ML20214W241
| ML20214W241 | |
| Person / Time | |
|---|---|
| Issue date: | 07/08/1982 |
| From: | Bernero R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Blond R, Cunningham M, Margulies T NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20213E209 | List:
|
| References | |
| FOIA-87-113, FOIA-87-60 NUDOCS 8706150173 | |
| Download: ML20214W241 (274) | |
Text
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b MEMORANDUM FOR: Those on Attached List l--
' Robert M. Bernero, Director
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i: FROM: l -
Division of Risk Analysis Office of Nuclear Regulatory Research B
SUBJECT:
MINUTES OF NRC/IDCOR TECHNICAL EXCHANGE MEETING,~
JULY 7-8, 1982
[
e i Enclosed for your information are the minutes for the subject meeting held in I Albuquerque, New Mexico on July 7-8, 1982. These minutes describe in brief terms the subjects discussed at the meeting and summarizes the agreements made between NRC and IDCOR.
Complete copies of the viewgraphs p esented at the meeting are now available.
Because of the bulk of the material, the number of copies made has of necessity been limited. However, all organizations with people in attendance at the
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meeting will receive at least one copy. Individuals receiving these copies ,
', are indicated with asterisks- on the attached distribution list. -
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- During the meeting, results were presented (and are included in the enclosed r
l material) from the first phase of Sandia's value-impact studies of risk reduction features. As was emphasized at the meeting, these results are very preliminary and do nct yet reflect competing risk or uncertainty effects. Please treat this material accordingly.
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J lMMf8g3 870MO Robert M. Bernero, Director SHOLLYS7-ds0 PDR Division of Risk Analysis U Office of Nuclear Regulatory Research i
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Enclosures:
.' As stated cc: T. Tyler, TEC-IDCOR i .
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[ MEETING MINUTES NRC/IDCOR TECHNICAL EXCHANGE MEETING
? ALBUQUERQUE, NEW MEXICO JULY 7-8, 1982 V(hj f,;r.>n[ Q _
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- 1. ATTENDANCE -
,; Attendees are listed in Attachment 1.
>. 2. GENERAL DISCUSSION I The meeting was organized and in general conducted according to the agenda provided as Attachm'ent 2.
g In Part I, overviews' of activities related to the severe accident decision-making process were p' resented by Mr. Bernero of NRC, Dr.'s Buhl and Fontana of TEC/IDCOR, and Dr. Sehgal of EPRI.
I Part II of the meeting was related to the development of risk codes.
Following overviews by Dr. Aldrich and Mr. Fuller, detailed presentations were made on thermal-hydraulic code development. Dr. Denning and S
Mr. Cybulskis discussed the efforts underway to improve the MARCH code.
Dr. Henry then described the approach being taken in the development of the MAAP code for IDCOR. Code development to improve capabilities
! in fission product transport were then discussed. The status and content of the MATADOR code was presented by Dr. Denning, followed by presentations t
on IDC09 work in this area by Dr. Sehgal and Dr. Burns.
In this area, an important topic of discussion related to how risk codes such as MARCH and MAAP can be validated. In particular, the role of detailed codes in this context was di:mssed. IDCOR contractors seem to favor the approach of using available experimental data to directly -
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provide reasonable assurance of their code's validity. NRC contractors t indicated a preference for using the detailed codes, which have been or
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{ will be validated'with experimental data, to in turn be used to validate f the risk codes. .
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k After the detailed presentations, issues relating to schedule, uncertainty v
i analyses, validation and benchmarking of these risk codes were discussed.'
l Certain agreements were made with respect to this code work; these are 8 summarized in Section 3.
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F Part III of the meeting dealt with the analysis of the values and' impacts
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of possible plant modifications to reduce risk. Following presentations
[ by Dr. Benjamin and Mr. Asselin on the general approaches being taken by NRC and IDCOR, respectively, work was discussed othbenchmarkNhe present level of risk in existing LWRs. It was indicated by the presenters (Dr. Benjamin and Mr. Young) that this work is being pursued to permit an as-accurate-as-possible characterization of the risk associated with current LWRs for comparison to possible quantitative acceptable risk criteria and as a starting point for risk reduction assessments.
l In the next portion of Part III, the methods for risk reduction analysis were discussed. Mr. Asselin presented the method by which IDCOR intends to pursue this area. Since this part of the 10COR program had not yet started, details and results of the work were not available. Dr. Benjamin, along with Mr. Hatch, Ms. Bennett, Mr. Drayer, Mr. Behr, and Mr. Burke, then presented the details of the NRC approach in this area, including ]
preliminary results, analysis of the financial risk of severe accidents, methods for' the characterization of value-impact results, and uncertainty analysis methods.
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In Part III, several specific topics of importance were discussed. ~0ne b related to the extrapolation of plant-specific PRA calculations to generic classes of LWRs. While it was generally agreed that generic reiults are preferable from a decision-making standpoint, no concensus was evident as to the feasibility of such extrapolations. A second important area of
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> discussion related to NRC's analysis of the financial aspects of severe accidents. Preliminary results from this work were presented and discussed
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O on the different portions of the financial risk: offsite health effects, I property damage, onsite property damage, etc. The relationship of such costs to the ALARA criterion in the proposed safety goal (i.e., $1000 per r
) manremaverted)wasalsodiscussed, j .
I With the completion of these presentations, a period of general discussion, I
with a sumarization of agreements,. ensued. During this discussion, Dr. Kerr expressed a concern that, while he had seen a large amount of s
work on the technical detail to support NRC's decision-making process, l
( work to support how the decisions are to be made was less evident. The ensuing discussion led to no firm conclusions; consideration of this matter was left for additional study and action by NRC. The specification of a date for the next meeting on this subject was deferred until the timing of benchmark calculations could be determined. Following these discussions, the meeting was adjourned.
f I 3. AGREEMENTS
- 1. A benchmarking exercise will be performed using the NRC risk codes ,;
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(MARCH-2, MATADOR) and the IDCOR codes. A set of standard problems will be defined following the NRC/10COR meeting on accident likelihood on August 24-25, 1982. (Cunningham, Tyler)
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p i 2. NRC (Cunningham) will arrange for IDCOR's (Fuller) early ac' cess to -
the MATADOR code.
,. 3. IDCOR (Fontana) will provide to NRC (Ross/Bernero): (1) List of IDC'ORcontractors;(2) Report on Task 2.1 (Ground Rules for the IDCOR Program); and (3) Report on Task 3.1 (Definition of Likely Sequences). .
- 4. NRC (Blond) will provide a copy of the Siting Report to IDCOR E (Tyler). .
lJ S. NRC (SNL - Benjamin) will provide additional preliminary risk reduction results for Calvert Cliffs and Sequoyah to IDCOR (Asselin).
) 6. NRC (Cunningham) will provide copies of ANL's financial risk work l toIDCOR(Tyler). -
- 7. An update of this meeting will be held as results of the benchmark l calculations are completed. (Cunningham, Tyler) ,
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Attachment 1
, Attendees NRC/IDCOR Technical Exchange Meeting 7 July 7, 1982 Name Organization Telepho'ne Number Allan S. Benjamin SNL (505)844-5258 5 Robert M. Bernero NRC-RES (301) 443-5936 i Roger M. Blond NRC-RES (301)443-5960
/ David C. Aldrich SNL (505) 844-9164 Jack V. Walker SNL 844-2876 t Timothy S. Margulies . NRC-RES 443-5960 t Mark A.Cunningham NRC-RES 443-5960 Peter Cybulskis BCL 424-7509
[ Richard S. Denning -
BCL 424-7510 t Jim Meyer NRC-NRR 492-4409 f Steven W. Hatch '
SNL 846-1975 Terry G. Tyler TEC-IDCOR 966-5856
} W. Kerr ACRS 764-6213 l Tony Buhl TEC-IDCOR 966-5856 Denny Ross NRC-RES 427-4338
[ e, D. L. Fuller IDCOR/EPRI at TEC 966-5856
- B. R. Sehgal EPRI 855-2719 M. H. Fontana TEC-IDCOR 966-5856 Stu Asselin TEC-IDCOR 966-5856 R. E. Henry FAI-IDCOR 323-8750 T. M. Howe EG&G 526-9409 b J. A. Dearien EG&G 526-9374 i C. M. Allison EG&G 526-9009
) Ian Wall EPRI 855-2935 John Carey EPRI 855-2105 M. C. Leverett EPRI 855-2936
, W. B. Murfin SNL/SAI 844-3462 l F. E. Haskin SNL 846-0276
! Vance L. Behr '
SNL 846-3086 Marc Kenton FAI 323-8750 f Jeff R. Gabor FAI 323-8750
) Bill Mims TVA 632-7263 1
Odelli Ozer EPRI 855-2089 Joe Rivard SNL 844-6374 N. D. Cox EG&G 526-9685 J. M. McGlaun SNL 844-4408 S. L. Thompson SNL 844-4406 7
G. G. Weigand SNL 844-5554 E. Gorham-h.. u m SNL 844-4065 Kevin Winegardn'er PNL/BNW 375-3839 k -
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l Name Organization Telephone Number
- Milton J. Clauser SNL (505) 844-7154 "
Ken Bergeron SNL (505) 844-2507 ~ -
Bill Xastenberg UCLA (213) 825-2045 Norman Evans SNL (505) 844-5960 Marshall Berman SNL (505).844-1545 Dave McCloskey SNL (505) 844-8870 Ron Iman SNL 505) 844-8834 Trevor Pratt. BNL 516) 282-2630 llans Ludewig BNL 516)282-2624 Ron Lipinski SNL 505) 844-5092 George Edgar IDCOR (202)872-5121 Dirk Dahlgren SNL (505) 844-1407 Bob Burns .
EDS Nuclear (415) 544-8000 9
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9 Attendees l NRC/IDCOR Technical Exchange Meeting f July 8, 1982 q ~n Name Organization Telephone Number i
s Mark A. Cunningham NRC/RES (301)-443-5960
[ Robert M. Bernero NRC/RES (301) 443-5936
- David C. Aldrich SNL (505) 844-9164 h Roger M. B1ond - RES/N.RC3\f\ (301) 443-5960 1 Allan S. Benjamin SNL (505) 844-426% : 13 p Jack V. Walker SNL (505) 844-2876 Peter Cybulskis BCL (614) 424-7509 l Richard S. Denning
- BCL (614) 424-7510
/ Jim Meyer NRC/NRR (301) 492-4409 Ivan Catton UCLA (213) 825-2040
- Terry G. Tyler , TEC/IDCOR (615)h66-5856
) W. Kerr ACRS (313) 764-6213
} Denny Ross NRC/RES (301) 427-4338
] John Raulston TVA (615) 632-3063 1
~
Mario Fontana _" *.
- JEC/IDCOR .1615) 966-5856
.t u w $tuTony Asselin Buhl TEC/IDCOR '(615) 966-5856
, TEC/IDCOR (615) 966-5856 L Jon Young EI/IDCOR (206)854-0080 3
Alan Kolaczkowski SNL (505) 844-8624
- M. C. Leverett EPRI (415) 355-2936 Bill Mims TVA (615) 632-7263 Trevor Pratt BNL (516) 282-2630 Hans Ludewig BNL (516) 282-2624
. Vance Behr SNL (505) 846-3086 4 F. Eric Haskin SNL (505)846-0276
? F. Sciacca EI/SNL (505) 844-2803 j Timothy S. Margulies NRC/RES (301) 443-5960
] W. B. Murfin SNL/SAI (505) 844-3462 b Dirk Dahlgren SNL (505) 844-1407 i
Marc Kenton FAI (312) 323-8750
- Jeff Gabor Ron Iman FAI (312) 323-8750 SNL (505) 844-8834 i R. E. Henry FAI (312) 323-8750 l P. R. Bennett SNL (505) 846-4927
?
Gary Sanders SNL (505) 846-0085 3 Fred Harper SNL (505) 846-1977 1 Darryl Orayer SNL (505) 846-3568 d Wallis Cramond SNL (505 844-5388 5 Kevin Winegardner BNW (509 375-3839 l Jack Hickman SNL (505 844-3874 d Ken Brienzo Consumers Power Co. (517)788-8174 L' B. R. Sehgal EPRI (415) 855-2719 L1 W. E. Kastenberg UCLA (213)825-2045 j N. D. Cox EG&G (208) 526-9685 -
p R. T. Curtis NRC/RES (301) 427-4252 -
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TECHNOLOGY for ENERGY CORPORATION ~
IDCOR PROGRAM MANAGER IDCOR PRESENTATION- ACCIDENT SEQUENCES .
i STUART V. ASSELIN, TEC, IDCOR TECHNICAL MANAGER FOR RISK ANALYSIS i JONATHAN YOUNG, ENERGY, INC., IDCOR TASKS 3.1, 6.1, 7.1 ;
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NRC/lDCOR MEETING ON ACCIDENT SEQUENCES l AUG.24-26,1982 :
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l AusuST 20, 1982 RFPORT ON IDCOR WORKSHOP DN CONTAINMENT STRUCTURAL CAPABILITY WORKSHOP HELD IN CHICAGO ON FEBRUARY 24 AND 25,.1982
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WORKSil0P DN CONTAINMENT STRUCTURAL CAPABILITY , REPORTS PRESENTED FOR ORIGINAL SIATlDN CONTAINMENT TYPE EDNSTRUCTION--MATEfUAL PRHMe - ZldN LARGE DRY PRESTRESSED CONC 8ETE fl7 INDIAN POINT' LARGE DRY f REINFORCED CONCRETE .l7 YANKEE R0WE LARGE DRY STEEL SPilERE 35 SE000YAll ICE CONT. ' STEEL, STIFFENED CYLINDER 12 BROWNS FERRY MARK I STEEL 56 LIMERICK MARK 11 REINFORCED CONCRETE 55 LASALLE MARK II PRESTRESSED CONCRETE fiS IIARTSVILLE MARK III STEEL 15 GRAND GULF l MARK III REINFORCED CONCRETE 15 l
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AUGUST 20, 1982 .
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WORKSHOP DN CONTAINMENT STRUCTURAL CAPABIL~ITY GENERAL 0BSERVATIONS
- l. PRESSURE LOADS ARE CONSIDERED AS STATIC LOADS.
- 2. HAND CALCULATIONS AND COMPUTER ANALYSIS ARE USED. ,
- 3. AS-Bu1LT MATERIAL PROPERTIES ARE INVESTIGATED.
- 4. AILURE IS DEFINED AS MATERIAL STRESS AND STRAIN BEYOND VALUES 3ERMITTED FOR DESIGN BY BUILDING CODES.
- 5. FAILURE PRESSURES ARE DEFIJED AS VALVES AT WHICH NO LEAKAGE OR DTRUCTURAL-COLLAPSE IS ;XPECTED. (5MALLDEFORMATIONS) 6'. THE WEAK STRUCTURAL LINK IS DEFINED.
- 7. TEMPERATURE.IS OF LITTLE CONSEQUENCE.
-8. (APACITY OF EQU PMENT HATClES, PERSONNEL LOCxS, PIPING PENETRATIONS, LECTRICAL 3 ENETRATIONS, AND VALVES ARE 3
ADDRESSED FOR OME LANTS. 1 a 1 e e
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I.; ; e PLANT FEATURE SURVEY y.- l 9 4 e PRA COMPARISON LIMITATIONS Z!j
- n . .
??b N .i
['. ! _\ -
}
;.4 (f j
*i l .
Blai
N 'P ' " . M M W 6 % . 6 . 7 2 :.Sei n... s . n ',G C & u a
'T w.' ;;pm 6
. .\ ,
.I J ,
^
ru
'r P
i J _ gup I f b _%s. 1
.)
C
':l e
.:4
/
t',4
.g g4 9
m SLBTASK 3.2 - RESULTS x . s.i.
. l,"
.3.-
,J
- ')
- ?.
C .. . S.$ 43 T 1,'i, s'
.I!
u '
.?
e b* o , rJ 71, h!
?
y 3
'N v -A . n> I 31 n
[f .
~'
J.al
.'1 I ' 'e . j m'J k,f
mentc>c eu. , .- - .zec- mm,mus-ar._ m ungumx m I ca , ., g . f~ a-u Total Frequency of Dorninant Core Demoge Sequences , r?- At Eoch Picnt - .
?ni N,
Is l Plant Core Plant Demoge Frequency
- l.*
- 7 i
lf Sorry 4.4E-05 i, l, 3!0 Oconee 7.3E-05
4 *;l Sequoyoh 5.7E-05 7.q:--
2 e? Zion 4.8E-05 sJ. P; i sq Pi Peoch Bottom 2.lE-05 il i
] Grand Gulf 3.5E-05 2
R;s Limerick l.4E-05
- q fi
- r.i l ig Q-44 k:.t 7;
v 2 74 _ dy [] .
'All frequencies are events / year.
m e . _ _-- 4- . . = - . - - ,- -
7we em> -wes=namrmems. -e-y s t*
,.i $ *. , i 7
FUNCTIONAL CLASSIFICATION SCHEME
- 5) --
(Q - fej ,
.?!
j ._ IDENTIFIER FUNCTION
.;V k
Gi 33 RIE II) . REACTOR INTEGRITY - EARLY RI L II L) REACTOR INTEGRITY - LATE jj CMoE (2) CORE INVENTORY MAKEUP - E .:.LY
-1.
gj C,ML (2L) CORE INVENTORY MAKEUP - L-7-. C,HE (3) CORE HEAT REMOVAL - EARL' 1 3, Cqo (3L) CORE HEAT REMOVAL - LATE [. CPNE (4) CONTAINMENT PRESSURE SUF:: ESSION - EARLY j CPNL I4L) CONTAINMENT PRESSURE SUP::ESSION - LATE j CHNE (5) CONTAINMENT HEAT REMOVA~ - EARLY i; ijg CHNL (5L) CONTAINMENT HEAT REMOVA_ ' ATE f CINE (6) CONTAINMENT INTEGRITY - EA :-.Y g ,m, CINL I6L) CONTA!NMENT INTEGRITY - LG
,C RNE (7) CONTA!NMENT RADIOACTIVIT' :EMOVAL - EARLY yi CRNL (7L) CONTAINMENT RADIOACTIVIT' :EMOVAL - LATE G
M 64 Il 9
~
hii h? +:t
.4 n
- d n6
.f- _
.r,,- . -v - ,. .,-
m-- .m w -- mm a. . = c .m;. ~ , ,masa s \.3 g?- g. 1 , Dominent Core Demoge Sequences - Surry L P,* gj ' - - - N 2- j W - P.'j Core Demoge - Q Sequence Frecuency Functional Foilure E. M:j SD2 9E-06 12 [' SH 2 6E-06 12 L 1
- TML . 6E-06 3 g V 4E-06 126
~.
% SgD 3E-06 12 d.1
-SgH 3E-06 12 L
'[ 1 TKO 3E-06 123
'il TMLB' 3E-06 ' 234S7 dj Sj AD 2E-06 12 @fa s AH . 2E-06 12 L
N tj SC 2 2E-06 147 a C
, ?>
TKMQ lE-06 123 i'.v 4 Iij AF IE-08 14LLL S7 f-1
,44 h, c j
- 4 h}
h.i t. TOTAL FREQUENCY 4.4E-OS Il
,.J h*j 7J B
3 a ta, [. ' -
'f.
41 m m _
_- va- - n ,~- - _ v - m n.
- il .
2 o, '* a i,) Dominant Core Demoge Sequences - Oconee* 1,? M. . 5j Core Demoge g Sequence Frecuency Functional Foiture i SH3 IE-05 12 L t f; T2MO-H IE-05 12 L v
;f T2KMU 8E-06 23 st s
i f T2MLUO 8E-06 23 M a SgD 7E-06 12 M sg T gMLUO SE-06 23 ':; T2MO-FM SE-06 12LL 7 S3FH AE-06 12LL 7 di V 4E-06 126 M 5D 2E-06 12 (:{. 2 Of Tg(83)MLU 2E-06 23457 .c 3 T gMLU 2E-06 23 9:
']g; T2MO-D 2E-06 12 ld S2FH IE-06 12LL7 Ut jj SD 3 IE-06 12 h,, T2MLU IE-06 23 x,
Y T3MLUO IE-06 23 6[ 4
.9 3
M II TOTAL FREQUENCY 7.3E-05 .:9 21 m .o
.n A +0ther seq ences specified in the IDCOR Task 3.1 Workshop report which contributed to (r.1 '1 less than one percent of the total frequency were not included.
k ',.h
'J L'~~~:-- v~. _, .
y . *
- ~ ..n .- _. -- n -__-~ ._ , . :..- . .= =,maux, 0}'
j}
- g. ..., - . .
.1 IE' Dominont Core Demoge Sequences - Sequoyoh
- t.; . ,.
p - NI 4 2 - n d Sequence Coce Demoge $,l Frequency Functional Failure , c. .; ' #j SH 2 2E-05 12 L
-d) SHg IE-05 12 t
A
,j SO2 6E-06 12 S2HF SE-06 12457 e LLLL
-y V SE-06 126
..i Ls4 SDg 4E-06 ,
12 si q SgHF 3E-06 124S7 ya LLLL f T 23ML 3E-06 3 pvl d AH SE-07 12 g L l' j , Tg38 MLB'l3 3E-07 23457 b;j AD 2E-07 12 h s. AHF 9E-08 12457 LLLL P. f3 gj TOTAL FREGUENCY 5.7E-05
- g. ;.
$I E1 n I,l q - E.' (. '4 ry >u '
'i .
M y [i n .
-w- ;- ,, , . - . . , -
.. r -
.. _,__,,._,, _wrnmemmw*
- (j .
Mj +. ,* rs . ~ k,. ; Dominent Core Demoge Sequences - Zion N.; 3 .
-n V.
c .. .
?
d - M rh Core Demoge
;g. Sequence Frecuency Functional Failure I
f 2E-05 12LL 7
% 2 '
6E-06 23457 I;lq 3 SE-06 12LL 7 - w{ .j 4 SE-06 12LL7
# 5
.;p 4E-06 3
;.i 6 3E-06 3 4'5.
7 2E-06
... 13LL 7 p 8 IE-06 12LLg 45 n."
9 IE-06
.. 12 pj 10 4E-07 12 93 j !! 3E-07
.:j 23 .
~
2: 12 2E-07 23 -ll fjj _ 13 2E-07 23 E 1..j 4 I4 2E-07 23457 ... ) 15 {. 2E-07 23 it w 16 2E-07 126 55 y TOTAL FREQUENCY 4.8E-05
?!
3 . ,.J r -t I m, y p3 .
..me w =, ,.mm m m p~
[e 6R 41 fij
$il Dominant Core Demoge Sequences - Peoch E'ettom yj _
11* e ._, I j . b g Core Demoge Sequence Frequency Functional Foilure yl
.-) % TC IE-05 23
- ,3 TW IE-05 S L
h
.y
. v: TOUV SE-07 2 - s c 5E g 2E-07 12
- x il AE IE-07 12 m
\y S2H1 IE-07 12 t
tl1j 3 5J2 IE-07 IS L b n R IE-07 12 bl J!s! SE2 SE-08 12 e i,-){ S HI 4E-08 12 L [j SJ 3E-08 .ISL $5) AHI IE-08 12L
.a AJ IE-08 h
~s IS L B! n 'c1
)j TOTAL FREQUENCY 2. l E-05 if
- e l
bl n k5 m 0 : (
.i
-lk
'rj . ~ .- - - . .
; , m3,. . - ,-. - - ,- _
~ - y. - - % .
mem mm Ny:! . . .. F.; y . k;. Dominant Core Domage Sequences - Crand Gulf q Q
.5-
- n. .
h* 1 [M Sequence Care Domcse Frequencz- Functional Failure lNl T 230W IE-05 S L 7, T gQW ' 6E-06 S L lf SI SE-06 15g T' T23C SE-06
- 23 l:{,j T23POI 4E-06 IS
. L g T gPOI 2E-06 -
ISL g. is TlQUV 2E-06 2
%i $ T23POE SE-07 12 TgPOE 2E-07 12 p .
I3 TOTAL FREQUENCY 3.5E-05 dl w., 4 f4
;d. -)h M -M a
n [t5 E'?) y .., e.; ' 'ei i[.) n
- t . . - , , - -
, v
#= -
_-.__y- .._._ ____.,- - m m 7, . _ , . . , e
- e. . :
I$ p ., Dominant Core Demoge Sequences - Limerick g,.: Mg l ; <, i 'd' A, .. h Sequence Core Demoge Frequency Functiono! Foilure %d rg ;. TEUV 6E-06 23457 3/
. .e TpOUX 4E-06 2 n
j TTOUX 8E-07 ,2
] ,,'
TEUX 7E-07 2 TgUX 7E-07 12 14 TTPW(P) 4E-07 IS t [ b; ; TTI CgPU 3E-07 123 !.[ tr. TFGW(0) 2E-07 S L $) e .;;p Tp2 CMPU 2E-07 123 . Tp2CgUUR 2E-07 23 2 2E-07 235 TpC gW12 x@;;' ip TgGUX 2E-07 2
?.7
?;j TFPW(P) lE-07 IS L ys TpOUV IE-07 2 2>d . T gC'UX IE-07 123 I TT CC g2 lE-07 23 u M 49 'Il fil TOTAL FREQUENCY 1.4E-05 -
;d -
Lid M.]
- 3j
')}
['g s; r- - , ,
- + - - ~ w - - w o - e - - e - >. m 7 4
I,.ti .*. ..* m4 Id M v.,'. Summary of Functional Group Contributions a . d To Total Core Demoge Frequency ._ 4.j, - I T . 7 9 PERCENT CONTRIBUTION e GROUP PWR BWR
.q 2 ,
I6.22 3 8.37 r W
,j* Sg 3I.58 s 12 17.27 3.02 a
li' . 12 g 26.48 .24 {j ISL l1.87 1 1 23 9.04 21.34
. 4 .I23 2.27 1.40 126 5.94 M]
g
.;r ..j g 12gL 7 19.05 )j, 13LL 7 1.04
}qi 147 1.14 !:1
.j 235 .47
*j. 12LLt45 .4 1
! 14LLLS7 0 4 124S7 tLLL 3.55 s.[
ej 23457 S.75 13.99
':j o ?
TOTAL 100. 100. - Jl l1 g 0
- h. .
c . ,- ,-. . . . . . - . _ -.. -.,,
.Am. -
x.m.- %%2c.,. _ . _ w, . - . .. u ,6 . - - . m .. ,
.m.2 2,;w.m,
;4
*e * *
.4 1
Function Contribution to Core Demoge
.(
LLj
. n
',.]
'Nj
.a
.j,
' ))
FUNCTION PWR BWR _OVERALL
;j (Percent) (Percent) (Percent)
;t
$l RI E 77.3 16.5 51.2 11 5.Q RIL 0 0 0 q
14 on CMo E 40.3 56.4 47.2
.M ?g4 - CMo L 49.6 0.2 28.4 C,HE 25.4 37.2 30.5 '}a .1
- ),j CA 1.0 0 0.6
,;j CPgE 6.9 14.0 9.9 1 #pj Cdl 4.1 0 2.3
['qIj CHNE 5.8 14.5 9.5
-u .p; ; i.
Cd 4.I 43.5 21.0 CINE 5.9 0 3.4 i. p P. T CINL 0 0 0
.s .
;. Csag 6.9 14.0 9.9 j
CR NL 23.6 0 12.9
.n 1
- 4
+H 5
R3 pd 9a y _
.. s; il
,t
-f<
9d
m.s ea. w m m w ceng=.a w - ,sem m2= i h
- M-kj -i , .
General Types of i_ight Water Reactor Plcnts h d 4 Ef1 ._ , - 4 t
'1 r:5 Type of Containment ** -
a 9 Recctor NSSS Operating Designed or j Type Supplier
- Plants Under Constructico 1
(J ij PWR B&W I, 5 2, 4 i C-E I, 3, 5 9 e a
;1 -
W I,2,5,6,7,8 3, 4 . N m-y
,;4 -
7, BWR GE 10,13 I I,12 5;u , AC 14 l:_ - .
.y'k a;
4 il ih
~;;
- t
.:n j ,NSSS Suppliers: 4 9 B&W - The Babcock & Wilcox Compcny (]j C-E _ - Combustion Engineering - 3.[I W ,
- Westinghouse Electric Corporation 3;) GE - General Electric Company h AC - Allis-Cbc!mers btj y . /.1 .li '
k:':;d , h, - a J-'t ,, Refer to fortowing pcge for description of containment type. Kj y 1 a , -
m.-mem -- -,% - - x,m - - mE! dj < ,= .. g.; w1 a -
" Type of Containment c.
f2fl Y2l PWR: v; 41 p' Concrete cylinder with steel liner - no secondary containment. l.
- 2. Concrete cylinder with steel liner (subatmospheric (9-11 psic)) - no secondcry
($e (j v containment. f,) 3. Concrete cylinder with steel liner - steel enclosure building cs secondary H containment. pi , d '4. Concrete cylinder with steel liner - concrete shell as secondary containment (slightly subatmospheric). d,, p; S. Steel cylinder - concrete shield building as secondary contninenent. ,
.e
- 6. Steel cylinder with ice condenser - concrete shield building as secondary b@tc contalment.
,r it.; 7. Steel spher: (above grade). . "4 8.' Steel sphere (partially below grade).
3 j 9. Steel sphere (above grade) - concrete shield building as secondary contcinment y ' (slightly sobotmospheric). . s 4-)
,] BWR:
'. i 1- 10. Mark I: Steel lightbulb and torus design with concrete shield - reoctor building
'j cs secondary containment i
e i.! I1. Mark II: Steel lined concrete conical frustrum drywell and cylindrical (q suppression chamber - reoctor building as secondcry containment, m W 12. Mark Ill: Drywe!' and suppression pool inside steel lined concrete cylinder - reactor building as secondcry containment, gj h vji
- 13. Steel sphere (particily below grade).
j I4. Steel cylinder (concrete shielding on inside). d ??A G'G -
.g:
y .
