-- 55- 

EVALUATION 

GAS COOLED FAST REACTOR 

0 

JUNE 1972 

DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY 

U.S. ATOMIC ENERGY COMMISSION 
Abi EVALUATIOU 

OF 

TW GAS COOUD FAST KEACTOK 

JUNE 1972 

DISCLAIMER 

This report was prepared as an account of work sponsored by an agency of the United States 

Government. Neither the United States Government nor any agency thereof, nor any of their 

employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or 

process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, 

manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views 

and opinions of authors expressed herein do not necessarily state or reflect those of the 

United States Government or any agency thereof. 

. AL&VISIQA OF UACTOB DEVELOPMENT AN0 TEMNOLOGY 

U.S. ATOMIC ENERGY COMMISSION 
DISCLAIMER 

Portions of this document may be illegibie 

electronic image products. Images are 

produced from the best available original 

document. 
IC, 

I1 

111. 

EV . 

V. 

VI. 

VIS e 

VfII * 

IX 

32 

32 

32 

35 

38 

40 

43 
Page 

- Tab le 

ry of Oxide-Fueled WFR Design Characteristics ............ 18 

2. Comparison of Oxide and Carbide Fueled GCFR Designs ............ 20 

Figure 

1. Development Time Scales ........................................ 11 

2. CutaW%y View Of PCRV ........................................... 26 

3. Schematic of Pueh Element Pressure Equalization System ......... 36 

4. Phases in Achieving LMFBR Program Objectives ................... 4% 

5, Utility Acceptance Requirements €or LMFBR Plants ............... 42 
AN GAS PAST REACTBK 

I 

th@ coaek?pt as le option to existing power production syetms and 

eo gain public aceepr. 

after thfs is 

h@%vy, long-tern te, of rerpources in funds, facilities and 

new forin of power production. It i8 only 

at the utlPlties and fndustry could make the 

personael to provide the traasition from the early experimental facilities 

and demonstration plants to fd.1 scale cmetcial reactor power plant 

systems. Consistent with the 'polbcy eeeablishecrl for all power reactor 
2 

engfneering, 

. A second phae %a \pP he engineering and manufae:trlrriglg 

capabilities are ~ev~lope~~ 

depth ~~~~~~~~~~g and proof tasting of first-ob-a-kind 

This Pncludee the conduct of inequipment ad systems, These would then be 

is sufficiently devlelopeh and confidence 

is attained, the next stage would be the 

construction of large demonstration plants. 
A third phaae in which the utilities mke Parge erccala cotmi 

to electric generating plants by developing the capability tgs 

mananage the d fgn, construction, test and operation of 

power pPants in a safe, reliable, economic, and ~VSP 

acceptaBLe manner. 

Significant experience with the Light Water Reactor (LWR) the High 

Tmperature Gas-Cooled Reactor (IiTGR) and the Liquid Metal Fast Breeder 

Reactor (LMFBR) haas been gain&? over the past two decades pertaining to 

the efforts that are required eo develop and advance nuclear reactors 

to the point of publfc and commercial acceptance* 

clearly demonstrated that a logical progressfon thtough each csf the three 

phasea is an exorePlarPg dffficul , time consuming and costly mdertak%ag, 

requiring the highest level of technical management, professional cmpeThis experience has 

tence and organitatfoa skills. 

rerat experience in the expanding LWR design, coasttuction and 

lieem iaag ac tivieieas 

base raehnolo 

although these were satisfactory in many cams for the first experiments 

and demonstration plants. 

advanced in its developaent, tracks closely this LWR experience and has 

This has agafn been demonstrated b9 the 

ich aPPghasizes cleatfy the need for men stronger 

and engineering efforts than were tn3tiallg provided, 

The LMFBR program, which is relatively well 

further reinforced tMs need as fts applies to the technologyy dwelopment 

and engineering lication areas. 

3 
4 

It should also be kept in mind that the large backlag of codttaents and 

the shortage of qualified engineering and technical management personnel 

and prooftest facifiries in the goverxunent, in industry arid in the 

utilities des it even more necessary that all the reactor system be 

thoroughly designed and tested before additional significant commitment: 

to, and construction of, commercial power phnte is initiated. 

The large scale ccmnnitments to th uranium-plutonium fuel cycle through 

purchases of the 

resources engendered by these COmmftments led the U. S. Atamis Energy 

and tho substantial investment of the Nationts 

Cammission (AEC), Division of Reactor Development and Technology to 

initfate reviews in 1966 of the technical status and the possible 

benefits implicit in the development of the various reactor concepts 

being considered in the civilian nuclear power program for meeting 

future power nee&, 

on making effective use of our national resources and help determine 

the requirements for, and allocations of, these resources, With regard 

to the CCF'R, the AEC evaluated this concept in 1969 and issued "An 

Evaluation of Gas-Cooled Fast Xesctors," WASH-1089, April 1969, along 

These revfews were needed to help provide guidance 

with a companion report "An Evaluacion of Alternate Coolant Fast Breeder 

Reactors," WASH-1090, April 1969. 

The GCFR designs evaluated in WASH-1089 were based, in large measure, on 

information provided by the single industrial developer of the GCFR, and 
therefore, the repott generally reflected this 

am. The infomation in WASH-1089 was used as 

viewpoint and enthusiinput to the mbseqwnt 

systems analysis (WSB-1698) md sost-benefit Strpdi- (WME-1126, USIIEC, 

"Cost-Benefit Analysis of the U.S. Breeder Reactom: Program," 1969 and 

5WH-1184, USAEC, "Updated (2970) Cost-Benefit Andysis of the U.S. 

Breeder Reactor Program," January 1972). 

WASH-1098, "Potential Nuclear Growth Patterns, '' December 1970, was 

prepared by the Systems Analysis Task Force (SATF) which was concerned 

with the development and applfcatian of a model of the U.S. electrical 

power economy, The model input data was provided by the individual task 

forces charged with the evaluation of the reactor concepts under consideration, including GCFR. 

to GCFR, indicated that largely because of the uncertainties in the cost 

The results of tlafs evaluation, as it applied 

estimates for both the LME'BR and the GCFR, it was not possible to draw 

a d@finitive conclusion concerning the soundness and desirability of 

conducting a parallel breeder program on the GCFR in addition to the 

WBR. Accordingly WASH-1098 concluded that "the current state of 

knowledge is not sufficient to support a definitive evaluation of the 

potential of the GCPR against that of the LPIFBR." 

