ML20148S639
| ML20148S639 | |
| Person / Time | |
|---|---|
| Issue date: | 11/21/1978 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1572, NUDOCS 7812040016 | |
| Download: ML20148S639 (133) | |
Text
.h1UTh$h ACRS SUBCOMMITTEE MEETING ON ADVANCED REACTORS WASHINGTON, DC
/7y
- - AUGUST 11 , 1978 8 Cit.5-/S 7 N .
?M ///vl77 i The ACRS Subcommittee on Advanced Reactors held an open meeting on August 11, 1978 in Room 1046, 1717 H St., NW, Washington, DC. The purpose of this meet-ing was to review matters related to the NRC-sponsored research on the safety of the advanced reactor design. Notice of this meeting was published in the l Federal Register on Thursday, July 27, 1978. Copies of this notice is in- g cluded as Attachment A. A list of attendees for this meeting is included as Attachment B, and the schedule for the meeting is included as Attachment C. .
Selected portions of the meeting handouts are included as Attachment D. A j
?
complete set of the handouts has been included in the ACRS Files. No written .--
2 statements or requests for time to give oral statements were received from members of the public. The meeting was attended by Dr. M. Carbon, Subcommittec I_,
Chairman; Dr. W. Kerr, Subcommittee member; Dr. J. C. Mark, Subcommittee i member; Dr. P. Shewmon, Subcommittee member; Dr. R. Savio, ACRS Staff; d and the ACRS consultants, Dr. R. Nicholson and Dr. R. Seale. The meeting .:
was opened at 8:30 a.m. with a short executive session during which
- - ,- j.
Dr. Carbon briefly summarized the schedule and the goals for the day's ((-
meeting. The meeting was held entirely in open session and was adjourned i; at 6:00 p.m. on this day. ,
,. .[
I-INTRODUCTION _- T. Murley and C. Kelber, NRC/RSR #G Dr. Murley gave a brief introduction. He indicated that he believes that ,
HRC's advanced reactor safety research program is a very productive program within the limits of the present funding and current uncertainty in the ;. ,,
~
national policy on advanced reactors. .Dr. Kelber summarized the NRC advanced reactor program with emphasis on the accomplishments of the program during '. . E the past year and the changes that had occurred since the ACRS last reviewed _
these pronrams. A summary of th,e accomplishments for FY 78 is given on pages 1-4 of Attachment D. A major accomplishment within the last year has been the release of the SIMMER 11 and SSC codes. Dr. Kelber indicated that the SIMMER 11 Code has all the basic capabilities needed to analyze the vessel ,- ,
disruptive accident problem. He noted that verification of the validity of .
the models used in the code had not yet been completed. The SSC is designed 7812040Olg .
7812010 .;
.--_-___x ____? j
- 4DVAliCED REACTORS 8/11/78 ,
for the analysis of the pipe break and the loss of forced circulation -
accidents. .
I i
The COMMIX-1 Code, a three-dimensional transients singic phase thermal code, j has been documented and has been released. This code has been tested against j the small-scale FFTF plenum experiments and has been found to give good -)
'}
~
agrecment with the experimental results. .
'3 The EPIC Code has been incorporated into the SAS-3D Code and replaces the j VEllVS portion of the SAS Code. This is an improved treatment of core dis- .$
ruptive accident and predicts a marked reduction (about 50%) in the calcu- [-
lated work potential available to do damage.to the reactor system. The ]
larger work potential calculated with the vel 10S Code was the result of .J unrealistic initial conditions used in the VEltVS calculation. -j yi The llRC work on advanced reactor test facilities will be terminated on '*j October 1st in response.to instructions from the Office of Management and -;
Budget. .Ik
.s The acrosol release and transport experiments formed in the HSPP vessel have :j been completed and have been checked against the llAARM and AEROSIM Codes. -5 Test data from the Containment System Test Facility (CSTF) obtained under _(5 DOE sponsorship has been used for large-scale verification tests of the ~~3 HAARM Code. The acrosol release and transport program is nearing completion. .,( -
ARSR hopes to be abic to do enough additional experiments to test their '. _]
analytical models against a wide range of conditions, but budget cutbacks ,-j ,
may limit this activity.
. y
{
W.T The FAST test facility has been completed and has been checked out and is .a now in operation. This vessel is essentially a 1/10 scale representative -f of reactors, such as the CRBR and will be used for studying problems -
of vapor bubble transport. It'is expected that these experiments will 's be a substantial contributor to the SIMMER verification.
e
- '" "" N4 +pp es. gypgr e.gre,,,
,. p.p., ,,,, , , ,_
ADVAf1CED REACTORS 8/11/78 A number of single fuel pin burst energetics experiments were completed in -
the ACpR prior to its shutdown for upgrading. Twelve tests with fresh oxide fuel and three with fresh carbide fuel were completed. All tests showed a low conversion of thermal to mechanical energy and that the initial pres-sure pulse decreases rapidly. In the case of the oxide fuel, the pressures were consistent with the fuel vapor pressurization alone. In the case of ..
the carbide tests, the thermal to mechanical energy conversion was a factor :
of .10 higher than the oxide tests, and it is believed that these pressure .j increases must come partly from fuel-coolant interactions. 4 T
A new model of fission gas release has been developed at Sandia and is -j believed to represent a significant improvement. 2
~
i
~1 The analysis of the debris bed experiments performed last year have been f._t.
compl eted . The results show that such beds operate stably with local sodium, Q dryout for the range of test times (to one hour), and that the local dryout
- fl does not appear to be a real coolability limit. Local dryout leads to a : ~;f rate of temperature rise that is slow enough to assure that there would be " .~2 significant delays between the onslaught of local dryout and debris bed 5;3 melting. .g;
. :;i M
Several proprietary codes from the General Atomics Company have been reviewed f in depth and a research information letter summarizing these reviews is .j scheduled for release during the next fiscal year. T
- i i The DASH, LARC-II, and SUVIUS Codes which were developed for NRC are now '.'i currently in use at LASL. These codes treat fission product diffusion ., j, and circulation within the primary system. The core seismic eveluation j code, OSCll, capabilities have been extended to three dimensions and the
. . {.:
testing of this code has begun. The liONSAP-C Code has been used to check -
the response of pre-test vessels and the predictions of this code have . .l .
been compared with model tests. The code appears to underestimate the ]
ultimate load carrying capability of these concrete vessels. Work in .
this area is continuing. ,
? i
_ p_f
8/11/78 ADVANCED REACTORS ,4-The prototype helium afterglow moisture monitor has been successfully tested for nearly a year at the Fort St. Vrain nu'cicar plant.
Dr. Kelber discussed some of the technical problems encountered during this year. The SIMMER Code predictions of channel blockage are not consistent with experiments. It is hoped that this problem will be .
resolved within the next six months and will not require major code revisions. The heat transfer correlation void in the SSC Code is not properly predicting the heat transfer between fuel and sodium in the turbulent and laminar flow regimes. It is believed that this is a numerical problem and can be resolved without any major change in the code structure.
The Monte Carlo calculations used to establish a benchmark for the neutronics verification of the SIMMER Code have been progressing more slowly and yiciding much less information than was initially hoped for. The SIMMER neutronics models have adequately predicted'the experimental measurements and show a systenatic discrepancy with diffusion theory calculations.
Dr. Kelber indicated that a Five Year Plan had been issued but it appears that the icvel of funding requested.will not be granted.
Dr. Kelber noted that the funding reductions in Fiscal Year 1979 resulted ~
~
in program cutbacks and a lower level of morale. Some loss of key con-tractor personnel is expected. Dr. Kelber indicated that in the FY 80, ~
FY 81, and FY 82 budgets, the highest priority would be given to the -
SIMMER Code development and to the work associated with the code verifi-cation.
The ARSR work in connection with USAP and INFCE has been largely confined to the review of papers, particularly within the areas of alternate systems.
G I
ADVANCED REACTORS ' 8/11/78 ANALYTICAL ADVANCED REACTOR SAFETY RESEARCH BRANCH - R. Curtis, RSR Dr. Curtis summarized the work being sponsored by the Analytical Advanced Reactor Safety Research Branch. He noted that more detailed ,
discussions of these programs would be given at the September 12th i and 13th meetings at Sandia Laboratories and the September 27th meeting .
i in the District of Columbia. The SIMMER work has been discussed with the ACRS Subcommittee on HCDA on July 27 and 28, 1978'. The work of h the Analytical Advanced Reactor Safety Research Branch presently i emphasizes the development of computer codes and the verification f testing of these codes. Major changes in the direction of these i programs within the last year have been: (1) the reduction of effort ~_k -
on the priting of the SIMMER Code and an increased emphasis on the ..
application and testing of the SIMMER Code, and (2) the phasing 3 out of the safety test facility work related to SAREF. .N
-i Accomplishments during FY 78 and the future plans for the Analytical .3 Advanced Reactor Safety Research Branch are summarized in pages 5-17 7.} ~
Some of the highlights of the accomplishments are: .;
of Attachment D.
(1) completion of the physics studies and partial completion of the : 3 whole core accident series on the conceptual 2000 MWe LMFBR; (2) the ,.
, C-completion of a review of U.S. and foreign containment analysis '~ T _*
methods; (3) the completion of event trees for the impact of various engineered systems in-core phenomenology associated with the HCDA and .j '
containment phases of LMFBR analysis; (4) the release of the version ._
of the COMM1X-1 Code (a single phase 3-dimensional thermal hydraulics .
code); (5) the development of the SSC-L Code into the operational phase; (6) completion of the critical experiments for the validation of the . 'I_
~
VIM and SIMMER codes; (7) completion of the SIMMER-II Code; and (8) completion of the ACPR upgrade.
~
l .
t
' ADVA'fiCED REACTORS 8/11/78 _
EPERIliEllTAL FAST REACT 0n SAFETY RESEARCH BRAf1CH - li. Silberberg, RSR lir. Silberberg summarized the work being sponsored within the Experi-
~
mental Fast Reactor Safety Research Branch. fir. Silberberg noted _
l that the ACpR upgrade had been completed and that the pulse and steady-state capabilities of the reactor had been increased by a factor 3.3.
The upgraded ACPR has the capability of testing 7 pin and possibly -:
19 pin test assemblies. A very much improved fuel motion diagnostic system has been evolved from the coded-aperature-imaging of fission gamma rays. This diagnostic capability should provide a u.nique facility for 3 studying advanced experiments in accident energetics and fuel debris !
behavior. The steady-state capability of the upgraded ACPR will allow -9 studies of the coolability limits of debris beds to a decay heat level .
equivalent to 4% of prototypical Lf4FBR power. ;
Prior to the ACpR upgrade, a series of tests,12 with fresh oxide fuel '7 -
5 and 3 with carbido fuel, were performed. The carbide tests, however, ,
showed thermal to mechanical energy conversions of an order of magni 2 -;
tude higher than what the oxide tests showed. The pressures experienced in the oxide fuel tests were attributed to the fuel vapor pressure and the " -
larger pressures experienced in the carvide tests were attributed to fuel- -
coolant interactions. It is intended that in FY 79 tests will be performed -p-with irradiated fuel .
In FY 78, three debris bed coolability experiments were performed in ACPR
'~
along with three small capsule molten fuel expe. %nts. It was concluded - ,' -
from these experiments that local bed dryout was not a bed coolability _
~'
limit. Additional debris bed coolability experiments are planned for FY 79. The impact of the expected budget constraints will be to reduce _
the number of debris bed experiments and to defer the large molten fuel capsule tests. .
The core melt program has been redirected in FY 78 as a result of liRR -
request to emphasize the testing of basalt concrete in support of the FFTF review.
g .:_ N ,,
/IDVNICED REACTORS o 7- 8/11/78 The accomplisignents in the molten core technology program for FY 78 included the completion of four large-scale (200 kg) stain 1 css steel melt / concrete interaction tests. The design of the large induction furnace melt facility (500 kg U0 ) design has been completed and con-2 struction has been initiated. Initial trial melts are scheduled for February of 1979 and the facility is expected to be fullycoperational by July 1979. The impact of budget constraints within this program will be to reduce the number of experiments to be perf6cmed and to defer the combined fuel, steel and s6dium interaction experiments.
Within the sodium containment structural integrity programs, the accomplishments for FY 78 have included the completion of three large-scale (up to 200 kg) tests with limestone concrete and the completion of six additional large-scale tests with limestone con-crete to study the effect of sodium pool depth, the effect of precrecked concrete, and the effect of flawed liners. A chemical I
model of the sodium / limestone-concrete reactions has been developed and the separate effects test facility designed, built and bro 0ght to an operational stage. The first test is to be performed in August of 1978.
The flRC acrosol release and transport program is directed towards pro-viding data and verified models for the radiological consequence assersment. In FY '78, a scrics of five sodium aerosol tests have been completed in the !!ucicar Safety Dilot plant (t!SPP) at ORNL following extensive modifications to this facility. The data obtained from the NSpp tests and tests conducted at HEDL in the Containment Systems Test Facility (CSTF) have been used in the verification of the aerosol codes. Experiments have been carried out BCL to determine the sodium acrosol agglomerate properties. Capacitor discharge vaporization tests have been used to establish size distributions of acrosols as a function of energy deposited in the fuel . The vessel to be used in the Fuel -
Acrosol Simulant Test (FAST) has been fabricated and inspected. Tests
ADVAtlCED REACTORS 8- 8/11/78 ,
are scheduled for this facility in FY 79. The impact of budget con-l straints on this work has been to reschedule the flSPP tests into ,
FY 1980, to defer the fission product aerosol experiments, and to -
defer sodium tests in FAST. The object of the research program ]
for structural integrity is to provide sufficient independent i experimental analytical to permit licensing of new types of reactors .'i-with an expected 40-year life. Currently the material property investigations are emphasizing creep / fatigue interaction effects. -
Accomplishments for FY 78 have been a study of biaxial effects ;(.
on crack growth, fatigue studies of 21/4 Cr-1 Mo weld metal steel, arid creep buckling study of pHC elbow. 8 A
EXPERIllENTAL GAS COOLED REACT 0P, RESEARCH BRANCH - R. Schamberger ;l' Mr. Schamberger summarized the activities of the Experimental Gas [
Cooled Reactor Branch. The Gas Cooled Reactor Associates (GCRA), 3 a utility group, has been formed to promote the commercialization ,g of the HGTR. Congress has also expressed interest in funding of .-f IITGR research. The objectives of this research are to provide .T licensing with independently validated methods and data for GCR I-h safety assessment and to identify and evaluate potential signi- .:
ficant contributors to the public risk. The individual tasks ?.$
within the program deal with: (1) fuel and fission products, (2) I.o4 primary coolant, (3) structural response (4) materials properties, -
' 2; (5) instrumentation, (6) accident delineation, and (7) analysis. . -[;
~-s .
Program changes in FY 78 are summarized on page 18 of Attachment D. -2
.-1
. ._.~
Accomplishments in FY 78 are summarized on pages 19-25 of Attach- D ment D. The highlights of these accomplishments are the improve-
'h -
ments in the understanding of fission product transport, the 2:
high temperature performance of graphite, and a development of a -25 fission product transport code. The helium afterglou detector has ', " )
been tested extensively at the Fort St. Vrain reactor. It is expected 3 that future programs will have the increased emphasis on the Fort St.
Vrain reactor licensing issues. .
fd :'
.h-
~.
cr
lA10VT0TELTMlCKCTORS -@- 8/31/78 j
- .j l
llTGR SAFETY RESEARCH AftD DEVELOPMEllT - T. Werner, DOE ,j l
fir. Werner summarized DOE's HTGR safety research and development . programs. .5
.r fir. Herner noted that the failure of General Atomics' initial commercial- ;;;,
ization ondeavor had resulted in reluctance on the part of vendors and :i suppliers to invest further in the llTGR commercialization. A previous ,}
United States utility commitment to the purchase of ten large plants has .$w been cancelled. A new HTGR initiative, however, has recently come into R place with the formation of Gas Cooled Reactor Associates (GCRA) in which . . _
nine utilities have formed a group whose purpose is the development of a ,h commercial HTGR. fir. Werner noted that the United States Government is 9
.u now faced with the issue of whether to support HTGR commercialization with @j national incentives. It would appear at this time that the utilities must _I) take a leadership role and that the supplier base for components must be -7 .-
broadened if HTGRs are to be successfully commercialized in the United y{5l States. ?-T
-G The present thinking on HTGR commercialization cost and rate sharing are s {P]' E summarized on pages 26 and 27 of Attachment D. In accordance with this [d thinking, the utility owners would absorb about 1/2 of the predictable h.N cost of a pilot HTGR and the government and suppliers would share equally , -Ns
-G in the remainder of the cost. It is additionally proposed that the suppliers ; . .~:
would bear about 1/4 of cost overruns and that government and the utility Jy owners would share equally in the remainder. M. u
.MY A sumrary of the world-wide HTGR plant experiments is given on page 28 .
,2$. . . -
of Attachment D. A summary of the DOE funding on HTGR is given on pages ,y.;g 29-31 of Attachment D. The fcnding for FY 77, FY 78, and FY 79 are Tf..d, respectively, $i7.2 million, $32 million, and $31 million. Funding for 'q;'
FY 1980 is still under review. The major DOE contractor will be General k Atomics. ORNL, GCRA, and General Electric will be major contractors '. j in this program. ~, .W '
- - t
". i
.1 .
.Nc6) '
. . ~, . i f
- G e
~. .*.
N* ire ep..es . .m, , ..w,.-w o.... , , . . . , . . , , , , . . . . . . . . . . -
.y ADVAtlCED REACTORS 10 - .8/11/78 i
. ::l d
N t
DEPARTliENT OF ENERDY'S GAS COOLED FAST REACTOR SAFETY PROGRAMS - D. Erb & D.Emon, DOE 17f i.
k Mr. Erb and Mr. Emon summarized the DOE gas cooled fast reactor safety .~Q programs. They noted that their programs strategy centered about a ,3 plan in which both the U.S. and foreign countries share the cost and M
d the benefits of a single design development. A single U.S./FRG design f is planned with the Helium Dreeder Associates managing the program.
N A
Funding for FY 78 under this program is approximately $18 million with ~!R the U.S. contributing $14 million and FRG contributing $4 million. J.
Expansion of this program to around $20 million for FY 79 is planned, y a..
General Atomics is expected to be the principal contractor for the U.S. ' ,*
- .1
- 3 work. A summary of the budget for the individual tasks for FY 78 and FY 79 is given on page 32 of Attachment D. Summaries of the individual j,'
program tasks are given on pages 32a-39 of Attachment D. 5'd U. i INTERNATIONAL ftVCLEAR FUEL CYCLE EVALUATION (INFCE) AND NON-PROLIFERATION ALTERNATE ASSESSMENT PROGRA!1 (NASAP) - J. Leary, DOE
'.;c.y .
]
Mr. Leary sunmarized the organization and work scope of the INFCE and the w?
HASAP. The If1FCE has 52 member countries with technical coordinating :3 committees involving 22 member countries and is split into 8 working groups SM
.o having the following work scopes: Working Group 1, Fuel Availability; .,.4j Working Group 2, Enrichnen ' vailability; Working Group 3, Supply .
/E.) : .