- 7. .;
y. if n e~ , . ._ _
_ ~ . ~ .
~. - ...u...-
y.. .. (J a TA '
. g. <
N COMPARISON STUDIES 33 - 31
' .1 .'I*l
- f. 5 4j';u.
.t O ,$). ? e CATEGORIZATION OF INITIATING EVENTS, -t; y,
- 4 6
d e TOTAL CORE DAMAGE FREQUENCY J" 1 8.-yc .,.
-, .e INTERFACING SYSTEMS LOCA (V) i:
1 I'j ii; .e TMLB' AND EQUIVALENT SEQUENCES s.r s . .
.:j -
a j e SAFETY SYSTEM CONTRIBUTION TO CORE DAMAGE
- M ix
.v d- e VESSEL STEAM EXPLOSION PROBABILITIES sv4 l 50 x p. Al a
- p. 4. 'C-d.4 .
r m. Vd.
"'a' 4
aa, bl .rg
#1
- y
3lN&W "*3 %2TH1d%2iM2Zi@s2LEGlGfiEi&Ls??Lii2HHh [2LM2fMM;)iTWihELG.")
~
1 ] Summary of Initiating Event Frequencies and Contributions to Core Damage l Combined % Transient Combined % LOCA 1
. Transient Initiated LOCA Initiated
. Plant Frequency Core Damage _ Frequency Core Damage _
Surry 10.2 29% 1.4E-3 71% Oconee 7.2 ' 37% 1.95-3 . 40% 23%** Sequoyah 7.2 5% 2.8E-3 95% Zion 14.0 34% 3.7E-2* - 66% Peach Bottom 10.2 98% 1.4E-3 2% Grand Gulf 7.2 86% 1.5E-3 14% . Limerick 9.0 ( 95% 1.2E-2 57** 0% j nt. 4 Zion also included an SG tube rupture frequency of 2.4E-2 per year. This was not a'significant contributor to risk. The value of 3.7E-2 per year does not include this SG tube rupture frequency. i i ** Transient induced LOCAs. iS q
*** Inadvertently open relief valve. '
, [j q
*I ll .
,.gwtx > . - r ,
- u. .s->. .~.t. u,.. c .- . ,vn , . s. .w. c ;.;, . . . .
. > naa.;uc,, , ..c;
- l,.-
q -t. *e
.a' s . i.a "G ,
u;- E4 .
- ~.-
'X SENSITIVITY STUDIES
- /
- 4.
- N1
' :h jy Tc w
g . w fa
- E,;,
a 2: W4 ?M e VESSEL STEAM EXPLOSIONS
+ .
e. pl - il-
.ry , e SYSTEMS OPERABILITY IN A FAILED CONTAINMENT q
M1-gj
% e DEBRIS BED COOLABILITY b'
e EFFICACY OF FEED AND BLEED *
- f6l.
N. k. .
*.4 H :q x;
*i i
't i
- 3. ..
- Y) h
)
?
Ci' ad , bei El
$*I L 4 u
f))- m.- .
u- ,> m eum warmvwww .mam. va.x,,
'].
rn
.?
?' , . .
a5 e;c M i ., b. 8s fj Estimated Impact of Assumed Feed and Bleed Viability for PWRs , 4 ll1t y1
.d' ..
- f Baseline Core Melt d Frequency with Core Melt %
M. ' ' Plant Frequency Feed and Bleed Chance Comments
,?'
Surry 4.4E-5 3.8E-5 -14% Reduction in releases
}- commensurete with reduction in core melt.
.a .
.3 -,
d; y Oconee 7.3E-5 7.3E-5 ' O Boseline assumes F&B is jj viable. s:
- M.f fi QJ Sequoyoh 5.7E-5 5.4E-5 -5% Reduction release commensurate with
#j reduction in core melt.
n..p! . sj Zion 4.8E-5 4.8E-5 0 Baseline assumes F&B is .. viable. 3 r,f i
-k
'fah 3
#'3 i1 m
S' v.i : j-4 . ! a A.
- m. ,
n
- di n -
k,4 - J(_j - MI i
- A.
G I'!). si m, . _. _. ._._ _ ., , . , . . -
; -. - . _ m..._ .. -- .
. .__ m__,.c.__,.
t
.;A +. .
J 3
'1 ,i1 1 , .t j a
- 1 SUBTASK 6.1 - RISK SIGNIFICANCE PROFILE FOR ESF
?
,. ;d, . r.o - N uq g 4. ?hl L'.;3 f.! *4.
'l r 4]
?! e MEASURES OF IMPORTANCE OF FEATURES 5 b, ' ,.
.:. 3
~Jq - TOTAL CORE DAMAGE
'a
+ .:;.
- BY INITIATOR TYPE vg ,.4 pv PJ f9 5
e MEASURES OF IMPORTANCE OF FUNCTIONAL GROUPS
.,-d s .:
l .s 31 e VARIABILITY ASSESSMENT l::: p-
,1 z
M sj e IMPACT OF UNSTUDIED EVENTS p' n.
~ '{
V'q ' '3
- ~4 .f
'M
[M
- *.1 r M. ,
VJ I'i k u's ':, ;t O p'- [$3 >.
- _4
!1 , ,d.
- l. '
l _ . - - - - - - .
.y,- v ., m_.mn --n,-mw_mwnm. g
.,n ,g a g g w m ;
E Th l
a?.26., ,, .
1
.'J. ' , Table 2-1 ' '9 .J System Risk Significance Summary ,,
M
;y-. . .. _
- 1j-
%j: i L,% ' (d L'N jj .?.4 !
Surry Oconee Sequoyah Zion Pecch Bottom Grand Gulf Limerici<
.a #,i % AFW PCS HPR LPR PCS RHR-SPC HPCI W
g (.41) (.59) (.61)' (.67) (.95) (.77) (.93) TM
,d,3.
HPl LPR LPR HPI/HPR HPSWS PCS RCIC M (.27) (.4 l} (.60) (.46) (.48) (.7 l) (.93) dq .
,w.
P LPR HPR AFW ESWS ADS Pl
,b (.25) (.40) (.29) (.48) (.46) .x..
8 n !? PCS S/RV RPS W (.23) (.23) (.48) b y-led i P HPR - h (.20) ' 1 ,l Y$b
*J ..q 1 -
4'T I ?; p.s
. -i
-hM
- g. >
L9 dli nh y.6.* ; 5-
L.3.m _x mm -- ;.~. > . . ..m.. .. .mn- %,.. z .. 4 m ., _ s,m,w . . wamem g. 4
- pj-sr , :
.g . t . . .s g.-..
,;c
.j Table 3-8
.e si < -
- Adjusted System importance Relative to Core Domoge .r qi, Surry - 4. _ -
it) . kb .4 - Adjusted h[k System importance Importance . Fj O Measure Measure .f) . c: e.' S .LPR . l.3E-02 .25 3.2E-03 Q B p HPI 8.6E-03 .27 2.3E-03 d: PCS 1.0E-02 .23 2.3E-03 HPR 9.0E-03 .20 1.8E-03 7f} _
- @ S/RV l.0E-02 .09 9.0E-04 LPI 4.7E-03 .14 6.6E-04 kl T3, AFW l.2E-03 .41 4.9E-04 M, CSI 2.4E-03 .05 1.2E-04
'9 Accumulotors 9.SE-04 .11 1.0E-04 .L RPS 3.6E-05 .09 9.0E-04 '6 '/l *U .
I ' aj l:n s - m # M b
.i i
b y.
*l .
bl n2 h[j . -
awnm m w w www= v maw; (/d h, 1 Table 3-9 h Adjusted System Importance Relative to Core Damage ik,If Oconee 4. . U.^ p
' Y, .-
J Adjusted Importance importance System O Measure Measure k 3 3:. S/RV . 5.0E-02 .23 1.2E-02 j PCS 1.0E-02 .59 5.9E-03 y HPR 9.lE-03 .40 3.6E-03 41 HPI (feed and bleed) 1.5E-02(I) .15 2.3E-03 [ HPI (injection) 1.5E-02 .14 2.lE-03
.$ -LPR 4. l E-03 .41 l .7E-03 IN' CSR 7.lE-03 .14 9.9E-04 LPI. 2.6E-03 .I8 4.7E-04 h
i:n AFW 6.5E-04 (2) .07 . 4.6E-05 O LPSWS 2.7E-05 y- .19 5.!E-06
'; RPS 2.6E-05 .II 2.9E-06 p
p t,s] - h'l}i- $lb II) Represents value for HPl unavailability when AC power is availcble. gh (2) Represents value for emergency feedwater unavailcbility during a Tg -type transient. N, .
'q e
n y q I/I
- c g$85n2
^ jq; c , , , - , _ . -, =.,
tgwy _ . . >. x m . ,~.a . ;_+.. ww w i~-wJu >uw -wmmnwsza.ecn -m ~exces : xem
4 - .,J -
M gj A ,- ..
<? .e .?q Table 3-10 Adjusted System importance Relative to Core Domoge 90 .o .
Sequoyoh *' vt
-1 4,1 -
p
- u. --
.;f >_. Adjusted -
G Importonce Importance [d System Q Measure Measure fd sa i.N
@j HPR '
8.0E-03 .6 I 4.9E-03 n. LPR 4.6E-03 .60 2.8E-03 41.7'
;Ii ' HPI 3.SE-03 .19 6.7E-04 p .
W
'; a PCS 1.0E-02 .06 6.0E-04 m; .AFW l.0E-0S .06 6.0E-07 n
y- CSS 3.3E-03 .14 4.6E- 04 gJ
;,.9 - LPI l.9E-03 .10 1.9E-04 ,M5 ' * ,4 -; p, n. ]$
q.s jjj . Id m
- i 1.i Ti J.1 J.$ ,.<
,d
<J il
.3 -
4 4 I 8T.h .a wt . e,.4
- .3
,- p q Ir,*,) n2 @l e _ _a _- F.A , - - = - -_ - --- - - . - - - - -- -
ya - . n.,. ~_ ... s .. w m _ w.s . . _c.sw , , .
. ;_n. : x. . . ,a v , ,. .gwge.,...>m. ,
4; m.7 M i
.ie .
G:r Table 3-1 I Adjusted System importance Relative to Core Damage 4dk
?fl ,
Zion ~ - -l l L'f w]-
?? ,
p; i
)
W Adjusted $) Importance importance SystemII) Q Measure Measure (x).<,
.z.
xw LPR l.8E-02 .67 1.2E-02 [q% g. . HPl/HPR 9.2E-03 .46 4.2E-03 11: RPS l.lE-04 .15 1.7E-05 +M LPI 4.4E-041 .03 1.3E-05
- s ;
3 AFW 4.2E-06 .29 1.2E-06 s %g HPI (feed and bleed) 2.8E-06 .08 2.2E-07 y 4-
/;
c y
-l
- >s (I) Zion study assumed PCS to be always unavailable, therefore PCS is not included on e.
F. this table.
.i .~ G' -
ss, ~ N
'6 a
e.
}}. .
l} 1 - 3
.Y s
f'In2
" -w.. wu.w ..
a., a.a . - m " - . w w.~ 1. :.. .;;., .. ;vs.~x~ ~ ~ :.xxaxw,. %;1 s
'.)
fl. ,. E s Tcble 3-12 [4: Adjusted System importance Relative to Core Ocmoge - 5 s1 .Pecch Bottom - ,4
- t w
i. tu bl Adjusted (!- Importance importonce Q System O Mecsure Measure (:.l L'!.4 idj
- gg PCS 7.0E-03 .95 6.7E-03 o
g HPCI 9.8E-02 .04 3.9E-03 0, RCIC 8.0E-02 .03 2.4E-03 y
.3 LPCI l.5E-02 .04 6.0E-04 q
9 HPSWS P. 4 4.3E-04 .48 2. l E-04 [ ESWS 1.2E-04 .48 5.8E-05 f ADS 5.0E-0.3 .01 5.0E-05 t.:. CSS 9.5E-04 .04 3.8E-05
- i. <
<] ' 'l t-L'j H:t m
o aj Km lii 1, 4 r.a b . I[*t f $:j L-r.h Ii}
-7 n ?b l'
f_ ,j f.1 SI - 1 - a . .t r4 l J \v
. U2 y 4
,,-- .a. - .- u. . ... w -
.m- .,
.w e.,c m. .
_mi . w,u 4,j . ..
,j),1 Table 3-13 1
Adjusted System importance Relative to Core Damage
,d gs. a Grand Gulf -
i,: z,
.3 - - -
fN l .[g Adjusted f.e3 Importance importance (.[j System Q Measure Measure My E%.:) J l0.' PCS 2.0E-0 I .7 I l.4E-0I a. pe S/RV l.0E-01
- .t; .19 1.9E-02 rs RCIC S.2E-02 .08 4.2E-03 F .. i K!,
e LPCI 4.lE-02 .08 3.3E-03 M LPCS
.e1 3.SE-02 .08 2.8E-03 y HPCS 3.3E-02 .08 2.6E-03
'q RHR-SPC 3.0E-03 .77 2.3E-03
.]
ADS S.0E-03 .08 4.0E-04 " ~;) RPS l.4E-OS .14 2.0E-OS d 0:^ c ,: la i y .,' .I y C q . o y
^*k .1 S
r.;I e i g Tl
..: n2 __ . _ _ . . ..
- ,, .- - , w - - vm - . m m m m aw~ ~ v w '+"-"= ma c Mu M **'-
, /.[,. g;---... . .. .
- y.
Jih m, i Table 3-14 Adjusted System importance Relative to Core Demoge
. Limerick ~- -
t',; 4, ,1 s s Q: - a.ej - g .. $[9$. Adjusted . Importonce Importance (($ - System (l) Q Measure Measure P 4l ' Yf S/RV l.0E-0I .07 7.0E-03 c) HPCl/RCIC
% 4.9E-03 .93 4.6E-03
'w
;y PCS 2.0E-02 .II 2.2E-03 .
NI ADS 2.0E-03 .46 9.2E-04
'/ 4 RHR or RHRSW 8.0E-OS .06 4.8E-06
}1 ,, RPS 3.0E-05 .07 2.lE-06 LPCI 7.7E-OS .01 7.7E-07 a HPCl/Feedwater 3.0E-07 .02 6.0E-09 kb n1; .. n - d4 El y yi d II)Some systems are grouped together for the purpose of obtaining an unavailcbility 9 from the Limerick PRA. b,); p* l. (*y $ u i' ^ sej . 6::1 1. 3 I.u'/, I} M
- q F '.f -
F01 hel - 7:, d
- n2 --
m ,s . m -
-a
p Klo
- w m e"t p 6 9isi e m *'mrNSV!?ISzcem w/s7
- o iii g
r.. -
~.
fw Table 2-2 j ~;g Protection Features and Contribution to Core Demoge, {. ' 'M
.i -
1 n
*Ij _
J' Protection Features PWR (%) BWR (%) 3 h
!!)
;) . Core Protection - Early 48 47 C
r Core Protection - Latg Si ** Containment Protection 10 63 Release Mitigation 34 9 5, 3 _k
- .3 p :,
f, ., I! h.
.4 1: . t,1 *The percent contribution to core demoge is weighted by core demoge @j frequency, see text for details of calculations.
e i s g **Less than l percent. M
.Y) :.
I
,)
't !.pj 0, .' r4 Q n m;h - .? r ~?
,i q3 -n22 - .s.
h ___ _
yveveccaJassmxNWe*W8535E'EJFONN
..r s
k,'- 3+ <j 1- Table 3-42 l Protection Features and Core Damage Frequency -
^
PWRs - 1 4 1 Core Damage Contribution to Protection Features Frequency Core Damaae (%) h.3
. j; M e.1 Surry 3
Core Protection - Earfy 3.lE-05 70 d Core Protection - Late I.lE-05 25 Comainment Protection SE-06 II
. Release Mitigation - 9E-06 20 'l
- d Oconee
{t y Core Protection - Early 4.3E-05 59
;j Core Protection - Late 3.0E-05 41 2 Containment Protection 2E-06 3 ? Release Mitigation 1.6E-05 22 h Sequoyah i'f Core Protection - Early 1.7E-05 32 N Core Protection - Late 3.9E-05 68 7 Containment Protection 8E-06 15 / Release Mitigation l.3E-05 24 0
Zion 4 l]% Core Protection - Early Core Protection - Late 1.6E-05 3.3E-05 33 69 d Containment Protection 7E-06 15 y Release Mitigation 3.8E-05 80 W PWR Average ?d d Core Protection - Early 2.7E-05 48 Core Protection - Late 2.8E-05 51 r Containment Protection 6E-06 10 ~ 1, Release Mitigation 1.9E-05 34 n _ M x; N n2h1 S
y mn... .c , , . . , - - ,, _.,._a. . ,- , .. a . . ,=
..a.u. w w _., ..o c . . ; . , , . . mam .-
t . i/: 1 .;f. . * ., ,. i.2, - 3,j p1 Table 3-43 g Protection Feature and Core Damage Frequency . jd . BWRs '
/
o.u 'I
.a -
'$} Core Damage Contribution to $1 Protection Features Frequency Core Damage (%) C c.1 - 'M 4.) Peach Bottom sy if d Core Protection - Early Core Protection - Late I.lE-05 2E-07 52 i W Containment Protection 1.0E-05 48 .9 Release Mitigation 0 0 m.b i;) Grand Gulf ,y
}1 Core Protection - Early 8E-06 22 7.} Core Protection - Late 0 0 fi. Containment Pr'otection ' 2.7E-05 77 211 Relecse Mitigation 0 0 h
3, 5: Limerick
.t L;, Core Protection - Early 1.4E-05 97 m Core Protection - Late 0 0 ? Containment Protection 7E-06 49-d:4 Release Mitigation 6E-06 43 ht T.] BWR Averace- ?;
$!- Core Protection - Early 1. l E-05 47 d- Core Protection - Late 7E-08
- y Containment Protection 1.5E-05 63
- N Release Mitigation 2E-06 9 T
Ti
*i c:1 *Less than I percent.
IY Y- - e U
. k,
'J
;A -) n2a fl , .,. , . . ,
4
'r Y"$. :YY $$? A *: -
h k- ~ E 1.1 $% k.g 1 . A;$
]* Table 4-1 '.
j Synopsis of External Events 2
.j ,
(VEIITS 4 Winds, Aircraft.. iC teactor Setssic Fire fleed Tornadoes Illssiles Sebetage . WA5H-1400 Estimated order of General estimate that Design basis fleed Virtually all tornedoes Considered to be
- magnitude elsk at fire seuld be 20% of protectlen considered Considered to be within miniscule risk from generic U.S. site. rist from all other sufficient to the design bests of the aircraft.
}
Average frequency causes. This results eliminate fleed plant. Alsk therefore .4 Turbtw mi 6 a.as 10'I per year for la C.M. frequency of laduced risk. censid m d min k l. ,,,,g,,,,,ssiles
,,,4 core melt. Range was 4-8 a 30-6/ r. Stace -
laslentricant risk r 10' to 10'I. Non- t s was we I within study uncertaintles* to th cm. 1 speClfte analysis which estimated II" " "*I P"''d Turbine missiles frequency of ground "' #* damaging fuel in Spent fuel peel accelerations and k corresponding
- 4 a 10' 5' probattllty of C.M.
e a 2l04 Estensive seismicity- Systematic, plant fragility analysis, Design bests fleed Tornadoes considered to Aircraft hit prob.
- specific fire analysts, protectlen constdered be within the design
. Mean core damage of Core melt frequency of suffletent to basis.
- 2 a 10
- 4 5.6E-6. Site spectitt 6 elleinste fleed seismic curves. plant specific analysis of "Ifg,[',"g laduced risk.
Core damage prob. estimated at c w ent fragil Hy result la contalament Fleed laduced core I a 10'I. and system failure g,gy,,,, g,,,9, g,,,gg,, models. H ghting equipment Turbine alssiles i estimated at below '
= 1 a 10-6 1 a 10*I.
LIM (alCK + * * * *
- 1l R55 map * * * * * * ~
s
,
- hot lactuded j
p. e I'
] .
O
'l i
e s , @
, p .L e_e ?: - :.u.x =: - e.us e 'co,hdw s sdi.i . O _ m,u.,.s.; - - ' + _ - - ^-s c.<uz.r.s h : .;.
..]
1 .
-K,i . .* .
if-Areas of Potential Limitations on Compcrisons
,a -
- s s
.j A ... .4 . .'a .. ~. .. I,
- l. Overall Study Scope
,[hj 13 s.1 '+: 2. Systems Analysis. Methodologies
. ... t Mws
"...s c.i 3. Success Criteria n
j 4. External Events Analysis l,[ S. Data Base Utilization and Development ,*1 Li 6. Human Reliability Analysis
.' e i
J
- 7. Accident Processes and Consequence Analysis m$
- b. : .
'd 8. Offsite Consequence Analysis
- ..r.
- l Bj 9. Treatment of ATWS.
r ... .- I i;:n F .4 1' i s'; ?%b .a tM ? YI ia! i t.f
- *il
>". y ij zu - -
- 5. ). -
,.,.f
*Y a
inamescmscaawreewm.+ewwwwnwra.swnser:1.w'~ ww+1sslywm:wn 3 q.t.
,,;j 'A y)i.
j -~.. ; y. [ v., SUBTASK 7.1 - BACJLINI RISK PROFILE
.p. .-
P']4
!b't 1
Y.: kh ' a ' ed gq-ca e METHODS FOR COMBINING AND/OR COMPARING RESULTS ~$ [ - PLANT DAMAGE STATE FREQUENCY gq , L- - RELEASE CATEGORY FREQUENCY fijj - COMPLEMENTARY CUMULATIVE DISTRIBUTION FUNCTIONS l., - fs Ws e NORMAllZATION OF RESULTS
- N 22 h
{1 e RELATIONSHIP TO SAFETY GOALS
,. t LJA i
0 - CORE DAMAGE f?. - EXPOSURE .C) y] - MORTALITY U l a w -
..] .
pIf ) PL M l* l 9il a m.---- _,. .- _ _ . , , . . . - . -- .- -
V mil F *;M21f35LVL M 2"H &h::dY MlL M M M W f M N
?NDOY%%$ . '
- g j ..
4 . Key Parameters of Sfgniffcant Release Categories I Release Release Energy l Time Duration Halogen (!) Alkali (Cs-Rb) Release Release k (hr) (hr) 0 Release (10 BTU /hr) rac+. ion Fracdon 7; _ 3 e PWR-1 2.5 .5 520 .7 .4 f PWR-2 2.5 .5 170
.7 .5 PWR-3 5.0 1.5 6
.2 .2 '
BWR-1 2.0 2.0 130
* .4 .4 BWR-2 30.0 3.0 30 . .9 .5 "
. BWR-3 30.0 3.0 20
.1 ..I Z-1 2.5 .5 l 520 .7 .1
; 2 2.5 .5 250
.7 .5 2R 12 3.0 . h 20 .7 a
.5
! 9.
L-1 2.0 .5 40 .4 f L-2 7.0 .4 2.0 36 '
.11 .09 L-3 1.5 2.0 .3 .26 L-4 1.5 .2 f, 2.0 .3
.73 .7
; L-5 1.5 2.0 .3 y
.07 .09
\ e i :
' j'-
s f i. ,
e ny w,en.neosmwmovmummmazwavmmumnasacnmmecscames: snstgryg;cpanysmym .M - Q
, q .. . .
a r.; m 1 , 3 ['d. , Revised Release Classification Porceneters - vr N w q -.
% Release Froction I t d
$j4 1 l H - High - 0.10 or more al
'i L - Low - 0.01 to 0.09
., VL - Very Low . Less than 0.01 4
/
1 i;] Release Stort Tirne
*dt
- 3 s
?1 E - Early - Less than 4 hours .y D - Delayed - 4 to 9 hours 1s .
L - Late - 10 to 19 hours
} VL - Very Late - 20 hocrs or more N
y. G
!$ Release Duratica .T te ge jj] S - Short - 0.5 hour
',j M - Mid - 1.5 to 4.0 hours
>j L' - Long - 10 to 12 hours Id 4,.
'M.
,h.,
A l} p -
.s <1 ih
%3
~
l1 A4 hj 31
,,1 }J a e
it ._.: - - -,.. - - , - _ - . - _ .. ~ . - . . - - .-
y"w m W 2h.x w w W m EWin w vaGD.twRFO?m3.*. p, . f ., , e ?% . )::* 3 ), O. 1. t:
'i .i .
3 :, -
..t 4
e 7 *2 . *: E 5 2 ::'r- .- E .7: -- 5 .:i*h e q ;* P > a
,,J 1
W 80 c- s:
- -ej B. s-
,1 v, i s. ..
Q., gg Ea I.r. ra. es r Q . 6 . .n
. er
%.3.. e e-r;
.:.w
,2~
e.:.:
- d. 9:n rs
?r
=:e p:
. 2se w e
- 1 A'n m . i .
' '4 9 d i d ,.' e - p Pp f Y YY .? .4 h 'i.,ff L,1- .x Se Q
I
.ns 1r n!!gI P= 8
~
r= :~ Il
~~
s2 il tt K ~:! *i.
= ;. ce
- r. <
k b O e. r..:A) c f 9 o y1 . ji
- E 2.
d TP P pp i I * *1 1 k P4 ~ =P.y oss ~ "t m e.4 .- y yH';
;, s.
a A s= e r 15., o g P.>. o .J M *
** e ..
9 v . -: 33 u 8 % -:*
,1 - ~ E : n .: .... s. .. -. -.
G
, ,h, - c 1 < o 1
- 4. .; LD T{
71
'11 5 '.rY
'. ? r-si r? i
;y u=
. 3 M 11. ~d =+
~e a n a .
+u .3 C x a x x a x x x x x x x .a a s 1
.ao-
./. 28 O -
;] 93 . . . . . v v . . . . . . e
.;. r-j, t"f 3 m = z z = = z a 3 s s W W W W W v:n* 2 *
,4 e : -
' I. 3 = = m = z z = = s s s .i W 4 W W 44
'Ji
.s ..
h- "- - _- a __ g vb ,,,
S . - O :.1 5 U i Nc.ilA312fl :DiL
... ?
9 1 3. Key Accident Sequences
~
Sequoyah Zion Peach Bottom Grand Gulf l Initial Set SH 2 1 TW , T230 S0 2 2 - TC T gQW V 5 TQUV T23C S HF 6 SE ' 2 g SI T23E 7 TQVW T 23 PQI i TB y3 B 9
- 13. T gQUV 10 T yPQI .
14 - i i [f [ Additional [Q Sequences - j SH g 16 AEG None
, SD g S C0Y ,
( THF - i - f' f .
_ m z z ;. aid D.i 2* "dIEN
~
b, SEQUOYAH XEY DOMINANT SEQUENCES
, ;t
- 5 Core Release Category Frequency d
Damage frequency 1 2 3 4 5 L RSSMAP Results* 5.7 x 10-5 1.8 x 10-7 1 x 10-5 2.6 x 10-5 1.6 x 10-5 4 x 10-6 Dominant Core Damages ' Sequences 5.7 x 10-5 1.8 x 10-7 1 x 10-5 2.64.x 10-5 1.65 x 10-5 4.2 x 10-6 Key Dominant (initial Set) 3.9 x 10-5 3.4 x 10'9 1 x 10-5 2.33 x 10-5 6 x 10-6 0 K2y Dominant Sequences as % cf Total 68% 2%- 100% 88% 36% 0
-i
?
c p Additional Key Sequences . 4 SHy 1 x 10-5 1 x 10-7 [ 1 x 10-5 T SDg 4 x 10~0 4 x 10-8 4 x 10-6 F
! Key Dominant as % total, with additional sequences 93% 80% 100% 88% 97% 95% ,
.i,.
g >d I
' p
*As calculated from Table 8-5. RSSMAP study, page 8-17. a R
I <
- 1
- c'- d
. ,, j
tno ,g.. .. t
> . _ . = . , c yq.ss
> r,
,,g._. . i a . _ g.y - - ,,~ gyg e>1 ; c, l< r-
- 4 i'r
,y J s ' . * 'o e y~. .u d - e.
o e. e e. e
= = . . %..] 3 . . = = 2
- e. o e.
- t'1 . e m.
n
- e e . -----
e 7e 7e . 2 -
.S. , 5 m a = 3
= n
*l - m ~
.y{ e' d d ss^ S S S
,. h' . = a = = = g g
;.s1q . .
c e
=
pj
+ . .. .
e
- o.
e y) i- * = = = 2 2 ,Jld - 7
,1 .. . .
'.,1 - e e . e
. - - - . w
. a rpret' . . m = . ..
e e. n n a e L f' . . s 4 . e o .
.ge . - .e.
g (, w g E = = m U e e. . E 2 4 4 4 w 6 e
. e, .e 2
W m. e e e w . s
.y. m . e a m .m.
e e
!< F- *t m
9 9 a x ,
?. < o. e. .
.v 2:e e e e g
l n 2 = = . g 3 y)4, s- 5 e. e. p
.rI w
g e e e
# . e.
. ..i, y W a = a JL z = = a 3 0 -: 3 2 3
} 13 .
N - - -
;] - ,.& . . . e e m, . = = = . .
c
.~ ~ ~
h.YI * ,q 9 m . - .
&W. e
'o
'e e
,p 4 &.
= . . -
e
= . . -
o et ~ u . * . e.
%) w ~. ~,. .
.e , e.
b yg . ,t-y9 s 7, I - I'*** .
.., e y
g
. 1-
. s. .
.t-
~ * "
M. V
= =
r, r 1 . *
. y b
~ .
- . y & = .
t
. ?. y Y & Y= !-- -
t. a 7 b 9 T
. s2 . g
- 9 W
+ .84 .
. e4 -
4J 1 e4 e
^*..J .I ,
w . 6
. m b .*
fI h %d 3 e 8'* 3
- 2
- e. em a. .
a g, 9 b.
