ML20238A092
| ML20238A092 | |
| Person / Time | |
|---|---|
| Issue date: | 08/19/1987 |
| From: | Vissing G Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-675A NUDOCS 8708280311 | |
| Download: ML20238A092 (84) | |
Text
9 -
l
[p yo UNITED STATES l
NUCLEAR REGULATORV COMMISSION l
,y WASHINGTON, D. C. 20555 I
gg
,j o, W August 19, 1987 Project No. 675
)
j I
MEMORANDUM FOR:
The record FROM:
Guy S. Vissing, Project Manager l
Standardization and Non-Power i
Reactor Project Directorate l
Division of Reactor Projects III, IV, j
V and Special Projects I
SUBJECT:
SUMMARY
OF MEETING OF THE NRR STAFF WITH COMBUSTION ENGINEERING CONCERNING SEVERE ACCIDENT ISSUES FOR 1
SYSTEM 80+, AUGUST 12, 1987 l
INTRODUCTION A meeting of Combustion Engineering (CE) and the staff was held on August 12, 1987, at the request of CE at the NRC offices in Bethesda, Maryland, to discuss severe accident issues for the System 80+ design. The purpose of the meeting was to provide the big picture of severe accident issues in the design cer-tification program, to discuss the overall approach to design certification and 4
describe the process leading to design certification. provides the attendance list for the meeting. Enclosure 2 provides CE presentation material which presents the overall approach to design certification for the System 80+
design. provides the I. T. Corporation presentation material which I
describes the Advanced Reactor Severe Accident Program (ARSAP). Enclosure 4 consists of a draft CE paper which presents in narrative fccm the CE program plan to address the severe accident issues in the CE design certification program.
DISCUSSION The System 80+ design will be an evolutionary design based on the current System 80 design. The Objectives of the System 80+ design is to enhance safety, improve performance, reduce total cost of a plant, use proven technol-ogy and to reduce licensing risk. The program is sponsored by DOE under the DOE Advanced Light Water Reactor (ALWR) Design Verification Program. The CE program is integrated with the DOE and EPRI programs which has as a goal design certification of the System 80+ through rule making in the 1990s. provides an overview of the design certification program and the CE plan for addressing the severe accident issues in the design certification program. Enclosure 4 is a narrative discussion of the CE plan to address severe accident issues.
l@gl The base line PRA will be a Level 1 PRA based on the current System 80 design.
The final PRA for the System 80+ design will be a Level 2 PRA. CE indicated that external events will not be included in the PRA. This was a concern to The base line ( g -
the staff and a subject which would need further discussion.
PRA will be an evaluation tool to support the evolutionary design process.
8708280311 870819 PDR PROJ 675A PDR
l August 19, 1987 The base line PPA will be complete by the end of Dr.tober 1987.
It will be submitted to the staff for information. Also a meeting is planned for CE to l
discuss the base line PRA after the staff has an opportunity to review it. CE indicated that they would use their codes for the PRA and the NRC would do their independent analyses.
CE will use their MAAP code to support the deterministic analysis. There was some concern expressed that they were not using the NRC codes for severe accident analyses. CE indicated that they would compare their MAAP code with l
the NRC codes.
It was uncertain how much staff review would be ir.volved.
CE indicated that the first submittal, Chapter 1, will be in early September with Chapter 10, the balance of plant, to be submitted late in October. The first major submittal will be in October.
1 The severe accident issues will be identified and handled though topic papers prepared under the DOE sponsored Advanced Reactor Severe Accident Program (ARSAP). Those topic papers which are applicable to the System 80+ design will be submitted under the System 80+ docket. The topic peNrs will be submitted in six sets. The first set (a single paper) will addr a s issues which have effectively been resolved. The remaining sets will consist of approxi-mately 26 papers which will need resolution for the System 80+ design. The first topic paper is scheduled for submittal in September and submittals will i
continue at approximately twn month intenals unt.il July 1988. CE requested that each topic paper be resolved within five months after submittal. The topic i
papers will be proposing criteria which has not been previously established, i
i CE recognized that, by their nature, many issues will not be completely l
l resolved until completion of hearing and design certification.
1 CE would like to meet with the staff again in September to discuss the first l
topic paper.
CE wants staff feedback on the list of topic papers to determine if they are I
on the right track. They requested feedback on their overall plan. They I
emphasized that for each issue they would need NRC guidance and they needed a conrnitment from the staff to complete the work. The staff indicated that i
the overall program is the first of a kind. The approach was a reasonable approach. The submittals will need staff review and resources will have to l
made available. At this point the schedule for review was uncertain.
Original signed by
)
Guy S. Vissing, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, l
V and Special Projects l
Enclosures:
DISTRIBilTION JEdordan ACRS(10)
As stated d Docket;Eile/
'LRubenstein NRC & Local PDRs GVissing JPartlow PDSNP Reading OGC-Beth NRC Participants cc:
FCoffman
,TCintula(AEOD)_TPratt(BNat'l Lab)
See next page q
E on GV ing:bd LRubens n
0
/87 08/// /87 08/q/87
____________J
1
~
l August 19 1987 0
2-I The base line PRA will be complete by the end of October 1987.
It will be submitted to the staff for information. Also a meeting is planned for CE to discuss the base line PRA after the staff has an opportunity to review it. CE indicated that they would use their codes for the PRA and the NRC would do their independent analyses.
CE will use their MAAP code to support the deterministic analysis. There was some concern expressed that they were not using the NRC codes for severe accident analyses CE indicated that they would compare their MAAP code with the NRC codes.
It was uncertain how much staff review would be involved.
/
CE indicated that the first submittai, Chapter 1, will be in early September with Chapter 10, the balance of plant, to be submitted late in October. The first major submittal will be in October.
The severe accident issues will be identified and handled though topic papers prepared under the 00E sponsored Advanced Reactor Severe Accident Program (ARSAP). Those topic papers which are applicable to the System 80+ design will be submitted under the System 80+ docket. The topic papers will be submitted in six sets, The first set (a single paper) will address issues which have effectively been resolved. The remaining sets will consist of approximately 26 papers which will need resolution for the System 80+ design.
The first topic paper is scheduled for submittal in September and submittals will continue at approximately two month intervals until July 1988.
CE requested that each topic paper be resolved within five months after submittal. The topic papers will be proposing criteria which has not been previously established. CE recognized that, by their nature, many issues will rot be completely resolved until completion of hearing and design certification.
CE would like to meet with the staff again in September to discuss the first topic paper.
CE wants staff feedback on the list of topic papers to determine if they are on the right track. They requested feedback on their overall plan. They emphasized that for each issue they would need NRC guidance and they needed a commitment f rom the staff to complete the work. The staff indicated that the overall program is the first of a kind. The approach was a reasonable apprcach. The submittals will need 5,taff review and resources will have to made available. At this point the schedule for review was uncertain.
,,-s.
/
.x.
/.,m
. Guy S. Vissing,f roject Manager Standardization and Non-Power Reactor' Project Directorate Division of Reactor Projects III, IV, V and Special Projects
Enclosures:
As stated cc:
See next page
Cortbustion Engineering, Inc.
Project No. 675 Advanced CESSAR I
cc: Mr. A. E. Scherer, Director Nuclear Licensing Combustion Engineering, Inc.
1000 Prospect Hill Road Windsor, Connecticut 06095 Mr. C. B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering, Inc.
7910 Woodmont Avenue, Suite 1310 Bethesda, Maryland 20814 I
l
/
4.
J ATTENDANCE LIST FOR MEETING WITH COMBUSTION ENGINEERING.
CONCERNING SEVERE ACCIDENT ISSUES FOR ADVANC"D SYSTEM 80 8/12/87 NAME-ORGANIZATION Guy S. Vissing NRC/NRR/PDSNP Mario H. Fontana IT/ARSAP
- Rick Summitt IT/ARSAP Mark W. Crw p C-E Paul M. Haas IT/ARSAP Bob Morris IT/ARSAP Kirby Dawson EGLE Idaho ARSAP Steve Long NRC/NRR/RAB Dean Houston NRC/ACRS/ Staff P. K. Niyogi NRC/RES/DRAA/PRAB Chuck Kling C-E Regis A. Matzie C-E George A. Davis C-E Mike Green C-E l
Frank Ross DOE Zoltan R. Rosztoczy NRC/RES E. Chelliah NRR/RES/PRAB Brad Hardin NRC/RES/PRAB Mark Rubin NRC/NRR/ DEST Larry Kopp NRR/RSB Jerry Holman-NRC/RES/SAIB Jacques Read NRC/RES/SAIB Gil Brown NRC/ACRS J. H. Raval NRC/SPLB/NRR C. G. Tinkler NRC/SPLB/NRR R. J. Porrett NRR/DREP/RAB
l.j excto ese 2 l
l l
F CESSAR DESIGN CERTIFICATION PROGRAM
- A MEETING WITH THE NUCLEAR REGULATORY COMMISSION -
~
" DEGRADED CORE ISSUES" i
i DATE: AUGUST 12,1987 J;
COMBUSTION ENGINEERING, INC.