Assurance; Working Group 4, Reprocessing and Recycle; Working Group 5, 'g Fast Dreeder Reactors; Working Group 6, Spent Fuel Management; llorking jQ. . ...
Group 7, Waste Management and Disposal; and Working Group 8, Alternate k;.2
- s ?.
Fuel Cycle Considerations. A more detailed summary of the work scope ' ';5, for each of these working groups is given on pages 40-42 of Attachment D.,
,y -:-
The United Statcs has members on all of these working groups. The working N group memberships are listed on pages 43-45 of Attachment D. , '$--j
, 5.*
.* z.-:,$-
l **
- f' .
)',k. '.
. ' : ?.;
~
e e-e i w+ we . , p e . ag - p e.. . g s e.,-4
- t ,
g/D /73 41-O lir. Leary also summarized the flASAP programs. The purpose of flASAP is to identify nuclear systems with high proliferation resistance, to identify institutional arrangements with increased proliferation resistance, to develop strategies, to implement the most promising schemes, and to provide technical support to If1FCE. The schedule for ifASAP is summarized on page 46 of Attachment D. The ffASAP schedule ;
calls for the identification of promising alternate systems by October of 1978. The key decisions within the flASAP program are summarized on pages 47 and 48 of Attachment D. The work to date :
suggests that institutional controls can be developed to reduce I proliferation vulnerabilities to comparable icvels for once through -2f systems, advanced thermal reactors with recycle, and breeder systems.
Key problems appear to center about domestic and international acceptability, the time scale for implementation, and the effectiveness -
- [
of these. counter measures.
THE DOE LMFBR SAFETY PROGRAM - J. Griffith, DOE lir. Griffith summarized the DOE safety programs. The DOE U4FBR safety
.4 programs are structures about four lines of assurance (LOA). They _3 are: (1) to prevent accidents, (2) to limit core damage if an accident
[
occurs, (3) to control the progression of the postulated accident, and ,. {.
(4) to attenuate the radiological consequences of the postulated accident. -
Budgeting for FY 78 and FY 79 is summarized on page 49 of Attachment D. ~'
The total budget 1.s $67 million for FY 78 and $44 million for FY 79. .
The funding within the individual tasks are summarized on pages 50-64 ,'i of Attachment D. -
. . ~ .
o EP O
e O
/
9 t
~
.ADVAftCED REACTORS 8/11/78 .~. .
~i 1
DOE LMFPR SAFETY RESEARCH AflD DEVELOPMENT - EXPERIMENT PLANNIllG - F. Gavigan, DOE .
[
Hr. Gavigan indicated that their strategy for experiment planning is . :.j centered about establishing goals for a particular line of assurance, -l establishing the requirements for success, induct a extensive planning ,ij and reevaluation program, and conduct testing as would be required. .}
A good deal of emphasis is given to a systematic evaluation of the l-j 6xperiments before they are performed. An example of this process S,
.for LOA 2 is given on pages 65-74 of Attachment D. ,-f
'r.:
TRANSITION PHASE ANALYSIS - H. Alter, DOE ..I~0
. ~.3 Mr. Alter made several comments on the role of large-scale computer M codes and LHFBR safety. He indicated that, along with experiments '...a2 -. 9l and the phenomenological modeling efforts, large-scale codes are '. .),y invaluable in resolving LMFBR safety and licensing issues. Large- ,
scale codes also serve as the essential link in translating under- ".c standing gained from experiments and phenomenological studies to - :. c-ed the study of postulated accidents .in reactors. Large-scale codes M .
<. e can give an appreciation of the range of accidents which can result '
. -3
~
from a possible initiator and an appreciation of the range of conse- ..
Til quences. He indicated that he believed that past the onslaught of 3-fuel disruption, large-scale codes will never be able to identify .f;3, Q 3
a signal limiting scenario and an associated well-defined set y of consequences as being the only possibic results of the postulated ~~ , ; .
core disruptive accident. Sb
- kl The schematic on page 75 of Attachment D summarizes the ANL system of '.',U IC analysis of codes. DOE-sponsored are summarized on page 76 of Attach- 'M.~} .
ment D. The key phenomenon in the various core disruptive phases are
',-y summarized on pages 77 and 78 of Attachment D. .
.* 5*.
l .--2
,.a x.~
* **2
'4
;Y
~
. . f.'C- .
(, .-
... _....,-._--...._.c..-...-.. . - - .
ADVAllCED REACTORS 8/11/78 - A comparison of the transition phase codes, TRAtlSIT-HYDRO, and SIMMER . capabilities is given on pages 79 and 80 of Attachmet D. A description l of the TPAftSIT-II and TRAllSIT-IIYDR0 Codes are given on pages 81-85 of 1 Attachment D. A summary of the features of the USEFUL, VEliUS-II, and d EFFECTS /VEf1VS-III Codes are given on pages 87-94 of Attachment D. The
-}
planned experimental verification of the TPAllSIT-HYDRO and EFFECTS-II/ 4 VEltVS-II Codes are given on pages 95 and 96 of Attachment D. A listing of the Allt structural response codes .(REXCO-HEP, ICEO, ALICE, STRAW / WHAM, __d SADCAT, ICEPEL) is given on page 97 of Attachment D. The planned experi- ff, mental verification of these codes is summarized on pages 98 and 99 of li Attachment D. A summary of the features of these codes is given on $ . . pages 100-108 of Attachment D. __ Mr. Alter also discussed the transition phase codes being developed at '$ i!EDL. These are the FUMO-T thermal hydrodynamic code, and the FUSS h couple space energy neutron thermal dynamics code system. The FUSS- kw system includes the FUMO-T. Code, a neutronics, and the SAS 3-dimensional -ig code. Development guidelines for the FUSS-FUM0-T programs are given .S.
~
on page 109 of Attcchment D. Comparisons between the capability of [3h these codes are summarized on pages 110 and 111 of Attachment D. ,[.$$ CC-Mr. Alter also discussed the APRICOT program. This is a voluntary b?:
.c -
cooperative international program for the comparison validation and y-benchmarking of the computer codes used to analyze the LMFBR structural {'- response to pressure loads from postulated core meltdown accident .7. enerDetics. phase oneMr this program has al'cdy been completed. .t_~
- s .-
In this phare, three simple computer problems were used to test .[g numerical methods of the basic code. Phase two is now underway and D is expected to be complete in January of 1979. In this phase, four 5 cc aputer programs having geometries which are protypical of reactor D . systems and which can be compared to existing experimental data are ,'{, being used to benchmark and verify the existing code systems. The !-E organizations participating in the APRICOT program are listed on 2-pages 112 and 113 of Attacia)ent D. ~~,- l .v: 1 . l
'1.'d .
g s,
t ADVAllCED REACTORS 8/11/78 e The meeting was adjourned at 6:00 p.m. on this day. 110TE: For additional details, a complete transcript of the meeting -l is available in the liRC Public Document Room,1717 H St., i1W. .-; Washington, DC 20555 or from Ace-Federal Reporters, Inc., 4 444 llorth Capitol St., NW, Washington, DC. ,i.
.5
- .=.
..v.3
~~
_s
*O s em
.- L-
..ws ye -
.=*
.3
.9;
-rs *
- te.
cM
- I*h
= v. .
'l$
* ,S.
. . ~
- 45
*e e .. . .+ ,
me3"
- ., O a
. \ .'
- k
,..~_-
e e
,- .s~ -:- - -
I
. '. E
- 3
.'e..
- , e ,
G G*- G i e
#e
~he commlist$'s inve Igatirn trydl. [4110-01] ThgNati:nal Econo Opportunity Advisory C: unci,@n is tutt 3trir.cd ,
t
.r! that 11.S. producers lost substaip. . DEPARTMENT OF JUSTICE by secti 05 of the Commjpttty Scr.
revrnues as th3 result of low ring vices Act advise the Pjesident ang! 7;. t Aneitewst Div1ste,, rea to meet. LTFV prices.of apa. - the Directo f the Cpmmunity Scr. . . e merchandise, and that such osses $UGAR CASES vices Administf tion licy matters 1 A lbuted materially to the educ- s i de Q t ton of g the net operating profi of the mPetitiv* ImPect Stoloment g
- 2. p ducers. Pursuant to the Antitrust',Proce and operation grams under the *.* y S.C. 16, act. . ,
Inz2.tHoon :r INJUR dures arid Penalties the Department Act,16on': of Justice U, July 11. Records s. 11 be ke of all proceed. ., j
;xamin g the question pf whether 1978. filsd an amendment to the Com- ings and spa!! be available for public ~~. :
ra is a Ikelihood of ury to the petitive Impact Statement in U.S. v. Inspectiop'at the office o'Ig the Nation. L the back. Great Western Sugar Company, et al, al Advisbry Council on Ec '1cmic Op- . M* '6
- estic industry aga 3und of th'p present pr lems of the Civil No. C74-2674 SW; U.S/v. Cal (for- portunity. / WA1.TER B. Qtn. CH,
- ustry as o\itlined abole. I have de- nfa and Hatecifen Sugar Company, eg EzecutiveDi (clor, , , y .,,
nined tha the conti$ued importa. al, Civil No.\ C74-2675 RPH; and U.S. / m '
- v. Utah. Idaho Sugar Codipany et cL, IFR Doc.78 20838,Fued 7-26-78; 8:qam3 a and sale o stainles/ steel pipe and ,
' .7[ r..I >a from Ja an at /less than fair Civil No. C74-2676 o SC./The f
amend. :,. ~.r.y
.us is likely lo inj e the domestic mentInisthe ascompetitive foll. ws: impact statement. [7590-01] '
E.a
.inless steel pipe tube industry.
section II Descriptioniof Defendants. . >.1 1 NUCLEAR REGULATORY .;:,a
.:ost purchase o welded stainless page 4, lines 7-a, whic)1 read **C&H is el pipa and tu e r ported that price the largest canet sugar refiner in the - ,
COMMISSION g ,; their prima (y consideration in United States with its principal,refin- ADVISORY COMMITTEE ON REACTOR SAFE- M)
.cing orders. C}u; rent conditions of GUARDS $UBCOMMITTEE ON ADVANCED e Japanese ecopbmy and Japanese c*[anged t read C i ane su "# '%
41 industry woqld encourage Japa. refiner with its p'rincipal refinery in Meeting 6 .: 3 je producers tg continue to offer Crockett, Californfi." . " [. ' - ,
.ded stainless teel pipe and tube The competitive \ impact statement The ACRS Subcommittee on Ad- , _t - sale in the ted States at less was published in the Fror. rat. REctsrca vanced Reactors wn1 hold an open .g in inir value.y n}p of the Japanese on June 23,1978/iri vol. 43, pp. 27257- meeting on August 11.1978,in Room a aducers whof gate assurances to 27260* 4 f 1046,1717 H Street NW., Washington. . -'~ ] .
casury in 19'p wefe found not to be Dated: July 171197 . D.C. 20555, to review matters related ;q ling at less tnan fa}r value and were to the NRC sponsored research on the :2
- 1uded froni this inding. The re- CHAat s F. B. McAr.r En. ty i!ning two firrns, de pite assurances, ,Specic! Assistantfor 8[tiee of t s eeing as p b i re found tp be sell g at less than J,udgmerl,t Negotiations. at 43 FR 2G162 and 30631. June 16 and OE
.7 value. TVese facts ead to a strong IFR Doc.78-20777 FHed.,T-2M8; 8:45 am3 July 17,1978, respectively. #
a.ithood of further p}, rice depression / 4 In accordance with the procedures I outuned in the FcDERAr. Recrstra on 7. f > d harm tg the U.S. iydustry. In the
.cence of.a.n affirmadve finding by [6820-41] r i
- October 31,1977 (42 FR 56972), oral or written statements may be presented fN is Commssion,it is Illtely that those NATIONAL. ADVISORY, COUNCIL ON by members of the public, recordings 2
gi , panese chepanies currently selling ECCNOMIC OPPORTUNITY wn1 be permitted only during those . .j less thsh fair value1wiU interpree c tc e Commission's decis141 as a Ifeense COMMITTEE fROGRES$ RE ORTS,seWELFARE e portions ,a ofW theesmeeting ns may.when aitran. increase the margins di LTFV sales, REFORM, HEALTH CARE, UNEMPLOYMENT, e EM @ b conshu, meders M Ge M- g salting int increased limports. In- YOUTH EMPLOYMENT, AND OTHER ISSUE 5 . et, ad sM. ;-.
-eases in Japanese exports to the AFFECTING POVERTY CONSTITUENCY Persons desiring to make oral. state- .. g .ited State's at less thsn fair value M** tine ments should notify the Designated c ould intenkfy the dif4culties that # ~ .,k,; ,
Trently ex et in the U.S.4ndustry. J 24.1978. practicable so that appropriate ar-Fursua'nt to section'10 $f the Feder- rangements can be made to allow the ,, 4 R'onct.UstoN \~ al Advisory Committee Act of 1972 necessary time during the meeting for . 26 On the basis of the infognation ob. notice is hereby given that.the Nation- suchThe statements. agenda for - subject meeting M Commission)investiga- al Advisory Councu on Economic Op- '
.Q i .ined sn. I concluin the)de that an 1.$dustry in portunity will hold a 2. day ' meeting on shall be as follows: -
l :e United States is likely3to be in- August 28 and 29,1978,' at the Councu gy,'j,,,,g yy, f 37g, g.30 a.m. unia the 5
- red by rea.non of the importation of office at 1725 K Street NW (Room conclusion o/ business di elded stainless steel pipe #and tube 405),iWeshington, D.C. Tlie meeting
-om Japan sold at less than Entr value win pegin at 9 a.m. and is open to the, g[g',N.g* h$"e[N$t,'N '
'v
. thin the I meaning "of the publ,1c. i may be present, to explore and exchange
- j ntidumping A The purpose of the meeting will be their preliminary opinions regardinc mat- i ters which t.hould be considered during the ;,
to di Issued: July 2 1978* The;scuss committeescommittee progress of the Council willreports. meeting and to formulate a report and reo- ., meeti on August 28 to discusdwelfare ommendations to the Ien committee. , By order of th Commission * . At the e netuston of the Executive Ses- 4, ! Kpxwcrn R. bsoN. ref ' health crre, unempioyment. sion, the ::ubcommittee will hear presents. .- Sce"etary. yo h unemployment. .and% other tions by n.nd hold diseunstons with represen- i I { iss es that affect the poverty constitu* tatives of the NRC Staff, and their consul. . ! IFR Doc. 78-20B43- 45 am) e y. The committees wal meet in the tants, pertinent to the above topfes. The led 7-2G-78; g C uncu's office at 9 a.m. g subcommittee may then caucus to deter- ; a s .$ ' .} FEDERAL REGISTER, VOL. 43, NO.145--THURSDAY, JULY 27,1978
- I.=.:
l m ~ ATTACHMENT A l
.-.- - :n .
. 6
t
- 32473 NOTICES
,' l Persons with questions may can or lated safety evaluation suppotting !!-
ense No. NPP-0;(2) the report of th2
- r. Ine whether tha mattirr id:nttit:d irr ths i 1.otal session have been adequately covered write Dr. Harry J. Watters or M7. R. dvisory Committee on Reactor Safe- '
and whither the project is ready for review La rence Vandenberg in the Office of ~ by tha !ull commutes M agement and Program An ysis, gt ards, dated April 12,1978;f(3) the regarding tel hone 301-402-7721. Of(ice of Nuclear Reactor (Reguis- I Further information safety evaluation j report ! topics tar be discussed, whether the Wijtten comments should e ad. tion's I meeting has been canceled or dresseti to the Director, office f Man. (NUREG-0308) thereto dated November and1977supplement and No.1 f rescheduled, the Chairman's ruling on agement. and Prograrn Anal) is, U.S. June 1978, respectively; (4) the Ucens- ! Nuclears Regulatory Coritmission, ' requests for the opportunity to pres. Washington, D.C., 20555. Comments ee's firpt safety analysis report and ent oral statements and the time anot. must . b received by September 15, amendments thereto: (5) th,e licensee's ted therefore can be obtained by a pre- environipental report and amend. paid telephone can to the Designated 1978.~ ments tnereto: (6) the draft environ-Federal Employee for this meeting. Dated t Bethesda, Md.; this 21st mental ) statement (NUREG-0070), A dated Mak 1976; and (7) th' e final enyl-Dr. Richard P. Savio, telephone 202- ' day of Juli.1978. y s G34-32G7, between 8:15 a.m. and- 5 c- p.m ' . For thet U.S. Nuclear # Regulatory ronmentals statement (NUREG-0254), e.d.t. Commission. a dated Jurie 1977; are available for . q .- [ public inspection at the(Commission's Dated: July 20,1978. - Lrz V. Gossica~ Publie Doc'ument Room at 1717 H Joros C.Horzz, Adtssory Committee
. Ezeeutitie Director for Operations-Street NWa Washington, D.C. 20555 and the Arkansas Polytechnic Conege,- .
-{
Management O// feen IFR Doc. 75- 0751 FIIed 7526-78; 8:45 ami RusseUville, Ark.72801.} [FR Doc. 78-20633 Filed 7-26-18; 8:6am). $- Copies of I(ems (1) and (2) may be -
"* obtained upon request addressed to ;
i the United Stiites Nuclear Regulatory [7590-01)
- l. -
Cadmission. Washington, D.C. 20555r, Attention: Director, Divistort of Proh ,
$U2VEY OF PouCES AND PROCEDURES AP. ND UGHT CO.,
cet Management. Copies of Items (3) i PtCA3tE TO THE gXPRESSION OF Dn it. . ARKANSAS P' OWE and (7) may be' purchased at current i 8 (ARXANSA! NUCLEAR.ONE,UNif 2y - 148 PRO 713510NAL OPIN60NS rN rates from the National Technical In-issunnes a .Fatit y Operating License formation Service.1 Department of
\ Request foe Cem,n nh Commerce, 3285'iPort Royal Road.
The Nuclear Regulatory Commblon' clear Notice 15 hereby given that the Nu,. Springfield, Va. 221 1.
