ORNL-TM-0733REV3.txt

 
  

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OAK RIDGE NATIONAL LABORATORY
operated by

UNION CARBIDE CORPORATION

NUCLEAR DIVISION w

for the
U.S. ATOMIC ENERGY COMMISSION

 

ORNL- TM- 733 (3rd Revision)

COPY NO. -

DATE - July 25, 1969

RECEIVED BY DTIE Aua 111969 W

MSRE DESIGN AND OPERATIONS REPORT

Part VI

OFERATING SAFETY LIMITS FOR THE MOLTEN-SALT

REACTOR EXPERIMENT

R. H, Guymon P, N, Haubenreich

NOTICE This document contains information of a preliminary nature
and was prepared primarily for internal use at the Oak Ridge National

Laboratory. |t is subject to revision or correction and therefore does
not represent a final report.

e et aE -
DISTRIBUTION OF TH!S DOTUMENT 1S LMITED
To Government Agsicies and Thair Contractors
- LEGAL NOTICE = -~ S L

This report was prepored as an account of Government sponsored work. Neither the United Siates,

nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the accuracy,
completeness, or usefulness of the information contained in this report, or that the use of
any information, apparotus, method, or process disclosed in this report may not iniringe
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any information, apperatus, method, or process disclosed in this report.

As used in the obove, “person acting on behalf of the Commission'" includes any employee or

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

or his employment with such contractor,

 
v -

 
 
 

    

  
   

 
 
 

LEGAL NOTICE

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A. Makes any warranty or representation, expressed or implied, with respect 1o the accu-
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NOTICE

This report contains information of & preliminary
nature and was prepa red primarily for internal use
at the originating ins tallation. It is subject to re-
yision or correction and therefore does not repre-
gent & final report, 1t is passed to the recipient in
confidence and should not be abstracted or further
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ployee OT coniractor O
such empioyse or contracter of the Commisaion, of employee of such contractor prepares,
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disseminates, OT provides access to, any snformation purspant to
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MSRE DESIGN AND OPERATIONS REPORT
Part VI

OPERATIN |
G SAFETY LIMITS FOR THE MOLTEN-SALT REACTOR EXPERIMENT

(Third Revision)

R. H.
Guymnon P. N. Haubenreich

is to include all th i relat to the h t
nose items directl
elated to e heal
Yy h and safety
fo J

of the publi i
¢c. Some items extend to the safety of the operat
ors and

the protection
of the Experiment i
against a severe and di i
isabling accident

safety 1limits requi
guires approval of the O
RNL management and
the AEC-0RO

Contract Administ
strator., Any violation of a safety limit shall be
reported

not later than the anext work day to the AEC~ORO

This document s a
Operating Saf .b?PErhedes MSRE Design and Operations Report, Part V
g Safety Limits for the MSRE, ORNL-TM-733 (2nd R ; I —
evision).

i. Fuel System

. R -

fuel
salt shall not exceed 1.0 ft°/min, If this rate i
is ex-

ceeded i
, gas flow into the drain tanks shall be stopped

exceed i
25 psig whenever fuel salt is in the reactor
vessel,

If this I1imit 1
imit is reached, the pump bowl shall be vented th ;
the charcoal beds to the stack e

 

SIS LMITED
Ty l"‘e.-n{. ’
Sownntractors

_lxq

1
1.3 Surge Volume — The total gas volume in the fuel-pump bowl and
the gverflow tank shall be at least 5,0 ft° whenever the reactor
is critical. If this limit is reached, the reactor shall be
taken subcritical until the volumetric inventory of fuel is
reduced.,

1.4 Excess Reactivity — The reactivity shall be such that the con-
trol rods must be withdrawn at least 50 percent of their total
worth to make the reactor critical at 1210°F., If this limit is
reached, the reactor shall be held subcritical uwatil the ab-
normality is corrected,

1.5 Power — The reactor shall not be operated at a steady power in
excess of 8 Mw. If heat balances indicate nuclear power above
8.1 Mw for more than 1 hour, the heat-removal rate shall be re-
duced to a heat-balance power of 8.0 Mw or less.

1.6 Addition of Fissile Material*“— No more than 120 g of fissile
material shall be added to the fuel in the pump bowl in any

single addition,

 

1.7 Reactivity Anomaly — The reactivity anomaly shall not exceed
0.5% 3k/k while the reactor is critical. "Reactivity anomaly"

is defined as the difference between the observed reactivity and
the reactivity predicted on the basis of measured reactor physics
characteristics and calculated effects of changes in operating
conditions, burnup and fission product accumulation. If this
limit 1s reached, the reactor shall be taken subcritical.

