ORNL-TM-9756.txt

 

'OAK RIDGE
'NATIONAL
LABORATORY

MARTIN MARIETTA |

      

* OPERATED BY

~ MARTIN MARIETTA ENERGY SYSTEMS, INC.
- FOR THE UNITED STATES '
 DEPARTMENT OF ENERGY

 

ORNL/TM-9756

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COVER

Extended Storage-in-Place
of MSRE Fuel Salt and Flush Salt

Karl J. Notz

DIETRIRUTION OF THIS DOSUW ST IS UNLIKIT S

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Printed in the United States of America. Available from
National Technical Information Service
U.S. Department of Commerce
5285 Port Royal Road, Springfield, Virginia 22161
NTIS price codes—Printed Copy: A07; Microfiche AO1

 

 

 

 

This report was prepared as an account of work sponsored by an agency of the
United States Government. Neither the U nited States Government nor any agency
thereof, nor any of their employees, makes any warranty, express or implied, or
assumes any legal liability or responsibility for the accuracy, completeness, or
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to any specific commercial product, process, or service by trade name, trademark,
manufacturer, or otherwise, does not necessarily constitute or imply its
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any agency thereof. The views and opinions of authors expressed herein do not
necessarily state or reflect those of the United States Government or any agency
thereot.

 

 

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ORNL/TM--9756

DE86 001720

Nuclear and Chemical Waste Programs

EXTENDED STORAGE-IN-PLACE OF MSRE FUEL SALT
AND FLUSH SALT

Karl J. Notz
Chemical Technology Division

With contributions from:

R. C. Ashline, Chemical Technology bivision

D. W. Byerly, University of Tennessee

L. R. Dole, Chemical Technology Division

D. Macdonald, Martin Marietta Energy Systems, Inc.,
Engineering

T. E. Myrick, Operations Division

F. Jo. Peretz, Martin Marietta Energy Systems, Inc.,
Engineering

L. P. Pugh, Operations Division

A. C. Williamson, Martin Marietta Energy Systems, Inc.,

Engineering

Date Published: September 1985

NOTICE This document contains information of a preliminary nature.
It is subject to revision or correction and therefore does not represent a
final report,

Work Sponsored by U.S. Department of Energy
Surplus Facilities Management Program

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37831
operated by
MARTIN MARIETTA ENERGY SYSTEMS, INC.
for the
U.S. DEPARTMENT OF ENERGY
under Contract No. DE-AC05-840R21400

 

&
DISTRIZUTION OF THIS DOCIITIENMT IS UMLIRITED
AB STRACT * * * * e . -* L] ® o * e

1.

EXECUTIVE SUMMARY < o « « &
l.1 HISTORY o o o o o o o &«
1.2 PROJECTIONS « ¢ ¢ o & &
1.3 CONCLUSIONS &« o « o o

INTRODUCTION « o ¢ o o o o

TORY e - e o . . . o * e
DESCRIPTION o« o o ¢ o &

PRIOR STUDIES « o ¢ o
OPTIONS « o o o o o o &

PROJECTIONS 4« o ¢ o o ¢ o
4,1 RADIOACTIVITY o ¢ o o
4.2 RADIOLYSIS o o s o o &
4,2.1 Pressure Rise .
4.2.2 Corrosion Rates
4.,2.3 Use of Getters .

HIS
3.1
3.2 SURVEILLANCE AND MAINTENANCE
3.3
3.4

4.3 INTEGRITY OF THE FACILITY

4.3.1 Cell Penetrations

4,3.2 Water Control .

»

iii

- *

4.3.3 Secondary Containment

4 GEOLOGY AND HYDROLOGY .
o> ENTOMBMENT + ¢« o o « o
6 UTILIZATION o o o o o
7 FINAL DISPOSAL « o o »
4e7.1 WIPP v o o o o o
4,7.2 Commercial Spent

4.8 IN-CELL ALTERATIONS .
4,9 CONTINGENCY PLANNING .

CONCLUSIONS o ¢ o o = o + @
5.1 EVALUATION OF OPTIONS .
5.2 SELECTION OF PLAN « « «

CONTENTS

Fuel Repository
4.7.3 Greater Confinement Disposal . .

Appendix
Appendix
Appendix
Appendix
Appendix
Appendix

A.
B.
C.
D.
E.
F.

BIBLIOGRAPHY AND REFERENCES o« ¢« ¢ ¢ ¢ ¢ o o o« &
SEMIQUANTITATIVE EVALUATION OF SIX MSRE OPTIONS
WIPP WASTE ACCEPTANCE CRITERIA .« & & 4 o ¢ o o
ANALYSIS OF MSRE CELL PENETRATIONS . « «. . . .
BUILDING 7503 DRAINAGE SYSTEMS « o ¢ o ¢ « & &
GEOLOGIC INVESTIGATIONS RELATIVE TO MOLTEN SALT
REACTOR DECOMMISSIONING e o o o« ¢ o ¢ o o o

lo
17
18

27
27
45
46
50
52
54
55
60
62
62
63
64
65
66

67
6/
68

71
71
73

75
79
85
93
111

125
EXTENDED STORAGE-IN-PLACE OF MSRE FUEL SALT
AND FLUSH SALT

 

Karl J. Notz

ABSTRACT

The solidified fuel salt and flush salt from the Molten
Salt Reactor Experiment (MSRE) have been stored at the QOak
Ridge National Laboratory (ORNL) since the reactor was shut
down in 1969. The fluoride salt eutectic, containing 37 kg of
uranium plus plutonium and fission products, is safely con-
tained in three heavy-walled Hastelloy tanks, which are located
inside a reinforced concrete cell. Removal of these salts to a
remote location is not feasible until an appropriate repository
has been identified, built, and placed in operation. Since
this may take many years, extended storage—-in-place was criti-
cally evaluated. The evaluation, which involved a preliminary
assessment of several options for enhancing the integrity of
in-place storage, including containment improvement, the addi-
tion of chemical getters and neutron poisons, and entombment in
concrete, showed that this approach was a rational and safe
solution to the problem for the short term. Entombment is
essentially nonreversible, but the other options are open-
ended; they do not 1limit the future selection of a final dispo-
sal option. Specific actions and improvements that would
enhance safe containment during extended storage and would also
be of future benefit, regardless of which disposal option is

finally selected, were identified.
1. EXECUTIVE SUMMARY

1.1 HISTORY

The Molten Salt Reactor Experiment (MSRE) was concluded in 1969
after several years of well-planned and highly successful work. This
homogeneous reactor concept was based on the thoriunrU-233 fuel cycle
and used a molten fluoride eutectic as the operating medium. This work
is thoroughly documented.

At shutdown, the fuel salt containing most of the uranium and
fission products was divided and drained into two separate tanks, thus
ensuring criticality safety. The flush salt, containing 1 to 2% of the
uranium and fission products, was drained into a third tank. The salts
were allowed to cool and freeze, thereby precluding any leakage and
decreasing further the already-low corrosion rate. The drain tanks are
made of heavy—-walled Hastelloy N, a special alloy created for the
program, which has superior strength at high temperatures and
outstanding corrosion resistance toward the eutectic fluoride system
used for the MSRE. These tanks are contained within a hermetically
sealed, stainless steel—lined, reinforced-concrete hot cell, located
below grade except for a double set of roof plugs.

A surveillance and monitoring program, which includes daily and
monthly measurements, has been in force since shutdown. There is also
an annual reheat (but not hot enough to remelt) to recombine any
fluorine that might have formed from radiolysis by (a,n) reactions on
the fluoride salt. There have been no adverse incidents or releases of
radioactivity since the reactor was shut down 16 years ago. Several
prior studies have been made of decontamination and decommissioning (D&D)
of the facility, based on removal and reprocessing of the salts and on
the assumption that a site or repository would be available to accept
this material. In fact, there is no such site or repository at present.
Nor is it feasible to reprocess the salts without major construction of
such capability, whflch would be very costly and would introduce a finite

probability of radiation exposure or release.
The present study focused on extended storage and any enhancements
to the storage mode that would benefit the storage period and also be
beneficial in view of eventual final disposal. Time frames of 1 to 20
years (short term), 20 to 100 years (near term), 100 to 1000 years

(midterm), and more than 1000 years (long term) were considered.

1.2 PROJECTIONS

The most important of these is the radioactivity projection, which
was carried out to 1 million years by using the ORIGEN2 code. (Prior
projections were truncated at 5 and 20 years.) This projection showed
two major aspects: decay of fission product (FP) activity, as antici-
pated; and decay/ingrowth/decay of actinide activity, which is somewhat
unusual and derives from the slow ingrowth of the U-233 decay chain. The
FP activity has declined by a factor of 50 since discharge, will decline
by another factor of 8 at 100 years after discharge, and will essentially
disappear at 1000 years. The fission products are the major source of
beta and gamma activity. The actinide activity will initially decay by a
factor of 4 at 1000 years but will then grow back in to about its origi-
nal level at 40,000 years before commencing final decay.

The actinide behavior is the net result of U-232 (half-life of 72
years) and plutonium decay, along with U-233 (half-life of 158,000 years)
ingrowth and then decay. Each actinide decay spawns an additional five
or six alpha decays (along with some beta—gamma activity). The alpha
activity is important for several reasons: it is a long—term source of
neutrons from (a,n) reactions on Be-9, F-19, and Li-7, which are major
components of the fluoride eutectic, and it contributes about 50 W of
decay energy for a very long time. The neutrons must be considered in
shielding calculations but do not pose an undue problem. The decay
energy is, indirectly, a problem, not because of heat but because of

radiolysis.
Radiolysis of fluoride yields fluorine plus free metal. At slightly
elevated temperatures (>150°F) recombination is rapid enough to preclude
a buildup of fluorine; however, at lower temperatures, free fluorine will
eventually form (after an incubation period of >5 years) and will then
continue to be produced., Unless the radiolysis problem is brought under
control, disposal of the salts in the fluoride form will present a signi-
ficant problem. One possible way to limit the formation of free fluorine
is by the addition of a getter - an active metal that reacts readily with
fluorine.

The physical integrity of the drain tanks and cell is of obvious
importance. This factor was considered in terms of penetrations and
control of water. No deficiencies that would jeopardize extended storage
are evident at this time in either area, but some additional work in

these areas would be beneficial.

1.3 CONCLUSIONS

At this time, extended storage of the solidified fuel salts is the
most prudent and rational course. Actions that can be easily taken out-
of-cell to enhance storage should be implemented. An eventual decision
to remove the fuel salts to a final disposal site will be tempered by
site availability in 20 years or so. A future decision concerning
whether to reprocess or not will be controlled by the ability to limit
radiolysis, which is an open question at this time but must be resolved

in the interim.
2. INTRODUCTION

The MSRE was a graphite-moderated, homogeneous—fueled reactor built
to investigate the practicality of the molten—salt reactor concept for
application to central power stations. It was operated from June 1965 to
December 1969 at a nominal full-power level of 8.0 MW. The circulating
fuel solution was a eutectic mixture of lithium and beryllium fluorides
containing uranium fluoride as the fuel and zirconium fluoride as a che-
mical stabilizer. The initial fuel charge was highly enriched 235U,
which was later replaced with a charge of 233y, Processing capabilities
were included as part of the facility for on—line fuel additions, removal
of impurities, and uranium recovery. A total of 105,737 MWh was accumu-
lated in the two phases of operation. Following reactor shutdown, the
fuel salt was drained into two critiéally safe storage tanks and isolated
in a sealed hot cell, along with a third tank containing the flush salt.

When the reactor was first shut down, the assumption was made that
the facility and the fuel and flush salts would probably be utilized
again at some later date. Therefore, the shutdown procedure was essen-
tially a mothballing operation followed by surveillance and maintenance
(S&M) procedures. These procedures were designed to ensure safe tem-
porary storage and to maintain the operational capability of the facility.

Some years later, it became evident that the facility would not be
restarted, at least not as a molten—salt reactor. A decommissioning
study was done in which two options were considered: (1) removal of the
fuel and flush salts, followed by complete dismantling of the hot cells;
and (2) removal of the fuel and flush salts, followed by entombment of
the structural compoments within the reactor and fuel drain cells. The
second option is far less costly than the first, but both are quite
expensive because of the removal of the radioactive salts, which must be
processed and repackaged. Obviously, both options require a repository
which will accept the processed and repackaged salts. No such repository

currently exists; nor is there any assurance that one will be available in
the reasonably near future, even though there are several possibilities.

Therefore, our task is to determine what steps should be taken to extend

the storage period safely and what measures, if any, should be adopted to

enhance the present storage conditions.

In dealing with these questions, there are several time frames that

represent various operational limits. These can be defined and described

as

1.

3.

4.

follows:
Short term: 1 to 20 years. This is the period during which in=-cell
operations can still be carried out with confidence. Final disposal
options will be more clearly defined at the end of this period.
During this time span, the fission product activities will decay to
about two—-thirds of their present level.
Near term: 20 to 100 years. During this period, institutional
control can be maintained; therefore, it is the logical time to
transfer the fuel (and flush) salts to the final repository. The
fission product activities will decay to about 15% of their present
level during this time span.
Midterm: 100 to 1000 years. This is a reasonable lifetime for a
man-made concrete structure. During this time, the fission product
activities will decay to essentially zero, while the U-232 and plu-
tonium activities will decay to about 3% of their present level;
however, the U-233 activity will grow in to about 180% of its pre-
sent level.
Long term: 1000 to 1l million years. Geology will be the

controlling factor for this period. During this time, only the
U-233 decay chain will have any significance. Its activity will
peak after 40,000 years at about 900% of the present level and will

then decay permanently.
3. HISTORY

3.1 DESCRIPTION

The primary reactor and drain system components are contained within
two interconnected cells; the coolant and fuel processing systems are
located separately within adjoining cells (Figs. 1 and 2). The reactor
and drain tank cells are sealed pressure vessels that serve as secondary
containment for the fuel. The reactor cell is a 24-ft-diam steel tank,
while the drain tank cell is a stainless steel—lined reinforced concrete
rectangular tank. Each cell has removable roof beams and shield blocks
with a stainless steel membrane seal that must be cut open for access.
The coolant system and fuel process systems are located within shielded
cells that are kept at a slightly negative pressure and are swept by a
containment ventilation system. Access is gained via removable top
shield plugs. The associated equipment is housed in a steel—concrete—
transite structure that has containment features. Both the containment
cells and the high-bay area are maintained under negative pressure, with
an active ventilation system consisting of centrifugal fans and roughing
and HEPA filters that exhaust through a 100-ft steel discharge stack.
The reactor heat dissipation system included a salt-to-air radiator
exhausting through a steel stack and a drain tank for storage of the
coolant salt, where this material now resides. This stored coolant salt
is essentially nonradioactive. Ancillary facilities at the site include
an office building, a diesel generator house, a utility building, a
blower house, a cooling—water tower, and a vapor condensing system.
These facilities have been described in detail in other reports.“’5’15’18
The three cells of most concern to this study (reactor, drain, and pro-
cessing) are described in terms of penetrations and "containment enve-
lopes” in Sect. 4.3

ffjor components of the material of construction,

« A fuel-salt drain tank is shown in Fig. 3; the

 

properties and the

Hastelloy N (also c4lled INOR-8), are listed in Table 1,19
The presence o!&the solidified, stored fuel and flush salts is the

most significant aspect of the MSRE. More than 4600 kg of fuel salt and

4300 kg of flush salt, containing about 37 kg of uranium (primarily U-233)
ORNL OWG 63-4347

  
       
          
  
 
     
    

  
    
 

 
  

 
 
  

   
         

 

 

 
    
  
  
 

  
   
 
    

 
 
  
 

 
   

    
  

    
   
   
  

 

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12

OQRNL-LR-DWG 61719

INSPECTION, SAMPLER, AND
LEVEL PROBE ACGESS

   
 
  
  
 
 
   
 
 
 
 
 
  
  
 
  
  
 
  

STEAM QUTLET
STEAM DOME

GONDENSATE RETURN

WATER DOWNCOMER INLETS

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STEAM RISER

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HANGER CABLE

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HOSE WATER DOWNCCMER

 

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GAS PRESSURIZATION
AND VENT LINES

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INSTRUMENT THIMBLE
FUEL SaLl 57YSTEM
FILL AND DRAIN LINE

SUPPORT RING

FUEL SALT DRAIN TANK

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THIMBLES (32) TANK FiLL LINE

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THIMBLE POSITIONING RINGS

FUEL SALT SYSTEM

FILL AND DRAIN LINE — TANK FiLL LINE

MSR Primary Drain and Fill Tank

Fig. 3. Fuel-salt drain tank.
Table 1.

13

Composition and properties of INOR-8
(also known as Hastelloy N)

 

Chemical compositon, 7

Ni

Mo

Cr

Fe, max
C

Ti + Al, max
S, max
Mn, max
Mn, max
Si, max
Cu, max
B, max
W, max
P, max
Co, max

Physical properties:

Density, 1b/in.3

Melting point, °

Thermal conductivity, Btu/(hefte«°F)

at 1300°F

Modulus of elasticity at ~1300°F, psi

Specific heat, Btu/lb+°F at 1300°F

Mean coefficient of thermal expansion,
70 to 1300°F range, in./in.*°F

Mechanical properties:

Maximum allowable stress,? psi

1000°F
1100°F
1200°F
1300°F

17,000
13,000
6,000
3,500

0.317

2470-2555

12.7
24,8 x 10

0.138

8.0 x 10°6.

 

8ASME Boiler and Pressure Vessel Code, Case 1315.
14

and 743 g of plutonium (primarily Pu-239) are present in the drain tanks.
Calculated fission product activities (mainly beta—-gamma) of these salts,
decayed to 1985, total about 32,000 Ci. The alpha activity from tran-
suranic isotopes and their daughters amounts to about 2000 Ci. These
isotopes are divided roughly 99%Z in the fuel salt and 1% in the flush
salt. The total alpha activity of the fuel salt is very high, about
400,000 nCi/g, while that of the flush salt is about 6000 nCi/g. The
total decay heat at present is about 200 W, with three—-fourths coming
from the beta-gamma component and the remainder from the alpha emissions.
The compositiogs of the fuel, flush,and coolant salts are given in Table
2, The latter salt, which served as a secondary coolant, is not radioac-—
tive and is stored in the coolant cell. It is of interest primarily
because it contains 338 kg of high-purity Li-7.

As expected, the radiation hazards associated with the stored fuel
are significant. Gamma and neutron dose rates within the reactor and
storage cells are in the 103 rad/h range. Some of this radiation
results from (a,n) reactions with the eutectic salt base. The high
radiation field also causes the generation of some free fluorine, which
slowly accumulates. Since recombination (with the metal simultaneously
set free in the radiolysis) is accelerated by increased temperature, the
salt is reheated periodically; however, it is not melted. The stored
salts are in a stable, noncorrosive state, as dry frozen solids.

In addition to the stored fuel, the principal areas of concern at
the MSRE are the reactor components and process equipment remaining in
the below-grade containment cells. These components are internally con-
taminated and, in some cases, highly neutron activated. Exposure rates
of up to 2200 R/h have been measured in the reactor vessel, attributed
primarily to Co—-60. The inventory of residual radioactive materials in
the reactor and fuel processing cells is estimated to be several thousand
curies, with the majority of that activity being associated with fission
and activation products. The remaining cells, process piping, and asso-—
ciated operating areas are known to be slightly contaminated. The
readily accessible areas of the reactor building (including the reactor
bay and office areas) are generally uncontaminated and are being used for

laboratory and office space, as well as for storage of various materials.
15

Table 2. Stored MSRE salts

 

 

 

Fuel salt Flush salt?® Coolant salt?
Total mass, kg 4650 4290 2010
Volume, ft3 at room 66.4 69.9 42.5
temperature
Density, g/cm3 2.47 2.17 2.17
Composition, mol %
LiF 6445 66D 66D
BeF 30.3 34b 34b
ZrFu 5.0 - -
UF1+ 0.13 — ——
Uranium content, kg
U-232 c c
U-235 0.85 0.09 -
U-236 0.04 0.00 -
U—238 20 01 0.].9 -
Total 36.46 0.49
Plutonium content, g
Pu-239 657 13 -
Pu-240 69 2 ——
Other Pu 2 0 -
Total 728 15 —_
Lithium composition, 7%
Li-6 - 0.009d 0.009
Li-7 - 99,9914 99.99]

 

4Trace—element analyses of 39 batches used for both salts gave
16 ppm Cr, 39 ppm Ni, and 121 ppm Fe. Twelve other analyses of the flush
salt gave 38, 22, and 118 ppm, respectively. (Note: Could the Cr and
Ni have been interchanged?) In another series of 22 batches, the
corresponding values were 19, 25, and 166 ppm.

bRreported values. Analytical data for batches 116—161 gave 63 and
37%, calculated from reported values of 12.95 wt % Li, 9.75 wt % Be, and
77.1 wt 4 F. For batches 101-130, the calculated composition was 64.3
and 35.7%.

