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;ffDATE- June 9, 1967:
(fifiSII REICES

   

SAFETY PROGRAM FOR MOLTEN-SALT BREEDER REACTORS

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Investigations required ‘in determining the safety characteristics

- of MSBR power plants are outlined, and associated safety program cost -
é,festimates are given.
~ in the MSBR are described; the favorable characteristics arise from the
ri;;;prompt negative temperature coefficient of reactivity, the low system
- pressures, the mobility of- fluid fuel, and the low excess reactiv1ty
‘available to the reactor at any time.-

The safety: features of the major plant systems

‘Reactivity additions which need

itfifdetailed study include those assoclated with net fuel addition to the

»lifcore region, those due to graphite behavior, those caused by changes

" ’in fluid flow conditions; and those due to control rod movement. Re-

Jlrjactivity coefficients which: require evaluation include those assoc1ated

:*7:wa1th temperature, voids, pressure, fuel concentration, and graphite con-

flr;ffcentration.
- conditions and also under circumstances where reactivity itself is not

‘The integrity of plant containment under’ react1v1ty inc1dent -

- involved need to be evaluated; included here are evénts such as mixing

- of water and steam with. coolant salt criticality in regions outside the
-.core, and flow blockage Within the- fuel or .coolant streamsi_

Stability .

.';siffanalysis of the reactor plant is required to determine the operating,
. ...control, and/or design requirements for Obtaining satisfactory plant

'fi'f"characteristics

Phy51cal behavior of materials and of equipment under

. _MSBR conditions, as _they relate to reactor safety, need to be determined

h_l;experimentally.

In order to dellneate and resolve the basic safety prob-

 lems associated.w1th MSBR- systems, it is ‘estimated that ébout $1.3 million

i required over = peried of ‘about eight years, with most of the effort

:i“;q;($0 9 million) occurring during the first four years._f'f“; | -

“UT":E Thrs documem confcms rnforrnahon of a prehmmory nalure

and was prepcred primerily for internal use at the Oak Ridge National - _
Laboratory ‘It is subject 1o revision or correction and therefore does

not represent a fmo! reporf T _ . THIS DOCUMENT HAS BEEN REV]EWED

NO INVENTIONS OF BATEHT IHTL.RL'ST

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~LEGAL NOTICE

This report was proparod os an account of Government sponsored work. Neithor the United Shhs, '

nor the Commission, nor ony person acting on behalf of the Commission:
A. Mckes any warranty or representation, expressed or implied, with respect te the accuracy,

completeness, or usefulness of the information contained in this report, or that the use of -

.any information, apparotus, mefhod or process ‘disclosad in this report may net lnfringo

privately owned rights; or

B. Assumes any licbilities with respect to fho use of, or for damages usulflng from the use 0‘-,-."

any information, apparatus, method, or process disclosed in this report.

As used inthe above, * person acting on behalf of the Commission® includes any employee or

controctor of the Commission, or employee of such contractor, 1o the axtent that such employes

- or contractor of the Commission, or employee of such contractor prepares, disseminates, or
provides access to, any information pursuant fo his employment or contract with the Cemmnulon, o

or hls -mploymom with such contractor,

 

 

 

 

 

 

 

 

 

 

 

 

 
 

TABLE OF CONTENTS

Page

| 1. INTRODUCTION . « & v v ¢ v 4 o 4 o o v o o o o v v o o o & 3
| 2. MAJOR PLANT SYSTEMS INFLUENCING REACTOR SAFETY . . . ... 6
| 2.1 Reactor System . . . ¢ & . 4 4 ¢ v 4 4 e 4 4 e e e T
2.2 Steam System . . & ¢ v ¢ ¢ ¢ v ¢ e v 4 e e e e e .. 10

2.3 Fuel Recycle Processing System c 4 s e s s e s e s 13

2.4 Off-Gas System . . . v v v 4 4 4 e e e e e e .. 15

3. REACTOR SAFETY ASPECTS . + & « ¢ + & « « o « o o« « « » « « 18

3.1 Reactivity Coefficients and Kinetics Parsmeters . . . 18

. . . O
5 02 ContI‘Ol _ROd Func tiOIl . . * o‘ . . . . . . . . . 2

7

oA | ce .. o2l
whe 5.3 Reactor Incidents . . . . . . . . ¢« « s ¢ o .

 

 

 

 

‘;‘ . . » » . - . 2
5 . 3 . l Reactiv1‘ty‘ Additlons - . . . . . » - . 3
3.3.2 Mechanical and Physical Integrity - 29
C on-tainment ® & & & s e .8 a4 @ ® e & e 9 o e
t . M . . . . 5 1
5.3.3 Miscellaneous Incidents . . . . .
’ - - . . e . . * 5 2
3 . LI' Reactor Stability . '.o ., . .. e 9 . e e . _ .
7 - | : N » o ’ e » 8 a # e s s o *. = ’ . . 5 5
,'l'l MSBR SAFETY PmeM . . L] » . - .
- . | 7 ' ) ) - * o o+ e w s s e s 33
b1 Summary . . ... ... St |
. | o e S - L .- L. 3 & -
7 ) . ) . o o . ¢ ¢ 0 e e o . 0 l' e 3 8
ACKNOWLEDGMENTS .« « v v v v o v v o . Trrrrrr
. LEGAL NOTICE |
This Teport was prepared as an sccount of Government sponsored work, Neither the United
Btates, nor the Commission, nor ARy person acting on behaif of the Commission .
A, Makes any Warranty or representation, expregsed or iraplied, with respect to the accu-
TAcy, completeness, or usefulness of the information contained in this report, or that the use
-of any mformation, Apparatus,, method, or process disclosud in this Teport may not infringe |
. privately owned rights; or - : .
- i B. Assumes any labliities with Tespect 1o the use of, or for damages resulting from the
: : Y , ; use of any information, apparatus, method, or proceas disclosed in thia report, i
; : . b _ As used in the shove, *‘person acting on behaf of the Commisaion** fncludes any em-~
- .k ;‘ployoe or contractor of the Commissio '

n, or employee of such contractor, to the extent that
i: such employes or contractor of the Commission, or employes of such contractor prepares,
- disseminates, or provides access to, any fnformation pursuant 1o his employment or contract
with the Commisaton, or his employment with guch contractor, ;

-

 

e OF THS DOCUHERT, 15 ONOMTED
DISTRIEUTION OF THIS O

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SAFETY PROGRAM FOR MOLTEN-~SALT BREEDER REACTORS

Paul R. Kasten

I. INTRODUCTION

The purpose'of this report is to discuss important aspects of
molten-salt breeder reactor-plants which are related to the opera-
tional and ultimate safety of such systems, and to present a program
for investigating reactor characteristics and associated cost require-
ments. In order to be relatively specific, the Molten Salt Breeder
Reactor plant (MSBR) described in Ref. 1 forms the basis for this dis-
cussion. However, general studies which also consider other design
concepts will need to be’performed' the general studies required will

come into better focus as MSER safety and design information is

_developed.

Briefly, the MSBR design concept concerns a two-region, two-fluid
system with fuel salt separated from the blanket salt by graphite tubes.
Circulating-fuel temperatures are high (~1300°F), and reactor pressures
are low (~100 psi)r The energy produced in the reactor fluid is trans-
ferred to a secondary coolant§salt-circuit, which couples the reactor
to a supercritical steam cycle. The fuel salt consists of uranium
fluoride dissolved in a carrier salt containing a mixture of lithium
and beryllium fluorides, while the blanket salt contains thorium fluo-
ride dissolved in a similar carrier salt. The blanket salt also cir-

culates through passages in_the;graphite moderator region of the core.

"The coolant salt is‘a mixture'of sodium fluoride and'sodium fluoroborate.