The consistent concPusion reached in the cost-benefit studies (WASH-I126 

and -3.184), viz., sufficient information is available to indicate that 
6 

the projected benefits from the LMFBR program can support 8 parallel 

breeder program, is highly sensitive to the assumptions on plant capital 

costs. 

ample experience exists, capital costs and especially small. estimated 

differences in costs are highly speculative for plants to be built 15 or 

20 years from now, it is questionable whether analyses based upon such 

costs should constitute a major basis for decision making relative to 

the desirability of a parallel breeder effort. 

- With the recognition that even for ongoing concepts on which 

In compliance with a request from the Office of Science and Technology 

for further review of the CCPR at: this time, the AEC has undertaken this 

internal assessment which examines the technical developments that have 

taken place in the continuing research and development and design efforts 

on the GCFR system. 

the initial research and development phase with emphasis on the development of basic GCFR technology. 

at Gulf General Atomic (SA) and has been supported by governmentsponsored research at a level of about $1 million pet year over the past 

This request recognized that the CCFR has been in 

The program has been carried out primarily 

several years. The utility industry is supporting research and 

development on this concept at a level of about $1.4 million per year. 

Total expenditures from 1963 to date on GCFR technology have mounted to 

approximately $17 million, 
The GCFR is still in tha: early phasw of an overall RbD effort 

notwithstanding tbe b 

LPiFBR programs, 

with the GCFR concept to move forward toward their satitlfaetory reaalution would require a substaratial feaearch and development program along 

ectad to accrue to it froa the ll'EGR aand 

An adequate investigation of the probPem associated 

the lines outlined herein with sufficient proof-testing of major components 

and systems to provide 

and investment in such plants for large-scale power generation. 

prerequiaites for such a progyt 

programs, e.g., same of the fuel development, physics and safety work in 

the Liquid Metal Fast Breeder W~CLQX- (LMIPPIR) program and coreponeat: 

development in the High Temperature Gas Reactor (HTGR) prograa, 

component and plaat operating and maintenance experience at the 330 %?e 

Fort St. Vtain reactor will provide important basic information. 

Hawever, most of th@ critical ad unique characteristics of the CCFR 

ces csaPaenrsurate with the large commitment 

Some 

arc provided by portions of najor ongoing 

Also, 

uately represented in research and development programs on 

other concepts e In dition, the flexibility for resolution of such 

t and engfineering problems normally required by the designers, 

constructors, 

due to the compact arrangement proposed for the GCFR. Consequently, 

if feasibility were to be confirmed through appropriate experimental 

reactor and other facility operation, the additional overall 

technical effort needed for such a full-scale program on the 

d operators of even the more proven concepts is restricted 
8 

6cFR could be comparable in magnitude to the efforts on ather major 

reactor developnent progt such * the LkfRt md 

Experience in reactor development progr 

demonstrated that different or 

cemi of introducing B P BLC tar depre lopmen e p rogr 

in this country 

tions in evaluating the projected 

d carving it Eoward 

to the point of Parge-a 

dlf f eren t est 

efforts, and the costs and time required to bring &at program to a stage 

of successful large scale application and public accepttdnce. 

rcial utilization, would arrive at 

s, scope of development cmd @ongPaeerin$ 

Based upon the extensive program required to bring a new comccept to 

commercial utilization, and 

accrue from progress in the 

the incremental cost to the government of a parallel breeder program of 

this type has be 

in8 into account the benefits expected to 

W and HTGR programs over the next 5 years, 

tad b9 the AEC to range up, to about $2 bfllfcn 

fa undiseomted dlirect costs, (WASH-1184). The CCBaR would requfre 

magnitude of fundin to this level in order to establish the necessary 

technology and engiaeering bases; obtain the required industrial capabblity; and advauca through a series of test facilities, reactor experfments, 

and demonstration planes to a commercial GCFR, safe and suitable to serve 

as a major energy option for central station power generation in the 

utility environment. 
9 

I1 

SimMARY 

z"ne GGFB concept uses helium as the coolant gas, which leads to several 

potentially favorable attributes of the GGFW. Helium is both optically 

and neutronically transparent and does not beccRne radioactive. The GCFR 

has a potentially high breeding ratio resulting largely from the coolant 

properties. Since the use of gas cooiing requires a high coolmt pressure 

to secure adequate heat transfer, the GCFR is subject to the poksibility 

of depressurization accidents as well as loss-of-Plow accidents. Tire 

potential for occurrence of these types of accidents requires resolution 

of significant reactor damage and safety questions which are unique eo 

the GCFR concept. In addition, there are other outstanding engineering 

problems aesociated with safety, reliability and maintainability which 

differ significantly fro= sidlier considerations for the HTGR and LMFBR. 