CPresent at 220 ppm, U-basis.

dFor batches 116~142. The values are 0.0065/99.9935 for batches
143-161.
16

The MSRE facility appears to be structurally sound and capable of
retaining the current radionuclide inventory. No significant spread of
contamination or personnel exposure has occurred since facility

shutdown.

3.2 SURVEILLANCE AND MAINTENANCE

A comprehensive maintenance and surveillance program is provided to
ensure adequate containment of the residual radioactivity at the MSRE. 11
Routine inspections of the containment systems and building services,
radiological surveillance of operating areas and ventilation exhaust,
stored~salt monitoring (temperature and pressure), and periodic testing
of safety systems are performed as part of this program. In addition,
the fuel and flush salt are reheated in order to allow recombination of
fluorine,and the containment cells are subjected to a leak test,
both on an annual basis. Facility maintenance includes general repairs,
exhaust duct filter changes, and instrumentation and controls main-
tenance. Consolidation of the surveillance instrumentation and periodic
heater and controls tests are planned as major improvements to the
current program.

The salt storage cells have been under a planned program of regular
surveillance and maintenance since the reactor was shut down in late
1969. This program includes daily observations of certain parameters by
the Waste Control Operations Center and monthly observations by the
Reactor Operations Group. The daily observations include measurements of
internal cell temperature, building—air radioactivity, pressure differen-
tials of the cell ventilation system, and stack off-gas radioactivity
levels in terms of alpha, beta-gamma, and radioiodine. These obser-
vations are recorded on log sheets (the logs have been kept since 1969).
The radioactivity data are now also computerized and coordinated through
the central Waste Control Operations Center. A primary input point to
the Center is located in the control room in 7503.

The monthly checklist involves a complete walk-through inspection
of the entire facility plus recordings of in-cell temperatures and sump
levels at seven locations. Any necessary maintenance items are noted and
added to the list of work to be scheduled. (These logs have also been
kept since 1969.)
17

On an annual basis, the three fuel and flush salt tanks are
reheated to recombine any elemental fluorine that may have formed from
radiolysis. The acceptable temperature range for the reheating is
>300°F (to ensure that the diffusion rate is fast enough) but <500°F
(well below the melting point of the salts). The reactor and drain tank
cells are also leak-tested annually, and a check is made of about 40
equipment items. In addition, special checks are made of the ven-
tilation system, two of the sump pumps, and the DOP efficiency of the
ventilation filter.

During the course of this study, it became apparent that data on
building groundwater were needed. Therefore, two additional items were
added to the S&M procedures: a periodic check of the radioactivity of
the building sump discharge, and a periodic check of the operating fre-
quency of the building sump pump. The radioactivity measurements, which
were initiated in June, have consistently shown no activity above
background. The sump pump monitoring was started on August 19, and
daily readings show an operating cycle of about 1 h/d. Rainfall during
the period was greater than normal. The pump capacity is about 35
gal/min. These measurements will be continued for at least 1 year in
order to establish a data base that covers omne complete annual weather

cycle.

3.3 PRIOR STUDIES

Decommissioning of the MSRE presents some unique problems because of
the presence of the fuel and flush salts. Plans for site decommissioning
will first have to address the issue of disposition of the fuel. 1In the
early studies, it was generally assumed that the stored fuel would be
removed from the MSRE cells, with or without reprocessing (fluorination)
to strip out the uranium and some of the fission products, and then sent
to a final repository or to retrievable storage; however, these operations
are complex and potentially hazardous and, therefore, expensive.13’15’1‘5,18
Thus far, no need for the recoverable U-233 or for the cell space has
been established. Therefore, there has been no incentive from that
direction to proceed with decommissioning. Further, there is no known
site that will accept the fluoride fuel salt for disposal in its present

condition, either now or in the near future. Consequently, it appears
18

that the solidified salts must continue to be stored in their present
mode and location, regardless of future possibilities, needs, or deci-
sions. Fortunately, this is possible because the salts, their contain-
ment, and the facility are in excellent condition. However, it is
essential to address two issues not directly considered in prior
studies:

l. Is it rational to continue storing these salts in their present
situation and, if so, what guidelines should be followed to enhance
this storage in terms of safety and stability for an extended
period of time?

2. What other options exist for the future, and what steps should be
taken in the near term to facilitate future decision—-making?

This study gives an affirmative answer to the first question and
identifies specific areas where improvements should be made. It also
pinpoints needs that should be addressed in the short term to provide
necessary information for future selection among available options.
Finally, a preliminary work plan is outlined to achieve the above

objectives.

3.4 OPTIONS

Prior studies of MSRE decommissioning have focused on the final
disposition of both the salts and the facility. Selection among options
involved many possible decision points, including the following:

A. Should the salt be chemically processed to separate the uranium

(and some fission products) from the bulk of the fluoride
salts?

B. Should the fluoride salts be chemically converted to another
form?

C. Should the salts (with or without processing and/or conversion)
be packaged and sent to an off-site repository or an on-site
storage/disposal area, or should they be disposed of in the
MSRE facility itself?

D. Should the MSRE facility be totally or partially dismantled and
portions of it entombed in concrete?

At present, there is no firm basis for selecting among these

options. Options A or B, if exercised, would require sophisticated chem-

ical processing of highly radioactive materials; new flowsheets and
19

processing equipment would also be needed (and the equipment would ulti-
mately have to be disposed of). The total cost of processing might be
as high as $10 to $20 million. No choice can be exercised on option (C)
at this time because there is no off-site or on—-site location that can
accept this material at present, and we do not have an adequate basis at
present to recommend permanent disposal in the MSRE facility itself.
Option D could cost as much as $10 million and requires a prior decision
on options A, B, and C.

In short, we need to defer a final decision. Continued storage of
the fuel and flush salts is forced upon us by external conditions. The
issues of concern, then, revolve around three questions that need to be
considered:

l. Is is reasonable and safe to plan continued storage-in-place for up
to 20 years or so?

2. Before deliberately embarking on such extended storage, what, if
anything, should be done to enhance the storage environment?

3. During this storage period, or prior to it, what should be done to
facilitate a future decision of a final nature?

In order to gain a sense of perspective, it is instructive to
define a range of options and systematically evaluate them against
various criteria. The six options that have been treated this way as
part of the evaluation are listed and defined in Table 3. (Option O,
which is essentially to continue the present practice, is not acceptable
for an extended period of time.) Options 1, 2, and 3 focus on the final
decision and bracket a wide range of possibilities; this provides a
broad, long~term perspective. Options 4, 5, and 6 focus on three
variations of enhanced storage-in-place in order to bring short-term and
near—-term needs into view. The steps used for assessing each option are
as follows:

l. List the process operations involved and compare them qualitati-
vely.

2. Evaluate them semiquantitatively on the bases of (a) process readi-
ness, (b) economics, (c) short—term hazard, (d) long-term hazard,
and (e) conservation (i.e., beneficial utilization of the buildings

and contained materials).
20

Table 3. Options for the Decontamination and Decommissioning of the MSRE

 

Option Action to Be Taken

 

0 Continue as is, with no decision. Requires continued
surveillance and maintenance, introduces maximum uncer-
tainty, and eventually still needs a decision between the

"real” options outlined below.

I Complete dismantlement of hot cells following removal of

fuel and flush salts.

2 Entombment of reactor and drain tank cells following remo-

val of fuel and flush salts.

3 Entombment of reactor and drain tank cells, with fuel and

flush salts in-place.

4 Enhanced near-term storage of solidified salts (which

still leaves all other options available).

5 Enhanced storage in-place, including remelt and addition
of getters (which still leaves optioms 1, 2, 3, and 6

available).

6 Enhanced storage in-place, including remelt, addition of
getters, and repackaging (which still leaves options 1, 2,

and 3 available).

 
21

3. Identify the limiting factor(s).
Of necessity, these evaluations are subjective and only semiquantitative
at this stage.

The first evaluation (Table 4) lists the major process steps that
would be required for the various options and indicates which steps are
required in each case. The steps were defined in such a way that each
represents a more—~or—less equivalent effort or cost. Evaluation then
simply requires counting the number of steps for the particular option
in question. Clearly, from this point of view, options 1 and 2 are pro-
hibitively complex (and costly) and would not be selected unless there
were an overriding reason for doing so.

The second evaluation involves "scoring™ each option for every
applicable process step, as identified in Table 4. These results are
tabulated in Appendix B. Scoring, which was done on five different
bases, was set up so that a low score is more favorable than a high
score. The available scoring ranges were selected to balance the rela-
tive importance of each of the five factors, on the assumption that
near—term hazard, long—term hazard, and conservation were each about
twice as important as either process readiness or economics. Scores for
each criterion, as well as totals, are summarized in Table 5. The
totals were converted to a more normal scale of O to 100, with a high
score more favorable, by taking reciprocals of each total and then nor-
malizing by dividing by the largest value (the 0.0161) and multiplying
by 100. On this basis, Options 3, 4, and 5 are distinctly'superior to
Options 1, 2, and 6.

The third evaluation (Table 6) identifies limiting factors that
would disqualify an option from further consideration at present. While
this evaluation is the least quantitative, it is, nonetheless, highly
significant because it seeks to identify controlling factors. The
limiting factors are those which, if not within an acceptable range,
would eliminate that option no matter how favorable the other aspects
might be. Thus, in terms of the five criteria already established,
Options 1 and 2 should not be considered further at this time. When two
additional criteria were introduced, based on anticipated public reac-
tion and on reversibility regarding future choice of options; Option 3

was also removed from consideration.
22

Table 4. Process steps required for MSRE options

 

Option

 

Process step 1 2 3 4

 

Melt salts (and add getters)
Remove salts

Build salt process facility
Process or convert salts/U
Package salts (and U)
Dismantle/dispose of facility

Transport packaged products

PO XK R KR oKW
|
I

Store/isolate products

Decontaminate cells/equipment

Dispose of liquid LIW

|
I
|

Remove equipment
Dispose of solid LIW
Dismantle cell structure

Dispose of solid LLW/rubble

PP PSP R K b e K R ) e e

Restore area

Seal pipes/penetrations

Internal entombment

i
KoK e
Ko s

t

{

External structures
Stabilize drain—tank cell - - - X

Continue surveillance - - - X

Number of more-or—-less
equal process steps
involved 15 11 3 3

 
Table 5.

Summary of semiquantitative evaluation

 

 

 

Option
Factor 1 2 3 4 5 6
A. Process readiness 47 41 15 15 19 23
B. Economics 54 47 15 15 20 28
C. Near-term hazard 66 47 14 16 21 31
D. Long—~term hazard 4 6 12 17 12 10
E. Conservation 14 6 6 6 6 6
Total 185 147 62 69 /8 98
Reciprocal 0.0054 0.0068 0.0161 0.0145 0.0128 0.0102
Normalized 34 42 100 90 80 63

 

tc
24

Table 6. Limiting factors for MSRE options

 

Factor 1 2 3 4 5

 

A, Process readiness

Decision now? XXX XXX
Decision in 5-20 years 777 7277
B. Economics XXX XXX
C. Near—term hazard XAX XXX

D. Long-term hazard

Minimal for all options

E. Conservation XXX

F. Institutional perception
of anticipated public

reaction
Decision now XXX
Decision in 5-20 yearsP 777
G. Reversibility/future XXX XXX XXX

choice of options

 

AThere is no identified location at this time where the removed salts
could be taken.

bpublic reaction may become more rational with time, or institutional
leaders may regain public confidence.
25

The results of all three evaluations are consolidated in Table 7.

In summary, in terms of these three evaluations, and at this time:
l. Options 1 and 2 are rejected on the basis of each evaluation.
2. Option 3 is rejected, based on limiting factors but recognizing

that public reaction may change in future years, and that lack

of reversibility may not be detrimental when judged against more

definitive criteria.

3. Option 6 deserves further consideration based on the semiquan-
titative evaluation.

This leaves Options 4 and 5, and possibly Option 6, available for
short-term selection. These options are addressed in Sects. 4 and 5.
Some topics relevant to the other options are also addressed; however,
much of the information presented is of broad and general applicability,

and is germane to all options.
26

 

 

 

Table 7. Options and evaluations for decontamination and decommissioning
sScored
Option Description ab B¢ cd

0 Continue as is, with no decision. Requires con-
tinued surveillance and maintenance, introduces
maximum uncertainty, and eventually still needs
a decision between the "real” options outlined
below.

1 Removal of fuel and flush salts; complete dis- 15 185 5
mant iement of hot cells.

2 Removal of fuel and flush salts; entombment of 11 147 4
reactor and drain tank cells.

3 Entombment of drain tank cell, with fuel and 3 b2 2
f lush salts in-place.

4 Enhanced near-term storage of solidified salts 3 6Y U
(which still leaves all other options available).

5 Enhanced storage in-place, including remelt and 4 78 v
addition of getters (which still leaves Options
1, 2, 3, and 6 available).

6 Enhanced storage in-place, including remelt, 6 98 0

addition of getters, and repackaging (which still
leaves Options 1, 2, and 3 available).

 

81n all cases, a low score is superior.
bNumber of process steps involved.
CSemiquantitative rating of process steps.

dNumber of Limiting factors.
27

4. PROJECTIONS

4.1 RADIOACTIVITY

The original decay calculations, made using ORIGEN and MSRE-specific

S Subsequently, these data have

input data, were truncated at 5 years.
been extrapolated to the year 2000.1¢ However, any decisions regarding
permanent emplacement require decay data out to a million years, as well
as a detailed knowledge of radiological properties in the 10— to 10,000~
year time span. Ideally, the entire calculation (generation and deple-
tion) should be repeated using ORIGENZ2, the updated and improved version
of ORIGEN. However, ORIGEN2 does not have a molten salt reactor model,
and construction of such a model would be prohibitively expensive.
Therefore, the approach was to take the old ORIGEN output data at
discharge and then make the decay éalculations with ORIGEN2. When these
two sets of data at 5 years after discharge were cross—checked, only
inconsequential or readily explainable differences were noted. This
gives us confidence in using the ORIGEN-generated results as our starting
point.

The truncated ORIGEN data were presented in four groupings: fission
products in both grams and curies, and actinides in both grams and
curies. At least one activation product (Zr-93, from the fluoride salt
mixture) was included in the old ORIGEN runs. The two sets of lists
(grams and curies) were generally mutually exclusive; that is, isotopes
listed on the grams output (because they were present in a quantity
greater than 0.1% of the total mass) were generally not included on the
curies output (where they appeared if contributing greater than 0.17% of
the total radioactivity). By combining these two sets of lists, we
believe that the input data for our ORIGEN2 calculations were acceptably
complete.

The summary results are given in Tables 8 through 11 (fission pro-

ducts and actinides; grams and curies) for these times:
RE
SR
SR
SR

Y

Y
ZR
ZR
ZR
ZR
ZR
ZR
ZR
NB
NB
MO
MO
yO

87
88
89
90
89
91
90
91
92
93
Sy
95
96
93
95
95
97
98

MO100

TC
aUu

g°
99

RU101
RU102
RU103
RU104
RU106
RH103
PD105
PD106
PD107
AG107
SB125
TE125

TE127M

TE128

TE1294

TE130
1127
1129

XEB129

Cs137

BA137

BA138

Lat139

CE140

CE1W

CE142

CE144

PR141

ND143

ND144

ND145

Table 8.
grams of fission products and daughters

DISCHARGE

8.480E+00
5.640E+00
S.573E+0Q0
9.893E+01
6.660E+01
T.458E+00
4,440E+00
9.730EB+01
1. 100E+Q2
1.190E+02
1.1B0E+02
9.306E+00
1. 140 E+02
0.0

4,397E+00
2.040F+01
3.250E+01
3.160E+01
2.850B+01
2.980E+01
0.0

1.950E+01
1.560E+Q1
2.292E+00
7.880E+00
2.238E+00
5.07T0E+01
1.820E+01
4.610E+00
5.680E+00
0.0

6.176 E-01
0.0

4.175E-01
9.46 QF+00
8.859E-01
2.0908+01
0.0

1.490E+01
1.287E+02
5.560E+00
T.440E+01
1.700E+02
1.890E+02
1. 439E+01
1.730E+02
3.979E+(1
1.700E+02
1.670=2+402
1. 130E+02
1.070E+02

SUMMARY

ORIGEN2 output for MSRE salts:

TABLE:

28

CONCENTRATIONS,

GRAMS

ENTIRE CORE (MSRE WITH U-233 FUEL)

1.01IR

8.480E+00
S.640E+00
3.705E-02
9.660E+01
7. 214E+01
9.846E-02
6.768E+00
1.047E+02
1. 100E+02
1. 190E+02
1. 180B+02
1.779E-N
1. 140E+02
3.979E-06
2. 194E-01
3.772E+01
3.250E+01
3.160E+01
2.850E+01
2.980E+01
9.697E-05
1.950E+0%
1.560E+01
3.659E~-03
7.880E+00
1.125E+00
5. 299E+01
1.820E+01
5.723E+00
5.640E+00
6.018E~07
4.809E-01
1. 404E-01
4.092E-02
9.460E+00C
4.T31E-04
2.090E+01
3.889E-01
1.579E+01
6.972E-07
1.258E+02
8.5C0E+00
T.440E+01
1.700E+02
1.890E+02
5.972E-03
1.73CE+02
1.633E+0
1.844E+02
1.670E+02
1.365E+02
1.070E+02

10.01R

8.480E+0Q0
S.64 0E+00
9.390E-22
7.798E+01
7.217E+01
1.201E-18
2.580E+01
1.048E+02
1.100E+02
1.190E+Q2
1.180E+02
6.080E-17
1.140E+02
1.379E-04
T-416E~17
3.812E+01
3.250E+ 01
3.160E+01
2.850E+01
2.980E+01
9.687TE=-04
1 .950E+01
1.560E+01
2.357E-28
7.880E+00
2.309E-03
5.299E+01
1.820E+01
6.846E+00
5.640E+00
6.018E-06
5.057E-02
5.767E~01
3.4188-11
9.460E+0Q0
Y.67T3E-33
2.090E+01
4.299E~-01
1.579E+01
6.972E-06
1.021E+02
3.211E+01
T.440E+01
1.700E+02
1.890E+02
2.183E-33
1.730E+Q2
5.395E-03
1.8B44E+02
1.670E+02
1.528E+02
1.070E+02

100.0%R

8.480E+00
5.64 QE+0Q
0.0

S.154E+00
T.217E+01
0.0

9.424E+01
1.04 8E+02
1.100E+02
1.190E+02
1.180E+02
¢.0

1.140E+02
4.393E-03
0.0

3.812E+01
3.250E+01
3.160E+01
2.850E+01
2-S7S5E+G?
9.696E-03
1.950E+01
1.56 0E+G1
0.0

7.880E+Q0
3.060E-3C
5.299E+01
1.820E+01
6.848E+00
5.64 0E+00
6.018E-05
8.368E-12
6.280E-01
0.0

9.450E+00
0.0

2.050E+01
4.299E-01
1.579E+01
6.972E-05
1.277E+01
1.215E+02
7.4 0E+01
1.700E+02
1.890E+02
0.0

1.730E+02
8.319E-38
1.844E+02
1.670E+02
1.528E+02
1.070E+02

1.0KY

8.480E+00
S.640E+00
0.0

4.552E-09
T.217E+401
0.0

1.034E+02
1.048E+02
1.100E+02
1. 189E+02
1« 180E+Q2
0.0

1.140E+02
5« 290E~=02
0.0

3.812E+01
3.250E+01
3.160E+01
2.850E+01
2-370E+0i
9.681E-02
1.350E+01
1.560E+01
0.0

7.88CE+0QQ
0.0

S5.295E+01
1.820E+01
6.848E+00
5.639E+00
6.018E-04
0.0

6.28B0E-01
0.0

9.460E+00
0.0

2.090E+01
4,299E-01
1.579E+01
6.972E-04
1.1888~-08
1.343E+02
7.440E+01
1.700E+02
1.890E+02
¢.0

1.730E+02
0.0

1.844E+02
1.670E+02
1.528E+02
1.070E+02

10.0KY

8.480E+00
5.640E+00
0.0

0.0

7.21TE+01
0.0

1.034E+02
1.048E+02
1.100E+02
1. 185E+02
1.180E+02
0.0

1.140E+02
5.369E-01
0.0

3.812E+01
3.250E+01
3.160E+01
2.850E+01
2.8858+01
9.541E-01
1.950E+01
1.56Q0E+01
0.0

7.880E+00
G.0

5.299E+01
1.820E+C1
6.84BE+00
5.634E+00
6.0158-03
0.0

6.280E-01
0.0

9.460E+00
0.0

2.090E+01
4.299E-01
1.578E+01
6.371E-03
0.0

1.343E+02
T.440E+ 01
1.700E+02
1.890E+02
0.0

1.730E+02
0.0

1.844E+02
1.670E+02
1.5288+02
1.070E+02

100.0KY

8.480E+00
S.640E+00
0.0

0.0

7.217E+01
0.0

1.034E+02
1.04BE+02
1.100E+Q2
1.137E+02
1.180F+02
0.0

1.180E+02
5.270E+00
0.0

3.812E+01
3.250E+0Q1
3.160E+01
2.850E+N
Z2e152E+01
8.27TE+Q0
1.950E+01%
1.560E+01
0.0

7.880E+00
0.0

5.299E+01
1.820E+01
6.848E+00
5.580E+00
5.986E-02
0.0

6.280E-01
0.0

9.460E+00
0.0

2.090E+0
4.299E-Q1
1.572E+01
6.9578=02
0.0

1.343E+02
7.440E+01
1.700E+02
1.890E+Q2
0.0

1.730E+02
0.0

1.844E+02
1.670E+02
1.528E+02
1.070E+02

1.0MY

802+00
UOE+CO

17E+0Q1

4
6
0
0
2
0

1.034E+02
1.048E+02
1.100E+02
7.564E+01
1.180E+0Q2
0.0
1.140E+02
4.335E+01
0.0
3.812E+01
3.250E+Q1
3.160E+01
2.850E+01
1.151E+00
2.865E+01
1.950E+01
1.560E+01
0.0
7.880E+00
0.0
5.299E+01
1.820E+0Q1
6.8482+00
5.069E+00
5.708E-01
0.0
6.280E~-01
0.0
9.460E+00
0.0
2.090E+01
4.295E-01 .
1.511E+01
6.821E-01
0.0
1.343E+02
7.440E+01
1.700E+02
1.850E+02
0.0
1.730E+02
0.0
1.844E+02
1.670E+02
1.528E+02
1.070E+02
ND146
ND14g
ND15G
PM147
Puidgy
SM147
smMtus
SM150
SM151
SM152
EU 151
EU152
EU153
EU154
EU155
GD154
GD155
SUMTOT

TOTAL

DISCHARGE

8.410E+01
4,630E+01
1.900E+01
4,.011E+01
4.912B-02
2.390E+01
0.0

2.710E+01
5.585E+00
1.460E+01
0.0

3.1108-02
5,450E+00
1.303E-01
7.629E=01
0.0

0.0

2.709E+03

2.7T09E+03

SUMMARY

Table 8.