- Regulatory Commission (the (NRC) has prepared a Survey of foll- ' Commission) hasiissued.faciuty oper- Dated at Bethes a. Md. this 13th cies and. Procedures Appucable to t.he ating license No.4NPF-8 to Arkansas day of July 1978. f Expressich of Differing Profes!.lonal Power & Light Co., authorizing the For the Nuclear , egulatory Com ,
Opinions.tThe concepts and broce. g dures described in the survey ,ivill be ,unit loading of fuel and maintaining in a operational mode 5 condition the ,
- ) JOHN F.Srotz, !
the basis fo'r subsequent development (cold shutdown condition).The Arkan- 1 of an NRC hency. wide procejiure for sas Nuclear One, Unit 2 plant is a pres- Chief, Light Wcter Reactors. Branch Nd.1, Diyisfort of Proh cm;:loyees tq raise differing profes. surized wated reactor' located at the 11 sicnal cpiniens. The NRCf requests censee's site in Pope County, Aric.. cet Manapement g commen:s- upon the surve7 findings, The Commission has.made approprt* [FR Doc.78-20755 Filed 7-25-78;8:45 pm) cnd also any c.dditional recommenda. ate findings!as recuired.by the Atomic < , tions or experiences regarding poucy Energy- Act; of .1954, as amended (the- i criteria, procedures, or other mecha. Act), and ;the Commission's regula- [7590-01] , i nisms that the " Commission should tions in 10lCFR ChaptertI. which are , consider before prdceeding with devel, set forth in the license.'The applica- IDocket nos.50-282 an 50-306) f5 . ' opment of an ageneg-wide procedure. tion for the license complies with the / NORTHERN 57 ATE 5 POYrER CO. The survey presentsf'a set of proce- standards land requirements of the Act I A dural steps comprising a system for "" gulau " *" *"" "8 raising dif f ering prof essio al opinions. T nct I herance ithe f k Examples of specifier procedures from ifcensinglaction encompassed in the The U.S.INuclear Regulatory Com-EcVeral orgs.nizations inre presented combined notice of receipt of appUca. mission (the Commission) has issued within a framework encompassing all tion fob facility operatinit license; of the procedural / steps, The survey notice of availability of applkant's en- amendment /Nos. 30 Nos. and 24 to facility also contains a description of overall vironmental report; and' notice of op. operating license DPR-60, issued to the Northern States DPRi42 and criteria that can be used to judge the portunity' for hearing published 1974 in the (39 Power Cd. (the llCensee) which revised effectiveness of' A system;for rafsing PrDERAtl REGIsT!21 on April 23, S ' technice.1 specifications. for operation ditfering professional opinions. FR 14371). of UnitiNos.1 and 2 of the Prairie Single copies;of..the survey can be The license is effective as of it:r date obtained by writtng to the U.S. Nucie- of issudnce and shall expire 6 months Islandlocated facilities) nuclear Generating in Goodhue Plant (the - County'. Er Regulatory Commission. Office of from e' aid date, unless extended for Minn. The amendments will become alysis. good c$use shown, or upon earlier issu. Management Washington /D.C / and20555. Program In At]dittorr. ad ance o'r denfal' of a subsequentlicens- effecthe as of the date') of t issuance.The amendments revisedithe a single copy of the survey is available. ing aciton technical specifications for the fa)ili-A copy of (1) facility operatinA 11 ties to incorporate surveillance re-
;-and may ,be Inspected. and copiedain censet' No. NPP-6 complete with ledh- quirements relating to the ECCS the Commission's Public: Documeht Room at.:1'i1*T H Street NW., Washinit- nicallspeelffcations (appendices A and ton D.C. Copiesof comments received B)' / attachment 1 ), preoperatlorUtl throttle-valves and delete fulfilled coh-testi and other items which must be ditioit 2.C.(3)b of the licenses regard-ing the spent fuel' pool' radiation are also.available.for inspection,at.the combleted prior to loading fuel (at-Washington. D.C. Public Document tachment 2) and the Commission's re-levels.\
Roont. FEDER AL RECISTEt; Vot. 43; NO: 145- THUR50AY, JULY 27,19r8 ....o , g e
- t. .
ACRS SUBCOMMITTEE MEETING ON ADVANCED REACTORS WASHINGTON, DC - AUGUST 11, 1978 i ATTENDANCE LIST ACRS NRC STAFF , l M. Carbon, Chairman' - C. Kelber
- W. Kerr L. Rib J. C. Mark M. Silberberg P. Shewmon B. Bursen R. Nicholson, ACRS Consultant T. Walker R. Seale, ACRS Consultant - . D. Sells R. Savio, Designated Federal Employee T. Murley R. Schamberger R. Curtis ARGONNE NATIONAL LAB.
P. Garner ' SANDIA LABS. D. Ferguson R. Coats J. Walker LOS ALAMOS SCIENTIFIC LAB. C. R. Bell DEPARTMENT OF ENERGY J. Griffith WESTINGHOUSE ELECTRIC CORP. J. Leary ' H. Alter H. Huang MEMBERS OF THE PUBLIC OAK RIDGE NATIONAL LAB. T. Wehner T. Kress BROOKHAVEN NATIONAL LAB. D. Schweitzer C. Sastre ATTACHMENT B
--e ,---- -. e r- , , -s 1 - -- - , - - * , w , r n , e
PROPOSED SCHEDULE FOR 1 !. AUGUST 11, 1978 MEETING. '., , T
~- -
OF THE ADVANCED REACTORS SUBCOMMITTEE j- l
. . j- }
- 1. Executive Session 8:30-9:00 a.m. ,
- 2. NRC Presentations - Overview of the NRC 9:00-1 :00 p.m.
sponsored programs - 4 hrs. . J
- a. Introduction - (Program summary.
, .[
Discussion of the RES report to the ;, ACRS on Advanced Reactor Research) 45 min. , s, T. Murley and C. Kelber , i
- b. Analytical Advanced Reactor Safety 1 hr. 15 min.
Research Branch - R. Curtis ,
- c. Experimental Fast Reactor Safety 1 hr.15 min. ~
i Research Branch - M. Silberberg ! i
- d. Experimental Gas-Cooled Reactor Safety Research Branch - R. Schamberger 45 min.
LUNCH 1:00-2:00 p.m.
- 3. DOE Presentation:; - Overview of DOE 2:00-5:00 p.m. i sponsual programs - 3 hrs. I
- a. Liquid Metal Fast Breeder Programs - 1 hr. 3 min.
. J. Griffith
- b. Gas Cooled Fast Breeder Reactor '
programs - D. Erb 45 min. 't C. Gas cooled Thermal Reactors programs T. Werner 45 min. ,
- 4. General Discussion and Conclusions .2 hrs. 5:00-7:00 p.m. -
.j ATTACHMENT C
.-..}-. 1 i
i a I I* L
w, , -
/ . -
' ^
~ '
- ! ADVANCED REACTOR SAFETY RESEARCH ACCOMPLISHMENTS Y July 1977 - AususT 1978 ie q,
~
J m FBR-ANALYSIS PROGRAM 3* l e SIMMER-II CODE FOR HCDA ANALYSIS DOCUMENTED AND RELEASED TO USERS e SUPER SYSTEM CODE (SSC) RELEASED TO USERS t 4
; e PRODUCTION VERSION OF FUEL PIN TRANSIENT BEHAVIOR CODE',' LAFM-1 COMPLETED
!! ~
i e 3-D THERMAL-HYDRAULIC TRANSIENT CODE (COMMIX) DOCUMENTED AND RELEASED TO USER 1 p b i . 9 S 9
. l
.i .
. e 6 ff ADVANCED REACTOR SAFETY RESEARCll ACCOMPLISHMENTS JUNE 1977 - AususT 1978
, GAS COOLED REACTORS ,
e MATERIALS TEST LOOP RELIABILITY DEMONSTRATED , e L0ll CYCLE /IIIGH CYCLE FATIGUE CORRELATI0tlS OBTAINED e CATALYZED PGX (STRUCTURALLGRAPilITE OXIDATI0tl 00AfiTIFIED . e NINE CODES DEVELOPED AND RELEASED FOR TESTING - e DEMONSTRATED VALIDITY OF PHYSICAL SCALIflG 0F GRAPillTE BLOCKS FOR CORE SEISMIC TESTI e llELIUM AFTERGL611 MOISTURE It0!!ITOR DEM0flSTRATED IN FSV e FORT ST, VRAIN ECCS AtlALYSIS FOR NRR COMPLETED e CONFIRMED GRAPillTE OXIDATION RATE INDEPENDENT OF TENSILE STRESS CONDITIONS b e
^
^- - -
. ~ - - .
r ..
* ~
.] -
ADVANCED REACTOR SAFETY RESEARCil ACCOMPLISHMEtlTS JULY 1977 - AususT 1978 FBR-EXPERIMENTAL PROGRAM . INITIAL FIRST-0F-A-KIND TESTS C0fPLETED IN THE ACPR e PROMPT BURST ENERGETICS TESTS COMPLETED ll/0XIDE AND CARBIDE FUEL. a CORE DEBRIS C00 LABILITY .
~
VISUAL OBSERVATION OF IRRADIATED FUEL DISRUPTI0tl TESTS ~
- i . .
- e .ACPR UPGRADE COMPLETED .
e TEN LARGE (200Ks) SODIUM-CONCRETE INTERACTION TESTS COMPLETED; (NRR USER REQUEST) - FFTF SUPPORT TESTS UilDEPJlAY , i e INITIAL LARGE (200Ks) SUSTAINED, MOLTEN S.S.-C0!! CRETE INTERACTION TESTS COMPLETED; (NRR USER REQUEST) I
- m
.- ~ .i . -
i- .
. I, l FBR-EXPERIMENTAL PROGRAM (c0NT.) ,
- e. S0DIUM AEROSOL PROPERTIES DATA 0BTAlilED FOR IMPROVED AEROSOL TRANSPORT CODE (HAARM-3). EARLY RELEASE OF CODE TO NRR FOR FFTF.
l , e FAST FACILITY COMPLETED AND CHECKED OUT (O'RNL - Sift 1ER QUALIFICATION SOURCE TEP11) i i .
.e FIRST MIXED AEROSOL- (FUEL-S0DIUM) TEST RUN IN NSPP j l
. ~
i t ! j. ! i i* .g
.i .
3
. l L
4
- - - - - - - ~
,, y . . . .
1
~
upon recent accomplishments, their implications and our future plans. A.. Reactor Safety Modeling and Coordination (7.1.1.13) (1) Accomplishments s Completed physics studies on conceptual 2000 MWe LMFBR e Completed the BOL core comparative studies with the l ECC Whole Core Accident (WAC) group.
.e Completed initial SAS-FRAX code comparison as a part of the UK/NRC. joint study.
o EPIC code has been incorporated into SAS-30 and has been provided to Sandia e Completed investigation of boiling pool problems.
.{
(2) Safety Implications e The very high Na voiding rates suggested by previous , investigators have not been encountered. ! l , e The lower smear density used in the US/NRC accident l t 1 studies appear to rule out LOF/ TOP scenarios found in certain CRBR accident studies. e A simple pin failure criterion after Moorhead (UK) )
~
which combines burst failure with melt-through appears ! to have promise. .
\
e Extended SAS-EPIC runs which run to neutronte shutdown ; i
- show lower fission energy levels than comparable SAS-VENUS runs since in the'1atter cases the velocities
- l and accelerations are set-to zero at the start of the l l 1
VENUS portion of the analysis (NRR has been aware of this conservatism) . O
.e ,-
- m. i,~~ ---.,~..,me.,- ..,--,.,m.s,,-w. ,,,,..-r._,.--,.,-.,% ,ey.w__,_ - , . _ , r,,,, - . - w.. .,-,.m.-., , .....c%. - ._,.,m.--.- . . . , . , , . - - - - - we-,,
l
= .--
l (3) Future Plans e Extend cooperative studies to cases with irradiated fuel e Extend cooperative studies to' a detailed comparison of fuel failure models e Extend EPIC to improve modeling to treat gradual meltdown accidents. This will mean replacing EPIC CLAZAS and SLUMPY with an integrated module to be called FCC-1 (F_uel, Clad, , Coolant). B. Containment Analysis (7.1.1.12) (1) Accomplishments '
. e Completed a review of US and foreiga containment l analysis methods. From this review has come an evaluation of models which are satisfactory for use in CONTAIN. Work has started on developing models for phenomena where satisfactory models could not be found.
o Code framework and module interfaces have been defined. (2) Safety Implications - This program has been closely coordinated with related experimental work. The initial safety imp 1tcations have come from the experimental program and will be discussed ! there.
,; . . a. . .
erf (3) Future Plans e Short term development will concentrate on a cell interaction model to handle concrete-sodium-core melt interactions within an enclosed cell. This module will be used for experiment analysis.. It will-then be a building block for the system analysis code. e First version of system analysis code with largely existing models "to begin operational testing next summer. C. Accident Delineation Study (7.1.1.11) (1) Accomplishment s Completed event trees for the impact of engineered systens, in-core phenomenology associated with HCDA and the containment phases of LMFBR accidents. (2) Safety Implications e While these event trees are clearly in a preliminary stage, we believe they can contribute to the *
. identification of dominent uncertainties. .
e This work supports the experimental program by providing a means to rank order the branches of the '
)
event tree and thus provide a basis for setting priorities among competing experiments. l l l 1>
* ~ ^
,9 .
.g .
t (3) Future Plans-e An interim report on the feasibility of applying these methods in the absence of a firm design is due in September. Our preliminary indications are that it will recommend continuing the study and to focus it upon the containment phase of HCDA as a tool to s develop CONTAIN and the associated experiments.. L. COMMIX, A 3-D Thermal-Hydraulics Code (7.1.1.10) (1) Accomplishments
- e. An exportable version of the single phase 3-dimensional ,
thermal-hydraulics code COMMIX-1 has been completed. The coce has been released on a trial basis' to the LMFBR community. Copies of the code along with . j documentation have been provided to West'inghouse' Corporation (WARD), General Electric Co. , Atomics, International, Babcock and Wilcox Co., MIT and the UKAEA. , e COMMIX-1 has been used to calculate the 1/15 scale FFTF outlet plenum experiment performed at ANL with good results. This is a comparison of measured and calculated outlet temperatures. ; e The code has been used to calculate flow in FFTF piping during a flow coastdown. The calculation , shows severe flow stratification and reversal l l indicating a 'need for experimental data in this area.
. O
- . . . ~ . . - . . . . . . - . . . . . . . - . . . . .. 9
- Unfortunately, the calculation neglected heat transfer from the sodium to the pipe which might alleviate stratification.
e Calculations h:ve been made on the ORNL 19-rod . bundle experiments with flow blockage. When the wire wrap is taken into account the agreement is good. e The two-phase vers, ion, COMMIX-2 is being developed. COMMIX-2 was used to simulate the high pressure jet l impingement experiment. At time t = 0, a _high pressure jet containing a mixture of steam (67%) and water (33%) enters into a stagnant atmosphere and impinges on a 1 vertical wall. The experimental set-up and its initial operating conditions are shown in the next slide. A comparison between the calculated results obtained by COMMIX-2 and the steady-state pressure measurements on the impinged wall, and results obtained from a sensitivity study of variations of interfacial drag coefficient (K) and evaporation rate (A]. Note that agreement is good between the experimental data on pressure distribution and the calculated results with-K = 2.0 x 1012 and A = 100. I. e
&^f ;
i
- l
. , . ,- .. . . .- . m. ,;...
1 1 (2) Safety Implications COMMIX is proving to be a very versatile and useful -] 1 tool. We expect it to become the standard for the l detailed analysis of the thermal hydraulic perfomance of reactor components. (3) ~ Future Plans e A boundary fitted coordinate system met)od to transform a hexagonal fuel bundle into X-Y geometry for solution of the 3-D thermal hydraulics equations is being developed (ANL-78-1). The methods should allow rigorous solution of the problem without assumed L
. mixing coefficients. The transfamation method also shows promise for piping elbows- and mixing tees.
e The W-1 experiment planned for the sodium Loop Safety facility will be pre-calculated with COMMIX-1 and after the fact with COMMIX-2. - E. Super Systems Code (SSC) (7.1.1.7 and 7.1.1.8) (1) Accomplishments e The loop version of the code SSC-L is operational and + t l calculation of test cases in progress. A study is being made to compare SSC with the IANUS and FLODISC
- codes on FFTF. To make a direct comparison, it was necessary to use only 2 core channels plus a bypass.
This model proved to be very sensitive to the friction - t factor vs. Reynolds nu:nber in the laminar flow' and i transition regions. A careful review of the literature ( l.
+ . . . . . . .-
i _. ._ . - . ~ . . . . . . . _ _ _ . . . _ _ .e ..' .
-s-- I on this subject leads to the conclusion that 'there i is little or no data on laminar flow thru wire- l wrapped fuel bundles other than the 217-rod bundle -
data being generated at HEDL. .i e A 'model' for CRBR (with 12 channels in' the core and blanket)hasbeenconstructed. We should have 4 i results with this model for loss of forced circulation and pipe-break accidents by the September 25 meeting. e Numerical methods using nested time steps which greatly reduce running time were developed as part of this program. The various components or. processes are i
+
advanced at different timesteps which are a rational fraction of some master clock time step. Components or processes that.have short time constants are calculated more frequently than slow-moving components / processes. The savings in CPU computer time is a factor of 2.7 to 5 using MSS. - e A workshop for potential SSC users was held April 6 an and 7,1978 at BNL. The workshop was attended by 32 individuals (non-BNL) from 17 organizations. In general, it was a constructive two' day discussion. The ACRS was represented by a consultant Dick Lyon from
- i ORNL.
O S E
.g.
. j i
(2) Safety Implications f We expect SSC to become our principal tool for the investigation of accidents other than HCDA, ! (3) Future Plans - 4
; 1 e Continue improvement of plant protection and control .
1 i systems in SSC-L. e Continue to cievelop component modules for SSC-L to ) analyze the FFTF reactor. e Complete and release a first version of SSC-P that will model a pot-type LMFBR. e Continue development on a first version of SSC-S to simulate the long-term decay-heat removal from an LMFBR. F. Safety Related Critical Experiments for Validation of the VIM Monte Carlo Code and SIMMER Neutronics (7.1.1.5) (1) Accompitshments - e The experimental portion of the program was completed on schedule in December 1977. Data reduction is still in progress but will be completed by the end of September 1978. The experiments were designed to cover a wide range of spectrum effects and leakage effects encountered in distorted geometries. (2) Safety Implications e Monte Carlo estimates of reactivities which result from core damage are conservative (i .e. , too high), while reactiv' ities computed from diffusion theory [9-
4 i I i are not. Diffusion theory misses the reactivity change from reference. to slump-in by 1.7f. K,ff or 5 dollars. e Errors in the Monte Carlo core-damage reactivities are far from negligible. Since in these calculations, the structure of the critical experiments has been modeled, in enomous detail (every plate) ANL feels these remaining errors can be attributed to errors in the cross sections. e In addition to the reactivity measurements, other - important safety related quantities such as material worths, reaction rate profiles, Doppler coefficients central control rod worth s/1 and 8,ff. Data reduction - on these measurements is still in progress. l
. l (3) Future Plans !
! I e Re%ction and. reporting of the experimental data will be ampleted in FY 1978. The VIM Monte Carlo calculations '
will be redone in FY 1979 using improved cross section sets to resolve the small errors in the present l calculations. , , e VIM Monte Carlo calculations of the experiments using I homogenized core regions will be made to allow a direct ; , I l comparison between VIM and transport theory codes. p e
. , s G. Fuel Pin Transient
- Behavior (7.1.1.4)
(1) Accomplishment l e The Los Alamos Failure Model (LAFM) computer code has been completed, published, and (to the exters feasible) verified. The code successfully predicts the time and location of fuel failure or the clad strain for pins which did not fail in, TREAT transient overpower experiments. (2) Safety Implications , e The analysis'of the LOF/ TOP accident depends heavily I upon the time and location of failure of fuel in the unvoided channels. This code correlates available , I experimental evidence and provides the means for t assessing this critical turning point in the accident scenario. , (3) Future Plans l This program ends with this fiscal year. Future - improvements or extensions will be undertaken as part of the SIMMER program as the need and opportunity presents itself. H. SIMMER (7.1.1.2 and 7.1.1.3) .