E
~ ~~
6
CESSAR DESIGN CERTIFICATION PROGRAM
-~
- A MEETING WITH THE NUCLEAR REGul.ATORY COMMISSION"
~
~
~~
" DEGRADED CORE ISSUES"
'~
~'
DATE: AUGUST 12, 1987 1
'~
COMBUSTION ENGINEERING, INC, l
-._,. 9._...J
I j
a d
i 4
\\
4 1
i 1
1 1
l PURPOSE i
i i
APPRISE NRC STAFF 0F C-E's DESIGN CERTIFICATION 1
i i
a j
PROGRAM PLANS FOR ADDRESSING SEVERE ACCIDENT i
(DEGRADED CORE) ISSUES 7
i i
I I
1 l
i 1,
5, l
]
i i
S, 1
v I
l l
1 l
i e
1 I
i l
i 1
somousTioN)sNCINEERING
-l
l l
1 i
AGENDA l
1 1
1 9:30 AM INTRODUCTION M. GREEN 1
\\
k 9:35 AM C-E ALWR OVERVIEW R. MATZIE l
9:50 AM OVERVIEW 0F C-E's DOE G. DAVIS 1
DESIGN CERTIFICATION I
PROGRAM 1
10:15 AM PLAN TO ADDRESS SEVERE M. CRUMP i
ACCIDENT ISSUES IN THE C-E h
. :l DESIGN CERTIFICATION PROGRAM i
12:00 PM LUNCH i
i i
i 1:00 PM SEVERE ACCIDENT TOPIC M. FONTANA i
i PAPERS l
f I
T 1:45 PM ADDITIONAL DISCUSSION, IF M. GREEN A-NECESSARY j ' ~ :.
)
i
)
I,
}
I coassusTion)sucamaansma
i e
l 1
h l
l INTRODUCTION l
1 1
CESSAR DESIGN CERTIFICATION (CESSAR-DC) l
~
i i
c
! j LICENSING APPROACH q
l 1
3 l l j
(
l 1
DR. M. D. GREEN I
i i
ol i
s. -:
l !, -
MANAGER, STANDARD PLANT LICENSING l
l 1
l l l
\\
l
,1
- seassusTlem) ENGINEERING
~
\\
- ~ ~
I l
4
l 1
i l
CESSAR-DC LICENSING APPROACH a
- ~... <
REVIEW SYSTEM 80 DESIGN WHICH OBTAINED FDA O
l l
o UPGRADE AREAS IDENTIFIED FOR IMPROVEMENT l
0 MODIFY DESIGN TO INCORPORATE RESOLUTION i
OF SEVERE ACCIDENT POLICY ISSUES
[
0 OBTAIN DESIGN CERTIFICATION i
l i l
i I
(
_... %......no.).
4 ed 1
i j
l 1
l C-E ALWR OVERVIEW 4
j I
!1 1
i
=
4 i
l
.I l
I DR. R. A. MATZIE 1
1
)
i i
i DIRECTOR, ALWR PROJECT OFFICE i
t i
b I
i I
i L.
l l
1 1
1 l
C-E ALWR PROGRAM OBJECTIVES O
TO DEVELOP AN ADVANCED PWR DESIGN WITH ENHANCED SAFETY l
IMPROVED PERFORMANCE l
REDUCED TOTAL COST j
PROVEN TECHNOLOGY i
0 OBTAIN NRC CERTIFICATION OF THE ADVANCED PWR DESIGN TO REDUCE LICENSING RISK RETAIN NUCLEAR AS A VIABLE FUTURE OPTION i
l i
i k
N...un.).
.)
i 1
e C-E AL!@. PROGRAM 1
1 i
i I
I i
l EPRI ALWR I
l j
I REQUIREMENTS I
,j DOCUMENT l
1 l q 1
l DOE I
I DUKE (DOE) l Q
~
I' ADVANCED I
CONSTRUCTABILITY -l-4H i
l I&C I
l PROGRAM l I l
PROGRAM i
j i
i h
.j jl
.]
I I
I i
1
.~
l C-E l
l DOE I
l PRODUCT ENG. l l ADVANCED REACTOR I
]
I & DEVELOPMENT
- SEVERE ACCIDENT I[
l I
PROGRAM i
PROGRAM i
l!
I I
I Il }
,g i
T j
,l l
j i
i i
I I
DOE ALWR I
I DESIGN VERIFICATION I
l PROGRAM I
I I
l I
I
'1 1
lI CERTIFIED SYSTEM EQ DESIGN
-sosseusTion)sucamammsme
J DESIGN CERTIFICATION PROCESS i
l
/\\
(_X.
1r 1P IP REVIEW SYSTEM 88 I
i f
MAINTAIN DESIGN / DES GN N DEFER INCORPORATE CHANGE 1r 1
PRELIM.
PRA DESIGN r\\
j CHANGE l
1r 1r i
OES.
COST
{
ANAL EST MAINTAIN DESIGN M-n PREPARE i
i CESSAR I
u 1
4 SUBMIT CESSAR 1
- COMBUST 10NhENGINEERING
- ... l
~
l
GN S
I S
N N
r e
- g E
t e
C O
h I
c S
L I
R.
c, T
g E
A E A u
a C
LC N
U N
p U
F DN s
A N
O tn e
I n
T o
S G
p s
m g
l i
o A
M N
o s
r s
I C
gt r
E I
R d
e s o r
d e i n N
Z T n E
a E
o n
ns C S a I
Y n D E
f s
a i
I N
w a
aged i
S w S
a s
n nn lc N
I e
E I
G C m
Ei r
t gs ga gg i
RN R e
a nt n a
P N.
t A
e cEni s
A E
E D s
s e er G E
E S e eonea ye nt n e G
C S J L R C
v S
di hcriung n
l t
e gcat gs e
R U
S u n e e ne s
N M
n O
f S
~
I M
ET S
Y S
S M
&[
R E
E c
h gJ.
N u\\
s A
a r
E e nt e
e o r /
D s
V ec e
E t'
Hna D m ;t.
e s et EH ycpH n a g:
L R
, P osd e H nle g
n E en A p C
e l
w e t
E E
pc p
a oec oa co C
nl nDa s
U C
V r
l s
e k e hcdv v tnuv U
s nd eos e N
N I aADCPD e
A
T re p
r p
e k
l a
r h n c
u K
S T
n T
I
.W l
m E
S l
t t
e i
l A
R r
R a
P h
US C
I kd cr G
n o ol I
C R
a nfl B
n r wi N
1 E.
S u
e aG e
h rc R
em R
i i N
T r
L S
. e GO A
Y P D L.
l E
S F
e b
EF R
WJP
,l l
U Ys S
t Tu IG LA n n
OI N n Ga NA y et e T AAM1
?
l e T
F t
i
- tt ICT
. O s
nt s
RUC
. F I
e Au PRR P A
. Gr e
T e
sm B T )a.
l e f t l S
rP
. Et i
E e
t H
'n uP
~
i O
AI
[C
.l I
R I M. G I
AI TE T
- Z WJF ee CI LOF t
N s
A ARO l
N P
Gt c'
s tO iL RE G
e A s l
Ni z
A v a
t t r 9 a R
ao a
l I
E.
e A
l A r a.
i iG t
.D P
R ta Cn S
e
&s k
E E
a D
y Ds A
1 I
EA E
CT L
g 08A.
K z_
wag S
A t
m Ai T
o w
t la I
L
[
D i
, 3, 3
i'
'N
')
['
/
5
^i
,A N
g
,i k
I
\\
t
. i i
q i
l I
OVERVIEW OF C.E's 4
i DOE DESIGN VERIFICATION
)
t b
i i
(;
PROGRAM 1
a e
i
.t 1
q s
)
l l b
- )
G. A, DAVIS F
-i 1
I 1
1 PROGRAM MANAGER, DOE ALWR PROJECT i
l l
i i
\\
y w.
\\
[ I 4
p
.{ '
L.....
~
T Y
N L
E
)
T D
R L N
A A I
, E C
S V R
C S O R
A E R 0
S C P r
E E
(
P R
S A
T E
Y T
O V
L RR N
E A
OO S
N PF S
A E
E R
R
)
U O
Y C
S S
T T
S R S
E C
E I N M
G d
F S
(
I A
N E
A Y
R C
A R
S LN G
R A O l
i O
N C
D D
NI R
E N
A S P
W N
C A
S
~
E G
N YI N
N I
A T
T M
)f S
V' N
EM J
D E
D E
F O I
N D
A M
AC A
A S
S C
N 0
E S
Y S G 8
O E
DR I
F T I D
S RO I
NS M
S AT E
MD T
O A
DA R
EE E
M NL V
E S
R X
A U R 0 Y
F S
TG N'
I 8 S
I S1 G
U N
R P
I QM T
O E
S EE N
I C
E R E
RT E
T C A S
I D
S M
A T
E N
RY E
M S
OL O
E WS L
R I
TC I
O L
P O
L U
T D
AN M
F
)
I S N S
N P B
L E
I I
I I A A
GO U
RD D
S B
NT Q
PE N
ER O
A '
ET A
T A R
HT C
A A
(
P CI R
YR Y
R M
N F O F
O M M
EB I P I
P A R
R U R
TR T
RR O
AS E
NO N
O G F
P W
EC E
C O R
ED S
DN D
NR E
H N N
I I I P P
PN A
KSA T
R 0a A
M p
f
~
- ll,.:
RO F
STNE 6i M
ER I
U M
Q E
M E
T E
R S
T Y
S L
E S
Y A
P S
N D
Y O
C L
R I S S
P E
T M P
T C E U
A N T E
S W
U S L
D R
F Y U
M E
E S
D A
E T
D O
E T
F N
ER J
M T
N E
Z A S
E Y
C I
E R
M C
DL E
R N
N L
R C I
E O
A U W
A O
E A
G R
D N P
L T
R T
N -
C N
E N
A N R
U O
M O
T O A
N C
E C
S N E
L CUN O
l-F
8 1
8 8
9 7
9 8
9 7
8 1
9 1
8 9
8 1
I A
9 1
8 N
9 R
RT 1
9 E
R 8
E A1 E R
1 B
E 9
H El D U
O C
r E
E f
SI T T
E t
B 1
H SWA S
B l
G M
E i
l r
CU l
C A
E E
U E
G T
R C
N l"
l S
l O
M S
D S
A F
O R
NE OT o
3 IP 2
4 l
1 TA 5
AH rC f
1 3
6 7
9 l I HH E.