TABLE:

29

(continued)

CONCENTRATIONS,

GRAMS

ENTIRE CORE {MSRE WITH U-233 FUEL)

1.01IR

8.410E+01
4.630E+01
1.900E+01
3.080E+01
1.069E-04
3.321E+01
4,9018-02
2.T10E+01
5.542E+00
1. 460E+01
4.285E~02
2.955E-02
5.450E+00
1. 202E-01
6.634E-01
1.00698-02
9.951E-02
2.709E+03

2.709E+03

10.0YR

8.410E+01
4.630E+01
1.900E+01
2.856E+00
1.166E-28
6.115E+01
4$.912E-02
2.710E+ 01
5.171E+00
1. 46 1E+0Q1
4.140E-01
1.868E-02
5.450E+00
5.820E~02
1.886E-01
7.210E-02
5.743E-01
2.709E+03

2.709E+03

100.0YR

8.410E+01
4.630E+01
1.300E+01
T.345E-10
0.0

6.401E+01
4.912E-02
2.7T10E+01
2.585E+00
1. 46 2E+01
3.000E+00
1.90 3E-04
5.450E+00
4.118E-05
6.453E-07
1.303E-01
7.629E-01
2.709E+03

2.709E+03

1.0KY

8.410E+01
4,.630E+01
t-900E+01
0.0

0.0

6.401E+01
4,912e-02
2.710E+01
2.524E-03
1. 462E+01
5.582E+Q0
2.286E~-24
S.450E+00
1.294E-36
0.0

1.303E-01
7.629E-01
2.709E+03

2.709E+03

10.0RY

8.410E+O
4.630E+01
1.9CO0E+O1
0.0

0.0

6.401E+01
4.9128-02
2.710E+01
1.582g~-33
1.462E+01
5.585B+00
0.0

5.450BE+00
0.0

0.0

1.303E-01
7.6298-01
2.709E+03

2.TC9E+ Q3

100.0KY

8.410E+01
4.,630FE+01
1.9C0E+Q1
2.0
0.0
6.401E+01
4.912E-02
2.T10E+01
0.0
1.462E+01
5.585E+00
0.0
5.450E+00
0.0
0.0
1.303E-01
7.629E~01
2.709E+03

2.709E+03

1.0H4Y

8.4 10E+01
4.530E+01
1.9G0E+01
.0

01E+0Q1
12E-02
10E+01

62F+01
85E+00

0
0
6
4
2
o
1
5
0
S5.450E+00Q
0

0

QOLFCUVMEONWYEFEC

1.303E-01
7.629E-01
2.709E+03

2.709E+03
SR
Sk

Y

Y
ZR
ZR
NB
NE
NE
TC

89
90
90
91
93
95
934
95
95¥
9s

RU103
RU106
RH103X
RH106
PD107
SN123M
SB125
TE125H4
TE127
TE127M
TE129
TE129HM
1129
CS5134
€5137
BA137H
CETul
CE142
CE1u4
PR144
PR1GYN
PMI4T
PM148H
SM151
E0152
EU154
EU1SS
SUMTOT

TOTAL

Table 9.

DISCHARGE

1.620E +05
1.350E+04
1.360E+04
1.830E+05
2.991E-01
2.000E+05
0.0

1.720E+05
4.149E+03
5.054E-01
7.400E+04
7-.491E+03
7.401E+04
8.949E+03
2.902E-03
1.390E+0Q2
6.380E+02
1.870E+02
3.2792+04
3.940E+03
9.790E+04
2.670E+04
2.631E-03
5.519E+00
1.120E+04
1.050E+04
4.101E+05
4.153E~-06
1.270E+05
1.280E+405
0.0

3.720E+04
1.050E+03
1.470E+02
5.380E+00
3.519E+01%
3.550E+(2
1.800E+06

1. 800E+Q6

SUMMARY

30

ORIGEN2 output for MSRE salts:
curies of fission products and daughters

TABLE:

RADIQCACTIVITY,

CURIES

ENTIRE CORE (MSRE WITH U-233 FUEL}

1. 01IR

1.077E+03
1.318E+04
1. 319E+04
2-416E+03
2-991E-01
3.824E+03
1. 412E-02
8.584E+03
2.837E+01
5.054E-01
1.181E+02
3.766E+03
1.065E+02
3.766E+03
2.902E-03
0.0

4.967E+02
1.216E+02
3.782E+02
3.862E+02
9.282E+00
1.426E+01
2.789E-03
3.944E+00
1.094E+04
1.035E+04
1. 702E+02
4.153E-06
5.211E+04
5.211E+04
6.254E+02
2.856E+04
2.286E+00
1. 459E+02
5. 113E+00
3.246E+01
3.087E+02
2.068E+405

2.068E+05

10.01R

2.729E-17
1.064E+0U4
1.064E+004
2.9U8E-14
2.991E-01
1.307E-12
1.135E~-01
2.901E-12
9.694E-15
5.054E-01
7.611E-24
7.730E+00
6.861E-24
7.730E+00
2.902E-C3
0.0

5.224E+01
1.275E+01
3.160E=-07
J.226E-07
3.281E-29
5.0418-29
2.789E=-03
1.9148-01
8.889E+(03
8.409E+03
6.220E-29
4.1538-06
1.722E+01
1.722E+G1
2.066E-01
2-649E+03
2.492E-24
1.361E+02
3.232E+00
1.572E+01
8.773E+071
4.159E+04

4.159E+04

100.0¥R

0.0
1.209E+03
1.269E+03
¢.0
2.99 1E-01
0.0
2.824E=-01

52E-01

OO OO0

20E-26

2U4E-26
02E-03

OO0 a0 OOVLDLHOD

48 E-09
09E-09

789E~03
1.389E-14
1. 111E+03
1.051E+03
0.0

4.153E-06
2.655E-34
2.6558-34
3.186E-36
1.247E-07
0.0

6.804E+01
3.292E-02
1. 11 2E-02
3.002E-04
4.730E+03

4.730E+03

1.0KY

0.0
6.212E-07
6.214E-07
0.0
2.990E-01

89E-03

34E-06
83E-07

43E-02
55E=-22
95E-34

= OEF PR CUUa4O]OO0ONOoOUOOOOORUOOOOOOOPO

1.159E+00

10.0KY

1.076E+00

100.0KY

0
0
0
0
859E-01

16E-01

50E-M

T1E-03

T6E-03

0.
0.
0.
OO
2.
0.
2.
0.
0.
3.
OQ
0.
0.
0.
2.
0.
00
0.
G.
0.
0.
0.
2.
0.
0.
0.
0.
4.153E~-06
0.

0.

0.

0.

Q.

0.

0.

0.

0.

9.

0
2
0
0
6
¢
0
0
0
8
0
0
0
0
0
0
0
7
0
0
0
0
1
0
0
0
0
0
0
0
0
0
2

B1E-01
9.281E-01

1.0MY

01E-01

06E-01

52E-02

08E-03

6BE-03

53E-06

WOOODOOOODO0ODOFOOOONOUCOOOONCOODOOa OO0 aOmOco0O
8 & & 2 4 r B R 84BN T F MY AR NN A s s A Y Y R
WOOOCOCOO0OO0OOOMO00CO0OOUOOOOO 000 QUOoOOEEQQUoOeO

56E-01

3.956E-01
HE 4
PB206
PB207
PB208
BIZ209
EA226
TH229
TH230
TH232
0232
0233
0234
U235
U236
7238
NP237
PU239
PU240
pPu2u41
AM241
SUMTOT

TOTAL

Table 10.

DISCHARGE

OO0 O0OOOCO

-
-
-
-
-
-
-
-
-
-

0
0
0
0
0
0
0
0
0
7

846E+00
3.232E+04
2.911E+03
9.860E+02
6.800E+01
2.370E+03
0.0

6.223E+02
7.501E+01
8.T42E+Q0
2.793E-01
3.937E+04

3.937E+04

SUMMARY

31

ORIGEN2 output for MSRE salts:
grams of actinides and daughters

TABLE:

CONCENTRATIONS,

GRAMS

ENTIRE CORE (MSRE WITH U-233 FUEL)

1.0YR

1. 0u45E-02
9.789E-15
8.4 14E-14
6.277E-02
5. 396E-06
3.587E-08
1.389E~-01
8.111E-03
1. 979E-086
7. T7T1E+00
3.232E+04
2.9%1E+03
9.860E+02
6.801E+01
2.370E+03
7.761E-04
6.223E+02
7.500E+01
8.331E+00
6.894E-01
3.937E+04

3.937E+04

10.0YR

1.051E-01
3.171E-10
8.777E-11
6.435E-01
5.926E-04
3.5828~06
1.388E+00
8.111E-02
1.980E-05
T.1268+00
3.232E+04
2.911E+03
Q.862E+02
6.808E+01
2.370E+03
3.273E-02
6.221E+02
TW493E+ 01
5.402E+00
3.586E+00
3.937E+04

3.937E+04

100.0YR

8 .354E-01
2.170E-06
5.077E-08
4.437E+00
5.9 1E-02
3.535E-04
1.382E+01
8.107E~-01
1.990E-04
2.996E+00
3.231E+04
2.910E+03
9.878E+02
6.878BE+01
2.370E+03
1.065E+00
6.205E+02
TLU22B+00
7.096 E-02
7.868E+00
3.937E+04

3.937E+04

1.0KY

4.4587E+00
4.088E-03
8.727E-06
7.207E+00
5. 7948400
3.108E-02
1.323E+402
8.064E+00
2.089E-03
5.161E-04
3.218E+04
2.903E+03
1.003E+03
7.542E+01
2.370E+03
7.028E+00
6.046E+02
6.T46E+0 1
1.085E-20
1.875E+00
3.337E+04

3.937E+04

10.0KY

T.195E+01
2.036E+00
8.893E-04
T.207TE+00
4.4C5E+02
1.217E+00
8.765E+02
7.647E+01
2.802E-02
1.207E-41
3.094E+04
2.830E+03
1.139E+03
1.162E+02
2.3708+03
8.847E+00
4,.666E+02
2.598E+01
0.0

1.015E~06
3.937B+04

3.937E+04

100.0KY

1.142E+03
2.208E+02
6.807E-02
7.207E+00
9.361E+03
9.393E+00
9.958R+02
4,560E+02
3.917E-01
0.0

2.08B7E+04
2.193E+03
1.56 3E+03
1.414E+Q2
2.370E+03
8.593E+00
3.492E+01
1.866E~03
0.0

0.0

3.937E+04

3.938E+Qu

1.0MY

3.6 19E+03
2.352E+03
1.303E+00
7.207E+00
2.861E+04
1.576E+00
1.947E+01
7.7T12EB+01
4.045E+00
0.0

4.081E+02
1.711E+02
1.596E+03
1.376E+02
2.370E+03
6.4 20E+00
1.525E-10

0.0
0.0
0.0
3.938E+04

3.938E+C4
TL207
TL208
TL209
PB209
PB210
PB211
PB212
PB214
BI210
BI211
BI212
BI213
BI214
P0O210
PO212
Po213
PO214
P0O215
PG216
Pc218
AT217
RN219
RN220
RN222
FR221
RA223
RA224
RA225
RA226
AC225
aca27
TH227
TH228
TH229
TH230
TH231
TH234
PA231
Pa233
PA234H
0232
0233
U234
0235
u23e
U238
NP237
PE238
PU239
PU240
PO241
Am24n
cnzu2
SUMTOT

TOTAL

Table 11.
curies of actinides and daughters

DISCHARGE

S1E+01

302402

30E+02

CUVOoOCOODUNOOOCOFO

BOE+01

30E+02

30E+Q2

30E+02

30E+02

a2 @ 4 & % ¥ B & & 5+ & 3 B % M B s & & B s ¥ s e % s s e

OO0 U0 R0O0 a0V a00OLOO0OWNO
OO0 COCUNOOLOONOCODOCUooCOoOUNOoOocOoO~NOoOo

1.680E+02
3.130E+02
1.820E+01
2.132E-03
4.401E-03
7.971E-04
0.0

1.000E+00
3.8702+01
1.710E+01
9.0102+02
9.590E~ Q1
2.450E+01
2.553E+03

2.553E+03

SUMMARY

ORIGEN2 output for MSRE salts:

TABLE:

ENTIRE CORE

1.0YR

7.084E~10
5.672E+01
6.385E~04
2.956E-Q2
3.591E-10
7. 104E-10
1.579E+02
3.547E-08
3.591E-10
7. 104E-10
1.579E+02
2.956E-02
3.547E-08
1. 164E-10
1.011E+02
2. 892E~-02
3.546E~-08
7.104E-10
1.579E+402
3.548E-08
2.956E-02
7.104E-10
1.579E+02
3.548E-08
2.956E-02
7. 104E-1¢C
1.579E+02
2.956E- 02
3.548E~-08
2.956E-02
7-104E-10
7.006E-190
1.573E+02
2.956E-02
1. 638E-04
2. 132E-03
7.97T1E~04
4.511E-08
5.473E-07
7.371E-04
1.664E+02
3.130E+02
1.820E+01
2. 132E-03
4.402E~03
7.971E-04
S5.473E-07
1.090E+00
3.870E+01
1. 710E+01
8.586E+02
2.367E+00
5. 193E+00
2.525E+03

2.525E+03

10.0YR

6.46TE-08
S.618E+01
6.382E-03
2.955E-01
3.395E~07
6.486E-08
1.564E+02
3.542E-06
3.396E-07
6.486E-08
1.564E+02
2.955E-01
3.542E-06
3.396E~07
1.002E+02
2.891E-01
3.542E-06
6.486F=08
1.564E+02
3.543E-06
2.955E-01
6.486F~08
1.564E+02
3.543E-06
2.955E-01
6.486F-08
1.564E+02
2..955E-01
3.543E-06
2.955E-01
6.U82E~08
6.396E-08
1.562E+02
2.955E~-01
1.638E-03
2.133E-03
T7.971E-04
4.5148-07
2.3088-05
7.971E-04
1.526E+02
3.130F+Q2
1.820E+01
2.1338-03
4.406E-03
7.971E-04
2.308E-05
1.040E+00
3.869E+01
1.708E+01
5S.567TE+02
1.231E+01
4.511E-06
2.206E+03

2.206E+03

32

RADIOACTIVITY,

CURIES

(MSRE WITH U-233 FUEL)

100.0%R

3.144E~-06
2 .36 8E+01
6 .354E-02
2.942E+00
1.94 1E-04
3.153E-06
6 .590E+01
3.495E-0Q4
1.94 1E-04
3.153E-06
6.590E+01
2.942E+00
3 .495E-04
1.941E-04
4,222E+01
2.878E+00
3.495E-04
3.153E-06
6.590E+01
3.496E-04
2.94 2E+00
3.153E-06
6.530E+01
3.496FE-04
2.94 2B+00
3.153E~06
6 .590E+01
2.942E+00
3.496E-04
2.94 2B+ 00
3.153E-06
3.110E-06
6.590E+0Q1
2.9 2E+00
1.637E-02
2.136E~-03
T.97T1E-04
4.511E-06
7.508E-04
7.97T1E-04
6.415E+01
3.129E+02
1.819E+01
2.136E-03
4.452E-03
7.971E-04
7.508E-04
5.108E-01
3.859E+01
1.692E+ 01
7-313E+00
2.701E+01
0.0

9.704E+02

9.704E+02

1.0KY

4.492E-05
4,.086E-03
6.07T9E-01
2.815E+01
3.073E-02
4,504E-05
1.137E-02
3.074E-02
3.073E~02
4.504E-05
1.137E=-02
2.815E+01
3.074E~-02
3.073E=-02
7.286E-03
2-754E+01
3.073E-02
4.504RE-05
1.1378-02
3.074E-02
2.8 15E+01
4.504E-05
1.137E-02
3.074E-02
2.815E+01
4,5048-05
1.137E-02
2.8B15E+01
3.074E-02
2.8 15E+01
4.504E-05
4.442E=-05
1.137E~-02
2.815E+01
1.628E~01
2.170E-03
7.971E-04
4,503E-05
4 .956E-03
7.971E~Q4
1.105E=02
3.116E+02
1.815E+01
2.170E-03
4.881E-03
7.971E=-04
4.956E-03
4.173E-04
3.760E+01
1.538E+01
1.118%-18
6.439E+00
0.0

6.1U9E+02

6.149E+02

10.0KY

4,396E-04
1.105E-09
4.029E+00Q
1.865E+02
1.204E+0C
4.4 Q9E-04
3.074E-09
1.204E+00
1.204E+00
4.4 09E-04
3.074E-09
1.865E+02
1.204E+00
1.2C4E+00
1.970E-09
1.825E+02
t.2C4E+0Q0
4.409E-04
3.074E-09
1.204E+G0
1.865E+02
4.409E~04
3.074E-09
1.204E+00
1.865E+02
4.409B-0U
3.074E-09
1.865E+02
1.204E+00
1.865E+02
4.409E-04
4.348E-04
3.074E-09
1.865E+02
1.544E+ 00
2.463E-03
7.971E-04
4.407E-04
6.239E-03
7.9712~-04
2.584E-40
2.996E+02
1.769E+01
2.463E-03
7.520E-03
7.971E-04
6.239E-03
5.541E-35
2.901E+01
5.922E+00
0.0

3.487=~06
0.0

1.857E+03

1.857E+03

100.0KY

2.795E-03
1.544E-08
4.577E+00
2.119E+02
9.287E+00
2.803E-03
4% .298E-08
9.289E+00
9.287E+0Q0
2.803E-03
4.298E-08
2.119E+02
9.289E+00
9.287E+00
2.754E-08
2.073E+02
9.287E+00
2.803E-03
4.298E-08
9.291E+00
2.119E+02
2.803E-03
4.298E-08
9.291E+00
2.119E+02
2.803E-03
4.298E-08
2.1T19E+02
9.291E+00
2.119E+02
2.803E~-03
2.764E-03
4.298E-08
2.119E+02
9.208E+00
3.3818-03
7.971E-04
2.803E-03
6.060E~03
7.971E-04
0.0

2.021E+02
1.371E+01
3.381E-03
9.1498-03
7.971E-04
6.060E-03
0.0

2.171E+00
4.25UE-Q4
0.0

0.0

¢g.0

2.006E+03

2.006E+03

1.0MY

3.442E-03
1.595E-07
8.9518-02
$.144E+00
1.558E+00
3.452E-03
4,%938E-07
1.558E+00
1.558E+00
3.452E-03
4,.438E-07
4.144E+00
1.558E+00
1.558E+00
2.844E-07
4.055E+0Q0
1.558E+00
3.452E-03
4.438E-07
1.559E+0Q
4.144E+CO
3.4528-03
4.438E-07
1.559E+00
4 . 144E+Q0
3.4522~-03
4.438E-07
4.184E+00
1.559E+00
4.144E+00
3.452E~03
3.404E-03
4.438E-07
4,144E+Q0
1.557E+00
3.452B-03
7.970E-04
3.452E-03
4.,527E-03
T.970E-04
0.0

3.9528+400
1.069E+00
3.452E-03
8.909E-03
7.970E-04
4,527E-03
0.0

S7E-11

8E+01

5.382E+01
33

Discharge (these are the input data, as taken from ORIGEN)
1 year

10 years

100 years

1000 years

10,000 years

100,000 years

1 million years

Data were also calculated for those times used in the original

ORIGEN rums:

384 d (1 year and 19 d, corresponding to January 1, 1971;
used in the old ORIGEN calculations)

749 d (2 years and 19 d)

1115 d (3 years and 19 d)

1480 d (4 years and 19 d)

1845 d¢ (5 years and 19 d)

Agreement between the two outputs was checked at the 1845-d output

and found to be excellent, with minor exceptions. A detailed listing is

given in Tables 12 through 15. In summary, the exceptions are as

follows:

Fission products (in grams): Of a total of 45 isotopes, 31 agreed
exactly and the other 14 differed by no more than 47%. The total FP
mass was 2710 g in both cases.