.- g ",., j
'yh e -
$ e
.* f ,**
m
, y.
.m. e O"*
gd . .$ M ti 't e
* - g -- .* -m a . , _ . __
meewei.w93. e p
mmm29.WJWMERES2iliffEI2Y # J Peach Bottom. Key Dominant 5,equences . j -
, {i.
} <
Core Release: Category Frequency - Damage Frequency 1
- 2 3 4 r
WASH-1400 Study Results y 2.1 x 10-5 3.1 x 10'I 3.2 x 10-6 2.1 x 10-5 3.5 x 10'9' Dominant Core Damage Sequences 2.1 x 10-5 3.1 x 10~7 3.2 x 10 2.1 x 10-5 0 Key Dominant Sequences j (initial Set) i 2.08 x 10-5 l 3.07 x 10~7 3.13 x 10-6 2.05 x 10'5 0 Ksy Dominant Sequences. ( as % of Total 98% .-
?
99% . 98% i-98% , 0 ". <
~
1 Additional Xey Sequences i i C r l AEG F 7 x 10-10 7 x 10-10 g E. Key Dominant as % Total, .. with additional sequences 87%' 09% 98% 98% 20%
,?.
i t fiw 4 Uri 1 e, ll e
M z M GidI2iN"P7JPLT@2f?Jil D $ O
- l. 4 I
GRAND GULF XEY DOMINANT SEQUENCES (*, I i '
)
Core Release Category Frequency Damage Frequency 1 2 3 4 RSSMAP Results* 3.5 x 10-5 1.1 x 10~ 3.4 x 10-5 1.2 x 10-6 1.4 x 10-6 . Dominant Core Damage Sequences 3.5 x 10-5 1.2 x 10~7 3.2 x 10-5 1.3 x 10-6 1.6 x 10-6 [g Key Dominant Sequences (initial Set) 3.4 x 10-5 1.1 x 10-7 3.2 x 10-5 1.0 x 10-6 l'.2 x 10 -6 Key Dominant Sequences, as % of Total 97% 100% 94% 83% 85% e
~
) Additional Key Sequences NONE ---- ---- ---- ---- ---- Key Dominant as % of total, with Additional Sequences 97% 100% 94% 83% 85% i. i ;
*These values may be lev than the values for the dominant core damage sequences. ;[
This is due to round-off error. ? i y
l , 11/18/81 T b to (2 OUTLINE OF SEVERE ACCIDENT DECISION-MAKING PROCESS m This outline attempts to define key elements of the entire severe accident decision-making process in order to' focus planning efforts for the severe accident technical studies. A. Objectives of the Severe Accident Decision-Making Process
- 1. To ensure stable, technically well-founded, and cost-effective severe accident regulatory criteria which afford a balance among accident prevention, management, and mitigation, and
- 2. To evaluate existing regulatory criteria, and develop new criteria, if needed, in a stable, predictable, and expeditious process.
B. Key Elements of The Decision-Making Process - PHASE I
- 1. Use of the term "rulemaking" process is a source of confusion. It is preferable to conceive of the process as a " decision-making" process which may, but will not necessarily lead to a rulemaking proceeding per se,. This would depend upon the results of NRC Phase I and IDCOR studies.
- 2. It is convenient to consider the decision-making
.- process in two phases. Phase I defines what is needed technically and procedurally. Phase II consists of implementation of Phase I decisions.
M
2
- 3. The severe accident decision-making process must not be a vehicle for broad-scale regulatory reform. .
The decision-making process must be focused upon regulatory criteria for severe accidents only.
- 4. The existing body of deterministic regulatory criteria should not be changed by the severe accident decision-making process -- by either addition or deletion of existing regulatory criteria -- unless specifically justified on technical and/or cost effectiveness grounds. At this time, there are no specific areas which IDCOR has identified which require change.
Obviously, it will be important to be alert for candidate areas. ,
- 5. The IDCOR program and NRC Phase I severe accident technical studies will be available by mid-1983.
The technical subjects which should be considered as the basis for decision-making on severe accident regulatory criteria are those described in the IDCOR ProFran Plan.
- 6. The decision-making process must have a well documented technical basis, which would include the IDCOR and NRC Phase I studies. A well documented technical basis is necessary to assure permanence and stability to decisions on severe accident regulatory criteria.
7 . ," The basis for assessing whether new regulatory criteria are needed shall consist of performance .. 6
-3.
criteria (safety goal). These performance criteria shall include: 1) primary criteria for individual, and population risk, and 2) secondary criteria for assessing cost-benefit balance of measures to ! achieve incremental risk reduction. In regard to i potential secondary criteria, it is neither necessary : nor technically logical to specify a core damage l frequency griterion for the purpose of the severe accident decision-making process.
- 8. The Phase I decision-making process will result in a threshold determination as to whether new or modified regulatory criteria are needed, and if so, would lead to a rulemaking proceeding to promulgate any such i regulatory criteria. If it is determined that none !
are needed, arulemaking proceeding to codify that decision may or may not reccle. Any such rulemaking proceeding would be-informal (See II C.1. below). , C. Key Elements of the Decision-Making Process - PHASE II
- 1. Any rulemaking proceeding must be informal in format (Notice and Comment; any hearings as may be held will be purely legislative).
- }
! 2. ,Dependi,ng upon the resu i ts of the Phase I studies, aserihsofregulatoryoutcomesorscenarioscanbe , t l defined. Five are defined below for discussion s .- f, ,rs. - -
- Le . */ It may be a valuable exercise to reduce 2.a - e. to a logic m ,: diagram. ,
t i
? Y f
~
i %
.4
*/
purposes as 2a. - e. The five (a. - e.) are:
- a. The Commiss' ion determines, on the basis of the ,
Phase I studies, that: 1) compliance with existing deterministic regulatory criteria will.- provide a level of safety which satisfies the primary performance criteria on a generic basis,
- 2) no changes in existing regulatory criteria are needed' The Commission issues a Notice terminating the process initiated by the Advance Notice of Rulemaking (ANR), explaining the basis for its decision, and to the latter end, incorporating the NRC and IDCOR Phase I studies by reference.
- b. The Commission makes the determinations described in 2.a.1) and 2.a.2) above, and also proposes a severe accident rule which provides that no additional severe accident regulatory criteria are needed.
The Commission issues a Notice of Proposed Rulemaking and proceeds to a conclusion per 1 above.
- c. The Commission makes determinations a.1) and l a.2), proposes the rule as per b. above, and also proposes the following exceptions and conditions i
to the general rule that no additional severe accident ! regulatory criteria are needed: e l l ii -wm--c. p-w -- -- r TT'-- '-'- T % n 'W== w --" T"'9- Ti f v -' T-' ' ?-* *-"**
., ._- _ ~ . . - . _ . -
5 1.) Individual licenses or license applicants - of a plant having unusual siting conditions or unique design features, such that there is substantial doubt as to whether the rule would be valid in its application to that plant, would be required to perform plant-or site-specific PRA studies to determine compliance with the primary performance criterion. (Alternatively, the Commission could rely on its existing 10 C.F.R. 52.758 standards for waiving rules only when the purpose for which the rule was adopted would not be served.) Any PRA studies carried out pursuant to 2.c.1) would be subject to certain prescribed criteria (see 2.d. below), and would be evaluate'd for acceptability in accordance with the primary performance - criterion (See 2.c.2). 2.) If the licensee or applicant having unique design features or siting conditions shows compliance with the primary performance t criterion pursuant to 2.c.2.) above, the analysis need not be extended to consider ( i
~
t ~ l 1
the secondary goal. In that instance, and , also for any other individual plant docket, '~~T no additional requirements would be imposed, unless t,he NRC Staff or a party to an individual licensing proceeding first shows that the-marginal reductions in residual risks associated with specific measures are justified ! with respect to the secondary criterion. In the event of such a showing by the NRC Staff or party, the liceneee or applicant , would be afforded the opportunity to show the contrary.
- d. The Commission either makes the determination in 2.a.1) above, or does not, but nevertheless determines that there is sufficient uncertainty in the analysis to preclude a valid generic determination that no additional regulatory criteria are required. In that case, the Commission would issue a proposed rule which l
would establish a generic framework under which individual plants.could be evaluated for c severe accident risk. The framework would consist of: 1) the performance criteria, and 2) a prescribed
-methodology (assumptions, sequences, methods of e
m l l
-7 r
analysis, etc.). In fashioning this framework, . special care must be exercised to avoid overly , prescriptive criteria, to minimize the calculational burden on individual licensing dockets, and to assure flexibility to accommodate new information. In individual licensing dockets, the individual plants would be required to show~ compliance with the primary performance criterion using methodology which complies with the prescription set forth in the rule. The validity of the performance criteria and methodology would not be subject to attack in individual 1(censing proceedings. Only iss'ues of compliance would be open. Further, the need to consider the secondary performance criterion would be subject to the conditions set forth is 2.c.2) above.
-e.
The Commission decides that more research is needed before a meaningful rulemaking can be undertaken.
- 3. The scope and future effect of any rule will follow from the underlying technical subject matter which supports the decision-making process. If a given subject is not within the scope of the Phase I studies, it would not be the basis for the rule, and it would be open for consideration in , future individual licensing cases.
4 Unfess specifically identified as high priority pay-off
~
l items (large increase in safety or decrease in cost), _ i V
-8 1
all existing rules and regulatory criteria would be left i undisturbed by the severe accident rule (SAR). There must be -
~
an urgent need for change if an item is to be included in the SAR. The severe accident decision-making process may identify a number of areas for which change may be needed, but those changes -- unless urgently needed -- would be undertaken and implemented by the NRC in separate individual licensing or generic actions, as appropriate.
- 6. The technical subjects now incorporated in the IDCOR program plan define the appropriate basis or scope of information to support a severe accident Phase II decision. It is important to note that many technical subjects which also have a distinct identity as prominent regulatory decisions would be dealt with as sequences or phencmena which affect sequences, but not as regulatory decisions per se. For example, LOCA sequences would be considered, but Appendix K would not. ATWS sequences would be considered, but the regulatory decision on ATWS would not. Pressure vessel overpressurization would be considered, but fracture mechanics requirements would not.
Accident management would be considered, but detailed human factors considerations, and emergency procedures per se would not. This list of examples can be extended, and the principle (include as a sequence or a phenomenon which affects a sequence, but exclude as a regulatory decision) should be appljedasageneralrule. a 0
,y, - , _ . . , , , - - - _ _ , _ , _ - . , _ .,_ .,
_9
. 7. There are two technical subjects which do not fit nicely within the foregoing general rule. They are: sabotage and external-events (earthquake, flood, tornado, etc.). A .
suggested approach to each is defined below.
.8. Sabotage should be excluded from consideration. A saboteur can cause severe accidents which are no different in kind from the severe accident sequences which will be considered in the analysis. Although the actions of a saboteur could 2
affect the probability of those sequences, these probabilities are the focus of attention under the existing body of deterministic regulatory requirements. There is little doubt that a plant meeting the existing body of regulatory 1 requirements is robust against sabotage, and one must hypothesize extreme values of a sabotage threat before there is an appreciable perceived risk due to sabotage. In addition, it should be emphasized that the severe accident decision-making process had its genesis in the' _ , ! TMI accident, and that the TMI incident did not bring the adequacy of the existing regulatory requirements for sabotage into question. The analytical tools for performing PRA's in regard to sabotage are limited, and it seems unlikely that one could at this time extend the analysis beyond the point discussed herein to a more specific set of generic conclusions. It thus seems logical to exclude sabqhageonthebasisoftheforegoingconsiderations,
~
and defer further consideration unless and until: a) $- specific evidence exists to bring the existing deterministic _ requirements into serious question; and b) ongoing research
. -10 yields more appropriate analytical tools. In the meantime, there should be no need to consider these issues in the context of individual plant severe accident reviews.
- 9. External events should be included, with considerationI as described herein. As with sabotage, it should be emphasized that there was nothing in the TMI sequence which brings the adequacy of existing regulatory requirements concerning external events into serious question. Nor is there any doubt that plants meeting these requirements are robust. In certain instances, however, external events could be dominant risk contributors, particularly where normally dominant sequences ,have been suppressed. Several
-recent PRA's reflect these considerations. Ist the same time, it would appear that the risk associated with external events does not become significant unless extreme values for those events are hypothesized. In many instances (e.g., extremely severe earthquake devastating an entire area including a nuclear plant), the population risk from the plant is very small compared to the total risk associated with the event. Since the analytical tools for performing PRA's are limited, the results will in many instances be site dependent, and IDCOR's existing resources are limited, it will be difficult to undertake extensive analyses which yielB a specific set of generic conclusions. Taking into account all circumstances noted above, IDCOR will incorporate -
the results of existing PRA's concerning external events
O 6 -11 . 4 in its analysis, and provide a threshold analysis of the .
~
potential contribution to risk from external events. -
~
IDCOR will treat external events with the foregoing threshold analysis, and recommends deferring detailed consideratEon of those issues unless and until: oa) specific evidence exists to bring the existing requirements into serious question; and b) ongoing research yields more appropriate analytical tools. In the meantime, there should be no need to consider these issues in the context of individual plant severe accident reviews, unless there is a substantial showing that the risks associated with these issues would
- predominate for a particular plant, and that the severe ,
accident rule would not-be valid for that particular plant.
~
- 10. The s'o-called minimum ESF and degraded core decision-making processes should be combined into one process.
From a technical standpoint, this is the most logical approach. From a regulatory standpoint, the decision-l making process described herein would obviate the need L for a specific and prescriptive minimum ESF rule. This process does not contemplate substantial changes to existing regulatory criteria; rather, it seeks to determine whether additional criteria are needed, and if so, what 1 L criteria. It does not contemplate the wholesale l subtraction of ESF's as a result of individual risk studies. e w- - r wg y--,---, ,,-w.-- w w- wy y- e- - - - - . . - .- - - + . - - - - - -weev-w--,.m-.y-ww --mt----*-w--,
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.-12 4
- 11. There is considerable cause for concern that NRC actions _
in related individual licensing dockets or rulemakings, which can affect the scope and nature of the degraded. core decision-making process, are being taken without careful regard for the future effect on that process. A recent example illustrates this point (numerous additional examples can be cited, such as'the McGuire H2 decision, ATUS deliberations, etc.). During the Commission's November 5, 1981, affirmation session on the Interim Hydrogen Rule a final draft rule was considered. There are two parts of the statement of considerations of that rule which will at least operate as a source 6f confusion for the degraded core decision-making process. These are:
- a. "The Commission will be considering futher modification of 550.44 during the long-term rulemaking effort relative to consideration of degraded or melted cores in safety regulation. Part of this long-term rulemaking will involve a thorough reevaluation of hydrogen generation and control." This could imply that the severe accident rule will develop a new 550.44, which will include detailed requirements for o
E w -- - - r e
/ .
13 .
~
hydrogen control. The severe accident decision-making process should consider phenomenology of hydrogen- . -+__ : generation, but it would not necessarily consider methods of control in detail. It may be that the. severe accident decision-making process will identify the need for additional regulatory criteria or methods of control.
- b. "Several codmenters have expressed concern that the various rulemakings currently being pursued by NRC should be integrated, i.e., safety goal, degraded core considerations, minimum engineered safety features, siting and emergency planning. The NRC shares this concern. On October 13, 1980, the Executive Director for Operations established a Degraded Core Steering Group to coordinate degraded cooling and related rules. This group has completed its work and prepared a plan to ensure furture integration of these activities." This statement implies that the Steering Group's plan will govern the relationship between the several rules. Of necessity, this would have the plan govern the planning and scope of the degraded core decision-making process. It is doubtful that this accurately portrays current i - NRC management thinking.
1 m 7 y % w ,- .9'- - - - - --w- -- iy,-e,:-ytw--- -c,g-,---ge-r1y ye,w- wa y,w ge mp,y ----w -,-w wga-w+--r ? e yM e*-* ew --'www -- e
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4 TECHNOLOGY for ENERGY CORPORATION 9 THE IDCOR PROGRAM MARIO FONTANA, DIRECTOR TONY BUHL, VICE PRESIDENT IDCOR POTENTIAL CONTRACTOR INFORMATION MEETING MAY 21,1981
~
HYA1T-REGENCY KNOXVILLE, TENNESSEE
i .. i TECHNOLOGY for ENERGY CORPORATION l THE ISSUE OF DEGRADED CORES IS CRUCIAL ! TO THE FUTURE OF NUCLEAR ENERGY A SEVERE ACCIDENT WdVLD CAUSE PUBLIC REACTION j-
- A COSTLY ACCIDENT WOULD AFFECT AVAILABILITY OF l FINANCIAL RESOURCES l
- UNREAllSTIC REGULATORY REQUIREMENTS COULD l INCREASE COSTS AND UNCERTAINTIES l AN AC'CIDENT ANYWHERE IS AN ACCIDENT EVERWHERE i
9
V TECHNOLOGY for ENERGY CORPORATION THE ISSUE OF DEGRADED CORES WILL BE ADDRESSED WITHIN THE NEXT 2 TO 3 YEARS l l
- CHANGES IN DESIGN AND OPERATIONS WILL BE ADDRESSED '
= NRC RULEMAKING HEARINGS ARE LIKELY
- A REALISTIC APPROACH IS NECESSARY
5 . TECHNOLOGY for ENERGY CORPORATION l .
! ISSUES OF IMPORTANCE TO NRC
/ RULEMAKING l, PHENOMENOLOGY:
- HYDROGEN GENERATION, BURN
- REACTOR VESSEL FAILURE CHARACTERISTICS a
FUEL / WATER INTERACTIONS f CORE / CONCRETE INTERACTIONS
- CONTAINMENT FAILURE MODES l
- SEVERE CORE DAMAGE
!
- RADIOLOGICAL SOURCE TEAM i SYSTEMS ANALYSIS:
- REACTOR SYSTEMS SIMULATION
- CONTAINMENT SYSTEMS SIMULATION
- RELEASE TO THE ENVIRONMENT
= EQUIPMENT SURVIVABILITY g 0
- i. s.
l 1 TECHNOLOGY for ENERGY CORPORAllON i i . ! MITIGATIVE FEATURES BEING CONSIDERED ! BY NRC i !
- CONTAINMENT HEAT REMOVAL I
4
- CONTAINMENT MASS REMOVAL
!
- INCREASED CONTAINMENT VOLUME
- INCREASED CONTAINMENT PRESSURE CAPABILITY i
- COMBUSTIBLE GAS CONTROL l-
- CORE RETENTION DEVICES ,
l
- MISSILE SHIELDS
- ADD-ON DECAY HEAT REMOVAL SYSTEM l
4 I
~* *T +-- T - ----
P _ _- - _ _ _ - - - _ _ _
4 TECHNOLOGY for ENERGY CORPORATION THE NUCLEAR INDUSTRY HAS STARTED THE IDCOR PROGRAM TO
- ADDRESS THE DEGRADED CORE ISSUES i
- ANALYZE THOSE ACCIDENTS THAT COULD LEAD TO SEVERE CORE DAMAGE
- ASSESS AND RECOMMEND IMPROVEMENTS IN DESIGN AND OPERATIONS FOR PREVENTION INTERDICTION MITIGATION
- PREPARE A TECHNICALLY SOUND INDUSTRY POSITION e
*' e . ,,
I r
! TECHNOLOGY for ENERGY CORPORAllON 4 ; IDCOR PROGRAM ORGANIZATION i 1 I IDCOR STEERING ATOMIC GROUP -----------------
INDUSTRIAL 7
- CORDELL REED, CHAIRMAN FORUM J. SIEGEL g
\ TEC RESPONSIBLE
- OFFICER
! A.R. BUHL, VICE PRESIDENT ) i IDCOR IDCOR PROGRAM OFFICE IDCOR ) TECHNICAL M.H. FONTANA, PROGRAM DIRECTOR LEGAL l ADVISORY GROUP R.D. MOORE, PROJECT MANAGER GROUP I M. LEVERETT, CHAIRMAN i ) SENIOR l . CONSULTANTS H.K. FAUSKE N. RASMUSSEN i R. SEALE W. STRATTON l ! I I I I . REACTOR 8: PLANT REGULATION CONTRACTS RISK ANALYSIS H NOMENA & LICENSING ADMINISTRATION SYSTEMS S.V. ASSEllN M.H. FONTANA C.R. NAULT E.P. STROUPE R.M. SATIERFIELD
TECHNOLO3Y fer ENERGY CORPORATION , TEC TEAM MEMBER RESPONSIBILITIES - -- ;
- M.H. FONTANA, PROGRAM DIRECTOR -
i
- OVERALL DIRECTION OF PROGRAM- .
TECHNICAL COST AND SCHEDULE PERFORMANCE
- A.R. BUHL, VICE-PRESIDENT, TEC SENIOR TEC CORPORATE MANAGER RESPONSIBLE TO STEERING GROUP FOR THE IDCOR PROGRAM.
i RESPONSIBLE FOR PUBLIC INFORMATION AND IDCOR l INTERACTION WITH LEGAL ADVISORS l
- R.D. MOORE, PROJECT MANAGER RESPONSIBLE FOR CONTRACTOR PERFORMANCE REVIEWS AND TEC TECHNICAL MANAGEMENT REVIEWS.
l l
- S.V. ASSELIN, MANAGER-RISK ANALYSIS RESPONSIBLE FOR DEVELOPMENT OF TECHNICAL DETAILS OF RISK ANALYSIS PROGRAMS AND MANAGEMENT OF CONTRACTOR TECHNICAL PERFORMANCE
- E.P. STROUPE, MANAGER-REACTOR AND PLANT SYSTEMS RESPONSIBLE FOR DEVELOPMENT OF TECHNICAL DETAILS OF REACTOR AND PLANT SYSTEMS PROGRAMS AND MANAGEMENT OF CCNYRACTOR TECHNICAL PERFORMANCE
- M.H. FONTANA, MANAGER-PHENOMENA ANALYSIS RESPONSIBLE FOR DEVELOPMENT OF TECHNICAL DETAILS OF PHENOMENA ANALYSIS PROGRAMS AND MANAGE-MENT OF CONTRACTOR TECHNICAL PERFORMANCE
- R.M. SATTERFIELD
- RESPONSIBLE FOR REVIEW AND COUNSEL ON '
REGULATORY / LICENSING ISSUES AFFECTING IDCOR
- C.R. NAULT RESPONSIBLE FOR CONTRACTS ADMINISTRATION
~ ~ , -
j i l THE IDCOR PROGRAM MUST ADDRESS THE DEGRADED CORE ISSUES IN A BALANCED,
- INTEGRATED, AND REALISTIC MANNER
= MUCH PREVIOUS WORK HAS BEEN NARROW OR SUPERFICIAL l
= ULTRA-CONSERVATIVE ESTIMATES ARE INADEQUATE AND l
! UNNECESSARILY RESTRICTIVE .
=, ISSUES MUST BE RESOLVED IN CONCERT WITH NRC AND OTHERS i .
- PROGRAM MANAGEMENT IS THE KEY TO SUCCESS I
QI PA3812-lO-I
.g O
IDCOR STRATEGY :
- DEFINE SAFETY GOALS 'i i
- 16ENTIFY DOMINANT SEQUENCES THAT INVOLVE SEVERE CORE DAMAGE
- DETERMINE REAllSTIC REACTOR BEHAVIOR THROUGHOUT THE COURSE OF THESE ACCIDENTS ,
i
- IDENTIFY AND EVALUATE POTENTIAL PREVENTIVE AND CORRECTIVE ACTIONS AND TIME WINDOWS AVAILABLE a RELATE RESULTS TO SAFE 1Y GOALS
- PREPARE TECHNICALLY SOUND POSITION gini.imi
l 1
- IDCOR SPECIFIC OBJECTIVES
~
i i
- IDENTIFY AND ASSESS ACCIDENT INITIATORS, THEIR PROBABill1Y OF OCCURRENCE, AND THEIR POTENTIAL FOR j CAUSING SIGNIFICANT DAMAGE
~
)
- IDENTIFY KEY SEQUENCES FOR IN-DEPTH ANALYSIS i
j
- ANALYZE KEY SEQUENCES WITH RESPECT TO l - DRIVING FORCE PHENOMENA
) - FISSION PRODUCT PHENOMENA l - TIMING OF KEY EVENTS l - PLANT DYNAMIC RESPONSE
- SYSTEMS NTERACTIONS l - EQUIPMENT PERFORMANCE l - OPERATOR PERFORMANCE
!
- COMPARE l(EY SEQUENCES WITH WHAT THE OPERATOR l PERCEIVES IN THE CONTROL ROOM l t
1 s q . i IDCOR SPECIFIC OBJECTIVES (CONT.) ! r i
- IDENTIFY PREVENTIVE AND CORRECTIVE ACTION KEYED TO 4
TIME WINDOWS, IDENTIFY REQUIREMENTS FOR lMPLEMENTATION, AND ASSESS SIDE EFFECTS l
- EQUIPMENT REPAIR
- - OPERATOR ACTION i - OFFSITE EQUIPMENT
- EMERGENCY ACTION l
- IDENTIFY SAFE STABLE STATES l' l = IDENTIFY INHERENT FISSION PRODUCT RETENTION '
I PHENOMEN.A ;
- EVALUATE EFFECT ON RISK OF PREVENTIVE, CORRECTIVE, AND i MITIGATING ACTIONS AND PHENOMENA
= PREPARE TECHNICAL POSITION FOR RULEMAKING
S k \ l ACCIDENT l i XCp%, E :
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i
' SELECTION I ,i
; g
. SELECT DOMINANT Spd het M ~s~ier'5- 6:- #
j SEQUENCES l C4 h ?ysy;? 4- p :y,p ._,__g .-; -g ; Y SELECT l l i! PHENOMENOLOGICAL I, c
- N EQUIPMENT / SYSTEMS ;
iptmaMi
.~
l
- i. i} 3
} - PHENOMENA I{ H2BURN CONTROL Y STEAM t l ~ OVERPRESSURE "" l '
'l l ]
CONTAINMENT y
}
- STRUCTURAL g l
f CAPABILITY i I g
} . HO2 IRATION ' j
'}~~
I
- EQUIPMENT G
SURVIVABILITY : tid . CORE DEBRIS ' _ BEHAVIOR & *- l l COOLABILITY "" a 11 g, N - l OPERATIONS CORREMI MM gN PORT *- _
& INHERENT RETENTION -
M Ilgg ' 7 INTEGRATED ANALYSIS IDCOR EXECUTIVE ANALYSIS PROGRAM - INSTRUMENTATION REeulREMEMS OPERATOR ACTION : - ' T l CONTAINMENT ANALYSIS Ik1 ACE ATMCSPHERIC
~ ~ ~
PA A SE ggig e
.' ALTERNATIVES :. - - - A
~ T= 7" C E '
HOW SAFE IS SYSTEMS i ig py i i -I g SAFE ENOUGH g l W SAFETY GCAL/ *~ CORE RETENTION b
- I l CRITERIA . DEVICES -
b d,i -
! -- d h RECOMMENDATIONS C T EFA R WHAT SHOULD BE DONE l RISK REDUCTION
- b l g POTENTIAL i
=~
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4 b h A e' bb.cbl
TECHNICAL TASK SCHEDULE 0 3 G 9 12 15 18 21 24 I . .l . . l . . l . . i . . 1. , I . , I , i l
- 1. SAFEIY GOAL /CR11ERIA APPLICAllONS
+
* *
- O Y 2. SELECilON OF DOMINANT SEQUENCES 2 2
. 3. SELECilON OF CONTAINMENT PilENOMENOLOGICAL I IT ' O SEQUENCES
- 4. SIEAM OVERPRESSURE PilENOMENA -
T ' 4 (IN-VESSEL CONIAINMENT) f 5. IlYDROGEN GENERAllON AND BURN T 4
- 6. SURVEY IfYDROGEN BURN CONIROL T 4 23 s 4s 4
% 7. EQUIPMENT SURVIVABilllY FOR ENVIRONMENT 4I I T 4 w w 4 (DEGRADED CORE) h '8. CORE DEBRIS BEllAVIOR AND COOLABillIY 18 34s 4s 4 s
- 9. CONTAINMENT SIRUCTURAL CAPABillIY I I I I ' O s
23ss 4 e ie
-t 10.EVALUA1E ATMOSPilERIC AND LIQUID I I I T O PAlllWAY DOSE 3 e
- 11. FISSION PRODUCT tlBERAllON,1RANSPORT, AND _
I I T 4 C INIERENT REIENilON (INIEGRATE WilH DOE /EPRI) 23 4s 4 il I I I I ' O
- 12. ALIERNA1E CONTAINMENT SYSTEMS 3
e v 4
- 13. CORE RETENilON DEVICES ius e i 1 2 3 s2s as a 4 7 e ss is to assisi 2
-h(14. RISK REDUCilON POIENilAL 2 as 4s 4 sin s
-K'15. INTEGRATED MODEL DEFNilON AND ANALYSIS . 2 2 2a 3 3e 4 7 s es 4 Als 7:e 4 7 s u sses
- 16. OPERAllONAL ASPECIS OF ACCIDENT MANAGEMENT I I I I I~ I I I II'I'
- AND CONIROL y DRAFT REPORI 4 FINAL REPORT
TECHNdi.OGY f@r ENERGY CORPORATION t TASK RESPONSIBILITIES ARE: . RISK ANALYSIS - S.V. ASSEllN TASK 1. SAFETY GOAL / CRITERIA APPLICATIONS TASK 2. DOMINANT SEQUENCES TASK 3. CONTAINMENT PHENOMENOLOGICAL SEQUENCES TASK 14. RISK REDUCTION POTENTIAL TASK 16. OPERATIONAL ASPECTS OF ACCIDENT MANAGEMENT AND CONTROL , i. 39PA3920e
i ., iECHNOLOGY fer ENERGY CORPORAllON l t l I . ! REACTOR & PLANT SYSTEMS - E.P. STROUPE TASK 6. HYDROGEN BURN CONTROL ! EQUIPMENT SURVIVABILilY TASK 7. TASK 9. CONTAINMENT STRUCTURAL CAPABilllY f ! TASK 12. ALTERNATE CONTAINMENT SYSTEMS i TASK 13. CORE RETENTION DEVICES
~
J 4 i l i i' . --
- y -
i . TECHNOLOGY for ENERGY CORPORATION 1 ! i i i PHENOMENA ANALYSIS - M. FONTANA i TASK 4. STEAM OVERPRESSURE PHENOMENA 4 TASK 5. HYDROGEN GENERATION AND BURN TASK 8. CORE DEBRIS BEHAVIOR & COOLABILITY i l
. TASK 10. ATMOSPHERIC & LIQUID PATHWAY DOSE TASK 11. FISSION PRODUCT LIBERATION,
! - TRANSPORT, AND INHERENT RETENTION TASK 15. INTEGRATED MODEL 6EFINITION
& ANALYSIS l
I I -
" / n . ' W-- -
W" TECHNOLOGY f:r ENERGY CORPORATION s. . [ ,- t PRIOR ,~., PRA'S _ us g
- J DOMINANT -
OVERVIEW : SEQUENCES 2.1,2.2 CONTAINMENT : PHENOMENA EVENT j DOMINANT - W ER r SEQUENCES -, PRA,S , y 2.2,2.3 - c PHENOMENA -i UPDAIE EVENT TREES : MillGATION ALTERNATES 3.2' DOMINANT SEQUENCES 2.2,2.3,3.2 RISK
~ POSliiON EVALUATION ,
2.2, 2.3, 3.2 , POSSIBLE gg 14.2 GOAL 1.1 POST-TMI CHANGES RISK 1 RISK REDUCilONS e BENEFIT ACillEVED POSHION POSR M 1.2 POSR M
TECHNOLOGY for ENERGY CORPORATION l TASK 1 SAFETY GOAL / CRITERIA APPLICATION OBJECTIVE: TO USE A QUANTITATIVE SAFE 1Y GOAL TO ADDRESS DEGRADED CORE RULEMAKING ISSUES SUBTASKS: 1.1 SAFETY GOAL DEVELOPMENTS 1.2 RISK / BENE' FIT CRITERIA 1.3 EVALUATION METHODS
l i
; TECHNOLOGY for ENERGY CORPORAllON i
i . e
~
! SUBTASK 1.1 , ! ADAPT SAFETY GOAL DEVELOPMENTS i i ! SUBTASK SCOPE: ADAPT SAFETY GOAL DEVELOPMENTS INTO IDCOR FRAMEWORK AND RELATE TO IDCOR POSITION PAPERS i ANTICIPATED MANPOWER: 2 MAN-MONTHS
! LATEST START DATErJUNE 1981 LATEST COMPLETION DATE: DECEMBER 1981 l
TECHNOLOGY for ENERGY CORPORATION i !