Fuel processing is performed: on-site, in a processing plant integral

-with the reactor plant - Figure 1 gives a flowsheet of the 1000-Mw(e)

MSBR power plant, ‘while. Figure 2 gives the associated processing flow-
sheet. Details of these flowsheets are discussed in References 1l and 2.
The safety of MSBR's has not as yet been investigated in detail

-[however, it can be .discussed in e qualitative manner, p01nting out areas o

and 1tems,which need to be investigated. The operating philosophy and
 

 

 

 

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REACTOR VESSEL o S48 -
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f . 100874 _ j_usu | | 1518.5h-540p -1000°
|
1918 58 ! r i H24h-35|5p-|000‘ i
B | 3600p- 000°*F | 24 E . HP GEN.
H 150 1%s0¢ {2011 Ysee i N } tureine [1'° TURBINE 527.2 Mwse
Fl95.7 1teee o BOILER | | oo Gross
R0\ _100°F { | SUPERHEATERS | seinl 4 __,
100 g | ¢ 4 Jeso°F oot r |
BLANKET SALT HEAT COOLA ! 85I7° L |
EXCH. AND PUMPS | PU:LSSA'-T | P 3 GEN.
‘ FUEL SALY HEAT o TURBINE [ TURBINE 5°Té'r’m“"
H25*F EXCH. AND PUMPS 13092} E | | | _
1125 F ' o
‘ beior REMEAT STEAM | |
AL PRENEATERS | (ot ¢ |
i T CONDENSER 8 FEEDWATER
o | 300p-T00°F| | poren| _ SYSTEMS
I | J | -" | . 1 I 759.2'! | 5 ' '. 35009"55'F | 3500 .
T Y "= o3ysy ! _,\__'____ _l - 2H] el
- e ——- e T66Ah L——_—*"# |
- SR ] | | __ Pumes S e -
| (- FUEL SALT  (COUNT SR . DoREoRwawce :
tboeny  \—ODAIN TANKS ~ \_DRAIN-TANKS | CONET OUTRUT - 1000 Mws
LEGEND R | S , ~ GROSS GENERATION 10349 v
CORUEL — S " ‘ . BF BOOSTER PUMPS - 92 Mwe
. BUANKET mmaem - o - STATION AUXILIARIES 257 Mwe =~
" COOLANT =—eoom U DR : - .- REACTOR HEAT INPUT 2.225 ¥t
| smu ,.A.'.....\.'_;:;.-g_-. AT o R o - NET HEAT RATE 7601 Bte/kwh
Mo T o o " NET EFFICIENCY ML
Y SR S |
Pareemaeftit :
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O ...Fraeze Valre | , |
'Fig. 1. Flowsheet of a 1000-Mw(e) MSBR Power Plant.

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“ Fig. 2. Fuel- and Fertile-Stream Processing for the MSBER and MSBR(Pa).

 
 

the organization for safety in MSBR power plants will have to satisfy =~ &“;
the licensing and regulatory requiremenfs which exist; also, MSBR plants
must satisf&ctorily pass safety feviews, inspections, and testing,
Plant operations will have to be safe and efficient so that the health
and safety of plant personnel and that of the general public will not ;
be endangered, and so that thé plant can operate economically on a long-
term basis. While it appears that the safety of MSER systems can be’
assured at costs as low or lower than the-safety-requiremsnt-cOsts of
other feactor power plants, a definitive evaluation cannot be made until
detailed safety studies have been performed. |
In discussing MSBR safety, credible 1ncidents which would normally
never occur must be considered. Plant systems involved are primarily N
the reactor system, the supercritical-steam system, the fuel processing
system, and the off-gas_system.' These are discussed below relative to
their influence and function on reactor safety. Also, a discussion is
includéd of possible events which can be described qualitatively, but |
which need detailed investigation to be evaluated adequately. These .
involve reactivity coefficients, control rod function, possible inci- fi#
dents, and reactor stability. Finally, a summary is given of the MSBER
safety program, along with estimates of the costs associated with re-

solving safety design questions.
2. MAJOR PLANT SYSTEMS INFLUENCING REACTOR SAFETY

The reactor system is the primary one of interest, but other systems
can also influence reactor behavior. For example, rupture of the supsr-
critical boiler-superheaters could lead to high pressures in the secondary
coolant system, which in tumm could lead to rupture of the prifia:y heat
exchanger if proper safeguards are not employed. Such a train of events
would influence the reactivity of the reactor core, and need to be cen-
sidered relative to the adequacy of reactor plant containment _ .

Another plant system of importance is the fuel recycle system, since
it is integrated with the reactor plant and operates "on-line." This

operation could introduce reactivity changes into the reactor system.

 
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'Also, the off-gas system is an important protective system relative to

| the reference of radiocactive gases from the plant site.

-~ 2.,1. Reactor System

As considered here, the reactor system contains the reactor core,

the primary and secondary circulating-salt loops, and associated pumps‘

'vthe_heat transfer equipment. Important!items in this system are indi-

cated in Figure*}
The reactor vessel is housed in a circular cell of reinforced con-
crete, about 36-ft-diam by 42- -ft-high. This volume also contains the four

fuel- and blanket-salt primary heat exchangers and. their respective cir-

_culating'pumps;tr The wallrseparating this cell from the adjoining cells

is L-ft-thick, and the removable bolt-down roof plugs totel 8 ft in

‘thickness. The pump drive shafts pass through stepped openings in the

special concrete roof plugs'tolthe drive motors which are located in
sealed tanks_pressurizedlabore'the'reactor cell pressure. The control
rod drive mechanisms.paSSfthrough the top shielding in a similar manner.
The coolant-salt pipes passing through the cell wall have bellows seals at
the penetrations. | , | N

The cell is lined with l/h to 1/2-in.-thick steel plate having
welded Joints, which, together with the seal pan that forms a part of
the roof structure, provides a cell leak rate less than 1% (volume)
per 24 hr. The cell is heated to above 1050°F by radiant heating sur-
faces’loCated at the hottom'of thefcell. The liner plate and the con- -
crete structure are protected from high temperatures by 6 in. or more
of thermal insulation and by_a.heat removal system. The reactor and
heat exchanger'support“structures are cooled as-required.

Thus, there are several barriers to protect against the escape of

radioactivity The first is the primary reactor piping and equipment

the second is the seal-welded containment vessel, and a thlrd is the
~ reactor building proper which is maintained at & negative pressure by

: ventilating fans which discharge through a stack-filter arrangement

All penetrations into the reactor cell, such as those associated with
instrument, electrical, and service lines, are equipped with sealing

devices.
 

 

 

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Fig. 3. General Arrangement of Equipment in the Reactor Cell

and Coolant Cells.

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The four cooling-salt-circulating.circuits are houeed-in individual -
compartments heving h;ft#thick reinforced‘concrete walls and bolted-do#n;
removable roof plugs. Each compartment contains four boiler-superheaters,
two reheaters, one coolant-salt pump serfiing'the boiler-superheaters, and
one coolant-salt pump_supplying the reheaters. All pipingrpassing into
these cells from the tfifbine_plant has-sealed'penetrations and valving
located outside the walls. The coolant-salt‘pufip drive_shafts extend -
through the roof plugs and the cells are sealed and heated in the same
manner as in the reactor cell., Normally the temperature need not be
maintained above T50°F, however. | |

The secondary .coolant lines are maintained at a higher pressure than
the reactor system (about 200 psi, compared with ~100 psi in the reactor),
so that in the event of a primary heat exchanger tube failure, leakage of.
radiocactive fuel salt-ihto the secondary circuit will be minimized. Ordi-
narily, the activity of_the;éoolant salt will be that due to N*° (formed
from the N,a reaction on fluorine and having & half life of 7.k sec) and
Na®% (formed by an n,y reectiOn'and~having”a half life of about 15 hr).

.In each case the neutron source for activation is the delayed neutron

emission in the primazy,heatiexchanger.

The'design~pressure-for the reactor cell and the four adjoining
compartments is expectedlto:be-about 45 psig.'.Pressure-suppression
systems are provided, the reactor cell system being separate from the
system used for the other compartments. These suppression systems would
contain water storage tanks sofithat.vapors released into a cell would

pass through'these.tankswendfbe7conden5ed,maintaining the.cell pressure

below the design value. Noncondensable gases would be contained until
- they could be'disposed“ofrby5paesage through the off-gas system. When the

~coolant salt is discharged into the water in the pressure suppression

system some HF will be produced, The quantity and the effects need to

| be evaluated. Studies made for the MSRE suggest that corrosion of the
‘steel llners and tanks by the HF will not be & ‘serious problem.

The fuel drain tanks malntainisubcritical storage of the fuel and
also remove decay heat for maintaining proper fuel temperatures. Evapora-
tive cooling is provided. The coolant drain tank is similar to the fuel

drain tank except no cooling is required. . An inert cover gas system is
 

 

10

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- provided to protect the molten salt from oxygen and moisture at all
times. In order to keep stresses within‘eQuipment low, normal heating
end cooling of the reactor will be done slowly at rates of 100°F/hr or
less, applying temperature differences less than about 100°F. However,
the reactor system should withstand several severe thermal shocks (such
as & rapid fuel-salt temperature rise of about 4O0°F) without'breaching.