Fission products (in curies): Of a total of 29 isotopes, 7 agreed
exactly, 6 differed by minor amounts (1 to 4%), 3 differed signifi-
cantly (17% less, 50% less, and 2507 more), and 13 differed by large
factors (20 to 800). All of the large differences were in minor
isotopes that contributed less than 0.l1% of the total activity,
which agreed very well (5.70 x 10% vs 5.71 x 10% Ci). The causes of
these discrepancies are twofold: revised half-lives, and a numeri-
cal error in ORIGEN for the very low activities.,

Actinides (in grams): Excellent agreement for all nine isotopes;

 

two agreed exactly, five were within 1%, one minor constituent
(Pu-241) was within 14%, and the major contributor (U-233) was

within 2%Z. The difference for U-233 was traced to a revision in the
34

Table 12.

Reconciliation of ORIGEN/ORIGEN2 Results
I: Actinides and Daughters, in Curiles
Curies after 5 yr
Isotope Half-life ORIGEN ORIGEN2 Comments
In secular equilibrium with U-232¢
In secular equilibrium with U-232
In

secular equilibrium with U-232
secular equilibrium with U-232°€

T1-208 3.05 min 5.82 El
Pb=-212 10.6 hr 1.62 E2
Bi-212 60.6 min 1.62 E2
Po-212 0.3 usec 1.03 E2 1.04 E2

R
—t
o

Po-216 0.15 sec 1.62 E2 a In secular equilibrium with U-232

Rn-220 55.6 sec 1.62 E2 a In secular equilibrium with U-232

Ra-224 3.64 d 1.62 E2 a In secular equilibrium with U-232

Th-228 1.91 yr 1.62 E2 1l.61 E2 In secular equilibrium with U-232

U=-232 72 yr 1.60 E2 a

7-233 158 E3 yr 3.13 E2 a

U~234 245 E3 yr 1.82 E1 a

Pu-238 88 yr 1.09 EO 1.08 EQ

Pu-239 24.1 E3 yr 3.87 El a

Pu-2450 6540 yr 1.71 El a

Pu-241 l4.4 yr 6.88 E2 7.07 E2 ORIGEN used half-life of 13.0 yr

Am-241 432 yr 6.97 EO 7.40 E0 Daughter of Pu-241; ORIGEN half-life
was 470 yr

Cm—242 163 d 3.81 E-2 0,97 E-2  Assumed ORIGIN included a precursord

Subtotal 2.37 E3 2.40 E3 Difference due mainly to Pu-24]

Total 2.38 E3P 2.40 E3 Difference due mainly to Pu-241

 

4Same wvalue as ORIGEN.
bThis total includes some minor contributors not included in the subtotal.

CBranching decay product of Bi-212.
dAm-242 m (152 yr); 0.04 Ci at discharge would explain the difference.
35

Table 13.

Reconciliation of ORIGEN/ORIGEN2 Results
I1: Actinides and Daughters, in Grams
Grams after 5 yr

Isotope Half-l1ife  ORIGEN ORIGEN2 Comments

U-232 72.0 yr 7.49 EQ 7.47 EQ

U-233 158 E3 yr 3.30 E4  3.23 E4 ORIGEN used half~life of 162 E3 yr
U-234 245 E3 yr 2.94 E3 2.91 E3 ORIGEN used half-life of 247 E3 yr
U-235 704 E6 yr  9.86 E2 a ORIGEN used half-life of 711 £6 yr
U~236 23.4 E6 yr- 6.80 E| 6.81 El ORIGEN used half-life of 23.9 E6 yr
U-238 4,47 E9 yr 2.37 E3 a ORIGEN used half-life of 4.51 E9 yr
Pu-239 24.1 E3 yr 6.31 E2 6.22 E2 ORIGEN used half-life of 24.4 E3 yr
Pu-240 6540 yr 7.74 E1 7.50 E1 ORIGEN used half-life of 6758 yr
Pu-241 14.4 yr 6.03 E0 6.86 EO ORIGEN used half-life of 13.0 yr
Subtotal 4,01 E4 3.94 E4 Difference due mainly to U-233
Total 4.01 E4  3.94 E4 Difference due mainly to U~233

 

2game value as ORIGEN,
I1I:

36

Table 14.

Curies after 5 yr

Reconciliation of ORIGEN/ORIGEN2 Results

Fission Products, in Curies

 

Isotope Half-1ife ORIGEN ORIGEN2 Comments

Sr-89 50.5 day 8.67 E-4 .016 E-4 Numerical error in ORIGIND

Sr-90 29.1 yr 1.19 E4 a

Y-90 64 hr 1.19 E4 a

Y-91 58.5 day 8.43 E-=3  .059 E-3 Numerical error in ORIGIND

Zr-95 64.0 day 7.31 E=2  .042 E-2  Numerical error in ORIGIND

Nb-95 m 87 hr 7.76 E-4 .03l E-4 Numerical error in ORIGINC

Nb-95 35 day 8.56 E-=2  ,096 E-2  Numerical error in ORIGIND

Ru-103 39 day 1.79 E~7 .005 E-7  Numerical error in ORIGIND

Rh=103 m 56 min 8.94 E-8 .049 E-8 Numerical error in ORIGINC

Ru—-106 368 day 2.30 E2 2.32 E2 ORIGIN used half-life of 366 days

Rh-106 30 sec 2.30 E2 2.32 E2 In secular equilibrium with Ru-106

Sn=123 m 40 min 7.99 E=2 0 ORIGIN used half-life of 129 daysd

Sb-125 2.77 yr 1.83 E2 1.80 E2 ORIGIN used half-life of 2.70 yr®

Te-125m 58 day 8,70 El 4.4 El Daughter of Sb-125; revised
branching ratiof

Te-127 m 109 day 1.09 EO .032 EQ Numerical error in ORIGINP

Te-127 9.35 hr 5.39 E-1 .31 E-1 Numerical error in ORIGINC

Te-129 m 33.6 day 6.40 E-10 .008 E-10 Numerical error in ORIGIND

Te-129 $9.6 min 2.05 E-10 .005 E-10 Numerical error in ORIGINC

Cs~134 2.06 yr 1.00 EO a

Cs-137 30.0 ¥r 9.95 E3 a

Ba-137 m 2.55 min 9.30 E3 9.43 E3

Ce~141 32.5 day 1.52 E-9 003 E-9 Numerical error in ORIGIND

Ce~144 284 day 1.41 E3 a

Pr-144 17.3 min 1.41 E3 a

 
37

Table 1l4. (continued)
Reconciliation of OQRIGEN/ORIGEN2 Results

II1I: Fission Products, in Curies (Concluded)

 

Curies after 5 yr

 

Isotope Half-life ORIGEN ORIGEN2 Comments

Pm—-147 2.62 yr 1.02 E4 .98 E4

Pm-148 m 41,3 day .61 E~8 .004 E-8 Numerical error in ORIGEND

Sm-151 90 yr 1.42 E2 a

Eu~152 13.6 yr 4,02 EO 4,16 EO ORIGEN used half-life of 12 yr
Eu-154 8.6 yr 2.83 El 2.34 E ORIGEN used half-life of 16 yr
Eu-155 4.96 yr 5.14 E1  17.52 El  ORIGEN used half-life of 1.8 yr
Gd-162 8.6 min 8,27 E-2 0 ORIGEN used half-life of 1,00 yr8&
Tb~162 m 7.6 min 8.27 E~2 0 In transient equilibrium with Gd-162
Subtotal 5.71 E4 5.70 E4

Total 5.71 E4 5.70 E&

 

4same value as ORIGEN.

bA11 of these had half-lives of 32 to 109 days. ORIGEN gave the correct
values at | yr, but then changed to a longer half-life (about 50% longer).
Later versions of ORIGEN do not have this error,

“Daughter of a precursor that suffers from the error described in ff "b".

doRIGEN reversed the half-lives of $n~-123 and Sn~123 m.

€0RIGEN used a longer half-life during the first year.

fORIGEN used 47%; later value is 23%.

8This is an erronecus value used by ORIGEN, later versions of ORIGEN used
10.4 min, before going to 8.6 min.

 
38

Table 15.
Reconciliation of ORIGEN/CRIGENZ Results

 

IA'H

Grams after 5 yr

Fission Products, in Grams

 

Isotope Half-life  ORIGEN ORIGEN2 Comments

Rb-87 4,7 E10 yr 8.48 EO a ORIGIN used half-life of 5.0 E10 yrP
Sr-88 stable 5.64 EQ a

Y-89 stable 7.24 E1 7.22 E1

Sr-90 29.1 yr 8.44 E1 8,77 El ORIGIN used half-life of 28.1 yr
Zr-90 stable 1.57 El a

Zr-91 stable 1.05 E2 a

Zr—-92 stable 1.10 E2 a

Zr-93 1.5 E6 yr 1.19 E2 a

Zr-94 stable 1.18 E2 a

Mo-95 stable 3.97 El1 3.81 El

Zr-96 stable l.14 E2 a

Mo-97 stable 3.25 El a

Mo-98 stable 3.16 El a

Te-99 2.1 ES yr 2.99 E1 2.98 El

Mo-100 stable 2.85 El a

Ru-101 stable 1.95 El a

Ru-102 stable 1.56 El a

Rh~103 stable 5.30 E1 a

Ru~-104 stable 7.88 EO a

Pd~-105 stable 1.82 El a

Pd-106 stable 6.77 EO 6.78 EO

Pd-107 6.5 E6 yr  5.64 EO a ORIGIN used half-life of 7.0 E6 yrP
Te-128 stable 9.47 EO 9.46 EO

I-129 1.57 E7 yr 1.58 El a ORIGIN used half-life of 1.70 E7 yrb
Te-130 stable 2.09 El a

Cs-137 30.0 yr 1.14 E2 1.15 E2

Ba-137 stable 1.97 El a

 
39

i Table 15. (continued)

Reconciliation of ORIGEN/ORIGEN2 Results
IV: Fission Products, in Grams (Concluded)

Grams after 5 yr

 

Isotope Half-life ORIGEN ORIGEN2 Comments
Ba-138 stable 7.44 El a
La-139 stable 1.70 E2 a
Ce-140 stable 1.90 E2 1.89 E2
Pr-141 stable 1.84 E2 a
Ce~-142 stable 1.73 E2 a
- . . . ; c
gg_%zz ggzbée i.Z% gfl i.gg EEI Assume ORIGEN included a precursor
Nd-144 2.1 E15 yx 1.52 E2 a ORIGIN used half-life of «P
Nd-145 stable 1.07 E2 a
Nd~146 stable 8.41 El a
Pm—-147 2.62 yr 1.10 El 1.06 El Assume ORIGEN included a precursord
Sm=-147 1.06 E11 yr  5.46 El 5.35 El ORIGIN used half-life of «P>€
Nd-148 stable 4,63 E1 a
Nd-150 stable 1.90 El a
Sm—150 stable 2.71 E1 a -
Sm~-151 90 yr 5.21 EO 5.37 ED ORIGIN used half-life of 87 yr
Sm=-152 stable 1.46 El a
Eu=153 stable 5.49 EO 5.45 EO
Subtotal 2.71 E3 a
Total 2.76 E3f 2,71 E3

 

a5ame value as ORIGEN.

bror very long-lived isotopes the difference in half-lives has no effect
over a 5-yr period.

Cpr-143 (13.6 days); 6 grams at discharge would explain the difference.
dNd-147 (11 day); 1.6 grams at discharge would explain the difference.
€ORIGIN had more precursor, Pm—-1l47.

fThig total includes some minor contributors not included in the subtotal.
40

half-life from 162,000 to 158,000 years. This accounts for the
U-233 and also for the small difference in the total actinide mass
of 4.01 x 10" g vs 3.94 x 10% g, The difference for Pu-241, a minor
constituent, also results from a revised half-life of l4.4 years vs
the prior value of 13 years.

® Actinides (in curies): Excellent agreement; of a total of 17 isoto-

 

pes, 11 agreed exactly, 5 were within 6%, and an explanation was

found for Cm~242, a minor constituent where ORIGEN2 was 75% low. A

minor precursor for Cm—-242, Am—242m (152—years half-life), would not

have been shown in the summary ORIGEN printout because it was below
the 0.1%Z cutoff; however, after 5 years, it would contribute the

major part of the total Cm—242, which has a half-life of only 163 d.

The old ORIGEN calculations were quite thorough, including allowance
for activation of the eutectic salt and rather precise modeling of the
U-235 and U-233 fuelings and operating cycles. Allowance was made for
the continuous gas sparging, which resulted in partial stripping of tri-
tium and rare gases; also, the fluorination processing after the U-235
operation, which removed not only U, but also H, He, Se, Br, Kr, Nb, Mo,
Tc, Ru, Te, I, Xe, and Np, was taken into consideration. Both Tc-99 and
I-129, of potential concern for permanent emplacement, were in the grams
printout. We were able to identify only two factors that were not
included: activation of corrosion products, and activation by neutrons
from (a,n) reactions on the Be-9 and F-19 in the eutectic salt mixture.
Both of these are very minor contributors.

A helpful view of the ORIGEN2 results is given in Figs. 4 and 5.
Figure 4 shows the total FP activity and the major contributors out to 1
million years. Figure 5 shows the total actinide activity and the major
contributors, also to 1 million years. Both drawings were made to the
same scale. The former is essentially all beta-gamma contributors, while
the latter is mostly alpha emitters. The actinide decay chain does
include T1-208 (from the U-232 chain), which is noteworthy because of a
very energetic (2.6 MeV) gamma accompanying the beta decay to 298ph. 1In
each case, a relatively few isotopes contribute most of the activity over
a rather distinctive time span. Two observations are particularly

noteworthy:
41

ORNL DWG 85-438

 

106

105

104_

103 —

1()2 |

TOTAL RADIOACTIVITY (Ci)

101 |

100 |

 

  
  
  
   

i —

I 1 l | |

TOTAL FISSION PRODUCTS AND
DAUGHTERS

—-— Ce-144 /Pr-144 AND Pm-147

---------- Sr-90/Y-90 AND
Cs-137/Ba-137m

\ ————— Zr-93/Nb-93m AND Tc-99

(The Zr-93 is an activation
product.)

— ——— - = w—
e W S SW
—

 

| | | | 1

 

10-1
10°

Fig. 4.

10 102 10° 104 10° 106
TIME AFTER DISCHARGE (years)

Fission product activity of MSRE fuel and flush salts.
42

ORNL DWG 85-439

 

 

 

 

 

106 T 1 I I I
105 - —
TOTAL ACTINIDES AND DAUGHTERS
—-— U-232 CHAIN
........... U_233 CHA'N
~ 104~ /  eeee- Pu-239, Pu-240, AND -
o Pu-241/Am-241
>.
=
>
-
o
<I
©
o
<
a
-
<{
-
O
-
10"1 | | | ] |

10° 101 102 103 104 10° 10°
TIME AFTER DISCHARGE (years)

Fig. 5. Actinide and daughter activity of MSRE fuel and flush salts,
43

l. Fission product activity dominates at present, by a factor of 15,
but will decline by a factor of 10 over the next 100 years and be at
a minimal level after 1000 years.

2. Actinide activity, presently controlled by plutonium and the U-232
chain, is decaying slowly, while the U-233 chain is very slowly
growing in. The result is a minimum in alpha activity at 1000
years, along with the depletion of the T1-208 gamma, followed by a
maximum in alpha activity at about 40,000 years; therefore, per-
manent decay occurs.

The total activity has decayed dramatically since discharge, from
1.8 million Ci to 34,000 Ci. Most of this decline stems from the deple-
tion of short-lived fission products, which accounted for 200,000 Ci at 1
year after discharge and now, after 15 years, amounts to 32,000 Ci. The
actinides have decayed slowly, from 2550 Ci at discharge, to 2520 Ci
after 1 year, to 2000 Ci after 15 years.

It can be seen that the gamma activity will continue to decay with
half-lives of about 2, 30, and 72 years and will essentially disappear
after 1000 years, while the alpha activity will fluctuate within one
order of magnitude for the next 500,000 years before declining per-
manently. Since alpha decay contributes about one-fourth of the decay
energy at present, this long—term alpha contribution is pertinent con-
cerning radiolysis to yield fluorine.

The neutron activity of the salts from (a,n) reactions is much
greater than normal because of the presence of Be-9, F-19, and Li-7.
These isotopes, particularly the Be-9 and F-19, are excellent targets for
(a,n) reactions. At present (15 years after discharge), there are about
6 x 102 neutrons/s from this source. Spontaneous fission also contribu-
tes some neutron activity. These values, listed in Table 16, follow the
actinide curve with time. Less than 1% of the alphas are converted to
neutrons.

The transuranic alpha activity of the salts is relatively high,
about 70,000 nCi/g in the fuel salt and about 1,000 nCi/g in the flush
salt. (The total alpha activity, including that from all the daughters,

is about six times greater.) Removal of the uranium and neptunium by
44

Table 16. Neutron activity of fuel salt

 

(a,n) neutrons/s

 

 

Time

(years) IBe 19 71i S.F.& neutrons/s

Discharge 1.8 E9 4,7 E9 1.6 E7 1.86 Eb6

1 1.8 E9 4.7 E9 1.6 E7 1.75 Eb

10 l.6 E9 4.1 E9 l.4 E7 1.69 E6

100 7 E8 1.8 E9 6 Eb 9.30 E5

1000 4 E8 1.1 E9 4 E6 5.23 E5

10,000 1.3 E9 3.5 E9 1.2 E7 1.74 E6

100,000 1.4 E9 3.7 E9 1.3 E7 1.89 E6

1,000,000 4 E7 1 E8 0.4 Eb 4,64 E4

 

8gpontaneous fission.
45

fluorination could lower the activity of the flush salt below the
threshold level for TRU waste, but the fuel salt would still be a TRU

waste because of the plutonium content.

4.2 RADIOLYSIS

One of the more singular side effects deriving from the use of a
fluoride salt eutectic as the matrix material for the MSRE fuel is the
production of fluorine in the cooled salt as a result of radiolysis.
This effect was first observed in 1962 in small test capsules, where a
fluorine overpressure was noted after removal from a heated environment ., 2
Subsequent experiments have confirmed that fluorine forms 1in cooled
fluoride salts as a consequence of gamma radiation and recoil from the
reaction ,

yp Ladiation, .,
or

M'F, radiation M'F +F

followed by

F+F » Fsy.
After an incubation period, gaseous F, may be observed. At elevated tem—
peratures, recombination (the reverse reaction) is fast enough to prevent
accunulation of Fj. The recombination reaction is temperature dependent
(activation energy of 19.4 kcal/mol) but relatively independent of
fluorine pressure. This is consistent with a solid-state diffusion
mechanism, which would account for the primary aspects of fluorine
generation, the incubation period, and the rate—controlling factors in
the recombination reaction.

Prior analyses suggested that for the MSRE fuel salt the production
rate for radiolytic fluorine (in 1984) would be about 26 em3/h (at STP)
and the induction period would be about 5 years.7’16 For this reason,
the annual reheat was instituted after shutdown. A temperature of about
350°F is utilized, which is well below the melting point of the salt and
well above the estimated temperature of 145°F where the recombination
rate equals the formation rate.

In any discussion of extended storage, fluorine production is of

possible concern for two reasons: pressure rise, and corrosion. The
46

following analyses show that neither is a concern for extended storage,
but both must be considered in terms of permanent disposal. In the
latter case, mitigation can possibly be achieved by the addition of a
getter — an active metal which will react readily with elemental
fluorine.

To calculate the estimated production rate of Fyp, we use
Haubenreich’ s data? for the net yield of Fo at ambient temperature:

0.020 molecule/100 eV, or 0.17 cm3 (STP)/W-h. The induction period is
calculated from: 3.0 x 1022 eV/g, or 1.3 W~h/g.