EP LE l
l P
i D
h E
1 I
I C
6 S
1 m.
B L
)
I A
A 5
T H
I
)
7
. 4 1
)
1 E
T l
I 1 Ff A
R A,
A
,X i
lW O
I" A)
))
I )
I
) I A
S PI
. l I
Tl T3 LD l
NI 7E
I S
OC 1 C I
I I AI A
I E I
I 5T TT PT P,
TP R
SR
,D R
( R
( 2 RP A
I A 0N
,A AA A
AA 1
S VS 1 E 4P PP 3P 9
P S
ES P
(
((
(
( W E
RE
,P
,I E
C C
1A
' 9 56 25 8I 6N 1
L A
N C
C O
I M
N I
I M
D E
O T
N S
E N
,T I
S O
R
,A MS T
D I L I
l l
E CM EY A
NS LA T
DV E,
TS T
AM I C A
NN
,TH S
M N
E BI C
,T AN I
C SNT Tf SS BH F
D S
TSSM I S I
MY OC I
I NR S
YE RLY SR ES RE T
OE YLST EOS T
T PT R
N I W SO S
TO
&SS SE E
O TO RGY ACG NM YT C
I PP rrNS W
I f WI DEI G
T A
ENS N
f I"
R,
OLE ERL I
.S GW SEN G
I CS A.C P L EOO SSY N
YMO t
I R
Sf O
MC FCO ErS I E LSI S
C Er OEAY C
Df LV AST E
S DD C
- a. SC YY C L D
NEA D
NM ASC f
I E
N E
CCN GNO i
D LE RLSR NNW NER A
SI E
F YAF V
E RI TVEN GGD DNN O
T I
I EU C
CO Rf LAO LI EKC E
t NQ ADOR 0 E f I RC ED FSE C
EM ENRO 9
R I A AI P E
GB HAPB 9 MI UR I
E S.
BA&
FR SRS H
1 A
B C
D E
F G
I Il
~
G C
N I.
I S
Z IO N
I P
O S
I A
T T
l i
N A
P F
L S
M D
U E
G I
l C
E G
C R
W A
D E
4 T
N I
E N
A V
R E
E E
R S
R V
R U
I A
C E
S S
C S
R R
T U
P N
C N
O R
A T
F F
F 2
H L
O O
O P
E S
e C
N N
NE s
E N
C O
OU s
R A
I I
i S e
U H
T T
TS m
T P
U E
EI u
U 0
L L
L s
F F
O P
PE sa N
S M
MR s
R O
E O
OO s4 1
O C
R C
CC et s
3 F
e?
A
)e I
st si R
O 0
0 0
e!
-n E
et T
wi I
s!
R e
't C
se maes r
, b i
1 iI
I l
l i
l l
l PLAN TO ADDRESS SEVERE t
ACCIDENT ISSUES IN THE C-E i
I DESIGN CERTIFICATION PROGRAM
)
o i
p i
l 1
1 l
i 1
\\
i i
l DR. M. W. CRUMP l
SUPERVISOR, NSSS NUCLEAR DESIGN
?
I l
i I
I l
i l
I.
~
somousnom)ancassanne %
?
i I
l PLAN TO ADDRESS SEVERE ACCIDENT ISSUES IN THE COMBUSTION ENGINEERING DESIGN CERTIFICATION PROGRAM l
O PLAN OVERVIEW O
DESIGN EVALUATION FOR SEVERE ACCIDENTS
?
o PROPOSED PROCESS FOR RESOLUTION OF I
SEVERE ACCIDENT ISSUES l
o SCHEDULE j
o
SUMMARY
I i
em n l
i
=== = =a4 = = = = = = -
=d
~
i 1
1 1
.j i
PLAN OVERVIEW 1.
OBJECTIVES 1
2.
APPROACH d
i 3.
SuffiARY OF ARSAP SUPPORT 2
4 l
I I
i 4
i, i
1 1
i I
)
i i
1 i
l 1
1 L...
l
~
l l
\\
1.
OBJECTIVES o
COMPLY WITH NRC SEVERE ACCIDENT POLICY CRITERIA FOR FUTURE PLANTS, DEMONSTRATING DESIGN ACCEPTABILITY OF SYSTEM 80+ WITH RESPECT TO SEVERE ACCIDENT (DEGRADED 1
,i CORE) ISSUES o
PROVIDE FEEDBACK FROM SEVERE ACCIDENT ANALYSIS TO THE DESIGN OF SYSTEM 80+
e I
j I
i i
i i
I
1 4
J l
1 1
j 2.
APPROACH
~
APPLY DETERMINISTIC - PROBABILISTIC METHODS o
i l TO ESTIMATE SEVERE ACCIDENT RISK IN SYSTEM -
l 80+ PLANT DESIGN o
DEVELOP TECHNICAL POSITIONS FOR RESOLUTION OF g o
ALL RISK SIGNIFICANT SEVERE ACCIDENT ISSUES i !
.l I
l 0
OBTAIN HRC STAFF REVIEW AND CONCURRENCE ON
(;
q
-)
TECHNICAL POSITIONS PRIOR TO CERTIFICATION
.i y
I REVIEW l
o USE DOE ADVANCED REACTOR SEVERE ACCIDENT PROGRAM (ARSAP) AS A MAJOR TECHNICAL RESOURCE I
i k
t 0
i i
i l
I i
sosasusnos)ENCNGEERfMG v ~~ ;
i 3.
SUW.ARY OF ARSAP SUPPORT o
DEVELOPMENT AND APPLICATION OF SEVERE I
ACCIDENT RISK ANALYSIS METHODS DETERMINISTIC METHODS (MAAP)
PRA TOOLS AND GUIDANCE j
o RESOLUTION OF SEVERE ACCIDENT ISSUES IDENTIFY, CATEGORIZE, PRIORITIZE ISSUES WRITE TOPIC PAPERS ON ISSUES INCORPORATE INDUSTRY /NRC FEEDBACK l
PERFORM ASSESSMENTS FOR ISSUE RESOLUTION i
j o
SYSTEM 80+ DESIGN EVALUATION l
RECOMMEND DESIGN FEATURES FOR EVALUATION REVIEW MODELS AND ANALYSIS RESULTS PERFORM SUPPORTING ANALYSES FOR PRA INCORPORATE RESULTS IN TOPIC PAPERS i
I l
L.....
..... _ s
-m____
-._m._______
.s
]
)
i i
SYSTEM 80+
i i
DESIGN EVALUATION FOR SEVERE ACCIDENTS J
a REFERENCE DESIGN o
ENHANCEMENT FEATURES l
o ANALYSIS OF SEVERE ACCIDENT RISK P
i 5
1 o
LOGIC FOR DESIGN EVALUATION s
\\
i !
1 I
l f
i 1
i 1
L...
...~
i, SYSTEM 80 REFERENCE DESIGN 1.
SYSTEM 80+ WILL BE A EVOLUTIONARY ADVANCED DESIGN BASED ON THE SYSTEM 80 STANDARDIZED DESIGN l
l l
2.
PRACTICAL ADVANTAGES OF USING SYSTEM 80 AS DESIGN STARTING POINT f
a MEETS ALL CURRENT REGULATIONS (FDA GRANTED IN 1983) i O
PROVEN DESIGN WITH OPERATIONAL FEEDBACK l
(PALO VERDE NUCLEAR GENERATING STATION) o BASELINE PRA NEAR COMPLETION l
0 MAJOR SYSTEM 80 FEATURES ARE CONSISTENT WITH INDUSTRY ALWR REQUIREMENTS AND WILL BE MAINTAINED i
l
~
commesnom):memesmus
SYSTEM 80+
j l
ENHANCEMENT FEATURES j
)
I.
REACTOR t
1.
15% OVERPOWER MARGIN 2.
MANEUVERING CONTROL WITHOUT SOLUBLE BORON l
3.
RING FORGED REACTOR VESSEL l
j 4.