- SUBTASK 1.2 j DEFINE RISK / BENEFIT CRITERIA FOR
- ALTERNATE EVALUATION l SUBTASK SCOPE
PROVIDE A FRAMEWORK FOR ASSESSING SAFETY MERITS OF ALTERNATE PREVENTION AND MITIGATION SCHEMES ANTICIPATED MANPOWER: 2 MAN-MONTHS LATEST START DATE: JUNE 1981 LATEST COMPLETION DATE: DECEMBER 1981 o _m _ __ _ ___ _ _ - -- e-- . - _ . _ _ _ _ _ _ _ .
l . ! 1ECHNOLOGY for ENERGY CORPORATION - l l ! SUBTASK L3 ! DEFINE METHODS FOR EVALUATION OF l DEGRADED CORE CONDITIONS I - l SUBTASK SCOPE: l DEFINE GROUNDRULES FOR ANALYSES OF SELECTED IDCOR ACCIDENT SEQUENCES
- ANTICIPATED MANPOWER: 2 MAN-MONTHS 1
i LATEST START DATE: JUNE 1981 LATEST COMPLETION DATE: DECEMBER 1981 1 1
). .
e ,_, i i , TECHNOl.OGY for ENERGY CORPORATION TASK 2 - SELECTION OF DOMINANT SEQUENCES 1 j OBJECTIVE: . j TO DETERMINE DOMINANT CONTRIBUTORS TO PLANT RISK TO l PERMIT EXAMINATION OF POSSIBLE RISK REDUCTIONS SUBTASKS: ' 2.1 DEFINE DOMINANT SEQUENCES 2.2 ASSESS DOMINANT SEQUENCES 2.3 UPDATE ^ l i i hm *"* a __.
l .,- 1ECilNOLOGY for ENERGY CORPORATION , I t l SUBTASK 2.1 DEFINE INITIAL DOMINANT SEQUENCES 1 SUBTASK SCOPE: SURVEY RISK ASSESSMENTS AND COMPILE AN INITIAL j CATALOGUE OF DOMINANT ACCIDENT SEQUENCES ANTICIPATED MANPOWER: 6 MAN-MONTHS LATEST START DATE: JUNE 1981 LATEST COMPLETION DATE: OCTOBER 1981 \ 1 {. .-
wm w m m> _ TECHNOLOGY for ENERGY CORPORATION i SUBTASK 2.2 ASSESS DOMINANT SEQUENCES l SUBTASK SCOPE: REFINE THE CATALOGUE OF SEQUENCES AND ASSIGN PROBABillTIES AND POTENTIAL CONSEQUENCES ANTICIPATED MANPOWER: 21 MAN-MONTHS LATEST START DATE: AUGUST 1981 LATEST COMPLETION DATE: FEBRUARY 1982
1, ' 1ECHNOLOGY for ENERGY CORPORATION ? i ! SUDTASK 2.3 .
- UPDATE TO INCLUDE PARALLEL STUDIES i
i SUBTASK SCOPE: INTEGRATE RESULTS OF CONCURRENT RISK ASSESSMENTS INTO CATALOGUE OF SEQUENCES
- ANTICIPATED MANPOWER: 6 MAN-MONTHS
~
LATEST START DATE: OCTOBER 1981 . l LATEST COMPLETION DATE: FEBRUARY 1982 l
i -. , TECHNOLOGY for ENERGY CORPORATION l TASK 3 SELECTION OF PHENOMENOLOGICAL l SEQUENCES l OBJECTIVE: TO PROVIDE MECHANISM FOR EXAMINING PHENOMENOLOGICAL
- EFFECTS ON DOMINANT ACCIDENT SEQUENCES l SUBTASKS
j 3.1 DEVELOP CONTAINMENT TREES
- 3.2 UPDATE SYSTEM / CONTAINMENT TREES '
i 1 l 1 i j .
l, ,, TECHNOLOGY for ENERGY CORPORATION , I ! l SUBTASK 3.1 l DEVELOP CONTAINMENT EVENT TREES SUBTASK SCOPE: . DEVELOP CONTAINMENT PHENOMENOLOGICAL EVENT TREES TO EXTEND EFFORT IN SELECTING DOMINANT SEQUENCES , ANTICIPATED MANPOWER: 18 MAN-MONTHS LATEST START DATE: AUGUST 1981 LATEST COMPLETION DATE: JANUARY 1982 l l 1
l TECHNOLOGY for ENERGY CORPORATION 1 i
- SUBTASK 3.2 UPDATE OF SYSTEM / CONTAINMENT EVENT TREES SUBTASK SCOPE
INCORPORATE IDCOR PHENOMENOLOGICAL RESULTS INTO DOMINANT SEQUENCES l ANTICIPATED MANPOWER: 8 MAN-MONTHS LATEST START DATE: JUNE 1982 l LATEST COMPLETION DATE: OCTOBER 1982 l .
~
l ' l . 1' TECHNOLOGY for ENERGY CORPORATION i I TASK 4 t STEAM OVERPRESSURE PHENOMENA 9 ! OBJECTIVE: - l TO PREDICT THE RATE OF STEAM GENERATION DURING DOMINANT ACCIDENT SEQUENCES AND ASSESS STEAM EXPLOSIONS , SUBTASKS:
- 4.1 ASSESS STEAM PRESSURE GENERATION i
~
) i i e e
j TECHNOLOGY for ENERGY CORPORATION SUBTASK 4.1 ASSESS STEAM PRESSURE GENERATION SUBTASK SCOPE: DEVELOP MEANS TO CHARACTERIZE INTERACTIONS BETWEEN
. CORE DEBRIS AND COOLANT TO BE USED IN PREDICTING STEAM GENERATION. ASSESS STATUS OF PRESENT KNOWLEDGE
! WITH RESPECT TO STEAM EXPLOSIONS AND DOCUMENT l RECOMMENDATIONS ANTICIPATED MANPOWER: 24 MAN-MONTHS l LATEST START DATE: JULY 1981 LATEST COMPLETION DATE: DECEMBER 1982 O 00 900000 9
' ~~ ' ~-' "
TECHNOLOGY for ENERGY CORPORAllON l TASK 5 l HYDROGEN GENERATION AND BURN l OBJECTIVE: ! TO PREDICT HYDROGEN GENERATION AND THE SUBSEQUENT HAZARD FROM OVERPRESSURIZATION AND COMBUSTION l SUBTASKS: 5.1 DETERMINE RATE AND AMOUNT OF HYDROGEN GENERATION 5.2 DETERMINE HYDROGEN DISTRIBUTION 5.3 DETERMINE COMBUSTION LIMITS OF H2 -AIR-STEAM-CO2 l i l
j ' ., l' , , l-TECHNOLOGY for ENERGY CORPORATION ( 1 I SUBTASK 5.3 DETERMINE COMBUSTION LIMITS OF H2-AIR-STEAM-CO2 SUBTASK SCOPE: i DETERMINE FLAMMABILITY LIMITS (INCLUDING DEFLAGRATION ' AND TRANSITION TO DETONATION)'OF HYDROGEN IN CONDI-TIONS REPRESENTATIVE OF DOMINANT ACCIDENT SEQUENCES. INCLUDE EFFECTS OF STEAM, AIR, CO 2, TEMPERATURE, PRESSURE, AND SIZE. ANTICIPATED MANPOWER: 8 MAN-MONTHS LATEST START DATE: SEPTEMBER 1981 LATEST COMPLETION DATE: DECEMBER 1982
TECHNOLOGY for ENERGY CORPORATIQN i 1 SUBTASK 5.2 l DETERMINE HYDROGEN DISTRIBUTION SUBTASK SCOPE: DETERMINE THE POTENTIAL FOR STRATIFICATION AND POCKETING, INCLUDING EFFECTS OF PASSAGE OF HYDROGEN-RICH MIXTURES THROUGH SUPPRESSION POOLS, ICE CONDENSERS AND SPRAY CURTAINS ANTICIPATED MANPOWER: 20 MAN-MONTHS LATEST START DATE: SEPTEMBER 1981 LATEST COMPLETION DATE: DECEMBER 1982 l l
l TECHNOLOGY for ENERGY CORPORATION j . ! SUBTASK 5.1 . DETERMINE RATE AND AMOUNT OF HYDROGEN GENERATION
~
! t l SUBTASK SCOPE: FOR DOMINANT ACCIDENT SEQUENCES, DETERMINE THE RATE AND l AMOUNT OF ZlRCONIUM-WATER AND STEEL-WATER REACTIONS PRODUCING HYDROGEN FOR BOTH INTACT CORE GEOMETRY i AND CORE DEBRIS-WATER INTERACTION PROCESSES ANTICIPATED MANPOWER: 15 MAN-MONTHS LATEST START DATE: SEPTEMBER 1981 LATEST COMPLETION DATE: DECEMBER 1982
TECllNOLOGY for ENERGY CORPORATION I TASK 6 HYDROGEN BURN CONTROL , OBJECTIVE: j TO EVALUATE MEANS TO PREVENT, SUPPRESS OR CONTROL ! HYDROGEN BURN SUBTASKS 6.1 SURVEY OF HYDROGEN AND OXYGEN DETECTORS l i 6.2 EVALUATION OF PRE-EVENT INERTING 6.3 EVALUATION OF FOGGING / SPRAY SUPPRESSION 6.4 EVALUATION OF CONTROLLED BURN SYSTEMS . , l e
4 ,, TECilNOt.OGY for ENERGY CORPORATION , t SUBTASK 6.1 l ~ SURVEY OF HYDROGEN AND OXYGEN l DETECTORS SUBTASK SCOPE: CONDUCT AN INDUSTRY SURVEY TO ESTABLISH THE CAPABILITIES OF CURRENTLY EXISTING EQUIPMENT FOR THE DETECTION OF HYDROGEN AND OXYGEN.
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ESTIMATED TIME REQUIRED: 4 MONTHS TO FIRST DRAFT,
. 6 MONTHS TOTAL ESTIMATED MANPOWER: 3 MAN-MONTHS LATEST START DATE: OCTOBER 1981 .
LATEST COMPLETION DATE: april 1982 1
TECHNOLOGY for ENERGY CORPORATION 1 l l- SUBTASK 6.2 l EVALUATION OF PRE-INERTING l SU.BTASK SCOPE: i CONDUCT A SURVEY OF EXISTING OR ONGOING STUDIES OF , l CONTAINMENT INERTING CONCEPTS AND EVALUATE THE BENEFITS i AND HAZARDS ASSOCIATED WITH EACH APPROACH. ESTIMATED TIME REQUIRED: 3 MONTHS TO FIRST DRAFT, 6 MONTHS TOTAL ESTIMATED MANPOWER: 6 MAN-MONTHS LATEST START DATE: SEPTEMBER 1981 l LATEST COMPLETION DATE: FEBRtlARY 1982
TECHNOLOGY for ENERGY CORPORATION SUBTASK 6.3 EVALUATION OF FOGGING / SPRAY SUPPRESSION SUBTASK SCOPE: REVIEW THE AVAILABLE INFORMATION ON HYDROGEN BURN SUPPRESSION SYSTEMS AND EVALUATE THE HAZARDS AND BENEFITS OF THIS TYPE OF HYDROGEN BURN CONTROL ESTIMATED TIME REQUIRED: 6 MONTHS TO FIRST DRAFT, .
- 9 MONTHS TOTAL ,
ESTIMATED MANPOWER: 9 MAN-MONTHS , LATEST START DATE: OCTOBER 1981 ! LATEST COMPLETION DATE: JUNE 1982 i t 1 -- - . ._ _ _ _ - _ _ _ _ _
TECHNOLOGY for ENERGY CORPORATION , SUBTASK 6.4 REVIEW CONTROLLED BURN SYSTEMS SUBTASK SCOPE: IDENTIFY AND ASSESS POTENTIAL CONTROLLED BURN SYSTEMS INCLUDING IGNITERS, POST-EVENT IN.ERTING CONCEPTS (HALON INJECTION), CATALYTIC COMBUSTERS. EVALUATE THE HAZARDS . AND POTENTIAL BENEFITS OF CANDIDATE SYSTEMS. ESTIMATED TIME REQUIRED: 6 MONTHS TO FIRST DRAFT, 9 MONTHS TOTAL . ESTIMATED MANPOWER: 18 MAN-MONTHS LATEST START DATE: OCTOBER 1981
- LATEST COMPLETION DATE
- JUNE 1982 i
i
l' 1ECHNOLOGY for ENERGY CORPORATION TASK 7 ' EQUIPMENT SURVIVABILITY FOR DEGRADED CORE ENVIRONMENTS i i OBJECTIVE: TO IDENTIFY MINIMUM SET OF EQUIPMENT AND TO EVALUATE THE CAPABILITY TO SURVIVE A DEGRADED CORE ENVIRONMENT SUBTASKS: 7.1 IDENTIFY' FUNCTIONS TO BE PERFORMED 7.2 IDENTIFY MINIMUM SET OF EQUIPMENT 7.3 IDENTIFY DEGRADED CORE ENVIRONMENTS i 7.4 EVALUATE EQUIPMENT CAPABill1Y l
=
I l { TECHNOLOGY for ENERGY CORPORATION l SUBTASK 7.1 I IDENTIFY FUNCTIONS WHICH MUST BE PERFORMED ! SUBTASK SCOPE: REVIEW SUCCESS CRITERIA DEVELOPED IN OTHER SUBTASKS AND l DEFINE THE FUNCTIONS WHICH ARE REQUIRED TO PREVENT l OR MITIGATE THE EFFECTS OF DEGRADED CORE ACCIDENTS AND l TO MONITOR REACTOR AND CONTAINMENT STATUS. ' ESTIMATED TIME REQUIRED: 4 MONTHS ESTIMATED MANPOWER: 16 MAN-MONTHS LATEST START DATE: FEBRUARY 1982 LATEST COMPLETION DATE: JUNE 1982 l
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l-TECHNOLOGY for ENERGY CORPORATION l r I i j l SUBTASK 7.2 l IDENTIFY MINIMUM SET OF EQUIPMENT SUBTASK SCOPE: REVIEW THE SAFETY FUNCTIONS DEFINED FOR SELECTED PLANT CONFIGURATIONS AND DEFINE THE EQUIPMENT REQUIRED TO DETERMINE THE STATE OF THE PLANT AND TO BRING IT TO A SAFE STABLE STATE. ESTIMATED TIME REQUIRED: 4 MONTHS
. ESTIMATED MANPOWER: 16 MAN-MONTHS LATEST START DATE: MARCH 1982 LATEST COMPLETION DATE: AUGUST 1982
l '
- TECHNOLOGY for ENERGY CORPORATION 1
i SUBTASK 7.3 IDENTIFY DEGRADED CORE ENVIRONMENT i SUBTASK SCOPE: - REVIEW THE ACCIDENT SEQUENCE ENVIRONMENTS FOR THE MOST LIKELY ACCIDENT SEQUENCES. IDENTIFY THE PARAMETERS WHICH ARE SPECIFIC TO THE EQUIPMENT LOCATIONS OF INTEREST. l ESTIMATED TIME REQUIRED: 4 MONTHS ESTIMATED MANPOWER: 16 MAN-MONTHS j LATEST START DATE: JUNE 1982 f ~ LATEST COMPLETION DATE: OCTOBER 1982 l
=
] TECHNOLOGY for ENERGY CORPORATION if , L SUBTASK ,7.4 ' EVALUATE SURVIVABILITY OF EQUIPMENT ' AND DEFINE TESTS SUBTASK SCOPE. FOR THE ENVIRONMENT ASSOCIATED WITH MOST LIKELY ! SEQUENCES, EVALUATE SURVIVABILIW OF EQUIPMENT NECESSARY l TO PERFORM MINIMUM FUNCTIONS. ESTIMATED TIME REQUIRED: 6 MONTHS f ESTIMATED MANPOWER: 12 MAN-MONTHS LATEST START DATE: SEPTEMBER 1982 f - LATEST COMPLETION DATE: MARCH 1983 6
4 , TECHNOLOGY for ENERGY CORPORATION l ! l TASK 8 , l CORE DEBRIS BEHAVIOR AND COOLABILITY i l OBJECTIVE-l TO PREDICT THE BEHAVIOR OF CORE MATERIALS AND THEIR l lNTERACTIONS WITH STRUCTURAL MATERIALS DURING DOMINANT l ACCIDENT SEQUENCES AND TO DETERMINE COOLAB!LITY LIMITS 3 SUBTASKS: 8.1 ANALYZE IN-VESSEL CORE MELT PROGRESSION 8.2 ESTABLISH. IN-VESSEL COOLABILITY, VESSEL l i PENETRATION, EX-VESSEL COOLABILITY 8.3 EVALUATE CORE DEBRIS-CONCRETE REACTIONS , i
'l
j ; I
- TECHNOLOGY for ENERGY CORPORATION l
l i )
- SUBTASK 8.1
! ANALYZE IN-VESSEL CORE MELT PROGRESSION i SUBTASK SCOPE: DETERMINE EVENT PROGRESSION FROM BOIL-OFF THROUGH MAJOR GEOMETRY DISTRUPTION TO PENETRATION OF DEBRIS . THROUGH THE CORE SUPPORT STRUCTURE. ANALYZE POTENTIAL SAFE STABLE STATES. '] ANTICIPATED MANPOWER: 50 MAN-MONTHS l LATEST START DATE: JULY 1981 LATEST COMPLETION DATE: DECEMBER 1982 l I I . _ ._ _ ._ _
e l TECHNOLOGY for ENERGY CORPORATION l SUBTASK 8.2 ! ESTABLISH IN-VESSEL COOLABILITY, VESSEL ; i PENETRATION, EX-VESSEL COOLABILITY l l SUBTASK SCOPE: l ESTABLISH LIMITS FOR IN-VESSEL COOLABILITY, MODES AND l TIMING OF VESSEL PENETRATION, AND LIMITS OF EX-VESSEL ' l COOLABillTY. ASSESS EFFECTS OF DEBRIS DISPERSION AND OF ' l LOCAllZED MELT-THROUGH EFFECTS. ANALYZE POTENTIAL SAFE l STABLE STATES. . l ANTICIPATED MANPOWER: 60 MAN-MONTHS l LATEST START DATE: JULY 1981 LATEST COMPLETION DATE: DECEMBER 1982 .
TECHNOLOGY for ENERGY CORPORATION ] l I SUBTASK 8.3
- . EVALUATE CORE
- DEBRIS-CONCRETE REACTIONS l SUBTASK SCOPE
- .
! CRITICALLY EVALUATE EXISTING MODELS. PREDICT RATE OF l CONCRETE PENETRATION AND OF GAS GENERATION; INCLUDE
- EFFECTS OF JET FORCES AND DEBRIS DISPERSAL DURING MELT-THROUGH. ANALYZE POTENTIAL SAFE STABLE STATES.
l ANTICIPATED NIANPOWER: 18 MAN-MONTHS LATEST START DATE: JULY 1981 LATEST COMPLETION DATE: DECEMBER 1982
TECHNOLOGY for ENERGY CORPORATION i l TASK 9 CONTAINMENT STRUCTURAL CAPABILITY
- OBJECTIVE
TO ESTABLISH A UNIFORM REAllSTIC' BASIS FOR DETERMINING ULTIMATE CONTAINMENT CA,PABILITY SUBTASK: 9.1 WORKSHOP ON CONTAINMENT STRUCTURAL CAPABILITY 9.2 EVALUATION OF CONTAINMENT STRUCTURAL' CAPABILITY PA4823 3S
. 4 TECHNOLOGY for ENERGY CORPORAllON SUBTASK 9.1 . ! WORKSHOP ON CONTAINMENT STRUCTURAL ! CAPABILITY - ) SUBTASK SCOPE: l CONDUCT A WORKSHOP WHICH WILL CONSIDER THE THREATS TO CONTAINMENT WHICH RESULT FROM DEGRADED CORE ACCIDENTS, ASSESS CRITICAL PROBLEMS FOR SPECIFIC CONTAINMENT 1YPES, DEFINE POTENTIAL FAILURE MODES, i i SUGGEST APPROPRIATE FAILURE CRITERIA AND REVIEW RELATION ! BETWEEN DESIGN AND ULTIMATE STRENGTH l ESTIMATED TIME REQUIRED: 6 MONTHS ESTIMATED MANPOWER: 2 MAN-MONTHS LATEST START DATE: SEPTEMBER 1981 LATEST COMPLETION DATE: MARCH 1982 B - 4
TECHNOLOGY for ENERGY CORPORATION { SUBTASK 9.2 EVALUATION OF CONTAINMENT STRUCTURAL CAPABILITY SUBTASK SCOPE: DEVELOP AND EVALUATE METHODS TO DETERMINE THE ULTIMATE CAPABill1Y OF THE CONTAINMENT TYPE AND DEFINE POTENTIAL FAILURE MODES UNDER SEVERE ACCIDENT CONDITIONS. , ESTIMATED TIME REQUIRED: 10 MONTHS ESTIMATED MANPOWER: 24 MAN-MONTHS LATEST START DATE: MARCH 1982 LATEST COMPLETION DATE: JANUARY 1983 e
TECilNOLOGY for ENERGY CORPORATION t l l TASK 10 l EVALUATE ATMOSPHERIC l AND LIQUID PATHWAY DOSE } OBJECTIVE: ; TO IDENTIFY MOST APPROPRIATE MODELS FOR EVALUATING RADIOLOGICAL CONSEQUENCES OF DOMINANT SEQUENCES AND PERFORM ANALYSES SUBTASK 10.1 EVALUATE ATMOSPHERIC AND LIQUID PATHWAY DOSE AND PERFORM ANALYSES O N
1 i. ! TECHNOLOGY for ENERGY CORPORATION l l . l SUBTASK 10.1
- EVALUATE ATMOSPHERIC AND LIQUID PATHWAY DOSE AND PERFORM ANALYSES l
- SUBTASK SCOPE
EVALUATE APPLICABILITY OF EXISTING ATMOSPHERIC AND
- LIQUID PATHWAY DOSE CALCULATIONAL TECHNIQUES. PROVIDE l
LISTING OF RECOMMENDED MODELS. PERFORM ANALYSES. l ANTICIPATED MANPOWER: 6 MAN-MONTHS LATEST START DATE: APRIL 1982 LATEST COMPLETION DATE: JANUARY 1983 l l
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(. ' I TECHNOLOGY for ENERGY CORPORATION
- . TASKTI j FISSION PRODUCT LIBERATION, TRANSPORT, AND INHERENT RETENTION i
- OBJECTIVE
- .
i TO PREDICT FISSION PRODUCT RELEASE FROM FUEL, TRANSPORT - l WITHIN CONTAINMENT, CHEMICAL / PHYSICAL FORM, AND ! ULTIMATE DISPOSITION DURING DOMINANT ACCIDENT SE-l QUENCES SUBTASKS l 11.1 EVALUATE RELEASE FROM FUEL ; l 11.2 IDENTIFY PATHWAYS
- 11.3 ASSESS TRANSPORT BEHAVIOR (INCLUDING EFFECTS OF ENGINEERED SAFETY FEATURES) 11.4 ASSESS CHEMICAL FORMS ,
11.5 ASSESS INHERENT RETENTION
j . l . TECHNOLOGY for ENERGY CORPORATION l SUBTASK 11.1 j EVALUATE RELEASE FROM FUEL SUBTASK SCOPE: l ASSESS DATA BASE AND RECOMMEND TECHNIQUES FOR PRE-DICTING RATE, QUANTITY, AND CHEMICAL / PHYSICAL FORM OF FISSION PRODUCTS RELEASED FROM FUEL DURING DOMINANT ACCIDENT SEQUENCES ANTICIPATED MANPOWER: 3 MAN-MONTHS LATEST START DATE: NOVEMBER 1981 LATEST COMPLETION DATE: DECEMBER 1982 <
. 1 ! TECHNOLOGY for ENERGY CORPORATION , 4 i ! SUBTASK 11.2
- IDENTIFY PATHWAYS
! SUBTASK SCOPE: l lDENTIFY PATHWAYS (IN ADDITION TO THOSE IDENTIFIED BY t l CONTAINMENT FAILURE ANALYSIS) BY WHICH FISSION PRODUCTS g COULD ESCAPE CONTAINMENT . . l ANTICIPATED MANPOWER: 3 MAN-MONTHS l LATEST START DATE: NOVEMBER 1981 i LATEST COMPLETION DATE: DECEMBER 1982 r l
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TECHNOLOGY for ENERGY CORPORATION SUBTASK 11.3 l ASSESS TRANSPORT DEHAVIOR ' l (INCLUDING EFFECTS OF ENGINEERED l SAFETY FEATURES) SUBTASK SCOPE: l ASSESS DATA BASE AND CALCULATIONAL MODELS FOR PREDICT- , l ING FISSION PRODUCT TRANSPORT BEHAVIOR WITHIN PRIMARY
- SYSTEM AND CONTAINMENT, INCLUDING EFFECTS OF ENGINEERED SAFETY FEATURES. RECOMMEND BEST TECHNIQUES;
] l IDENTIFY DEFICIENCIES i ANTICIPATED MANPOWER: 12 MAN-MONTHS LATEST START DATE: NOVEMBER 1981 . l LATEST COMPLETION DATE: DECEMBER 1982 i 0
I' . . , i '- 4 l TECilNOLOGY for ENERGY CORPORATION
! i
) SUBTASK 11.4 . ! ASSESS CHEMICAL FORMS i SUBTASK SCOPE: l ASSESS DATA BASE AND TECHNIQUES.FOR PREDICTING CHEMICAL FORMS OF FISSION PRODUCTS DURING DOMINANT ACCIDENT l SEQUENCES. RECOMMEND BEST TECHNIQUES; IDENTIFY l DEFICIENCES l ANTICIPATED MANPOWER: 6 MAN-MONTHS l LATEST START DATE: NOVEMBER 1981 l LATEST COMPLETION DATE: DECEMBER 1982 l -
l isennotosv nor eninov coneon41 ion i SUBTASK 11.5 l ASSESS INHERENT RETENTION I SUBTASK SCOPE: ASSESS DATA BASE AND TECHNIQUES FOR PREDICTING EXTENT OF l INHERENT FISSION PRODUCT RETENTION FOR DOMINANT SEQUENCES. RECOMMENDED BEST TECHNIQUE; IDENTIFY l l DEFICIENCIES . ANTICIPATED MANPOWER: 6 MAN-MONTHS LATEST START DATE: NOVEMBER 1981 i i LATEST COMPLETION DATE: DECEMBER 1982 l ) l (*"" * * -_
- TECHNOLOGY for ENERGY CORPORAllON 1
! TASK 12 SURVEY OF ALTERNATE CONTAINMENT i SYSTEMS l ! -OBJECTIVE: TO REVIEW ONGOING STUDIES OF ALTERNATE CONTAINMENT SYSTEMS AND TO EVALUATE THE POTENTIAL FOR RISK REDUCTION l OR UNDESIREABLE SIDE EFFECTS .