*
1.8 System Test at Elevated Pressure — The fuel circulating system

 

and fuel drain-tank system shall be pressure-tested at least
once a year at a minimum pressure (measured in gas) of L5 psig,
a minimum temperature of 1150°F, and flush salt being circulated
by the fuel pump.

1.9 Corrosion — The chromium concentration in the fuel salt shall
not exceed 1000 ppm. If this limit is reached, steps shall be
taken to minimize the corrosion rate and to reduce the chromium

concentration in the fuel to less than 500 ppm.

 

*
This 1imit will not be reached by any spontaneous change of a pro-
cess varlable, so no operator response is specified.
v -

Control Rods and Safety System

 

2.1

2.2
2.3

2.h

2.5

2.6

2.7

2.8

*
Scram Circult Tests -~ All scram circuits shall be shown to be

 

operating properly by testing before each fill of the reactor
vessel with fuel salt.

*
Scram Tests — The scram time for each control rod shall be

- measured before each fill of the reactor vessel with fuel salt.

Scram Time*—~ The reactor shall not be taken critical if the
scram time of any control rod is greater than 1.3 sec,

Rod Speed — The reactor shall not be taken critical if the
motor-driven speed of any control rod is less than 0.45 in./sec
or more than 0.55 in./sec. If a rod will not move, the reactor
shall be taken subcritical.

Control Rod Cooling — Any control rod that is not fully with-

 

drawn shall be supplied with copling air whenever the reactor
is operating at powers above 15 kw. Temperatures in the rod
drive whose cooling air supply is connected to that for the rod
shall be accepted as evidence of air flow through the rod.

Instrument Shaft Water — The nuclear instrument shaft shall be

 

filled with water whenever fuel salt is in the reactor vessel,
If for any reason the water level cannot be maintained at
8Lg-ft elevation or above, the fuel shall be drained.

Nuclear Startup Instrumentation — One neutron count-rate channel

 

shall be in service throughout the filling of the fuel loop with
fuel salt and whenever the reactor is being teken critical. If
an instrument failure occurs during filling, the fuel shall be
returned to the drain tank.

*%
Flux Instrumentation —— A minimum of two flux safety channels

 

shall be in service during nuclear operation.

 

*
This limit will not be reached by any spontaneous change of a pro=-
cess variable, so no operator response is specified.

**A fuel £ill shall not be started if this limit is not met. If the

limit is violated after fuel is in the core, the reactor shall be taken
subcritical immediately by full insertion of all control rods and shall
not be taken critical until the requirements are met.
 

 

 

 

 

 

 

 

 

 

*H ol
2.9 Period Instrumentation -— A minimum of two period safety
channels shall be in service during nuclear operation.
2.10 Fuel Temperature Instrumentation**—— A minimum of two reactor-
fuel-outlet temperature safety channels shall be in service
during nuclear operation.
2.11 Flux Trip Point**—— The reactor power which will cause a safety-
rod scram trip shall be 12 Mwt or less during nuclear operation.
2.12 Flux Trip Point, Fuel Pump Off**—— The indicated reactor power
which will cause a safety rod scram trip shall be 12 kwt or less
during nuclear operation when the fuel pump is not operatingli-l
2.13 Period Trip Point**—— The shortest positive reactor period‘fihat
will be tolerated without causing a safety rod scram trip shall
be no shorter than one second during nuclear operation.
2.14 Fuel Temperature Trip Point**—— The reactor outlet temperature
which will cause a safety rod scram trip shall be 1300°F or
less during nuclear operation.
3. Coolant System .
3.1 System Test at Elevated Pressure*—— The coolant circulating o
system and coolant drain-tank system shall be pressure-tested
at least once a year at a minimum pressure (measured in gas) of
45 psig, a minimum temperature of 1150°F and coolant salt being
circulated by the cocolant pump.
L. Containment
4.1 Cell Shield Blocks — All reactor cell and drain-tank cell
shield blocks shall be in place and secured by hold-down devices
whenever fuel salt 1s in the reactor vessel,
L.2 Cell Oxygen Concentration — The reactor cell and drain-tank
cell shall contain a nitrogen-air mixture having an oxygen con=-
centration below 5 percent whenever fuel salt is in the reactor
**A fuel fill shall not be started if this limit is not met., If the
limit is violated after fuel is in the core, the reactor shall be taken
subcritical immediately by full insertion of all control rods and shall
not be taken critical until the requirements are met. - -
*This 1limit will not be reached by any spontaneous change of a fi-"

process variable, so no operator response is specified.
» —

.2

4.3

L.k

L.5

L.6

L.t

(continued)

vessel. If this limit 1s reached, the nitrogen purge into the
cell shall be increased to bring the oxygen concentration below
5 percent as guickly as is practical.