REDUCED TH i
i II.
l i
1.
LARGER PRESSURIZER h
i 2.
10% SG TUBE PLUGGING MARGIN o
3.
INCREASED SECONDARY INVENTORY d
{
4, 1000 PSI STEAM f
1 III. SAFEGUARD SYSTEM l
j i
~
1 1.
4 TRAIN SAFETY INJECTION d ~ ~~
2.
DIRECT VESSEL INJECTION WITH 1 PUMP / TRAIN 3.
INCONTAINMENT REFUELING WATER STORAGE TANK 4.
4 TRAIN EMERGENCY FEEDWATER SYSTEM 5.
SAFETY}RESSURESHUTDOWNCOOLINGSYSTEM J PRESSURIZATION SYSTEN HIGHER L i
6.
l I'/.
AUXILIARY SYSTEMS i
1 1.
NON-SAFETY CVCS i
2.
CENTRIFUGAL CHARGING PUMPS
?
l 3.
TWO STAGE LETI)0WN V.
INSTRUMENTATION AND CONTROL SYSTEMS i
1.
ADVANCED CONTROL CENTER (NUPLEX 80) l 2.
ADVANCED DATA COMJNICATIONS 3.
IMPROVED CONTROL SYSTEM PERFORMANCE i
AND RELIABILITY 4.
MICR0PROCESSER BASED COMPONENT CONTROL SYSTEM coassurTION)ENceNEERING
I l
SYSTEM 80+
ANALYSIS OF SEVERE ACCIDENT RISK Il i
1 1.
DETERMINISTIC ANALYSIS 4
2.
PROBABILISTIC RISK ASSESSMENT (PRA) 1 i
1 L
l
\\
I i
1 1
somuusTion)smanessasses
h i
1 i
i i
1.
DETERMINISTIC ANALYSIS g
3 O
APPLICATION TO DESIGN EVALUATION l
\\
CALCULATE SYSTEM 80+ PLANT RESPONSE FOR
-l j
DOMINANT CORE DAMAGE ACCIDENTS
^
i k
l l
CALCULATE EFFECTS OF DESIGN CHANGES l
j l
i ON PLANT RESPONSE h
o RESULTS OF CALCULATIONS F
J 4
i I
EXTENT OF CORE DAMAGE i~' l~
l
[l l.
FISSION PRODUCT RELEASE IN CONTAINMENT i
l CONTAINMENT LOADING R
l 1
1 l
RESPONSE OF CONTAINMENT AND ESSENTIAL i
1 EQUIPMENT EX-CONTAINMENT SOURCE TERM
(
l i
1 I
COM54fSTIONhENC8MEERING
~ f 1
4 l
I 1.
DETERMINISTIC ANALYSIS (CONT.)
o MAAP CODE (VERSION 3B) PROPOSED AS MAJOR TOOL f
FOR DETERMINISTIC ANALYSIS MODULAR ACCIDENT ANALYSIS PROGRAM (MAAP) f ORIGINALLY DEVELOPED AND USED FOR IDCOR MAAP 3B INCLUDES MAJOR MODELING IMPROVEMENTS IN RESPONSE TO IDCOR REVIEW RESULTS AND ONGOING RESEARCH PROGRAMS l
l EPRI USERS GROUP FORMED f
C-E HAS EXPERIENCE WITH EARLIER MAAP VERSION 0
ARSAP SUPPORT FOR MAAP ANALYSIS j
CONTINUED DEVELOPMENT OF CODE AND DOCUMENTATION l
PROVIDE USER TRAINING i
l ASSIST WITH SYSTEM 80+ MODEL DEVELOPMENT REVIEW C-E CALCULATION RESULTS FOR CONSISTENCY, VALIDITY j
PRESENT DETAILS OF MAAP METHODOLOGY TO I
i NRC AND RESPOND TO QUESTIONS 1
l 6.....
3.
O
2.
PROBABILISTIC RISK ASSESSMENT (PRA)
I o
APPLICATION OF PRA TO DESIGN EVALUATION l
ESTIMATE BASELINE CORE DAMAGE FREQUENCY IDENTIFY DOMINANT CORE DAMAGE CONTRIBUTORS
~
~~I (INITIATING EVENTS AND ACCIDENT SEQUENCES)
ESTIMATE RISK REDUCTION BENEFIT OF DESIGN l
CHANGES l
i l
o PRA DEVELOPMENT FOR DESIGN VERIFICATION I
LEVEL 1 (BASELINE) PRA FOR A GENERIC SYSTEM 80 PLANT H -'
LEVEL 1 PRA FOR SYSTEM 80+ EVALUATION LEVEL 2 PRA FOR FINAL SYSTEM 80+
PLANT DESIGN i
o ARSAP SUPPORT FOR PRA EXTERNAL EVENTS ANALYSIS E '~?'
RISK QUANTIFICATION FOR A STANDARD PLANT ~ SITE :
REVIEW C-E PRA RESULTS FOR COMPLETENESS
~~
AND CONSISTENCY WITH OTHER PRAs coassesTson)sucemasasme
l
{
r=
K 0
S E8 INT E
1
?
TN N
ROS D
s IO NEE e
NA IM LA ETC GGI Y
El 2
LEA ER ZC INF MC 2
i ETR 1
LDS EHT Li SAI E
AY L
AEV DCS PG
=
l BS NR U
MI A
J ISED
=
^
o N
8e Y
~
N
~
0G 8I) 7 SR ED o
MEA SE N
TNV N
EDS n:
4 O
T S
NOO I
SSE APR m
^
T YSC LSP A
S S (-
PEM
^
U RI U
N
~
L
~
A
/ '-
M V
~
m E
x
~
N
=
G I
S N
ED R
OF T
N NOD C
N tG ATI I
G 0I L
C G
I O
S) 8S PEC G
L EE E
SA N
DK 3e MD EN s :
8 E
ZOE R
U PD TT YPR O(
SN LSE E
B YA AEV E
SL NRE M
P A
S f
G M
=
5
)
e l
=
s o
r S
T N
R s
T GR N
E u
Y E
ION P
R M
SFO e
A E
I THE L
P a
SWR DET r
'8 4I I I e ULI RA DAU TUU C
oc I
N Q
CTL P
I E
EAA R
LEV O
EFE T
- b S
Ill l!
i,e 1
l I
\\
?
i PROPOSED PROCESS FOR RESOLUTION OF J
1 SEVERE ACCIDENT ISSUES 1.
PROCESS PROPOSED TO NRC BY ARSAP l
1 2.
TOPIC PAPERS r
l 3.
EXPECTED FORMS.0F ISSUE RESOLUTI0ff 3
1 4
4.
RISK SIGNIFICANT ISSUES GlVEN PRIORITY l
5.
LOGIC 1
1 l
)
I l
f I
O l
8 8
- i 9
i 4OM5USTIONhENGINEERING m:
--am n m
l l
1.
PROCESS PROPOSED TO NRC BY ARSAP AT JULY 16 AND AUGUST 3, 1987 MEETINGS o
C-E WILL " SPONSOR" ARSAP TOPIC PAPERS FOR SEVERE ACCIDENT ISSUES RELATED TO SYSTEM 80+ CERTIFICATION Y
]
o-C-E/ARSAP WILL PROPOSE IDENTIFICATION OF
)
ISSUES, TOPIC PAPERS, AND SCHEDULES FOR 2
PRESENTATION TO NRC g
l t o
C-E/ARSAP WILL PRESENT TOPIC PAPERS ON CESSAR l
DOCKET FOR NRC REVIEW l
[
o NRC TO SUPPORT REVIEW AND RESOLUTION OF I
ISSUES b
....on )....
.J
~
u
J i
i 4
l l
2.
TOPIC PAPERS WILL
'l o
DEFINE ISSUES AND IMPORTANCE TO SAFETY 0
PRESENT TECHNICAL STATUS 0
PROPOSE REQUIRED ACTIONS FOR ISSUE RESOLUTION f
1 i
3.
EXPECTED FORMS OF ISSUE RESOLUTION l
i i
o ISSUE RESOLVED WITHOUT FURTHER ACTION i
i
-i 1
a' ISSUE RESOLVED BY ACTION, SUCH AS IMPLEMENTING A DESIGN FEATURE OR
~
- l PERFORMING OTHER CONFIRMATORY ANALYSIS 1
o ISSUE RESOLVED BY ACTION BUT SUBJECT T0 p
1 CONFIRMATORY R&D i
}
i
\\
i I
-A....
w '
~
1 I
4.
RISK SIGNIFICANT ISSUES HAVING POTENTIAL DESIGN IMPACT FOR SYSTEM 80+ WILL BE GIVEN PRIORITY i
o CONTAINMENT PERFORMANCE o
DIRECT CONTAINMENT HEATING BY EJECTED CORIUM f
a EX-VESSEL HEAT TRANSFER FROM CORIUM TO CONCRETE o
DEBRIS C00 LABILITY j
0 HYDROGEN GENERATION, IGNITION, AND BURNING i
)
i
%... mi..>.
