'#n
-,.e. erg %m-_
% e a
d-
-*i'-'e-1
[
_~.
APPROACH i
Provide guidance on what severe. accident vulnerabilities j
need to be considered, reasonable measures that should 1
be taken to. address these issues and required documents-i 1
Reg. guide will:
j l
List specific severe accident vulnerabilities which evolutionary designs must address j
Provide guidance on form a,nd content of PRAs i
j i
i bnl 4
aui l
1 o
~,
O.-
7~
APPROACH (Cont.)
Reasonable measures are intended to demonstrate:
Core damage frequency below 10-5 per year y
Frequency of large release below 10-6 per year 4
1 Provide reasonable assurance that containment can effectively mitigate a range of severe accidents i
I Aim is to provide balance between prevention and mitigation I
bni 4
aui
~
3
,a
- ~.,..,
. < - ~
.. = ~-
- e-c--n.
APPROACH (Cont.)
I Prevention:
)
i Functional performance requirements were developed 1
to drive core damage frequency to 10-5 or lower-
]
i l
-b 1
Performance requirements derived for key safety l
l I
functions-1 i
J Reactivity control 1
l Inventory control i
1 Addressed'by requiring levels of redundancy, j
diversity, automation, and independence bnl I
aui i
i
.)
y APPROACH (Cont.)
l
~
Focus of today's meeting is on mitigation Aim is to develop containment performance re.quirements l
that give. reasonably. assurance (approximately 90%
chance) that containment remains intact given a core
~
melt accident Significant challenges to containments at existing plants were identified Performance requirements.were developed for those-challenges found.to be important contributors-to l
uncertainty in containment performance i
I bnl-aui
1 p
c Potential-Containment LVulnerabilities{ for Existing Plants..-
-1
../
Identified' by Previous Studies
~.
~
Large Ice
' Potential. Vulnerability Volume Cond Mark I Mark II Mark IIII
~
Containment Bypass:
Interfacing systems - loss-of-coolant accident
- YesI Yest
.Yes t yes!
Yest
- Failure to isolate containment Yes!
Yes!
Yes!
Yes!
Yes =i
- Steam generator tube rupture Yesl Yesl N/A N/A N/A Early Structural Failures:
Overpressurization with high temperatures:
- due to noncondensible gases andI steam Yes2 yes2 Yes -
Yes Yes-2 Yes No No Yes due to combustion processes Yes
- due to dirdet containment Yes Yes Yes Yes Yes heating Missiles or pressure loads:
No,
No2 Yes3 No2 due to steam explosions No2 2
Melt-through:
due to direct contact between core debris and containment No No Yes No No Late Structural Failure:
Overpressurization with high temperatures:
due to noncondensible gases and steam Yes Yes Yes Yes
-Yes
- ' due to combustion processes Yes Yes No No Yes Melt-through:
due to basemat penetration by core debris-Yes Yes Yes Yes Yes Notes:
N/A = Not' applicable.
IRelatively low probability but potentially high consequence.
bO!
2 ow probability.
(
L
, 3 ossibility of steam explosion in downcomers of some. Mark H designs.-
- SUl P
N
.m
+
~
IMPORTANT CONTAINMENT = CliALLENGES --
~
l PWR LARGE VOLUME. CONTAINMENTS i
o First draft -NUREG-1150:
Probability of early ' failure relatively large for high pressure core meltdown accidents (direct containment heating, H2 i
combustion, and induced steam generator tube rupture)
Probability of late, failure large if CHRS fail because of long-j term. pressure buildup Second draft of NUREG-1150:
1 Probability of early failure relatively low because of.
depressurization.of primary system by various mechanisms a
(stuck open valves, high temp. failure of hot leg, etc.) ~
l Because of low probability bypass events' are important-
)
contributors bnl aui-g y
-w
-.c-y 1 p.-c 3
y v-r- ' '** * '^; '
r**--
'~
"# --4>
c'-'-
+ - - -
2m m-sv---.
------ rm-
-mmm-mm
=.