The homogenebus and fluid nafiure_of molten-salt fuels permits réady
transport of material from one system to another.. From the viewpoint of
safety, it is important that the fissile fuel'remainhomogeneqqsly'diétri-
buted in the carrier salt. This has been demonstratedrrepedtedly‘under
both nonirradiation and irradiation conditions; in additibn,'chemical
stability of the fuel salts improves with increasing temperatufe,'a
favorable relation. Also, the fuel salt expands with increasing tempera-
ture, effectively leading to expulsion of fuel from the core region and
leading to a negative temperature coefficient of reactivity. Because of
the ease of fuel addition and removal, very little excess reaétivity is
provided within the reactor during normal operating conditions. - o s

Fission gases are continuously removed from the reactbr"core on &
very short cycle time (less than one minute) by sparging the salt with -
inert gas. Fuel processing takes place on about a 30-day cycler(for the
fuel salt), so that the fission product content of the reactor.systém
is always relatively low.

Since the fuel salt does not wet the container material or the moder-
ator, drainage of the fuel salt plus flushing the system with carrier
salt should remove a large fraction of the fission products from the
circulating-fuel system. The actual behavior will need to be studied
experimentally. '

2,2. Steam System

The steam system is indicated in Figure 4 and consists of the

- coolant-salt heat exchangers, boiler feed pumps, feedwater heaters, the
'turbine-generator, and associéted equipment. The steam~power system
uses Steam conditions of 3500 psia --.1000°F/1000°F, which are repre-

sentative of modern steam power plant practice. The feedwater enters |
 

 

 

 

 

 

 

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12 | | |
" the boiler at T00°F so that little or no freezing of the secondany f : \o/
coolant salt takes place. | |
~ The 16 boiler-superheaters consist of U-tube-U-shell hea't exchangers,
| which transfer heat from the 1125 F coolant salt to the 700°F feedwater
and generate steam at71000°F and 3600 psia. Variable-epeed, coolant-~

{

salt pumps are used to permiticentrol of the outlet steam temperapuré.
There arejeiéht shell -and-tube heat exchangers which'function as re-
Jheeters and transfer heat from the coolant salt to 570 psia steam from
the highfipressure turbines exhsust, raising its temperature to 1000°F.
Reheat steam preheaters are used to heat this exhaust steam to. about
600°F before it enters the reheaters. - S .
The heat-exehange equipment is located within containment cells

which communicate with the reactor cell by means of coolant-salt lines,

and with the turbine room by means of steam and water lines. In addi-
tion, these eells communicate with the fuel proeessing area by'means of

- small coolant-salt lines and with the control and service areas through

penetrations for gas, cooling water, instrumentation lines, etc. These ¥
cells also communicate with a vapor suppression volume through a large | |
conduit equipped with a rupture disc. The vapor-condensing system pro-
vides pressure control of the coolant-salt cell in the event Qf'a rupture
of the_steam or water circuits. Biological shielding is provided.for the
cells, and a controlled inert gas atmosphere is maintained. |

Molten salts do not undergo a significant chemical reaction with

water; however, high-temperature steam is produced when water contacts
molten salt. In order to provide for accidents producing steam, or for
leakage of high-pressure steam into the coolant-salt cells, a vapor-
suppression system is used to provide pressure relief, and maintain -
pressures below the containment design value of about L5 psig. Auto-
matic block valves are provided in the steam lines to reduce the likeli-
hood of draining the water in the steam system into these cells in the
event of a rupture.

_ To protect against high pressures in case of failure of a super-
heater tube in the heat exchanger, rupture discs are provided on the

shell Side'of the superheaters‘and»reheaters for venting the coolant

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15

system into the vapor condensing,system. These rupture discs protect
against overpressure in the coolant-salt circuit and thus protect the
reactor'system; which is separated from the coolant salt by the tube

walls of the primary heat‘erchanger.

2.3. Fuel Recycle Processing System

 

\ The floWsheet for the MSBR processing system has been given pre-
viously in Figure 2. The core fuel is processed by the fluoride vol-

| atility process to separate the uranium from the carrier salt and fission

products. The valuable carrier selt 1is separated from the rare-earth
fission products by_the vacuum-distillation process. The fuel salt is
reconstituted by absorbing UFs in uranium-containing‘carrier salt,
followed by reduction in the'liquid phase bycbubbling hydrogen through
the melt. Excess uranium fromithe‘reactor is sold as an equilibrium
mixture of the fuel isotopes. Fuel salt is returned to the reactor as
needed. | - '
The blanket salt is processed by the fluoride volatility process
along with a Pa-removal process in which Pa is extracted by liquid bismuth
containing dissolved thorium. The same process also removes uranium.
Small side streams of fuel salt and blanket salt are continuously
withdrawn from the reactor circulating systems and. routed to the process-
ing plant located within the same building. At the same time, makeup

streams. are returned to the fuel and blanket systems at the same rate

they are removed. Theserratesiare:lowVenough thatino significant reac-

'tivity additions to the reactor should normally be possible.

The fuel-recycle processing‘plant is located in two cells adjacent

,'to the reactor shield one contains the high radiation 1evel Qperations,r
“and the other contains the lower-radiation-level qperations. Each cell
'_is designed for top access through a removable biological shield having
‘& thickness equivalent to 6 ft of high-density concrete. A general plan
'of the processing plant and a partial view of the reactor: cell is shown

~in Figure 5. The highly radioactive operations in the fuel-stream proc-

essing are carried out in the smaller cell (upper left). The other cell
houses equipment for the fertile stream and the fuel-makeup-stream

operations.
 

14

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15

The highly radioactive.cell contains only fuel-stream processing
equipment consisting primarily of theffluorinator, still, waste receiver,

NeF and MgF, sorbers, and associated vessels. The other cell houses the

‘blanket processing equipment,and fuel- andlfertile-stream makeup vessels.

The processing plant will use hydrogen and fluorine gases in the
treatmentrof the salts. Care must be taken in utilizing these gases
because of the hazards assoclated with obtaining explosive mixtures of
hydrogen and oxygen, or fluorine. Thus, hydrogen must be isolated from
the fluorine and from the reactor cell. Also, fluorine must be isolated
from the reactor system, and:organicgluhricants must not enter the fluorine
system. - | | ‘

The processing plant will utilize the same off-gas disposal system
as the reactor plant. This combined use should not introduce operating
hazards. The integrity of the cooling systems needed for cooling of
processing equipment must be assured, both during continuous processing
and during storage of waste. | o

Criticality considerations must be considered, such that recovery
of fissionable material constitutes no criticality hazard; however, due
to the relatively small quantities of fissile fuel held up in the proc-
essing plant and the character of the materials handled, no difficulty
is anticipated. |

Reactor fuel additions will be done primarily through the return
line from the processing-plant. The associated components would be of

all-welded construction and would be maintained by remote maintenance

‘procedures.

‘g 2;4.s OfffGas.System

‘ Xenon and . knypton as well as tritium are stripped from the fuel

,salt in the reactor circulating system by sparging with an inert gas,

such as helium. This gas along with. the gases generated are treated

in the off-gas system. - | |
The flowsheet for the off-gas system in shown in Figure 6. After

pass1ng through a decay tank, the fission product gases are passed

through water-cooled charcoal beds where xenon is retained for 48 hr.
 

 

 
 

Cm OUnLETY

ORNL-Dwg . -67-6316

 

\‘\

49 hey MOLDUP NENON WATER
~ COOLED CHARCOAL ADSORECRS

Fig. 6. Flowsheet for the Off-Gas System.

 
 

 

 

 

»

i)

i

17

In addition to removing the *®BXe, this system of circulation effectively
transfers arlarge fraction of the other gaseous fission products to areas
where the decay heat can be removed more readily. '

About 0.1 scfm of the gas stream leaving the initial charcoal beds
(or 0.4 scfm total for the four fuel-salt circulating loops) is passed
through additional charcoal beds and then through a molecular sieve
(operated at liquid nitrogen temperature) to remove 99% or more of the
85Kr and other gaseous,products. The effluent helium can be recycled
into the reactor system or passed through filters, diluted, and dis-
charged into an off-gas stack. . The molecular sieves canh be regenerated,
and the radioactive gases that are driven off can be sent to storage
tanks.

Concentration and storage of the tritium will probably require
additional equipment; this operation needs additional study.

A helium system provides cover gas for the blanket pump bowls,
the drain tanks, fuel-handlingtand processing systems, etc. Essentially
all helium will be recycled to the cover-gas system. Any discharged |
cover gas passes through charcoal adsorbers and absolute filters, is
diluted with air, and discharged through the off-gas stack.

Relative to the off-gas processing of the fuel recycle system, most
of the facilities are located 1n the processing plant proper. In the
processing plant, off-gas comes primarily'from the continuous fluorinators,
while smaller amounts are formed in various other processing vessels. The
gases are processed to prevent'the'release of any contained fission prod-

ucts to the atmosphere. . Excess fluorlne used in the fluorinators is re-

'cycled through a. surge chamber by a. pos1tive displacement pump, and a
small side stream of the recycling fluorine is sent through a caustic

scrubber to prevent gross buildup of fission products. Each of the

'processing vessels and holdup tanks has off-gas llnes which lead to the

"scrubber for treating HF, fluorine, and volatile fission products.