J
1 l
I l
j PROPCSED PRCCESS TCR RESCLUTION OF SEVERE ACCIDENT ISSUES 1
1 l
1 l
I DEFINE,CATEGCRIZE l
_.i.....PRICRITIZE ISSUES.
l
)
)
. ASSEMBLE BEST AVAILABLE INFC.
CCNTINUCUS I
,. TRACFJ:NG 1
CF, S / A.. INFC..;.:
q PRODUCE l
l ARSAP POSITION i (TCPIC) PAPER n
{
'~
,.8 v
~
C
/ NDU TRY b
]
MCDIFY PAPER
/
i o
i 5
ARSAP/C-E f
j PRESENT l
[
l i
1 TCPIC PAPER RECCMMEND l
d P '~ ~ ~~ ~ '
TC NRC CESIGN 1.
J EVALUATION 7-No 1 l 4
CBTAIN
~ CONDUCT ~~I ~
L-:Yes ANALYSIS /
NRC LETIER i
. l...CF AGREEMENT ASSESSMENT g
..g.
y Yes j
l 1
o ISSUE CNFIRMATCR RESCLUTICN RESCLV?.D7 R&D REQ. ANALYSIS /
l l
No REQUIRED 7 No ASSESSMENTS' No 4
i.'
l Yes Yes
'~~~T NRC WRITES SPECIFY I
POSITION PAPER, CONFIRMATORY NUREG(S)
R&D REQTS.
)
j l
j ACTICH BY NRC
.[
l a
1 i
"a J
l L...
.n 1
i SCHEDULE FOR SYSTEM 80+ DESIGN EVALUATION q
MILESTONE TARGET t
COMPLETE BASELINE PPA 9/87
~CCMPLETE INSTALLATION AND TRAINING FOR MAAP 3B 11/87 COMPLETE REFERENCE MODEL FOR SYSTEM 80 2/88 COMPLETE DESIGN FEATURE EVAll!ATIONS FOR SEVERE ACCIDENTS 9/88 i
FINALIZE DESIGN 10/88 COMPLETE LEVEL 2 PRA 6/89 I
40mSU5710NhENGINEERING -
i l
SUMMARY
d i
o PROPOSED PLAN ADDRESSES THE NRC SEVERE ACCIDENT POLICY i
(
STATEMENT CRITERION FOR COMPLETION OF NRC REVIEW OF NEW
~
f H
."i PLANT DESIGNS FOR SAFETY ACCEPTABILITY, EMPHASIZING 4
DEGRADED CORE ISSUES i
d a
o SYSTEM 80+ CERTIFICATION WILL ESTABLISH PRECEDENCE FOR l l
)
REGULATORY REVIEW 0F ALWR SEVERE ACCIDENT ISSUES.
4 THEREFORE, DOE ARSAP HAS AGREED TO PROVIDE MAJOR DIRECT SUPPORT FOR gy SEVERE ACCIDENT RISK ANALYSIS
~
I RESOLUTION OF SEVERE ACCIDENT ISSUES I
I o
NRC STAFF TO PROVIDE AGREEMENT ON ANALYSIS METHODOLOGY (MAAP)
T PRUCESS FOR ISSUE RESOLUTION l-LIST OF ISSUES i
CRITERIA FOR RESOLUTIOR OF ISSUES'
~
I a
commemon) memaanma
\\.
1
.h*
ENCLOSURE 3 ADVANCED REACTOR SEVERE ACCIDENT PROGRAM q'ARSAP)
SUPPORT FOR CE CERTIFICATION i
M. H. FONTANA
~
P.M.HAAS R. L. SUMMITT PRESENTED TO THE
~
~
U. S. NUCLEAR REGULATORY COMMISSION AUGUST 12,1987 m
DOE ARSAP OBJECTIVE Assist in the early resolution of risk significant severe accident issues so that they will not be major issues for evolutionary advanced light water reactors during the 1990's s
W 9
e
^
a DOE ARSAP MISSION ya 1.
Support vendors in DOE design verification program by severe accident assessment and analysis.
o Design Evaluations o
Safety Analysis Report input / Preparation o
Defense Before NRC j
2.
input to, review, and support EPR'l requirements document in sbvere l
accident areas, i
q 3.
Achieve more genenc resolution of severe i
accident issues -- than may occur through support of vendor certification or of EPRI requirements document -- such that they
~
are not issues for ALWR's.
\\
i 1
e-1 i
I h
4
c y
/
ARSAP PARTICIPANTS DOE: Supporting Agency EG&G Idaho: Prograrn Manager IT Corporation: Technical Director Fauske & Associates: Subtler Contractor for Analysis Methods Development IEAL: Subtier Contractor for Regulatory issues
(
~
I O
e N
ARSAP SUPPORT TO INDUSTRY l
Development and Application of Severe o
Accident Analysis Mett. ads t
Resolution of Severe Accident issues o
Direct Support to ALWR Vendors o
\\
Support for the EPRI ALWR Requirements o
Assess Regulatory Compliance Alternatives o
and Address Severe Accident Relevant l
Certification issues I
e 6
--_h
TOPIC PAPER SETS 3
Set 1.
Resolved IDCOR/NRC issues -
Applicabilityto ALWRS Set 2.
Plant Response Under Severe Accident Conditions Set 3.
Probabilistic Methods Set 4.
Risk Reduction Methods Set 5.
Risk Results Set 6.
Applications of Methods l
e l
TOPIC PAPER Set 1: RESOLVED IDCOR/NRC ISSUES -
APPLICABILITY TO ALWRS (Single Paper)
I o
Reactor coolant system natural circulation (IDCOR issue 2) o in-vessel steam explosions and alpha mode failure (IDCOR issue 7) o Ex-vessel heat transfer models from molten core to concrete (IDCOR issue 10) o Fission product release prior to vessel failure-(IDCOR issue 1) o Release model for control rod materials (IDCOR issue 3) o Fission product and aerosol deposition from primary system (IDCOR issue 3)
O a
b
TOPIC PAPER-SET 1: RESOLVED IDCOR/NRC ISSUES -
APPLICABILITY TO ALWRS (2) i o
Ex-vessel fission product release (during core
- concrete interaction.s) (IDCOR issue 9)
Fission product and aerosol deposition in o
containment (!DCOR issue 12)
~
o Amount and time of suppression pool bypass (IDCOR issue 13a)
Revaporization of fission products (IDCOR-11):-
o
~
o Secondary containment performance (IDCOR lssue 16) (Resolved by design)
)
o Modeling of emergency response (IDCOR issue 14) e 9
TOPIC PAPER SET 2: PLANT RESPONSE UNDER SEVERE ACCIDENT CONDITIONS o
in-vessel hydrogen generation (IDCOR issue 5) o Core melt progression and vessel failure (IDCOR issue 6) o Direct containment heating by ejected core materials (IDCOR issue 8) o Containment performance (capability, failure modes, isolation, bypass) (IDCOR issue 15) o Hydrogen ignition and burning (IDCOR issue 17) o Fission product release during high pressure core ejection i
G
TOPIC PAPER SET 3: PROBABILISTIC METHODS External events -- seismic (Fire and flood o
resolved by design Human factors -- required operator actions o
o Human factors -- unexpected operator actions
~
with potentiai adverse effect o
Human factors -- quantification of human error probabilities
~
Success criteria -- partial success and mission o
time o
Common cause failures o
identification of dominant sequence.s i
i
TOPIC PAPER SET 4: RISK REDUCTION MEASURES o
Essential equipment performance (IDCOR issue 18) o Severe accident management -- plant equipment /information system capability o
eevere accident management -- conditions for safe stable states o
Mitigation features
TOPIC PAPER SET 5: RISK RESULTS Consensus on integrated severe accident o
1 analysis code capability, validation, and application 4
I Safety goal implementation --interpretation of o
goals and usage of PRA results in comparison.
with goals, including interpretation of 4
uncertainties o
Uncertainties in plant risk -- effects of system analysis uncertainties o
Uncertainties in plant risk -- effects of uncertainties in severe accident analysis (Phenomenology, plant damage states, methodology) o Uncertainties in plant risk -- treatment of-
~ ~ ~
propagation of uncertainties
~ ~~
Uncertainties in plant risk -- completeness of o
choice of sequences and cutoff probabilities
TOPIC PAPER SET 6: APPLICATIONS OF METHODS o
Effect of severe accident issues on regulations -- probabilistic accident design bases o
Effect of severe accident issues on l
regulations -- assessment of regulatory compliance alternatives o
Effect of severe accident issues on regulations -- effectiveness of technical specifications e
SUMMARY
CE/ DOE /ARSAP are committed to complete o
resolution of severe accident issues, consistent with the NRC Severe Accident Policy Statement,in support of certification of evolutionary advanced light water reactors NRC is requested to provide the resources in -
o licensing and research to assure that seve.re accident issues are resolved consistent - in technology and schedule-with ALWR certification j
o NRC is requested to participate in high level management discussions with CE/ DOE /
ARSAP as appropriate to review progress, apply midcourse corrections as necessary,
~
and assure that program objectives are be.ing met
~
j
e i
e APPENDIX ARSAP PROGRAM STRUCTURE AND STATUS l
~
j O
e-
ARSAP PROGRAM STRUCTURE The ARSAP Program is designed to support resolution of severe accident issues, keyed to the NRC Severe Accident Policy Statement WBS 1. Management and Planning Policy Development 4
Program Management-Organization Interfaces WBS 2. Severe Accident issue Resolution _
Issue identification
~
Topic Paper Development Issue Resolution and Documentation WBS 3. Severe, Accident Analysis Methodology Selection and Development of Analysis Tools
~
?~
Revise of Experimental Data and.