~
IMPORTANT CONTAINMENT CHALLENGES --
'BWR MARK I CONTAINMENTS Both drafts of NUREG-1150:
1 Probability of early failure high because of:
l i
3 Direct contact of core debris with containment boundary -(drywell shell meltthrough) 1 Rapid pressure buildup (steam pressurization and DCH)
Late failure can also occur i
t i
.I bni 1
aui
c.
,E s.
l l
CONTAINMENT PERFORMANCE REQUIREMENTS _
1 Based on the results of.NUREG-1150;and related studies L
performance requirements were developed to address the-following five. challenges:
[
w Hydrogen combustion High pressure meltdown Containment bypass 4
t Core debris / containment tioundary interactions a
1 Long-term decay heat removal 4
1 m
bnl
)
aui.
i i
~.
S..
._1
.._;1, 3.-. _ _.,, _...22_=..,.
l. '
a y,,,.
J.
q 3
CONTAINMENT CONCEPTS i
..t PRESENTATION'TO THE 2
ADVISORY COMMITTEE'ON REACTOR SAFEGUARDS BETHESDA, MARYLAND 13 DECEMBER 1989 i
i R. J. LUTZ, JR;
. FELLOW ENGINEER WESTINGHOUSE ELECTRIC CORP.
'I s
- ~ -
f CONTAINMENT PERFORMANCE o
LARGE DRY PWR CONTAINMENTS DESIGNED TO THE EXISTING REGULATORY CRITERIA GENERALLY HAVE AN I
LARGE MARGIN FOR ACCOMMODATING SEVERE ACCIDENTS.
[
CONTAIN:iENT INTEGRITY IS MAINTAINED DURING l
THE EARLY DYNAMIC PORTION OF A SEVERE j
ACCIDENT l
CONTAINMENT INTEGRITY IS MAINTAINED IN THE LONG TERM DURING A SEVERE ACCIDENT IF ACTIVE CONTAINMENT HEAT REMOVAL IS AVAILABLE, EVEN FOR HEAT REMOVAL CAPABILITY WELL BELOW THE DESIGN BASIS.
CONTAINMENT FAILURES AS A RESULT OF FAILURES OF ACTIVE CONTAINMENT HEAT REMOVAL ARE PREDICTED TO OCCUR AT TIMEB GREATER THAN 1 DAY
+
LATE RECOVERY OF CONTAINMENT HEAT REMOVAL OR ALTERNATs HEAT REMOVAL MEANS HAVE GENERALLY NOT BEEN CREDITED IN PRA ANALYSES.
,l CONTAINMENT PERFORMANCE o
LARGE DRY CONTAINMENTS CAN GENERALLY ACCOMMODATE BOTH SHORT TERM AND LONG TERM SEVERE ACCIDENT PHENOMENA.
SUFFICIEllT VOLUME TO MAINTAIN HYDROGEN LEVELS BELOW DETONABLE LIMITS, CONTAINMENT GE0 METRY IS NOT CONDUCIVE TO TRANSITION TO DETONATION l
CONTAINMENT GEOMETRY IS NOT CONDUCIVE T0 i
DIRECT CONTAINMENT HEATING R CAVITY GEOMETRY IS CONDUCIVE TO CORE
~
+
CAN GET WATER TO THE CAVITY REGION 4
+
SUFFICIENT AREA FOR QUENCHING AND HEAT REMOVAL
l< l FUTURE CONTAINMENT DESIGN o
ENC 0URAGE THE ACCEPTANCE AND USE OF THE EPRI DESIGN GUIDANCE o