The scrubber operates as’'a. continuous, countercurrent packed bed

with rec1rculating aqueous KOH. A small side stream of KOH sOlution'is

~ sent to. Waste, and the scrubber off-gas is contacted with steam to hydro-

lyze flss10n products such as tellurlum. A filter removes the hydrolyzed
products. The noncondensable fission products are sent to the reactor

off-gas facility.
 

'718 :

The off-gas gystem must be designed to handle the very radioéctive ,
gases and to provide'cooling of these gases. Also, while the wvapor
pressure of molten salts is very low, MSRE experience;indicates that
some particulate matter can be carried into the off-gas stream. Cold
‘trapping or.filtering must be provided in the off-gas lines for removing .
these mist-like'particles. Any oill leakage and -associated decomposition'
‘prbducts entering the off-gas system must be removed by a filter system.

The off-gas system primarily removes fission products, recirculates
sparge gases back to the reactor system, and holdslup fission products
until they have decayed sufficienfly for disposal. If fission products
are not held up sufficiently, radioactive gases are discharged prematurely,
leading to high activity levels.

3. REACTOR SAFETY ASPECTS

In operating a reactor power plant there always exists the possi-
bility that reactivity can be inadvertently added to the system, lead-
ing to a system disturbance. If this disturbance is very small, no ill
effects result. Increasing the degree of disturbance can lead to con-
ditions which affect reactor operation (operating safety) and eventually
to conditions which affect the safety of the general public (ultimate'
safety). In this section the MSBR operations are discussed from the
viewpoint of items which need to be evaluated from a safety standpoint
such as reactivity coefficients, control rod funétion,VPOSSible reactivity
events that could cause reactivity additions to the reactor, and thé
stability requirements of the reactor power plant. In general, the
specific situations which need to be evéluated are dependent upon the

design and operational features of the system.
3.1. Reactivity Coefficients and Kinetics Parameters
A number of reactivity coefficients are associated with an MSER

system. These include those associated with temperature, voids, pressure,

fuel concentration, graphite concentration, xenon concentration, fuel
 

 

 

 

ȴ

.

19

burnup,‘fuelrflcw'rate,:and involve the fuel and blanket fluids separately

and together. From the'viewpoint of‘reactor?safety, the most important

coefficients,appear to be the temperature coefficients of reactivity for

' the fuel salt, the blanket salt, and the graphite moderator, and the fuel

concentration coefficient of reactivity. There are special circumstances
where others are also of importance. All of these need to be determined
specifically. o |

Molten-salt reactors have, in general, a relatively large negative
fuel temperature coefficient of.reactivity, due to the expulsion of fuel
from the core region with increasing temperature; :The'value for MSBR
systems will be in the range of =1 x 107® Ake/°F to =5 x:107F Ake/°F,'the,
value being a function of design and operating conditions. This coef-
ficient gives inherent control and safety to molten-salt systems, since
any increase in power level tends to decrease the reactivity and thus
decrease the power level. Since MSBR's will normally operate with only
low values offexcess'reactivity available, the temperature coefficient
appears sufficientcfor controlling the reactor without excessive tempera-
ture variations. This inherent control feature permits use‘of control
rod mechanisms which have relatively slow action. -~ ' '

- Increasing the prompt temperature coefficient of reactivity generally
improves the safety and stability margins of reactor operations, provided
that-the reactivity is added by means other than the temperature coeffi-
cient. HOyever, the temperature coefficient itself can add reactivity

by means of '"cold slug" type occurrences. . Such an occurrence in an MSBR

- would be normally associated with en increase of fluid flow rate; however,

1ncreasing the flow rate tends to- decrease reactivity due to the associated

‘increased loss in delayed neutrons. The effective value for the delayed
‘neutron fraction in 223y. fueled reactors 1is about 0.003 in fixed fuel

systems; in MSBR systems, the effective value for beta durlng fuel cir-
culation would be about 0. OOl. e o T | '
Reactivity coefficients need to be determined in order to prqperly

"'evaluate the safety of MSBR systems.» Primary values appear ‘to be the

temperature coefficients associated with the fuel and blanket- fluids and
with the graphite; the void coefficients associated with both the fuel
 

 

20

and blanketrfluids; concentration coefficients associated with the fissile
and fertile salts in the core; reactivity coefficients associated with
loss of-fuelfflow;,effective,delayed neutron fraction as a function of
- flow and power conditions; and the reactivity effects associatedrwith
g&aphite shrinkage, graphite breakup, and fuel soakup by graphite.

The reactivity cdefficients need to be consistent with the kinetics

model used in the:safety evaluations, and time- and space-dependent criti-

cality effects need to be included in such studies. These time- and’
space-dependent effects should include consideration of the'different
heating and flow, rates within the reactor, afterheat generation, and the
change in the effective delayed neutron fraction during a power pulse..
Other parameters needed in the kinetics“analysis include the prompt

. neutron lifetime and xenon poisoning effects.

3.2. Control-Rod Function

 

One or more control rods are provided in the MSBR in order to provide
flexibility in reactor operations, and to control reactivity additions
such that fuel temperatures and associated temperatures do nbt become
excessive. As mentioned in Section 3.1, inherent control is provided |
through the negative temperature coefficient of reactivity, which pro--
vides prompt protection ageinst reactivity additions. At the same time,
if reactivity additions take place over a long-time interval, the total
regctivity added may lead to undesirably high fuel temperatures if'only
the temperature coefficient is utilized (hcwever, such temperatures may
be permissible for relatively short times -- order of hours). Installa~
tion of control rods which are slow acting (response time of about one
second) appears sufficient for controlling maximm fuel temperatures, and
would pemit reactivity control independent of fuel temperature. Control
rods provide an easy means of controlling reactor power at low power
levels where the temperature coefficient is a poor operational control;
during power operation, control rods would normally be fully withdrawn

from the core. S o . - oy

.
 

 

 

»

[ 1]

21

The required reactivitY'worth of control rods is a function of shim
and shutdown margin requirements, and needs to be investigated in detail.

Control-rod.wbrth as a function of fuel concentration, power conditionms,

~and reactor deeign'should be studied. In particular, use of "control

rods" which use fertile blanket salt as absorber material need to be
evaluated. | 7

The action and position'of'cantrol rods during reactor stertup need
to be examined. It appears reasonable that the rods be fully inserted
prior to start of fuel circulation, with eriticality achieved by with-
drawal of the rods. o

In general, the control’rods of the MSBR need not be used for shim

'requirements (e.g., change.in steady-state Xe level, or fuel temperature);
rather, associated reactivity changes can be made by adjusting the fuel

concentration. Reactor shutdown can be obtalned by insertion of a con-
trol rod, or by stopping & fuel pump which leads to fuel drainage from .
the core region. | _

It does not appear that control rods need to control large amounts
of reactivity (probably less ‘than 1/2% in reactivity) or to have fast
response times (response times of about a second are probably sufficient).
chever, detailed studies need to be performed relative to specific re-
gquirements as a function of eore'design. The results obtained will be
used to determine general considerations concerning'control rods and MSBR

safety.

3.3, Reactor Incldents

 

'Items'to be considered hene‘eOHCern physical~events9which influence .

,isystem reactivity, as. well as some which do not influence reactivity

per_se. QperatiOnal safety, or the abllity to continue reactor Qperation

after'abnormal events, 1s involved as well as ultimate safety where con-

_ tainment of gross radioactivity and public safety are. the important con-

icer.ns These definitions are illustrated below.
 

 

22

As normally considered, a reactor ificident involves a core reactivity
‘addition. If the reactivity addition is small enough, thére is primarily
a small disturbance in reactoi power, with no deleterious effects to the
reactor plant. Under these circumstances, operational safety is main-
tained. If the reactivity addition is large enough, a graphite tube
separating the fuel and blanket fluids may break because 6f the pressure
rise, with no other untoward effects. Under these circumstances the
reactor plant has produced no public hazard, but must be shut down for
repairs. Under these circumstances operational safety has not been
maintained, but ultimate safety has not been involved. If the reac-
tivity addition is so large that the reactor vessel ruptutes and gen-
erates a disruptive force which results in penetration of the reactor
containment, both operational and ultimate safety may be violated.

Reactor plant incidents can also occur without'the'reactof itself
being involved. For example, if mechanical failures occur which permit
water or supercritical steam to contact secondary coolant salt within
the cell containing the steam generators, high pressures could occur in
the cell and lead to rupture of this containment. Release of steam con-
taining particles of radiocactive coolant salt could involve personnel
hazard and ultimate safety.