The additional technical work dosle since the publication of WASN-1089 and 

WASIt-1098, both by the &C and private industry, and fur'ther definition of 

problem areas aided by discussions with the Regulatory staff and the 

Advisory Camittee mn Keactor Safeguards (ACRS) have served to reinforce 

I the earlier conclusion of WASH-1089 that a substantial program would be 

needed to support the commercial introduction of the GCF'R. This is 

particularly so since only limited effort cm the GCFR is underway outside 
10 

development work and related large 8c 

countries meh as is the CBWI %or the concept. Ba 

InforaPation ~~r~~~~y avdlable, 8~bl discussed in this report, there 

have bean no si the ~~~~~~~~~% stat-, 

relative to the total effore P red for its cia1 application, 

Although the proponents of then GCFlt c cept would inread to depend hmvtly 

program and, to le, the techno1 

the GCFR CUI be c 

overall development ~~o~~~. This uld not be surprising, abaca the 

goverrmant an8 indostrial fund 

total leBS th $28 million. With a 

betng in only the early stages of 

~~~ on this concept through FY 1972 

eriescs with the LWR, HTGR ornd 

, Fig. 1, With the size 02' the major goveraPac?nt and industrial 

%nvestmenk, recent experience wlth LWR'e cm problems arfsiag in multiple 

cdtments to fall size power pl tes provides ample @videnee of the 

necessity to provfdr a broad industrial base and to conduct astmmive, 

in-depth development and testing efforts addreaped to knll critical conponezits 
11 

Figure 1 

DEVELOPMEM TIME SCALES 

1 JAN 1 IAN 1 JAN 1 JAW 1 IAN 1 JAN 
12 

and systems throughout all project phases. 

on the GCFR would be expected to cover those phases that are currently 

under consideration for the LMPBR Program. 

p. 41). 

development programs (in the range of two billion dollars from the 

An overall full-scale proRram 

(See Figure 1 rand Figure 4, 

Suck a program would involve costs comparable to other reaceor 

Government) in addition to large outlays and commitments by industry and 

the utilities. Such a GCFR program would have to take into account any 

overlapping or concurrent comitwents and availability of resources with 

the HTGR comercialfzatfon efforts. This, of course, is a problem faced 

by other reactor programs as well. 
1x1 

BASIS FOR W~~~-~O%9 AND ASSOCIATED EVALUATIONS 

GCFR development was initiated in Nov BF 1963 by the AEC under a contract 

with Gulf General Atomic (SA)* to fnvescigate the concept wh9ch had 

XErR studies. The AEC-sponsored 

started with the objective of a 

Wt which was to lead to a demonfull-scale plant. An outcome of 

was a conceptual desiw for a 

test bed for fuel development, 

and constructing a special 

evolved frosa earlier privately 

work outlined a development program t 

gas-cooled fast reactor experiment of 54) 

sttation power plant 818 a step towards a 

the next year of AEC sponsored R&D (1964) 

reactor experiment which would serve as a 

and would provide eriemce in designing 

prestressed concrete reactor vessel (PCRV). 

In the period 1965-68, the C and W continued studies of the NFR. A 

conceptual design for a 1600 MWe GCFR power plant was evolved which 

featured a horfzoneal P instead of the original vertical arrangement 

and also differed in other important relspects from the original concept. 

This effort incorporated ideas of the utility companies, particularly a$ 

to the fayout and design of the nuclear steam supply components from the 

viewpoint of operation, intenance and safety. Also, a new design for a 

reactor experiment was developed to reflect engineering aspects of the new 

large plant design. 

- 

Wulf General Atomic formerly was the General Atomic Division of General 

Dynamics. 
The 1000 We GCFR reference conceptual design prepared by GGA for study bp 

the AEC Fast Breeder Reactor Alternate Coolant Task Force 29394 was an 

extension of the above conceptual design. 

reported fn WASH-1089 dated April 1969. 

The task force analyses were 

A br$eP description of the GCFR 

plant designs considered in this study is included in the next section. 

A more detailed description may be found in WASH-1089. During 1967, to 

* 

satisfy the needs of the Kiternate Coolant Task Force as well as to meet 

AEC contractual requirements, a Preliminary Development Plan far the GCFR 

was also prepared. 5 

WASH-1089, WASH-1126, and WASH-1184 

The conclusions of WASH-1089 were based upon the 1000 MWe plant conceptual 

design provided to the Alternate Coolant Task Force. The development plan 

and CQS~S projected in WASB-PO89 were associated with bringing the concept 

to the stage of commercial introduction including construction and operation 

of a reactor experiment to help demonstrate the adequacy of the fuel 

perforname. 

As noted at the time WASH-1089 was published, the designs evaluated were 

based on information provided by developers of GCFR and generally reflected 

their viewpoints and opinions on achievable technology. 

Based upon the information in IJASH-1089, the cost of development of a 

parallel breeder was estimated and examined in the context of the overall 

LMFBR development program. As noted fa. WASN-1126, “Cost-Benefit Analysis 

14 
of the U.S, Breeder Reactor Program," April 1969, if the WFBR wet@ to be 

introduced in 1984 or ear11er, it was concluded that sufficient benefits 

* would be aenerated to support the cost of a parallel breeder program and 

still maintain B benefit-cost ratio in excess of one. 

Although the WBR introduction date was deferred to 1986 in WASH-1184, 

"Updated (1970) Cost-Benefit Analysis of the U,S. Breeder Reactor 

Frogram," January 1972, the concl.usims remain essentially unchanged, viz,. 

a tentative case ean be made to expand the industrial breeder base by 

establishing a parallel breeder program. 

It should be stressed that the cost-benefit ratios derived Pn both 

WASH-1126 and WASH-1184 depend heavily upon the assumed capital costs for 

the various power plants, especially the breeders. 

results to a slight increase in the capftal cost of the CCFR was 

illustrated in WASH-1098 (page 6-50), where an increase of 7.7% Pa the 

assumed cost of a GCFB from 130 to 140 $/We permitted the LMFBR to 

compete at an assumed cost of 150 $/We. 

over the past two years Rave confirmed the tenuousness of utilizing 

assumed small capital cost differences such as these. Variations in design, 

plant location, labor productivity and cost, escalation, interest rates and 

periods of construction can readily produce capital cost differences of 10 

to 15%. Capital costs for all of the breeder designs considered in past 

studies were only best estimates at the time, normalized to some extent 

The sensftiviey of the 

The capital CoBt studies performed 

. 

t costs, At ent, while a reasonable capability to predict costs of 

IV 

GCFR DESIGNS PRESENTED IN WASH-1089 

P. OWL Working Group 

The gas-cooled fast reactor designs reviewed in WASH-1089 were developed 

by GGA. 2*3'4 

National Laboratory (OWL) working group for evaluation. 

2'wo oxide fuelled designs were admitted to the Oak Ridge - 

A carbide 

fueled design was also prepared by GGA, but it was submitted too late in 

the study to allow evaluation. 

Oxide Fueled CCFR DES~KXUS 

The two oxide fueled designs were designated "reference" and "derated" 