- 4 (1) Accomplishments s SI)NER-II has been completed and documented. i I
. 4' l i
f
e A three day tutorial on the new version of the code was held in April with attendance from Japan, Europe and essen'tially, all the U. S. organizations involved in fast reactors. e Completed a program plan for SIMMER qualification testing. e A sensitivity study to examine the imptet of changes . to 25 key parameters used in SIMMER exchange functions will be completed in September. e A comprehensive documentation of the literature searches, experimental evidence and other bases used to select f the models and correlations now in SIMMER will be completed by Septe.mber. (2) Safety Implication,s e Evidence that the. low energy conversion to work predicted last year is correct has grown during l l the last year. Many more variations have been run J and the processes are better understood. ' e Preliminary runs to study transition phase indicate that SIMMER is a useful tool and that an attack on , I the problem of the scope of the last two years work
~
energetics should yield substantial new insight into the transition phase. l l l l 1
1 (3) Future Plans e Concentration of near term qualification testing on establishing the credibility of vessel problem snergetics predictions.. e Concentration of model and code development on improvements identified in studies of subassembly meltdown propagation and the transition phase. III. Safety Test Facildtf es (Section 7.2) A. ACPR Upgrade (7.2.1.2 and 7.2.1.3) . l (1) Accomplishments e Reactor upgrade has been completed on schedule and within budget. _ _ _ . e Coded-aperture gamma ray imaging system to measure test fuel motion is ready for installation.
~
(2) Safety Implications - e Increas.e capability. to perform and diagnose experiments on prompt-burst energetics, fuel dynamics and post accident heat removal. (3) Future Plans This facility is dedicated to the ekperimental program in material interactions and will be discussed'briefly by l M. Silberberg and in more detail on your visit to Sandia.
\G o,,-y,.. - - . , s . - v-, ' ' ' ' '
. ? 1 I
l l B. Review of STF and Fuel Motion Diagnostics (7.2.1.4 and 7.2.1.5) These programs have been very productive in the past, however, in view of circumstances, bo'th will be terminated on September 30. Early this year, when it became apparent that the DOE SAREF program would be substantially deferred, they were told to bring the wort to an orderly conclusion and to prepare a final report. Where work on instrumentation could be directlp' applied to existing experimental programs, it was to be transferred and ! justified as part of the experiment program. IV. Summary Analysis is the core of any reactor safety research program. It ! provides the means to integrate the many competing phenomena and extrapolate to reactor conditions. O e
. l l
l
\'1
l ! ADVANCED CONVERTER PROGRAM ADJUSTMENTS FY 1978 " i 1 ITEliS ASIS OF ITEMS SASISOF
, TASK REDUCED IEDUCTION INCREASED INCREASE -
4
~
! FUE' AND FISSION FyEL BE];1AVIOR: COMMERCIALIZATION P RISK I YRODUCTS UPER. IEMP. SCHEDULE FUELANDF.2$00K BEHAVIOR:> CONTnInuTION i PRIMARY COOLANT PRIMARY SYSTEM FUNDING DEPRESSURIZATION: PgTENTIALRISK " " JET MIXING i i [ MTERIALS CORE GRAPHITE KNOWLEDGE STRUCTURAL FSV CONNECTION i
,, GRAPHITE .-
! STRuCT' URAL RESPONSE CORE SEISMIC FUNDING !. $ i l! ' COOLANT IMPURITIES ITEM COMPLETION . i INSTRUMENTAT10N FSiNOISEANALYSIS FSV t GRAPHITE NDT FUNDING ' , i . ! CORE IEllP. > 2500- K FUEL AND F.P. ACCIDENT DELINEATION ' TRANSFER TO: : ET MIXING PRIMARY COOLANT ' ANALYSIS ALL EXCEPT FSV FUNDING
/
r . . w*
-- - -- -_- --- --- _ -------_a----- - - - . __ - - - _ - . - - - - - - . - - - . - - - - - . _ - - - - - - _ - _ . - - - - . - - - -
a b
.. ! u
- i. l l
ADVANCED CONVERTER: ACCOMPLISHMENTS [ FY 1978 l l1 i !
. l FUEL BEHAVIOR AND FISSION PRODUCT TRAN,SPdRT ,
O PROGRESS IN THE IDENTIFICATION OF FISSION PRODUCT CHEMICAL SPECIES.AND WHERE THEY DEPOSIT AS A FUNCTION OF TIME, TEMPERATURE AND DEPOSITION SURFACE. O INITIAL RESULTS ON THE TRANSPORT OF FUEL AND FISSION PRODUCTS AT VERY HIGH TEMPERATURES. IN H II51 (Cone) . GRAPHITE. l I l
. l
ADVANCED CONVERTER: ACCOMPLISHMENT FY 1978 PRIMARY COOLANT 0 IDENTIFICATION OF MATERIAL GAS PHASE INTERACTIONS (IN Hil) AND THE POSSIBLE CONNECTION WITH H2 DIFFUSION IN STEELS. , l 0 SMALL SCALE DEPRESSURIZATION TESTS - JET MIXING IN THE l SECONDARY CONTAINMENT. e . e 4
- m- .. ==:
9
? .
I
~
ADVANCED CONVERTER: ACCOMPLISHMENTS FY 1978 ; MATERIALS BEHAVIOR , O CORRELATION OF STRUCTURAL GRAP-HITE STRENGTH REDUCTIOR WITH WEIGHT LOSS FR0!! OXIDATION: DEPENDENCE ON STRESS, TEMPERATURE, OXIDANT AND OXIDANT LEVEL. O EXTENSION OF HIGH TEMPERATURE METALS DATA BASE. 5 l st
..:.. ==; =='
t ' l ADVANCED CONVERTER: ACCOMPLISHMENTS FY 1978 , STRUCTURALRESP0 HSE O 3-D SEVEN FUEL COLUMN SEISMIC RESPONSE CODE O PCRV FINITE ELEMENT CODE (NONSAP-C) DOCUMENTATION AND TESTING. O VALID SCALING OF GRAPHITE WITH PLASTIC - SEISMIC RESPONSE l 4 e 9
. . - . . . . . . . . . . . -.. ..-. . . - . . _. . l
; i I
o e ADVANCED CONVERTER: . ACCOMPLISHMENTS
. l FY 1978 ;
INSTRUMENTATION -
- O CORRELATION OF-GRAPHITE STRENGTH WITH SONIC VELOCITY.
O TESTING OF HELIUM AFTER-GLOW DETECTOR AT FSV, S 0 0 n 0 e 33
- ~.-, w . ~ , , , _*
9 ADVANCED CONVERTER: ACCOMPLISHMENTS - FY 1978 *
- I, ACCIDENTDELINEATiON !
O ASSESSMENT OF COMBUSTION HAZARD IN DEPRESSURIZATION i ACCIDENTS. . . . O GRAPHITE AEROSOL GENERATION O OBSERVATION OF MATERIAL TRANSPORT IN GRAPHITE AT VERY 1 HIGH TEMPERATURES. 0 l 3 P e i AP'
- l .
t I ADVANCED CONVERTER: ACCOMPLISHMENTS
.FY 197'8 ANALYSIS e FSV ECCS ANALYSES e ECCS ACCIDENT CODE QUALIFICATION STUDIES e VENDOR CODE EVA'LUATIONS ,
e FISSION FRODUCT IRANSPORT CODE 3
- i
. . 1 e
. l 9 (* i
~
j PROPOSED HTGR COMERCIALIZATION COST AND RISK SHARING DDE SUPPLIERS UTILITIES
~
SUPPORTING R&D X t i4 EXT PLANT DESIGN AND LICENSING X X X i i l'
~
l 11 EXT PLANT ENGIdEERING X X .X - I
\
t I' NEXT PLANT iiANUFACTURING X . l . !' NEXT PLANT CONSTRUCTION X X i ADDITIONAL (COMERCIAL) PLANT X , X l 1 DESIGN, LICENSING, MANUFACTURING, . ! AND CONSTRUCTION
~!! '
h i s ;' e __-__m
---- _____m . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _
e
- .. . e p
c HIGH TEMPERATURE GAS COOLED REACTOR : - ( j. ! POSSIBLE COST SHARING
- i. ARRANGEMENTS FOR COMMERCIALIZATION -
l (THROUGH THE FIRST LARGE COMMERCIAL PLANT) e ! OPEN ENDED R&D COSTS l . ! RISK ' (PLANT & R&D) " EQUIVALENT" GOVERNMENT - l o PLANT COSTS (D.O.E.)
= >. '
! b PREDICTABLE COSTS - l, y $2 BILLION 2 1 SUPPLIER I-UTILITY ESTABLISHMENT ! OWNERS COSTS . h SUPPLIERS
. l!IGH TEMPERATURE GAS C001FD REACTOR WORLDWIDE PLANT EXPERIENCE :
OPERATil4G EXPERIENCE , lial3E SIZE I g .- PEACH BOTTOM 40 MWE 7 YEARS (1967-74) I. FORT ST. VRAIN 330 MWE STARTUP TESTING NOW-MAX.. POWER TO DATE 6E 700 -1100 MWE FIRST PLANT NEAR HAVING CONSTRUCTION .; 10 COMERCIAL PLANT '
~
PERMIT, THEN-ALL CANCELLED 1974 . COMITHENTS FEDERAL REPUBLIC M-fe -= = #=~.-<-~- 4 e s-=5 . 0F GERMANY AVR 15 MWe 12 YEARS (1966-78) 300 MWe UNDER CONSTRU(TION 98 -81
- OPERATIVE l'
[' THTR I
- hf c.. -w ,
=== O Y 4 L N$ &hM '
!; lb.En. ' ' 11 YEARS (1964-75)' I i DRAGON 20 MWT . h
FUNDING HISTORY . THERMAL GAS C001FD REACTORS-BUDGET AUTHORITY $ (MILLIONS) FY 1976 FY 1977 FY 1978 FY 1979 'FY 1980 OPERATING 17.5 16.6 31.0 28.5 EQUIPMENT 0.6 0.6 1.0 2.5
~
18.1 17.2 32.0(1) 31.0 (2) TOTAL ~
. q
- i t
I (1)DOEHASREQUESTEDCONGRESSIONALAPPROVALTOREPROGRAS.$1.2MI , I' ADDITIONAL FUNDS TO HTGR (2) THE LEVEL OF FUTURE GOVERNMENT SUPPORT TO BE DECIDED ON BASIS OF l , OCTOBER 1978 PROGRAM REVIEW TO BE CONDUCTED BY THE DOE ASSISTANT
- SECRETARY FOR ENERGY TECHNOLOGY R
i
FY 1978 HTGR PROGRAM FUNDING TO NEAREST $100K .
~
- l. B/0 ($000) .
l HTGR l GENERIC AND STEAM CYCLE ADVANCED HTR v . i FUELS 4000 0 ;
~
MATERIALS 3700' - 1400 COMPONENT AND SYSTEM R&D 8100 300 SAFETY 900 0 i DESIGN 2050 2800 i' - + 3250 -* 0 UTILITY INITIATIVE. ! 22000 4500 .l ! EQUIPMENT 800 l i . 22k00 4500 i u 27300 , j-TOTAL 1
.h 6 j ll
4 HTGR .
~
FY 1978 PROGRAM FUNDING BY CONTRACTOR B/0 $(000). GENERAL ATOMIC 16395 (INCLUDES 470 EQUIPMENT) .-~ o ORNL 4720 (INCLUDES 330 EQUIPMENT) ;- GCRA . 3250 . t GENERAL ELECTRIC 2510 ' , l L LASL 150 ; 275 i MISCELLANE0US i ; TOTAL 27300 G i [. 3
- 1 _..T ' i .:..... _ _ _ _
.. j FY 1978 AND FY 1979 GCFR BUDGET gyy .
CONTRACIQR IfEl IASK id- *k SAFETY
;/
r
.!l ANL 01345 FUELS & MTERIALS $ 900 $ I,000 ;
01346 ZPR PLANNING & ANALYSIS 100 140 / 01353 SAFETY , 610 680 J l SUBTOTAL . I,610 1,820 . ORNL 01350 PCRV- 150 200 01351 SHIELDING 600 780 01352 CFTL 2,350 3,500 X SUBTOTAL 3,100 4,480 LASL 00354 SAFETV - 540' 810 / , EG8G 00520 GRIST 600 620 / 00616 RELIABILITY 45 50 / , SUBTOTAL 645 670 . GA. 00581 MNAGEMENT 200 150 00582 CORE ELEMENT I,100 I,150 00583 FUELS AND MTERIALS 700 850 00584 NUCLEAR ANALYSIS 450 500 00585 SYSTEMS ENGINEERING 210 400 00586 COMPONENT DEVELOPMENT I,749 3,305 00587 CIRCULATOR 220 0 00615 NUCLEAR ISLAND 300 300 00638 DYNAMICS 265 325 00759 ALTERNATE DESJGN 700 300 00761 ALTERNATE FUEL CYCLES 200 300 ; 00588 IN-PrLE LOOP IESTS 200 220 / 00589 REACTOR SAFETY 580 580 / 00617 RELIABILITY 30 30 / SUBTOTAL 6,904 8,410 GESW 00705 GE INDEPENDENT DESIGN STUDY 50 100 00708 W INDEPENDENT DESIGN STUDY 50 100 SUBTOTAL 100 200 HBA 00663 GCFR MNAGEMENT 270 700 OTHER CAPITAL EQUIPMENT I,200 900 DIRECTOR'S RESERVE 0- 850 HEDL VAT 0 3,360 SPS 131 0 / TOTAL B0 $14,500 $22,200 DATE BUDGET DEVELOPED 5/78
) >;
m ._ _.- _ _ _ _ _ _
3 I i -
~
j ANL SAFETY PROGRAM - PROGRAM: GCFR SAFETY ASPECTS OF FUEL AND CORE ll FUNDING LEVELS: (ACTUAL AND PROPOSED) y mza , nZz 19za mza - ll -$660K $700K ' $6IOK- $680K
- i jl MAJOR ACTIVITIES (1977-1979)
l ii 19zZ o ANALYZE THE EFFECTS OF HIGH BURNUP AND ABSORBED HELIUM ON ACCIDENTS. . IJ -o ANALY2E THE POST-ACCIDENT CORE DEBRIS BEHAVIOR (FUEL-GRAPHITE AND FUEL-CONCRETE INTERACTIONS).
- l. ,
j o COMPLETE HIGH PRESSURE, FLOWING HELIUM, DEH-TEST CHAMBER. . 4 i . BZR o DEMONSTRATE EFFECTS OF FUEL SWEEP 0UT DURING HIGH-RATE TOP ACCIDENTS. . ! o CDA ANALYSIS FOR A GCFR DEMO PLANT DESIGN.
~
I [ . EZR o INTEGRATION OF EARLY LASL AND ANL TEST RESULTS INTO ACCIDENT ANALYSES. .
- o DETAILED DESIGNS AND TEST SPECIFICATIONS FOR THE GRIST-2 TEST TRAINS.
k - )j~. J i#1
-- __--_- _ _ _ _ _ _ - __ - __-__*_____-__.__m_s _ ____m_ c_ ' " ' _e " _ .._I -"_ - T '~ -'
'"5- ~ - - - - - ' - - - - ' - - - -
i l}-, i: . !j . CFTL PROGRAM FIFMENTS
~
PURPOSE: ~ TO PROVIDE GCFR CORE ELEMENT ENGINEERING CONFIRMATIONS l 0F HEAT TRANSFER, PRESSURE DROP AND STRUCTURAL. BEHAVIOR . UNDER fl0RMAL, OFF-DESIGN AND DBDA OPERATING CONDITIONS. jl,
.)x ll
! METHOD: DIRECT ELECTRICAL HEATING (DEH), OUT-0F-PILE TESTS AT ll 4 ORNL.
~
PROGRAM ElFMENTS: ' ) ~ LOOP - DESIGN AND CONSTRUCTION OF THE FACILITY TO PROVIDE THE TEST BEDi INSTRUMENTATION AND DATA ACQUISITION EQUIPMENT. . TEST BUNDLES - DESIGN AND CONSTRUCTION OF THE SIMULATED CORE
. . ELEMENTS FOR TESTING.
i . [ OPERATIONS - EXECUTINE THE EXPERIMENTS; EXAMINING THE TEST BUNDLES; MAINTAINING THE TEST LOOP. I ! ANALYSIS - PLANNING THE TESTING; PROVIDING THE.RESULTS IN A l USABLE FORM. th
.~-
4
.1 A
L-- _ - - _. . _- . ._ . . . - _ _
- l. ,
~
CFTL MILESTONES : i TEST LOOP- . !; I. COMPLETE CONCEPTUAL DESIGN - SEPTEMBER 1977-
- j. 2. COMPLETE TEST LOOP CONSTRUCTION AUGUST 1980
- 3. COMPLETE ACCEPTANCE TESTS OCTOBER 1980 l
p TEST BUNDLES l ' l I. COMPLETE CONCEPTUAL DESIGN APRIL 1978 !
. 2. COMPLETE FABRICATION OF IsT BUNDLE DECEMBER 1979
- 3. INSTALL IsT BUNDLE IN LOOP AUGUST 1980 TEST OPERATIONS I. START TESTS WITH BUNDLES APRIL 1981
,.. 2.n COMPLETE TEST PROGRAM AUGUST 1983 AEE80XIMATE COSTS TEST LOOP CONSTRUCTION AND EQUIPMENT $27M R&D 4M TEST BUNDLES 8M
. TEST OPERATIONS _ZM ESTIMATED TOTAL $46M-l^ _ .
~
{- : i LASL SAFETY PROGRAMS < PROGRAM AND FUNDING FY 1977 FY 1978 FY 1979 l , .
! DUCT MELT THRU AND DROP OUT TESTS (DMFT) $270 $500 $700 DEPRE$$URIZATION A'CCIDENT (DAC) 11 0 110 , ; MAJOR ACTIVITIES (1977-1979) 19ZZ e DESIGN TEST FIXTURES. .
l . ,. e MATERIAL CHARACTERIZATION UNDER TEST CONDITIONS.
. j 19Z9 e EFFECT OF He PRESSURE AND NATURAL CONVECTION ON FUEL MELT-DOWN.
e CLADDING AND DUCT WALL MELT-DOWN BEHAVIOR.
- e MODIFY TEST FIXTURES FOR DEPRESSURIZATION ACCIDENT TESTS.
L9ZR e RUN'FULLSIZEDMFTANDDACTESTS.
~
r s l, ~ . .
/
i "~ .
'- r.