PRA MUST BE USED AS A TOOL FOR THE DESIGN OF CONTAINMENT AND CONTAINMENT SYSTEMS SUFFICIENT CONTAINMENT VOLUME TO ACCOMMODATE HYDROGEN GENERATION WITHOUT ACHIEVING DETONABLE MIXTURES SUFFICIENT REACTOR CAVITY AREA TO ACCOMMODATE LONG TERM COOLING BY A WATER COVER REDUCTION IN PROBABILITY AND/0R CONSEQUENCES OF CONTAINMENT BYPASS SEQUENCES ENHANCEMENT OF ASSURANCE OF CONTAINMENT ISOLATION E0VIPMENT SURVIVABILITY IN SEVERE ACCIDENT ENVIRONMENTS (E.G. FAN COOLERS)
I o
NO CHANGES IN REGULATORY REQUIREMENTS OR REGULATORY PRACTICE ARE REQUIRED FOR FUTURE CONTAINMENT DESIGNS, EXCEPT THE INCONSISTENCY BETWEEN PRA AND DETERMINISTIC SOURCE TERM
,'1 1
PROBLEM AREAS i
o ONE PROBLEM IN CONTAINiiENT DESIGN HAS BEEN IDENTIFIED WHICH WOULD REQUIRE A CHANGE IN THE REGULATIONS OR REGULATORY PRACTICE TO INCREASE THE i
LEVEL OF SAFETY AFFORDED BY THE CONTAINMENT SP/90 LONG TERM COOLING VS, DOUBLE C011TAINiiENT FOR OFFSITE DOSE (4'ITERION PRESENT DESIGN INCLUDES TRADITIONAL DOUBLE CONTAINMENT TO MEET 10 CFR PART 100 CRITERIA
+
SECONDARY CONTAINMENT MUST BE " CLOSED" TO MAINTAIN LEAKAGE COLLECTION FUNCTION
+
PROHIBITS ADDING WATER ON EXTERIOR 0F STEEL CONTAINMENT SHELL FOR DIVERSE LONG TERM HEAT REMOVAL o
OTHER CONFLICTS MAY ARISE AS A RESULT OF USING DETERMINISTIC METHODOLOGY WHICH IS NOT CONSISTENT WITH CURRENT PRA METHODOLOGY 1
i '
4
~ _. _.
g 4-
\\
e 9
M E4
=
W W
W E
=
4 w
W W
J O
W W
D W
N EW E
O W
M
.J
=
E C
3 E
E O
W H
W E.
W Ch E
D W
W J
w W
M W
m O
d SD M
K W
O W
G M
m W
W E
4
=J M
m O
4 W
4 M
W 4
J H
4 E M
W W
M O
E
=
D E
W W
W 4
4 E
E H
N.
W G
W M
W 4 O
M M
M W
D W
O W
W G
W E
M S
p E
M 4
J B
W C
M O
O M
m 4
4 W
H W
E kJ 4
9 4
W E
H E
4 E
H W
E O
M W
4 W
E W
E-E i*
E
=
W
v-p PRESENTATION OBJECTIVES:
i PROVIDE PERSPECTIVE FROM.THE POINT OF VIEW 0F THE ADVANCED LIQUID METAL COOLED REACTOR (ALMR) PROGRAM j
i I
.s l
BASIC APPROACH I
TRADEOFFS j
UNIOUE CHARACTERISTICS RECOMENDATIONS i
i i
l I
i 7
l q
1 I
.i 1
1 1
c <
r
}
s i
i l
I i
i
=w W
i 4
D O
i M
s O
H E
i n
ti M-M g
J M J
4 W
E M
W M
W e
m J
4 W
t 4
M e
E @
S W
O W
W M k
E 4
M W
A O
W W
i E
g W'
M H
O W
3 p
w >
W H
E M
C W
4 g
W O
H W
W M
W 4
4 W
W =
W W
i M
W 4
E g
W M
M O
O O M m
W E
E a
g g
g E
O 4
O N
H m
O H
W W
E E O
M W
W H
O W g
H H
O W
E E
M h W
(
W O
O W
W W
W W
t-W W
E W
D u
3 m
i E
W M
E R m O
W l
M E
M W
E D W W
E M
W W
.E h
w g o
O E
W W
D M
O M
O 4
O W
G W
g l
H M
3 W
M g
w E O W
4 M
N H
H W
W E
t D
W W
D M
M E W
r g
~
l W.