- 2. RV OF ALTERNATE CONTAINMENT SYSTEMS i
I t
i l TECHNOt.OGY for ENERGY CORPORATION SUBTASK 12.1 SURVEY OF ALTERNATE CONTAINMENT . i SYSTEMS i .. SUBTASK SCOPE: l REVIEW ONGOING STUDIES BY INDUSTRY AND NRC OF ALTERNATE ! CONTAINMENT SYSTEMS AND EVALUATE HAZARDS, BENEFITS AND l COSTS OF CANDIDATE SYSTEMS IF NECESSARY. ESTIMATED TIME REQUIRED: 12 MONTHS ESTIMATED MAN _ POWER: 20 MAN-MONTHS l LATEST START DATE: OCTOBER 1981 f LATEST COMPLETION DATE: SEPTEMBER 1982 1 i l I f
f f
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TECilNOLOGY for ENERGY CORPORATION TASK 13 SURVEY OF CORE RETENTION DEVICES OBJECTIVE: TO REVIEW ONGOING STUDIES OF CORE RETENTION DEVICES AND TO EVALUATE THE POTENTIAL FOR RISK REDUCTION OR UNDESIREABLE SIDE EFFECTS l SUBTASKS: . 13.1 SURVEY OF CORE RETENTION DEVICES i tr
I ,, 4 . TECHNOl.OGY for ENERGY CORPORATION I,
! I 1
l - SUBTASK 13.1 , i SURVEY OF CORE RETENTION DEVICES l j SUBTASK SCOPE ' REVIEW THE EXISTING OR ONGOING INDUSTRY OR NRC STUDIES - OF CORE RETENTION DEVICES AND EVALUATE THE POTENTIAL HAZARDS, BENEFITS AND COSTS ASSOCIATED WITH REAllSTIC SYSTEMS IF NECESSARY. ESTIMATED TIME REQUIRED: 6 MONTHS ESTIMATED M'ANPOWER: 6 MAN-MONTHS LATEST START DATE: JULY 1981 - LATEST COMPLETION DATE: FEBRUARY 1982 , I S GO PQ4823 40 _ _ _ _ _ _ _ _ _ _ _ _
.i .,
3 TECHNOLOGY for ENERGY CORPORAllON i
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, t l i l TASK 14 i RISK REDUCTION POTENTIAL i i i OBJECTIVE: ! TO EXAMINE REDUCTIONS IN PLANT RISK DUE TO CURRENT I ACTIVITIES AND POSSIBLE FUTURE ACTIONS SUBTASKS: 14.1 POST-TMI CHANGES 14.2 FEATURES AND PHENOMENA l I i i l . . i - . ._ _ _ _ _
- 1. . ,,
i 1 TECHNOLOGY for ENERGY CORPORATION i i ! u l ! SUBTASK 14.1 EXAMINE POST-TMI CHANGES i l SUBTASK SCOPE:
- EXAMINE DETAILED RISK STUDIES FOR IMPACT ON RISK OF i
! POST-TMI CHANGES IN DESIGN AND OPERATIONS . 3 j ANTICIPATED MANPOWER: 2 MAN-MONTHS l LATEST START.DATE: SEPTEMBER 1981 LATEST COMPLETION DATE: MARCH 1982 o ,- --
TECilNOLOGY for ENERGY CORPORATION I i SUBTASK 14.2 l
. ANALYZE MITIGATION FEATURES AND PHENOMENOLOGICAL ISSUES i SUBTASK SCOPE: ,
i USE IDCOR PHENOMENA MODEllNG TO ASSESS RISK REDUCTION l POTENTIAL OF PREVENTIVE OR MITIGATIVE SCHEMES j ANTICIPATED MANPOWER: 16 MAN-MONTHS l LATEST START DATE: SEPTEMBER 1981 l LATEST COMPLETION DATE: FEBRUARY 1983 i ,
i 1ECilNOI.OGY for ENERGY CORPORAllON , TASK 15 l , , ' INTEGRATED MODEL DEFINITION I AND ANALYSIS l OBJECTIVE: ! TO INTEGRATE RESULTS OF OTHER SUBTASKS INTO A COMPRE-l HENSIVE ACCIDENT ANALYSIS CAPABILITY AND PREDICT Pl. ANT BEHAVIOR FOR DOMINANT DEGRADED CORE ACCIDENT SEQUENCES l SUBTASKS: I 15.1 DEVELOP ACCIDENT SEQUENCE TIME-LINE CHARTS ! AND IDENTIFY POTENTIAL SAFE STABLE STATES ' 15.2 ASSESS AVAILABLE CODES, DEFINE USES, AND FOLLOW AND SUPPORT ONGOING ACTIVITIES 15.3 DEVELOP IDCOR EXECUTIVE ANALYSIS PROGRAM , 15.4 1NTEGRATE PHENOMENOLOGY MODELS WITH ANALYSIS 15.5 PERFORM CONTAINMENT ANALYSIS
,. , .9 - . - - - - -
TECHNOLOGY for ENERGY CORPORATION l i SUBTASK 15.1 DEVELOP ACCIDENT SEQUENCE TIME-LINE . ! CHARTS AND' IDENTIFY POTENTIAL l SAFE STABLE STATES ! SUBTASK SCOPE: l PREPARE ACCIDENT SEQUENCE TIME-LINE CHARTS FOR DOMINANT SEQUENCES TO ASSIST IN PROVIDING PROGRAM INTEGRATION l AND IN IDENTIFYING PHENOMENA, TIMING, AND POTENTIAL CORRECTIVE AC. TION. IDENTIFY POTENTIAL SAFE STABLE STATES. ANTICIPATED MANPOWER: 9 MAN-MONTHS LATEST START DATE: JULY 1981 LATEST COMPLETION DATE: DECEMBER 1982
l' i. r 1ECHNOLOGY for ENERGY CORPORAllON SUDTASK 15.2 ASSESS AVAILABLE CODES, DEFINE USES, AND FOLLOW AND SUPPORT ONGOING ACTIVITIES SUBTASK SCOPE: IDENTIFY AND ASSESS EXISTING CODES AND ONGOING DEVELOPMENTS THAT CAN BE USED IN THE OVERALL IDCOR INTEGRATED ANALYSIS PACKAGE. SUPPORT DEVELOPMENT OF , THE IDCOR EXECUTIVE ACCIDENT ANALYSIS PROGRAM ANTICIPATED MANPOWER: 6 MAN-MONTHS LATEST START DATE: JULY 1981 LATEST COMPLETION DATE: FEBRUARY 1982
! .. +.
TECHNOLOGY for ENERGY CORPORATION { SUBTASK 15.3 i DEVELOP IDCOR EXECUTIVE ANALYSIS PROGRAM l SUBTASK SCOPE: - l DEVELOP MODULAR IDCOR EXECUTIVE ANALYSIS PROGRAM FOR l ANALYZING FOR DOMINANT ACCIDENT SEQUENCES THE FOLLOW-i ING: 1) INTACT GEOMETRY CORE HEATUP, 2) IN-VESSEL MATERIALS MIGRATION, 3) COMBINED PHENOMENA IN CONTAINMENT. ~ MAKE APPROPRIATE USE OF AVAILABLE MODELS AND TECHNIQUES AND OF PARALLEL ACTIVITIES. l ANT!CIPATED MANPOWER: 53 MAN-MONTHS _ LATEST START DATE: JULY 1981 l LATEST. COMPLETION DATE: SEPTEMBER 1982 . f i l l -
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) TECllNOLOGY for ENERGY CORPORAllON 1 ! SUBTASK 15.4 INTEGRATE PHENOMENOLOGY MODELS ! WITH ANALYSIS l SUBTASK SCOPE: ! ASSESS RELATED PHENOMENOLOGICAL DEVELOPMENT i ACTIVITIES AND PROVIDE ASSURANCE THAT THE BEST MODELS ARE INCORPORATED INTO THE IDCOR EXECUTIVE ANALYSIS PROGRAM ANTICIPATED MANPOWER: 12 MAN-MONTHS ! LATEST START DATE: AUGUST 1981 l LATEST COMPLETION DATA: SEPTEMBER 1982 l l l i I v l l
l
' - " T5CHNOLOGY for ENERGY CORPORAllON
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TASK 16 !- OPERATIONAL ASPECTS OF ACCIDENT
- l. -
MnNAGEMENT AND CONTROL 1 OBJECTIVE: TO CHARACTERIZE THE POTENTIAL FOR OPERATOR RESPONSE , i IN MANAGING DEGRADED CORE CONDITIONS l SUBTASKS: - ! 16.1 CORRECTIVE ACTIONS / TIME WINDOWS I 16.2 EFFECTS OF CORRECTIVE ACTIONS 16.3 INSTRUMENTATION REQUIREMENTS ! 16.4 ROLE OF THE OPERATOR l 16.5 MAN / MACHINE BENEFITS CRITERIA
TECllNOLOGY for ENERGY CORPORAllON il SUBTASK 16.1 ! IDENTIFY CORRECTIVE ACTIONS AND TIME ! WINDOWS AVAILABLE l ! SUBTASK SCOPE: ! DETERMINE WHERE ACTIONS CAN PREVENT OR ARREST l ACCIDENT CONDITIONS AND TIME AVAILABLE ! ANTICIPATED MANPOWER: 4 MAN-MONTHS'
. LATEST START DATE: AUGUST 1981 LATEST COMPLETION DATE: JANUARY 1983 1 -
1 i 4 j
l TECHNOLOGY for ENERGY CORPORATION 1 1 I SUBTASK 16.2
- IDENTIFY EFFECTS OF CORRECTIVE ACTIONS i
l l SUBTASK SCOPE: l ASSESS EFFECTIVENESS OF ACTIONS THAT ARE USED TO PREVENT . i OR MITIGATE DEGRADED CORE CONDITIONS ! ANTICIPATED MANPOWER: 12 MAN-MONTHS LATEST START DATE: OCTOBER 1981 LATEST COMPLETION DATE: OCTOBER 1982 6
8 t TECHNOLOGY for ENERGY CORPORAliON SUBTASK 16.3 IDENTIFY INSTRUMENTATION REQUIREMENTS SUBTASK SCOPE:
ASSESS EFFICIENCY OF EXISTING INSTRUMENTATION UNDER DEGRADED CORE CONDITIONS AND EVALUATE POTENTIAL IMPROVEMENTS ANTICIPATED MANPOWER: 12 MAN-MONTHS LATEST START DATE: OCTOBER 1981 LATEST COMPLETION DATE: OCTOBER 1982
a O i j TECHNOLOGY for ENERGY CORPORATION i i i i - SUBTASK 16.4 - l CHARACTERIZE ROLE OF OPERATOR i SUBTASK SCOPE: i INTEGRATE TASK 16 AND CHARACTERIZE THE OPERATOR'S ROLE l UNDER POTENTIAL DEGRADED CORE CONDITIONS ANTICIPATED MANPOWER: 25 MAN-MONTHS l LATEST START DATE: OCTOBER 1981 LATEST COMPLETION DATE: JANUARY 1983 i i 1 .
j , j. TECllNOLOGY for ENERGY CORPORATION I ! SUBTASK 16.5 l . CHARACTERIZE BENEFITS OF t MAN / MACHINE CRITERIA . SUBTASK SCOPE: DEFINE MAN / MACHINE INTERFACE AND ASSESS HUMAN FACTORS CRITERIA WITH RESPECT TO DEGRADED CORE CONDITIONS ANTICIPATED MANPOWER: 9 MAN-MONTHS LATEST START DATE: APRIL 1982 LATEST COMPLETION DATE: OCTOBER 1982
' si , .
.a Sandia National Laboratories uf date: June 8, 1981 Albuquerque. New Mexico 87185 to:. - Distribution -
g h. ^=YA
- hani A. S . Benjamin, 4414 l
1 .s wbka: IDCOR Potential Contractor Information Meeting t \. On May 21, 1981, about 150 representatives from across the nuclear industry gathered at the Hyatt Regency Hotel in Knoxville, Tennessee, to attend the IDCOR Potential Contractor Information~ Meeting. The purpose of the meeting was to ! acquaint the participants with the IDCOR (Industry Degraded l- Core) project and to solicit responses from organizations interested in providing contract support. In addition to, industry representatives, each of the national labs sent at least one observer. I have attached copies of the agenda, the vugraphs used by
- (
, Technology for Energy Corporation (TEC), and the task descrip-tions as they now stand.
The structure and scope of the IDCOR program has not changed l significantly since the last meeting, which was previously ! reported,* except for one notable addition. A task has now been'added to develop an "IDCOR Executive Analysis Program" for analyzing accident sequences from initiation to completion. The ' program will incorporate models that treat: (1) intact geometry core heatup, (2) in-vessel material migration, and (3) combined phenomena in containment, making use of available models as far as possible. The IDCOR Executive Analysis Program appears to be similar in scope and intent to the NRC Integrated ,e Severe Accident Code, which is a line item in NRC's' degraded d' core cooling rulemaking-related research program. A summary of IDCOR tasks and projected manpower requirements is given in Table I. Basically, the IDCOR approach is to procure data from existing programs sponsored by NRC, EPRI, and others, to interpret the data from their own perspective, and to prepare a technically sound position on which they can I stand during rulemaking. Toward this pursuit, there will be four areas of emphasis:
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*D. A. Powers, et al., Report on the Meeting with the IDCOR Project Group, April 13-16, 1981, Germantown, MD.
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June 8, 1981 Distribution 7 . L
- 1) Identification and removal of conservatisms
- from existing accident analyses,
- 2) assessment of potential human interfaces during -
the progression of accidents,
- 3) characterization of safe stable states and determination of the time windows for achieving them, and
- 4) identification of minimum engineered safety features and evaluation of their survivability during severe accidents.
The emphasis on these areas was apparent during the discussions and can also be observed from the manpower allocations in Table I. On the contrary, the tasks dealing with development and analy-sis of new systems for accident prevention or mitigation are clearly de-emphasized. The philosophy is to wait and see what NRC plans to do in this area, and to address new systems only when it appears likely that NRC will attempt to require them. The industry's position on new safety systems seems to have already been determined, and it is predominantly defensive. After attending the last two IDCOR meetings, I have two recom-mendations regarding Sandia's activities in support of NRC on degraded core rulemaking. Both recommendations concern the need for coordination. First, it is apparent that the IDCOR program runs parallel to Sandia's programs in severe accident phenomenology, systems analysis, and risk assessment, and that the industry will be critiquing our results and interpreting them in their own way, whether we like it or not. To avoid a potentially embarrassing situation in which we are surprised or even ambushed during rulemaking hearings, we must stay abreast of their thinking and be aware of their conclusions. I have therefore suggested to Mario Fontana, the IDCOR Program Director for Technology for Enercy Corporation, the following three points:
- 1) That Sandia be represented on their peer review group,
- 2) that IDCOR participate in the industry review process that we have initiated in our systems ,
related programs in degraded core cooling, and _
- 3) that Sandia be invited to attend their general contractor review meeting, held semi-annually.
U
. - 3, Distribution June 8, 1981 Mario's response was that my name had been included in their first 1ist of people to comprise their peer review group.
T' h e second recommendation is that we institute a steering group within Sandia to oversee the various activities at Sandia associated with degraded core issues. In the face of what appears to be a well-organized and unified program by the' industry to formulate positions for degraded core rulemaking, it is important that we also develop an overview of our programs and formulate an integrated position on the issues that will
' ~
be emerging. ASB 4414:ls Distribution: 4400 A. W. Snyder 4410 D. J. McCloskey 4412 J. W. Hickman 4414 G. B. Varnado [ 4414 J. L. Darby 5 4420 J. V. Walker 4421 J. B. Rivard 4422 D. A. Powers 4440 G. R. Otey 4441 M. Berman 4442 W. A. Von Riesemann 4443 D. A. Dahlgren .
4e ( Table I. IDCOR Tasks and Anticipated Manpower Allocation Tasks Man-months Risk Analysis: 145 ,_ Safety Goal / Criteria Applications 6 - Dominant Sequences 33 Containment Phenomenological Sequences 26 Risk Reduction Potential 18 Operational Aspects of Accident Management 62 and Control Rcactor and Plant Systems: 148 Hydrogen Burn Control 36 Equipment Survivability 60 Containment Structural Capability 26 r L Alternate Containment Concepts 20 . Core Retention Devices 6 Phenomena Analysis: 374 Steam Overpressure Phenomena 24 Hydrogen Generation and Burn 43 Core Debris Behavior and Coolability 119 Atmospheric and Liquid Pathway Dose 6 Fission Product Liberation, Transport, and 30 Inherent Retention Integrated Model Definition and Analysis 152 Total 667 l
(Enclosure 3) -a'
)-.- ' I.D .)
t:5-RAT-DKG-015 J men,reenmecronoma Westinghouse Water Reactor Bectric Corporation Divisions i,a emowenewsmwanutsno December 3,1980
.TO: MEMBERS OF AIF DEGRADED CORE SU8 COMMITTEE SU3 JECT: INITIALPROGRAMPLAE The initial program plan for the industry effort in preparation for degraded core rulemaking has been prepared and reviewed by the Task Team. This program plan is attached for your review and coment. Overall, the program is esti-mated to take approximately 2 years and 7 MS. It must be emphasized that' both schedule and cost estimates are representative but that significant changes could develop in the estimates for any task. For planning purposes, a 10 M5 budgetary base is raccrende.d. spread over 2 years.
In all of the major tasks, the program is structured to provide more detailed planning first prior to proceeding. This should provide the program manager with the complete base for decisions. However, no effort is currently included for iteration with the NRC. Please provide any comments that you may have on this to D. K. Goeser, (412)373-4001. Sincerel ,
^
s V D., . Goes , Manager robabili tic Risk Assessment uclear afety Department ib O l l O.
- p. , .
INITIAL PROGRAM PUW r This initial program plan identifies the integrated efforts believed necessary ___ to generate sufficient technical information to develop rational positions on j key degraded core issues and to provide the base for successful degraded core rulemaking hearings. - The program has been scoped to incorporate the results of known programs-- industry, NRC, foreign--and includes only those efforts necessary prior to participation in the rulemaking. Based on current knowledge, it is unlikely that any simple solutions are available; there are gaps in the technological information which must be filled to develop and defend rational positions, and there is a need for effective understanding of the complete picture regarding degraded core issues and strong annagement of an integrated program. The program is focused on obtaining sufficient information, not defaloping information for the sake of information. The program is defined in related segments. These segments have been derived from an overview of the degraded core situation as follows. 1 The industry will develop rational positions related to the characteriza-
~
tion of residual risk from degraded core conditions and the potential alternatives for further preventi.on or mitigation of residual risk. A safety goal or criteria that provides a measure of acceptability is 'a necessary condition for effective evaluation of residual risk.
- The release of radioactivity to the environment is cause of deleterious I censequences. Releases can occur only if both large quantities of r dio-activity are produced and released from the primary system and the con-tainmenc is ineffective.
The relert' of radioactivity :o the environment is the cause of deleta , celettricus c:nsecuences. R'eleases can occur only if both large quan-tities of raoicactivity are procuced ano releaseo from*tne ;rimary system and the c:ntainment is ir. affective. The attanuation prior to c'entainment failure must also te c:nsidered.
The separa.te tasks are: -
- 1. Safety Goal / Criteria Application
- 2. Selection of Dominant Sequences
- 3. Identification of Phenomenological and Containment Transient ~
Critical Sequences
- 4. Steam Overpressure Phenomena (In-Vessel, Containment)
- 5. Hydrogen Generation and Burn
- 6. Hydrogen Burn Control 7 Equipment Survivability for the Degraded Core Environment
- 8. Core Debris Coolability
- 9. Containment Structural Capability .
- 10. Evaluation of Liquid Pathway Dose
- 11. Fission Product Liberation and Removal
- 12. Vented Containment Systems
- 13. Core Ladle
- 14. Residual Risk Reduction Eva,1uation
- 15. Integrated Mode 1' Definition and Analysis For each task, the following sheets identify the relationship of the task to overall objectives, the information known to be available, the scope of the recorr.anded actions, the schedule and budgetary estimates.
- 1. SAFETY GOAL /CRITERIAL APPLICATION Relationshio to Objectives: A safety goal / criteria is a necessary condition for proceeding with the rulemaking. There must be:
- a. Acceptability limits for individual and population health effects 1 including both probability and consequences,
- b. Risk / benefit criteria for alternative evaluation
- c. Definition of methods for evaluation of degraded core conditions
- (realistic analysis of transients, no arbitrarily postulated equipment I failures, use of realistic ultimate containment pressure capability, etc.)
tiithout these items, there is no common basis for decisions regarding acceptability of current designs, the relative benefits of alternative _ features, and the analytical results obtained for containment transients. c,w ~..,-.---,,,ne. n --- - - v-..,-.-w n . - - --- .wr,-,,a y,-e,- , ,< , rw.--m,.,, , -,, . - .,-, ,--,7, - - - --, , - ,- , , ,,,- - -. - - --
V .. 5-a relatively higher than the. remainder. By selecting the dominant sequences, _ the analytical and evaluation efforts are focused on areas where significant gains can be achieved. Available Information: For defining dominant accident sequences, a con-siderable technical base is available from similar " studies, either complete or due to be completed in the near term. These include: the Reactor Safety Study (WASH-1400), the NRC Reactor Safety Study Methodology Application Program (RSSMAP) which includes studies on four additional reactors, the individual plant studies including Zion-Indian Point Mint-WASH 1400 and complete study TVA Sequoyah studies, the NSAC/ Duke Oconee study, the Philadelphia Electric Limerick study, and others. In addition, studies on Auxiliary Feedwater Reliability, the IREP Program, and Inadequate Core Cooling studies each provide additional information useful in defining dominant accide,nt sequences. For the early phases of this program, the assimilation and integration of this available information is necessary. Although the information is useful, direct application without informed scrutiny would not provide the proper base for proceeding of particular importance is the inclusion of operating plant experience in defining the dominant sequences. In the later phases of this program, integrating the res'ults of major on-going studies is necessary, since no resource is provided in the program for performing studies to this level of detailed information. Scope of Recomended Actions: The scope of this task is to define the dominant sequences and rational for selection and provide documentation. This scope is divided into three parts: (1.) Initial definition of "likely" dominant sequences based on available material, an initial ranking of sequences in terms of probability and -
~
consequences and a definition of currently available preventive and mitigative (e.g., containment heat removal) systems would be prepareo.
~
i.. , 7
- (3.) Update to include detailed study results
- After detailed studies .
~
(Zion-Indian Point, NSAC/ Duke, Limerick, TVA, *etc.) are completed - these results would ~be integrated into the dominant accident sequence selection and rationale. The detailed studies should provide-the in-depth technical basis for the dominant accident sequence selection. 4
,, This task is estimated to require 6 N1 of effort assuming personnel familiar with the detailed studies are available.
Schedule /Budeet: Given above. 4
- 3. . SELECTION OF CONTAINMENT PHENOMEN0 LOGICAL SEOUENCES Relationship to Objectives: Identification and understanding of the criti-cal phenomena of degraded core conditions which lead to containment
~
challenges is necessary to evaluate the inherent plant margins and the effects of proposed mitigating features. The identification of the contain-ment sequences in conjunction with the dominant accident sequences focuses the analytical and experimental efforts on the critical areas related to
- system or containment failure challenges. A comprehensive identification of containment sequences assures that analyses or features are considered only if broadly significant.
Available Information: The Reactor Safety Study (WASH-1400), NRC studies , on mitigating features, and on-going industry studies (Zion-Indian Point, Limerick TVA/ Duke /AEP) each provide information on core degradation
- phenomena and the importance with respect to containment challenge.
! Further information in specific areas (hydrogen generation and burn, core degradation progression, e'tc.) is identified in tasks below. Scoce of Recornw.nded Effort: The scope of this activity is to generate , contain:.ent phenomenological event trees (or equivalent) for the major reactor systems. The containment event tree (or equivalent) would start where the dominant accident sequence effort terminates, i.e., conditions for core degradation have been postulated to occur. The event tree would identify the significant branch points the:eafter until recovery or
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- 4. STEAM OVERPRESSURE PHENOMENA (IN-VESSEL, CONTAINMENT) .
Relationship to Objective: Production of large energies or quantities of steam from the interaction of core debris with water is,one of the major postulated mechanisms for causing containment failure. Sinceawiderange of steam production rates can be calculated depending on the assumptions used in the analysis, it is necessary to develop technically based positions on the phenomena involved and the uncertainties in the theoretical and experimental bases for these phenomena. For realistic evaluations of residual risk, bounding calculations on the production of steaming are misleading at best. The proper approach is to evaluate the phenomena realistically based on current knowledge, specifically identify uncertain-ties, and provide an integrated assessment of the potential for containment ! challenge resulting from core-debris-water interaction. Since the amount of debris, rate of interaction, and initial conditions of debris-water-system complex, are all interrelated, this phenomenological evaluation includes the determination of appropriate models for core degra-dation melt progression, (included in task 8) necessary conditions for steam explosion initiation, dynamics of mixing and steam generation between core debris.and water, and conditions for missile generation and analyses of results of missiles.
~
Available Information: The Reactor Safety Study (WASH 1400), NRC studies i related to mitigating features, vapor explosion research in the metals and LMFBR areas, and Zion-Indian Point /Li. 2 rick industry studies all include infonnation relating to these phenomena. Extensive research programs in these areas have been or are being carried i out at Sandia, Argonne, Hedl, Brookhaven, Battelle-Columbus, Purdue ," University, and German facilities, as well as various other sites.
.Scoce of Recomended Actions: The overall scope of the activities listed below is to provide sound phenomenological models for the progression of - , ~
core melt for the identified dominant sequences, the conditions necessary for occurrence of steam explosions and effects of resultant explosions i _ -- ~ .. _ _ .- - _ . . _ _ _ _ _ _ ._,- _ ...,_ _ - _ ___ -_ ___-._.-- _ --
l.
~
11
~ '
established. Included would be the potential effects of the energy
~ ' . required for mixing and mechanisms for supp1'ying the energy, local thermal interactions, hydrodynamic dispersion effects, particulari-r zation of the debris, crust formation among others. Based on the information developed, mechanistic models for steam generation would be developed.
Depending on the results of this effort and the assumed pursuit of the current NSAC and planned EPRI program with ANL, experi-mental work broadening the technical base may be desirable. This could include: varying core debris injection mode, varying initial pressure conditions, and simulant material tests to establish specific mechanisms. This effort should presently be considered contingency. (3.) Structural Response of Equipment Based on the results of (1.'), bis task would establish the conditions for missiles or ex-vessel steam explosion phenomena. Establishing the necessary conditions for missile generation for physically reali-
. zable steam explosions includes identification of the means for rapid momentum transfer to the structure. Based on analyses to date, it is likely that such transfer is improbable at best. Ex-vessel steam explosion phenomena and resultant structural loadings will be assessed.
It is assumed that the current study results provide a reasonable assessment and that extensive structural analyses will not be required. Budaet/ Schedule: For each. task above, the following sequence would apply--
- a. Literature survey and precise definition of program and results (1-3 months after initiation)
- b. Interim Report covering Items A-E in detail (6-9 months after initiation)
- c. Final Report including models and experimental data (12-18 months after initiation) , ,
O __.____m_
3 . Scope of Reconnended Actions: The scope of this task is to provide sound ' ~~ phenomenological models for the generation, distribution, ignition and
- combustion of hydrogen as related to LWR conditions. The task is divided into parts. For each part, the intent is to develop the following information:
- a. Description of phenomena involved
- b. Relationship of phenomena to containment challenge
- c. Description of the physics underlying the phenomena
- d. Available theoretical, models and experimental evidence
- e. Best estimate progression path
- f. Range of uncertainties that should be considered
- g. Recommendedmodel(s)foranalysis
- h. Recommended experimental efforts (1.) Generation Rate and Amount of Hydrogen For the dominant ' accident sequences, detennine the rate and amount of zirconium-water or stainless-steel-water reaction producing hydrogen for both intact core geometry and core debris-water inter-action progressions. Currently available experimental data and models whould be used and improved to determine the best estimates of hydrogen production rates and amounts and to identify the appro-priate rang'es of uncertaincies. - .
(2.) Hydrogen Distribution This part includes two areas: a position paper compiling previous work on the large scale mixing characterisites of hydrogen and the conditions necessary to prevent significant stratification or pocketing, and system (or small scale) analysis of distribution and mixing of hydrogen. . For the second area, determine analytically the potential distributions assuming hydrogen rich releases (characteristic of the dominant accident sequences) into air atmospheres for conditions character-istic of releases into containment volumes, through. suppression j
- e The estianted resources to complete the tasks are: *
(1.) Generation Rate and, Amount of Hydrogen ~ 15 m (2.) Hydrogen Distribution , ,
~
30 m (3.) Combustion Limits of H 2-Air-Steam 8E For budgetary purposes, a contingency of 200K for experimental work in this area 'should be included: )
- 6. HYOROGEN BURN CONTROL Relationship to Obfective: Since the burninp of hydrogen provides a poten-tial challenge to the containment, means to prevent or control this burning must be defined and evaluated. The evaluation of these means would include feasibility, potential advantages and disadvantages particularly focused to evaluation of the benefits (negative, neutral or positive) with respect to risk reduction, and the impacts of including the alternative devices.
l Available Information: The Sandia and NSAC hydrogsn compendiums provide a general source for information. The on-going efforts, particularly TVA, provide a substantial further source of information with, respect to , pre-inerting; significant industry effort has been expended and data is available. It has been assumed that present activities being carried out by TVA/ Duke on ignitors and Halon injection will be completed and will form the basis for development of sound industry . positions related to the evalua-tion of these means for prevention or mitigation of burn. Scoce of Recommended Activities: There are three areas in this task: survey of hydrogen detectors, evaluation of pre-inerting for containments not' aiready evaluated, and evaluation of fogging / spray suppression. , e o 9 9
a -
. . j 7- , - e .- must operate must be defined and the capability of the equipment to survive must be developed.