Table 17 gives the thermal power for actinides (and daughters) and
for fission products, as calculated by ORIGEN2. These results, which are
graphed in Fig. 6, show that there is a rapid decline in thermal power
during the first 100 years and actinide activity is the controlling fac-
tor thereafter. These curves are, of course, an analogue to the radioac-
tivity curves shown previously, except that the actinide contribution is
relatively greater since the average energy per alpha decay is greater
than for beta-gamma decays. This also introduces conservatism in the
calculations. Haubenreich's data? were obtained for gamma radiation, but
alpha radiation has a much higher linear energy transfer (LET).
Therefore, the free radicals produced by alpha decay will be much closer
together, and recombination is more probable. The induction periods and
net fluorine yield rates were calculated using Haubenreich's values;
these values are also given in Table 17. (The actual F, production
rates are about twice the net yields, but some recombination is occurring
even at ambient temperature.)

The data in Fig. 6 were interpolated to estimate induction periods
and Fy yield rates during the 10- to 100-year period; the results are
given in Table 18. 1In 35 years (50 years after discharge), the Fj
yield rate will fall to half its present value and will continue dropping
until it reaches a minimum at about 1000 years. It will remain at an
average value of about 8 cm3/h for the next 500,000 years before com—

mencing a final decline.
4.,2.1 Pressure Rise

The pressure rise in the fuel salt drain tanks can be calculated
from the Fy yield rate and the void volumes in the tanks. Each tank (50
in. diam by 86 in. high) has a capacity of 80 ft3, and the volume of
47

Table 17. Thermal power and calculated fluorine yield rate

 

Thermal power (W)

 

Induction Fluorine

 

Time Actinides Fission period yield
(years) and daughters products Total (years) (cm3/h)
Discharge 49 5,655 5,704 a a
1 50 682 732 0.9 124
10 49 116 165 4 28
100 29 14 43 16 7
1,000 17 - 17 40 3
10,000 48 - 48 14 8
100,000 51 - 51 13 9

 

1,000,000 1.3 - 1.3 530 0.2

4531t is molten; recombination is very fast.
THERMAL POWER (W)

ORNL DWG 85-1098

 

200 —1v

160 —

120 —

®
o
l

DH
@
e

 

|\|II|II|II|FI|II

 

 

100

Fig. 6.

 

TIME (years)

Thermal power of MSRE fuel and flush salts.

8%
49

Table 18. Radiolysis data for the 10- to 100-year period

 

 

Time after Estimated Induction Fluorine
discharge? thermal power period yield
(years) (W) (years) (em3/h)

10 165 4.2 28

15 140 4.9 24

20 122 5.7 21

30 98 7.0 17

40 83 8.3 14

50 72 9.6 12

60 63 11.0 11

70 57 12.1 10

80 51 13.5 9

90 47 14,7 8

100 43 16.0 7

 

8The reactor was shut down on December 12, 1969,
50

solidified fuel salt is 66.4 ft3. Since the fuel salt is divided between
two tanks, the void volume of each tank is 47 ft3. This provides con-
siderable room for expansion. Assuming that the 1985 reheat, completed
in June-July of this year, is the last one, an incubation period of 5
years would yield gaseous fluorine starting 20 years after discharge.

The calculated pressure rise per year in the 20th year would be 2.0 psi.
Table 19 gives the calculated values out to 100 years. The annual
pressure increase, although quite small, is not insignificant enough to

be neglected over a period of years.

4.,2.2 Corrosion Rates

 

Corrosion data for INOR-8 have been obtained only at temperatures
above the melting point of the salt because this was the obvious area of
interest for reactor operation. Corrosion by the molten fluoride eutec—
tic is extremely slow and, below the melting point, would be even slower.
In the solid state, the corrosion rate is limited by diffusion processes.
In addition, as a corrosion product builds up, it serves as a diffusion
barrier and may act as a protective coating. Chromium is the most
vulnerable component, which was predicted theoretically and confirmed
experimentally.8 After initial reaction with any impurities and oxide

films is completed, there is a very slow reaction with UF,:
Cr + 2UF, » CrFo + 2 UF3 .

Several years of MSRE operation yielded an estimated corrosion rate of
only 0.1 mil/year at operating temperature (1100-1400°F).8

Corrosion by elemental fluorine, in the molten state, is far more
severe. During reprocessing operations, which used F, to strip out ura-
nium by conversion to UFg, a corrosion rate of 0.1 mil/h was estimated
while in the molten condition (50°F above the liquidus temperature or
about 800°F).® Again, chromium is the most vulnerable component by vir-
tue of its larger negative free energy of formation for the fluoride
salt relative to the other components of INOR-8. The attack on nickel
was estimated at 0.04 mil/h, while that for chromium was 0.14 mil/h.
This selective attack on the chromium is in keeping with the chemical

properties of the two metals.
51

Table 19. Calculated pressure rise in fuel salt drain tanks,
assuming last reheat was in 1985

 

 

Time after Pressure rise Total pressure
discharge?® (years) rate (psi/year) buildup (psi)
20 2.0 2
30 1.6 20
40 1.4 35
50 1.2 48
60 1.1 59
70 1.0 70
80 0.9 80
90 0.8 88
100 0.7 95

 

4The reactor was shut down on December 12, 1969,
52

The situation for the solid state is quite different. This is
clearly evidenced by the fact that free fluorine can even be formed in
the solidified salt since, for this to happen, elemental Li, Be, or Zr
must also be formed. All of these metals are more reactive chemically
than Ni, Mo, Cr, or Fe, the major components of INOR-8. Further, they are
dispersed throughout the salt on a molecular scale as a result of the
radiolysis process. The implication is that elemental fluorine is quite
unreactive with metals when solid—-state diffusion is the rate-controlling
process. Even in the exposed parts of the tank, where gaseous diffusion
allows free access by the liberated fluorine, any initial attack would
immediately build up a layer of metal fluoride, after which solid-state

diffusion would become rate controlling.
4,2.3 Use of Getters

The preceding discussion suggests that radiolysis can be regarded as
a minor factor over the next 10 to 20 years but must be considered for
prolonged storage. The addition of chemically active metals, to serve as
getters, has been proposed. Comparison of the amount of F, that might be
released and the void space available in each tank for the addition of
getters shows the idea to be feasible. Assuming an average F, yield rate
of 15 cm3/h (Table 18), the calculated production over the next 80 years
is 470 mol. Active metals that are easily available are listed in Table
20, along with the reactions and the densities of the metals. The volume
of metal required to react with 470 mol of Fj, was calculated from this
and was found to be quite small. Considering the declining radiolysis
rate, there is enough void volume in each tank for metal to consume the
total F, yield over the next 10,000 to 100,000 years.

At 100 years, the calculated 470 mol of F, represent about 0.67% of
the fluoride inventory in the tanks. There would be about 3000 mol of
F,, or about 4% of the total fluoride, at 1000 years. In any event, as
fluorine is consumed by reaction with getter, the concentration of unre-
combined free metal in the body of the solid salt would increase to the
same degree and would, therefore, increase the rate of the recombination
reaction. Thus, a time would come when recombination equaled production

and the net Fy yield became zero.
53

Table 20. Potential fluorine getters

 

Volume of metal

 

Density of required, to
Active metal consume 470 mol
metal Reactions (g/cm3) of Fo (£t 3)
Be Be + Fy, + BeFj 1.85 0.082
Ca Ca + Fo > CaFo 1.55 0.43
Zr Zr + 2F5 » ZrFy 6.4 0.12
Ti CTi + 2F5 > TiFy 4.5 0.09
Al Al + 1.5F, » AlF3 2.7 | 0. 24
Mg Mg + Fp » MgFp 1.74 0.23

 
54

Other factors to consider are the manner and the degree of contact
between the solid salt and the getter metal. Since the F, has found its
way to where pressure can be measured, simply adding metal flakes, chips,
or sponge to the void space above the salt may be adequate. If finely
divided metal were also added to the bulk salt (while in a molten con-
dition and then allowed to freeze), recombination within the solid salt
would be enhanced. Metal with a density nearly identical to that of the
salt is required in order to achieve good dispersion throughout the salt.
This can be achieved by alloying Be, Mg, or Ca with Al, Zr, or Ti in
suitable proportions.

Before the addition of getters is undertaken, additional data are
needed in two areas: the storage tanks themselves, and the performance
of candidate getter metals. Concerning the former, a means for monitor-
ing for Fj release should be included. This could be done either via
measurements of pressure rise in the sealed storage tanks or via periodic
analyses for F; in swept samples. Meanwhile, tests should be conducted
with various metals to determine the actual efficacy of each in reacting

with F, at ambient temperature and in various configurations.

4.3 INTEGRITY OF THE FACILITY

The apparent integrity of the hot cell and the contained fuel salt
and flush salt is very high. In many ways, the facility already has the
characteristics of an engineered repository. The uranium is in a solid
form and is being held in a configuration that is safe against criticali-
ty. The containers are made of heavy-walled Hastelloy N (INOR-8), which
is highly resistant to corrosion by fluoride. They are stored in a
heavily shielded hot cell made of reinforced concrete with 3-ft-thick
walls (largely underground) and a double~layer 5-ft-thick roof. The
cell, lined with a welded skin of 1/8-in. stainless steel, is gas tight
and leak-tested annually; it is monitored for temperature and pressure on
a regular basis. However, the facility was not designed for permanent or
extended emplacement, and three factors need attention — cell penetra-
tions, control of groundwater, and secondary containment — before this
can be seriously considered. These factors are addressed in Sects.

l].. 30 ].""'40 3.3.
55

4.3.1 Cell Penetrations

The hot cells were built for complex operations involving material
transfers and extensive use of electrical heating. All functions were
fully instrumented with redundant power supplies, control circuits, tem-
perature and pressure readouts, etc. Therefore, the three main operating
cells (reactor cell, drain tank cell, and fuel processing cell) have
numerous penetrations for electrical signals plus a large number of
material transfer lines through the outer cell walls. There are also two
large cell-to-cell openings: a 36-in. opening between the reactor and
drain tank cells, and a l4-in. opening between the drain tank and repro-
cessing cells (the latter is sealed). These openings carry insulated,
heated lines for the transfer of molten salt, plus other lines that
almost fill these openings. Tables 2] and 22 list the penetrations for
the reactor cell and drain tank cell, respectively. A more comprehensive
tabulation, including the processing cell, is given as part of Appendix D.
In summary, these penetrations can be grouped and described as follows:
Reactor Cell

Numbers I—XXIV:

Penetrations are from 4 to 36 in. in diam (many 8 and 24 in.); most

are fitted with multiple lines, for materials, electrical, ther-—

mocouples, and off-gas. Only one is a "spare;" four carry dual
water lines.
Drain Tank Cell

Numbers 1—30:

Single penetrations are from 3/4 to 14 in. in diam; they are used to
move material [steam, water, helium, and molten salt; nine were used
by the Chemical Technology Division (Chem Tech) for reprocessing]
and are located as follows:

- South wall: 18 lines, mostly l-in.;

- West wall: 2-3/4-in. lines;

—_ North wall: the nine used by Chem Tech and the l4-in. salt
transfer opening; eight of the nine are 1-1/2-in.,

and one is 4-in.
i1l
Iv
y
1!
Vil
Yill
Ix
I
Xl
ill
Xill
{1V
Y
XVl
X¥I1
WL
X1X
XX
ixl
iE1
FX1
X

Huitiple &0
Hultiple 44
fultiple 44
Hultiple &0
Muitiple &%
Gpen

Open

Huitiple 8
Blanked in Cell
Multipie
Hultiple i
Single

Furnace

Hultiple
Hultiple
Hultiple
Heltiple
Hultipie

Furnace

Hultipie
Muitiple

fipen Duct i
Hultiple 38
Open

Pl G = P B P Y PR e e Ll O

Table 21.

Reactor Leak DBetectors
Electrical
Electrical

Thermocouples

Instrumentation

Samplier ffgas

Sampler

Fuel Pump Aux: Fiping
Neutron lastrusent Tube
Fuel Puap Liquid Level
Fuel Puep Aux. Piping
Coaponent Coolant Air
Coolant Salt to HX
Water Lines

Spare

Water Lines

Water Lines

Water Lines

Coolant Salt to Radiator
Ottgas

Otfgas

Cell Exhaust Duct
Thermocouple

Brain Tank Cell intercon
Sump?

Cell Location

Ft-in., Degrees

836
634
834
834
836
847
B47
836-9
834-5
B44-4
B36-9
829-10
840-10
839-9
839-%
B39-9
§39-9
839-9
837
839-9
839-%
824-10
836

.B25-2

Bottos Center

15
30
43
b9
75
1o
13
125
143
153
160
143
1790
183
200
203
210
220
220
22
230
243
323
330

MSRE reactor cell penetration list

e ————— - —————

 

Sub-unit sizes

 

TH3/4% 2B%1/2" 2583 /B* IPS
60 € 3/B*1PS

Iel/2%, §4374%, I91VIFS

Arcess frea Size Reference Drawings
Generals EGGD-40704,41487,41489,41450
South ES5A 24" DKKD-40974 EBED-418463 EBBB-41B&4 DJJD-35494 DIID-40495 60 % 1/4" tubing
South E5A 24" DKKD-40976 EBBD-4i863 EBBD-41B64 EMNI-56230 ENMI-5624h 6 % 1*1FS, 38 ¥ 3/4"1PS
South ESA 24" DKKD-40976 EBED-41863 EBBD-41B64 EMMI-56230 EMMI-54246 & * 1“1PS, 38 % 3/4°IPS
South ESA 24" DKKD-40976 EBED-41863 EBBD-41864 DHHB-35567
South ESA 28" DKKD-40976 EBRD-41863 EBBD-41B44 DHHB-35567
High Bay 4 DKKD-40%73 DKKD-30974 {2 #1/2%)
High Bay 4" DKKD-40973 DKKD-40974 DBBC-4133%
Servite Tunnel (B" DKED-40717 EKKD-40735
High Bay 36" DKKD-40716 EKKD-40715 EHHA-4179%

Special Eq. Ra.
Special Eq. Re.
Special Eq. Ae.
Coolant Celi
Coalant Cell
Coolant Cel}
Coolant Cell
Coclant Cell
Coolant Cell
Coolant Cell
Coolant Cell
Coolant Cell
{07 Tunnel

West Tunnel
Drain Tank Cell

24"
b
6!
30"
28"
3&*

DKKD-40973 DKKD-40%75 EGBD-55411 EJID-535428
DKKD-44718 EKKD-40737 EGBD-35411

DKKD~-40714 EGGD-55411
EKKD-40711 EGBI-3349B
DKKB-40746 DKKD-40741
DKKD-40740 DKKD-40741
DKKB-40740 DKKD-40741
DEKD-40740 DEKD-20741
IKKR-40740 DKKG-40741
DKKD-464712 EBBI-35498
DEKB-40740 DKKD-40741
DKKD-40740 DRKD-40741
DKKD-4471G £EKKD-40749

DKKD-40976 EBED-41863 EBBD-4:B4&4 DHHB-55367

EEKE-40713
ERBI-55493

10 % lon Chasber Buides
6 ¢ 1/4* tubing
2+ 1"1PS, 11 ¥ 1/2*1P5

Jet"
2 ¥ 1-18

38 ¢ 3/8"1PS

 

 

 

9¢
Table 22, MSRE drain tank cell penetration list

 

1L, Type Subsets lsage Lell Locatian Accesc frea Size Reterence Drawings Sub-unit sizes
1 Single | Steam fros Domes South Wail South ESA 3" in 4° DDKD-40945 ERGD-55425
2 Single 1 Stear froam Domes Couth Mall Scuth ESA 3" in &° DDKD-30945 EBGD-554%5
3 Single | Water to Steam Domes South Wall South ESA I DOrD-40948 EGGD-35425
4 Bingle 1 Water to Steam Doaes South Hall South ESA 1" DDKD-40948 ERGD-5542
3 Single | Water ‘ South Wall South ESA 1" [DED-4094E DKXB-412533
& Single 1 Hater South Kall South ESA 1" BOED-40948
7 Single 1 Spare South Wall Bouth E54 Iy DiKD-40948
B Singie ! Spare South Hall South ESA I DBED-40948
g Single { Heliua South Wail South ES8 1/2% in 1° DI¥D-40548
10 Sirgle ! Heliua South Wall South ESA 1/2* in 1" DDKD-40948
it Single 1 Heliua South Wail South €54 172" in 1" DDET-4G948
{2 Single { Heliue South Wall South ESA 1/2% in 1" DDER-A094E
i3 Single I Helium South Wall South ESA 142" in 1*  DDKB-40948
14 Single 1 Heliua South Wail South ES4 1/2" in 1° DDKD-40948
15 Single 1 Heliua South Wall South ESA /2" in 1" DDKD-40748
14 Single ! Helium Gouth Wall South ESA §/2" in 1" DDKD-40948
17 Single I Sump Discharge South Wall Waste Cell 314 DOKD-40948 DKKB-41280 DKKB-412681
18 Single 1 Sump Discharge South Wall Waste Cell 34" DDKD-40948 DEKB-412B0 DKKR-41281
19 Single 1 Air to Sump West Wall West of Bldg. 3/4" DDKD-40948 DEKB-31280
s Single 1 Air to Sump West Wall West of Bldg., 3/47 DDKD-40948 DKKE-412B0
24 Single {Chea. Tech.) North ¥all Fuel Proc, Cell 1-1/2° DKED-40949
22 Singie {Chen. Tech.) North #all Fuel Proc. Cell i-1/2" DriD-40949
23 Single (Chem. Tech.) Narth Wall Fuel Froc, fell 1-1/2° DEKD- 40949
24 Singie (Ches. Tech.!} North #all Fuel Proc. Cetl 1-1/2° DKED-40%49
5 Single {Chem. Tech.} North Wall Fuel Proc. Cell t-1/2° DEXD-40349
z Single {Chea. Yech.) North Wall Fuel Proc, fell &7 DKED-40249
27 Single {Chem. Tech,) North Kali Fuel Proc., Cell 1-1/2° DKKD-40749
Z Singie {(Chee. Tech.) North Wall Fuel Proc. tell 1-1/2° DKED-40949
29 Single {Chem. Tech.) North Mail Fuel Froc. fell 1-1/2° DKKD-40949
in Furnace Salt Transfer North Wall Fuel Froc. Leil 14" DKED-4094%
Ai-1 to 34 BHank 29 Instrueentation East Wall Narth ESA 3ig" DKKD-40947 [HHB-35347
B-1 to 34 Bank 36 Instrumentation East Kall North ESA 374" DKED-40947 DHHB-55347
£-1 tp 34 Bank 36 Theraocouples East Wall Korth ESA 34" DKkD-40947 DHHE-33567
D-1 to 34 Bank 16 Thereocouples East Wall North ESA 3/4" DKKD-401947
E-1 to 36 Bank 14 Thermocouples East Wall North ESA iy DExD-40947
f-37 to 50 Bank 24 Electrical East Wall Rorth ESA 316" in 3/4" DEKD-40947 EMMI-31636
§-37 to &0 Bank 24 Electrical East Wall North ESA /16" in /4" DEKD-80947 EMNI-S1636
£-37 tp &0 Bank 24 Elertrical East Mall North E5A 346" in 3/4" DKKD-40947 EMMI-51456

p-37 to b0 Bank 24 Electrical East Wall North ESA 316" in 374" DEED-40947 EMMI-T1656

LS
o 60 Bapk
n & Bank

— T T o e

Hultiple

[~

Hultiple
Hultiple
Single

Multiple
Hultiple
Aultiple

3R} T o I X T O

r-J

4 Electrical

4 Electrical

Spare

Spare

Cover BGas

Spare

Spare

Spare

Component Coolant Air
Component Conlant Air
! Component Coolant Air
4 BP Cell

12 Leak Detector

12 Leak Detector

-

d

d

(2

East Hall
Bact Wail
East Hall
Eact Wall
East Wall
East #all
Eact Kall
East Wall
East Wall
East Wail
East KWall
East Wall
East Mall
East MWail

Table 22.

North EGA
North ESA
North ESA
Merth ESA
North ESA
North E54
Horth E54
Morth ESA
Korth ES4
North ESA
North ES3
North ESA
South ESA
South ESA

34167 in 3747 DYKD-40947 EMMI-D1ASH
316" in 374" DEED-40947 EHMI-51656

6"‘
b"
b‘!

in

o
o
:,
b
R
1-1/2" in 4"

(continued)

DEXD-40947

FRED-40947

DEKD-40947 EGGE-41584 3% 1/27IFS
DXED-40747

BEKD-409487

DEED-40747

DEKD-40747 EGBE-31684 DJJA-41877 DIJA-41BBO 3% /4IRS
[XkD-40947 EGGE-41884 DIJA-41579 DJJA-41BRO 3% 3/401P8
DKKD-40347 EGGE-41894 DJIA-41879 LJIAR-41880

DYED-40547 EGGE-1684 DJJA-41879 BJJA-41880 4 1/2°1PS

DKKD-40947 ERED-41863 EBBD-41BAT DJID-354%4 DJJD-55495 12 * 1/4" tubing
DKKD-40947 EBER-41BAT EEBD-41863 DJID-33494 DIJD-55495 12 # 1/4" tubing

8§
59

Banks A~l! to -60, B-1 to -60, C~1 to -36, D-1 to =36, and E-1 to
-36: All are 3/4-in. penetrations of the east wall and are used for
for instruments, thermocouples, and electrical supplies.