Cell Pressure -—— The pressure in the reactor cell and drain-tank
cell shall be maintained bétween -1 psig and -4 psig whenever
fuel salt is in the reactor vessel. If either limit is reached
and the pressure cannot be brought back into limits within one
hour, the fuel shall be drained.,

Cell Temperature — The average temperature of the atmosphere

 

in the reactor cell and drain-tank cell shall not exceed 350°F,
If this limit is reached the fuel shall be drained.
Cell Leak Rate During Operation — The leak rate of air into

 

the reactor and drain-tank cells shall be determined once per
week during reactor operation. It shall not exceed 7O scfd at

the normal operating pressure of -2 psig and temperature of

'130°F. If a measurement indicates a leak rate in excess of this

1limit, the leak-rate data shall be analyzed without delay and

if the analysis does not indicate that the rate is actually
within limits, the fuel shall be drained.

Cell Leak Test at Elevated Pressure*- The reactor and drain-tank
cells shall be leak-tested at least once per year at a minimum
pressure of 20 psig. The leak rate at this pressure shall not
exceed 280 scfd,

Reactor Cell Annulus Water — The water level in the reactor

 

cell annulus shall be maintained above elevation 844 ft - 9 in.
If this limit is reached, steps shall be taken without delay
to raise the water level, If the specified level cannot be

attained within L hours, the reactor shall be taken subcritical.

 

*
This 1imit will not be reached by any spontaneous change of a

process variable, so no operator response is specified,
L.8

4.9

4,10

k11

L.12

4.13

L1k

Vapor-Condensing System Pressure — The maximum vapor-condensing

 

system pressure shall not exceed 3 psig whenever fuel salt is in
the reactor vessel, If this limit is reached and the pressure
cannot be brought below 3 psig in one hour, the fuel shall be
drained.

Vapor-Condensing System Water Volume — The volume of water in

 

the vapor-condensing tank shall be between 8000 gallons and

9300 gallons whenever fuel salt is in the reactor vessel, If
the volume of water cannot be held in that range, the fuel shall
be drained.

*
Vapor-Condensing System Test at Elevated Pressure — The vapor-

 

condensing system shall be pressure-tested at least once per
year at a minimum pressure of 20 psig.

Ventilation Filters Test*—— The high~efficiency particulate
filters ("absolute" filters) at the stack shall be tested in

place at least once a year and after each change of filter ele-

 

ments., Filters in service shall have an efficiency of 99.9% or
greater for 0.3-micron dioctylphthalate particles. v

Ventilation Through Open Cell — When openings are made into the

 

reactor cell or drain-tank cell, a flow of air shall be wmain-
tained through each opening from the operating area intec the

cell, If a net inward flow cannot be maintained, the opening
shall be closed or be reduced in size to meet the requirement.

Fuel System Gas Supply Pressure — The pressure in the header

 

supplying cover gas to the fuel system shall not be less than
28 psig. TIf this limit is reached, appropriate block valves
shall be closed immediately to guarantee containment.

Ieak Detector Header Pressure — The pressure in leak detector
headers connected to flanges in the fuel and fuel offgas systems
shall be at least 10 psi above the pressure inside any of the
connected flanges whenever fuel salt is in the reactor vessel.
If this limit is reached, appropriate block valves shall be

closed immediately to guarantee containment. .

 

*
This limit will not be reached by any spontaneous change of a process
variable,

s0 no operator response is specified.
*
4,15 Block Valve and Check Valve Test — All block valves and check

 

valves that are part of the primary containment of the fuel
cover gas and fuel offgas shall be leak-tested at least once
a year.

4,16 Thermal Shield Water Flow — A cooling water flow of at least
15 gpm shall be maintained through the thermal shield whenever

 

the reactor is critical. If the flow drops below this limit
while the reactor is critical and cannot be restored within one
hour, the reactor shall be taken subcritical.