Recommendations
. -~
~
Benchmarking and Validation and Verification
ARSAP PROGRAM STRUCTURE (2)
WBS 4. PRA. Methodology and Application Criteria i
Guidance for use of PRA in Design Guidance for Regulatory Applications of PRA Development of EPRI Functional PRA's l
Development of EPRI PRA Ground Rules and Assumptions WBS 5. Regulatory Compliance Alternatives Support of Risk Based Assessments of ineffective Requirements Proposal of Alternatives Support of Vendor /EPRI Efforts with i
NRC WBS 6. Certification Process Development and Support
~
Regulatory Assistance to Vendor Certification Efforts WBS 7. Lessons Learned Notebook Annual Updates
ARSAP PROGRAM STRUCTURE (3)
WBS 8. EPRI Requirements Document Review, Comment, and input to EPRI Requirements Document in Severe Accident Areas WBS 9. EPRl/ Vendor 600 MWe Plant Design and
~
/
Certification 1
GE Desig.n (SBWR) Review Westinghouse Design (SPWR) f Review WBS 10.CE System 80 Plant Design and Certification Review of Severe Accident issues Relevant to CE Design Definition of Severe Accident i
Sequences for Analysis Performance of Severe Accident
.q 2
- ' ' ~ " ~
~
Analysis Support to System 80 PRA Assessment of Severe Accident Mitigation Features Support of Final Design Approval and NRC Review Process
ARSAP PROGRAM STRUCTURE (4) l l
WBS 11.GE ABWR Plan Support MAAP User Support Assistance in Developrnent and Applications of MAAP Model Analysis of Fission Product Particle Size --
Effect on Suppression Pool DF Support for PRA Seismic Analysis Support for ABWR ATWS Analysis WBS 12. Westinghouse APWR Plant Support Monitoring of Activities WBS 13.Chernobyl Related issues e
4
' ARSAP STATUS (2)
White Papers Written and Under Review for o
Regulatory Compliance Alternatives and For Certification Process Development Report on IDCOR Lessons Learned Published o
o First Draft of Lessons Learned Notebook Under Review o
EPRI ALWR Requirements Document input Provided for Chapter 1 and 5 3
1 Requirement of Document Chapters 1,3,4,5,
]
o Reviewed; involved in planning of Chapter 6 o
Provided PRA Related input External Events (Seismic)
Ground Rules and Assumptions l
Functional PRA Models Containment Capability and Mission T_ime
~
l l
k I
ARSAP STATUS (3) o Agreements in place to Support CE System 80+ Certification Review Severe Accident issue Relevant to CE Design Define Severe Accident Sequences for Analysis Perform Severe Accident Analysis Assess Severe Accident Mitigation Features Support Final Design Approval and NRC Review Process o
Defined Support for ABWR Certification Provide MAAP User Support Develop MAAP Model for Secondary Containment Determine Aerosol Particle Size and Effect on Suppression Pool DF Support PRA Seismic Analysis Support Support ABWR ATWS Analysis Support o
Participated in U.S. Involvement in IAEA Chernobyl A.ctivities
1 ARSAP STATUS o
Program Pian and implementation Plan Complete o
Interfaces with EPRI and Vendors Defined j
and Agreed I
o IDCOR/NRC issue Resolution Experience Summarized With Relevance to ALWR's Major Severe Accident issues identified, Topic' o
l Papers and Schedule Identified o
MAAP and MELCOR Chosen as Prime Analytical Integrated Codes j
o Use of PRA for Design Guidance Being
~
)
Developed o
Severe Accident Groundrules and Assumptions Developed o-
-High Level Functional PRA Models Developed.
4 m.-**
ma-m
'p'*
931(84T7)/SF-1 ENCLOSURE 4 LRAFT Plan to Addre::s Severe Accident Issues in the Combustion Engineering Design Certification Program l
l
)
J i
l 1
August 7, 1987 Prepared by Combustion Engineering, Inc.
931(84T7)/SF-2 W
l 1.
Introduction The design of the next generation of U.S. Light Water Reactors (LWRs) will be required to consider plant response to postulated severe accidents and to demonstrate an acceptably smal'i level of risk to the public and property as a prerequisite for licensing by the U.S. Nuclear Regulating Comission (NRC). This requirement is implied by fundamental criteria put forth in the NRC policy statement on severe accidents. ( }
Plant evaluations for severe accidents, using a combination of deterministic and probabilistic methods, have not been performed in the design process for past plant designs. Consequently, there is a need to define a process for addressing and resolving severe accident issues, including design specific evaluations to identify the impact of design changes in plant response, and select design changes based on quantified cost versus benefit considerations.
Important to this process is a validated and accepted methodology for modeling severe accident phenomena and plant response, and a detailed probabilistic risk assessment (PRA) basis for identifying severe accident vulnerabilities and for evaluating the risk reduction benefit of design changes. The Industry Degraded Core
]
Rulemaking Program (IDCOR) made major contributions to the. current industry state-of-the-art for severe accident analysis. More recently, the U.S. Department of Energy (DOE) has formed the Advanced Reactor Severe Accident Analysis Program (ARSAP) for the purpose of assisting the U.S. industry in implementing the NRC policy statement on severe accidents for new plant designs.
The purpose of this paper is to propose a process to be used by Combustion Engineering for addressing severe accident issues for certification of the system 80+ standardized design.
The process ircludes major technical support activities which will be provided by ARSAP, the methodology proposed to severe accident analysis, and a means for providing positions for severe accident issues for NRC review and approval.
031{8417)/SF-3 hDAT Gi II.
Design Certification Program Combustion Engineering is conducting a program to develcp the System 80+
standardized PWR design and to obtain cerM fication of the design through the rulamaking process provided by the U.S. NRC revised Standardization Policy.
It is the intent that C-E's System 80+ design will comply with the high level design criteria and requirements of the Electric Power Research Institute (EPRI) Requirements Document, and with the criteria of the NRC policy statement on severe accidents for new plant designs.
The EPRI Requirements Document is being developed under an industry-wide program conducted by EPRI to develop requirements for the next generation i
of U.S. Advanced Light Water Reactor (ALWRs).
The industry ALWR requirements have as major objectives improving substantially the level of safety to the public and property, protecting utility investment through high equipment and systems reliability, and assuring
)
licensability by compliance with all applicable regulatory requirements, j
including resolution of NRC Unresolved and Generic Safety Issues. The j
development of the EPRI Requirements Document is being performed by industry contractors including vendors, architect-engineers, and consultants under the guidance of a Utility Steering Committee, with review being conducted broadly within the U.S. industry and by the NRC.
The NRC policy statement on severe accident establishes additional criteria which will be applied for licensing review of new plant designs, as discussed in Section III below.
The U.S. Department of Energy has established the DOE Design Verification Program to demonstrate the implementation of new plant design requirements for U.S. PWRs. As PWR contractor for this program, C-E is receiving support for certification of the System 80+ design. The basic approach for implementation of new requirements in the C-E design is described below.
931(84T7)/SF-4 o
Oh-.
1.
Reference Design C-E has chosen to use System 80, as defined in CESSAR-F, as the base design for the Design Verification Program because it is considered by j
C-E to be a state-of-the-art design containing many advanced design features identified by the EPRI Requirements Document.
System 80 has j
already received a Final Design Approval (FDA) by the NRC, and is currently operating at the Palo Verde Nuclear Generating Station (PVNGS).
Therefore, as a starting point for the advanced design, it has the benefits of complete design detail, meeting all current NRC regulations, i
and having operational feedback.
The System 80+ will incorporate improved design features based on the EPRI Requirement Document, operational feedback, and applicable regulatory requirements for new designs. The design changes will be implemented by submittal of a series of CESSAR-F change packages for NRC review. The sequence of change submittals will follow the planned submittal schedule of EPRI Requirements Document chapters for NRC review.
In this manner the EPRI Requirements can be directly addressed in the j
System 80+ design, including identifying and justifying cases where exceptions will be taken to the EPRI Requirements.
The paced submittal schedule for CESSAR-F revisions is also intended to facilitate the NRC review by limiting the scope of changes in each submittal package and highlighting the incremental effects of dienges relative to the reference design.
2.
Balance of Plant The scope of the current CESSAR-F is limited to the Nuclear Steam Supply System (NSSS), with the balance of plant (B0P) represented by interface requirements and reference to the applicant's SAR.
Complete BOP designs exist for the three System 80 units at PVNGS.
BOP designs have also been developed for other proposed System 80 units including Washington Nuclear Plant-3 and the Cherokee-Perkins units.
b k
931(84T7)/SF-5 The Duke Power Company, which was architect-engineer for the Cherokee-Perkins project, is serving as subcontractor to C-E for tha DOE Design Verification Program.
The scope of the Design Verification Program has been extended to include major 80P design, including emergency feedwater, control room, and containment systems.