Available Information: Information currently exists on the definition of necessary paramaters as part of the TMI-actions. Information on enif ron-ments either exists through on-going studies or will be developed as part of thi industry program. Scoce of Recommended Activities: For five plant configurations fBWR, each NSSS vendor PWR, and ice 'condensor), the following tasks are recomended:
- a. Identify the minimum set of functions which must be performed or equip-ment which must not operate as a consequence of the environment to permit termination of core degradation sequences or to monitor the status of the plant and return containment.
- b. Eased on "a.", identify necessary minimum set of generic equipment
- c. Identify environments associated with the dominant sequences
- d. Evaluate the survivability of the equipment in "b." for the conditions in "c." and define tests if necessary.
- e. Develop recommendations on equipment survivability criteria and document results of complete task.
Budoet/ Schedule : The budget for this effort is estimated to be 4 3/4 MM ! per type, or 24 MM total . For testing 100K should be included in contingency. This effort should start after the containment sequences are defined. 1
- 8. CORE COOLABILITY Relationship to Objective: The progression of a postulated core melt is the dominant factor in determining the potential for containment challenge.
- The amount of the core involved coherently (on a relatively short time ,
l scale) will effect the generation of steam pressures, the release of ! hydregen, the vessel failure modes, the distribution of material subsequent to vessel failure, the ultimate coolability of the molten rIterial, and 1
.- . fg- ,.
l
. I (1.) Analysis of In-Vessel Core, Melt Progression and Raccolability For each of the dominant accident sequences, determine the progres- ,
sion of the event from boiloff through major geometry disruption to penetration of the debris through the core supporting structure. The progression models should account for prope/ transfer of ti1e heat of zirconium-steam reaction into the fuel, interassambly radia-tion, fuel relocation thermal propagators, melting / refreezing, quenching and particularization, and power profiles. Ex-Vessel Coolability (2.) In-Vessel Coolability, Vessel Penetration The scope of this effort is to establish the limits for in-vessel , coolability of debris, the modes and timing of vessel genetration, and tfut limits for ex vessel coolability. The models developed ' would account for de5ris Sed quenching, heat transfer between core debris and structures, structural failure predictions, conditions
~
for and impact of crust formatien, particularization of debris, limits for coolabtitty of debris beds including power density-- height--particle size--porosity considerations--and other effects which would substantially increase or decrease the debris coolability. (3.) Core Gebris Concrete Reaction ' The scope of this effort includes the critical evaluation of cur- . rently existing models, development of rate of concrete penetration and ncn-condensible gas generation by refining current model estimates, and assimilating and following Sandia and foreign work in this area. i I Budcet/ Schedule : For each. task above, the following sequence would apply--
- a. Literature survey and precise definition of program and results (2-3 months after initiation)
- b. Interim Report covering Items A-E in detail (9-15 months after initiacien)
- c. Final Report including models and experimental data (15-24 months after initiation) .
+
- . - - - - - , .,n. .- , - -,,..-c n, , - - ---- . . , - - - - - - - - - , , , , - - - - , . - - - , - - . - - - , , . ,
k . . . 3.....
/ Budget / Schedule: , ,
- a. 2 m to provide documentation assuming utility contribution of -
personnel to seminar _
- b. 2 m to define loadings; 20 m to perform analysis The seminar should be scheduled early in the program. The inertial loading activity would be initiated following better definition of initial conditions from Task 6 above. '
- 10. syALUATION OF LIQUID PATHWAY DOSE Relaticnship to Objective: The potential for increased risk due to transport of fission products through liquid pathways and the needs for interdiction of these was raised by the NRC. Generally, studies have shown that this is not a major risk contributor. * -
Available Information: NRC Liquid Pathways Study, Sandia Study, W Offshore Power Systems Liquid Pathways Study, and Battelle work all provide infor-mation. In addition, werk at Hanford on leading and transport may be available. Recommended Scoce of Activities: If this issue is raised, integrate available information and provide scoping information on the feasibility time span and cost of source interdiction. Budcet/ Schedule: This work should not be initiated immediately, and in fact may not be necessary. - The effort for the above scope is 3 E. e I
~ *
- j23-8
Establish the major design parameters.and syst,em design (without developing fine detail on design--sizes of filter beds 'are not required). , Define the operational aspects of the system with particular attention I to alternative valve / control arrangements for containment penetration. Evaluation of the benefits and hazards including probability of successful and spurious operations. Define in conjunction with Task 14 below the risk benefits (or lack thereof)'. Budoet/ Schedule: This activity should be initiated later in the program unless NRC activities dictate that useful results would accrue by earlier l performance. L . The estimated resource for this is 20 Mt.
- 13. CORE LADLE ,
Relationship to Objective: One mitigating feature currently being pursued by NRC is the core ladle. It is necessary to identify the advantages and ' disadvantages, real contributions to risk reduction (positive, negative, neutral) and impacts of addition of this feature. Available Information: NRC/Sandia Studies and industry e(f.o rts (Zion-Indian Point, Floating Nuclear Platform) plus significant National Lab and , foreign studies all include material related to core ladle design require-ments, etc. Recommended Scoce of Activities: An evaluition of the real impacts of retrofitting combined with assessments in conjunction with Item 8 that l cooling inherently provided by water loss is more effective is recorrended. Budget / Schedule: This effort should not be initiated immediately unless NRC activities provide another rationale for doing work early. i The estimated resource for this activity is 8 Mi.
3
, ,Recernended Scoce of Activities: The scope of this task is to provide the integrated analyses of the dominant sequences for representative _
~
plants and integrate the models developed in the previous tasks into this integrated analysis. There are five tasks defined below. It has been assumed that the approach will be the modification of the MARCH / CORRAL system augmented by the EPRI/NSAC TMI Heatup Code. This assumption results from the facts that MARCH /CCRRAL is immediately and widely available, provides capability for effective sensitivity analyses if properly " calibrated"7 and will be used by NRC and must therefore be understood in detail by industry. (1.) Define MARCH / CORRAL Use Convene personnel experienced with MARCH / CORRAL application and , degraded core analyses to the minimum mandatory modifications for MARCH / CORRAL prior to use, plans for baselining PARCH / CORRAL for industry use, plans for intr ing phenomenological model development
. into analyses and identif3 L . westrability and directions for major upgrade of MARCF/ CORRAL.
(2.) Containrent Analyses perform analyses for representative containment / system types for dcminant accident sequences and necessary variants. ~ (3.) MARCH /CORPAL Improvements Define program for major developmcnt of PARCH / CORRAL model improve-ments and implement if desired. (4.) Integrate Phenomenology Models into Analyses Include the results of the phenomenological development activities into the integrated transient anlaysis. i - , (5.) EPRI/flSAC TMI Code , Qualif, and dccument the pWR damage progression' code. Develop,
, qualify and document the 3WR damage progression equivalent code.
TASK RESOURCES i ~ M CRU K$ K$
- 1. .2 16 13 4 105
- 2. (1.)
(2.) 18 6 146
,,(3. ) 6 48
- 3. l '. 88 4, (1. ) 18 40 184 24 75 267 -
(2.) 10 20 36 (3.) ,
- 5. 15 20 126 (1. )
30 20 246 (2. ) . . 8 64 (3.) 3 24
- 6. (1.)
12 96 (2.) 3 24 (3.) 24 10 195 7. 50 100 430
- 8. (1.)
60 50 100 595 (2.) 18 144 (3.) 2 16
- 9. (1. )
22 20 182 (2.) 3 24 10. 11. 20 160
- 12. _
8 64
- 13. ,
- 14. 14 20 118
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' l Tfi{Pr INDUSTRY DEGRADED CORE RULEMAKING PROGRAM
. CHARTER
- I. GOALS' AND SUPPORTING OBJECTIVES The Nuclear Regulatory Commission initiated on October 2, 1980, a "long-term rulemaking to consider to what e xt en.t , if any, nuclear power plants should be designed to deal effectively with degraded core and core melt accidents." The NRC's rulemaking proceedings would address the objectives and content of a degraded core-related regulation, the design and operational improvements under consideration, the effects on other safety considerations, and the costs and benefits of design and operational improvements.
In conjunction.with the NRC's rulemaking, the nuclear industry initiated the Industry Degraded Core Rulemaking i (IDCOR) Program. The goals of the IDCOR Program are to develop technical data necessar needed, and y toif determine whether so, to aid changes inregulation in establishing regulationsthat are are: o Consistent with an overall nuclear safety goal o Based on thoughtful ' analysis which carefully considers the costs and benefits of the design or operational improvements which may result from the implementation of any new regulations o Expressed so as to minimize uncertainties with regard to its interpretation and implementation. In support of these goals, the objectives of the.IDCUR Program are to develop concise, logical and well-documented technical bases for use in the degraded core rulemaking proceedings and to coordinate, where applicable, the presentation of the industry's technical bases. II. MANAGEMENT STRUCTURE The management structure of the IDCOR Program builds on the
; goals and objectives of the Program. The organizational
! entities include the: o IDCOR Policy Group o IDCOR Steering Group o Atomic Industrial Fo. rum
- Working Charter; to be approved by Policy Group _l
-l '
l h
EXHIBIT II IDCOR Policy Group, _
-Organization Representative
~
Arizona Public. Service Co. ' The Babcock 6 Wilcox~Co. Mr. D. E. Guilbert Baltimore Gas 6 Electric Co. Mr. Richhrd C. L. Olson Bechtel Power Corp. Mr. Samuel A. Bernsen Black 6 Veatch, Consit. Engr. Mr. M. John Robinson Boston Edison Co. Brown & Root, Inc. Burns & Roe, Inc.* . Mr. Tom A. Hendrickson C.-F. Braun & Co. Mr. Roger N. Moore Carolina Power 6 Light Co.* The Cincinnati Gas 6 Elec. Co. Mr. James D. Flynn The Cleveland Elec. Illuminating Co. Mr. Lawrence 0. Beck Combustion Engineering, Inc. Mr. John M. West Commonwealth Edison Co. Mr. Byron Lee, Jr. Censolidated Edison Co. of N.Y., Inc. Consumers Power Co. Mr.' James W. Cook i Dciryland Power Cooperative
- Mr. Ernest B. Tremmel Daniel International Corp. . Mr. H. W. McCall The Detroit Edison Co.
Duke Power Co. Mr. Warren H. Owen " Duquesne Light Co. Mr. Roger Martin Ebasco Services Inc. Mr. Ed O'Donnell Exxon Nuclear Co. Inc. Mr. Samuel Beard ! Florida Power Corp,.* Florida Power 6 Light Co. Mr. Robert E. Uhrig Fluor Power Services, Inc. Mr. D. M. Leppke
- Gsneral Electric Co. Dr. A. P. Bray'
! Gibbs 6 Hill, Inc. Mr. D. C. Purdy Gilbert Associates In Mr. Sherman D. Goodman Gulf States Utilities.c. Dr. E. Linn Draper Houston Lighting 6 Power Co. Mr. G. W. Oprea i Illinois Power Co. Mr. Leonard J. Koch - Indiana & Michigan Electric Co. Mr. R. S. Hunter Iowa Electric Light 6 Power Co.* l Jersey Central Power 6 Light Co.* Mr. Robert Keaton ! J. A. Jones Construction Co.* Kaiser Engineers, Inc.* Kansas Gas 6 Electric Co. Long Island Lighting Co. Mr. Milliard S. Pollock Metropolitan Edison Co.* Mr. Robert Keaton Middle South Services, Inc. .
. Dr. D. Clark Gibbs e
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l EXHIBIT II (continuad)
. Organization Representative -
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Nebraska Public Power District Mr. Jay M. Pilant Niagara Mohawk Power Corp. LNortheast Utilities Services Co. Mr. William G. Counsil Northern Indiana Public Services Co. - Northern States Power Co. Mr. Robert O. Anderson Omaha Public Power District
- Pacific Gas 4 Electric Co. Mr. S. Taggart Rogers Pennsylvania Power 4 Light Co. Mr. Edward M. Nagel Philadelphia Electric Co. Mr-. Vincent S. Boyer Portland General Electric Co.
Power Authority of the State of N.Y. Mr. John Leonard
~Public Service Co. of Colorado *
," Public Service Co. of Oklahoma Mr. Vaughn L. Conrad Public Service Electric 4 Gas Co. Mr. Robert L. Mitti Public Service Indiana Mr. Seth W. Shields l Puget Sound Power'4 Light Co.
- Rochester Gas 4 Electric Co. Mr. John E. Arthur Sacramento Municipal Utility District Sargent 4 Lundy Mr. John E. Ward South Carolina Electic 4 Gas Co. Mr. Thomas C. Nichols Southern California Edison Co.
Southern Company Services, Inc. . Mr. Ruble A. Thomas Stone 4 Webster Engineering Corp. Mr. William J. Kennedy Tennessee Valley Authority Mr. Dwight R. Patterson Texas Utilities Generating Co. Mr. J. B. George The Toledo Edison Co. Mr. Charles R. Domeck Union Electric Co. United Engineers 4 Constructors, Inc. Mr. Gunnar E. Sarsten Virginia Electric 4 Power Co. Mr. James T. Rhodes l Washington Public Power Supply System Mr. G. D. Bouchey Westinghouse Electric Corp. Mr. John J. Taylor Wisconsin Electric Power Co. Mr. Sol Burstein Wisconsin Public Service Corp.* ' i Yankee Atomic Electric Co. Mr. Don Vandenburgh e I 0 Tentative l j I - l
- .. _ , - - ...__.--.m.. - . . . _ , _ . , _ , _ . . __ _ , . - . . - _ - , _ _ _ . _ _ _ _ , . _
o IDOCR Progran Manager
-o Technical Advisory Group o Legal Group .
o Technical Contractors - Exhibit I depicts the interrelationships of the I,DCOR ~~- management structure.
- 1. IDCOR POLICY GROUP _
The IDCOR Policy Group is responsible for establishing overall policy and funding direction for the IDCOR Program and selecting Steering Group Members. The Policy Group will also act on recommendations concerning key program decisions which are proposed by the IDCOR Steering Group. Voting membership on the Policy. Group includes all companies that contribute funds to the IDCOR Program. The current organizational membership is shown in Exhibit II. Each
. member of the-Policy Group has one vote. The individual designated by each organization as the Policy Group Representative is generally at the senior management level.
Be is responsible for representing his company and, when required, casting his company's vote. In order to ensure continuity each member company should be represented at every scheduled, meeting. As currently planned, the Policy Group will meet semi-annually, or more . often if required, to review IDCOR Program plans, activities and conclusions.
- 2. IDCOR STEERING GROUP The IDCOR Steering Group is responsible for developing and implementing programs consistent with the overall policy developed by the IDCOR Policy Group. Execution of t'hese responsibilities involves:
I o Presenting major program and policy recommendations to l the Policy Group for approval. o Providing operating principles for the Program. o Selecting and directing the IDCOR Program Manager, the Technical Advisory Group and the Legal Group. o Reviewing and acting-o,n the Program Manager's recommendations. o Providing liaison with the nuclear industry and the NRC.
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o Providing the financial management of the IDCOR Program. . o Authorizing the initiation, modification, completion or termination of contracts. ~- - The Steering Group consists of not more than 12 voting i members selected from organizations represented on the ! Policy Group. The Steering Group members should be experienced in the organization and execution of large, multi-functional programs and knowledgeable of utility issues in general, as well as those related to the proposed degraded core rulemaking. Non-voting ' members of the Steering Group include the AIF l IDCOR Project Manager and Chairman of the Technical Advisory Group and the Lea'd Lawyer of the Legal Group. Other non-voting members of the Steering Group sit at the discretion of the Chairman of the Steering Group.
- 3. ATOMIC INDUSTRIAL FORUM The AIF is responsible for staff support to the IDCOR Policy Group, Steering Group. Groupincludes; Technical Advisory Group, and Legal This support o Planning and analysis -- providing periodic assessment of Program performance and activities, resulting in operational procedure and policy recommendations, and -
overall planning and documentation of Program activi ties. o Communication -- providing ongoing liaison with the Program Manager on program activities, other AIF committees, and other organizations as directed by the Steering Group. o Administration -- providing financial, budgeting and control, accounting and contracting services. - The above functions are administered by the AIF -IDCOR Project - Manager. The AIF IDCOR Project Manager reports to the IDCOR Steering Group.
- 4. IDCOR PROGRAM MANAGER The IDCOR Program Manager is responsible for managing the day-to-day operations.of the, Program in accordance with the
tercs of his contract. Exscution of this responsibility ' requires that.the Program Manager , o Recommend' technical programs for Steering Group approval . , o Plan and organize the Program into appropriate sub-programs and tasks 4 o .Byaluate and recommend technical contractors foi Steering Group approval. o Direct, monitor and evaluate the technical work of contractors and consultants through the application of formal and informal management systems , i o Integrate rechnical and legal efforts and formulate i strategy options for the Steering Group o Produce, distribute, and present progress reports l' o Provide technical inputs or guidance for, participate in, or coordinate the development of testimony by - qualified individuals for the rulemaking proceedings
, including any preparation sessions, pre-hearings and hearings.
r r . l The Program Manager reports to the IDCOR Steering Group.
- 5. IDCOR TECHNICAL ADVISORY GROUP 4 l
l The objective of the IDCOR Technical Advisory Group (TAG) is ! to advise the IDCOR Program Manager and Steering Group on ,
! the technical elements of the IDCOR Program. Principally, i j the TAG works constructively with the Program Manage,r in the ! identification of problems and methods of solution regarding technical i issues associated with the scope of the degraded core rule ,
making process. Through its Chairman, the TAG also advises t j the IDCOR Steering Committee of technical program j developements. 'l The membership of the TAG includes industry gen (ralists, - ! knowledgeable of overall degraded core issues, and technical l experts, familiar with the specific technical aspects of the
; IDCOR Program. Membership in the TAG is not limited to members of the IDCOR Policy Group.
i j The scope of TAG activities includes advising on all technical aspects of the IDCOR program, including IDCOR i technical strategy. The TAG scope is not primarily one of j in-depth peer review of ind.ividual pieces of technical work f : i I , J .
done in the program, but such work may be included under special circumstances when re or the IDCOR Steering Group. quested by the Program ManagerIn addition to general s to the Program Manager, areas in which the TAG focuses on: o The understanding of technical issues and the r-- development of technical strategies, o Knowledge and utilization of non-IDCOR degraded core ., programs, o Completeness of technical program; prioritization of program components. l o Reviewing recommendations on suitable technical contractors. o Quality of' technical contractor work; technical responsiveness to probable degraded core issues, o Technical content of program milestones and their accomplishment. o Program integration. The TAG documents its activities with formal meeting minutes, prepared by the TAG Secretary, which reflect the TAG's key conclusions and recommendations.
- 6. IDCOR LEGAL GROUP The IDCOR Legal Group (LG) is responsible for providing legal support on IDCOR activities. The LG provides support to and takes the lead from the Program Manager. The LG shall be comprised of a primary Legal Firm, consisting of a Lead Lawyer and supporting lawyers. The Lead Lawyer may,'at his discretion, convene meetings of a larger set of industry lawyers. Through its Lead Lawyer, the LG also advises the IDCOR Steering Group of program developments of a legal nature.
III. IDCOR POSITIONS - -
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- 1. POLICY GROUP CHAIRMAN The responsibilities of the Policy Group Chairman are to:
o Schedule and chair all Policy Group meetings 6
o .Nagotiato participation end funding for the IDCOR Program o Act on Steering Group recommendations to the Policy . Group . __ s - o Communicate with senior management in government and industry on the activities of the IDCOR Program.
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- 2. STEERING GROUP CHAIRMAN /VICE-CHAIRMEN 4
The responsibilities of the Steering Group Chairman are tc: o Schedule and chair all , Steering Group meetings
- o. Report the activities of the Steering Group to the Policy Group o Communicate with the NRC and industry the activities
! of the IDCOR Program. o Take on any other responsibilities delegated to him by the Steering Group. The Vice Chairmen will assume the authority and responsibility of the Chairman in his absence. 9
- 3. ATOMIC INDUSTRIAL FORUM PROJECT MANAGER The responsibilities of the AIF Project Manager are to:
o Provide staff support to the IDCOR Policy Group, Steering Group, Technical Advisory Group and Legal Group , o Provide planning and analysis support for IDCOR Program activities o Provide ongoing liaison with the ID,COR Program Manager and the technical contractors - o Document the plans and activities of the -IDCOR Policy - Group, Steering Group, Technical Advisory Group and
. Legal Group o Oversee IDCOR's financial administration including financial planning, budgeting and control and accounting -
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e 4
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- 4. TECHNICAL ADVISORY GROUP CHAIRMAN /VICE-CHAIRMAN The responsibilities of'the Chairman of the TAG are to: .
o Schedule, prepare agendas for (in consultation with' Program Manager), and conduct meetings of the TAG I o Appoint a Vice-Chairman and Secretary _ o Manage the TAG's review of technical issues and report findings to the Program Manager o Appoint subcommittees or task groups, for special assignments. . o Report the. activities of the TAG to the Steering Group o Establish the TAG's working procedures and policies o Forward minutes of TAG meetings to members of the TAG, Steering Group, the Program Manager, and the AIF IDCOR Project Manager. The Vice-Chairman will assume the authority and responsibilities of the Chairman in his absence. IV. PROCEDURES
- 1. VOTING
- 1) Policy GrouE '
o Member organizations contributing funds to the IDCOR Program are eligible to cast one vote , o A quorum is one-half of the voting members. A simple majority vote of the voting members determines each ' issue before the Policy Group. I ! o A simple majority vote of the voting members - i determines each issue before the Policy Group
- 2) Steering Group -
I o Full members of the Steering Group are eligible to j cast one vote o A quorum is two-thirds of the voting members. ! o A simple ma'jority vote of the voting members [ , determines each issue before the Steering Group.
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. ~ -- , e w e mw n e w . . , - - m-<-v . - . , . - - -.~-~n-+-mwm-~rmew~mm~~~n-
s, .
- 2. FUNDING Funding for the IDCOR Program is based on the following principles:
o Shares among contributing groups are: Utilities 80% NSSS Vendors
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10% Architect Engineers / ' Contractor 104 - o For each utility contributor, the following formula applies-One-half of total utility amount shared on an equal basis by all nuclear utilities One-half of total utility amount shared on a per reactor basis, with a limit of six reactors for a individual contributor Reactors include those in operation, under construction, or an active order (i.e., NTCPs). Where applicable, the maximum total ~ share for any holding compang operation will not exceed the maximum utility share ',1.e., for a six reactor utility). o For each vendor contributor, the following formula applies: One-hal'f of total vendor amount shared on an equal basis by all vendors One-half of total vendor amount shared on 4 per reactor bases Reactors include those in operations, under construction, or on active order (NTCPs) -- o For each architect engineer / contributor, the following formula applies. One-half of total fee shared on an equal basis by all One-half of total fee shared on a per reactor basis Reactors include those 1.n operation, undar construction, or on active order (NTCPs), I S 4
5
- 3. CONTRACTOR PROCUREMENT The following guidelines govern the selection process for IDCOR contractors: __.
o The contractor selection process will be systematic, objective and documented
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o Competitive proposals will be solicited for contracts exceeding $50,000.00 unless the specific situation warrants otherwise, e.g. if:
- The time urgency of the work is so great that formal solicitation of proposals would be precluded
- A person or organization are known to be uniquely qualified because of prior work assignments, special facilities, etc. ,
o' The authorization for a potential contractor to proceed to contract to perform work or modify work within the scope of the IDCOR Program requires the, approval of the: Steering Group, fo.r individual contracts or modifications in excess of $50,000. Chairman of the Steering Group, for individual contracts or modifications up to $50,000; the cumulative total of such contracts or modifications will not exceed $250,000.00 in any 1-year period. AIF Project Manager, for individual contracts or modifications less than $10,000; the cumulitive total of such contracts or modifications will not exceed $50,000.00 in any 1-year period j 4. REPORTING REQUIREMENTS: REPORT DISTRIBUTION l
-l l The following principles serve as guidelines for IDCOR reporting requirements: - --
o The Steering Gr'oup will present formal program status reports to the Policy Group on a semi-annual basis o The Program Manager will present: Informal progress , reports, with supporting , documentation, to the Steering Group on a monthly basis
Formal, fully documented Program status reports to the Steering Gr,oup on a quarterly basis
- 5. AMENDMENTS . _
This Charter must be approved, and amendments ratified, by the Policy Group. 4 e 6 O O 4 e
- e en k ,
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. i i
g j ,. .-:' t i INDUSTRY DEGRADED CORE RULEMAKING GDCOR) PROGRAM JOHN SIEGEL, MANAGER SPECIAL LICENSING PROJECTS ATOMIC INDUSTRIAL FORUM PRESENTED ATTHE LICENSING INFORMATION SERVICES CONFERENCE l CLEARWATER BEACH, FLORIDA i -
. APRIL 9,1981
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CONTENTS e INTRODUCTION - BACKGROUND AND ' PROGRAM GOALS AND OBJECTIVES
- PROGRNM MANAGEMENTSTRUCTURE e PROGRAM TECHNICALSTRUCTURE e CURRENTSTATUS l
- NRC ACTIVITIES .
l e
*4 8I
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1 1 N O I N o T O I C T U P I D N R . O I G C < R I S . T R E o N O D o I M M . I A A . R R . G G o O O R P R P o. P e e . 4 l ! 1
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. . j PROGRAM ORIGIN i I
e THE IDCOR PROGRAM WAS INITIATED IN RESPONSE TO RECENT NRC LICENSING ACTIONS - SPECIFIC PLANT , ACTIONS AS WELL AS A PROPOSED RULEMAKING ON DEGRADED CORE ISSUES - SUGGESTING THAT NRC WAS ' MOVING IN A DIRECTION OF SIGNIFICANT MITIGATION FIXES
- SPECIFIC PLANT LICENSING ACTIONS INCLUDED THE ZION AND INDIAN POINT FACILITY ASSESSMENTS i
- ON OCTOBER 2,1980, THE NRC INITIATED A LONG-TERM RULEMAKING TO CONSIDER TO WHAT EXTENT, IF ANY, NUCLEAR POWER PLANTS SHOULD BE DESIGNED TO >
DEAL EFFECTIVELY WITH DEGRADED CORE AND CORE MELT ACCIDENTS o \ k t l
** s1 . {
? PROGRAM DESCRIPTION e THE U.S. NUCLEAR INDUSTRY BELIEVED THAT THE RULEMAKING COULD HAVE A LARGE IMPACT ON THE VIABILITY OF THE NUCLEAR OPTION i
- THE SUGGESTED MITIGATION FIXES INCLUDED SUCH DESIGN OPTIONS AS CORE LADLES AND FILTER-VENTED .
CONTAINMENTS
- SUCH FIXES COULD HAVE SIGNIFICANT FINANCIAL AND '
4 OPERATIONAL IMPACTS ON NUCLEAR POWER PLANTS, WITH UNCERTAIN SAFETY BENEFITS
- THE U.S. NUCLEAR INDUSTRY THEREFORE, INITIATED THE
$10 MILLION IDCOR PROGRAM, A COLLECTIVE, EXTERNALLY - FUNDED EFFORT t
i i l
l PROGRAM GOALS AND OBJECTIVES i e 'THE GOALS OF THE IDCOR PROGRAM ARE TO DETERMINE WHETHER CHANGES IN REGULATIONS ARE NEEDED, AND IF i SO, TO AID IN ESTABLISHING REGULATIONS THAT ARE: . I
- CONSISTENT WITH AN OVERALL NUCLEAR SAFETY GOAL
- BASED ON THOUGHTFUL ANALYSIS WHICH CAREFULLY CONSIDERS THE COSTS AND BENEFITS OF THE DESIGN OR OPERATIONAL IMPROVEMENTS WHICH MAY RESULT FROM THE IMPLEMENTATION OF ANY NEW REGULATIONS
- EXPRESSED SO AS TO MINIMIZE UNCERTAINTIES WITH REGARD TO ITS INTERPRETATION AND IMPLEMENTATION.
- IN SUPPORT OF THESE GOALS, THE OBJECTIVES OF THE IDCOR PROGRAM ARE TO DEVELOP CONCISE, LOGICAL AND WELL-DOCUMENTED TECHNICAL BASES FOR USE IN THE
{ DEGRADED CORE RULEMAKING PROCEEDINGS AND TO COORDINATE, WHERE APPLICABLE, THE PRESENTATION OF THE INDUSTRY'S TECHNICAL BASES. e i i O
$1 .
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II. PROGRAM MANAGEMENT STRUCTURE THE MANAGEMENT STRUCTURE OF THE IDCOR PROGRAM BUILDS ON THE GOALS AND OBJECTIVES OF THE PROGRAM. THE ORGANIZATIONAL ENTITIES INCLUDE THE: e IDCOR POLICY GROUP e IDCOR STEERING GROUP e ATOMICINDUSTRIAL FORUM e IDCOR PROGRAM MANAGER e TECHNICAL ADVISORY GROUP e LEGAL GROUP e TECHNICAL CONTRACTORS EXHIBIT I DEPICTS THE INTERRELATIONSHIP OF THE IDCOR MANAGEMENT STRUCTURE. 1 t jl
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EXHIZIT I - 1 12CCR MANACEMENT OTRUCTURE I i
; IDCOR POLICY GROUP e Overallpolicyanddirection j
j IOcOR
; ..............................' AIF
- STEERING GROUP ' " " " " * " " " " * " * " * " " * " -
=
l j e Generalpolicyanddirectiotr ! e Planning. analysis andliaison j j e Overalloperation : e Financialadmuustration j : j
- Revsew/approvalof Program j i Manager recommendations l :
- Contract authorization j l l e Lisesonwithindustry/NRC l
i : : l l I TECHNICAL IDCOR LEGAL ADVISORY GROUP PROGRAM MANAGER G8BOUP j O Consultation /reviewof e Management of day-to-day e Consultation and support technical program development operations onlegalissues
- and issue analysis e Technicalprogram development
, e Contractorselectionand management i e Technical monitoring and reporting e Integration /presentationof technicalofforts i
- TECHNICAL j
CONTRACTORS i t e Analysis / reporting of specific technicalissues
- Dotted knes indicate lieeson with Steering Group gl
l MANAGEMENT STRUCTURE (Continued) l l l e POLICYGROUP - I THE POLICY GROUP IS RESPONSIBLE FOR ESTABLISHING OVERALL POLICY AND FUNDING DIRECTION FOR THE IDCOR PROGRAM. e STEERINGGROUP ' THE STEERING GROUP IS RESPONSIBLE FOR DEVELOPING AND IMPLEMENTING PROGRAMS CONSISTENT WITH THE OVERALL POLICY DEVELOPED BY THE IDCOR POLICY GROUP.