Numbers GR:
All are 6—-in. penetrations of the east wall, most with multiple
lines for cover gas, cooling air, and leak detectors; five are
"spares.”

Reprocessing Cell

Numbers 31-54:
Specific penetrations have not been identified, but these functions
are provided:

- material transfer: at least 4 (HF, H,, F,, off-gas);

>
- H-series line heaters: 28; ‘
- equipment heaters: 57;
- thermocouples: 94;
— limit annunciators: 15,

These penetrations are well-engineered and have proven to be
reliable through many years of operation and monitored storage. They,
along with the reactor and drain tank cells, were pressure—~tested to 45
psi (gage) prior to placing the MSRE into hot operation. All external
penetrations are welded into place and were offset or shielded to mini-
mize radiation leakage. The numerous electrical penetrations were sealed
at installation. The transfer lines are, or can be, sealed on both sides
of the cell wall with Gray-loc closures or by welding. There is no cause
for concern for the short term, but these penetrations should be reviewed
from engineering and corrosion aspects before extended storage is adopted
as a planned policy. Some of the transfer lines probably should be
capped, rather than just being valved off. Many of the electrical
penetrations are of no further use and could also be sealed. A more
stringent analysis is appropriate for extended storage. The large cell-
to-cell opening also deserves a stringent analysis for long-term storage
if that is to be pursued. Again, for the short-term, the two—cell system

appears to be more than adequate, but extended storage might benefit from

specific modifications.
60

An engineering study of the penetrations, closure options, and

possible containment envelopes was started late in FY 1985 but could not

be completed with the available funds; the results obtained thus far have

been reported and are included as Appendix D. It is anticipated that

this work will be completed in FY 1986, or as soon as funding is pro-

vided. The important features of this study to date are:

1.

A sequence of three possible containment envelopes was defined, for
future consideration: the drain tank cell; the drain tank plus
reactor cells; and the drain tank, reactor, and reprocessing cells.
Eight types of cell penetrations were identified and catalogued.
Three types of closure options were identified for these eight types
of penetrations: (1) capping/welding of unneeded lines that emerge
from a multiple penetration; (2) total capping of a multiple
penetration where none of the individual lines are needed anymore
and are cut off behind the new seal-plate; and (3) use of a
threaded, pressure-tight plug on electrical lines, as was done on
the original design with spare lines.

Individual penetrations were catalogued in terms of containment
envelope and closure options. The identification of possible future
uses for each of the penetrations was started but not completed.
Finally, it was suggested that a visual inspection be made of all
accessible penetrations to determine the present condition of each.
This will be done as soon as possible because of the obvious impli-

cations if any penetrations were to be found in poor conditiom.

4.3.2 Water Control

4.

If water should enter the cells, it could cause major problems such

increased corrosion rate of the Hastelloy tanks;

slow reaction with the fluoride salts or rapid reaction with any
fluorine (from radiolysis) to yield HF;

radiolytic production of Hj and 05, which is an explosive mixture
that could cause pressure buildup or could increase oxidative corro-
sion; and

movement of radioactive material along transport paths provided by

water.
61

Water entry has never occurred over the past 20 years. Each cell
does have a sump where any water would be collected (if it did enter) and
could then be pumped out or jetted out. The cells are leak—-tested
annually and are known to be leak tight; therefore, water entry is
theoretically not possible, even if the level of the groundwater were to
rise above that of the cell floor. Even so, groundwater is a long—term
concern since the water table is not very far down at the MSRE site.
However, the building was constructed with French drains below the
footers and has a building sump with a minimum elevation of 812 ft. The
lowest cell level is the drain tank cell, at 814 ft for the cell floor,
which is above the building sump level.

Although the building sump pump is known to operate occasionally,
the frequency of operation and the volume of water pumped have not been
recorded or measured previously. As a part of this study, a pump monitor
was recently installed and has been in operation since August 19, 1985.
Readings were taken daily, except on weekends. During the first three
weeks, the west pump operated an average of 1.06 h/d, with a daily high
of 1.5 h and a daily low of 0.6 h. The east pump, a backup system, did
not come on during this period. It is planned to continue gathering
these data as part of our ongoing S&M.

Because of the importance of water control for extended storage, an
engineering overview was prepared. This report is attached in its
entirety as Appendix E. The information contained within it was obtained
from engineering drawings and reports; highlights from the report
include:

l. A description of the radioactive liquid waste system.

2. A description of the nonradiocactive drainage, which includes ground-
water collected by the building French drain system and water
entering floor drains outside the cells, which is then routed to the
building sump cited earlier.

3. A description of the storm water system, which collects water from
roof drains and discharges to a separate catch basin.

4., A description of the sanitary wastewater system, which is discharged

to a septic tank and drain field system.
62

4.,3.3 Secondary Containment

In context of this discussion, secondary containment refers to the
external building structure surrounding the hot cells (i.e., the outer
shell, the building ventilation system, the high-bay area and overhead
crane, and the control rooms). This is the containment that is relied on
during open—cell maintenance or in the event of a release from one of
the cells. Secondary containment will be required and must be fully
functional in case it should become necessary to open one of the cells
for modifications, additions, removal of extraneous equipment, or treat-
ment of the salts. Secondary containment is not required during standby
mode, which is the usual condition of the facility.

The present status of the secondary containment system is functional.
The high-bay area can be closed and ventilated through the cell ventilation
system, while maintaining a negative preséure inside the containment zone.
The overhead crane is still operational. The remote maintenance control
room has deteriorated somewhat, and the zinc bromide viewing windows have
been drained (but are still usable). The main control room contains most
of the instrumentation being used for monitoring and surveillance.

The useful lifetime of the secondary containment system is definitely
limited in the absence of a planned program to maintain this capability.
Any operations that may need to be done prior to committing to extended
storage should be carried out in the reasonably near future. In doing
so, it would be possible to still utilize experienced personnel who have
a firsthand knowledge of the facility from prior experience when the MSRE

was an active project.

4.4 GEOLOGY AND HYDROLOGY

A geologic and hydrologic overview of the MSRE site was prepared by
a consultant in this field. A copy of his report is included as Appendix
F. The more significant of his findings can be summarized as follows:

1. The general geologic conditions at the MSRE site are favorable for
continued storage. The area has a low seismic risk and has been
relatively stable for millions of years. Other waste storage facili-
ties (solid LLW, TRU waste, hydrofracture) have been sited in
Melton Valley because of its suitability for such applications.
63

2, Water is a significant concern because the water table is not far
below the surface and because water is a potential problem for two
reasons: it can hasten corrosion, and it can provide a pathway or
mechanism for movement of any released radioactivity.

In planning for the future of MSRE and evaluating various options,
the time scale must be considered since each time frame has its own
characteristics. The various time frames are defined in Sect. 1. For
purposes of this evaluation, two key characteristics are: the types of
physical control that can be applied, and the nature of the radiocactive
decay which is occurring.

Seismic factors are of minor concern to us through the near term
(up to 100 years) since the reinforced concrete cells and heavy-walled
Hastelloy tanks will provide ample strength during the early years.
Water intrusion, however, must be cdnsidered even for the short term.
Data are needed on water behavior in the immediate environment. As
listed in the geologist's analysis, data are needed on the following:
(1) characterization of the earthen materials at or near the site; (2)
the site design itself, including all engineered improvements; and (3)

site hydrology.

4.5 ENTOMBMENT

Entombment in cementitious material has been surveyed by L. R. Dole
of this Laboratory, whose complete report was given as Appendix E in the
evaluation study that preceded this report.29 His survey identifies
several aspects of this concept that could be detrimental, thus indi-
cating a need for caution since entombment is nonreversible for all prac-—
tical purposes. It also points out the potentially harmful effects of
water, which suggest restraint at present and also support the need for
more data in this area (as noted in the previous section).

While entombment is very attractive and might, in fact, become the
option of choice at a future time, it is premature at present because we
are not in a position to make a permanent commitmént. This will continue
to be the case until we have definitive data concerning the radiolysis of
fluoride. The negative aspects of entombment include the preclusion of
direct surveillance, the possible net weakening of the overall structure

from shrinkage or expansion mismatch, and the introduction of water from
64

the grout mixture itself. In addition, the solidified grout could act as
a wick to transport moisture to metal surfaces and thereby increase
corrosion. It also fills up void space that might better be used as a
sink for intruded water or as a reservoir for a desiccant (e.g., unslaked
lime or Portland cement).

On the other hand, a properly engineered grout could add strength
and provide a diffusion barrier against the movement of water either in
or out of the cell. The projected lifetime of a grout can be measured in
millennia in a dry climate. A more-detailed survey is required to iden-
tify any areas where specific data need to be gathered, or generated, in
order to have the necessary technical basis for a future decision
regarding entombment.

Radiolysis of the water contained in grout is also a possible
drawback. The resulting oxygen and hydrogén could lead to excessive
pressure or an increased corrosion rate; however, inhibitors or catalytic
recombiners could be used to mitigate this effect.

The use of other entombment materials, such as lead or sulfur, has

been suggested, but these materials have not been examined in detail.

4.6 UTILIZATION

The MSRE building facilities are being used extensively, par-—
ticularly the office areas and the storage space in the high-bay and
receiving areas. Approximately 50 people from the Health and Safety
Research Division are housed in the office complex, and this number is
expected to grow. The history of the Laboratory clearly shows an ever
increasing need for office space. Therefore, the office facilities
should be retained in any event.

The need for storage space for low-level radioactive samples and
materials has also continued to grow. At present, the building provides
space for soil samples (about 100 55-gal drums) and Cs-137 sources (about
260 in 200-1b lead pigs), and miscellaneous equipment (e.g., about 20
remote manipulators). These storage facilities definitely need to be
retained. Even a conservative estimate suggests an additional 20-year
lifetime for the offices and the storage space. This is commensurate
with the anticipated short-term extended storage of the fuel and flush

salts.
65

Looking beyond utilization already extant, the hot cells themselves
are a valuable asset whose future utility should be preserved. Because
of activation, the reactor, drain, and processing cells are likely can-
didates for entombment unless a suitable major project comes along. Even
with entombment of these three cells, however, they could first be filled
with waste removed from other cells. The remaining cells could easily be

used for storage purposes or, with modification, for a suitable project.

4.7 FINAL DISPOSAL

The long—~term alpha activity of the U-233 decay chain may require
eventual removal of the fuel and flush salts, even if the radiolysis
problem can be controlled with getters. In the latter case, it would be
possible — and highly desirable — to leave the salts in their present
Hastelloy N tanks since these tanks should serve as excellent primary
containers. There are three possible options for permanent disposal of
the tanks and contents:

1. the WIPP (Waste Isolation Pilot Plant), which is now under construc-
tion in New Mexico;
2, a spent fuel repository, which is planned for a western location; or
3. on-site greater confinement disposal (intermediate—depth burial),
which has been considered for 0Oak Ridge.
These three options are discussed in the following sections. In the
event that the radiolysis problem cannot be controlled with getters, thus
making reprocessing necessary, the separated uranium and TRU waste would
still be candidates for the same three final options. However, in this
case, with the fluoride removed and the residue repackaged, acceptance
criteria should be easier to meet. It should be recognized that, ini-
tially, all disposal sites will be conservative in specifying their
acceptance criteria but that, with the passage of time and accumulation
of experience, the potential for some relaxation in these criteria is
very real, especially for a one~time interment that could be given addi-

tional overpacks.
66

4.7.1 WIPP

Since the WIPP is actually under construction and has been planned
for many years, the waste acceptance criteria (WAC) for WIPP have been
the subject of extensive study, review and comment, and revisions. Even
so, these WAC are not yet final and, as already pointed out, tend to be
conservative. Appendix C summarizes the WAC for remotely handled (RH)
waste, the Certification Compliance Requirements (CCR), and some general
comments. The WAC and CCR are not always in agreement. Those require-
ments that concern the MSRE fuel salt are briefly discussed below.

Size limit: Diameter of tanks too large (50 vs 26 in.) but

length satisfactory (86 vs 121 in.)

Weight: Each tank (~1300 kg) plus salt (~2300 kg) is just
within the limit (3636 kg); addition of getter
would make them overweight.

Surface dose: Surface dose probably greater than the 100-rem/h
limit; also, neutron dose much higher than 75
mrem/h, as limited by the CCR (although the WAC
requires only that neutron doses >75 mrem/h be
reported).

Thermal power: Well below the limit of 300 W/package; drain tanks
are about 100 W each.

Criticality: About 10 times the limit of 1.9 g/L; actually,
about 17 g/L.

Pu-239

equivalent: Probably about the limit of 1000 plutonium-

equivalent Ci/package.
4,7.2 Commercial Spent Fuel Repository

This offers some attractive features., Certainly, the MSRE fuel and
flush salts qualify in a generic sense. The MSRE D&D is funded through
the commercial branch of the SFMP since the MSRE was intended for com—
mercial application. The fuel salt is a spent fuel, and the flush salt
contains some spent fuel. Most of the WIPP restriction would not be
applicable to a spent fuel repository, which would be designed to handle
higher doses, thermal power, criticality, and Pu=-239 equivalent. One
possible problem is the tank diameter. A repository might not be

designed for a package diameter of over 4 ft, even on an exception basis.
67

4.,7.3 Greater Confinement Disposal

Every major DOE site (Oak Ridge included) has potential problems
with oversized, overweight, or otherwise out-of-specification TRU waste,
for which intermediate~depth disposal is being considered. The MSRE fuel
salt 1s an obvious candidate for such disposal since the large tank
diameter would not be a problem in this situation. Shipment would also
be minimized if an Oak Ridge site is developed in the future.

4,8 IN-CELL ALTERATIONS

A number of in-cell alterations might be performed in the short term
(within 10 to 20 years) to enhance the condition of the fuel salts during
extended storage and also to help set the stage for eventual removal. As
a step toward the eventual removal of the salts, it would be highly
desirable to place the cell and the salts into a suitable condition
during the short term so that the removal operation can be carried out as
simply and easily as possible. The following possibilities are presented
and then evaluated in terms of both enhanced storage and eventual
removal:

l. removal of steam domes (and addition of neutron poison);

2. addition of fluorine getter, without remelting;

3. remelting of salt, with addition of fluorine getter;

4. repackaging of salt into smaller containers, after remelting with
addition of fluorine getter;

5. disconnection of lines to and from the drain tanks; and

6. installation of additional instrumentation.

It is assumed that the fuel salts will eventually have to be removed
to a more permanent location because of the long—-term activity from the
U-233 chain. It is also assumed that it will be feasible to do so with
the salts packaged in their present containers. With these assumptions,
it is clear that the steam domes must be removed eventually; therefore,
this should be done in the short term while experienced personnel are
still available. Once the steam domes with attached bayonet tubes are
removed, the thimbles (32 per tank) can be used to insert neutron
poisons. These can be in the form of Cd, B, or Ge metal rods.

The earlier discussion on radiolysis is a strong argument for the

addition of a fluorine getter. In fact, if an effective getter is not
68

added, disposal of the fuel salt as the fluoride will probably not be
acceptable. The simplest way to include getter is to add it to the void
volume in each tank. This might require cell entry; of course, the pre-
ferred method would be to supply getter through the addition tube used
previously for adding salts. 1If cold laboratory tests show that the
getter must be dispersed within the solid salt to be effective, then
remelting would be required. While this operation is possible, it might
cause some further corrosion to the tanks. Dispersal of added getter
could be accomplished via sparging with dry helium through the dip tubes
already provided for this purpose.

In the event that the salt needs to be melted in order to add
getter, this would be the logical time to repackage into smaller con-
tainers (if this step is needed). However, the incentive to repackage is
small since the controlling factor is radiolysis and, if radiolysis can
be controlled by the addition of getter, this can be done just as well in
the present drain tanks.

The lines for the vessel should not be disconnected until there is
assurance that the salt need not be repackaged or reprocessed. Such
assurance cannot be provided until actual data are in hand to verify the
addition of getter as a viable method for limiting fluorine formation.

This may require the installation of additiomnal instrumentation.

4.9 CONTINGENCY PLANNING

The MSRE has had an exceptionally clean history following shutdown
of the reactor in December 1969. Because of this long period of trouble-
free surveillance and maintenance, formal contingency planning and/or
emergency planning has received virtually no attention. However, since
it is now becoming clear that extended storage in the present location
will be required, such planning is appropriate in the immediate future.
Obviously, no major contingencies are expected; nor are they likely to
occur., On the other hand, it is prudent to carry out a structured analy-
sis in order to systematically investigate the possibilities, to identify
any weaknesses that could or should be remedied in advance, and to be
prepared in case the unexpected happens. A preliminary list of possible

events includes, but is not necessarily limited to, the following items:
5.1

71

5. CONCLUSIONS

EVALUATION OF OPTIONS

Evaluation of the options, data, and constraints discussed in the

body and appendixes of this report led to the following conclusions:

1.

2.

3.

Extended storage is not only feasible but preferable at this time

because

Ae

b.

the fuel salt is presently in a safe place;

there is no available repository or disposal site to take it if it
were removed, either with or without reprocessing, at this time;
the full range of future disposal options or ldcations is not yet
known ;

any enhancement actions taken at this time would probably benefit
whichever option is finally exercised; and

extended storage would provide time to collect additional data
that would be of direct help in selecting a final option at a

future date.

Entombment should be deferred because:

de

b.

such action is not necessary at this time;

it is essentially nonreversible and would be premature at this
time; and

additional technical data are needed before such a decision could

be supported.

Some enhancements in direct support of extended storage should be

evaluated in the near future. For example:

Ae

cell penetrations should be reviewed by Energy Systems
Engineering, inspected by the Operations Division and, where con-
sidered necessary, sealed permanently;

cell internals should be reviewed to identify interfering struc—
tures, if any, and plans made for their removal;

remelting of the fuel salt, with possible addition of a fluorine
getter (and possibly a neutron poison) should be evaluated for
implementation since this would eliminate both the need for annual

reheating and long—-term concerns about radiolysis-produced fluorine;
4.

d.

72

repackaging (if needed, it should be done in conjunction with
remelting) should also be evaluated;

enhancements seem prudent since the alpha activity (and the
resulting neutrons from a,n reactions) will decline only slightly
in future years, even though the gamma activity will drop signifi-
cantly;

these enhancements would not interfere with the exercise of
options at a later date and would, in the interim, improve on the
already sound and secure condition of the facility; and

these enhancements, while not necessary for continued short-term
storage, should be made prior to initiating extended storage while
the assured capability to do so still exists, in terms of both
people with firsthand knowledge and equipment that is still func-

tional.

The following pertinent data (or studies) should be obtained (or

made) over the next few years:

Ae

b.

Coe

observations of groundwater levels around the site;

possible improvements of a civil engineering nature, to lower the
groundwater level;

Hastelloy N corrosion behavior in the presence of moisture plus
alkali, fluoride, or both;

suitability of candidate materials for use as a fluorine getter,
in—cell desiccant, tailored grout, and neutron poison; and
contingency plans developed, along with a review and update of S&M

procedures.

Beneficial utilization of the facility should be encouraged and

aided for the following reasons:

A

b.

to help fill directly our own facility requirements;

office space at the site is fully utilized today, and this need is
expected to grow;

the high-bay area is being used to store bulky, mildly radioactive
samples,and this need is expected to continue;

the three more active hot cells (reactor, drain tank, and repro-
cessing) could be used for storage of other radiocactive wastes;
the remaining cells could be used for hot operations or radio-

active waste storage; and
69

(1) ventilation failure; (2) extended power failure; (3) water ingress
from leakage; (4) electric heater failure; (5) thermocouple failure;
(6) overheating during annual reheat; (7) loss of instrument air; (8) a
catastrophic event such as a tornado, earthquake, plane crash, or flood;
(9) sabotage or terrorist action; (1Q0) possible radiation exposure via
plume or ingestion; and (l1) possible radiation exposure to site opera-—
tors. In addition to the above incidents, a contingency analysis should
also cover three broad categories: (1) information sources, (2) iden-—
tification of equipment, and (3) physical condition of components.

In the event of a contingency, needed information must be easily and

T

quickly available. A concise "contingency plan,” with readily available
backup information in various specific areas, is probably appropriate.
In addition, easy and positive identification of components, controls,
etc., may be essential. The recently completed color coding scheme is a
significant step in this direction. The coding employed is pink for
routine maintenance and surveillance, whether daily, monthly, or annual,
and yellow for possible future use in remelting, transferring, etc. (If
these colors seem unusual, they were chosen to match the markup of
drawings used to do this work.)