These systems will be added to the scope of CESSAR-F.
In addition, present interface requirements for other B0P systems will be replaced by more detailed, standardized functional requirements.
This has the advantage of defining the B0P systems which incorporate major safety related features of the EPRI Requirements Document, and enabling more detailed evaluation of plant performance for design basis events and for response to severe accidents.
The participation by Duke Power Company also provides the benefit of knowledge gained as principal contractor for the DOE Constructability Program and as 1 co-contractor with C-E for the EPRI ALk'R Program.
III. NRC Policy Stateinent The NRC policy statement on severe accidents re existing plants is contained in NUREG-1070.(1) garding future designs a The statement includes the conclusion that existing plants do not pose an undue level of risk to the public and, therefore, proposes no imediate action on generic rulemaking or other regulatory changes to deel with severe accident risk in existing plants, It recognizes, however, that the level of safety may be improved at reasonable cost in new designs, making use of advances in basic knowledge and anasytical methods for severe accident analysis and lessons learned from existing plant evaluations. Safety improvement is viewed as essential both for near term public acceptance and utility investment in new plants and for long term public success and confidence in their operation. The fundamental criteria put forth as n:aessary to show that new designs are acceptable with respect to severe accident concerns are:
A.
Demonstration of compliance with the procedural requirements and criteria of current regulations;
931(84T7)/SF-6 i
B.
Demonstration of technical resolution of all applicable Unresolved Safety Issues and the medium and the high-priority i
Generic Safety Issues; C.
Completion of a Probabilistic Risk Assessment (PRA)-and i
consideration of the severe accident vulnerabilities the PRA exposes along with the insight that it may add _to the assurance of no undue risk to the public health and safety; and D.
Completion of a staff review of the design with a conclusion of safety acceptability using an approach that stresses deterministic engineering analysis and judgement complemented by PRA.
The policy statement also addresses the applicable criteria for approval or certification of reference designs including those previously granted a Final Design Approval (FDA).
In this case reference design applicants must request that their FDAs be amended to permit their designs to be referenced in new Construction Pennit (CP) and Operating License (0P) i applications.
The request must either include the information needed to satisfy each of the criteria stated above, or provide suitable interface requirements to ensure that CP and OL applications referencing the design will satisfy each of the criteria.
For the Combustion Engineering Design Certification Program it is intended that amendment to the existing FDA will be applied for by providing the information needed to satisfy each of the criteria, including completion of the severe accident review.
Further, since the expanded scope of System 80+ will cover a greater portion of the complete facility design, it is intended that a comprehensive PRA will be completed for the System 80+ plant design, in order to enable a one-step CP/0L licensing process without the requirement for the CP/0L applicant to submit a plant-specific PRA.
The policy statement allows'the applicant to choose the option of Design Certiffr.ation by undergoing a rulemaking proceeding in addition to the above criteria and procedural requirements.
System 80+ will undergo such a proceeding, although the procedural requirements are uncertain at this time.
931(84T7)/SF-7
'e o
IV.
Design Evaluation for Severe Accidents 1.
Deterministic-Probabilistic Approach The approach for evaluation of severe accidents in the System 80+ design will stres: deterministic engineering analysis and judgement, complemented by PRA. This is. consistent with the NRC approach for severe accidents, as described in Reference 1.
Deterministic methods will be used for engineering analysir, of plant safety performance, the likely response of the plant to the postulated core-melt accidents, and the perfomaNe.
potential of candidate design features under evaluation (e.g.
for mitigation of severe accident consequences). The role'of PRA in this evaluation will be to aid in identification of accident sequences to be analyzed, and as a tool for estimating risk reduction benefit.1 The results of this evaluation process will be used in a process for resolution of severe accident issues as part of the design System 80> plant design, and to demonstrate safety acceptability consistent with Criterion D in Section III above.
2.
Accident Sequences Leading to Core-Melt The starting point for calculation of core damage frequency for the System 80+ severe accident evaluation will be a baseline Level 1 PRA for the existing System 80 NSSS design.
The results of the Level 1 PRA will provide calculated core-melt probabilitiesforpostuistedaccidentsequencesspecificiothe current System 80 designi, and will therefore contributs to i
initial insights on severe accident vulnerabilities for the System 80+ plant design.
It will also provide the starting point for a PRA model for the System 80+ plant design.
The baseline PRA model will be extended to incorporate NSSS design E
931(84T7)/SF-8 l
hi changes derived from the EPRI Requirements Document.
These changes will include major modifications to the Safety Systems which are expected to contribute to significant improvement in the overall level of safety compared to the existing System 80 design.
Examples include a four-train emergency core cooling system, a manual safety depressurization system, and location of refueling water storage tank in containment.
l The PRA model will be modified to incorporate these changes for System 80+, to estimate the benefit for postulated accident sequences, and to identify dominant accident sequences leading to core-melt for the System 80+ design.
3.
Consequences of Core-Melt Accidents q
Dominant accident sequences leading to core melt will be analyzed with deterministic methods to estimate the accident
)
progression and consequences including potential containment failure and release of fission products. The major method proposed for deterministic analysis of severe accidents will be the most recent version of the MAAP/PWR code, which was originally developed by Fauske and Associates. Inc. (FAI) and applied in the IDCOR Program.(2) C-E has experience with use of an earlier version of MAAP in severe accident investigations for pre-System 80 plant designs.(3,4)
The MAAP analysis will model major systems of the System 80+
plant and treat the following major severe accident i
phenomenology as defined in Reference 1:
\\
progression of core melt in the Reactor Coolant System; loading of the containment;
(
h
\\\\-
931(84T7)/SF-9 rl response of the containment and other essential equipment; and fission product and transport in containment.
The results of the MAAP analysis for severe accidents will estimate the extent and severity of consequences in the plant which may threaten public health and safety; in particular, whether or not containment failure is predicted to occur and,.
~
if so, the ex-containment fission product source tenn and time, parameters. The phenomenology of ex-containment fission product transport and consequences will be based upon generic modeling of a standard plant site.
4.
Evaluation of Risk Reduction Benefit The design process for System 8C+ will include evaluat(on of severe accident risk reduction benefit versus cost for design features not already incorporated as a result of inobstry ALW requirements.
This may include, for example, changes to7the reactor cavity design to improve confinement and coolability of core-melt debris.
The IDCOR program evaluated risk reduction benefit for numerous design feitares as potential backfits for existing plants.(5) Some of these featurgs, although not cost-beneficialasbackfits,couldbeevaluatedfotbSystem80+
as potential features for resolution of,'ALWR severe accident issues.
Analysis of plant response to tevere accident sequences will be performed using the severe accident methodology (MAAP), for both the reference Systen 80+ design and with the feature under.
evaluation, in order to estimate the potential reduction in consequences. The estimate of risk reduction benefit will be based on the consequences (e.g., source term), the probability of the accident occurring as calculated using FRA,.and the T
I
, c '[f 931(84T7)/SF-10
)
quadtified impact of consequences on the public and property.
The analysi! of risk reduction benefit versus cost will be used as a decision tool for implementing design changes to reduce severe accident risk.
1 i
V.
APSAP Support for C-E Design Certificate cn 1.
General 1
The ARSAP program, initiated,in 1986, and is being directed to support ALWR certification with respect to severe accident issues.(6) The scope of the ARSAP program includes activities to aid the U.S. industry in successfully addressing the NRC a
policy on severe accidents for new plant designs. These activities include.a general role for identifying and resolving AWR severe accidnt issues. Acether major role'is validating and applying severe accident analysis methodologies.
In eddition, ARSAP activities are identified for ' support of individua7. vendors in assessing the adequacy of new plant designs for severe accident safety and activities design certification.
s 2
-Resolution of Severe Accident Issues The resolution of ALWR severe accident issues is of fundamental importance to certification of System 80+.
C-E and ARSAP have proposed a process for resolution ALWR severe accident issues described in Section VI.
'nis process provides a means to identify severe accident itsues, develop an approach for-resolution of each issue, and to conduct issue resolution.
For Leach issu'e which may impact certification.and is of significance to severe accident risk, a Topic Paper will be prep _aredwhichprovidesinformationonthe-issueandproposes an approach for.. resolution.
The Topic Papers will be used to presenttechnicalinformationandapropo$edresolution approach for each severe accident issue to the NRC, and-to seek
931(84T7)/SF-11 concurrence for the resolution approach for each issue.
It is expected that the resolution approach in many cases will identify analysis or other fonn of confirmatory work, and in some cases, design evaluation for System 80+.
ARSAP will then support C-E in carrying through the issue resolution and presenting results to the NRC as part of the certification process.
It is important to the System 80+ design and certification process to identify and develop positions early for ALWR severe accident issues with the potential to impact design. This will enable issues to be addressed directly for System 80+ by 3
performing design evaluations.
ALWR ravere accident issues having potential design impact include:
containment performance; direct containment heating by ejected core materials; ex-vessel heat transfer models from molten core to concrete; debris coolability hydrogen generation, ignition and burning; and mitigation features ARSAP will perform support activities for resolution of severe accident issues for the System 80+ design, including activities to:
recomend design modifications for evaluation; assist in defining criteria for acceptable plant response;
i 931(84T7)/SF-12
+
review deterministic results for consistency and validity; provide cost 'venefit analysis as needed for new design features; and develop supporting technical analyses in severe accident Topic Papers.