The design of an MSBR plant must consider both operational and
ultimate safety aspects; the resulting reactor plant must have opera-
tional safety assured under nearly all credible circumstances, and
ultimate safety assured under all credible circumstances. Items which
need to be considered in such safety design studies are discussed below
and are separated into those which involve reactivity addifiions to the
reactor proper, those associated with mechanical and physical'integrity,
and items not covered in either of the above categories. In nearly all
cases, these events require malfunction of equipment or reactor-qperation
as indicated below. | |

C.
 

 

&}

By

23
)
5.53.1 Reactivity Additions
| Reactivity can be added to the MSBR byemechanisms and events similar
to those.considered for the‘MSBE;s in addition, the use of two fluid

streams separated by graphite tube walls and the supercritical steam-

 

power cycle requires that several other events be-considered. Possible
reactivity additions need to be investigated in detail.

The protective devices available to the MSBR are similar to those
in the MSRE. “Prompt" protection is afforded by the negative temperature
coefficient of reactivity and '"delayed" protection is provided by the
control rods and also by drainage of fuel salt from the core region:
Since all reactivity changeS-involve rates of addition rather than re-
activity steps, an important factor in protection is the minimum neutron
source strength which can exist_inithe core. The MSBR fuel contains an
inherent neutron source oflnearly;lof-n/sec“due to the a,n reactions re-
sulting from the alpha decay of *22U and ®2*U in the fuel salt. An addi-
tional neutron source exists from the 7,n reaction resulting from the
decay of fission products W1thin the fuel salt; the photoneutron source
is greater than 10 n/sec_for_about four months after reactor shutdown
following a month's operation at power. Thus a strong internal neutron
source is always present if reactivity 1s added at low rates, multipli-
cation of this source results in a significant increase in reactor power
before large‘amounts.of,reactivity,can_be_added to the system, which in
turn permits\the temperature.coefficient=to become effective after rela-
tively small gross reactivity;additions. _ _

Net Fuel Addition to Core. . Probably the largest reactivity addition

 

that can take place in-an MSBR is that associated W1th breakage of one or

'more graphite tubes with net. addition of fuel salt to the core region.
,However, spec1al c1rcumstances have to exist for this to take place since
| ‘the blanket region qperates at pressures higher than the fuel region, and
i.:tube breakage under normal conditions ‘would add fertile salt to the fuel

region and reduce reactiv1ty -Thus, to add reactivity, ‘the fuel pressure

- would have to rise higher than the blanket pressure at the time of, or

"ishortly after, breakage of a graphite tube. This is possible if the

high pressure of the supercritical steam system is at least partially
 

 

2y

transmitted to the fuel-salt system, or if there is a decrease in the ' \Ej
blanket pressure without a concufrent decrease'in‘the core-fuel pressure.

Failure of the'tubing in the boller-superheater could allow the
high-pressure steam to enter the coolant-galt system. - To protect against
a buildup of pressure in the coolant system, rupture discs are provided
in the steam generator and reheaters, and also could be provided'on the
shell sides of the fuel and blanket heat;exchangers.’ If these rupture
discs fail to operate, or fail to operate quickly enough, it is con-
ceivable that a buildup of pressure in the coolant system could cause
failure of the primary fuel heat exchanger. The likely means of failure
would be rupture of the shell or éollapse of the tubes, neither event
transmitting the pressure increase to the fuel fluid. However, if-there'
were localized weakness in a fuel-heat-exchanger tube, due to a defect
in manufacture, fretting corrosion, etc., failure of & tube could occur
leading to & buildup of pressure in the fuel system. Alternatively, loss
of overpressure in the blanket region could permit operation with fuel
pressures higher than blanket pressure. If a graphite tube failed under ®
such operating conditions, there would be a net fuel addition to the
core region. The reactivity addition would depend upon the pressures
and flow passages involved and their variation with time.

If steam does contact coolant salt, no éxothermal reactions of any
consequence are involved. Mixing of steam with coolant salt would oxidize
the coolant salt, but no safety hazard would be introduced beéause:of this
action. However, the corrosiveness of the mixture to the container ma-
terial needs determination. There are no fission products in the coolant
salt, and the induced activity present would decay (the primary activity
is associated with Na®** and N'®, having half lives of about 15 hr and T
sec, respectively). Cleanup of the system and repair or replacement of
damaged equipment appears possible. | P

The coolent salt is compatible with the fuel salt, so leakage of
coolant salt into the reactor system does not involve safety; any-such
leakage would reduce reactivity The BF3 added to the reactor fuel could
be readily removed by heating the salt, with the BF; removed as & gas.
¢

22

Contacting fuel salt with steam would oxidize the uranium, but
probably would not cause“any;problems other than those associated'with
subsequent cleanup~of,thejfuelts HOwever,_possible~reactivity effects
due to fuel precipitation need to be studied specifically.

At this time it appears reasonable that engineered safeguards, such
as:installing rupture discs within the heat exchangers of the coolant
system, and providifig*strehgtheoed.primary system heat exchanger tubes
can eitherfprotect againstpsuch-sh_accideht, or-keep_the_amount'of fuel
salt added to the core“regithsmall enough that ultimate”safety'is not
involved.  However, detailed-studies are needed to examine this situation.

Reactivity Changes Due to Graphite Behavior. In addition to the

 

case discussed'sbove‘in whichibreakagehof_graphite tubes was'sssumed

to teke place, other graphitelbehavior can effect reactivity changes.
For example, shrinksge of graphite during radiation exposure can effec-

 tively influence fuel concentrations' however, the associated reactivity

changes should take place at: rates such that they can be readily com-

_pensated by adding or removing fuel through normal- operations.:

Reactivity can be sdded if part of the graphite inside a fuel tube
were to break awayrfrom the.tube'proper'and.be swept out of the core
region. ;Only.smsll amounts of reactivity,could'be involved so long as
this actidn'took'place in. single_tubes,'snd noudifficulty‘for-this situa-
tion would be anticipated..'Alternatively, if graphite were removed from

- the blanket portion of the core: region, it would be displaced by fertile

salt leading to a decrease in reactivity such that safety is not involved.
- Graphite is compatible with molten salt but fuel penetration into

“the- gr&phite could. take plsce with time. Here again, the time element

involved'would make such events insignificant from a safety viewpoint.

';_ij on the other hand a pressure rise tock place in the ‘core which caused
| the fuel to penetrate and. fill voids in the graphite, perhaps a signifi-
'rcant reactivity addition could be obtained. The actual addition is de-
'pendent upon ‘the physical properties of the graphite employed. If the
‘7pressure rise. occurs because of a previous reactiv1ty addition, the

- pressure buildup itself would expel fuel salt from the core and tend

to decrease reactivity
 

26

Fuel-salt penetration in graphite appears to present little problem
during normal operation, but may present difficulties during emergency
shutdowns which require fuel-salt drainage. Fuel remaining in the- graphite
would generate decay heat which could lead to undesirably high temperatures
(temperature distributions and levels influence thermal stresses and creep
rates, which can affect the mechanical integrity of the graphite). I
ebility of blanket salt to remove this decay heat needs inVestigefion.

Reactivity Changes Associated With Changes in Flow Conditions. In a
circulating-fuel reactor, an appreciable fraction of the delayed neutrons
can be emitted external to the-core“nnder normal flow canditiqne. In-
creasing flow thus tends.to:1gwer”the,contributidn of delayed neutrons
t0 the fission chain and also decreases the average neutron lifetime of
the reactor. While lowering the delayed neutron fraction (beta) is nor-
mally considered detrimental to safety, this is in the context of systems
having instrument control. Lowering the value of beta in a system having
inherent control under the condition that reactivity additions take place
at relatively low rates does not significantly decrease the ultimate safety
of the system. Also, the effective value of beta increases during a rise
in power, a favorable condition. Lo

Since delayed neutrons are "lost" because of fuel circulation, stop-
page of flow due to pump power failure would tend to add reactivity to
the system. However, in the MSBR the reactivity addition would only be
about 0.002. In addition, stoppage of flow leads to drainage of the core,
which would make the reactor subcritical. The fuel temperature rise due
to afterheat during drainage of the core may be.the most signifieant vari-
able, and needs detailed study.' Also, time delays invfuel"drainage“frdm
the core following pump stoppage needs to be investigated experimentally,
and the results interpreted relative to reactor safety.

Another reactivity incident possible with systems having a negative
temperature coefficient of reactivity is that of the "cold slug" accident.
Such an accident could occur by starting the fuel-eirculating pump at a
time when the fuel external to the core has been cooled well below that .
of the fuel in the core. The cooler fuel would add reactivity when it

entered the core; this addition could exceed the reactivity decrease due
 

 

 

&

-y

o7

to the "loss" of delayed neutrons associated with fluid transport. By
going critical,only'with.thefpump.on;fmaking use of the control rod for
thiS'purpose; would avoid the "cold slug" incident. - The seriousness of

- the cold slug incident and the control mechanisms needed under various

_ c1rcumstances needs investigation.