designs. Sane princjepal characteristics of the two oxide fueled designs 

are given in Table P which has been modified from Table 2.1 off WMR-1089 

to show the parameters for the latest available 1008 MWe GCPR plant 

design. Design parameters are also shown for a 300 We demonstration plant. 

The entire primary cooling system for the oxide designs wa8 housed within 

a horizontal PCRV which contained the centrally located vertical reactor 

core, four stem generators and their associated circulators located in 

internally isolated compartments at the enda of the vessel. The PCRV hid 

double containment of all penetrations, individual standpipea above each 

17 

core and blanket clement, and a watercooled, but uninsulated, steel liner. 

The whole primary c uit flow path was confined within insulated 

ducting and shells surrounding the reactor and the steam generators; 

thus the main internal compartments were at substantially room temperature. Theme compartments were continuously purged with clean helium. 
ll,vl..led ;.if0 K,.b. 1972 

Table 1 Sunrary of Oxide-Fueled GCFR Design Cr.aracteristicsa 

h'A:'Ii - 1095 - .-- 

GCFR-4 Reference Des GCFI(-LD Derated Design 300 Pdc 111 no rlethtC ion0 tIr P!.VC~ 

Pwer Reactor ntcl-ar power, Mu 

Net electrical pwr, !+d 

Net tnemodynamic efficiency, '$ 

COI~JS~ tion 

Core inlet pressure, psia 

Qeactx pressure drop, psi 

Flov rate. Id? 

Teu.perr~tures, F ("C) 

Fieactcr inlet 

Core iniet 

Core cc-let 

Resct,ur outlet 

Steam ?Lar,t ccnditions 

Leai7ir.g ~~prrhcite: 

Pressure, psi8 

Tarper+t,rr, OF 

Coolant 

Cor.der.;er prces'xe, in. 33 

Peak lines? heat rating, W/ft 

&zxin L: c1sddir.g surface temperhture, 'P ("C) 

~axinm interior fuel temperature,b *P ('2) 

Core C-~?XS.I prrfomnca (at 1OSa pader) 

X372.d 

Local hot spot 

Iio2:nal 

Local ha; spst 

Peak-to-sverage core power ratio 

Xeaa core pcvc dersity, &(ti)hiter 

Wean Pieslle fuel retlng, Mw(th)/kg 

Active core volune, litera 

Active core length, cn (io. ) 

Active core ilnmeter, CI (in.) 

blmket thickness, ca (i 

Rejial blf?2er thkkness, ca ( 

Core 1ecgrh-zo-dlaz.eter ratio hei raterial (core) 

Fuel w:er:al (blanket) 

Fuel d:strl'wtlor. 

Core acd bla-kt deecription 

%abr;Lated values are resuits of ORE review. 

2530 

1W 

39.5 

bCalculated fuel tezperatures hbove zelting point 0: the oxide (4950'F) indicate tbt, 

'raten frca CA-L306'+ dJ:ed Ily 15. 1970 

accordirg L,O OW: cslcdet:or.s, local neltira vi11 probably occur. 

'lok..cn from GA-IOG~~ dated JUAC 8, 1971 

Kalium 

12 50 

i2.t 

12.53 X 

629 (332) 

635 (3%) xe3 (639) 

1193 (M3) 

1515 

lW3 

1.5 

17.8 

1315 (713) 

1L35 (779) 

4435 (2724) 

5205 (2824) 

i.L7 

283 

0.86 

8510 

lL8.2 (58.C) 

269.5 (1W6.0) 

to (23.6) 

depLeteb UOp (W$ T.D.) 

(W$ T.D. 1 

Ily; zoned radially to give 

1.2 redial maximum-to-meen ratio 

Reliuro 

1250 

42.6 

14.41 x iob 

589 (3%) 

595 (313) 

ioea (587) 

1094 (590) 

1850 

9Go 

1.5 

16.3 

E2 6 -* 

311 10C.7 

37.6 3.0 

nc11ua 

1250 

5.47 x 106 S.A. 

40 COC4.877, 

290'0' 

875 

1.75 

12.5 

.- 

1290 

1205 (652) 

1315 (713) 

L765 (2629) 

5039 (2787) 

l'L7 

260 

0.82 

9717 

15L.8 (60.9) 

60 (23.63 

51 (23.1) 

0.55 -- 

Cepleted *a2 (90$ T.D., 

Uniform axially; zoned rsdialL3 to give Depleted 

281.5 (iiO.Sj 

-- 

Mixed Pu and depieted yo2 (%$ T.3.) *- 

Wxad Pu h d;plered UQ 

1.2 radl.61 W15Uia-VO-ineeU ratlo 

21100 

921 

-- 

IS 

-- 

(750) 
19 

The reactor core was asseabled inside a vertical cylindrical steel core 

barrel, which spanued the PCRV from top to bottom. The core barrel served 

the multiple purpose of supporting the core grid plate, contahing and 

directizrg the flow of coolant, and acting as a them1 shield. The core 

consisted of bundles of small diameter fuel rods contained in thinwalled 

metal boxes. These fuel boxes were cantilevered downward from the 

deep section upper grid plate. 

Both of the oxide fueled designs had free standing stainless steel-clad 

pins about 0.3 in. in di eter. The cladding surface was intentionally 

roughened over approximately BO eo 70% of the active core length to 

enhance the heat transfer caefficieut in the downstream portton of the 

sore and thereby prevent excessive cladding surface temperatures. 

esigns a180 included an alternate fuel element design that had 

a pressure equalizing collection system to vent the fission product gases 

to a receptacle that was isolated from the coolant system. The objective 

of this ~~~i~o~~~n% system was to eliminate the large gas pressure 

differential across the cladding while preventing the fission product gases 

from escaping into the coolant. 

Carbide Fueled GCFR Design 

An advanced GCFR design to exploit the high density and excellent 

ity of carbide fuel wag also prepared by GGA. The 

carbide fuel design was submitted too late in the study period to be 

subjected to a technical evaluation. However, the design data were 

reviewed. (See Table 2). More details on this design are contained in 

WASH 1089. 
20 

T&le 2 Comparison of Oxide and Carbide Fueled GCFR Designs" 

Reference Advmc & 

Reactor Reactor 

Oxide Fueled Carbide Fueled 

Common conditions 

Coofwt 

Total pumping power fraction, $ of 

Design maxfmum cladding temperature, "C 

Cladding (stainless steel) thickness-toControl-rod and structure blockage, 46 

Net electric output, Mw 

thenaal 

diameter ratio 

-~ 

Average enrichment of fresh core fuel, 

239pu + "42puju + pu 

Helium Relim 

4 4 

700 m 

0.04 0.04 

10 10 

1, I, cxm 

0.127 0.127 

ApproFriatc Ocore proportions 

Core length-to-diameter ratio 

Core volurne, liters 

Active length, cm 

Active diaixeter, cm 

Fuel volume fraction 

Coolant t.olLme fraction 

Rod dinmeier, cm 

Number of rods 

Opernticrlg conditions. 

Core pressure, pi 

Gas inlet temperature, ZF 

Gas outlet temperature, F 

Steam conditions 

Outlet temperature, OF 

Reheat temperature, OF 

Feed temperature, OF 

Pressure, psi 

Net plant efficiency 

Maximum rod heat iced, kw/ft 

Specific power, total Mw( th)/kg core 

Perfonnance 

fissile at startup 

Power density, kw/liter 277 

Conversion ratio, average 1.51 

Assumed maximum burnup, Mnd/MT 

Core life, years 

Out-of-pile time, years 

Fractional increase in fissile plutonium 

Geometric doubling time, years 

per cycle 

2.29 

1.0 

0.55 0.40 

8,510 4,030 

148 93 

270 2 34 

0.29 0.30 

0.55 0.53 

0.805 0.736 

4Q,m 37, ooo 

1,250 I, 750 

644 56 7 

1,175 1,057 

1000 900 

900 

275 

1,800 

0.373 

Po00 

375 

2,400 

0.395 

18 30 

0.90 1-50 

2.49 

0.5 
21 

hecause of the potential advantages of tne carbides and nitrides as 

fuel for the LMFBH or the @CFR, research has continued, The current 

advanced fuel technology program at Los Alamos Scienrtif ic Laboratory 

involves an effort of approximately $750,000 per year and includes the 

irradiation of carbide and nitride fuel pins in the EBR-II reactor and 

transient rests in the TWAT facility. T%e currerit status of techndilogy 

indicates that low density carbide fuels have promise for high burnup 

and high power density reactor applications. The major apparent 

problem involves swelling due to retention of fission gases. The 

retention of fission gas followed by a sudden release of this gas in 