.- INEL SAFETY PROGRAM
- j. PROGRAM: IN-PILE TEST LOOP ENGINEERING (GRIST-2) 4 '
l1 - ll' FUNDING LEVELS: (ACTUAL AND PROPOSED) FY 76 i FY 77 FY 78 FY 79
$525K $500K $600K $620K
[; + . l MAJOR ACTIVITIES: (1977-1979)
- }
19ZZ o COMPLETE THE LISTING 0F GRIST-2. TEST-REQUIREMENTS
, o COMPLETE THE LISTING OF GRIST-2 HANDLING REQUIREMENTS o COMPLETE THE TEST LOOP CONCEPTUAL DESIGN REPORT-
[ - o COMPLETE THE GRIST-2 PROJECT MANAGEMENT PLAN l l 19ZR o BEGIN . PRELIMINARY (TITLE I) DESIGN OF. TEST LOOP . ! o BEGIN PREPARATION OF SDD'S OF MAJOR-SYSTEMS l; o PREPARE DRAFT PROCUREMENT PACKAGES OH LONG LEAD ITEMS
- o PERFORM PHYSICS. ANALYSES FOR EXPERIMENT DESIGN AND SAFETY ,
f , EVALUATION l 19Z9. o COMPLETE TITLE I WORK; PREPARE TITLE I COST ESTIMATE AND
-CONSTRUCTION SCHEDULE -
l t ! o BEGIN FINAL DESIGN (TITLE II) AND INITIATE PROCUREMENT l- . ACTIVITIES ON LONG LEAD ITEMS y[
~
o PERFORM VARIOUS SYSTEM CONTROL AND SAFETY ANALYSES l'- _
i GRIST-2 MILESTONES . 3 1 I TEST LOOP . I. COMPLETE CONCEPTUAL DESIGN SEPTEMBER 1977 r NOVEMBER 1980
- 2. COMPLETE FINAL DESIGN l
NOVEMBER 1981-i, 3. COMPLETE CONSTRUCTION
- 4. COMPLETEhCCEPTANCETESTS JANUARY 1982 .
}
HANDLING AND EXAMINATION EQUIPMENT SEPTEMBER 1977 j I. COMPLETE CONCEPTUAL DESIGNS ' NOVEMBER 1980 !
- 2. -
COMPLETE FINAL DESIGNS 1
-e NOVEMBER 1981-
- 3. C.0MPLETE CONSTRUCTION ,
e j , TEST OPERATIONS I. START TESTS APRIL 1982 ;
- 2. COMPLETE TEST PROGRAM SEPTEMBER 1984 A i
i APPR0XIMATE COSTS '
$27M-TEST LOOP CONSTRUCTION 2M l R&D HANDLING AND EXAMINATION EQUIPMENT 3M-f
_ff : TEST OPERATIONS ESTIMATED TOTAL $38M
-rg l
o q
i t i .
.i ! i
- ! GENERAL ATOMIC COMPANY SAFETY PROGRAMS 1
- j~
PROGRAM: GCFR SAFETY TEST PROGRAM SUPPORT l FUNDING LEVF1S: (ACTUAL AND PROPOSED) 1926 19ZZ 19ZB 19Z9. l $100K $125K $200K $220K MAJOR ACTIVITIES (1977-1979): 19ZZ o C0 ORDINATE THE GRIST-2 TEST AMONG GA, ANL AND EG&G.
~
~
. o DEFINE THE DMFT PROGRAM AT LASL.
i l . 19ZR o PROVIDE PRELIMINARY GRIST-2 TEST ASSEMBLY DESIGN. j o PLAN Tile DMFT AND DAC TESTS AT LASL. - l 1922 o DEVELOP DETAILED GRIST-2 fEST PROGRAM. ! o L AND INTERPRET DMFT AND DAC TEST RESULTS. ANALY~E l . e
- l,. ., .
W - s e 4/ 7 v y lj -
#c-77 GENERAL ATOMIC COMPANY SAFETY PROGRAMS T
i GCFR REACTOR SAFETY, ENVIRONMENTAL AND RISK ANALYSIS
- h. PROGRAM:
!I - l; FUNDING LEVELS: (ACTUAL AND PROPOSED)
.lSZE 19ZZ .ISZa 1912 .
1
~$205K $660K $580K $580K ll !
lI MAJOR ACTIVITIES (1977-1979): '
. i , - 19ZZ o OVERALL GCFR SAFETY PROGRAM PLAN. p I
o RELIABILITY ANALYSIS.0F THE DECAY HEAT REMOVAL SYSTEM. . j o ASSESSMENT OF POST-ACCIDENT FUEL CONTAINMENT (PAFC) WITHIN THE PCRV. -l f '1928 o PRELIMINARY ANALYSIS OF THE LOSS OF DECAY HEAT REMOVAL ACCIDENT. . o EVALU4TIONOFCORECATCHERCONCEPTS. - o ASSESSMENT OF PAFC EXTERNAL TO THE PCRV. l 1923 o IDENTIFY P0TENTIAL DESIGN IMPROVEMENTS BASED ON RISK ANALYSIS RESULTS. ; l' o IDENTIFY R&D PROGRAMS TO REDUCE RISK UNCERTAINTIES. l 0 DEFINE DESIGN REQUIREMENTS FOR PAFC. !
' ~.....' . . : .. . ;~. 7~:: ; _ . . . .,_.,; g_ = _
T
, . g .y ,.,
4 l
*dsEEEMERMERREMBRI
.ee.?
1 , , s. ~ INFCE
- it ,
2 . g
- 2. Enrichment Availability 3. Supply Assurances
!l 1. Fuel Availability
- Nuclear Energy
- Fuel Cycle Strategies .
.
- National Needs -
! u Demand
- Different Enrichment Consistent With '
3 I
- Resource Non-Proliferation -
e
- Technolog.ies .
, Requ..irements
- Incentives <
- Proliferation Risks
- Uranium Ava.i lability -
- Timely Deliveries
- Safeguards
- Heavy Water i i '
- Plutonium
! ;( Availability, A Fuel Cycle Centers . Creditor , ,
~
;
- Thorium Availability +LDC Needs Exchange . -l ,
h'th '
~
/ '+LDC Needs A Multinational
, l[f,;di . .
Arrangements ij A cmsscuts +LDC Needs ;.
> p. .
5
L-* . . a . ,T
- .-.. . . -- ars7ssr. ..
j ji , , f i 1
;g . .. . . _
3
; i '
Technical and. Economic
~
i 4 li ! i < !Ij$
"4 Reprocessing, PU Handling - 5. Fast Breeders lI j . .
and Recycle l
- Technological, Economic
- Technological Economic
- j. - Environmental and Energy
- Environmental arid Energy l-
"
- Safeguards
- Safeguards -
; ~
i A Fuel Cycle Centers .
- Reprocessing Modes i
i
^
- Disposal Strategies A-Fuel Cycle Centers 4 :
- Economics
+ LDC Needs '
- Plutonium Handling i
t .!i ' ~
- Intemational Control of Plutonium - -
!y l j
- Alternative Handling Methods ,
- Recycle in Thermal Reactors
} .
~
+ LDC Needs ,
;di ,
1
- , i .
6
- = -
g .. {
}4 .-
is fs j. n% y
.; SCOPE -
9
;l< u! $g ' 8. Advanced Fuel Cycle ; 6. Spent Fuel 7. Waste Management [g Management and Disposal . Concepts }
g( kh.g
~
- Storage Strategies
- Handling and
- Fuel. Utilization /
g -
- Costs Disposal '
Present Thermal Reactors kg
- Short Term / inter-
- Spent Fuel .
{j mediate S.torage .
- Repositories gugh Fuel
; A Institutional Safeguards
- Siting Problems Utilization
~
?
i,h and Safety
- Recovery Risks
- Research Reactors i Y Ii .-
e Legal Matters A Institutional and
- Thorium U-233
- Spent Fuel Storage Safety
- LW and Thorium f i
L] 7 [fj
Spent Fuel Capacity
~
- Legal Matters Breeder 'l k + LDC Needs +LDC Needs
- High Temperature I@'
P Breeder 4
+ i
- Other Advanced i Y
.j
~~ .
Concepts ! Wl tl
- Proliferation Risks I' '
- Commercialization . '
l
-
- Safety Problems
'i .
. 4LDC Needs == ~ a d{
'A > : ..
.ee.~ c., .c_ x - 2 7 3 q
,Q
~
- p. a ~. . . ,-c {
g- .=e.. .g1-,o - y O. .
- c. . o .. -
i y ..
- 1
- \
v .
. ~
1 1 . t
~ ~
';i l,1 ly INFCE Participation Working Groups 1-4
! lm! e. R.D. Nininger ' George Quinn L.T. Scheinman Charles Hebel
- < a, ti Alternate: Alternate: Attemate:
G.R. Bray #.A. Leary
- y J.L. Schwennesen W.R. Voigt, Jr. .
!}, S.G. Coord: S.G. Coord: S.G. Coord: S.G. Coord:
l J.A. Patterson W.R. Voigt, Jr. F.F. McGoldrick; G.R. Bray #.A. Leary l 3' i si ' F. F. McGoldrick f E.B. Steinberg M.' Rosenthal E.B. Steinberg l #i[ J.P. Boright R. Speir D.C. Thomas J.L. Schwennosen l($f E. Noble . E.B. Steinberg , G.G. Opfinger H.Rowen j A.W. Reynolds .R.E. Dierlam . T. Greenwood . H. Lowenberg i l R. Garvin J.R. Patton K. Larson R.E. Rosenthal ] j' D.M. Sikes D.M. Sikes T.H. Isaacs - R. Fuhrman y 1 W.C. Ramsay - G. Bunn H.B. Curtis ,
?
f
~ ^
R.J. Bettauer E. Perchonok S.M. Bologa J t
.' D. Ferguson
- i y,fj _.
~
- s. . -
31
5 ,%
~
k
?
On - F -
*f w INFCE Participation Working Groups 5-8
~
.f fk kg !'
M.J. Lawrence C.W. Kuhlman G.W. Cunningham R.G. Staker ,
'l r.1 '
36
~,j
~~
Alternate: Alternate: , d i' Alex F. Perge ._
~
i s S.G. Coord: S.G. Coord: S.G. Co~ord: .p ! s S.G. Coord: . S.T. Brewer ' M.J. Lawrence Alex F. Perge E.J. Hanrahan , {N
- J.P. Boright '
L.C. Simpkins C.R. Cooley M.K. Moss ; i . O.J. Sheaks J.P. Roberts G. Bowman K.M. Black - ! M.K. Moss S.N. Ceja L.C. Simpkins S. Strauch *
^
N F M.J. Lineberry R.B. Chitwood J. Martin C.N. Kelber ~
- j' F. Hoffman L R. Livingston B. Newmark D.T. Hafemeister f R.P. Denise B. Newmark R.J. Catlin A. Mowery e p V.W. Lowery ,
A. Platt D. Smith , R.W. Cochran '- E.A. Delaney R.J. Catlin ~~ '
?? W.O. Harms -
H. Lowenberg ' 'i$ ;) C.E. Till M. Stevenson - T.H. Isaacs F. von Hippel ' { L. e T.H. Isaacs
- K.D. Dance 3 H. Lowenberg -m.a ...
. $:.. A . 1 4 h *
- - = - - - - - - -
(
; .a .
l !k} 4 INFCE Crosscut Participation . i )' i, I . i! Proliferation !Fr Safeguards Institut.ional Technical -
; ; Res.istance Econom.ic ,
Samuel C.T. McDowell Robert Rochlin Philip Farley E. Hanrahan I i W. Hagis J. Boright H. Curtis i R.J. Catlin - H. Curtis E. Milenky G. Bray
; W.A. Higinbotham .
T. Greenwood [ 4 { L. Gallini A. Mowery .' G. Inman ] D. Hafemeister J. Glasgow J.Sheaks 4, > H. Kendrick H. Kendrick D. Hafemeister S. Brewer T. Nero S.C.T. McDowell L. King D. Mathes l ;h. '} M. Rosenthal T. Nero V. Lowery i*y S. L. Williams , R. Rosenthal F. McGoldrick . i L.F. Wirfs T. Sherr J. Patterson - l ;; B. Liimatainen R. Speir R. Rosenti al
, R. Simpkins' -
J. Schwennesen '.
. , E. Steinberg ,,,..
~'
- s
.g .
f i i
. . . .. - *=
(gS { a<
i -- . . . - . _ . . -. .-
- o. .
+ .. ;i .
t' l 5
'l~ j i,
. SELECTION OF NASAP WINNERS
- . '.s 3
JAN 1978 - l i
! ) APRIL 1978 -
i ( N ,
"k
- ! 1l OCT 1978
'l ; il l ' TECHNICAL ASSESSMENTS In EG4 3 JULY 1979; 4 i .
- Prcliferation Resistance 47EO i
'
- Technical Performance 4 ALL
- Economics of Power
- Safety Acceptability AsgESSep T- --
i . i Production
- Environmental Accepta-I -
ALTERNATIVE
- Resource utilization bility ,
SYSTEMS
- i (REACTOR AND .
ASSOCIATED NON-TECHNICAL ASSESSMENTS l l. ' FUEL CYCLES)
- Commercial Feasibility -
i
;'
- International Acceptability Analysis ,
- Institutional Analysis KEY OUTPUTS gzA ,
i , , -
'. 0 h First Screen of Systems -
i p ggO 3999 @
- a..
Preliminary Systems identified - ESC
- psO s Promising Alternative Systems identified f '
- i. " 4 Recommended Strategy Options r
t *
- I.
44
t
~
KEY DECISIONS OR lSSUES 1 2 U.S. R&D PROGRAM q . SCOPE OF U.S. BREEDER PROGRAM
- NEED FOR AN ADVANCED CONVERTER .
t REPROCESSING TECHNOLOGIES :
!' USE OF BARNWELL 7 .
NEED FOR THORIUM CYCLE DEVELOPMENT. . SPIKING /Pu IRRADIATION , i SAFEGUARDS TECHNOLOGY ' DOMESTIC . ENRICHMENT EXPANSION EXPORT / IMPORT OF URANIUM . REPROCESSING AND RECYCLE b . i
~ . .y s
11 l'~ - --
.._.~
l . -
. KEYDECISIONSORISSUES(CONTINUEDt .
INTERNATIONAL i i .
~
i' TECHNOLOGIES IN USE OR IN DEVELOPMENT ,
- . , REACTORS - ENRICHMENT - .
p FUEL CYCLES . REPROCESSING ~ !
~
~
SPENT FUEL STORAGE [ . , .. FUEL ASSURANCES .
~
i URANIUM REPROCESSING .. i
. a.
ENRICHMENT WASTE MANAGEMENT i.. ;
~
i - l FUEL CYCLE CENTERS 0R OTHER ARRANGEMENTS 'i.! ! LEVEL OF SAFEGUARDS RULES OF TRADE . i, . i i s. f . .5- [ a n .
- t .
3 db
~
- . - ~ . . . _ _ _ _ _ ._. . - _ _ _ _ _ . _ - _
REACTOR SAFETY BUDGETS BY B&R CATEGORY- ,
' ~ ~
', (in Thousands)
FY 78 FY 79 . LOA - I PREVENT ACCIDENTS 1,960 3,639 . a
~
~
LOA-2 LIMIT CORE DAMAGE 23,562 16,026
^ -
LOA-3 CONTROL ACCIDENT PROGRESSION 9, 13 9 7, 10 5 LOA-4 ATTENUATE ACCIDENT CONSEQUENCES I,749 1, 401
. LOA SUPPORT AND INTEGRATION 5, 13 9 6,333 FACILITY DEVELOPMENT 10,278 -
6 11 OPERATING BUDGET (BO) 51,827 3 5,115 s- +
- FACILITY DEVELOPMENT 4 CONSTRUCTION BUDGET (BA) -
15,550 9,000
- O s t
- R
- n.
l
~ ~~
L"
~
ACCIDENT PREVENTION FUNDING (in Thousands) -
~
FY 78 FY 79
~ "
. .. REllABILITY TESTING I,670 3, 314 -
LOCAL FAULT PREVENTION 290 325 - OPERATING FUND, 1NG (80) I, %0 3,639 - l
~
.g I
j =. i 1 I
~
, s A, . .
CORE DA. MAGE LIMITATION F'UNDING (in Thousands)
. . i li .. n i . t i , i n g; i no ,1
- sii, s i n.. .- : io . ) FY 78 FY 79
- 8,901 7,421 5
,- . TOP Wl0 SCRAM ACCOMMODAT10N ,, : . . 5,880
~
' TUC WIO SCRAM ACCO,MMODKIONII' 9,!857.t! , lt
~
LOSS-OF-SHRS $C,CO,MM0y,AT1,0{,;,.g ,,,;; 151 150 ; LOPi- ACCOMMODATION, , ,, ,,3, , i p , , ,, , , , ; , ,, i, 518 ,, ;, (550 i LOCAL FAULT ACCOMMODATION 3 , 13 5 1,025 i .
~
OPERATING FUNDING (BO) 23,562 16,026 -
- 't
- Does not include $2M which will be contritiuted by the Japanese upon completion of ongoing negotiations i
- Does not include $iM from Japanese l.' -
l . ., n -
~
. f ACCIDENT PROGRESSION CONTROL FUNDING
~
. (in Thousands) !
FY 78
~
FY 79 ENERGEriCS ACCOMMODATION 4,515 3, 510 ~,
' CORE DEBRIS ACCOMMODATION 2,580 2,005 .
~
j . DESIGN OPTION EVALUATION 2,044 . 1,590 -.. I OPERATING FUNDING (BO) 9, 13 9 7,105 Y e 1 4 . 4 .
- d S i )
g 4
- I 1
a., 1
.J ACCIDENT CONSEQUENCES ATTENUATION FUNDING
- (in Thousands) .
i . [ ^ FY 78 FY 79 ATTENUATION INSIDE FAULTED CONTAINMENT 437 350 . ATTENUATION IN ENVIRONMENT I,175 941 4 . . ATTENUATION MECHANISM EVALUATION 13 7 11 0
~
. OPERATING BUDGET (80) I,749 1, 401 lL' .
.s
'i
~
O.. M
]
. - y FACILITY DEVELOPMENT FUNDING
- (In Thousands)
FY 78 FY 79 . OPERATING FUNDS ,
' ' ~ 6,800 202 4
SAREF FACILITIES ' 500 EBR-il SRM
+ SLSF SECONDARY COOLANT UPGRADE 400 ,
1,978 - SAFE SHUTDOWN TEST FACILITIES ..
~
600 409 FUEL MOTION MONITORING SYSTEMS ~: - OPERATING FUNDING (BO) 10, 278 611
~
l CONSTRUCTION. FUNDS 1 . , SAREF I4,050 .. 4',000 EBR-Il SRM I, 500 5,000 ! CONSTRUCTION FUNDING (BA) 15, 550 9,000 -
- j. .
r
LOA l---PREVENT ACCIDENTS . s 9
- LOA I .
4
~
ACTIVITY CONTRACTOR $ - ARD 1564
-CONTROL ROD REliABILifY TESTING l ' -
t GE . 1750 i t ARD 100 ! -LOCAL FAULT PREVENTION
- i. ~
ANL .. 225 . i
+
- -N -
l , i rl
LOA 2---LIMIT CORE DAMAGE . LOA 2 1
~ LOPl , ,,,,, ------$1550K ACTIVITI l CONTRACTOR
- i
-SLSF LOPl EXPERIMENT ii.
GE----TEST DESIGN- AND ANALYS IS HEDL--TEST TRAIN FAB. .. j i aiit iu 6. . .li i.il Ir nin ti 4 : ii.
--CONDUCT AND EXAM TEST
--FUEL FAB. i.i.