O O
H w
O W
E O
M M
W M
M M
p r
M M
E W
m 4 m i
E O
O M
M W
D H
W H
4 W
E F
O E
m E E O
W E
4 W
6 3
O W D
C W
M J
4 6
O W
g E
W O
W W
E W
W E
E M
N H 4
W m
M 4
W D
E M
E E W M
4 W
M d
M W O
M g
b M M E
4 E
E r
w W
E W
E E
M R
H J
g O
h g
E M
W x
q W
E W
O E 4 O
W O O
M W
W W O
m O
E H
H E
E H
E W
O e M
O E
4 O
O E
O O H
w m
a W
W M
W W W D
D W
M W
E O
E m
a J
l m
H E
m E 4 m w w
D O
E E:
W M
4 g
4 m
)
Wl E
E2 W
H E
E O
m M
W O
$ W m
W W
O W
M E
M E
M M
E g H
W W
H 4
W W 4 e
u W
M H-O E W
E W
g 4
M E
E W O
W >
W
, O M
O W 4 m a
W E
W W
W 4
i H g M
H m H
w W
i M>
4 O
O O
O O
l i
r I
.~.
_____.m..-
.c j
)
l i
f
)
l E
t M
Eer 4
5 l
H h
.J M
ad
?
E W
4 i
4 M
H O
f W
H
@ E f
E W
M W
D W
h H
M E
H l
J 4
W H
W M
W 3
O M
O W
AC M
et W
W O
M
(
b H
E H
U W
l W
et O
w w
er W
H J
E r
a W
M M
M eg W
W W
O W
M M
M H O 7
4 O
E M
W O
J W
M I
.J O
E H
M D
W es DC 4
4 W
6 kg b
E W
3 eg M
4 5
E M
T C
W W
M O
W H
h H
M E
er W
H C
+
W H.
W W
O M
M @
I H
W X'
W D
E 4
> M i
W M
et an M
t=
M E
M er W
M e
E M W 6
O E
O E
+
C E
eg W
W er O
H
.J O
M O
O M
e.
3 J
J H
H O E er M O H
4 O
4 4
4 O O
W W 4
et er W
W M
6 er et E
H W
M J W H
E 6
E E
W W
O O
M M W W
88 J
E SC W
H H
m' H
am W
O ag E W M'
M
= 4 3
W H
W at M
M U
M
(
W 4
M J
M O
W M
H 88 C
=J W
E E
E SC E W
O b
H eg M
4 M
H M
W M
M H
H M
M M
t H
E 4
6
-O 4
W M
M O 3 W
e O
W q
M W
W 6
E er M W W
W M
O At C
O W W eg O
at W
M U
6 E
6 eg O
W 4
H O
H W
W es E
M O O
4 R
J E M
M
.J B
=
'~
er W
W W W
M M
E E
6 W W W
M M
O-4 W
W W af
(
(
M H
eg M
E E
4 4 b
E M
E NEJ<l e
o o
o o
e o
~
.~
~ ~.,
..,c ALMR -
SUMMARY
OF RADIATION RELEASE CRITERIA BASED ON CURRENT NRC STAFF AND ACRS RECOMMENDATIONS NRC STAFF KEY ISSUES PAPER SECY-88-203 ACRS LETTER ON SAFETY GOALS MAY 13, 1987 PROBABILITY LESS TNAN:
PER PLANT YEAR TO EXCEE0:
I 10-2 10CFRS0 APPENDIX I II 10-4 10% OF 10CFR100 III 10-0 10CFR100 AND PAG FOR 36 NOURS IV 10-1 FOR LARGE RELEASE FOR THE FULL SPECTRUM OF SEVERE CORE DAMAGING EVENTS OBSERVATIONS:
THESE CRITERIA ARE MUCN MORE RESTRICTIVE TNAN THE SAFETY GOALS THE 36 NOUR PAG LEVEL IS ESSENTIALLY LIMITING, IMPLIES A HIGH LEVEL OF PREVENTION
,3 A_LMR - HIGH EMPHASIS ON PASSIVE SYSTEM FEATURES OBJECTIVE IS TO ACHIEVE HIGH RELIABILITY WITH PASSIVE FEATURES IN:
DECAY HEAT REMOVAL, REACTIVITY CONTROL, AND CONTAINMENT FUNCTION; l
VIRTUALLY IMMUNE TO OPERATOR ERRORS l
A STRICT DEFINITION:
A FEATURE WHICH RELIES ONLY ON THE LAWS OF NATURE AND THE l
i l
STRUCTURAL INTEGRITY OF MATERIALS, REQUIRING NO SENSING, SWITCHING, MOTIVE POWER, OR HUMAN ACTION, AND WNICH IS NOT DEFEATED OR IS j
DIFFICULT TO DEFEAT BY NUMAN ACTION.