the event of a temperature transient presents a potential safety problem 

that could result in clad distortion or failure. 

The multi-cavity YCRV which was used in this design consisted of a 

vertical cylindrical block of prestressed and reinforced concrete 

containing multiple cavities in which the reactor core and the steam 

generators were individually housed. The cavities were interconnected 

by passages in the concrete through which the helium coolant flowed. 

‘Lhe helium circulators, both main and auxiliary, were mounted in double 

containment penetratisns in the vessel. The general arrangement of the 

components in the PCKV is shown in Fig. 2 of Section VI. 
" Subsequent to the: issuance of WASB-1089, certain changes in the reference 

GCFR fuel elenrent design evolved, 

fuel venting as a paeans for equalizing the gas pressure inside and outside 

One change was a modified design for 

of the fuel cladding. GCFR fuel venting as first conceived2 used tight. 

seals between the fuel elemntra md the grid plate and required the 

recovery and storage of the released fission-gas effluent in gas cylindersr 

Further design iteration led t~ a proposed system in which fission gases 

would be removed from the element by a saaall differential pre'ssure and 

diPuted by 8 small 5trem of reactor coolant helium at the element-togrid-plate mnnecti e This approach would eliminate the requirement for 

t joint where the fuel element Joins the grid plate, The 

mixture of fission gases and helium diluent would be passed through a 

system of craps si Par to those used in an HTGR helium pu~ificati~n 

system, The resid 

the =in helium circulator, 

adoption of fuel ~enti~g as a reference conceptual GCFR design feature 

by GGA. 

eliminates the noble gas fission product containwnt feature regarded as 

art inportant advantage of non-vented fuel, with its attendant implications 

on other plant characteristics such as maintenance and fuel handling. 

Another major change was a lowering of the cladding hot-spot temperatures 

purified helium would be returned to the suction of 

This Ckwge in the venting design led to the 

It should be noted, hewever, that the use of vented fuel 
c 

1 

by modified core coolant orificing design and physics design including 

partial reloading. 

23 

Possible Ute mate Development Approach 

With the &creased cladding temperature ad the use of fuel venting in 

mind, an alternate development approach was chosen by the designer, 

namely, that the first GCFB facility could be a power-producing prototype plant based a8 closely as possible on WBR fuel and HTGK plant 

technologies rather than a reactor experiment. In the latter half of 

1968 a conceptual design study of a 330 We GCFR Demonstration Plant 

wm prepared. This design was based on the proposed use of fuel venting 

and a cladding hot-spot temperature of 1382'F, which was lower than for 

earlier designs. This revised new plant design also incorporated the 

use of a PCRV of the cylindrical vertical multi-cavity type being 

developed for large tiTGK plants, which is quite different from the type 

used in the WASH-1089 reference and derated designs. 

of the multi-cavity PCRV is the easier accessibility of primary systen 

6 

The prisnary virtue 

components. 

An associated draft development program plan was also prepared, based 

initially on this conceptual design. Much of the overall GCFIZ effort 

subsequent to WASH-1089 tias been addressed to the demonstration plant 

design rather than to a lo00 We plant design. 
Further Plant Design ~~if~ca~io~ 

Additional design work has been performed which is addressed to a 

7 demonstration plant design. A reference dmonstration plant design was 

CI 

adopted b9 WA in March 1970 based on a further 90'F reduction in cladding 

hot-spot taageratute (1292'F) and improved predicted plant performance, 

resulting from more sophisticated reactor physics design and from 

resuperheating the steam after expansfew through the circulator drive 

turbines. 

d 

(See Table 1 of Section IV) . 

Subsequent to completion of the 300 We demonstration plant conceptual 

design, effort has bean conducted on a more detailed ssfety analysis 

which wa8 developed for this concept in collaboration with utility 

operating engineers and nuclear safety epecfalists representing the 

utility sponsors. 

Information 

Staff including the Advieosg $: fttee for Reactor Safeguards (ACRS) 

A numbep of meetings on this document have been held curd are continuing. 

This work has resulted in a Preliminary Safety 

which is nw being reviewed by the AEC Regulatory 

c 

The latest GGA draft development program plan (Draft GA-A10788) is based 

on a systematic review of the 300 MWe GCFR demonstration plant conceptual 

fine the probl s that would require development work for a 

demonstration plat project. The scope of the plan is limited to the 

development needs of one GGHa demonstration plant by a single industrial 

partfclpant, and thus encompasses a lesser program than the work 

described in WASH-1089 which was intended to cover an entire program with 

a competitive industrial capability through full commercial introduction of 

the GCFR. 

24 
VI 

DESCRIPTPO8 OF DEMQNSTRATION PLANT CWG1PPTUAL DESIGN 

- 

The demonstration plant design is described because it is a mora recent 

v design than the latest msflable 1000 We G@FR plant design. The 

deacrlption emphasizes the nuclear stem strpplp system. The remainder 

of the plant is typical of modern high-temperature steam-turbine practice. 

The design is barsed on the utflfzation of LMFBR physics asad fatal 

technology to the extent possible and on the continuing development of 

the conrpoment technology that forma the basis of the 40 MCJe prototype 

HTOR at Peach Bottom, Pmnsylvania, and the 330 MWe Fort St. Vrafn RTGR 

Generating Station in Colorado and follow-on comercial HTGR's. Tabla 1 

of Section IV gives the principal parmeterr. 

The plant is housed in a reactor building, a fuel service building, and 

a turbine building. The reastot building contains the BCRV, which in 

turn contabs the core, the helimn primary coolant syetem and the steam 

generators. 

containment structure 

In addition, the reactor building functions BS a secondary 

also includes the fuel handling area and some 

r plant process and semfce systems. The fuel storage pool is in 

a fuel service building adjacent to the reactor building. 

The configuration of the reactor and its associated primary circuit 

components, all of which are housed within the PCIRV, is shown in Fig. 2 of 
26 

Y 

AUX 1 L I ARY 

CIRCULATOR 

XUXI LI ARY 

HEAT 

EX CH AN C E R 

THERMAL 

SHIELDING 

MAIN 

CiRCUCATOR 

PCRY 

STEAM 

GENERATOR 

REACTOR 

_- 

Figure 2 Cutaway view of TCKV showing the principal cogponents 

of the nuclear stsan supplv svstem 
27 

is section. The P is siuxilar in principle to the PCRV's that have 

already been built in Europe and Great: Britain and is also similar to SA'S 

large commercial BTGR design. 

tight. 

independent blow-limiting features where necessary to limit the 

depressurbzation rate in the ev@nt of primary barrier failure to a value 

that would not jeopardize core coolin b 

by concrete plugs designed so as to always be in compression and held in 