EG&G-ASSEMBLY TEST- ~
~
t -ETR OPERATION n.. l.. .. i .ii i . i i s i . ;i i. t i '5- ANL---PARTIAL
=
PIE
,.i
-B0ILING STABILITY EXPT. IN THORS ORNL l
4 T 4
~
I __
~
O. .. 4Y
r
. F.
t 4 LOA 2---LIMIT CORE DAMAGE . LOA 2 . LOPl ---- ------ - - -- --- -- ----- -- - - - - - --- -- $ 1550 K ACTIVITY- . CONTRACTOR . Sl'SF LOPl EXPERIMENT GE----TEST DESIGN AND ANALYSIS HEDL--TEST TRAIN FAB. ..
--CONDUCT AND EXAM TEST j
--FUEL FAB. l EG&G-ASSEMBLY TEST '
' l
! -ETR OPERATION + - ANL---PART,lAL PIE ,
-B0ILING STABILITY EXPr. IN THORS ORNL I
4
. l
. j
f j . .
. LOA- 2 -- LIMIT CORE DAMAGE ..
LOA 2 , 1
-TOP ------------------------------------- $ 742 i K . , j ACTIVITY CONTRACTOR GE------CAPSULE FAB,
-CAPSULE TESTS IN TREAT 31 LURE TIMEILOCATION i ON FULLLENGTH FUEL IRR. IN FAST REACTOR ------TEST CONDUCT AND . ANAL - }
l (PFR) HEDL----CAPSULE FAB. :
----TEST CONDUCT AND- ANAL
----CODE DEVELOPMENT i
F -CAPSULE TESTS IN TREAT INTERNAL FUEL MOTION HEDL- --CONDUCT TEST AND ANAL. LASL----DEVELOPMENT OF NEW FUEL l ' MOTION VIEWING SYSTEM L -
-LOOP TESTS IN S'LSF ON SLOW RAMP -RATE HEDL----TEST TRAIN FAB.
L
----TEST CONDUCT AND ANAL.-
- l. TRANSlENTS i EG&G ---LOOP ASSEMBLY I
---ETR OPERATION .
- i. -
GE------FUEL FAB. [ ' HEDL i
-DEVELOPMENT OF ANALYTICAL CODES FOR USE IN i ACCIDENT ANALYSIS (MELT)
ANL
-0UT-0F-REACTOR TESTING TO DETERMINE Y
BEHAVIOR OF COOLANT, CLAD,. AND FUEL IN ,
- ~
! . ACCIDENT CONDITIONS 9
~
LOA 2---LIMIT CORE DAMAGE i LOA 2 . j
-TUC -- - -------- -- - - - - - - ----- -- - - - - -$ 7880 K ,
ACTIVITY CONTRACTOR . l I
- TREAT LOOP TESTS ON FUEL DlSPERSAL ANL----TEST TRAIN FAB. PRETEST
~
AND POSTTEST ' ANAL AND CONDUCT EXP. .
~
j - SLSF LOOP TEST ON EARLY FUEL MOTION ANL----POSTTEST EXAM AND REP 0Ri
- SLSF TEST ON FUEL BEHAVIDR WITH EXTENDED ANL----TEST DES IGN . AND ANAL ~
FUEL MOTION TEST TRAIN FAB. . i GE-----FUEL FAB. .
- 0VT-0F-REACTOR EXPf'S. ON ROLE OF FISSION ANL----DEH EXPI'S l HEDL---INDUCTION HEATING EXPT'S l GAS IN FUEL DISPERSAL
~
PROOF OF PRINCIPLE TESTS l- - INHERENTLY SAFE CORE DESIGN GE ARD AT ENERGY TECH. ENGINEER-l , Al ING CENTER
-PREPARATION FOR LOOP TESTS IN TREAT AS ANL----DES IGN AND FAB. NEW j
PART OF USIUK PROG. MARK.3 LOOP ! HEDL---FAB. FUEL FOR 1RR. IN PFR t I-
- ANALYTICAL CODE DEVELOP (SAS) ANL
- OUT-0F-REACTOR- EXPI'S .0N VOIDING BEHAVIOR ANL m i
~
ORNL-
~
l n
1 t LOA 3---CONTROL ACClDENT PROGRESS 10N
- LOA-3 4
ENERGETICS ---- ---------------- ----$3 510 K
- l. ACCOMMODATION ACTIVITY CONTRACTOR TASK q ..
-PROBABILISTIC ~ ASSESSMENT ANL CODE DEVELOPMENT AND PHENOMIN0 LOGICAL
- . OF ENERGETICS SOURCES .
TESTING ' ~
. HEDL TRANSITION PHASE MODELING ,
ORNL lMPROVED NEUTRONICS ANALYSIS L '
-WORK-ENERGY ATTENUATION LASL ATTENUATION MECHANISM
" MODELING
- i. '
' -SR1 PHENOMIN0 LOGICAL TESTING -
i I
-CONTAINMENT R5PONSE .
ANL PRIMARY SYSTEM RESPONSE CODE DEVELOPMENT
^
Al SODIUM FIRE ANALYSIS AND TESTING ' ~ SAI INTERNATIONAL CODEm, e .- - BENCHMARKING
- 1 r
' - sD L __ _ __ . . _ . .
i
- LOA 3---CONTROL ACCIDENT PROGRESS 10N
~
LOA 3 i CORE DEBRIS F . ACCOMMODATION -- ----------------- ----- --$2005 K ACflVITY CONTRACTOR TASK . 1
-lN-VESSEL ACCOMMODATION '
. ANL PARTICULATE BED C00 LABILITY TESTING -
l GE ANALYSIS OF INHERENT ! - CAPABILITIES FOR
- - IN-VESSEL C00 LABILITY
-EX-VESSEL ACCOMMODATION ANL SMALL-SCALE TESTING OF-l.- ,
EX-VESSEL MATERIAL t' .JNTERACTIONS l - Al, - LARGE-SCALE TESTING OF , l HEDL- EX-VESSEL MATERIAL l INTERACTIONS.' i-GE ANALYSIS OF INHERENT ! . CAPABILITIES FOR EX-VESSEL C00 LABILITY 2 - O. .s w -
t
= l ].- LOA 3---CONTROL ACCIDENT PROGRESSION 4
LOA 3 . DESIGN OPfl0N -------------------------$ 1590K EVALUATION 1 l . ACTIVITY CONTRACTOR TASK .
-CONTAINMENT AIR CLEANING HEDL SYSTEM ANALYSIS AND TESTING
-ENERGETICS MITIGATION ANL ENERGY ABSORFil0N L6 - SYSTEM ANALYSIS
~
( ! COST-BENEFIT ANALYSIS ~ GE, HEDL DESIGN OPTION ANALYSIS AND RANKINGr
. . 2.y
.- n.,
l :. .
~
* .~..
l o 5 _ _ _ _ _
d + i LOA 4---ATTENUATE ACCIDENT CONSEQUENCES ' LOA 4 ATTENUATION ~
-- ---~---------------------$3 50 K IN.- FAULTED .
. CONTAINMENT ,
CONTRACTOR TASK ! ACTIVITY i -ENERGETIC RELEASES TO LASL ANALYSIS OF ENERGErlC CONTAINMENT CORE MATERIAL TRANSPORT . \ AI HIGH DENSITY AEROSOL
- t. -lNHERENT ATTENUATION IN CONTAINMENT . ' TESTING AND MODELING 9 1
) $ e O
./,7 '
- A 4
3. [ i
,. i , .i
~
LOA 4---ATTENUATE ACCIDENT CONSEQUENCES 9 , a* 1 ;< LOA 4 - I l - , , i, . b ATTENUATION i IN . ,----------------------------$941K . ENVIRONMENT . i TASK
! ACTIVITY CONTRACTOR
, it
-RADI0 ACTIVITY TRANS PORT ORNL HEALTH EFFECTS MODELING i l'"Id k AND EFFECTS IN THE i .' i ' ' ' 'l
- n. ENVIRONMENT i ii . :o . t liis
.l. . ii . i .it.itii!i I ' !'b' '
13 . A1, GE EX-CONTAINMENT AEROSOL , DEPLET10N MODELING 4 4 e e ~ A,
4
~
LOA 4---ATTENUATE ACCIDENT CONSEQUENCES - - LOA 4 4 AT.TENUATION ,
- - 1 MECHANISM 1 10 K ,
' ~' EVALUATION . , i . > ACTIVITY CONTRACTOR TASK COST-BENEFIT ANALYSIS HEDL PARAMETRIC AND - ' SENSITIVITY ANALYSIS 10F ATIENUATION [ MECHANISMS i . 1 i 5 .. l--
^
. 4 i
~
LOA-2 ACHIEVEMENT OF LOA-2 (TOP /W/0 SCRAM ACCOMMODATION)WITH 99% CONFIDENCE s e SHUTDOWN FROM FUEL REMOVAL. AMOUNT OF FUEL RELOCATED DISTRIBUTION OF FUEL RELOCATED e THE CORE AFTER T0P/W/0 SCRAM IS C00LABLE FAILURE LOCATION , FAILURE TIME FUEL EXPULSION RATE FCI ENERGY. CONVERSION EFFICIENCY POST TRANSIENT COOLANT FLOW 9 4 9 e
TOP TEST MATRIX PILOT STUDY
- e. PLANNED SERIES OF TRANSIENT OVERPOWER (TOP) TESTS
' TO DEMONSTRATE ACCOMMODATION OF TOP W/0 SCRAM.
e 141 TESTS IN SERIES e 25 CAPSULE TESTS 4 e 16 LOOP TESTS , e INDEPENDENT VARIAB,LES: o PIN STEADY STATE POWER e PIN BURNUP , e TRANSIENT RAMP RATE e . TEST MATRIX COVERS WIDE RANGES OF THE INDEPENDENT VARIABLES. 4 e e
. l OG
, ; ^
j , 4 i PRINCIPLE EXAMPLE .i j -e SAFETY R&D PROGRAMS DIRECTED TOWARD e SHOW THAT TOP W/0 SCRAM WILL TERMINATE ' QUANTITATIVE PROBABILITY GOALS WITH CORE PARTIALLY DAMAGED, BUT
; C00LABLE IN-PLACE WITH 99% PROBABILITY,
- . , (PROPOSED TOP BEST MATRIX (LOA-2) e ACHIEVEMENT OF G0ALS REQUIRES ACCEPTANCE OF
e SIMPLEMODELSTHATREf_ATEACHIEVEMENT e " SUCCESS" MODEL FOR LOA-2 TOP EVENT WILL SHOW WITH 99% PROBABILITY i
! 0F G0ALS TO THE KEY PHENOMENA i
SHUTDOWN FROM FUEL REMOVAL N0 AUT0 CATALYSIS IN-PLACE C00 LABILITY OF DAMAGED CORE e TEST DATA THAT ESTABLISH REQUIRED e TOP DATA REQUIRED.
- ' PROBABILITY LEVELS FOR KEY PHENOMENA -
99% LOWER CONFIDENCE B0UND ON - FAILURE LOCATION . 1 i ! 99% UPPER CONFIDENCE B0UND ON FCI EFFICIENCY l: i - 99% LOWER CONFIDENCE BOUNDS ON FUEL REMOVED FROM CORE REGION 99% UPPER CONFIDENCE BOUND ON FUEL EJECTED AND REMAINING IN ACTIVE I CORE REGION I e- 'i. . 1.
..g. e+w. ,.
..-p v
=Wh =***- 9 i ,
i %'
. w
; PRINCIPLE . EXAMPLE l
e~ "GEDANKEN" EXPERIMENTS ASSESS THE G'EDANKEN EXPERIMENT FOR TOP TEST MATRIX [! TECHNICAL RISK OF PROPOSED TESTS IS ONG0ING ,i AND PROVIDE GUIDANCE FOR IMPROVE-l' MENT OF SUCCESS PROBABILITY RESULTS S0 FAR INDICATE: 1 - DESIRABILITY OF IRRADIATION OF AN -, i! ALTERNATE FUEL PIN DESIGN (THIS HAS i! BEEN DONE) l: IMPROVED MEASUREMENT TECHNIQUES ,-; (E.G., FAILURE LOCATION) SHOULD BE INVESTIGATED 1
, SOME VARIATIONS OF THE SUCCESS MODEL
.l -
MAY YIELD BETTER STATISTICAL RESULTS 3 .. 4 . . I i l. !.~ i
- j. -
j f' -
- "GEDANKEN" EXPERIMENT
^ '
TOP TEST MATRIX llq CONSULTATIONS WITH EXPERIMENT i l , DESIGNERS, AT, ANL, HEDL a 3 i! MEASUREMENT f k UNCERTAINTIES l,l
'~
g ' FAILURE TIME il -
& LOCATION t
; I i .
l NdMINAL-(MEDIAN) I
~ ' NOMINAL PREDICTIONS OF
- (MEDIAN) FUEL REMOVAL PREDICTIONS OF AND VOIDING FAILURE HISTORIES i
4 j BEHAVE /SST CODE -) LOCATIONS
- PLUTO CODE i
DISTRIEUTION OF EXPERIMENTAL < EXPERIMENTAL VARIATION OF l VARIABILITY OF FUEL REMOVAL FAILURE AND VOIDING ANALYST LOCATION HISTORIES " JUDGMENT ANALYST l
.VvvvVMONTE CARLO SIMULATION
? . - l
. ._ . ; _ . . . . _ .1. .. - - -- T
. SIMULATION -
RESULTS UNCERTAINTY DISTRIBUTIONS . (ONE SET PER TEST):
, PROB y ONE SAMPLE PER TEST INPUT G 90% TO REGRESSION MODEL:
. S o 70% ~~.,
4
., x p
U 50% '9 REGRESSION E L 20% goggt IME- f lIr> 1 _Pa0s f 90% / i v i g 0 f 70%~ -- i 1 y v I CONFIDENCE o 50% f BOUNDS L f ' l 20% I s TIME 1' I . l P . ! R O j # B i 1 T - H A T . L 0 c o.
< LOCATION
-k h o
,-.no p.,c . - , , , w-,-
-,--....- - , + - , , , , ----,...n.--,,,gn,,,m -
,ws,, --.,..n.,,c-n-..nn--,-,,l-
~ - -
. . . . . . . . j
~
- i !
\
\
./\ .
.! SOURCES OF EXPERIMENTAL VARIABILITY
! e FAILURE LOCATION ;
e CONTROL OF RAMP RATE
](
e CONTROL OF PRE-IRRADIATION POWER LEVEL j
- e FUEL-CLAD GAP CONDUCTANCE I -
e FUEL STRENGTH . e CLADDING STRENGTH l e FUEL THERMAL EXPANSION I e FUEL PELLET DENSITY ~ , e FUEL REMOVAL AND VOIDING -
- , e FRAGMENTED FUEL PARTICLE SIZE I-e TEST OVERDRIVE -
l 1 i
~
_ _ _ _ . _ _ - _ . . _ _ . . _ _ . . . _ _ _ _ . . . . _ _ = _ _ _ . _ _ _ _ . _ _ _ _ _ _ . ,_____2
j .- !
. i 1, ,>
qv: i ti . s
' QUANTIFICATION 0F-VARIABILITY .
1; 4 ii I;' l
- e DERIVATIVES IN THE VARIABILITY MODEL WERE
- ESTIMATED BY COMBINED JUDGMENT OF ANALYSTS l AND SENSITIVITY STUDIES l V
~..
-e PROBABILITY DISTRIBUTIONS OF THE~ VARIABILITY .,
I SOURCE WERE ESTIMATED BY JUDGMENT OF THE l ANALYSTS , i
# 4 T-e I' -
i
!_- t
y 6 ep = *eww--- - ww w . mgge ' j. ' [ MEASUREMENT-UNCERTAINTIES OBJECTIVE l e REPRESENT THE VARIATIONS IN MEASUREMENT PROCESSES WHICH CAUSE RAND 0M
- i DIFFERENCES BETWEEN THE MEASURED TEST OUTCOMES AND THE ACTUAL OUTCOMES.
~
l FAILURE LOCATION e METHODS , e SEQUENCE OF THERM 0 COUPLE RESPONSES \ ! e POST TEST EXAMINATION e MAJOR UNCERTAINTIES .l 'l e POST TEST EXAMINATION NOT ALWAYS USEFUL (MULTIPLE FAILURE; EXTENSIVE RIP) e POOR RESOLUTION DUE TO TC SPACING FUEL REMOVAL ~ e METHODS l' , e H0DOSCOPE ~
- e MAJOR UNCERTAINTIES ,
e INSENSITIVE TO SMALL MASS CHANGES e MULTIPLE PIN FAILURES MAY CONF 0UND RESULTS
. VOID VOLUME ,
l e METHOD l e INTEGRATION OF DIFFERENCE BETWEEN INLET AND OUTLET FLOW METERS e MAJOR UNCERTAINTIES
~
- e DRIFT IN FLOW METER SENSITIVITIES l .
i AV LESSONS AND-CONCLUSIONS l , i i e THE "GEDANKEN" EXPERIMENT (ANALYSIS IS ONGOING) HAS PROVIDED INSIGHTS USEFUL T0 i l TEST DESIGN. t i [, e BASED ON EARLY RESULTS, ADDITIONAL PIN DESIGN WILL BE IRRADIATED (INCREASES CHANCE OF MEETING LOA-2 SUCCESS CRITERIA). ! -e FIRST ITERATION REGRESSION ANALYSIS SHOWS LOA-2 G0AL WOULD NOT BE MET (99% CONFIDENCE ! INTERVALS TOO WIDE). 4 i !j e SENSITIVITY ANALYSIS WITH MONTE CARLO SIMULATION POINTS OUT MAJOR CONTRIBUTORS TO . 1 i EXPERIMENTAL VARIABILITY: FUEL-CLAD GAP VARIABILITY (I.E., FABRICATION TOLERANCE). . e CHANGES IN MODEL AND MEASURED " PRODUCTS" ARE UNDER CONSIDERATION: l VOIDING MORE EASILY DETERMINED THAN FCI EFFICIENCY, j - BENEFICIAL TO MEASURE FUEL REMAINING IN COOLANT CHANNEL j - IN ADDITION TO THAT REMOVED FROM CORE, l - MEASUREMENTS OF FAILURE TIME AND FUEL EJECTION RATE - l - ORIGINALLY-PLANNED - NOT NEEDED, 1 SIMPLISTIC AUT0 CATALYSIS MODEL (BASED ON FAILURE LOCATION)
- l. MAY NOT BE ADEQUATE. .
i-
- i. . - --
SAS3A',' SAS3D', SAS4Ai MELT-IIIA SAS3A, SAS3D', SAS4A',' MELT-IIIA ACCIDENT INITIATION EARLY TERMINAT' ION INTAbTSUBASSEMBLIES IN-PLACE COOLING l~~ ~ ~ ~ ~ ~ ' ' ~ ~ ~ - I CORE DISRUPTION PHASE
--i I
I
. I I TRANSIT ' I
.I I
i . TRANSITION PHASE t 1 I .' ' CORE DISRUPTTON g i IVENUS-II i ' I .. i FX2/ VENUS-III TRANSIT I I SIMMER SIMMER ~ I I I MECHANICAL DISASSEMBLY GRADUAL FUEL REl10 VAL i I '
;. ENERGETibDI__SPERSAL Meth-OuT, BOIL-0UT g L _. _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . , _ . _ . _ , __, ,_' ._ ._ _ _ a .
i RFxCO ICEC0 ' DAMAGE EVALUATION WORK-ENERGY, SYSTEM-RESPONSE . POSTACCIDENT HEAT REMOVAL COMPREHENSIVE MECHANISTIC ACCIDENT ANALYSIS -
, PATH STRUCTURE AND COMPUTER CO, DES PHICH ANALYZE THE VARIOUS PHASES '
~
1('
(
, i DOE /RRT - SPONSORED CODES WHICH ANALYZE THESE ACCIDENT PHASE.S .
e TRANSITION PHASE , TRANSIT-II AND TRANSIT-HYDRO FUM0 ( ggod e GRADUAL FUEL REMOVAL PHASE
- TRANSIT CODES k O USEFUL /* .