EXAMPLES:
THERMAL EXPANSION, HEAT CONDUCTION, NATURAL CIRCULATION l
LESS STRICT DEFINITIONS ALLOW:
l ACTIVE INITIATION OF PASSIVE SYSTEMS (ONE-TIME VALVE MOTION) l DC POWERED YALVES AND INSTRUMENTS l
OPERATOR INTERVENTION IT IS RECOGNIZED TNAT THERE IS A WIDE RANGE OF ROBUSTNESS IN PASSIVE SYSTEMS
(
IN TERMS OF VULNERABILITY TO STRUCTURAL FAILURE OR HUMAN INTERVENTION l
. - ~..
.. ~...
- n.. ',
l
' o E
=d M
et J
W 4
i 8
d O
H i
h N
D E
W i
O O
6 M
E M
4 i
er w
H M i
M M
at U
H 6
H 4
W W
4 6
M W
J E
W O
J W
H E
W M
W M
- e M
O M
6 bd i
M 4
E H
O M
g
=
5 5m 5
O
.J E
W E
J W
eg W
W as O
W b
H W
.J 4
O H
.J J
W 4
W 6
E
= m W
J M
.J O
as W
H M
M 4
W G
i E
at E
M as; M
er W
M W
4 er W M W
W M
E M
er E
D E
O J
W D
H W
D H
W M
R W
L' O
ag W
W M
E M
W D
3 as E
E 4
O W
at E
4 6
W 4
W H
E CD w
O E
E 4
E 4
E E
W Og H
H O
O 4
M O
W as O
4 at W.
e.
M am 6
as H
A M
4 M E
H
.J
+
E A
E M
J W
4 O
.J 4
E W
4 E-O at 4 E
E M
M O
E M
W O W
M W
3 i
eg M
W W
H H @
er H
H W
H 6
M M
M M
M H
W 6
M M
eg H
3 E
W M
E H
H w
W W
H O
4 er 3
O 4
eg W
W W
O O
3 W
6 O
E E 4 E
R H
.J O
H W
.J O 4 M
4 H
H as E
4 sh M M H
N H
E E
W E
W R
H E
H J
W W
W M
6 W
W W W
J W
eg E
E E
W E 4
kg W
R R
H W
W O
.J H
W E
M E
M M
D O
W W
4 W
H 4
4 O
M et M
E
.J H
H M
W 4
W W
m 5
E M
M M
M H
M E
4 M 4
O E
4 E
M E 3 at E
W E
W W
M M
e6 M
W W
H M
F M
M H
H H
l-M O
W ag O
W W
H W E M
W M
W I~
E R
N O
.J N
W W
W N
.J N
O M
M W
W 4
M 4
M M
M M
E M
ac W
6 m W
E en E
H M
4 3
O 3
eg M
M
.J 4
4 M
E O
M O
W M C
ao e
(
O
.J L
J O
W W
G W
E W
O W
E en E
J E
W E
O W
W eg W
W 4
W 4
4 W
er 4
M W
M eg E
H
\\
I MW W
W E
SC M
C O
O O
O O
O O
O
- - - -., _ _ -. -. _....,....... ~. _., _ _,. _. -., -. -..., _,. -,.,. _ ~,..
..,,,m-,,x.--.-
r--
c t
CONTAINMENT DESIGN CONSIDERATIONS FOR l
THE SYSTEM 80 PLUSm STANDARD NUCLEAR POWER PLANT DESIGN
~
I f
COMBUSTION ENGINEERING, INC DECEMBER 13, 1989 h.
)
9
---n m
v-
---*----e-
~-
'. ' e. -
t i
i
-i i
i j
CONTAINMENT DESIGN APPROACH i
- MAINTAIN TRADITIONAL DESIGN BASES (DOUBLE ENDED GUILLOTINE BREAKS, ETC.)