5. Radiation

5.1 Building Radiation Monitors — A minimum of two radiation
monitors shall be in operation at all times, one in the high-bay
area and one in the office — control-room area. If a failure
should occur, steps shall be taken without delay to restore the
system. Until the normal system is again operable, equivalent
protection shall be provided by use of portable instruments and
special procedures.

5.2 Building Air Monitors — A minimum of two air activity monitors
shall be in operation at all times, one in the high-bay area
and one in the office — control-room area. If equipment failure
should occur, steps shall be taken without delay to restore the
system.

5.3 Stack Release of Radioactivity — The rate of release of radio-
active materials from the ventilation stack, averaged over any
12-month period, shall not exceed 0.62 yc/sec of iodine,

79 me/sec of noble gases, and 36 pc/sec of other mixed fission
products. If tais limit is reached, operations shall bhe re-
stricted to minimize further releases,

5.4  Stack Monitors — A system capable of monitoring release of
iodine, particulate B - y emitters, and particulate o emitters
shall be in service on the ventilation stacks at all times, If
equipment failure should occur, steps shall be taken to minimize
the possibilities for undetected release and to restore the sys-

tem as soon as possgible.

 

*
This limit will not be reached by any spontaneous change of a process
variable, so no operator response is specified,
10

6. Staff and Procedures
6.1 Minimum Staff — Whenever fuel salt is in the reactor vessel,
the minimum staff shall consist of one Supervisor or Chief
operator and two technicians.
6.2 Control Room*—— Whenever fuel salt is in the reactor vessel,
the main control room shall be attended by a certified Super-
visor, a certified Chief Operator or a certified Operator.

*
6.3 Reactivity Controls — Whenever fuel salt is in the reactor

 

vessel, manipulation of the control rods or reactor power con-
trols shall be done or directly supervised by qualified
personnel certified by the Director of the Reactor Division,
ORNL.

6.4 Procedures*—— The reactor shall be operated in conformance
with current MSRE Operating Procedures and test procedures and
instructions approved as specified in the Operating Procedures.
In no case shall these authorize exceeding the safety limits

applicable at the existing reactor conditions.

 

*
This 1limit will not be reached by any spontaneous change of a process
variable, so no operator response is specified.
-

O O~ O W N

72-86,
87-96.

97-98.
99.

fi]?:fi:fi?d?>§20lbcflt*tjciilm’ficfic4E}@iwffi HOEnEHOYG

11

ORNL-TM-733
Revision 3

Internal Distribution

 

G, Affel 35. P. R. Kasten

L. Anderson 36. A. I. Krakoviak
F. Baes 37. M. Iundin

S. Bettis 38, R. N. Lyon

E. Beall 39. H. G. MacPherson
S. Bettis Lo, R. E. MacPherson
Blumberg L1, H. E. McCoy

G. Bohlmann 42. H. C. McCurdy

J. Borkowski 43, L. E, McNeese

B. Briggs L, A, J. Miller

R. Bruce k5, R. L., Moore

B. Cottrell 4. E, L. Nicholson
A, Cox 47. A. M. Perry

L. Crowley 48, M. Richardson

L. Culler 49-50. M. W. Rosenthal
J. Ditto 51. A. W, Savolainen
P. Eatherly 52. D. Scott

R. Engel 53. M. J. Skinner

E. Ferguson 5k, TI. Spiewak

M., Ferris 55. R. C, Steffy

K. Franzreb 56. D. A, Sundberg
P. Fraas 57. R. E. Thoma

H, Gabbard 58. D. B, Trauger

R. Grimes 59. A. M. Weinberg
G. Grindell 60. J. R. Weir

H. Guymon 61l. M. E. Whatley

H. Harley 62. J. C. White

N. Haubenreich 63. Gale Young
Houtgzeel

L. Hudson

Central Research Library (CRL)

¥-12 Document Reference Section (DRS)

Laboratory Records Department (LRD)

Laboratory Records Department — Record Copy (LRD-RC)

External Distribution

Division of Technical Information Extension (DTIE)

H. M., Roth, Division of Research and Development, AEC, ORO

T. W. McIntosh, Div. of Reactor Development & Technology,

U. S. Atomic Energy Commission, Washington, D. C. 20545

Milton Shaw, Director, Division of Reactor Development and
Technology, U. $. Atomic Energy Commission, Washington, D. C. 20545