)
e MECHANICAL DISASSEMBLY FX2/ VENUS-III A# . USEFUL F# . ICECO-II/ STRAW-WHAM 8 DAMAGE EVALUATION REXC0 c0 des 6# ICECO-I.I ICECO-IUSTRAW-WHAM . e y I
KEY PHENOMENA IN, CORE DISRUPTION AND. DAMAGE EVALUATION PHASES _ e TRANSITION PHASE HYDRODYNAMIC BEHAVIOR OF BOILING FUEL / STEEL MIXTURE IN DISRUPTED SUBASSEMBLIES AXI AL BLOCKAGE FORMATION AND REMOVAL IN DISRUPTED S.UBASSEMBLIES , ! CRUST FORMATION AND REMOVAL ON HEXCAN WALLS AND MOLTEN / BOILING REGION BOUNDARIES . FAILURE OF *tlYACT HEXCAN WALLS FROM THERMAL AND MECHANICAL LOADS RESPONSE OF MOLTEN / BOILING REGIONS TO MATERIAL RE-ENTRY PROPAGATION OF MOLTEN / BOILING REGIONS INTO INTACT SUBASSEMBL:ES
~
e GRArJAL FUEL REMOVAL PHASE ULTIMATE REMOVAL OF AXI AL BLOCKAGES IN FUEL SUBASSEMBLIES BLOWDOWN OF BOILING FUEL / STEEL MIXTURE THROUGH ESCAPE PATHS TO Pl.ENA ,. , OTHER FUEL REMOVAL MECHANISMS . n1[
.I
~
, KEY PHENOMENA - CONT,'D 8 MECHANICAL DISASSEMBLY PHASE, HYDRODYNAMIC BEHAVIOR OF MOLTEN / VAPORIZED CORE MATERIAL (INCLUDING SELF-MIXING AND LIQUID-VAPOR SLIP)
NNERGY.ANDMOMENTUMINTERCHANGEBETWEEN FUEL AND NON-FUEL CORE COMPONENTS (STEEL AND SODIUM) - 4 INTERACTION BETWEEN' MOLTEN / VAPORIZED CORE MATERIAL AND REMAINING STRUCTURAL COMPONENTS IN CORE - EXTENDED MOTION OF LIQUID / VAPORIZED CORE , MATERIAL THROUGH UPPER SUBASSEMBLY INTERNALS 8 DAMAGE EVALUATION PHASE , THERMAL INTERACTION OF MOLTEN / VAPORIZED CORE MATERIAL WITH UPPER SUBASSEMBLY INTERNALS MECHANICAL RESPONSE OF UPPER SUBASSEMBLY INTERNALS AND UPPER CORE SUPPORT STRU',;TURE TO LOADINGS FROM' DISASSEMBLING CORE l THERMAL INTERACTION BETWEEN MOLTEt' .<IZED
, CORE MATERIAL AND SODIUM IN UPPER PLdhuM
' ~
~-
MOMENTUM' TRANSFER FROM EXPANDING MOLTEN / VAPORIZED CORE MATERIAL.TO SODIUM IN. UPPER PLENUM ,
- STRUCTURAL RESPONSE OF VESSEL, HEAD, AND VESSEL INTERNALS TO APPLIED LOADINGS , ,
y g;
COMPARISON F TRANSIT-HYDR 0 AND SIMMER CAPABILITIES ' FOR TRANSITION PHASE ANALYSIS MODEL AaE&_ TRANSIT-HYDRO SIMMER
. GEOMETRY ,
- 3-D HEXAGONAL-Z 2-D CYLINDRICAL R-Z-SUBASSEMBLY EACH REPRESENTED RING OF SUBASSEMBLIES REPRESENTATION INDIVIDUALLY WITH REPRESENTED AS AN CAN WALLS AND GAP ANNULUS
~
STRUCTURE EACH HEXCAN WALL PASSIVE STRUCTURE " FIELD" REPRESENTATION MODELED SEPARATELY HEXCAN WALL FAILURE' BASED ON FA LURE BASED ON FAILURE MODELING COMBINED THERMAL / THERMAL LOADING MECH,ANICAL' LOADING , HYDRODYNAMICS 3-FLUID MODEL (FUEL 2-FLUID MODEL (VAPOR, ! i MODELING LIQUID, STEEL LIQUID) 4 STRUCTURE
. LIQUID, VAPOR)
DETAILED SOLUTION DETAILED 2-D SOLUTION OF IN AXIAL DIRECTION COMPRESSIBLE HYDRODYNAMICS EQUATION APPROXIMATE TREAT-MENT OF RADIAL MOTION t . 1 J ..
. J g
/)
l . .
COMPARISON OF TRANSIT-HYDRO AND SIMMER CAPABILITIES - FOR TRANSITION PHASE ANALYSIS CONT'D . f.0 DEL AREA TRANSIT-HYDRO SIMMER OTHER COMMENTS DETAILED TREATMENT 0F BLOCKAGE FORMATION AND REMOVAL AND CRUST - FORMATION AND REMOVAL. MAIN FOCUS DETAI' LED ANALYSIS OF DETAILED HYDRODYNAMIC T.HERMAL AND STRUCTURAL ANALY.Sil .QE MOLTEN ASPECTS WHICH CONTROL REGION RATE OF MOLTEN REGION PROPAGATION AND ULTIMATE FUEL REMOVAL 0 0 9 0 9
+ demums Sea B
6 8 e
f.. TRANSIT CODES , e TRANSIT-II AND TRANSIT-HYDRO ARE DESIGNED TO DESCRIBE THE TRANSITION PHASE OF A CDA IN AN LMFBR FROM THE TIME THAT EXTENSIVE WITHIN-SUBASSEMBLY MATERIAL MOTION BEGINS UNTIL SUFFICIENT FUEL IS REMOVED TO RENDER THE CORE PERMANENTLY SUBCRITICAL. e THE TWO CODES. DIFFER ONLY IN THAT TRANSIT-HYDRO HAS A MORE SOPHISTICATED MOLTEN / BOILING MATERIAL MOTION MODEL. c AXIALLY, THE CODES MODEL FIVE REGIONS: SHIELD ORIFICE BLOCK AND LOWER REFLECTOR, LOWER AXIAL BLANKET, CORE, UPPER AXI AL BLANKET, AND UPPER STRUCTURE. e RADIALLY, EVERY SUBASSEMBLY IS MODELED, INCLUDING INDIVIDUAL HEXCAN WALLS AND THE' GAPS BETWEEN HEXCANS.
. e ALLOWANCE IS MADE FOR THE USUAL CORE SYMMETRIES.
a e me e e G W' . , e- - , - m-v w ,m-~ v n w W( .
... i. .
MODELS.0F SUBASSEMBLY INTERNAL CONDITIONS PRIOR TO COMPLETE MELTDOWN e INTERNAL CONDITION OF INTACT SUBASSEMBLIES FOLLOW COOLANT UP TO VOIDING - INCLUD,E PIN?TO-WALL RADIATION ALLOW CLAD TO MOVE PRIOR TO PELLET MELTING e CONTROL AND SAFETY SUBASSEMBLIES
. FOLLOW COOLANT UP TO VOIDING DO NOT MODEL CONTROL RODS THEMSELVES e ATTACK ON INTACT SUBASSEMBLY FROM ADJACENT BOILING REGIONS MODEL ATTACK OF MOLTEN FUEL / STEEL MIXTURE ON INTACT PINS , ,
e S 4 W 9 s
r ,
. .. e, .
, ., o
) .
PHYSICAL BOUNDA, RIES MODELS e SUBASSEMBLY WALL AND CRUST MODELS-LINEAR TEMPERATURE PROFILE IN SUBASSEMBLY WALLS FOLLOW GAP SIZES AND COOLANT TEMPERATURES ALLOW dRUSTS TO FORM AND CALCULATE THICKNESS 6 e MECHANICAL FAILURE OF DUCT WALLS S e BLOCKAGE FORMATION BASED ON ABLATION MODEL AXIAL TEMPER'ATURE PROFILE IN MOVING SLUG 9 BLOCKAGE MELTING AXIAL TEMPERATURE PROFILE IN BLOCKAGE RADIAL ANb AXIAL HEAT TRANSFER FROM BLOCKAGE BLOCKAGE CAN FALL INTO CORE REGION IF IT . MELTS OUT e
MOLTEN / BOILING MATERIALS MOTION MODEL e TRANSIT-II USES A SIMPLE RELAXATION MODEL FOR MATERIAL
~
MIXINGJ ASSUMES CONTINUOUS BOILED-UP STATE e- TRANSIT-HYDRO UTILIZES A MULTI-COMPONENT TWO-PHASE HYDRODYNAMICS-MODEL FOR TREATING MOLTEN / BOILING REGIONS e GEOMETRY IS HEXAGONAL-Z (ORIGINAL POSITIONS OF HEXCAN I WALLS FORM MATHEMATICAL CELL. BOUNDARIES) o FUEL AND STEEL VAPOR ASSUMED TO MOVE IN EACH CELL WITH SAME AXIAL VELOCITY
~ T.
9 LIQUID STEEL AND LIQUID FUEL COMPONENTS EACH HAVE THEIR OWN AXIAL VELOCITIES o RADIAL MOTION FROM CELL TO CELL CAN OCCUR WHEN HEXCAN WALLS SEPARATING THEM HAVE MELTED AWAY. HERE, THE GOVERNING EQUATIONS OF MOTIONS ASSUME THAT [4 AVERAGE VELOCITY CALCULATED ON THE BASIS OF PRESSURE DIFFERENCES IS ADEQUATE TO DESCRIBE RADIAL MOTION OF ALL MATERIALS. l . e a 9
* >/ *
- h -(
l -
8,* ', e n D, EXCHANGE MECHANISMS IN TRANSIT-HYDRO HYDRODYNAMICS MODEL , 0 MOMENTUM TRANSFER SLIP BETWEEN LIQUID STEEL AND VAPOR AND BETWEEN LIQUID FUEL AND VAPOR SLIP BETWEEN LIQUID FUEL AND LIQUID STEEL LIQUID AND VAPOR COMPONENTS EXPERIENCE DRAG ON REMAINING SUBASSEMBLY BOUNDARIES e MASS IRANSFER MOLTEN STEEL FROM HEXCAN WALLS IS ADDED TO LIQUID STEEL COMPONENT FUEL CRUSTS CAN FORM ON AND REMELT FROM SUBASSEMBLY HEXCAN WALLS VAPORIZATION AND CONDENSATION OF EACH COMPONENT IS MODELED , e ENERGY TRANSFER ENERGY TRANSFER BETWEEN LIQUID FUEL AND LIQUID STEEL IS TREATED ENERGY TRANSFER BETWEEN EACH COMPONENT AND SUBASSEMBLY HEXCAN WALLS IS TREATED . ( D
l ji , ADDITIONAL FEATURES OF TRANSIT-HYDRO
~
o ENERGY EXCHANGE COEFFICIENTS 1 SEVERAL OPTIONS BUILT INTO CODE CAN ALSO BE SUPPLIED BY USER 1 4 MOMENTUM EXCHANGE COEFFICIENTS a MUST BE SUPPLIED BY USER ARE PERMITTED.TO BE FLOW REGIME-DEPENDENT e FUEL REMOVAL MECHANISIMS BLOWDOWN THROUGH BOTTOM OF CONTROL ASSEMBLIES RADIAL REMOVAL THROUGH GAPS BETWEEN RADIAL BLANKET AND RADIAL REFL5CTOR SUBASSEMBLIES
~
BLOWDOWN THROUGH UPPER CORE SUPPORT STRUCTURE IF UPPER AXIAL BLOCKAGES MELT OUT e NEUTRONICS FEEDBACK WILL BE COUPLED WITH FX2 TWO-DIMENSIONAL TIME-DEPENDENT DIFFUSION THEORY CODE IN NEAR FUTURE (COMPLETE BY 6/79)-
. h.
e e
.= .
@y
, . - - .,---*w-, -,w,
, USEFUL CODE
~
e' DEVELOPED IN ORDER TO INVESTIGATE PHENOMENA RELATED TO FREEZINs AND PLUGGING OF A MOLTEN FUEL / STEEL MIXTURE AS IT.IS EJECTED INTO PIN BUNDLES OR GAPS BETWEEN HEXCAN e TREATSINCOMPRESSIBLEFLbWOFALIQUIDINTOANNULI, PIPES, OR BETWEEN PARALLEL. PLATES (FLOW IS ONE-DIMENSIONAL) e PROVIDES DETAILED MODELS OF CRUST FORMATION AND ABLATION - e AS THE LIQUID FLOWS, IT CAN FREEZE AND FORM CRUSTS 4 e SIMULTANEOUSLY, A MELT FRONT CAN DEVELOP IN THE UNDER-LYING STRUCTURE e IF CRUST BREAKS UP DUE TO TURBULENCE OF FLOW, ABLATION OF THE UNDERLYING MELTED STRUCTURE CAN OCCUR e IHE FREEZING AND PLUGGING MODELS IN TRANSIT ARE BASED ON EXTENSIVE STUDIES OF THESE PROCESSES WITH USEFUL 1 ( l . h
l
* + * , l a O O KEY ASSUMPTIONS IN ORIGINAL VENUS-II HYDRODYNAfjlCDISASSEMBLYCODE e
e REACTOR MATERIALS BEHAVE LIKE A HOMOGENEOUS ISOTROPIC l FLUID (MIXTURE) e POINT KINETICS AND FIRST-ORDER PERTURBATION THEORY ARE APPLICABLE e d e THE RELATIVE POWER LEVEL OF EACH MESH CELL REMAINS CONSTANT e HEAT TRANSFER FROM THE FUEL CAN BE NEGLECTED e FUEL VAPOR AND MATERIAL COMPRESSION ARE THE ONLY SOURCES OF PRESSURE O e O 0 e O O 8 e e 4
' APPLICABILITY AND LIMITATIONS OF VENUS-Il e
BASIC ASSUMPTIONS BETTER SATISFIED WITH ENERGETIC , EXCURSIONS - e PROBLEMS ARISE WITH MILDER EXCURSIONS (< 20 $/SEC.) l 4 - - LOWER TEMPERATURES AND PRESSURES MAKE STRUCTURAL CONSIDERATIONS IMPORTANT LARGER TIME-SCALES MAKE HEAT-TRANSFER CONSIDERA MORE SIGNIFICANT. e ' MODEL RESTRICTED TO SMALL DISPLACEMENTS LIMITATIONS ON LAGRANGIAN MESH DISTORTION DIFFICULTY IN MODELING INTERACTION (PENETRATION) 0F MOLTEN CORE MATERIALS WITH SURROUNDING STRUCTURE e THUS, THE NEED WAS RECOGNIZED FOR CODES WHICH COULD FOLLOW THE EXTENDED MOTION OF CORE MATERIALS AND TREAT MOMENTUM AND ENERGY INTERCHANGE BETWEEN THE CORE COMPONENTS AND WITH THE SURROUNDING STRUCTURE. THIS HAS LED TO THE l DEVELOPMENT OF THE SIMMER AND FX2/ VENUS-III CODES 9 9 k ( . . a G 9 5 0 0 9
^-
CONSIDERATIONS GUIDING'DEVELOPMENTLOF FX2/ VENUS-lII EXTEND MOTION HYDRODYNAMICS / SPACE-TIME NEUTRONICS CODE e FOR HYPOTHETICAL LMFBR DISASSEMBLIES, SUFFICIENT ENERGY CAN BE DEPOSITED TO MOVE CONSIDERABLE MATERIAL THROUGH EXTENDED-DISTANCES, NECESSITATING EULERIAN HYDRODYNAMICS e WHEN FOLLOWING EXTENDED MATERIAL MOTION, TIME SCALES BECOME EXTENDED SO THAT INTERCOMPONENT MOMENTUM TRANS-PORT CAN OCCUR; NECESSITATING A TWO-FLUID HYDRODYNAMICS MODEL o EXTENDED TIME SCALES PERMIT INTRACOMPONENT AND INTER-COMPONENT HEAT TRANSFER, WHICH CAN AFFECT PRESSURE DRIVING FORCES AND GENERALLY LOWERS FUEL TEMPERATURES e TREATING EXTENDED TIME SCALES ALSO NECESSITATES TREAT-MENT OF HEATED MATERIAL INTERACTION WITH STRUCTURAL COMPONENTSs INCLUDING BOTH FLOW DIVERSION AND ENERGY SINK EFFECTS, WHEN THE FULLY HYDRODYNAMIC FLOW MODEL BECOMES INVALID e . G e e e-o ( , L se j
OBJECTIVES 0F FX2/ VENUS .III DEVELOPMENT EFFORE - --- e IREAT EXTENDED MATERIAL MOTION EULERIAN HYDRODYNAMICS S. PACE-TIME NEUTRONICS AS k' ELL AS POINT KINETICS o ALLOW FOR RELATIVE MOTION BETWEEN VAPOR AND LIQUID FIELDS ,
~
TWO-FLUID HYDR 0 DYNAMICS , e ALLOW MASSs MOMENTUM AND ENERGY TRANSFER TO OCCUR ~~ BETWEEN COMPONENTS IN THE TWO FIELDS e ALLOW ENERGY TRANSFER BETWEEN COMPONENTS IN EACH FIELD - e MODEL THE INTERACTION OF THE TWO DYNAMIC FIELDS WITH A STATIC STRUCTURE FIELD e PROVIDE A DETAILED INITIAL CONDITIONS TO DAMAGE
. EVALUATION CODES SUCH AS ICEC0-II e KEEP MASS, MOMENTUM AND ENERGY EXCHANGE MODELS SIMPLE ENOUGH TO KEEP RUNNING TIME WITHIN REASON
.. - r- !
STEPS.IN DEVELOPING.THE HYDRODYNAMICS CAPABILITY FOR FX2/ VENUS-III e STEP-1: VENUS-III (COMPLETE) TWO-DIMENSIONAL HOMOGENEOUS FLOW EULERIAN HYDRODYNAMICS IMPLICIT TREATMENT OF MASS, MOMENTUM AND ENERGY CONSERVATION EQUATIONS USE MODIFIED ICE APPROACH TO OBTAIN POISSON EQUATION FOR PRESSURE FIELD ADIABATIC FUEL ENERGY EQUATION o STEP 2: VENUS-III (RM) (COMPLETE) TWO-DIMENSIONAL's TWO-FLUID EULERIAN HYDRODYNAMICS MOMENTUM EXCHANGE BETWEEN FIELDS IWO-LEVEL GAUSS-SEIDEL/ NEWTON-RAPHSON SCHEME DEVELOPED TO SOLVE EQUATIONS ADIABATIC FUEL ENERGY EQUATION , e , m e.e w. esee eq= =
. O 1 . /.