I i
0
" LICENSING" ANALYSES
- SUPPLEMENT WITH SPECIFIC SEVERE ICCIDENT
-MITIGATION FEATURES O
"BEST ESTIMATE" ANALYSES i
r I
. ', c..
l i
i I
SYSTEM 80 PLUS CURRENT CONTAINMENT DESIGN BASIS
- CODES AND STANDARDS o 10 CFR 50, REG. GUIDE 1.57, ASME e
SECTION III
- DESIGN CONDITIONS
- INTERNAL / EXTERNAL LOADS
- PRESSURE / TEMPERATURE
- NATURAL PHENOMENA
- CONSTRUCTION LOADS
- HDRODYNAMIC LOADS
- LOADING CATEGORIES i.
O SERVICE LEVEL A THROUGH D t
,v-.
7
, r e.,.
i 4
j s
i i
ACCEPTANCE CRITERIA FOR CONTAINMENT DESIGN
\\
l
- ASME CODE CRITERIA
- SRP SECTION 3.8.2, " STEEL CONTAINMENT" -
- STABILITY (BUCKLING) INCLUDES REQUIRED SAFETY FACTORS
- SERVICE LEVEL A AND SERVICE LEVEL B ARE CHECKED TO SAME STRESS INTENSITY LEVELS
- NO ACCEPTANCE CRITERIA FOR ULTIMATE CAPACITY
- c t..
i j
i
?
+
i CONSERVATISMS IN CURRENT DESIGN BASES i
- USE OF LEAK-BEFORE-BREAK WOULD REDUCE PEAK ACCIDENT PRESSURES BY APPROXIMATELY 50%
- ULTIMATE PRESSURE CAPACITY OF CONTAINMENT IS APPROXIMATELY 4 TIMES GREATER THAN DESIGN PRESSURE LIMITS
- MARGIN EXISTS BETWEEN DESIGN PRESSURES AND ACCIDENT PRESSURES (GDC 50)
- SERVICE LEVEL B LOADING COMBINATION COMBINES EFFECTS OF PEAK ACCIDENT PRESSURE WITH PEAK OBE
- ANALYSIS GENERALLY PERFORMED WITH STATIC PRESSURE AND RESPONSE SPECTRA APPROACH FOR PEAK DESIGN PRESSURE AND EARTHQUAKE LOADINGS RESPECTIVELY (IN - LIEU OF TIME HISTORY APPROACH)
- CURRENT SOURCE TERMS CONSERVATIVE BY ORDERS OF MAGNITUDE
.', r ;..
1 I
1 i
ADDITIONAL TECHNICAL ISSUES:
SEVERE ACCIDENTS j
l I
COMBUSTIBLE GAS CONTROL i
O METAL / WATER REACTION OF CLADDING 0 ADEQUATE MIXING IN CONTAINMENT O DETONATION LIMITS i
4 CORE DEBRIS COOLABILITY O ADEQUATE AREA IN REACTOR CAVITY O' RELIABLE METHOD TO COOL CORE DEBRIS
- DIRECT CONTAINMENT HEATING O POTENTIAL FOR HIGH PRESSURE EJECTION O PATHWAY TO CONTAINMENT-VOLUME
- CONTAINMENT VENTING i
1 0 MARGIN TO FAILURE L
0 CRITERIA FOR INITIATION i
mm,,.-_.-
-.p
__ c
m t-y.
i f, -)-
LARGE, STEEL SPHERICAL CONTAINMENT Dual Containment i
200 Ft. Diameter
/g N'
N Increased Space c._
~
"\\
l For Maintenance
& Access
! E i. p + "A A"
- = -
Designed To
-Q -tr
. C' r r mf -
.t.
9
. =
=.
=
m Mit.igate Severe
\\
o
- c:D
=
. 'o'
-y
- e u,ces:,h=rf.
_ aint.g_iJ.,., :.e.
i.
Core Damage iur u.
~
I =
f u.m 2 = s-
.I Shadow Area
.mtAf e
Pr.:, 1 s
{
Houses Safeguard
-cas h e ~~ i
+.