- ATOMICINDUSTRIAL FORUM THE AIF IS RESPONSIBLE FOR STAFF SUPPORT - PLANNING AND ANALYSIS, LIAISON, AND FINANCIAL AND CONTRACT ADMINISTRATION - TO THE IDCOR POLICY GROUP,
) STEERING GROUP, TECHNICAL ADVISORY GROUP AND ! LEGAL GROUP. ) 9 1 5
MANAGEMENT STRUCTURE (Ccat/nued) e PROGRAMMANAGER THE PROGRAM MANAGER IS RESPONSIBLE FOR MANAGING THE DAY-TO-DAY OPERATIONS OF THE PROGRAM. EXECUTION OF THIS RESPONSIBILITY REQUIRES THAT THE PROGRAM MANAGER:
- RECOMMEND TECHNICAL PROGRAMS FOR STEERING GROUP APPROVAL AND ORGANIZE THE PROGRAM INTO l APPROPRIATE SUB-PROGRAMS AND TASKS
- EVALUATE AND RECOMMEND TECHNICAL CONTRACTORS FOR STEERING GROUP APPROVAL
- DIRECT, MONITOR AND EVALUATE THE TECHNICAL WORK l OF CONTRACTORS AND CONSULTANTS THROUGH THE APPLICATION OF FORMAL AND INFORMAL MANAGEMENT SYSTEMS
- INTEGRATE TECHNICAL AND LEGAL EFFORTS AND '
FORMULATE STRATEGY OPTIONS FOR THE STEERING GROUP e THE STEERING GROUP SELECTED TECHNOLOGY FOR ENERGY CORPORATION TO BE THE PROGRAM MANAGER. DR. ANTHONY BUHL WILL HEAD TEC'S MANAGEMENT TEAM e
'8 i .
I MANAGEMENT STRUCTURE (Continued) e TECHNICAL ADVISORYGROUP THE TAG ACTS AS AN ADVISOR AND RESOURCE TO THE IDCOR PROGRAM MANAGER. DR. MILES LEVERETT OF EPRI18 THE TAG CHAIRMAN e LEGAL GROUP THE LG IS RESPONSIBLE FOR PROVIDING SUPPORT ON LEGALISSUES AND STRATEGIES e TECHNICAL CONTRACTORS THE TECHNICAL CONTRACTORS WILL BE RESPONSIBLE FOR CONDUCTING THE DETAILED TECHNICAL ANALYSIS FOR THE PROGRAM ! i' I
III. PROGRAM TECHNICAL STRUCTURE ' e AS CURRENTLY PROJECTED, THE IDCOR PROGRAM WILL CONSIST OF 16 TECHNICAL TASKS TO BE CONDUCTED OVER THE NEXT 12-20 MONTHS e THE MAJOR TASKS AREAS ARE:
+
- RISK ANALYSIS ~
! - REACTOR AND PLANT SYSTEMS l i ; f
- PHENOMENA ANALYSIS I
4 ) j . . I i i I 4 i i TECHNICAL TASKS (Continued) , i 1 i THE RISK ANALYSIS TASKS INCLUDE: e TASK 1 - SAFETY GOAL / CRITERIA APPLICATIONS e TASK 2 - DOMINANT SEO.UENCES e TASK 3 - CONTAINMENT PHENOMENOLOGICAL SEO.UENCES
- TASK 14 - RISK REDUCTION POTENTIAL '
j. i i
1 : a 1 l TECHNICAL TASKS (Continued) THE REACTOR & PLANT SYSTEMS TASKS INCLUDE: e TASK 8 - HYDROGEN BURN CONTROL e TASK 7 - EQUIPMENT SURVIVABILITY i e TASK 9 - CONTAINMENT STRUCTURAL CAPABILITY l e TASK 12 - ALTERNATE CONTAINMENT SYSTEMS e TASK 13 - CORE RETENTION DEVICES e TASK 18 - OPERATIONAL ASPECTS OF ACCIDENT MANAGEMENT AND CONTROL e
r - TECHNICAL TASKS (Continued) THE PHENOMENA ANALYSIS TASKS INCLUDE:
- TASK 4 - STEAM OVERPRESSURE PHEN'OMENA
!
- TASK 5 - HYDROGEN GENERATION AND BURN I
!
- TASK 8 - CORE DEBRid BEHAVIOR &
COOLABILITY e TASK 10 - ATMOSPHERIC & LIOUID PATHWAY I DOSE
- TASK 11 - FISSION PRODUCT LIBERATION, TRANSPORT, AND INHERENT RETENTION
- TASK 15 - INTEGRATED MODEL DEFINITION &
ANALYSIS i 1
1
- IV. PROGRAM STATUS i
CURRENT PROGRAM ACTIVITIES FOCUS ON THE l FUNCTIONS OF THE IDCOR: - l 1 e STEERING GROUP i 1 e PROGRAM MANAGER ) i l . l l I
)
i
- s e,
i l . l PROGRAM STATUS - STEERING GROUP i l
- THE STEERING GROUP HAS DETERMINED A SCHEDULE l FOR SELECTING THE IDCOR LEGAL GROUP. THE l SELECTION PROCESS SHOULD CULMINATE WITH A l RECOMMENDATION TO THE IDCOR POLICY GROUP IN
! JUNE,1981 , e THE STEERING GROUP'S EFFORTS TO COORDINATE INDUSTRY'S FINANCIAL SPONSORSHIP OF THE IDCOR PROGRAM INDICATE THAT SPONSORSHIP WILL TOTAL APPROXIMATELY 90% OF FUNDS REQUESTED, OR l $9 MILLION l ) e THE STEERING GROUP HAS TAKEN UNDER 1 CONSIDERATION THE QUESTION OF FOREIGN l SPONSORSHIP OF THE PROGRAM 7 i l l
'I e
t i ! PROGRAM STATUS - TECHNICAL EFFORTS l . TECHNOLOGY FOR ENERGY CORPORATION IS CURRENTLY DEVELOPING THREE KEY MANAGEMENT TASK AREAS IN THEIR
. PROGRAM MANAGEMENT STARTUP PLAN. INVOLVED ARE:
i e AN ANALYSIS OF STUDIES, PROGRAMS, AND RULEMAKINGS l RELATED TO THE SCOPE OF THE IDCOR PROGRAM. CURRENT PLANS CALL FOR: -
- INFORMATION EXCHANGES WITH EPRI, GENERAL ELECTRIC, NRC, DOE, UTILITY COMPANIES (TVA, DUKE, COMMONWEALTH EDISON, CONSOLIDATED EDISON, ,
PHILADELPHIA ELECTRIC), THE NATIONAL LABS (ORNL, HEDL, ANL, SANDIA, BNL, LASL AND INEL) AND FOREIGN PROGRAMS.
- FOREIGN VISITS ARE SCHEDULED WITH KFK,IKE, GRA AND KWU IN THE FEDERAL REPUBLIC OF GERMANY AND EDF, CEA AND FRAMATONE IN FRANCE.
AS PART OF THIS TASK, TEC - WITH DOE'S ASSISTANCE - IS PLANNING A 2-DAY INFORMATION EXCHANGE MEETING FOR APRIL 14-15,1981 IN WASHINGTON, D.C.
i_ { PROGRAM STATUS: TEC ACTIVITIES . . . ! e THE DEVELOPMENT OF A DETAILED PROGRAM PLAN. j SCHEDULED TO BE AVAILABLE FOR STEERING GROUP REVIEW l IN LATE MAY, THE PLAN WILL BE DEVELOPED IN PARALLEL l WITH THE ABOVE ASSESSMENT OF RELATED TECHNICAL EFFORTS. .
- THE DEVELOPMENT OF TECHNICAL SCOPE STATEMENTS AND l THE SELECTION OF CONTRACTORS TO CONDUCT THE
) TECHNICAL ELEMENTS OF THE IDCOR PROGRAM. 1 - THE TECHNICAL SCOPE STATEMENTS WILL BE AVAILABLE FOR REVIEW BY THE STEERING GROUP IN j LATE APRIL l i ! - INTEGRAL TO THE TECHNICAL CONTRACTOR ! SELECTION PROCESS, TEC PLANS TO PROVIDE A LIST OF POTENTIAL CONTRACTORS FOR STEERING GROUP REVIEW, AND TO HOST A CONTRACTOR INFORMATION MEETING IN KNOXVILLE ON MAY 21,1981. - d.
** l '.
1 i i i i V. NRC DEGRADED CORE ACTIVITIES 1 1 THE NRC'S EFFORTS IN DEGRADED CORE ISSUES j HAVE INVOLVED THE ACTIVITIES OF THE: I e ADVISORY COMMITTEE ON REACTOR SAFEGUARDS .
- NRC STAFF I
i I l 1 i I 8
- 4 ,
!. ACRS ACTIVITIES - i i i l THE ACRS SUBCOMMITTEE ON CLASS 9 ACCIDENTS CONVENED l ON: e MARCH 10-11,1981 TO DISCUSS THE CONCLUSIONS OF AN NRC STAFF REPORT ON SOURCE TERMS, TECHNICAL
- BASES FOR ESTIMATING FISSION PRODUCT BEHAVIOR j DURING LWR ACCIDENTS *', AND TO DISCUSS T.HE REPORT'S REGULATORY IMPACTS.
- THE STAFF'S REPORT APPEARS TO BE AT ODDS WITH '
EARLIER INDUSTRY REPORTS SUGGESTING THAT IODINE ATTENUATION FACTORS IN REGULATORY ANALYSES ARE OVERLY CONSERVATIVE.
- REGARDING POTENTIAL REGULATORY IMPACTS, THE STAFF SUGGESTED THAT A WIDER SPECTRUM OF ,
ACCIDENTS (E.G., CLASS 9) MIGHT HAVE TO BE
- CONSIDERED IN THE FUTURE.
j e MARCH 24,1981 TO DISCUSS INTER ALIA THE STATUS OF ' i DEGRADED CORE RULEMAKING ACTIVITIES WITH THE ) NRC'S STEERING GROUP ON DEGRADED COOLING. I
NRC STEERING GROUP ON DEGRADED COOLING AT THE MARCH 24,1981 ACRS SESSION, THE NRC STEERING GROUP ON DEGRADED COOLING PRESENTED A PROPOSED APPROACH TOWARD DEVELOPING A DEGRADED CORE RULE. THE STAFF: l
- IS ATTEMPTING TO FORMULATE A RESEARCH PROGRAM e PROGRAM COULD INCORPORATE VALUE-IMPACT ASSESSqWIENTS l
- PROGRAM COULD INVOLVE A PRELIMINARY PHASE TO .
l SCREEN ALTERNATIVE DESIGN FEATURES (4/81-1/82) AND A SPECIFIC DESIGN FEATURE ANALYSIS PHASE (7/81-4/83)
- PRELIMINARY SCHEDULE INCLUDES A DRAFT RULE TO THE COMMISSION BY MARCH,1982 AND A FINAL RULE BY JUNE,1983.
i
i NRC STEERING GROUP (Continued) THE NRC STEERING GROUP: i e APPEARS TO BE EMPHASIZING A MITIGATION APPROACH l
- IS CONSIDERING RECOMMENDING TO NRC"S i EXECUTIVE DIRECTOR FOR OPERATIONS THAT THE ESF AND DEGRADED CORE RULEMAKINGS BE CONDUCTED IN PARALLEL e MAY BE PLACING ADDITIONAL EMPHASIS ON OPERATOR ACTIONS e
I i4 ei +
FUTURE NRC ACTIONS
- THE NRC STEERING GROUP ON DEGRADED COOLING WAS DISBANDED AT THE END OF MARCH e RESPONSIBILITY FOR DEGRADED CORE ISSUES WILL PASS TO THE RECENTLY REORGANIZED OFFICES OF STANDARDS AND REGULATORY RESEARCH, POSSIBLY TO THE l
DIV!SION OF RISK ANALYSIS l
- AN ACTION PLAN ON AN APPROACH TO THE DEGRADED l CORE RULEMAKING, ORIGINALLY SCHEDULED TO BE SENT l TO THE COMMISSIONERS ON APRIL 1, WAS FORWARDED TO THE EXECUTIVE DIRECTOR'S OFFICE IN EARLY APRIL l
.. .' . . l
O
~
- Agreement L D C. = .4.- r- dd?S
- 1. Meet with 30 days Tech Progra Revieg Agenda -
% GJcJ
- a. Decision criteria as related tof, timing of research and-as-related-
--to-whc4her/how the require changes.
+
- b. Research plans - gaps and overlaps
- c. Future meetings; DOE role
- d. ACRS Interaction (August)
- e. Nominal schedu'le that leads to issuance of a proposed rule
- f. Disscmination of meeting minutes
- g. Proprietary info
- h. Commission briefing e
8)ag _ = = = ~ . . - -
6 INDUSTRY DEGRADED CORE PROGRAM -
- I D C O R --
PRESENTED TO NRSD AWARDS LilNCHEON AMERICAN NUCLEAR SOCIETY BAL HARBOR HOTEL BAL HARBOR, FLORIDA JUNE 10, 1981 SY ANTHONY R. BUHL VICE PRF,SIDENT 8 BOARD OF DIRECTORS TECHNOLOGY FOR ENERGY CORPORATION 10770 DUTCHTOWN ROAD KN0XVILLE, TENNESSEE 37922 615/955-5856 , 9
o. INDUSTRY DEGRADED CORE PROGRAM 1 *
'- IDCOR - .
ANTHONY R. BUHL VICE PRESIDENT TECHNOLOGY FOR ENERGY CORPORATION LIGHTENING RODS!
"THE MORE POINTS OF IRON ARE ERECTED AROUND THE EARTH, TO DRAW ELECTRICAL SUBSTANCE OUT OF THE AIR, THE MORE THE EARTH MUST NEEDS BE CHARGED WITH IT. AND THEREFORE, IT SEEMS WORTHY OF CONSIDERATION, WHETHER ANY PART OF THE EARTH BEING FULLER OF THIS TERRIBLE SUBSTANCE MAY NOT BE MORE EXPOSED TO MORE SH0CKING EARTHOUAKES. IN BOSTON ARE MORE ERECTED THAN ANYWHERE ELSE IN NEW ENGLAND; AND BOSTON SEEMS TO BE MORE DREADFULLY SHAKEN. OH!
THERE IS N0 GETTING OUT OF THE MIGHTY HAND OF GODI IF WE THINK TO AVOID IT IN THE AIR, WE CANNOT IN THE EARTH: YET IT MAY GROW MORE FATAL: AND THERE IS NO SAFETY ANYWHERE..." PUBLIC SAFETY HAS ALWAYS BEEN AN IMPORTANT TOPIC. MY'SERM0'N WAS FIRST DELIVERED BY THE RIGHT REVEREND THOMAS PRINCE IN THE OLD SOUTH BOSTON CHURCH AGAINST BENJ AMIN FRANKLIN'S USE OF THE LIGHTENING ROD IN 1751--THE FIRST INVENTOR AGAINST THE USE OF ELECTRICITY. WE HAVE OFTEN FOUND SAFETY ISSUES WROUGHT WITH EMOTION. . WAY BACK IN 1889, THOMAS EDISON WAS TROUBLED. HE WR0TE AN ARTICLE FOR THE SCIENTIFIC AMERICAN WARNIN0 THE PUBLIC ABOUT WHAT HE PERCEIVED AS A MAJOR PUBLIC DANGER.
"MY PERSONAL 0,ESIRE WOULD BE TO PROHIBIT ENTIRELY THE USE OF ALTERNATING CURRENTS," EDISON WROTE. "THEY ARE UNNECESSARY AS THEY ARE DANGEROUS. I CAN THEREFORE SEE NO JUSTIFICATION FOR THE INTRODUCTION OF A SYSTEM WHICH HAS NO ELEMENT OF PERMANENCY AND EVERY ELEMENT OF ^
DANGER TO LIFE AND PROPERTY." , i
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NOW, FROM THE VANTAGE POINT OF OUR ALTERNATING CURRENT WORLD' 90 YEARS LATER, IT IS APPARENT THAT THIS GREAT PERSON EITHER WAS INEXPLICABLY; WRONG IN PRINCIPLE, OR HE FAILED TO ANTICIPATE THE TECHNOLOGY THAT PUT A.C. ELECTRICITY INTO NEARLY UNIVERSAL USE. I TODAY WORLD LEADERS AND THE PUBLIC ALIKE ARE E00 ALLY CONCERNED ABOUT NUCLEAR SAFETY--BUT LIKE THE INDIVIDUALS IN THESE HISTORIC EXAMPLES WE DO NOT NEED OVER REACTIONS. REACTOR LICENSING IN THE UNITED STATES HAS ALWAYS BEEN AN ADVER-SARY PROCESS IN WHICH THE APPLICANT FOR A PLANT LICENSE PRESENTS A CASE TO NRC SHOWING THAT PLANT ,0PERATION- POSES NO REAL DANGER TO THE HEALTH AND SAFETY OF THE PUBLIC. THE SUM TOTAL OF ALL i SAFETY AND ENVIRONMENTAL ISSUES ARE ADDRESSED THROUGH A LENGTHY
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LICENSING PROCESS IN WHICH THE PLANT DESIGN AND SITE CHARAC-TERISTICS ARE DOCUMENTED IN SAFETY ANALYSIS REPORTS, ENVIRONMEN-1 TAL REPORTS AND OTHER DOCUMENTS, AND AFTER MUCH INTERACTION WITH , i NRC, MAY BE AIRED IN PUBLIC HEARINGS. PART OF THIS PROCESS INVOLVES TECHNICALLY DEMONSTRATING THAT THE i PLANT IS DESIGNED TO WITHSTAND CERTAIN S0-CALLED DESIGN BASIS ACCIDENTS (DBA). ARGUMENTS REGARDING THE MERITS OF PREVENTION VERSUS MITIGATION OF ' ACCIDENTS HAVE CONTINUED OVER THE YEARS - '
- THE QUESTION BEING "SHOULD NUCLEAR PLANTS BE DESIGNED TO WITHSTAND DEGRADED CORE ACCIDENTS?" ALTHOUGH SPECIFIC ISSUES, SUCH AS WHETHER CORE CATCHERS ARE NEEDED, HAVE BEEN FACED BY
. . _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . _ _ __ _ .a
~"~2 SEVERAL APPLICANTS FOR PLANTS DURING LICENSING, LIGHT-WATER REAC-TORS (LWR) IN THE U.S. HAVE NOT BEEN REQUIRED TO CONSIDER DEGRADED CORE ACCIDENTS AS PART OF THE DESIGN BASIS FOR THE PLANT.
ORIVEN BY EVENTS FOLLOWING THE ACCIDENT AT THREE MILE ISLAND, THE NUCLEARREGULATORYCOMMkSSION (NRC) DECIDED TO RE-EXAMINE THE RULES COVERING ACCIDENTS INVOLVING PEGRADED CORE CONSIDERATIONS. ON OCTOBER 2, 19 80, , NRC INITIATED A LONG-TERM RULEMAKING TO CONSIDER TO WHAT EXTENT, IF ANY , NUCLE AR POWER PLANTS SHOULD BE DESIGNED TO DEAL EFFECTIVELY WITH DEGRADED CORE AND CORE MELT
~
ACCIDENTS. THE NRC'S RULEMAKING PROCEEDINGS WILL ADDRESS THE THE OBJECTIVES AND CONTENT OF A DEGRADED CORE-RELATED REGULATION; DESIGN AND OPERATIONAL IMPROVEMENTS UNDER CONSIDERATION; THE EFFECTS ON OTHER SAFETY CONSIDERATIONS; AND THE COSTS AND BENEFITS OF DESIGN AND OPERATIONAL IMPROVEMENTS. IN PARALLEL'WITH THESE ACTIVITIES, THE NUCLEAR INDUSTY HAS DEVEL-OPED THE INDUSTRY DEGRADED CORE PROGRAM -- IDCOR - TO PREPARE AN ADEQUATE POSITION ON THE MANY TECHNICAL ISSUES SURROUNDING THE DEGRADED CORE QUESTION AND TO REACH SATISFACTORY RESOLUTION WITH THE NRC ON THIS MATTER. THE GOALS OF THE IDCOR PROGRAM ARE TO DEVELOP TECHNICAL DATA NECESSARY TO DETERMINE WHETHER CHANGES IN REGULATIONS ARE NEEDED, AND IF S0, TO AID IN ESTABLISHING REGULA-
~
TIONS THAT ARE: _ a,- - - - > -
'r- y r ,w4--g,y -we g - , , . , - -,.w-y , - , , ,wy- y_. ,c-, -, wv----,,..,,-3 ,
- CONSISTENT WITH AN OVERALL NUCLEAR SAFETY G0AL
- BASED ON THOUGHTFUL ANALYSIS WHICH CAREFULLY CONSIDERS THE COSTS AND BENEFITS OF THE DESIGN OR OPERATIONAL IMPROVEMENTS WHICH MAY RESULT FROM THE IMPLEMENTATION OF ANY NEW REGULATIONS e EXPRESSED SO AS TO MINIMIZE UNCERTAINTIES WITH REGARD TO THEIR INTERPRETATION AND IMPLEMENTATION THE OVERALL OBJECTIVE $F THE IDCOR PROGRAM IS TO PROVIDE A STRONG, DOCUMENTED TECHNICAL BASE AND TO DEVELOP INTEGRATED TECHNICAL AND LEGAL OPTIONS TO SUPPORT INDUSTRY PARTICIPATION IN THE ANTICI-PATED NRC DEGRADED CORE RULEMAKING PROCESS.
THE FOLLOWING GENERAL STRATEGY IS STRUCTURED TO ACHIEVE THIS OBJECTIVE--THAT IS TO ASSURE THAT A RULE, IF DEVELOPED, WILL BE BASED ON TECHNICAL MERITS AND WILL BE ACCEPTABLE TO THE NUCLEAR INDUSTRY. THE GENERAL STRATEGY IS TO:
- 1. DEVELOP AND EXECUTE AN EARLY ACTION PLANNING PROCESS TO REVIEW EXISTING IDCOR RELATED INFORMATION, TO DEVELOP ,
. TECHNICAL SCOPE STATEMENTS, TO OBTAIN BROAD TECHNICAL AND ORGANIZATIONAL REVIEWS, AND TO PRODUCE AN IDCOR PROGRAM PLAN BY JUNE 1, 1981.
- 2. CONDUCT THE TECHNICAL PROGRAM AS PLANNED TO ASSURE THAT AN ADEOUATE DOCUMENTED TECHNICAL BASIS EXISTS TO SUPPORT A CAREFUL EXAMINATION OF THE DEGRADED CORE ISSUES.
- 3. DEVELOP INTEGRATED TECHNICAL AND LEGAL OPTIONS CON-SISTENT WITH THE ELEMENTS OF THE ANTICIPATED RULEMAKING PROCESS.
- 4. WORK WITH NRC TO REACH CONVERGENCE ON THE TECHNICAL MERITS OF THE DEGRADED CORE ISSUES AND TO REACH AGREEMENT ON ANY PROPOSED RULEMAKING PROCESS S0 THAT ANY POTENTI AL RULEMAKING WILL BE BASED ON TECHNICAL MERIT ~
AND NOT DEVELOP INTO A PROTRACTED PROCEDURAL BATTLE.
~
MANAGEMENT STRUCTURE THE MANAGEMENT STRUCTURE OF THE IDCOR PROGRAM INCLUDES THE.- FOLLOWING ORGANIZATIONAL ENTITIES:
- IDCOR POLICY GROUP e IDCOR STEERING GROUP
- ATOMIC INDUSTRIAL FORUM e IDCOR PROGRAM MANAGER e TECHNICAL ADVISORY GROUP e LEGAL ADVISOR
- SENIOR CONSULTANTS
- TECHNICAL CONTRACTORS s
FIGURE 1 SHOWS THE INTERRELATIONSHIPS OF THE IDCOR MANAGEMENT 1 STRUCTURE. IDCOR POLICY GROUP
'THE IOCOR POLICY GROUP IS RESPONSIBLE FOR ESTABLISHING OVERALL
- l POLICY AND FUNDING DIRECTION FOR THE IDCOR PROGRAM AND SELECTING STEERING GROUP MEMBERS. THE POLICY GROUP WILL ALSO ACT ON RECOM-MENDATIONS CONCERNING KEY PROGRAM DECISIONS WHICH ARE PROPOSED BY THE IDCOR STEERING GROUP.
THE POLICY GROUP IS CHAIRED BY MR. JOHN SELBY AND INCLUDES ALL COMPANIES THAT CONTRIBUTE FUNDS TO THE IDCOR PROGRAM. THE INDI-VIDUAL DESIGNATED BY EACH ORGANIZATION AS THE POLICY GROUP REPRESENTATIVE IS GENERALLY AT THE SENIOR MANAGEMENT LEVEL. AS CURRENTLY PLANNED, THE POLICY GROUP WILL MEET SEMI-ANNUALLY,
. 0R MORE OFTEN IF REOUIRED, TO REVIEW IDCOR PROGR AM PL A'NS , ACTIVI- ,
TIES AND CONCLUSIONS. r- -. , - + - - . . -.-n, , , , . . , - _ . , . , . , , , .,7-,,. - - . .-..y.__.n.--.v- -,.,-,_,.,n- - . - . . , . . - . .
- TECHNOLOGY far ENER3Y CORPORATION -
t l POLICY GROUP ' J. SELBY, CHAIRMAN - 4 IDCOR STEERING ATOMIC GROUP ----------------- " INDUSTRIAL C. REED, CHAIRMAN i FORUM 3 J. SIEGEL l IDCOR
- j. TEC RESPONSIBLE LEGAL i
OFFICER ADVISOR 1 A.R. BUHL, VICE PRESIDENT j IDCOR IDCOR PROGRAM OFFICE l TECHNICAL M.H. FONTANA, PROGRAM DIRECTOR } ADVISORY GROUP R.D. MOORE, PROJECT MANAGER { M. LEVERETT, CHAIRMAN j SENIOR j CONSULTANTS i H.K. FAUSKE i N. RASMUSSEN t R. SEALE W. STRATTON j . l RISK ANALYSIS REACTOR & MANT ' SYSTEMS PHENOMENA
& LICENSING j S.V. ASSELIN ED FULLER ADMINISTRATION E.P. STROUPE R.M..SATTERFIELD C.R. NAULT l
Floure 1 IDCOR PROGRAM ORGANIZATION i , I .,
!.i nmp er a ii
- IDCOR STEERING GROUP -.
THE IDCOR STEERING GROUP IS RESPONSIBLE FOR DEVELOPING AND IMPLE-MENTING PROGRAMS CONSISTENT WITH THE OVERALL POLICY DEVELOPED BY
'THE IDCOR POLICY GROUP. EXECUTION OF THESE RESPONSIBILITIES INVOLVES:
t e PRESENTING MAJOR PROGRAM AND POLICY RECOMMENDATIONS TO THE POLICY GR0uP FOR APPROVAL e PROVIDING OPERATING PRINCIPLES FOR THE PROGRAM e SELECTING AND DIRECTING THE IDCOR PROGRAM MANAGER, THE TECHNICAL ADVISORY GROUP AND THE LEGAL ADVISOR e REVIEWING AND ACTING ON THE PROGRAM MANAGER'S RECOMMEN-DATIONS e PROVIDING LI AISON WITH THE NUCLEAR INDUSTRY AND THE NRC e PROVIDING THE FINANCI AL MANAGEMENT OF THE IDCOR PROGRAM e AUTHORIZING THE INITIATION, MODIFICATION, COMPLETION OR TERMINATION OF CONTRACTS THE STEERING GROUP IS CHAIRED BY MR. CORDELL REED AND CONSISTS OF NOT MORE THAN 12 VOTING MEMBERS SELECTED FROM ORGANIZATIONS REPRESENTED ON THE POLICY GROUP. THE STEERING GROUP MEMBERS ARE EXPERIENCED IN THE ORGANIZATION AND EXECUTION OF LARGE, . MULTI-FUNCTIONAL PROGRAMS AND KNOWLEDGEABLE OF U.TILITY ISSUES IN GENERAL, AS WELL AS THOSE RELATED TO THE PROPOSED DEGRADED CORE RULEMAKING. NON-VOTING MEMBERS OF THE STEERING GROUP INCLUDE THE AIF IDCOR PROJECT MANAGER AND CHAIRMAN OF THE TECHNICAL ADVISORY GROUP AND THE IDCOR LEGAL ADVISOR. OTHER NON-VOTING MEMBERS OF THE , STEERING GROUP SIT AT THE DISCRETION OF THE CHAIRMAN OF THE STEERING GROUP.