In the event of a serious contingency, it might be of considerable
benefit if the frozen fuel were in a condition whereby it could either be
sealed off and left in place for a long time or, alternatively,
transported elsewhere. These factors would, of course, be enhanced if
the fuel had previously been remelted and getters added, or remelted and
repackaged, or if interfering structures (e.g., steam domes) had been
removed. '

At this time, work is scheduled for the next fiscal year on a con-

tingency plan, along with a review and update of S&M procedures.
f.

73

the Li-7 content of the cooling salt is nonradiocactive and may

have enough value to justify recovery.

5.2 SELECTION OF PLAN

Based on the above considerations, an outlined plan and approxi-

mate schedule are proposed:

1.

2.

3.

4e

d.

b.

d.

1986 — complete (or expand on) work in progress:

penetrations — complete the engineering evaluation and provide
recommendations;

groundwater — continue data collection on radioactivity and sump
pump operation; complete an on-site survey and inspection of
drainage system and components;

radiolysis — conduct a peer review of the overall problem and
potential solution; prepare a technical plan for pressure or
fluorine "sniffer” measurements on the two drain tanks; make a
decision if reheating is to be discontinued (in order to obtain
direct data on radiolysis); and conduct preliminary laboratory-
scale tests with potential fluorine getters;

review and update the S&M procedures and prepare a contingency

plan.

1987 — upgrade facility and prepare for new work:

ae
b.

Coe

seal or improve cell penetrations, as appropriate;
install fluorine radiolysis test equipment; and
prepare a technical plan for removal of steam dome and other

internals in order to evaluate the work involved.

1988-95 — collect data:

ade

b.

fluorine radiolysis in the drain taunks; and

laboratory—scale tests on fluorine getters.,

1996 — Make the following decisions:

a.
b.
Ce
d.

€.

to remelt and add getter, or not;

to remove steam dome and internals, or not;

to reprocess, or not;

to repackage, or not; and

selection of final disposal site, based on information available

at that time.
75

Appendix A. BIBLIOGRAPHY AND REFERENCES
76

Appendix A. BIBLIOGRAPHY AND REFERENCES

The MSRE Project is extensively and thoroughly documented by
reports, drawings, photographs, and miscellaneous documents. A com—
puterized listing of these has been prepared by Park Owen and Nancy Knox,
of the ORNL Remedial Action Program Information Center. There are
approximately 1450 entries in this file, categorized as follows:

Type of entry No. of entries
Progress report ~140

Technical report

 

CF memo ~370
TM report ~240
ORNL report V70
Miscellaneous 14
Journal article 32
Drawing 19
Photograph ~500
Patent 4
Conference paper §50
Correspondence 11
Total ~1450

Copies of most of these documents are in the MSRE file, which is
stored in the basement of Building 7503. The Information Center also
has copies of many of the reports. These documents are a merger of
files previously maintained by individuals who were closely associated
with the project. (Unfortunately, several extensive files have been lost
over the years.) Most of the photographs were supplied by Luther Pugh.
There is also an excellent collection of construction drawings main-
tained by Martin Marietta Energy Systems, Inc., Engineering on file in
Building 1000.

A relatively small number of reports have provided all the
background information used for this study. A listing, in chronolo-
gical order, is provided on the following pages. The chronological num-

bers are used as reference call-outs in the main body of the report.
1.

2.

3.

7

8.

9.

10.

11.

12,

13.

14.

15.

77

Ruth Slusher, H. F. McDuffie, and W. L. Marshall, Some Chemical
Aspects of Molten-Salt Reactor Safety: (1) Dissolution of Coolant
and Fuel Mixtures in H50, (2) A Portion of the System LiF-BeF,-H,0
at 25, 60 and Near 100°C, ORNL-TM-458, December 14, 1962.

P. N. Haubenreich, Inherent Neutron Sources in Clean MSRE Fuel Salt,
ORNL-TM-611, August 27, 1963.

J. R. Engel and B. E. Prince, Criticality Factors in MSRE Fuel
Storage and Drain Tanks, ORNL-TM-759, September 14, 1964.

R. C. Robertson, MSRE Design and Operations Report — Part I —
Description of Reactor Design, ORNL-TM-728, January 1965.

R. B. Lindauver, MSRE Design and Operations Report - Part VII -

Fuel Handling and Reprocessing Plant, ORNL-TM-907R, December 28,
1967.

R. B. Lindauer, Processing of the MSRE Flush and Fuel Salts,
ORNL-TM-2578, August 1969.

R. C. Steffy, Jr., Inherent Neutron Source in MSRE with Clean U-233
Fuel, ORNL-TM-2685, August 10, 1969.

We R. Grimes, "Molten—-Salt Reactor Chemistry”, Nuclear Applications &
Technology 8, 137, February 1970.

M. J. Bell, Calculated Radioactivity of MSRE Fuel Salt, ORNL-TM-2970,
May 1970.

P. N. Haubenreich, Fluorine Production and Recombination in Frozen
MSR Salts after Reactor Operation, ORNL-TM-3144, September 30, 1970.
R. H. Guymon, MSRE Procedures for the Period Between Examination and
Ultimate Disposal, ORNL-TM-3253, February 10, 19/1.

 

Roy E. Thoma, Chemical Aspects of MSRE Operations, ORNL-4658,
December 1971.

P. N. Haubenreich and R. B. Lindauer, Consideration of Possible
Methods of Disposal of MSRE Salts, ORNL/CF-72-1~1, January 28, 1972.
E. L. Compere et al., Fission Product Behavior in the Molten Salt
Reactor Experiment, ORNL-4865, October 1975.

C. D. Cagle and L. P. Pugh, Decommissioning Study for the ORNL
Molten-Salt Reactor Experiment (MSRE), ORNL/CF-77/391, August 25,
1977.

 
16.

17.

18.

19.

20,

78

EBASCO Services, Inc., Technical Report - Feasibility Study:
Disposal of MSRE Fuel and Flush Salts, Letter Report (unnumber),
October 1980.

D. R. Simpson, Preliminary Radiological Characterization of the
Molten Salt Reactor Experiment (MSRE), ORNL/CF-84/92, September 20,
1984.

F. J. Peretz, Preliminary Decommissioning Study Reports — Volume 5:
Molten Salt Reactor Experiment, X-0E-231, Vol. 5, September 1984.
T. E. Myrick, The ORNL Surplus Facilities Management Program Long-
Range Plan, ORNL/TM-8957, September 1984.

K. J. Notz, Feasibility Evaluation of Extended Storage-—-in-Place of
MSRE Fuel Salt and Flush Salt, ORNL/CF-85/71, March 1985.
79

Appendix B. SEMIQUANTITATIVE EVALUATION OF SIX MSRE OPTIONS

 

Table No. Evaluation Basis
B.1 Process Readiness
B.2 Economics
B.3 Short-Term Hazard
B.4 Long~Term Hazard

B.5 Conservation
80

Table B.l. Semiquantitative evaluation of MSRE options:
A. Process readiness?®

 

Option

 

Process step 1 2 3 4 5 6

 

Melt salts (and add getter)
Remove salts

Build salt process facility?
Process or convert salts/U2
Package salts (and U)
Dismantle/dispose of facility

Transport packaged products
b

Loow s N W
|
i
i
N

Store/isolate products

Decontaminate cells/equipment

Dispose of liquid LLW

Remove equipment
Dispose of solid LIW
Dismantle cell structure

Dispose of solid LLW/rubble

R W U = W= W W NN N W

Restore area

Seal pipes/penetrations

Internal entombment

i
v e W
v o Wb

|

i

I

External structures
Stabilize drain-tank cell - - - 5 5

Continue surveillance - - - 5 5 5

Total 47 41 15 15 19 23

 

aRange: 0 to 5. (A low score is superior.) The rating considers
the cost of developing the process, as well as the uncertainty of success-
ful development. The time delay involved is not a part of the rating
since ample time is presumably available.

bNo process flowsheet has been defined at this time.

CNo facility which will accept the packaged products has been iden-
tified at this time.
81

Table B.2. Semiquantitative evaluation of MSRE options:

B.

Economics@

 

Process step

Option

 

3

 

Melt salts (and add getters)
Remove salts

Build salt process facility
Process or convert salts/U
Package salts (and U)
Dismantle/dispose of facility
Transport packaged products
Store/isolate products
Decontaminate cells/equipment
Dispose of liquid LLW

Remove equipment

Dispose of solid LIW
Dismantle cell structure
Dispose of solid LLW/rubble
Restore area

Seal pipes/penetrations
Internal entombment

External structures

Stabilize drain-tank cell

Continue surveillance

Total

54

Ww =~ U = P =~ U1 NN W

L o W

47

15

15 20 238

 

4Range: 0 to 5. (A low score is superior.)

relative cost of each process.

The rating considers the
82

Table B.3. Semiquantitative evaluation of MSRE options:
C. Short-Term hazard?

 

Option

 

Process step 1 2 3 4 5 6

 

Melt salts (and add getters) 3 3 - - 5

w
|
i
(

Remove salts

v
A%
|
1
|
i

Build salt process facility

ot
o

Process or convert salts/U 10 - - - -
Package salts (and U)
Dismantle/dispose of facility

Transport packaged products

e NN
|
)
!
i

Store/isolate products
Decontaminate cells/equipment
Dispose of liquid LIW

Remove equipment

Dispose of solid LLIW
Dismantle cell structure

Dispose of solid LLW/rubble

— = N = 0~ 0 W N~ U
|
|
1
!
1

Restore area

Seal pipes/penetrations

Internal entombment

|
— BN
w & Wun

1

|

1

External structures
Stabilize drain-tank cell - - - 8 8 8

Continue surveillance - - -

Total 66 47 14 16 21 31

 

3Range: O to 10. (A low score is superior). The rating considers
the expected operator exposure, public exposure, probability of an adverse
release, and possibility of sabotage on-site, during transport, or in the
final location. Short term means less than 100 years, nominally 5 to 20
years.
83

Table B.4. Semiquantitative evaluation of MSRE options:
D. Long-Term hazard?®

 

 

 

Option

Process step 1 2 3 4b 5 6

Melt salts (and add getters) - - - -~ - -
Remove salts - - - - - -
Build salt process facility - ~ - - - -
Process or convert salts/U - - - - - -
Package salts (and U) - - - - - -
Dismantle/dispose of facility - - - - - -
Transport packaged products - - - - - -
Store/isolate products 1 1 - - - -
Decontaminate cells/equipment - - - - - -
Dispose of liquid LLW 1 - - - - -
Remove equipment - - - - - -
Dispose of solid LIW 1 - - - - -
Dismantle cell structure - - - - - -
Dispose of solid LLW/rubble 1 - - - - -
Restore area - - - - - -
Seal pipes/penetrations - 2 5 5 5 5
Internal entombment - 2 5 - - -
External structures - 1 2 1 1 1
Stabilize drain~tank cell - - - 10 5 3
Continue surveillance - - - 1 1 1
Total 4 6 12 17 12 10

 

aRange: 0 to 10. (A low score is superior.) The rating considers
the expected public exposure, probability of an adverse release, and
possibility of inadvertent intrusion. Long term means greater than 100
years, nominally 1000 years or longer.

bnot really a viable long-term option; would eventually have to be
supplanted by one of the other options.
84

Table B.5. Semiquantitative evaluation of MSRE options:
E - Conservation®

 

Option

 

Process step 1 2 3 4 5

 

Melt salts (and add getters) - - - - -
Remove salts - - - - -
Build salt process facility - - - - -
Process or convert salts/U - - - - -
Package salts (and U) - - - - -
Dismantle/dispose of facility - - - - -
Transport packaged products - - - - -
Store/isolate products 1 1 - - -
Decontaminate cells/equipment - - - ~ -
Dispose of liquid LIW 1 - - - -
Remove equipment - - - - -
Dispose of solid LIW 1 - - - -
Dismantle cell structure - ~ - - -
Dispose of solid LIW/rubble 1 - - - -
Restore area (land) - 5 5 5 5
Seal pipes/penetrations - - - - -
Internal entombment - - - - -
External structures (buildings) 10 - 1 1 1
Stabilize drain-tank cell - - - - -

Continue surveillance - - - - -

Total 14 6 6 6 6

 

aRange 0 to 10. (A low score is superior.) The rating considers
near—term utilization of the buildings and long-term utilization of the
land.
85

Appendix C. WIPP WASTE ACCEPTANCE CRITERIA

C.l. WIPP WASTE ACCEPTANCE CRITERIA FOR REMOTELY HANDLED (RH) TRU WASTE

C+.2. WIPP CERTIFICATION COMPLIANCE REQUIREMENTS

C.3. WIPP: GENERAL COMMENTS
86

C.l. WIPP WASTE ACCEPTANCE CRITERIA FOR RH TRU WASTE
Source: WIPP-DOE-069, Rev. 2, Draft C, dated August 1984.

Combustibility: Any combustible TRU waste shall be packaged in a

noncombustible container.

Immobilization: Required if more than 1 wt % is in form of particles <10
microns, or if more than 15 wt % is <200 microns in diameter, Par-

ticulate desiccants (e.g., Portland cement) are exempted.

Sludges: Shall be packaged such that internal corrosion of the container

does not occur.

Liquid Waste: Free liquids are not allowed; minor amounts in cans or

bottles are acceptable,

Explosives and Compressed Gases: TRU waste shall contain no explosives

or compressed gases, as defined by 49 CFR 173, subparts C and G.

Pyrophoric Material: No more than 1 wt % may be pyrophoric forms of
radionuclide metals, and these shall be generally dispersed in the
waste. Pyrophoric materials other than radionuclides shall be ren-

dered safe by mixing with stable materials (e.g., glass or concrete).

Radioactive Mixed Waste: No hazardous waste is permitted unless it is
also co-contaminated with TRU waste. Reactive material shall be
identified. Corrosive materials must be neutralized, rendered non-—
corrosive, or packaged in a manner to ensure container adequacy

through the design lifetime.

Container: May be the RH container itself, or it may be the container

inside an overpack. It must meet design conditions for "Type A"
87

packaging, 49 CFR 173.412(b), specified by DOT. Container (and
labeling) shall be certified for a design life of at least 20 years

from date of certification.

Package Size and Handling: Maximum size 26" 0.D. (0.66 m) and 121" long
(3.1 m) including the pintle. Must have an axial pintle, of a
design acceptable to WIPP, and no other lifting devices.

Waste Package Weight: Shall not exceed 8000 lbs (3636 kg).

Surface Dose Rate: Not greater than 100 Rem/hr at any point. Neutron

 

contributions greater than 75 mRem/hr shall be reported in the data

package.

Surface Contamination: Smearable contamination no greater than 50 x 10712

 

Curies per 100 cn? for alpha and 450 x 10712 per 100 cm? for beta-

gamma.
Thermal Power: Not greater than 300 watts per package.

Nuclear Criticality: Shall not exceed 1.9 g/liter of fissionable isotopes

 

(averaged over 5 liters with a maximum 50% void space). If such
reasonable distribution cannot be assured, then the canister is
limited to 240 g total (in Pu—-239 fissile gram equivalents). The
canister may be loaded with DOT 17C drums which will provide inter-
nal partitioning and increase (sic) the limits to 100 g for each 30

gal drum and 200 g for each 55 gal drum.

Pu-239 Equivalent Activity: Packages shall contain no more than 1000 PE-

Ci (Plutonium -~ Equivalent curies).

Labeling: Each package shall be uniquely identified with a permanently

attached number at least 2" high.
88

Data Package: Certified data provided prior to shipment shall include:

Package identification number;

Dated certification statement that waste content and packaging are
in accord with the WIPP WAC, and the waste is unclassified;

Waste generation site;

Date of packaging;

Maximum Surface Dose Rate;

Weight;

Container type;

Physical description of waste form;

Assay information, including PE-Ci and Pu-239 fissile gram equivalent
contents;

Hazardous materials (non~radionuclide) content: identification and
quantity;

Measured or calculated thermal power;

Date of shipment;

Carrier identification;

Other information considered significant by the shipper.
C+.2. WIPP CERTIFICATION COMPLIANCE REQUIREMENTS

Source: "Draft A" of WIPP-DOE-158 dated August 1984: For unclassified, RH,
TRU waste, both newly-generated and retrieved-from—storage; all other

applicable DOE orders must continue to be met.

Container: May be the RH container itself, or it may be the container
inside an overpack. It must meet design conditions for "Type A"
packaging, 49 CFR 173.412(b), specified by DOT. The container (and
labeling) shall be certified for a design life of at least 20 years

from date of certification.

Package Size and Handling: Maximum size 26" 0.D. (0.66 m) and 121" long
(3.1 m) including the pintle. Must have an axial pintle, of a

design acceptable to WIPP, and no other lifting devices.
89

Immobilization: Required if more than 1 wt % is in form of particles <10
microns, or if more than 15 wt % is <200 microns in diameter. Par-

ticulate desiccants (e.g., Portland cement) are exempted.

Liquid Waste: Free liquids are not allowed; minor amounts in cans or

bottles are acceptable.

Sludges: Shall be packaged such that internal corrosion of the container

does not occur.

Pyrophoric Material: No more than 1 wt % may be pyrophoric forms of
radionuclide metals, and these shall be generally dispersed in the
waste. Pyrophoric materials other than radionuclides shall be ren-—
dered safe by mixing with stable materials (e.g., glass or

concrete) .

Explosives and Compressed Gas: Not greater than trace quantities of
explosive compounds or no vessels capable of being pressurized

greater than 7 psig.

Radioactive Mixed Waste: No hazardous waste is permitted unless it is
also co-contaminated with TRU waste. Reactive material shall be
identified. Corrosive materials must be neutralized, rendered non-
corrosive, or packaged in a manner to ensure container adequacy

through the design lifetime.
Waste Package Weight: Shall not exceed 8000 lbs (3636 kg).

Nuclear Criticality: Fissile content no more than 50 g/ft3, averaged
over 5 ft3 volume and 50% void space. If uniform distribution can-
not be assured, the container is limited to 240 g total (in Pu-239

fissile equivalents).
90

Pu-239 Equivalent TRU Activity: Shall not exceed a value (in Curies) TBD.

 

The Pu-239 equivalent is based on MPC values, DOE Order 5480.1A,
Chapter XI.

Surface Dose Rate: Not greater than 100 Rem/hr at any point. Neutron

 

levels shall be limited to 75 mRem/hr at the container surface(!).

Surface Contamination: Smearable contamination no greater than 50 X 10712

 

Curies per 100 cm? for alpha and 450 x 10712 per 100 cm? for beta-

gamma.
Thermal Power: Not greater than 300 watts per package.

Combustibility: Any combustible TRU waste shall be packaged in a noncom-

bustible container.

Gas Generation: All RH TRU waste canisters shall be vented, through a

Rocky Flats—type carbon filter or equivalent.
Color Coding: No criterion specified.

Labeling: Each package shall be uniquely identified with a permanently

attached number at least 2" high.

Documentation/Compliance: The shipper shall have operating procedures

 

with methods for determination of all required data. The data
package shall be transmitted to WIPP prior to shipment, in a format
acceptable to WIPP., A certificate of compliance shall be provided
by a designated individual from the certifying facility_and be main-

tained by the shipper.
C.3. WIPP: GENERAL COMMENTS

Source: (Excerpted from WIPP-DOE-069)
91

The facility is not yet authorized to be permanent. Therefore, retrieval

must be allowed for up to 20 years under present criteria.

Some experimental packages will be allowed. Out of 170 acres, 10 acres

is for experimental purposes.

RH TRU has a surface dose rate >200 mRem/hr, but may not be greater than

Pu-239 Fissile Gram Equivalent is based on Ky ff, assuming an optimally

moderated infinite array.

Pu-239 Equivalent Activity (expressed in PE-Ci) is characterized by:

k

i=1

where there are k isotopes, A; is maximum activity of isotope i, CFy
is the MPC correction factor for isotope i, obtained by multiplying
the MPC in 10 CFR 20, Appendix B, Table 1, column 1 for soluble
materials by 5 x 101! ml/Ci, to normalize relative to Pu-239.

TRU Waste is defined as defense waste contaminated with certain alpha-
emitting isotopes of atomic number greater than 92 and half-lives

greater than 20 yrs, in concentrations greater than 100 nCi/g.

Waste Container is the disposable containment intended for emplacement at
WIPP, including any integral liner or shielding. Package is the

container and contents.

Waste volume percent is the material volume, excluding trapped void
space, of that form compared to the total waste volume, but not com—

pared to the package volume.
93

Appendix D. ANALYSIS OF MSRE CELL PENETRATIONS

Prepared by

D. Macdonald and A. C. Williamson

Martin Marietta Energy Systems, Inc., Engineering

August 21, 1985
94

D.1 INTRODUCTION

The objective of this study is to review and comment on one aspect
of the integrity of the MSRE facility to continue to safely store the
fuel and flush salts presently located in the drain tank cell of
Building 7503. This study is concerned with penetrations through the
cell walls. These penetrations were evaluated in the context of
overlapping containment eanvelopes composed of various cell groupings.
The more than 100 penetrations into the reactor, drain tank, and fuel
processing cells (Fig. D.1) associated with each proposed envelope are
described by location, type, current condition, and potential for future
use. The identification numbers shown on Fig. D.l indicate cell
penetrations and refer to identification numbers shown on facility
drawings; they are also referenced tb Table D.1 of this study. Possible
closure options are proposed for those penetrations not needed for faci-
lity maintenance or for future fuel transfer. These evaluations of the
condition of the penetrations are incomplete at present and warrant
further study to establish the requirements for any recommended course
of action with regard to long-range plans for safe storage of the fuel

and eventual final decommissioning of the facility.