Drainage of .the reactor fuel sy stem begins automatically due to

- gravity forces when the fuel pump stops. Fuel from the core drains by

gravity into the sump tank of the fuelrpumpoWhere afterheat is removed
by cooling coils. Convective circulation may be assisted by flow of gas

used to sparge xenon from the fuel salt. However, as pointed out above,

~ fuel and graphite temperatures also-need to be studied during fuel drain-
age from the core. In general, the ability and need for afterheat re-

- moval requires detailed studies.

Changes in Fuel Concentration. Reactivity"can-be-added by increasing
the concentration of fiss1le_material within the'fuel-fluid; examples of

 

possible events are filling:the'fuel tubes with salt containingrabnormally
high fissile eoncentrationsf"and returning salt having sbnormally high -

'fuel concentrations from the processing system to the reactor system.

The reactor would initially be "filled" by adding fissile material
to the'carrier-salt while the latter was circulating. If, however, follow-
ing criticality and drainage of fuel salt from the reactor core, the

'fiss1le concentration'infthe drained fuel salt were increased inadvertently,

refilling the core could result in a supercritical reactor. Such'an event
1s highly- unlikeLy, since fuel ‘would. not be added in large amounts to the
-'dralned system,.also, partial freezing of the fuel salt does not appear '
”to 1ead to Significant increases of fissile concentration in the fluid

'*',portion of the fuel. Specific cases need to be evaluated, however.

The rate of return of fuel from the processing plant is low, and it

'will be difficult to add react1v1ty at a high rate through the processing
'1ines because of- the limited rate at which fuel can be added , A more
:frlikely way to increase fuel concentration above the normal value'would
ipfltbe to fill the core with fuel having a temperature 1ower than the critlcal

temperature. A reactlvity added by this means would correspond to & low-

'mrate addition and should cause no difficulty.
 

- 28

_If fuel were to accumulate outside the core. region, and inadvertently
“return to the core, reactivity could be added rapidly to.thé reactor.
Since the fuel is homogeneous and chemically stable, this event does not
appear to be likely; also, eny such possibility would be indicated by a
previous reactivity loss.. Nonétheless, the consequences of:urahium
precipitation or accumilation outside the core and its-subséquent'additiOn
to the core region requireé general evaluation. Such studies will help
determine operating procedures consistent-with reactor safety.

While none of the above events eppears to constitute an operational
or ultimate safety hazard, all should be considered in detail. -

Reactivity‘Addition by Control-Rod Movement. The presen¢e5of a con-
trol rod permits reactivity addition to the reactor by rod movement.

- Normally the reactor would be critical with the control rod completely
removed, - but there could be conditions whg:e criticality 1s achieved
_wii:h the rod partially or completely inserted. The amount and rate of
reactivity addition associated'wifh control rod movement.under'tfiese.con—
ditions would be limited by the control rod worth (which will probably
be under 0.005 Ake) and the rate of withdrawal (which will be limited to
a low value). As with the MSRE, no difficulty is  foreseen, particularly
if rod withdrawal does not continue after the power level reaches an
initial peak value as & result of rod movement. .

Reactivity Addition Due to Positive Pressure Coefficient.' The'MSBR
design specifies use of helium as & sparge gas to remove xenon from the
circulating fuel. As a result of this operation, some gas will undoubtedly
circulate through the MSBR core, resulting in a positive préssure coef-
ficient of reactivity. The importance of this coefficient on safety is a
function of the gas content of the core, which in turn is related to the
ease of stripping xenon from the fuel salt and the efficiency df the'gas
separator used to remove sparge gas before it enters the core region. An.
increase in pressure would decrease the fraction gas in the core'énd'in-'
crease reacfivity. Experience'with'the MSRE indicates that the above is’
not & serious problem, but it needs to be evaluated_specifically-fbr the
MSER. -

1

W

 
 

 

 

C

29

33,2 -MEchanical-andsPhysical Integrity — Containment

- This subject is related to the reactivity additions discussed above.
Here, the discussion is concerned with containment relative to events
which do not necessarily require or result in reactivity additions to
the reactor systemr, Some offthe,questions which arise are: What are
the'consequences of havinglwaterrand salt in a cell if these materials
accidentally make contact? What are the-cooling conditions required if
there is-mixingrofVSalt,and'waterfr What are the consequences of fuel-
salt leakage or diffusion into;the_coolant-salt.system? ,How-practicable
is it to maintain low leakage.fromra-containment-cell'at the temperatures
involved (leakage of no more than 1% of the containment volume per day)?
What are the consequences of a major spill of fuel salt within the reactor
rcell?.. .

' The containment of the?reaC£or plant has to be assured even though
there is rupture of, or‘leakageifrom, the primary and secondary salt
systems. Rupture and/or leskage may result from overheating, overstress-
ing, corrosion, or other unekpected'material failures. The severity of
the containment problem will depend on the amount offsalt.spillage, the
rate at which water mixes with hot salt, end the amount of water added
to the cell. ConsequenCes'of'a'spillfaccident are heat generation,
pressure buildup,-and release of fission products into the cell, and
these will'need’to be evaluated:for specific cases. Problems associated
with a major spill of fuel salt within the reactor cell must be considered
in the detailed design of MSBR. systems and must also be studied experi-

mentally If water is present' corrosion of steel by HF must be cons1dered.

 The effects of local.thermalhexpansion or energy deposition due to hot

salt‘spiliage-needs evaluation. - Provision should also be made that oil

from the: pump lubrication system does not .contact hot components, al- -

though if this does occur, there normally would.not be sufficient oxygen
i'to support combustion in the- cell atmosphere of inert gas (nitrogen)
- In order to assure contalnment knowledge of the very long-term creep

ebehavior of materials under plant operatlng conditions is needed. Infor-

mation. is also needed on’ the conditions required to produce steam ex-"

.plosions" upon mixing of salt and water; similar 1nformat10n is needed

for the mixing of oil and salt.__'
 

30

The containment of fissionrpro&upts‘shquldfibe assureg,,and release
of these through the off-gas system mst not constitute a safety hazard.
This involves the amount of volatile materisl which is to be released
and the amount of fission products carried in very small,-mistelike.Salt |
particles. Ih-any case, the release of material through the off-gas
 system should be controlled so that exposure of individuals is not ex-
cessive. This can be accomplished'by-filtration and retention systems-
as required. Beryllium afid fluorine hazards, as well as radioactive
iodine, must be considered relative-to_permissible release rates during
normal operation as well as follcwihg & severe incident. rThe'réleasezof
fission productS'upbn mixingrof fuel salt and water, or of salt and oil
elso needs detérmination. A fission'product'flow and ifiventory shéet
will be made as MSER design studies are made in more detail. Also, in-
vestigation of the plating out of fission products throughout the reactor
system is an important part of the chemical development program. The
implication that fission product plating have upon reactor safety needs
to be considered.

In designing the reactor system containment, consideration must be
given to the possibility of earthquakes. The effect of such an event on
reactor containment is, of course, dependent upon its severit&, which in
turn is a function of local conditions.' The possibility of.flooding and
associated consequences is also dependent upon local conditioms.

The most likely method of rupturing the'secondary containment is
through sabotage, missile damage, acts of nature, or excessive pressuré.
The generation of missiles in the reactor cell is not likely, since the
reactor pressure is low. Missile damage and high pressurés'are more
1ikeLyrin.the coolant cell and steam plant, and, although massive'concrete
shielding is provided, such events need further investigation. The con-
tainment cells will be protected by vapor-suppression systems, whichf
~should prevent the pressure from exceeding the containment design figure
(§h5 psig for present MSBR design) in case of buildup of steam pressure.
In désigning the vapor-suppression systems, it is necessary.to'consider
the amount of salt and water that can come together and/or the leakage
of high-pressure steam into the containment volume. ‘Valves are located

O

 
 

 

 

 

 

 

« ME [ 1

A

31

in the steam lines which can be closed to prevent draining all the steam '

system into the coolant cell., The reservoir of condensing water should
be adequate to keep the cell pressure below the design containment pressure.
Also, the supercritical steam systems contain relatively small amounts of

water in comparison with subcritical systems.