Systems i
i i
i
i 1:
.. -o s
j i
j CONTAINMENT TECHNICAL DATA CONTAINMENT:
CONTAINMENT TYPE STEEL SPHERE STEEL TYPE SA-537 CL. 2 INTERNAL DIAMETER 200 FEET.
WALL THICKNESS 1.75 IN 3
FREE VOLUME 3.4 x 10-6 CU. FT.
DESIGN PRESSURE 49 PSIG SHIELD BUILDING TYPE CONCRETE i:
INTERNAL DIAMETER 210 FEET l
WALL THICKNESS 3 FEET i
g
e_
{
-j i
i
._p,.. ;
1, u 4
1 d,
j in.e 4 4. =
n
[
h*
y a,.:T
}
(
(
l u
a a
b l 5 M
F]u 3-e p
E N,
u
-h
,w6ii 4
l r-1 sh, t.- y u.tt
~
'~
y 7
/
j
\\
L y
b sy,.st. p l
=
a n
I-g IL I
=
s M em s.n eme g
.,. m i j
- =O
-- 'd '
~2
-UI.":EMJA J.
h, l='
. ~-
g.
%eple=::=mme-g M
---lw
,g.m
.m
{
m&
{,
%j X~
. y m' rr sqi.'
-(
)
- sr3..
gs n.m
.. n.=
SECTION VIEW 90
- 270 SYSTEM 80+
w
.----e--,..--,-v-.r
,v n---w--
vv
- + - - - -
t SAFETY DEPRESSURIZATION SYSTEM PROVIDES SAFETY GRADE PRESSURIZER AND REACTOR VESSEL POST-ACCIDENT VENTING OF NON-CONDENSIBLE GASES PROVIDES SAFETY GRADE RCS DEPRESSURIZATION AND COOLDOWN WHEN NORMAL PRESSURIZER SPRAYS ARE UNAVAILABLE PROVIDFS FOR RCS DEPRESSURIZATION TO INITIATE BLEED An'U FEED FLOW IN UNLIKELY EVENT OF SUSTAINED TOTAL LOSS OF FEEDWATER FLOW PROVIDES FOR CONTROLLED RCS DEPRESSURIZATION DURING SEVERE ACCIDENT SCENARIOS
- l**a e
1.
T
,PSVs, a6 dr i
bb dr h
h h
h ILEED LEED SD 2 SD 1 30,3 30 4
[
h I
h PRESSURIZER i
l RCGV RC 106 RC 105 l
@@1:
- KD4 I
)
l RC 106 i RC 107
~
l RCGV l
RV.103 RV.104 k
- KD-6 RV 101 RV.102 f
ORIFICE RDT Ry RWST l
FIGURE 8 SAFETY DEPRESSURIZATION AND VENT SYSTEM Y-Mail l-
L 4
! : ':i t.
METHODOLOGY FOR SEVERE ACCIDENT ANALYSIS
- USE PRA TO ESTABLISH SIGNIFICANT ACCIDENT SEQUENCES
- USE BEST ESTIMATE, DETERMINISTIC MODELS (MAPP) TO ANALYZE CONSEQUENCES.
- OVER, WHELM UNCERTAINTIES BY:
O DESIGN FEATURES 0 CONSERVATIVE ASSUMPTIONS O SENSITIVITY ANALYSES
- USE EPRI GROUNDRULES AND ASSUMPTIONS AND ARSAP STUDIES AS BASES FOR ESTABLISHING ACCEPTABLE METHODS l
)
o ' J A., o i
1 I
e CONCLUSIONS
- CONTINUED USE OF " TRADITIONAL" DESIGN BASES IS APPROPRIATE.
]
- COMBINED WITH " LICENSING" TYPE METHODOLOGY, THESE i
DESIGN BASES RESULT IN RUGGED, CONSERVATIVELY DESIGNED 1
CONTAINMENTS.
~ IT IS PRUDENT TO INCLUDE FEATURES TO MITIGATE SEVERE ACCIDENTS, IF BASED ON A DETERMINISTIC, BEST ESTIMATE APPROACH.
a i
e I
V l
1 o
- -