3.
Severe Accident Analysis Methodology The development and validation of a severe accident analyses methodology for ALWRs is within the framework of ARSAP. ARSAP I
has chosen to use the MAAP code as the basic deterministic method. MAAP 3 and a successor version MAAP 3B have been
)
developed for IDCOR by FAI by incorporating major modifications to the MM P 2 version developed under the IDCOR program.
MAAP 38, which will be available by about September 1987, is recommended by ARSAP for existing and new plant analysis.
The new model features in MAAP 33 have been provided to address many of the severe accident modeling issues identified by NRC review of IDCOR results and by other severe accident issues research programs.
Technical support activities for the MAAP methodology which will be provided for C-E by ARSAP include:
Transfer of the MAAP source code and documentation, plus consultation as required for successful loading and execution on C-E's NOS/VE system; Training in use of PAAP 38:
931(84T7)/SF-13 "g
Support for introducing p' ant-specific modeling features; and Review of C-E models and analysis results for consistency and validity.
It is expected-that frequent interaction between C-E and ARSAP will be required to fulfill these technical support needs.
Training in use of MAAP 3B may be fulfilled by C-E attendance in a series of formal courses being prepared by ARSAP for the EPRI' sponsored MAAP Users Group, scheduled to begin-in Fall 1987; or, alternatively, by informal training.
4.
Regulatory Acceptance of Severe Accident Methodology Regulatory acceptance of the ALWR severe accident analysis j
methodology based on MAAP has an important' bearing on the ability to justify and defend the acceptability of the plant design for severe accidents during rulemaking.
Experience indicates that early and periodic interactions with the NRC.
staff is the best means to gauge acceptance and adjust methods without seriously impacting the program schedule. Therefore, C-E and ARSAP will:
I Present technical details of the severe accident analysis methodology to the NRC staff, specific to the use-for System 80+
provide responses to questions, resolution of topic papers, and, if necessary, modifications to the j
methodology.
=
931(84T7)/SF-14 gy 5.
Consideration of Probabilistic Risk Assessment Regulatory acceptance of the System 80+ plant design will require that PRA results be provided for demonstration of
)
overall safety acceptability for the final design. A Level 2 I
PRA will be perfortled after completion of design evaluation for severe accident issues and evaluation of the design. Technical support from ARSAP for PRA will include:
providing PRA technoiogy transfer providing external events analysis for the C-E PRA; providing models for calculating ex-containment fission product transport and consequences for a standard plant site; and reviewing PRA results for completeness and consistency with other PRA's 1
VI.
C-E/ARSAP Process for Resolution of Severe Accident Issues The process proposed by C-E and ARSAP to address severe accident issues for the System 80+ design is represented by the logic in Figure 1.
Early activities in this process will identify severe accident issues and propose an approach via Topic Papers for resolution of each issue. ARSAP will provide the lead for these activities with C-E concurrence on' the. -
proposed approach for resolution of each issue.
The resolution approach either will or will not require design evaluation.
If
~
design evaluation is required, it will be conducted by C-E as described in Section IV above, with ARSAP support as described in Section V.
The outcome of the design evaluation will provide confirmation of design acceptability as a basic for
931(84T7)/SF-15 O
Ob.
resolution of the severe accident issue.
If the resolution approach does not require System 80+ design evaluation, then ARSAP will conduct other analysis or confirmatory work required if needed as a basis for resolution of the severe accident issue.
In order to support the design certification schedule, priority will be given to initiating resolution for issues with highest risk significance and potential impact on certification.
Remaining activities in Figure 1 are to conduct review of the Topic Papers and achieve issue resolution.
These activities cover technical assessment for issue resolution and interactions with the NRC to achieve approval of the issue resolution. ARSAP will conduct technical assessment with concurrence by C-E and ARSAP Industry Technical Advisory Group (ITAG) on the technical a proach.
C-E will have primary responsibility for interactions with the NRC.
The resolution positions will be presented to the NRC by submittal of the Topic Papers on the CESSAR document.
VII.
NRC Support for Resolution of Severe Accident Issues The design certification of System 80+ will establish precedence for regulatory review of design acceptability for safety considering severe accident issues.
NRC staff support for the proposed C-E/ARSAP process for resolution of severe decident issues is indicated in Figure I.
NRC activities include review of submitted Topic Papers and provision of NRC j
staff position on issue resolution. After agreement is achieved on the criteria for issue resolution, it is expected
{
that the NRC staff will document it's positions in the fonn of a NUREG or other regulatory document.
931(84T7)/SF-16 NRC staff support required for the overall C-E plan presented in this paper will include providing agreement on:
The analysis methodology (MAAP);
The process for issue resolution; The list of severe accident issues; and c-iteria for resolution of issues.
VIII.
Schedule l
Schedule dates for the conduct of severe accident design evaluation are given in Table 1.
The completion dates are consistent with the design certification program schedule.
IX.
References 1.
NRC Policy on Future Reactor Designs - Decisions on Severe Accident Issues in Nuclear Power Plant Regulation, NUREG-1070, July 1985.
2.
IDCOR Technical Report 16.1A, " Review of the MAAP PWR and 8WR Codes, December 1983.
3.
C.L. Kling, et al., " Evaluation of Hypothetical Severe Accident Consequences for a Representative C-E PWR",
presented at International Meeting on Light Water Reactors - Severe Accident Evaluation, August 28 -
September 1, 1983, Cambridge, Mass., TIS-7464.
4.
C.L Kling, et al., " Degraded Core Considerations for PWRs with Top Mounted In-Core Instrumentation", presented at ANS Winter Meeting, San Francisco, CA., October, 1983, TIS-7509.
931(84T7)/SF-17 gb
~
i 5.
IDCOR Technical-Report 21.1, " Risk Reduction Potential",
June 1985-I
- 6. -
ARSAP, Advanced Reactor Severe Accident Program Plan, June 1986.
J l
1 i
i a
___._.__________m.____.____._J
!4l Mm
.=
ly g
0 g
E8 IIT
?
I 1
ECS ID Em N
s T
L gTC MGI IN e
N VM LEA Y
E E
gC I
P M
B ETR I
yU SAI E
SSP E
LDS DCM EN L
AY L
AEV SS ggR A
M se Y
N 0G 7
3I)
ED N
SR SE 6EA TIeV 3EDS
,g8 030O T S SSE APR LSW YSC PEI SS(
RI N
S eu s
s I
t ne d
T i
N M
NOD
+
ATI C
G 0I L C C
I 8M I
PEC A
, ate -
E ZOE J
r YN T
YPR PD e
S LSE O(
v B
AEV e
S NRE S
P A
S f
o n
o i
tu lo S
T N
1 s
t GR s
e Y E IOs s e
R RtM SFo r
TUE E
I u
r SWR
,e8 DET g
o ULI fRA i
f DAU TtU F
N Q CtL S
I E
EAA s
R LEV EPE e
S cor P
PAS A
io R
b
/E D
/
E S
C O
SR T
IS S
F IE
- N NST d
OYN e
ONI EA IM a
ILE UI RP
_~
8S Y
TAM s
NKA IS UNS o
IC/
EC LE LAS p
TAS CI AS O E O
NR UP NS S
.S r
OTF DO AA EOS P
C O OT REA R
R P
i seY
'o N
?
Y R
R.
D s
o O
?
OS s
N o N
T D
YTT E
sB IT A E FAC et PIA MDR IME oSU REI CRR cEL ISU EI BDA F Q 8
PFD R V E
e Se&
l E
k Y
OR C
oN ES ZE IU O
t Y
R C
e OSSN IE B
TE R
RN E
.R 1
EE 7
s SP TW E B EIA IUA
?. A C NTI D
e EAI R
ITe EE AZEL UO Y
TPS EY N
ILB
,e DPI E WIY 8
1 C. tea I
CB I
.IO N
TLo SL leg 5
s C
B n SO WcE
/. e O C E/F 5
EIsL RPI I
I OCa IS IR Gc NREI PAP ED PPT VE R
E CTU EI IOSA SO O
O EBr R
O RIN RT EtgAA A(
IIS C
DP O
A P
7' I
931(84T7)/SF-19 t
OEdI i
Table 1 j
Schedule Dates for C-E Severe Accident Analysis System 80+ Design Evaluation Date Complete Baseline PRA 9/87 l
Complete Installation and Training for MAAP 38 11/87 Complete Reference Model for System 80+
2/88
)
Complete Design Feature Evaluations for Severe Accidents 9/88 Finalize Design 10/88 Complete Level 2 PRA 6/89 i
Submittal of Topic Paper Sets 4
1.
Resolved IDCOR/NRC Issues 9/87 Applicability to ALWRs 2.
Plant Responst 'V!er Score Accident 11/87 Conditions 3.
Probabilistic Methods 12/C7 1/78 4.
Risk Reduction Methods 3/88 5.
Risk Results 5/88 l
6.
Application of Methods 7/88 l
1 i
i
_ _,