~.
m ATOMIC INDUSTRIAL FORUM (AIF) THE AIF IS RESPONSIBLE FOR STAFF SUPPORT TO THE IDCOR POLICY _ GROUP, STEERING GROUP, TECHNICAL ADVISORY GROUP, AND LEGAL ADVISOR. THIS SUPPORT INCLUDES: o PLANNING AND ANALYSIS -- PROVIDING PERIODIC ASSESSMENT OF PROGRAM PERFORMANCE AND ACTIVITIES, RESULTING IN OPERA-TIONAL PROCEDURE AND POLICY RECOMMENDATIONS, AND OVERALL DOCUMENTATION OF PROGRAM ACTIVITIES o COMMUNICATION -- PROVIDING ONGOING LIAISON WITH THE PROGRAM MANAGER ON PROGRAM ACTIVITIES, OTHER AIF COMMITTEES, AND OTHER ORGANIZATIONS AS DIRECTED BY THE STEERING GROUP e ADMINISTRATION -- PROVIDING FINANCIAL, BUDGETING AND CONTROL, ACCOUNTING AND CONTRACTING SERVICES i THE AB0VE FUNCT' IONS ARE ADMINISTERED BY THE AIF IDCOR PROJECT MANAGER, MR. JOHN SEIGEL. THE AIF IDCOR PROJECT MANAGER REPORTS T0'THE IDCOR STEERING GROUP. IDCOR PROGRAM MANAGER IDCOR PROGRAM MANAGER, TECHNOLOGY FOR ENERGY CORPORATION, IS f RESPONSIBLE FOR MANAGING THE DAY-TO-DAY OPERATIONS OF THE PROGRAM IN ACCORDANCE WITH THE TERMS OF HIS CONTRACT. EXECUTION OF THIS l RESPONSIBILITY REQUIRES THAT THE PROGRAM MANAGER: e RECOMMEND TECHNICAL PROGRAMS FOR STEERING GROUP APPROVAL e PLAN AND ORGANIZE THE PROGRAM INTO APPROPRI ATE SUB-PROGRAMS AND TASKS e EVALUATE AND RECOMMEND TECHNICAL CONTRACTORS FOR STEERING , GROUP APPROVAL _ e DIRECT, MONITOR AND EVALUATE THE TECHNICAL WORK OF CONTRACTORS AND CONSULTANTS THROUGH THE APPLICATION OF FORMAL AND INFORMAL MANAGEMENT SYSTEMS
.g.
o INTEGRATE TECHNICAL AND LEGAL EFFORTS AND FORMULATE STRATEGY OPTIONS FOR THE STEERING GROUP
- PRODUCE, DISTRIBUTE, AND PRESENT PROGRESS REPORTS ~
o DEVELOP THE INDUSTRY'S TECHNICAL POSITION o PROVIDE TECHNICAL INPUTS OR GUIDANCE FOR, PARTICIPATE IN, OR COORDINATE THE DEVELOPMENT OF TESTIMONY BY QUALIFIED INDIVIDUALS FOR THE RULEMAKING PROCEEDINGS INCLUDING ANY PREPARATION SESSIONS, PRE-HEARINGS AND HEARINGS e COORDINATE THE INTERFACE WITH NRC THE PROGRAM HAS BEEN ORGANIZED INTO 24 TECHNICAL PROGRAMS AND TASKS AND A MANAGER ASSIGNED TO EACH AS SHOWN IN FIGURE 1. THE PROGRAM MANAGER REPORTS TO THE IDC~0R STEERING GROUP. IDCOR TECHNICAL ADVISORY GROUP THE IDCOR TECHNICAL ADVISORY GROUP (TAG) IS RESPONSIBLE FOR ADVISING THE IDCOR PROGRAM MANAGER AND STEERING GROUP ON THE TECHNICAL ELEMENTS OF THE IDCOR PROGRAM. PRINCIPALLY, TAG WORKS CONSTRUCTIVELY WITH THE PROGRAM MANAGER IN THE IDENTIFICATION OF PROBLEMS AND METHODS OF SOLUTION REGARDING TECHNICAL ISSUES ASS 0-i ! CIATED WITH THE SCOPE OF THE DEGRADED CORE RULEMAKING PROCESS. THROUGH ITS CHAIRMAN, TAG ALSO ADVISES THE IDCOR STEERING ! COMMITTEE OF TECHNICAL PROGRAM DEVELOPMENTS. TAG IS CHAIRED BY DR. MILES LEVERETT AND INCLUDES INDUSTRY l GENERALISTS, KNOWLEDGEABLE OF OVERALL DEGRADED CORE ISSUES, AND TECHNICAL EXPERTS, FAMILIAR WITH THE SPECIFIC TECHNICAL ASPECTS _ 0F THE IDCOR PROGRAM. MEMBERSHIP IN TAG IS NOT LIMITED TO MEM-BERS OF THE IDCOR POLICY GROUP. i .
~75 THE SCOPE OF TAG ACTIVITIES INCLUDES ADVISING ON ALL TECHNICAL ~
ASPECTS OF THE IDCOR PROGRAM, INCLUDING IDCOR TEC.HNICAL STR ATEGY. TAG'S SCOPE IS NOT PRIMARILY ONE OF IN-DEPTH PEER REVIEW Oi INDI-VIDUAL PIECES OF TECHNICAL WORK DONE IN THE PROGRAM, BUT SUCH WORK MAY BE INCLUDED UNDER SPECIAL CIRCUMSTANCES WHEN RE0 VESTED BY THE PROGRAM MANAGER,0R THE IDCOR STEERING GROUP. IDCOR LEGAL ADVISOR THE IDCOR LEGAL ADVISOR IS RESPONSIBLE FOR PROVIDING LEGAL SUP-PORT ON IDCOR ACTIVITIES. THE LEGAL ADVISOR PROVIDES SUPPORT T0, TAKES THE LEAD FROM, AND REPORTS TO THE PROGRAM MANAGER. IDCOR LEGAL SUPPORT IS COMPRISED 0F A PRIMARY LEGAL FIRM, CONSISTING 0F A LEAD LAWYER AND SUPPORTING LAWYERS.- THE LEGAL ADVISOR ALSO ADVISES THE IDCOR STEERING GROUP OF PROGRAM DEVELOPMENTS OF A LEGAL NATURE. SENIOR CONSULTANTS THE SENIOR CONSULTANTS ARE HIGH LEVEL CONSULTANTS FROM OUTSIDE THE IMMEDIATE INDUSTRY, WHO, AT THE SPECIFIC REQUEST OF THE IDCOR PROGRAM MANAGER, CAN ASSIST IN ASSURING THAT THE PROGRAM IS TECH-NICALLY SOUND, PROPERLY SCOPED, AND INTERNALLY CONSISTENT. THEY WILL ASSIST IN DEVELOPING A SOUND, INTEGRATED POSITION FOR THE INDUSTRY. TECHNICAL CONTRACTORS THE TECHNICAL CONTRACTORS WILL PERFORM THE WORK SPECIFIED IN THE IDCOR TECHNICAL SCOPE STATEMENTS. THE CONTRACTORS WILL BE CHOSEN l _. . -. _. _.
BY THE PROGRAM MANAGER BASED ON DEMONSTRATED EXPERTISE AND PERFORMANCE. THE PROGRAM MANAGER WILL RECOMMEND CONTRACTORS TO
. THE STEERING GROUP FOR APPROVAL, WILL NEGOTI ATE ALL CONTRACTS, AND APPROVE ALL INVOICES PRIOR TO PAYMENT. CONTRACTS WILL RE ISSUED BETWEEN THE CONTRACTOR AND AIF. THE AIF WILL DISPENSE FUNDING TO THE TECHNICAL CONTRACTORS BASED ON THE PROGRAM MANAGER'S APPROVAL.
TECHNICAL PROGRAM THE OVERALL PROGRAM PLAN CAN BE CONSIDERED AS A SERIES CF FOUR MAJOR WORK AREAS OR GROUPS OF TECHNICAL TASKS. THE WORK AREAS ARE SHOWN IN FIGURE 2 AND ARE LISTED BELOW: e SAFETY GOAL AND GROUND RULES
- CURRENT STATUS AND ACCIDENT PREVENTION e CONTAINMENT EFFECTIVENESS e ACCIDENT CONTROL AND EVALUATION SAFETY GOAL AND GROUND RULES TASK 1, SAFETY GOAL, SERVES AS A CRITERION BY WHICH RESULTS CAN BE ASSESSED FOR RECOMMENDED IMPLEMENTATION. THE SAFETY GOAL FRAMEWORK MUST BE FORMULATED SUCH THAT IT CAN BE APPLIED IN THE CONTEXT OF THE DEGRADED CORE RULEMAKING AS WELL AS IN THE BROADER AND LONGER-TERM DECISION MAKING RELATIVE TO SAFETY. TASK 2 SETS THE GROUND RULES FOR THE ANALYSES. -
4 s w- ----wre-,-- ,r,-- --~ r , ww ,, ,.--p.- - - - , , -e,e----
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CURRENT ST ATUS & CORE D AMAGE PREVENTION x -
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Figure 2 TECHNICAL APPROACH TO DEGRADED CORE PROGRAM 1
?
l CURRENT-STATUS AND ACCIDENT PREVENTION ~ ,__ THE 0BJECTIVES OF NUCLEAR POWER PLANT SAFETY DESIGN ARE T0 (1) PREVENT ACCIDENTS, IF'THEY DO OCCUR, (2) TERMINATE THEM, AND (3) MITIGATE THEIR CONSEQUENCES. CURRENTLY, THE OBJECTIVES OF SAFE DESIGN AND OPERATION ARE ACHIEVED BY A COMBINATION OF THE
" DEFENSE-IN-DEPTH" APPROACH, SAFETY ANALYSES, MULTIPLE REVIEWS, 00ALITY ASSURANCE, SUR'VEILLANCE, TRAINING, AND INSPECTION. THESE FEATURES, WHEN ADEOUATELY IMPLEMENTED, GREATLY REDUCE THE PROB A-BILITY OF ACCIDENTS CHARACTERIZED BY SEVERE CORE DAMAGE.'
, ALTHOUGH THE SEQUENCE OF EVENTS AT- THREE MILE ISLAND WERE OUTSIDE THE CURRENT DESIGN BASIS CONCEPT,. THE " DEFENSE-IN-DEPTH" APPROACH $ WORKED EFFECTIVELY IN PROTECTING THE PUBLIC HEALTH AND SAFETY. THEREFORE, THE PRIMARY IMPETUS FOR NUCLEAR POWER PLANT SAFETY ! DESIGN, IN LIGHT OF THE THREE MILE ISLAND ACCIDENT AND AS SUP-PORTED BY THE CONCLUSIONS OF THE MANY INVESTIGATIONS CONDUCTED IN ITS AFTERMATH, SHOULD BE THE DEVELOPMENT AND IMPLEMENTATION OF MEANS TO PREVENT ACCIDENTS LEADING TO SEVERE CORE DAMAGE.
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THE TASKS IN THIS SECOND AREA FOCUS ON THE SYSTEMATIC ANALYSIS OF NUCLEAR PLANTS TO ESTABLISH THE CURRENT STATUS OF RISK AND THE l PREVENTION OF ACCIDENTS LEADING TO CORE DAMAGE. ! THE EMPHASIS IS ON EVALUATING THE . ABILITY TO COPE WITH REALISTIC ACCIDENT PHENOMENA AND REVIEWING POTENTIAL IMPROVEMENT IN DESIGN. IN ORDER TO ESTABLISH CURRENT STATUS, PRESENTLY AVAILABLE INFOR-l - MATION ON RISK ASSESSMENTS WILL BE REVIEWED AND SUMMARIZED IN THE l
- 1. ,
- l i
~ -
SECOND WORK AREA. THIS AREA ENCOMPASSES TASK 3, SELECTION OF DOMINANT' SEQUENCES, AND TASK 4, SELECTION OF PHENOMEN0 LOGICAL
, SEQUENCES. THESE TASKS FOCUS ON THE SYSTEMATIC ANALYSES OF I
NUCLEAR POWER PLANTS UTILIZING PROBABILISTIC RISK ASSESSMENT TECHNIQUES TO ANALYZE SEQUENCES. ; l ALSO INCLUDED IN THIS ' AREA ARE TASK 5, EFFECTS OF HUMAN ERROR AND TASK 6, EQUIPMENT RISK CONTRIBUTION, WHICH PROVIDE A BASIS FOR TASK 7, THE ESTABLISHMENT OF A BASELINE RISK PROFILE FOR CURRENT
~
r PLANTS. THE DISCUSSION OF IMPROVEMENTS MADE SINCE THE THREE MILE ISLAND ACCIDENT, TASK 8, AND A REVIEW OF POTENTI AL IMPROVEMENTS IN DESIGN AND OPERATION WHICH PRE' VENT CORE DAMAGE, TASK 9, COMPLETE THIS SECOND AREA 0F THE PROGRAM. TASK 9 IS DIRECTED TOWARD A REVIEW OF POTENTI AL IMPROVEMENTS IN DESIGN AND OPERATION WHICH PREVENT CORE OAMAGE. A COMPARISON OF THE BASELINE RISK PREVIOUSLY ESTABLISHED AND THE SAFETY GOAL WILL INDICATE IF THESE l l IMPROVEMENTS ARE NECESSARY OR DESIRABLE. > I I L CONTAINMENT EFFECTIVENESS THE PREVENTIVE MEASURES NOW TAKEN OR PLANNED DECREASE THE PROBA-BILITY OF A FULLY DEGRADED CORE ACCIDENT FROM OCCURRING. l i HOWEVER, THE CAPABILITY OF PLANTS TO COPE WITH SOME DEGREE OF I DEGRADED CORE CONDITIONS MUST BE EXAMINED TO PROVIDE ANOTHER LEVEL OF SAFETY. WHAT IS NECESSARY IS A SYSTEMATIC ANALYSIS OF PLANTS TO IDENTIFY THE MOST LIKELY ACCIDENT SEQUENCES LEADING TO _ i DEGRADED CORE CONDITIONS. IN ADDITION, SUCH A SYSTEMATIC ANALY-
$15 SHOULD INCLUDE THE CHARACTERIZATION OF THE RELEVANT PHENOMENA
(HYDROGEN GENERATION, FISSION PRODUCT BEHAVIOR, CORE C00 LABILITY, ETC.) ASSOCIATED WITH THE SET OF SEQUENCES IDENTIFIED IN TASKS 3 AND 4. THE CAPABILITY OF THE PLANT TO COPE WITH A SUFFICIENT RANGE OF DEGRADED CORE ACCIDENTS CAN THEN BE ASSESSED BY CON-SIDERING IN A REALISTIC MANNER THE ACUTAL DESIGN CAPABILITY OF THE PLANT TO COPE WITH THESE RELEVANT PHENOMENA. THE . THIRD MAJOR WORK AREA IS DIRECTED TOWARD AN EVALUATION OF CONTAINMENT EFFECTIVENESS. YHIS EFFORT WILL INCLUDE THE ASSESSMENT OF EXPERIMENTAL DATA, DEVELOPMENT OF PHENOMEN0 LOGICAL MODELS AND OF AN INTEGRATED CONTAINMENT MODEL, AND THE EVALUATION OF REACTOR AND PLANT SYSTEMS. TASKS 11, 12, 14 AND 15 PROVIDE AN
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UNDERSTANDING OF KEY PHENOMENA 0F FISSION PRODUCT BEHAVIOR, HYDROGEN GENERATION AND BURN, CORE DEBRIS BEHAVIOR AND STEAM OVERPRESSURE. THE ANALYTICAL MODEL TO BE DEVELOPED IN TASK 16
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WILL INTEGRATE THE RESULTS OF THE RHEN0MENA TASKS AND BE USED TO
' PERFORM CONTAINMENT ANALYSIS, TASK 23. THESE TASKS PROVIDE INI-TIAL AND DETAILED DESCRIPTIONS OF ACCIDENT ENVIRONMENTS FOR THE REACTOR AND PLANT EQUIPMENT AND SYSTEMS. TASK 10 ESTABLISHES THE
. CAPABILITY OF THE CONTAINMENT TO WITHSTAND THE LOADS DUE TO DEGRADED CORE ACCIDENTS. TASK 18 COMPUTES THE ATMOSPHERIC AND GEOLOGIC DISPERSAL OF FISSION PRODUCTS. IMPROVEMENT IN PERCEIVED RISK WILL BE ADDRESSED IN TASK 21. IN ADDITION, CONSIDERATION WILL BE GIVEN TO HYDROGEN BURN CONTROL SYSTEMS, TASK 13, AND ,"
EQUIPMENT SURVIVABILITY, TASK 17.
4 4 o _r__ THE NUCLEAR REGULATORY COMMISSION HAS INITIATED PROGRAMS TO DETERMINE THE POTENTI AL BENEFITS AND HAZARDS ASSOCIATEU WI.T.H ALTERNATIVE MITIGATION SYSTEMS. C0'NSIDERATION OF ALTERNATIVE CONTAINMENTS, TASK 19, AND CORE RETENTION DEVICES, TASK 20, WILL BE PERFORMED TO THE EXTENT NECESSARY. ACCIDENT CONTROL AND EVALUATION THE LAST WORK AREA DEALS WITH ACCIDENT CONTROL AND EVALUATION. THE IMPACT OF OPERATIONAL ASPECTS OF ACCIDENT MANAGEMENT, TASK 24, AND THE RECOGNITION OF SAFE STABLE STATES, TASK 22, WILL PLAY AN ESSENTIAL ROLE IN THE RESPONSE TO A SEVERE ACCIDENT. THE RESULTS OF STUDIES OF DESIGN AND OPERATION FEATURES OR ACTIONS WILL ALLOW A DETERMINATION OF POTENTIAL RISK REDUCTION. THIS INFORMATION ALONG WITH COST, FEASIBILITY AND PRACTICABILITY WILL BE CONSIDERED IN TASK 21, RISK REDUCTION, BASED ON THE EVALUATION OF PRESENT AND ALTERNATIVE APPROACHES AND COMPARISON WITH THE SAFETY GOAL, THE RESULTS OF THE OVERALL PROGRAM WILL BE INTEGRATED INTO A TECHNICAL POSITION FOR RULEMAKING. AS YOU CAN SEE, WE HAVE DEVELOPED A BALANCED PROGRAM WHICH FOCUSES ON PREVENTION AND ADDRESSES MITIGATION AS WELL. WE HAVE , COMPLETED AN INTENSE 3 MONTH PLANNING PHASE AND ARE JUST BEGINNING TO LET CONTRACTS TO GET THE TECHNICAL WORK DONE. LAST - WEEK WE SELECTED OUR LEGAL ADIVSOR--GEORGE EDGAR OF MORGAN, LEWIS
& BOCKIUS--T0 HELP US THROUGH THE PROCESS.
4 - - - _ . - - - , , _ . - - - _ _ - . - . e- - , - - .___._y_u.-yc -- . . - - - _ - - - - _ ,- - - - - - - - - - - , - .
IN OUR ATTEMPT TO ADDRESS AND RESOLVE THE MAJOR ISSUES SURROUNDING THE DEGRADED CORE QUESTION, WE EXPECT MANY TO THOSE OF YOU IlHO _ CRITICS--BOTH HELPFUL AND NOT SO HELPFUL. WOULD HILP, I ENCOURAGE YOU TO GIVE ME YOUR COMMENTS, VIEWS, AND SUGGESTIONS--WE NEED THEM! TO THOSE WHO WOULD HAMPER OUR EFFORTS, I RECALL THE00,0RE R00SEVEl.T:
"IT IS NOT THE CRITIC WHO COUNTS; NOR THE MAN WHO POINTS OUT HOW THE STRONC MAN STUMBLED, OR WHETHER THE DOER OF DEEDS COULD HAVE DONE THEM BETTER. THE CREDIT BELONGS TO THE MAN WHO IS ACTUALLY IN THE ARENA, WHOSE FACE IS MARRED BY DUST AND SWEAT AND BLOOD; WHO STRIVES VALIANTLY; WHO ERRS AND COMES SHORT AGAIN AND AGAIN; WHO KNOWS THE GREATER ENTHUSIAMS, THE GREAT DEV0TIONS: ,
WHO SPENDS HIMSELF IN A WORTHY CAUSE: WHO AT BEST, KNOWS IN THE END TRIUMPH OF HIGH ACHIEVEMENT, AND WHO, AT THE WORST, IF HE FAILS, AT LEAST FAILS WHILE DARING GREATELY, 50 THAT HIS PLACE SHALL NEVER BE WITH THOSE TIMID SOULS WHO KNEW NEITHER VICTORY NOR DEFEAT." (T. ROOSEVELT) e e G 40 k
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UNITED STATES
. 8 't, NUCLEAR REGULATORY COMMISSION g E WASHINGTON, D. C. 20555
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..... NOV 2 61984 ~ .
MEMORANDUM FOR: Distribution - FROM: Themis P. Speis, Director, Division of Safety Technology Office of Nuclear Reactor Regulation ._ Ffobert M. Bernero, Director, Division of Systems Integration Office of Nuclear Reactor Regulation
SUBJECT:
NRC/IDCOR MEETING ON BASELINE RISK PROFILE AND OPERATOR PROCEDURES FOR ACCIDENT MITIGATION Representatives of IDCOR will meet with the staff and their contractors on December 13-14, 1984, to discuss IDCOR methods and results for baseline risk profile and operator procedures for accident mitigation. The agenda for the meeting is given in Enclosure 1. This meeting is the fifth in a series of technical exchanges aimed at understanding the bases for the IDCOR methodology and comparing the NRC and IDCOR technical positions. The earlier meetings concentrated on phenomenological models for severe accident progression, containment loading and radiological releases. Although those meetings pro-vided a great deal of insight into the IDCOR models, it was clear that a true understanding of the IDCOR methodology requires an examination of how the models are integrated and applied to calculations of severe accident sequences at existing plants. By November 27, 1984, the Technology for Energy Corporation (TEC) will provice all meeting participants with a set of IDCOR documents related to the meeting topics. Enclosure 2 contains a list of documents which will be forwarded and a list of previously provided documents which will be referenced at the meeting. The meeting will take place at the Ramada Inn, 1251 West Montgomery Avenue, Rockville, Maryland. A number of rooms have been set aside; call the Inn directly (301-424-4940) for reservations, identifying that you are a partici-pant in the NRC/IDCOR meeting. Reservations should be made by December 3, 1984. Questions concerning the meeting agenda should be addressed to J. Mitchell (301-492-9402). ._ I y p,- g O CT - J ~Q l u p skN# U Themis P. Speis, Director Division of Safety Technology g Office of Nuclear Reactor Regulation I Robert M. Bernero, Director Division of Systems Integration - l Office of Nuclear Reactor Regulation -
Enclosures:
- 1. Meeting Agenda l
- 2. List of Relevant IDCOR Documents
ENCLOSURE 1 s--- AGENDA NRC/IDCOR MEETING ON BASELINE RISK PROFILE AND OPERATOR PROCEDURES DECEMBER 13-14, 1984 Thursday, December 13 10:00 - 10:30 - Coffee and Informal Discussion 10:30 - 10:45 Welcome, Ground Rules, Schedule (NRC)
' Introduction (IDCOR) '
10:45 - 11:45 - Issues Raised at Prior Meetings (IDCOR) 11:45 - 12:30 - Lunch 12:30 - 1:00 - Introduction to Risk Methodology (IDCOR) 1:00 - 1:45 - Sequences and Consequences (IDCOR) 1:45 - 2:00 - Break 2:00 - 2:30 - Baseline Risk Profile - Peach Bottom (IDCOR) 2:30 - 3:00 - Baseline Risk Profile - Sequoyah (IDCOR) 3:00 - 3:30 - Baseline Risk Profile - Grand Gulf (IDCOR) 3:30 - 4:00 - Baseline Risk Profile - Zion (IDCOR) l l _, . - . - , . , - - - - . - - - . . . -._.-,,--,,,,-,,n. , - . , . . . - . .
o Friday, December 14, 1984 8:00 - 8:30 - Prevention / Mitigation Study (IDCOR) - 8:30 - 9:15 - Core Damage Prevention (IDCOR).
- 1. Potential Use of Existing Equipment
- 2. Potential Benefit of Modifications 9:15 - 10:00 -
Operator Action Review for Severe Accidents (IDCOR) 10:00 - 10:15 - Break , 10:15 - 11:45 - Confirmation of Sequence Termination by Operator Action (IDCOR) 11:45 - 12:45 - Lunch , luation, ity 2:00 - 2:30 - Break and Caucus -
- 1. Risk Profile l
- 2. Operator Actions l 3. Issues Raised at Prior Meetings 3:15 - 4:00 -
Summary and Comments (IDCOR) 4:00 - 4:30 - Parting Remarks (NRC, IDCOR) er _. ..--, , , . -__ y7 _ .
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o s ENCLOSURE 2 LIST OF IDCOR DOCUMENTS RELEVANT TO THE DECEMBER 13-14, 1984 NRC/IDCOR MEETING I. Documents to be provided to Participants by TEC 9.1 Core Damage Prevention
-21.1 Risk Reduction Potential 18.1 Evaluate Atmospheric and Liquid Pathway Dose 24.4 Operator Response to Severe Accidents 23.1 Integrated Containment Analysis Peach Bottom Sequoyah
~
Grand Gulf Zion II. Documents Previously Forwarded to NRC Which Will Be Referenced at the Meeting 11.1 15.2a 11.4 15.2b 11.5 15.3 11.3 16.1 11.6 16.la 14.la 16.2 14.lb 16.3 15.la 23.1 15.lb 9 _erwre-- --n--w-w,w-
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NRC/NRR NRC/RES ACRS - H. R. Denton R. Minogue R. Tripathi E. Case 0. Bassett S. Seth R. Bernero G. Arlotto R. Cushman R. W. Houston R. Curtis G. Quittschreiber - L. G. Hulman T. Walker R. Vollmer G. Marino Z. Rosztoczy J. Glynn TEC B. Sheron W. Morrison J. Rosenthal J. Han A. Buhl C. Tinkler , R.. Wright M. Fontana R. Palla &M7 Cunningham E. Fuller W. Lyon M. Silberberg J. Carter, III A. El-Bassioni T. Lee H. Mitchell R. Barrett C. Peabody S. Asselin T. Speis P. Niyogi K. Meyer J. Mitchell T. Eng C. Allen B. Aggarwal EPRI P. Easley R. Meyer S. Acharya B. Burson M. Everett J. Read J. Larkins B. R. Sehgal F. Akstulewicz J. Telford D. Squarer W. Pasedag P. Baranowsky R. Vogel F. Gillespie C. Fuller Other NRC R. VanHouten J. Martin Battelle Columbus J. Conran, DEDROGR L. Chan M. Taylor, DEDROGR L. Soffer P. Cybulskis J. Austin, OCM J. Murphy R. Denning H. Thompson, DHFS M. Ernst J. Gieseke D. Ziemann, DHFS J. Jenkins P. Owczarski C. Ryder K. Wiegardner DOE FAI EG&G Idaho F. Witmer H. Fauske S. Behling R. Henry J. Broughton J. Gabor R. Gottula AIF M. Kenton E. Krantz J. Siegel NRC PDR e l i
2-Sandia National Lab Other Affiliations - D. Dahlgren R. Breeding, El M. Berman D. Moore, EI S. Thompson J. Young - R. Cole M. Lloyd, Middle South Services J. McGlaun D. Paddleford, W J. Hickman L. A. Wooten, W-K. Bergeron P. Nakayama, Jiycor D. Kunsman L. Azarello, Duke Power D. Aldrich W. Mims, TVA J. Sprung M. Cosella, Coned J. Walker J. Meincke, CPC0 J. Griesmeyer W. Iyer, NYPA F. Harper J. Davis, NYPA D. Powers A. Marie, PECO V. Behr G. Krueger, PECO J. Linebarger H. R. Diederich, PECO S. Dingman R. Smith, Scandpower A. Camp J. Engstrom, OKG AB/ Sweden A. Benjamin J. Liljenzin, CTH/ Sweden S. Webb L. Rib, LNR Associates C. Leigh J. Metcalf, Stone & Webster A. Peterson C. Ader, Stone & Webster D. Williams M. Corradini, University of Wisconsin P. Mast I. Spiewak, American Physical Society E. Haskin S. Niemczyk, UCS R. Habert, UCS Oak Ridge National Lab T. Theofanous, Purdue University J. Kelly, University of Virginia S. Hodge K. Araj, Harvard University I. Catton, UCLA . Brookhaven National Lab R. Seale, University of Arizona S. Beal, SC&A W. T. Pratt R. Paccione, Long Island Lighting Co. M. Khatib-Rahbar K. Holtzclaw, GE R. Newton R. Smith, NuCon Corporation T. Ginsberg A. Pressesky, AWS G. Greene M. Ryan, Inside NRC R. Jaung P. O'Reilly, NUS Wen-Shi Yu P. Fulford, NUS H. Ludewig G. Kaiser, NUS S. Blazo, Bechtel e
T- p Atenede ladustried Forene, ins. Change required by 71o1 Wisconsin Avenue P**W F*988883**** Westungton, D.C. 20014 Telephone:(3o1) 654-9260 7101 Wisconsm Avenue Sotheeds. Maryland 20614 TWX 7108249602 ATOMIC FOR oC
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l - p John n. si.e. speew ucen a9 Pmtects Manager August 16, 1984 TO: Distribution FROM: John R. Siegel
SUBJECT:
Additional Transmittal of Advance Material for IDCOR/NRC Meeting 38. Enclosed is Draft Technical Report: FAI Aerosol Correlation for the IDCOR/NRC Technical Exchange Meeting 38--Integrated Plant Analysis--which will be h' eld August 28-29, 1984, at the Holiday Inn Crowne Plaza in Rockville, Maryland.
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Copyrighted Document Addressed Under FOIA qe; ti For hard copy, )( refer to PDR Folder:FOIA - D $ l l
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$. DISTRIBUTION: - Yh - 31 , H. R. Denton, NRR G. Marino "L a ' E. Case, NRR J. Telford % R. Bernero R. Houston R. Meyer D. Muller R. Wright T. Speis ' J. Han B. Sheron J. Larkins W. Butler C. Ryder
, J. Hulman C. Peabody W. Pasedag L. Chan
, Z. Rosztoczy R. Curtis R. Barrett B. Burson r A. Thadani T. Lee l J. Mitchell ACRS i J. Read PDR 1 J. Rosenthal D. Garner, OCM -; R. Minogue, RES J. Meyer }, . i 0. Bassett J. Austin p;] D. Ross S. Chesnut M. Silberberg Tech. Ass't, Comm. Zech G. Arlotto S. Niemczyk, Union of Concerned Scientists m 2 B/sa}}