3.3.3  Miscellaneous Incidents

 

Included here are possible incidents which are not covered in the
above sections. These involve vessel criticality, heat removal, and heat
addition conditions. |

Studies are needed relstive to the possibility of attaining super-
critical conditions in fuel drain tanks and in vessels of the processing
plant, along with consequences of such occurrences. Also, criticality
conditions might occur as a result of fuel spills.a In general, tanks
which hold fuel should store it indefinitely in a suberitical condition.
Accumulation of SPilled'fuel salt should be in regions which cannot attain
criticality. -

The afterheat conditions which can exist within the reactor plant
particularly need to be studied in detall, and cooling and heating pro-
vided and assured as needed. The temperatures occurring in the core
following fuel drainage need to be evaluated as a function of fuel re-
tention by the graphite. The influence of air contact on fuel salt needs
study for conditions assoclated with core maintenance operations. The
effects of salt freezing and melting in various parts of the primary and
secondary salt circuits require evaluation, with equipment designed to
minimize undesirable:effectsx(e.g.,_rupture of equipment).

fTheconsequencesofflov_blockage~withthe reactorVSYStem require
investigation. A partiallyrplngged‘fueljtube would normally not be
deteotedqand could lead'to,ealt-bOilingrand“tenperature gradients which
may:affect*the'mechanioal'integrity'of*the'fnel tube. Flow blookage”may -

 also lead to inability to remove all the fuel salt from the core, ‘which

'may lead to afterheat problems and/or maintenance difficulties.'
 

 

 

32 | s

- 3.4, Reactor Stability.

Although usually treated separately, reactor safety and stability-

" are intimately related. Reactor safety normally considers‘reiatively
large reactivity additiéns and their influénce on system behavior for
small time intervals, while reactor stabiiity studies normslly consider
small reactivity additions and determine whether they result in & buildup

' of oscillations to the point where reactor safety is involved.

For the MSBR, investigations of stabllity are required to study
the influence of inherent characteristics on instrumentation and control
system }equirements. Altfiough the MSBR has a-negativé temperature éoef-
ficient of reactivity, this ifi itself,is‘not sufficient to insure stability,
particularly if the system has time delays. The MSBR has a number of
builtin time delays which can either help or»hinder feactor stability,
éuch as the tiine lags associated with heat transport from the graphite to
the fuel, with fuel and fertile salt transport, and with delayed neutron

_production. Because of the complexity of the three-loop system from a
dynamics enalysis point of view, & preliminary linearized analysls should
first be made to evaluate the current design and aid in establishing ap-
propriate means of system control. ’

It is estimated that an adequate preliminary analysis for the com-
plete system (reactor core to turbogenerator) would involve about 60 first-‘
order equations (about 1k for’{he fuel stream, 14 for the fertile stream,
7 for nuclear kinetics, 15 for coolent streams, and 10 for the steam
system). These equations would comsider fuel and blanket nodes, transport
deleys, heat exchanger nodes, and fuel leakage effects. Work is required
in formulating the specific equations and in compiling and_e#aluating the
physical parameters. Present computer codes could be utilized in the
initiel analysis. In particular, codes are available for pérforming a
dynamic enalysis utilizihg a general linear model. These can be used to
give system eigenvalués, system freduency response, and/or gystem trans-
ient fesponsel R - | S

Some of the important items to be investigated in stability énalysis
would be the significance of heat transfer lags between various parts of

.

bt
 

 

 

 

 

c

wh

o

55

the reactor system, the, relative importance of the fuel temperature and
graphite temperature coefficients of reactivity, and the influence of
delayed neutron fraction and flow effects on reactor system behavior.

The effect of fuel flow on_thejeffective-delayed neutron fraatibn may
reduce this value from 0.003 to less than 0.00l.  This reduction is not
necessarily bad from either the viewpoint of inherent safety or inherent
stabiliiy. 'In-fact,rlowering the delayed-neutron‘fraction can increase
the degree of stability, as was the case for the Homogeneous Reactor Test
(HRT). Also, the tendency (in circulating-fuel systems) of the effective

delayed neutron fraction to.increase‘during-fhe power rise portion of a

power pulse tends to aid stability. Plans for the MSRE call for operating

that reactor with 233U fuel beginning in the fourth quarter of FY 1968.
Studies will be made of the stability of the MSRE with the ®33U fuel and
the results will be used where applicable in the analysis of the stability
of the MSBR. ,

More detailed stability,analysis'studies would be dependent upon
the results.obtained from the initial evaluation but presumably would
include investigation of nonlinear effects and their relative influence

on results.
4. MSBR SAFETY PROGRAM
4.1. Summary

The studles and. investigations associated with MSBR safety are

summarized here in terms of. general and detailed: studies which need

to be done in order to evaluate ‘MSBR safety; these: constitute investi-

gations which will be carried out in the MSBR Safety Program..

~ * The’ favorable safety characteristics of MSBR systems arise from the
1bw excess react1v1ty-available,to,the reactor, the prompt negative tem-
pefature‘doefficient*of'reaCtivity,ithe low system pressures, the low

level of f1551on gases and. fission products retained w1th1n the reactor,

. the mobility of fluid fuel, and the ease of fuel drainage from the re-.

actor. At the same tlme,_there are a number of p0551ble incidents and
 

3l

safety aspects which need detailed investigation; these aspects are re-
lated to the specific plant design and involve both mechanical and nuclear
design features. Plant systems which have a major influence on MSBR f
reactor safety are the reactor system proper, the steam system, the fuel-
recyclefiprocessing system, the coolant systems, and the off-gas system
These are described above (Section 2), along with safety features that
vere incorporated in the plant de51gn ,

Safety analysis requires a study of possible incidents, their conse-
quences, and their avoidance.A,TypeSIOf accidents whichkcan take place

include those due to reactivity additions. Reactor behavior under such

circumstances. is influenced by reactivity coefficients and kinetics param--

eter values. Reactivity coefficients which will be considered include
those associated with temperature, voids, pressure, fuel concentration,
and graphite concentration, and involve the fuel and blanket fluids .

separately and together. The function and design of control rods will

be fully investigated; these studies will deteiudne the number, reactivity 1

worth, placement, and response requirements of control rods, as well as
the ability to utilize blanket salt as a control rod. Possible reactor
incidents will be evaluated as to their probability and their consequences,
also, the influence of design changes (including alternate core designs)
on safety aspects will be obtained.

Under normal operating conditions, the MSBR should be load-following
and self-controlling because of the prompt, negative temperature coeffi-
cient of reactivity associated with the fuel salt. The temperature coef-
ficient also protects against excessively high reactor temperatures and
pressures in case of reactivity incidents. This situation is partially
due to the large inherent neutron source strength present in-the fuel
salt (nearly 10’ n/sec due to the a,n'reaction),'which permits thertem-
perature coefficient to become effective as & reactivity cofitrol agent
soon after initiation of rate additions of reactivity. '

Reactivity additions and their safety implications which will be
considered in detail involve: breakage of graphite fuel tubes and the
possible net fuel addition to the core region; other types of graphite .
| behavior; changes in fluid-flow conditions; changes in fissile con-
centration within the fuel fluid; abrupt changes in fission product

“t3
 

 

wi

alk

25

concentrations;'chahge-in control rod position; and the effect of pressure
increases on reactivity. Relative to graphite tubes, & study of the cred-
ibility and the consequences of single and multiple failures of graphite
tubes in the reactor Wiil be made. '

The integrity of plant containment under both reactivity incident
conditions and finder circumstances where reactivity itself is not involved
will be evaluated for & number of physical possibilities; these include
events such as mixifig of coolant salt with water or steam; spills of
fuel or coolént salt and associated thermal, chemical; corrosion, -and
criticality affects; temperature changes due to afterheat generation;

container damage due to high temperature and/or corrosion; criticality

in regions outside the core; flow blockage within the fuel or coolant

streams; and blockage of flow in the off-gas system. The consequences
of credible accidents will bé determined in all cases.

. The application-df pressure-suppression systems. to molten-salt
reactor plants will be 1nvestigated, and-problems'associated with their
use will be analyzed. Additional design studies will be performed to
better define such systems and their operation in detail.

There are a number of areas which will be investigated experimentally
in order to determine the general safety problems of molten-salt reactors.
Some of these are closely related to areas studied as-part of the
engineering-development and research programs of the MSBR Program. These
include determination of the effects of reactor operating conditions on
the physical behavior and properties of graphite, of graphite-~-to-metal
joints,“and-of'Hasteiloy‘N., Tfiéilong-term creep properties of Hastelloy N

 and graphite need to be xnown end understood; also, the physical and

éhemicalJproperties‘of'Saltsééfid‘of salt-water mixtures need to be known.

. In addition, the ability to drain the fuel from the core under credible
| conditions,needs study;-alsofthe'desirébility of alternative'core-designs:

--relatiVE“to=afterheat-removai{will:be evaluated.