~
STEPS IN DEVELOPING THE H.YDR0 DYNAMICS CAPABILITY FOR FX2/ VENUS-III - CONT'D e STEP 3: VENUS-III (HMT) (COMPLETE) TWO-DIMENSIONAL HOMOGENEOUS FLOW ALL COMPONENT MASS AND ENERGY EQUATIONS TREATED IMPLICITLY HEAT TRANSFER BETWEEN COMPONENTS AND PHASE CHANGE TREATED IMPLICITLY e STEP 14: VENUS-III (RMHT) (PRESENT EFFORT) TWO-DIMENSIONAL, TWO-FLUID EULERIAN HYDRODYNAMICS DYNAMIC MOMENTUM INTERACTION BETWEEN THE TWO FLUIDS AND WITH STATIONARY STRUCTURE FIELD NON-LINEAR GAUSS-SEIDEL/NENTON-RAPHSON SCHEME PROVIDES FOR ITERATIVE IMPLICIT SOLUTION OF EQUATIONS ALL COMPONENT MASS AND ENERGY EQUATIONS INCLUDED IN' IMPLICIT ITERATION HEAT TRANSFER,' PHASE CHANGE, AND MASS SOURCES . AND SINKS TREATED IMPLICITLY
~
e COUPLING WITH FX2 SPACE-IIME .NEUTRONICS .'
- ALL CODES INCLUDING VENUS-III.(RMHT), EXIST AS .
BOTH STAND-ALONE VERSIONS AND ALSO COUPLED TO
. FX2 CODE
.T
SUMMARY
< OF FX2/ VENUS-III DEVELOPMENT EFFORT
~
e FX2 CODE IN PRODUCTION USE
~PROVIDES OVERALL PROGRAM STRUCTURE .,
, PROVIDES A POWERFUL DATA MANAGEMENT CAPABILITY (AN IMPORTANT CONSIDERATION) e VENUS-III (RMHT) STATUS 6
NEARING COMPLETION POWERFUL IMP (ICIT SOLUTION TECHNIQUE HANDLES THE EXCHANGE MECHANISMS WELL IMPLICIT SOLUTION TECHNIQUE LESSENS DIFFICULTY OF TREATING TRANSITION,FROM TWO-PHASE TO SINGLE-PHASE CONDITIONS \ e COMPLETION OF FX2/ VENUS-III EXPECTED IN FY1979 L
. WILL PROVIDE CAPABILITY SO'MEWHAT ANALOGOUS TO
~
I SIMMER-II l MODELSFORPHASETRkNSITIONANDFORSTRUCTURE l FIELD ARE LESS SOPHISTICATED THAN SIMMER-II FEWER COMPONENTS TREATED THAN IN SIMMER-II - B BECAUSE OF MODEL SIMPLIFICATIONS AND IMPLICIT '
" ~' '
SOLUTION TECHNIQUES, SHOULD RUN FASTER THAN . SIMMER-II . . l . -
-in0ULD COMPLEMENT SIMMER-II IN THE STUDY OF CORE
. , , O 3 ASSEMBLY PHENOMENA OVER EXTENDED TIME SCALES . -
._ ~... . . . . . _ - - - . , . . . . . . . _ _ _ , . . . . _ . . - . - . . - , . . . , . . . . - . . . _ _ . . . .
EXPGMMAL VERIFICATICN OF TRANSIT-HYRDRO AND FX2/VDUS-III CCCES Phenomena ' Applicable Experiments Performed of Interest to or in Planning Itydrodynamic behavior TRANSIT-HYDRO out -of-pile tests - of boiling fuel / steel using microwave heat-mixtures' ' ' ing - TRFAT RX test . d Axial blockage forma- TRkNSIT-ifYDRO thermite ejection into tion and removal VENUS-III bundles
'IREAT RX tests ,
Fuel crust formation TRANSIT-HYDRO TREAT RX tests out-of- ; and removal pile simulant mater.ial tests Failure of intact 'JRANSIT-HYDEC SRI hexcan tests thermite , hexcan walls from ejection into various thermal and mechan- hexcan geometries ical loads Response of molten / 2RANSIT-HYDRO . boiling regions to
.s.erial reentry W $ '$
1 ( e- e- , , , s,,>y n -- q--,e,-- e-- - e ny ,en , -r- ,- _
. . hW .
9 EXPERMEh'IAL VERIFICATION OF TRANSIT-HYRDRO AND FX2/ VENUS-III CODES , Phenomena
~ Applicable Experiraents Performed to or in Planning of Interest. ,
l Blowdown of molten / TRANSIT-HYDRO SRI upper plenum tests vaporized core VENUS-III thermite ejection tests material through- > escape paths to upper and lower , plenum h Hydrodynamic behavior FX2/ VENUS-III Out-of-site simulant of molten / vaporized . material tests core material'at high temperatures (including self-
'. mixing and slip) e Energy and momentum FX2/ VENUS-III Out-of-site simulant interchange between material tests fuel and non-fuel core components Within-core structure FX2/VENUhIII SRI tests responzs to energetic disassad ly .
4 Extended notion of FX2/ VENUS-III SRI upper plenum tests liqui 6/ vaporized core rateria'. t.hrough upper . subassembly internals -M
- N'
ANL STRUCTURAL RESPONSE CODES %
\
f e FOR CONTAINMENT RESPONSE REXCO-HEP
-- 2D, LAGRANGIAN, FINITE DIFFERENCE, EXPLICIT l ICECO . -- 2Di EULERIAN.,' FINITE DIFFERENCE, IMPLICIT / EXPLICIT ALICE -- 2D, ARBITRARY LAGRAtiGIAN-EULERIAN, IMPLICIT / EXPLICIT e FOR COMPONENT RESPONSE STRAW-WHAM -- 2D, LAGRANGIAN, QUASI-EULERIAN, FINITE ELEMENT SADCAT
-- 3D, LAGRANGIAN, FINITE ELEMENT e FOR PIPING LOOP RESPONSE ICEPEL
-- 2D, EULERIAN, FINITE DIFFER.ENCE e
4 e e a t'
/
* - - h
. EXPERIMENTAL SUBSTANTIATION OF HYDRODYNAMIC - STRUCTURAL CODES
~
o REXCO-HEP , EXTENSIVE VERIFICATION 1 - EARLY BRITISH, FRENCH, AND PROCTOR EXPERIMENTS 2 - EXPERIMENTS IN SUPPORT OF FFTF 3 - ANL/ SRI EXPERIMENTS d 4 - CURRENT COVA EXPERIMENTS e ICECO SUBSTANTIAL VERIFICATION 1 - ANL/ SRI EXPERIMENTS 2 - SNR-300 EXPERIMENTS 3 - CURRENT COVA EXPERIMENTS o ALICE MINIMAL VERIFICATION e STRAW-WHAM SOME VERIFICATION ANL/ SRI EXPERIMENTS Ch r ., , , ,- . . - . - - , - .-l
EXPERIMENTAL SUBSTANTIATION OF HYDRODYNAMIC - STRUCTURAL CODES CONT'D . o SADCAT LIMITED VERIFICATION ANL/ SRI EXPERIMENTS e ICEPEL - LIMITED VERIFICATION ANL/ SRI EXPERIMENTS EXPERIMENTAL WORK IS IN THE PLAN ING STAGE FOR ULTIMATE VERIFICATION AND/OR MODIFICATION OF THESE COD S.' , , l . l
. l
. 1 c(f
y . ICECO: (.lMPLICIT CONTINUOUS-FLUID EULERIAN [0NTAINMENT . c EULERIAN HYDRODYNAMIC FORMULATION 1 0 IMPLICIT FINITE-DIFFERENCE SCHEME 9 REATS WAVE PROPAGATION, SLUG IMPACT, BUBBLE MIGRATION, AND COOLANT SPILLAGE a e STRUCTURAL RESPONSE CAPABILITY (FINITE-ELEMENT STRUCTURAL DYNAMICS PROGRAM) 0 LARGE MATERIAL DISTORTION, LONG-DURATION CALCULATION e 9 # @ g e I
- me w ,eme
{0
PENETRATION HOLES
. (INITIALLY CLOSED)
( ' HOLDDOWN REACTOR / BOLTS
' .46 24 cm - '
.. a +
#[ COVER HEAD '-# _ _ .
"I :?!.: ::
te n p
, //
INITIAL ' FREE-SURFACE
...I. '
POSITION . . :: : : G ' o :: : L-CORE BARREL
/ j-
%, : i : '.j/ AND RADIAL SHIELD
/p" :
so - g GAS BUBBLE g
. :l ': . : '
/ '.
/
Po : 20.4 MPo g '" v g
.:l: '
f :. ~ CORE - SOPPORT i !l! / :ili!!! / STRUCTURE
- 8 d /VVVVV//
- ;l:, . .
5, A
- j :.. .,
.- .._, -v ' REACTOR VESSEL p -
10 @ 38. lcm:3BI cm l Typical ICECO Hodel for Analysis of Structural Response to Vapor Bubble L:pansion l .- I . w w -- < < , - -
.,, -,,,~,+m-,. ,
\
- i STRAW-WHAM FINITE EUiENT CODE .
l 6 NONLINEAR ANALYSIS CAPABILITIES STATIC , MODAL TRANSIENT e GEOMETRIES TWO-DIMENSIONAL AXISYMMETRIC THREE-DIMENSIONAL e ELEMENT TYPES STRUCTURAL ELEMENTS CONTINUA . e ELEMENT FEATURES . TREATS MATERIAL NONLINEARITIES TREATS NONLINEARITIES ARISING FROM LARGE DISPLACEMENTS 9 IREATS BOTH THERMAL AND MECHANICAL LOADS o- EMPLOYS A COROTATIONAL COORDINATE SYSTEM WHICH ROTATES
^
- AND TRAi4 SLATES WITH THE ELEMENTS 4
s p
- s .
- i l
STRAW-WHAM: ' ADDITIONAL CAPABILITIES AND SPECIAL ' MODELING' TECHNIQUES .
, i e 'IYPES OF LOADING .
IMPULSE LOADING j PRESSURE-TIME PRESSURE-VOLUME SPECIAL ELEMENT FOR SOURCE ' GENERATION A) FUEL-COOLANT INTERACTIONS B) ISOLATE-PARAMETER CHANGE INFLUENCES SLIDING INTERFACES l e O IMPACT-INTRODUCED THROUGH SLIDING HAVE SEVERAL SURFACE' ' l INTERACTION POSSIBILITIES BONDING TOGETHER. , BONDING-DEBONDING . BONDING-SLIDING e QUASI-EULERIAN ELEMENT o- AXIAL FL0w ELEMENT ACCOUNT FOR AXIAL COMPLIANCE .
.- COUPLED TO P. LANE HYDRODYNAMIC ELMENTS
-- NO FLOW ABOVE INTERACTION PLANE (35>
, . _ -..e. -c.. .-
l . . L e ' FUEL PIN MODELS - W USE QUASI-EULERIAN FOR PLUID VARIABLE MODULUS FOR PINS AND WRAPS s emen. e O
- 6 t
e e 4
*
- O l.
e e
- e. 48 e
0 I 9 4 6Sem e 9 4
- 9 0
. y (o.
e -
A i* ' . , 1
. ADJACENT ACCIDENT SUBASSEMBLY __.
Z - -
. / ,-
n
/
. . y'/
\
I D '
\ I
( - j f( .L - , {I s s-i TD ' AXIAL FLOW
.{
s e- . li . Q j'.s es-l, s
, s
-s
- I ;/
'N \ l
- { p,: , y .
; g s-ys -- .
,9 , ._
l g N, s -
, s o : -- \
o 1 f ., c, - SOURCE . h
? s
// .
s-
. . i (s s p s-t 4 : s- ' '^
sf Q:: 2D MODEL IN '
\
x y PLANE JV 7 . INTERFACE FLUID
.Cr
( AXISYMMETRIC - 3D MESH - STRAU-UHAM Ffodel of Subassembly to be Subjected to Large
- Pressure Pulse and Adjacent Subast.embly -
I~ t,
/
- s' ge h . .
+-
\
s, .
\
3 '*, '
, s
. . \
TIME = 0 ms TIME: 0.06 ms
' ~' \ - * \
\
w s
\a
. \. g.. .
i; s \ N s . y
'h '
. sN, x.
s g . 4 TIME = On ms TIME = 0'.i5 ms. \*
. . )
ICEC0/ STRAW-WHAM CODE SYSTEM e COMBINE ICECO AND STRAW-WHAM INTO A SINGLE CODE o OBTAIN ENERGY SOURCE FROM FX2/ VENUS-III CODE 4 e o PROVIDE ANALYSIS OF: SUBASSEMBLY DEFORMATIONS DEFORMATION OF STRUCTURES SURROUNDING CORE SUCH AS RADIAL SHIELD, CORE BARREL, CORE SUPPORT STRUC-TURE, AND UPPER INTERNALS
~
e TREAT EFFECTS OF SODIUM SPILLAGES THROUGH PENETRATIONS AND FAILED SEALS e ANALYZE MOVEMENT AND DEFORMATION OF REACTOR VE3SEL AND HEAD COVER , o SPECIFICALLY AIMED AT ADDRESSING QUESTION OF UPPER INTERNA RESPONSE TO HCDA LOADINGS 8 O O O e G
\.;t
'. a . .
. j
. PENETRATION HOLES !
(INITIALLY CLOSED) ( HOLDDOWN REACTOR BOLTS ._. COVER HEAD 24.46 cm-
.a u s U i. :. i. i.
- h. u ..
.. .. /// -
INITIAL .. .. FREE-SURFACE I UPPER INTERNAL. l: POSITION
' l:""
/-
STRUCTURE
/ // f f, ? :: -
E
" / / / / :l: : /
, CORE BARREL-u :: :' M
-i /".y AND RADIAL SHIELD N
-Tilii
/
...l..
.:::.. /j ;;;;,.,
g CORE g - g . i.. / .. .. ..
'e :: :- .: .: "-
CORE- SUPPORT O . :i: ::: ili:! / ;iliili- / STRUCTURE
/VVl/VV//
.:: .:: .:! ::. .:: ... : .: i
! :!: ::i i! i i:. ". !!/ -
.. ... ::l::- . ,:A
- .: ::l:.- . .gr ' REACTOR . ,
VESSEL 1 U , _10 -@ 38.l cm : 381 cm _ T"00 'Model of LMFBR with Upper Internal Structure.
# l j
.* e
. . - - - . . . _ i_ . , , . , _ , , , , , , _ _ _ _ , _ , . _ . , _ , . , . , , , , _ _ . , , _I
a FUS S--FUM0-T - DEVELOPMENTAL GUIDELINES ,
~
INTERFACE WITH EXISTING SAFETY CODES
~
. MODEL HETER 0GENEITIES OF FFR
. DEVELOP EXPLORATORY TOOL FOR SIMULANT ..
EXPERIMENTS
- PROVIDE SAFETY ANALYSIS CAPABILITY FOR IN-PILE TRANSITION PHASE EXPERIMENTS
. - DEVELOP ECONOMICAL CODE FOR PRACTICAL -
APPLICATIONS . HEDL l 1 l
~
f
~
i i,
~
! CAPABILITIES i ) FUMO-T, FUSS S IMMER j ! SAS INTERFACE
~
EXACT ,! NONE l , l NEUTRONICS MODELS 3-D DIFFUSION' l 2-D TRANS PORT . l, 2-D DIFFUSION . FLOW REGIMES CHURN TURBULENT. j BUBBLY i~ . DROPLET ! DROPLET l . l l GE0 METRY ! ! HYDRODYNAMICS HETEROGENE0US l 2-D I-D CHANNELS j R-Z- , 1 , j NEUTRONICS 3-D j' 2-D i HEX-Z OR X, Y, Z j R-Z . I 8 1 i HEDL 8 9
l . i CAPABILITIES (continued) . FUMO-T, FUSS
, S IMMER
- WHOLE CORE SUITABILITY HETEROGENEOUS (FfR) l SYMMETRIC (CRBR)
~
PRIMARY CAPABILITY MEDIUM ENERGY HIGH ENERGY ! BOILING ! DISRUPfl0N . MAXIMUM PROBLEM ~5 MIN ! ~ l SEC? TIME (BOILING) ! l l ! l HEDL i 1 a 1 .
a APRICOT PARTICIPATION FUNCTION ORGANIZATION COUNTRY KEY PARTICIPANT -
- COORDINATION SCIENCE APPLICATIONS, INC. US D. BERNSTEIN
[ REVIEW SANDIA LABS. US W. HERRMANN i LAWRENCE LIVERMORE LAB. US
. M.~ WILKENS i NAVAL SURFACE WEAPONS CENTER US R. LORENZ ATOMIC WEAPONS RESEA.RCH ESTABLISHMENT ENGLAND N. HOSKINS .
CALCOLATION ~ ARGONNE NATIONAL LAB. (ANL) US S. FISTEDIS S ANDIA LABS. (SLA) 4 US L. BAXTON , PHYSICS INTERNATIONAL CO. (PI) US BIRNBAUM SCIENCE APPLICATIONS, INC. (SAI) US M. GROSS ATOMIC ENERGY ESTABLISHMENT (AEE) ENGLAND R. POTTER EUR TO JOINT ESEARCH -lS PRA . (EU) TAL P. ASO l-STELLA t INTERATOM (IA) W. GERMANY D. DOERBECKER j INSTITUTE FOR REACHTORSICHERHERT (IRS) W. GERMANY ULLRICH POWER REACTORS AND NUCLEAR FUEL DEV. JAPAN C. KINOSHITA CORP. (PNC) ' POWER RESEARCH INSTITUTE (EGU) CZECH. V. ADAMIK - - 1I
c .
> t 1
PARTICIPATION: CODES / USERS CODE NAME USER SPATIAL ALGORITHM TIME INTEG. . ARES IAllRS L F-D SURFACE INTEGRAL EXPLICIT . CEFRA EGU L " F-D SURFACE INTEGRAL. CSQ SLA .E F-D " DONOR CELL DIFF. EURDYN IH EU L ' F-E CONST. PRES. QUAD. " PISCES 2DL PI, PNC, HSE L F-D SURFACE INTEGRAL " A I PISCES 20 ELK PI MIXED F-D SURFACE INTEGRAL " , REXC0 ANL L F-D MIDPOINT " l SOURBOM CMPl EU E F-D CENTERED DIFF.
- i. STEALTH sal L F-D SURFACE INTEGRAL, "'
T00DY 3 SLA L F-D SURFACE INTEGRAL " i
- L = LAGRANGlAN E = EULERIAN F-D = FINITE DIFHRENCE F-E = FINITE ELEMENT F -
(A)"LAGRANGIAN -REGIONS ARE COUPLED ACROSS MUTUAL INTERFAC5S TO l EULERIAN REGIONS. ~ t' , v- .sey+--
- __ _ - _ - _ _. __ __}}