An inner steel liner makes the PCRV PeakThe penetrations are provided with steel closures and structurally 

The large penetrations are closed 

place by two structurally independent and redundant means. 

plugs would not normally be rembved and thus are joined to the vessel liner 

by seal welds. The liner is insulated by a thermal barrier, and cooling of 

the liner and penetrations is provided by cooling tubes on the concrete side 

of the steel liner. 

membera that are inspeeeabla and replaceable to preclude a gross failure 

of the pressure vessel. 

These concrete 

This BCRV is designed with redundancy of tension 

Secondary Containment 

To limit the conrsequeunces of a possible penetration closure faflure, 

separate secondary cantaimeat is provided, similar to that proposed for 

rcieil HTGR's. It perfoms two functions: it ensures a 

miniwra coolant pressure (-2 atm) for core cooling following an 

accidental primary system depr surization and it confines fission 

products that potentially could be released from the fuel. 

Primary Reactor System 

The reactor coolant system contains three main loops, each with independent steam generators and circulators, and three auxiliary loops, each 
28 

with its own circulator amd fieat-removal system which are used as a backup for 

the main loops for shutdown coaling. 

associated circulators are housed in vertical cavities %n the walls of 

the PCRV surrounding the reactor cere. 

of about 1250 pafa, flows downward through the core, where it is heated 

to a temperature of about 10PO°F at rated capacity. 

downward aero81 the tube banks of the helically coiled, once-through 

The steam generators and their 

The helium coolant, at a pressure 

"he flow is also 

stem generators to accotlpmodate the use of upflow boiling. 

rates appropriate reversals in the gas flow path. 

arb obtained by directing the core curit gas up through a central hole 

in the boiler and dorm through the tube bundles, up ae;a%n around the 

This neceasiThese flow reversals 

boiler shells and then to top-mounted circulators frm which ehe gas is 

discharged to the reactor top plerpltln at a temperature of about 595OF. 

Main helium is circulated by three steam turbine-driven circulators of the 

type deaveloped for the HTGW, which 

driven by a single impulse stem turbfne stage that is in series with the 

stem ganerrator. 

centrifugal cirtc~l~rors powered by separate and fndividually driven 

loss a single axial compression stage 

Auxiliary circulation is provided by electrically driven 

alternators. 

Reactor Core 

The reactor core is conpoaed of 211 hexagonal fuel and radial blanket 

elements containing the fuel and blanket rods. The elements, which are 
29 

10 ft. in total length and about 6 1/2 in. across the flats, are supported 

from a top-mounted single grid plate to which each element is rigidly 

attached; they are clamped to the grid plate solely at their "cold" 

en&. The grid plate, which consists of an 11 ft. diameter? 2 €E. thick 

disc with closely spaced &in. diameter holes to accommodate the citcufar 

extensions of the fuel elements, is itself top supported by a surrounding 

cylinder connected to the liner of the top penetration in the PCRV. The 

claraping of the fuel elements to the grid plate, as well as the mms far 

operating a variable orifice within each fuel eleraent, is facilitated by 

providing individual small penetrations above each element in the top 

access plug. 

In the reactor core there are 27 control fuel eleaents, each of which 

contains a movable control rod. Twenty-one of these rods are operated 

as conventional control rods for n017&81 control, burnup, and shutdown 

requirements; each has a reactivity worth of approximately $0.85. Some of 

these rods will normally be fully or partially inserted in the reactor 

core during power generation. The other 6 control rods provide additional 

shutdown capability and are a backup emergency shutdown system. 

rods have a reactivity worth of $1.60 each. They will always be fully 

These G 

withdrawn from the core during power generation; therefore, they are not 

subject to burnup or eo irradiation heating as are the conventional 

control rods 

All control rods are physically identical except that surface roughening 

is used to enhance the heat transfer from the 21 standard control rods. 
30 

The 6 secondary shutdown rods have smoth surfaces. 

material is B4C. 

The neutron absorber 

The worth of the rods 2s established by adjustments in 

BIO isotope enrichment. 

the mmeO 

The mechanical design of all the absorbers is J 

The absorber section of the control rod is vented to the SUPrUUnding 

coolant to provide pressure equalization, 

filter at the upper end of the absorber section allow the gas to escape 

to the fuel eleanent inlet coolant stream but preverrt particulate matter 

from feaving the absorber ssectioa. 

A vent hole and a metallic 

The coolant flaw in each fuel element is orificed so that the same hot. 

spot cladding teaPpcaratura is reached in each element, and on-line 

adjustnmt of all orifices is provided. 

Preseure in the fuel r~ds and bl at rods is equalized to that of the 

reactor coolant by collective venting, and the fissian gases pass 

through the vent nurnffold to the helium purification repseen. 

Figure 3). 

element with failed cladding. 

(See 

This system provides means €or detecting and locating a fuel 

This is done by using radioactivity 

monitors on the vent lines. 
31 

GRID 

TO 

TRAPS 

UPPER 

BLANKET 

&-- I 

/ 

Figure, 3 
Vff 

IMPACT OF MOW BECENT GCFR DEVELOPMENT 

QN WASH-1089 EVAEUATPQN 

Malor Technical Changes Subsequent to WASR-1089 

The previous discussion of GGFB developmeat effort s the issuance of 

WASH-1089 is summarized as follows: 

(a) The modification of certain system parameters notably in fuel 

cladding rmparature. 

(b) Further effart ~d fuel venting (pressure equalization). 

(c) A closer alignment of the GCFR design with the HTGR deeign to 

obtain as much benefit as possible from the Fort St. Vrafn 

project and the dmrelopruent program for comercia1 HTGR plants. 

This work has served to further identffy the basic problem associated 

with the developmeat of the GCFR. These are detailed in the draft 

development program plan (Draft GA-A10788) and arc outlined below. 

Component Development Needs 

While the develop 

the successful operation, maintrrinabflity, and reliability of Fort 

St. Vrafn (PSV) and large HTGR's, there are many major problem areas 

unique to the GCFR design that require extensive developmental effort 

f the GCFR depends to a significant degree on 

32 
33 

and proof testing, over and above HTGR needs. The fact that such efforts 

are proposed to be conducted concurrently with increased engineering scaleup efforts on the HTGR introduces special considerations. In addition, the 

use of a compact arrangenrent in the PCRV will impose design constraints 

extremely difficult and costly to handle in such first-of-akind engineering development programs. 

must be addressed in a detailed and disciplined manner based upon a 

carefully conceived plan and adequate test facilities. 

Sit& a complex development effort 

1. Contairment of the entire primary coolant system within the PCRV 

is fwdamntal to the GCFB concept. Additional model testing will 