1~*Experimental-informationfwill5berobtained'on salt permeation of

grephite, fission-product deposition in reactor systems, the ability to

remove fission products from surfaces, and the ability to remove after-

heat generated in the5fuel'salf;“fExpérimentsrwill be performed concerning
 

 

36

- the release of fission gases from solid as well as molten fuel salts,
end concerning the fission-product flow and inventory throughout the -
reactor.system. Retention of - fission products as well as tritium in
off-gas systems will be demonstrated.

- Measurements will be made concerning the conditions required to.
produce "steam explosions' when molten salt and:water are mixed., The
release of fission pro&ucts from fuel salt upon mixing it with water,
or with oil, also will be measured.

Investigations will elso be. performed concerning the reactivity
effects associated with prec1pitation of uranium, rapid movements of
fission products, "cold slug" accidents,'fuel-lesksge-into'thevblanket
region via a plennm chamber, boiiing of blanket salt within the core |
region, and buckling of a fuel-plenum.wall with associated change in
graphite distribution. '

A stability analysis of the reactor plant systems will be made to
determine the operating, control,,and/or design requirements for obtain-
ing satisfactory plant characteristics. Items to be considered are time-
lag events, spatial-distribution effects, the effect of fuel tube
oscillations upon reactor behavior, and the relative importance of

various parameter values upon system behavior.
4,2. Cost Estimates

The planned safety studies are to resolve the basic safety problems
associated with MSER plants. This means that enough information will be
obtained to know which problems are the most important ones and how they
 can be overcome or eliminated (e.g., by chenging either the reactor de-
sign or methods of operation). These studies will also provide experi-
mental information which is necessary in order to resolve safety problems.
On this basis, the cost estimates required to carry out the investigations
indicated above are those given in Table 1. These estimates take into
consideration the efforts planned in other parts of the MSBR Program .
which are related to reactor safety, but do not include costs of such
studies. However, the MSBR safety program depends on these other in-
vestigations for major contribufions. Information which will be obtained

Ty
 

Table 1. Cost Estimates for the MSBR Safety Program%.

 

Cost, in millions of dollars'

 

 

"fInvéstigations ' — : —_—
o S “FY 1968 1969 1970 1971  1972° 1973 1974 1975 Total
Reactivity-related = 0.05 0.05 0.05 0.05 -- -- - - 0.20
events - ‘
Physical and chemical @~ 0.10 0.1 0.05 0.05 0.05 0.05 0.05 0.05 0.50
behavior of materials
Equipment-failure | 0.10 0.1 0.10 0.05 0.05 0.05 0.05 0.05 0.60
-events : o :
Total 0.25 0.30 0.20 0.05 0.10 0.10 0.10 0.10 1.30

 

*Does not include costs of safety studies carried out as part of other programs.

Le

 
 

38

from other parts of the Program include the physical and chemical prop-
erties of salts and structfiral materials, the characteristics of pressure-
suppfession systems and containment structures, and the behavior for
relisbility 6f reactor components. | |

The safety program outlined above includes the specific mathéfiatical
and physical formulation of the various problems, the compilation and
evaluation of parameter values, and-determination of reactor plant be-
havior under the postulated conditions. Experimental studies will be
performed in conjunction with other MSBR investigations, which will in-
volve modifying or initiating nevw experiments so as to give pértinent
'safety information. The objective is to détermine design and cperating
conditions which are compatible with reactor safety and economic power
production; The present MSBR design would serve as a starting point in
these studies; however; general safety information related to molten-

salt reactors will be obtained as problems become more clearly defined.
ACKNOWLEDGMENTS

The assistance of E. S. Bettis, T. W. Kerlin, W. B. Cottrell,
S. J. Ditto, and J. 0. Kolb in providing information for this report
is gratefully acknowledged. Thanks are also given to R. B. Briggs
and R. C. Robertson for their review of this report.
 

 

 

 

T

-«

)

REFERENCES

P. R. Kasten, E. S. Bettis, and R. C. Robertson, Design Studies of
1000-Mw(e) Molten-Salt Breeder Reactors, USAEC Report ORNL-3996,

 

‘Oak Ridge National Laboratory (August 1966).

W. L. Carter and M. E. Whatley, Fuel and Blanket Proce381ng Develop-
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Osk Ridge National Laboratory (June 1967).

S. E. Beall et al., MSRE Design and Operations Report, Part V,
Reactor Safety Analysis Report, USAEC Report ORNL-TM-T752, Oak

Ridge National Laboratory (August 1964).

 

 

 
 

 

 

Q.

 

 

 

 

 

 
 

 

C

1-50.

51.
52,
55.
54.
55.
56.
57
58.
59.
60.
61.
62.
63.
64,

66.
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70,

1.
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Th.
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76.
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82.

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89. 7

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91.

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9.
9L,

Internal Distribution

41

 

MSRP Director's Office

Room 325, 9204-1
R. K. Adams

G. M. Adamson
R. G. Affel

L. G. Alexander
R. F. Apple

C. F. Baes -

J. M. Baker

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W. P. Barthold
H. F. Bauman

S. E. Beall

M. Bender

E. S. Bettis

"F. F., Blankenship

R. E. Blanco
J. 0. Blomeke
R. Blumberg

E. G. Bohlmann

C. J. Borkowski'_

G. E. Boyd

~J. Braunstein

M. A. Bredig

R. B. Briggs

H. R. Bronstein
G. D. Brunton
D. A. Canonico
S. Cantor

W. L. Carter

G. I. Cathers -
~J. M. Chandler.

E. L. Compere
W. H. Cook
L. T. Corbin -
J. L. Crowley
F. L. Culler

'J. M. Dale '
D. G. Davis

S. J. Ditto-

A. S. Dworkin
J. R. Engel

E. P. Epler =
D. E. Ferguson
L. M. Ferris

A. P. Fraas

95.
96.
97.
a98.
99.
100.
101.
102.
103.
10L.
105.
.106.
107.
108.
109.
110.
111.
112.

113.

11k,
115.
116-120.
121.
122.
123
124,
125.

- 126,
127.
128.
..129_
130.
131,

- 132.
133,

134,

135.
136.
137.

138.

139,
- 140,

1.

1k2.
143,
1hk.

B' A.

P. H.

D. G.
C. S.
P. N.
F. A.
P. G.
J. R.
H. W.
R. W.
T. L.

ORNL-TM-1858

Friedman
Frye, Jr.

. Gabbard
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Goeller |

« Grimes
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. Guymon
‘Hannaford

Harley
Harman
Harrill
Haubenreich
Heddleson
Herndon
Hightower

- Hoffman -

Horton
Hudson

H. Inouye

W. H.
P. R.
R. J.

H. E.
Ho F.
C- _K-

Jordan
Kasten
Kedl

. Kelley
. Kelly
. Kennedy

Kerlin

. ‘Kerr

Kirslis
Krakoviak.r

. Krewson
. Lamb
. Lane

Lindauer

. Litman
Lundin
. Lyon

MacPherson
MacPherson
Martin

" Mathews
. Matthews

MCClung

‘McCoy

McDuffie
McGlothlan
 

 

ORNL-TM-1858
145. C. J. McHargue 171. O. L. Smith
146. L. E. McNeese 172. P. G. Smith
147. A. S. Meyer 173. W. F. Spencer
148. R. L. Moore 17k. 1I. Spiewak
149, J. P. Nichols 175. R. C. Steffy
150. E. L. Nicholson 176. H. H. Stone
151, L. C. Oakes 177. J. R. Tallackeon
152. P. Patriarca 178. E. H. Taylor
155. A. M. Perry 179. R. E. Thoma
154. H. B. Piper 180. J. S. Watson
155. B. E. Prince 181. C. F. Weaver
156. J. L. Redford 182. B. H. Webster
157T. M. Richardson 183. A. M. Weinberg
158. R. C. Robertson 184, J. R. Weir
159. H. C. Roller 185. W. J. Werner
160. H. C. Savage 186. K. W. West
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162. Dunlap Scott 188. J. C. White
*163. H. E. Seagren 189. L. V. Wilson
164. W. F. Schaffer 190. G. Young
165. J. H. Shaffer 191. H. C. Young
166, M. J. Skinner 192-195. Central Research Library
167. G. M. Slaughter 194-195. Document Reference Section
168. A. N. Smith 196-205. Laboratory Records (LRD)
169. F. J. Smith 206. Laboratory Records (LRD-RC)
170. G. P. Smith
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226. H. M. Roth, Atomic Energy Commission, ORO.
227-228. M. Shaw, Atomic Energy Commission, Washington
229. W. L. Smalley, Atomic Energy Commission, ORO
230. R. F. Sweek, Atomic Energy Commission, Washington
231-245, Division of Technical Information Extension (DTIE)
246. Research and Development Division, ORO g
247-248. Reactor Division, ORO