be needed to validate the design at the higher GCFR pressure (1250 

psi vs 700 psi for HTGR), although some of the PCRV development 

needed for GCFR might be accomplished as parr of the HTGR development program, depending on exact timing and nature of needs. Such 

efforts rrmst also be integrated closely with all of the major 

component development and first-of -a-kind engineering efforts. 

2. There are a number of unique first-of-a-kind components that 

represent a significant engineering extrapolation from other firstof-a-kind components, some of which have yet to be designed and 

built and others of which have yet to be prooftested under actual 

operating conditions or operated in a reactor plant. These 

onents and their associated maintenance equipment must be 

developed, fabricated and tested along with related development 

and proof-test facilities. 
3. 

4. 

stability for the GCFR at normal load range and at flows 

as low as 2% of normal for the shutdown cooling phase has 

to be provided. 

Reactor mechanisms such as control drives, orifice drives, and 

clamping devices are unique to the GCF’R and have to be operatel 
5. 

6. 

7. 

The soccesafrrl development and proof test of the GCFR 

refueling system and components also require firat-of-a-kind 

engineering plus testing in the operating environment to 

dt?akonstrate the adequacy of the design. 

Special reactor instrumentation--some continuously operable, 

other for initial test purposes-is required to monitor the 

gas exit temperature fr each fuel element, to measure 

vibrations and other parameters in the first-of-a-kind plat, 

and to assure design performance of dl components under 

normroll and abnormal operating conditfona. 

reliability of such fnstrtmelrntatiorp in a high fast flux and 

high temperature environment pose very difficult development 

Performance and 

problems. 

The GCFR core, because of the we of gas cooling with its 

heat transfer ULmPitintBons, has a large core void fraction 

wtrdch leads to neutron streaming and leakage problems and 

may introduce problema relating to the internal shielding 

of components. 

Fuel and Core Development Needs 

Xa order to Qaxfmize the benefits to be gained from other ongoing 

activities, the GCFR effort in this category should utilize to the 

extent possible the spin-off technology from the large scale efforts 

35 
c 

1. 

2. 

3. 

36 

being carried out under the top priority LMFBR program. 

there are major areas unique to the GCFR concept that will require 

extensive and costly development. 

However, 

While fast reactor physics methods and fundamental data developed 

by the expertise and from facilities at AIL for the WBR program 

will benefit the GCFR program, considerable additional effort will 

be required to meet GCFR needs. Further work is required on 

reactivity effects due to the presence of steam introduced into 

the core as a consequence of a steam generator tube failure. 

The fuel rods proposed for the GCFR are different from those 

planned for WBR design& in having a larger diameter, roughened 

outer surface, and venting of the fuel rods and assembly. 

Surface roughening may affect the strength of the cladding 

and irradiatiort testing Will be requfred to evaluate Such effects. 

Fuel ventfnlg introduces a number of questions that would 

require substantial development effort, including the rate of 

release of fiasfon products from the fuel pellets, their diffusion 

rate through the laug 

effects of breathing at the juncture of the fuel assembly vent 

and the grid plate that occurs with changes in plant load, fission 

product plateout throughout the vent system, charcoal behavior 

of the rod to the charcoal traps, the 
4. 

37 

under fast flux irradiation, the maintenance of alignment of 

seals under bowing and vibration stresses, and lastly the 

operation of the venting system as a whole under pressure 

transients with and without cladding leaks. The vented-fuel. 

concept, which is attractive in principle for the WBR as 

well, has not been adopted in amy current reactor system. 

It presents major design difficulties during plant operation 

and shutdown and during fuel handling, and it weakens the 

defense-in-depth safety concept by removing one barrier to 

the release of noble gas firssion products. Little test 

information on this concept is available as yet and much more 

remains to be done before proof-of-feasfb3firy can be 

established, including a variety of integral in-pile proaf 

tests of prototypical fuel subassemblies and assemblies and 

safety tests relating to the vented concept, under a range 

of operating, transient and shutdown conditions. Fuel handling 

tests would also be needed. 

There are heat transfer questions to be resolved as to the 

effects of rod spacers, fuel-element box walls, and possible 

rod bowing. Knowledge of the heat transfer over the whole 

range of flow from full power conditions on down is 

required for transient and safety analysis purposes. 
5. 

6. r 

7. 

38 

Flow induced vfbratfon of the fuel rods and of the whale Puel 

element 1s a potential problem to be overcome including the 

possibility of excessive seismically induced loads. 

There are questions respecting the behavior of the interfaces 

at the spacer/fuel rod and the fael element/grid in the hot 

helium environment. 

DiffLcult problas, co to any fast reactor, are those 

related to obtaining the desired fuel butnup of 100,000 M/T 

and coping with irradiation-induced swelling and creep far all 

metal parts in the core. 

these areas. 

Much reliable data are needed in 

Safety Needs 

In addition to many of the PPQ lam that have been raised on other 

reactors, there are a number of critical safety questions which have 

been identffied for GCFR's beyond those noted above. 

1. More detailed aesurance including test data is needed to assure that 

adequate reliabilfty of core cooling in potential emergency and 

k faulted conditions can be provfded, 

to have several main loops and auxiliary loops. 

fail, shutdown cooling of the core would have to be maintained by 

the auxiliary loops, each of which would require its own electricmotor-driven circulator and water-cooled heat exchanger. Startup 

requirements and adequacy of reliability of the auxiliary coaling 

The plant design would need 

If all main loops 

loops must be further analyzed and demonstrated. 
39 

2. 

c 

3. 

4. 

5. 

Depressurization of the primary coolant system has been considered as 

the design basis for engineered safeguards in the GCFR. The maximum 

allowable depressurization rate would have to depend on some type of 

flow limiting devices in the large penetrations. 

being reviewed by the Regulatory staff ta assess whether the system 

design can accommodate the proposed depressurization accident. 

Assurance must be given that the design provides adequate margin for 

This subject is 

a more rapid depressurization and for other concurrent failures. 

Although analysis indicates that the reactivity change of the core is 

small and negative for all conceivable steam concentrations resulting 

from steam generator tube failures, additional analyses end critical 

assembly experiments must be carried out to confirm this. 

The use of a core support system in which the fuel elements are 

tightly clamped at one end into a thick grid plate with no additional 

radial restraint, along with the potential deleterious effects of such 

materials phenomena as radiation damage and stainless steel creep, has 

led to derailed questioning by Regulatory and ACRS on the integrity of 

the system and the reactivity effects under transients and earthquakes. 

These problems will have to be resolved. 

The adequacy of the proposed protection provided by control system 

actions backed up by two independent shutdown rod systems against 

anticipated transients mist be proven. The consequences of failure of 

protective action in anticipated transients must be analyzed to 

establish their acceptability. 
40 

kvelopment Approach 

The current GCFR draft development plan proposed by GGA does not call for 

many of those aspects of the reactor development program identified in 

Figure 4. For exBuoPple, past experience in other reactor progr- has 

demonstrated the vital need for a relatively small reactor experiment in 

addition to other teat facilities to provide an essential contribution to 

the evidence of technical feasibility, to facilitate meaningful evaluations 

prior to a commitment for a large scale development program and a demnstration plant program. Cansliderable weight nust be given to the many problem 

area8 which historically have been discovered and solved as an outgrowth of 

specific operating experience on such experimental reactors. 

are the many problems on saall plants such ci8 STR, EBWR, HRE, SIR, SRE, 

Crases in point 

Saxtan, EBR-I, EBR-11, Hallam, Fermi, MSRE, EBOR, UIPTREX, EGCR, Piqua and 

BONUS, as well as the British experience with Dounreay. 

WASH-1089 pointed out the need for a GCFR reactor experiment (1) to provide more 

informtion concerning the physics and safety of GCFR's, (2) to assist in the 

developsamt and testing of GCFR components and (3) to provide a facility for 

fuel element dewlopaant testing. Experience since publication of WASH-1089 

appears to confirm &e fact that a sound engineering approach requires a GCFR 

reactor experiment along with other related efforts outlined in Figure 4 to 

help establish the technical feasibility and requisite technical basis for this 

concept prior to full scale comitmnt to a demonstration plant program. 

Moreover, such accomplishmnts appear to be eoaential to permit meaningful 

evaluation or acceptance by the utility industry and others. (See Figure 5) 
8 4i 

b. 

v) 

w 

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bc 

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4 tf h 

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cx 

a 

v, 

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VEX1 

CONCLUSIONS 

Progress has been made in further defining 

since the anal is reported in WA§H-1089. 

the problems of the GCEX concept 

While much of the privatelysponsored effort ha5 been abed at a 300 Ne demonstration plant, the 

overall full-scala plant system design has r-tned largely unchanged and 

the major developmental and safety issues specific to the GCFR as ourlined 

herein requltre resolution, The further definition of these problem areas, 

aided by experience to date in the LWR, HTCR and LMFBR programs and 

discussloaa with the Regulatory staff and the ACm, has served to 

reinforce the earlier conclusions that a substantial program af development and proof testing would be required to resolve adequately all of the 

major autstrndfinlg techndcal issues on this concept. 

Bared upon the information currently available, it is concluded that ow a 

relative baais there hava been ne signifbcant changes in the GCFR devalopsent statua to justify apprecjtably altering the conclusfom of WASH-1089 

that considerable research and development effort muld be needed to provfde 

the required aseurance of safe, reliable and economic operation at the 

design perfsmancs levels, 

outlined fot the 

test facilitfes, critical experiments, a reactor experiment affording 

extensive in-pile fuel testlag capability, safety tests, and a dmonstratfon plant program would be required to bring the GCFR forward as an 

option in the coarcarcial rtage. 

A measured, orderly program similar to that 

B including fuels and materials testing, component 
44 

The GCFR concept Is at an early stage of development with respect to tts 

engineering base, reactor operating experience, industrial participation 

and cammitmeut, and funding level. ft does not have the benefit of 

considerable existing fast: reactor experience both in the United States 

and abroad, such as enjoyed by the LMFBR. 

the! basic and major technological pp~~blt~t~~ could be developed, achievement 

Assuming that solutioacs to 

of the potential of the concept as a commercial power plant would require 

a greatly expanded Cove~wr~ernt effort with significant industrial participation and comitaaent. 

If additional 

introduction of the WBR were to be made available, they could be emplayed 

in a worthwhile manaelc to mve toward a GCFR experiment and other key test 

facilities to help establish the technical basis prior to camm3tting to 

conduct a parallel breeder development program on the GCFR. 

funding for such a program described in earlier reports would need to be 

reevaluated on a current basis, raking into account status and commitments 

on the EETGR, LWB and LWBR program. 

esources beyond those needed for the development and comercia1 

The timing and 

In the absence of any additfonal resources, the GCFR technology will continue 

to benefit from the substantial efforts on the LMFBR and HTGR technologies, 

as well 88 the ongoing development work on the GCF'R sponsored by the AEC and 

private industry. The cumulative direct expenditure on GCFR development by 

through PI 1972 ha8 been approlrirastely $9.0 million, while private 

support for the concept frm utilities over the four year period through 
FY 1972 has been $5.5 

few years for increasing the degree of effort 0x1 the GCFR ahbuld 

aatianal energy priorities a~d the degree of progress on the LWB, 

LMFBR and IFTGR so dicta$e* 

kPicm, The optton remains open over the next 

45 
44 

IX 

REFEREHGES 

&35eSS1~@nt ofEnergy Technologies - Report on Step 1, AEC-7, 

November 1971 e 

Staff Report, A 1000 We Gas Cooled Past Reactor Study for the 

Alternate Coolant Twk FOPC~, USAEC Report GA-7804, Genera1 

Atomic, July 21, 1967. 

1. 

2. 

3. General Atomic, Addendum to 1000 Wt? Gas Cooled Past Reactor Study 

€or the Alternate Coolant Task Por~e, USALSC Report GA-7804 (Add.) 

Sept. 28, 1967, 

4, General Atomic, Addendum 2 to PO00 We Gas Cooled 'Fast Reactor 

Study for the Afternsnts Coolant Task Force, USAEC Report GA-7864 

(Add, 2), November 12 1967, 

5, General Atomic, Preliminary Development Plan for the Gas G05led 

Fast Reactor (GCFHZ), GA-8257, Oct. 1967, 

Gulf General Atomic, Design Study of a 330 MWe Gas Cooled Fast 

Reactor Demonseration Plene, Feb. 1969 (private G.A. date) 

6. 

7. Gulf General Azomic, Design sf a Nuclear Stem Supply System for 

GCFR, GA-10064, May 15, 1970 (private GOA. data) 

8. Gulf General Atomic, Gas Cooled Fast Breeder Reactor, Preliminary 

Safety Information DocwOatp GA-10298 (2 Vols.), Feb. 15, 1971.