D.2 CONTAINMENT ENVELOPES

While the MSRE was in operation, the fuel salt was circulated in
the reactor, drain tank, and fuel processing cells. The fuel and flush
salts have been stored in three critically safe storage tanks in the
drain tank cell following facility shutdown in 1969. Since the three
cells are interconnected to varying degrees, this study was organized in
terms of three containment envelopes (Fig. D.2). Envelope 1 is composed
of the reactor, drain tank, and fuel processing cells. These cells are
grouped together because they contained most of the radioactivity while
the MSRE was operating. Envelope 2 is made up of the reactor and drain
tank cells, which are still connected via an open penetration (the
penetration to the processing cell is sealed). Also, the processing

cell is more accessible and more likely to be utilized in the future
REF. DWG. X3E-12598-001

 

v

P PPV o a0 Ty pe AP @0, K P oM gyt

 

1-V

v &

 

 

y

Fv v a

o

|

-
T S
ov.n by

&N

P

 

  

37-42 A

 

  

REACTOR CELL

 

 

 

 

    

FUEL_PROCESSING |

I CELL
Brib-’////L —

DRAIN TANY CELL

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

43-53

 

I-18

Fig. D-1. Identification and location of cell penetrations at the MSRE.

<6
96

 

REF. DWG. X3E-12598-001

 

 

 

 

 

 

 

 

N

 

 

 

 

ENvELOPE T

 

 

1|

 

|

 

N\=r7

 

 

 

 

 

EnveLore 11

 

 

 

 

N

 

EnveLore 1L

 

 

Fig. D-2. Containment envelope options.

 
97

than is the reactor cell. Once the processing cell is accessed, the
penetrations associated with Envelope 2 (between the fuel processing and
drain tank cells) can be examined and sealed. Finally, Envelope 3 is
the drain tank cell itself, where the fuel and flush salts are actually
stored. Once the reactor cell has been opened, the large penetration

between the drain tank cell and the reactor cell can be sealed.

D.3 DESCRIPTION OF PENETRATIONS

During reactor operation, the molten fuel salt had to be maintained
at a temperature above 840°F (the melting point of the salt) in order to
prevent a plug of solid salt from forming. This was achieved by
surrounding salt transfer lines with electrical heaters, insulation,
thermocouples, pressure reading ins;ruments, and redundant instrumen-—
tation. This complex array of electrical and instrumentation lines
necessitated numerous penetrations through the cell walls. 1In addition
to the cell wall penetrations, there are two large cell-to—cell ope-
nings: a 36—in. opening between the reactor and drain tank cells (Fig.
D.3), and a 14-in. opening between the drain tank and reprocessing cells
(Fig. D.4). These two openings carry insulated, heated lines for salt

transfer.

D.4 REACTOR CELL PENETRATIONS

Typical reactor cell penetrations (see Figs. D.5 through D.9) are
electrical, thermocouple, water, off-gas, and salt transfer lines. All
of these penetrations pass through the sand and water—-filled annulus.,

In addition, the reactor cell penetrations are provided with some mecha-—-
nism to allow movement relative to the cell walls. The 12-ft radius
indicates the reactor containmment cell wall, which was constructed of
stainless steel plate ranging from l-1/4 to 4-in. in thickness. The
15-ft radius indicates the reactor tank liner, which was constructed of
3/8-in.~thick stainless steel plate. The lines shown inside the larger
penetrations are presently connected or were used as spares and left

capped.
 

 

  

 

' 242

 

REF. DWG. X3E-12598-002

 

36°0D. 3 WALL THK.

 

 

 

 

 

 

 

  

 

 

 

/HEATER { INSULATION

 
 
 

 

CELL SIDE

 

 

 

CEL o
& 2] T %
A a1 e Vi
it WO L. N
= 36 CONCRETE=H] |=— 30" =% -
/ S WAL /) JED ONCGRETE VY| 13 PLATE
i 1 WAL Y I5°R

/15 SCH. 40 PIPE

ReacTor-Drain Tank INTERCONNECT

Fig. D-3.

 

 

\ 3?75 0D, 2 WALL THK.

A~ 2-12'R

 

Existing interconnecting penetration between reactor cell and drain tank cell.

86
REF. DWG. X3E-12598-003

g BARYTES BLOCK
gfLELL LINER

[

T 20 CONCRETE-—=1"T5
b9 s 7 WALL - 2

 

v, ;’;glfi |4 SCH. 80 PIPE
b v il
Y o v v v 2
»oa & 1o v D :'P
‘ L ELANGE

(BOTH ENDS) ta

 

Fig. D-4. Existing sait transfer 11ine penetration between drain tank cell and fuel processing cell.

66
REF. DWG. X3E-12598-E002

e

 

-
N
o~

vl

  
   

  

N
ANNULU S

 

2" SCH.40 PIPE

30 DIA ; o
25 MIN. OFFSET CELL SIDE

 

X

  

t

4 PLATE 'R

3 PLATE

Fig. D-5. Typical existing electrical, thermocouple and instrumentation pentrations into reactor cell.

00T
101

REF. DWG. X3E-12598-002

 

 

 

 

 

48
CELL SIDE

 

6 SCH. 160 PIPE

15 SCH. 40 PIPE 2R

 

 

 

 

 

SN Ao o
U g
p o)

 

 

/
r?
A

»

L6 SCH. 40 PIPE

 

5
S
<" CCoNCRETE (

2 3 pATE

8

 

AND § WA

ANNULUS

2’ PLATE

2 I'PIPE BRACES

5 (60" APART)

 

Sameier anp Sampier Orfons LiNes

Fig. D-6.

reactor cell.

Existing sampler and sampler offgas line penetrations into
_CELLSIDE

 

£ SCH. 40 HMPE
{TVP)

Fig. D-7.

REF. DWG. X3E-12598-002

39

 
  
 
  

SEAL PLATE (BOTH ENDSY)

 
  

4" PLATE ' 3 PLATE

| _ 8" SCH. 40 PIPE
WATER LiNes ano OFroas Lines

Existing water and offgas penetrations into reactor cell.

¢0T
TrsTRUMENTATION ELEcTRICAL Lines

PLUG IF NOT USED
;
3'/3 TYp I i
{

 

 

Eet : : £ v & :,‘:’;::9 Rop g b Che e e

B b = DV e T T e e

& aba F b- <1 a T -',.&».‘--'b-.v". e ¥oopi:
: AR

 

36" CONCRETE WALL

sl s B
segee BT

 

LA
v

 

 

R
v

8

- g

8"R et

/% SCH. 40 PIPE

& T g Sra aArtl e
o Kl e e
- LT o
T - b

 

X AN <5

 

 

 

 

t CEL\_ L\NER | |
¥, THREADED

Dran Tank Ceul

 

Fig. D-8.

COUPLINGS

REF. DWG. X3E-12598-003

Cooiant AR\ Cover (5AS LINES

L pLATE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

* A
a-. ~F :I 7T :1/_ g SCH. BO PIPE
B ',?.'k."ff;f-': jf,'_-‘-‘,\ A T SN S TN
S J" | ot T e e
H N
\
[ N
N \
36' CONCRETE WALL \\
"&& :: l %
\g PLATE
‘. -?." '-,'b‘ T A"i“,.-;'p" -1 { e T A gTLer v <
A e SIS
) .
[‘;\_s CELL LINER \6 SCH 80 P\PE

(SSTY

Drain Tank CeLL

 

 

:3" 'SCH. 40 PIPE

Typical existing penetrations into drain tank cell.

€0T
REF. DWG. X3E-12598-003

7 / ' SCH.40 PIPE
S L iJg SCH. 40 PIPE

l<— 2 CELL LINER (5ST)

 

 

 

 
  

 

Deay ANk CELL

 

 

 

36" CONCRETE WALL -

 

 

 

HeLium\ WATER LINES

Fig. D-9. Typical water and helium penetrations into drain tank cell.

01
105

D.5 DRAIN TANK AND FUEL PROCESSING CELL PENETRATIONS

Figures D.8 and D.9 show typical electrical, cover gas, coolant
air, helium, and water transfer lines. All of these lines, except for
those interconnecting the two cells, pass through the concrete cell wall
into access areas. As in the reactor cell, the lines are either con-

nected or were used as spares and left capped.

D.6 TABULATION OF PERTINENT INFORMATION

Much information about the cell penetrations has been collected and
tabulated (Table D.1). This table has been divided into three sections,
one for each of the main cells (reactor, drain tank, and fuel pro-
cessing). Table headings are: (1) penetration identification number,
(2) type/usage, (3) present coundition, (4) location in the cell, (5)
access area, (6) size of the major and interior penetratioms, (7) deter-
mination as to whether the penetration is needed now or in the future,
(8) closure options, and (9) envelope involved. The present condition
of the penetrations (connected, capped, or empty) is yet to be deter-—
mined; however, a detailed on—site inspection would provide most of this
information. Some of the penetrations used in the facility, and their
need for maintenance or for future salt transfer, were not readily

available at the time this study was completed.

D.7 CLOSURE OPTIONS

Numerous penetrations are currently needed for facility surveillance
maintenance. The solid salt is heated annually to recombine fluorine,
in addition to the periodic measurements of temperature and pressure.
Besides the penetrations needed for surveillance and maintenance, the
facility must maintain the capability for transferring the fuel and
flush salts out of the drain tank cell in the event this may be required
in the future. Several closure options are available for those penetra-
tions that are not needed now or in the future. Penetrations with
interior lines can be sealed in one of two ways (Fig. D.10). Each
interior penetration can be capped and left in place, or the end seal-
plate can be removed, the interior lines cut inside the larger penetra-

tion, and the seal-plate replaced with a new, solid plate. The option
Table D-1.

TYPE/USAGE

MSRE reactor and drain tank cell penetration 1ist.

 

 

1.D. PRESENT CONDITION CELL LOCATION ACCESS AREA SIZE NEEDED CLOSURE OPTIONS ENVELOPF
Drain Tank Cell
1 STEAM FROM DOMES (1) South Wall South ESA 3" in &" 1 1
2 STEAM FROM DOMES (1) South Wall South ESA 3" in 4" 1 1
3 WATFR TO STEAM DOMES (1) South Watl South ESA " 1 1
4 WATER TO STEAM DDMES (1) South Wall South ESA " 1 1
5 WATER (1) South Wall South ESA " 1 1
6 WATER (1) South Wall South ESA 1" ] 1
7 SPARE (1) South Wall South ESA " NO 1 i
A SPARE (1} South Wall South ESA 1" NO 1 1
9 HELIUM (1) South Wall South ESA 1/2" in 1" 1 1
10 HELTIM (1) South Wal)l South ESA 1/2" in I" 1 1
11 HELTUM (1) South Wall South ESA 172" in 1" 1 1
12 HELTUM (1) South Wall South ESA 172" in 1" 1 1
13 HELIUM (1) South Wall South ESA 1/2" 4n 1" 1 1
14 HELIIM (1) South Wall South ESA 172" in 1" 1 1
15 HELIUM (1) South Wall South ESA 1/2% in 1" 1 1
16 HELIUM (1) South Wall South ESA 172" in I" 1 ]
17 SLMP DISCHARGE (1) South wWall Waste Cell 3/4" 1 1
ig SUMP DISCHARGE (1) South Wall Waste Cell 3/4% 1 OR 2 1
19 AIR TO SIMP (1) West Wall West of Bldg. /4" 1 OR 2 l
20 AIR TO SUMP (1) West Wall West of Bldgp. 3/4" 1 OR 2 1
21 North Wall Fuel Proc. Cell 1-1/2" 1 OR 2 2
22 North Wall Fuel Proc. Cell 1-1/2" I OR 2 2
23 North wall Fuel Proc. Cell 1-1/2" 1 OR 2 2
24 North wWall Fuel Proc. Cell 1-1/2" 1 OR 2 2
25 North Wall Fuel Proc. Cell !-1/2" 1 OR 2 2
26 North Wall Fuel Proc. Cell 4" I OR 2 2
27 North Wall Fuel Proc. Cell 1-1/2" 1 OR 2 2
28 North wWall Fuel Proc. Cell 1-1/2" 1 OR 2 2
29 North Wall Fuel Proc. Cell 1-1/2" 1 OR 2 2
30 North wWall Fuel Proc. Cell 14" p 2
A~-36 INSTRUMENTATION (29) East Wall North ESA 374" 3 1
B-36 TINSTRUMENTATION (36) Fast Wall North ESA 374" 3 1
C-36 THERMOCOUPLES (36) East Wall North ESA 3/4" 3 1
D-36 THERMOCOUPLES (36) East Wall North ESA 3/4" 3 1
E-36 THERMCCOUPLES (36) East Wall North ESA 3/4" 3 1
A-24 ELECTRICAL (24) East Wall North ESA 3/16" in 3/a" 3 1
B-24  ELFCTRICAL (24) East Wall North FSA 3/16" in 3/4" 3 1

90T
Table D-1, continued.

e e e e e e o o S st

 

C-24  ELECTRICAL {(24) East Wall North ESA 3/16" in 3/4" 3 1
D-24  ELECTRICAL (24) . East Wall North ESA 3/16" in 3/4" 3 1
E-24 ELECTRICAL (24) East Wall North ESA 3/16" in 3/4" 3 1
F-24  ELECTRICAL (24) East Wall North ESA 3/16" in 3/4" 3 1
G SPARE (1) East Wall North ESA 6" NO 1 OR 2 1
H SPARE (1) East Wall North ESA 6" NO 1 OR 2 1
I COVER GAS (3) East Wall North ESA 6" (3 1/2") 1 OR 2 1
J SPARE (1) East Wall North ESA 6" NO 1 OR 2 1
K SPARE (1) East Wall North ESA 6" NO 1 oR 2 1
L SPARE (1) East Wall North ESA 6" NO 1 OR 2 1
M COMPONENT COOLANT AIR (3) East Wall North ESA 6" (3 3/4") 1 OR 2 1
N COUMPONENT COOLANT AIR (3) East Wall North ESA 6" (3 I/4") 1 OR 2 ]
o COMPONENT COOLANT AIR (1) East Wall North ESA 6" (ONE 1 1/2') 1 OR 2 1
P DP CELL (4) East Wall North ESA 6" (4 1/2™) 1 OR 2 1
Q LEAK DETECTOR (12) East Wall South ESA 6" (12 1/4") 1 OR 2 1
R LEAK DETECTOR (12) East Wall South ESA 6" (12 1/4™ 1 OR 2 1
Reactor Cell Ft-in., Degrees
I REACTOR LEAK DETECTORS (60) 836 15 South ESA 24" (60 1/4™) NO 2 0R 3 1
Il ELECTRICAL (44) 834 30 South ESA 24" (6 1";38 3/4") NO 2 0rR 3 1
111 ELECTRICAL (44) B36 45 South ESA 24" (6 1™;38 3/4") NO 2 OR 3 )
v THERMOCOUPLES (60) 834 60 South ESA 24" (7 3/4";28 1/2";25 3/8'")HNO 2 OR 3 1
v INSTRUMENTATTON (60Q) 836 75 South ESA 24" (60 3/8™) NO 2 0R 3 1
VI SAMPLER OFFGAS 847 110 High Bay 4" (2 1/2") NO 2 1
V1l SAMPLER 847 115 High Bay 6" NO 2 1
VIII FUEL PUMP AUX. PIPING (8) 836-9 125 Service Tumnel 18" (3 1/2";4 374";1 1) NO ? 1
X NEUTRON INSTRUMENT TUBE 834-5 145 High Bay 36" (10 ION CHAMBER GUIDES) NO ? 1
X FUEL PUMP LIQ. LEVEL (6) 844-6 155 Special Eq. Rm. 4" (6 1/a™) NO 2 ]
X1 FUEL PUMP AUX. PIPING (13) 836-9 160 Special Eq. Rm. 18" (2 1";11 1/2") NO ? 1
X1t COMPONENT COOLANT AIR (1) 829-10 165 Special Eq. Rm. 6" NO ? ]
XIIT COOLANT SALT TO HX (1) 840-10 170 Coolant Cell 24" NO ? i
X1V WATER LINES (2) 839-9 185 Coolant Cell 8" (2 1" NO 1 OR 2 1
XV SPARE (2) 839-9 200 Coolant Cell 8" (2 2") NO 1 OR 2 1
XV1 WATER LINES (2) 839-9 205 Coolant Cell 8" (2 2") NO 1 0R 2 1
XVII WATER LINES (2) 839-9 210 Coolant Cell 8" (2 2™ NO 1 OR 2 1
XVIII  WATER LINES (2) 819-9 220 Coolant Cell 8" (2 2™ NO 1 OR 2 1
XIX COOL. SALT TO RADIATOR (1) 837 220 Coolant Cell 24" NO ? 1
XX OFFGAS (3) 839-9 225 Coolant Cell 6" (3 1™ NO 1 OR 2 }
XXI OFFGAS (2) . 839-9 230 Coolant Cell 6" (TWO 1 1/4") NO . 1 OR 2 1
XX11 CELL EXHAUST DUCT (1) 824-10 245 CDT Tunnel 30" NO 2 1
XX1IT  THERMOCOUPLE (38) 816 325 West Tunnel 24" (38 3/8™) NO 2 0R 1 1
XX1v DRAIN TANK CEL1. INTERCON. 825-2 330 Drain Tank Cell 36" NO 2 3
XXv SLMpP Bottom Center NO ? 1

L0T
Table D-1, continued.

et P A

FUEL PROCESSING CELL

31
32
33
34
35
36
37
a8
9
40
4]
42
43
Q4
45
46
47
48
49
50
51
52
53
54

DRAIN

STEAM LINE

STEAM JET DISCHARGE
STEAM LINE

STEAM LINE

DRAIN
DRAIN

ULTRASONTC PROBE
INSTRUMENTATION
INSTRUMENTATION

INSTRIMENTATION
SALT ADDITION LINE

Tt AL bt e, gk e - e o

NORTH WALL
NORTH WALL
NORTH WALL
NORTH WALL
NORTH WALL
NORTH WALL
EAST WALL
EAST WALL
EAST WALL
EAST WALL
EAST WALL
EAST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
WEST WALL
ROOF

DECON. CELL
DECON. CELL
DECON. CELL

DRAIN TANK CELL

DECON., CELL
DECCON. CFLL
SPARE CELL
SPARE CELL
SPARE. CELL
SPARE CELL
SPARE CELL
SPARE CELL
HIGH BAY

ABSOR.
ABSOR.
ABSOR.
?

?

?

INSTR.
INSTR.
ABSOR.
INSTR.

CUBE.
CUBE.
CUBE.

CUBE.
CUBE.
CUBE.
CUBF..

HIGH BAY

6"
3"
374"
34"
374"
3/&"
6“
12“
12"
12"

6"

1 172"

 

YES

YES
YES
YES
YES
YES
YES

Gt g bt gt Gt e i et et e et Bt et et Gt T e e ed b e

OR
OR
OR
OR
OR
OR
OR
CR
OR
OR
OR
OR
OR
OR
OR
OR
OR
OR
OR
OR
OR
OR
OR

B R R RN NN RN NRNNRNRRNRNNRNNLDNRNDN

s it (b Wi bt Bt e S e e e ems Mme i e ot A mae R e s e

 

80T
Needed 1ine (connected

Unneeded Tines (capped)—
. / f

01d seal plate

 

 

<s ] K S S

 

 

N
S S D —

S
SO

N

 

 

 

 

 

 

 

 

 

 

TS
¥
S

 

 

CLOSURE OPTION 1

 

NOTE: Closure option 3 is a threaded plug.

ORNL-DWG 85-1194

;;é;:is cut inside large penetration
\
A

SN

 

 

 

 

 

 

 

 

 

 

 

\
\Qi?\b\ New seal plate
NS
\\\\\\‘\."\\.
\\l \\
lL\ \\ NS NN NN
N \\\::\\_
/] ] A
. - e
a. M .:'." AL AT
- . 4w e -
b SR ? |-

[

/]

" 1
%

A

L/ L/
“

¢ ¢

/ ;

"

] o o a) b ;

HM'F

 

 

 

 

CLOSURE OPTION 2

Fig. D-10. Closure options for multiple penetration assemblies.

60T
110

chosen depends on the use of the interior lines. if any of the interior
lines are needed, then the sealing of individual lines would allow the
needed lines to remain connected. If none of the interior penetrations
is required, the method of replacing the seal-plate with a new, solid
plate would eliminate exposed lines. Electrical lines can be sealed
with a threaded, pressure-tight plug, such as those used for sealing

spare electrical lines in the original construction.

D.7 SUMMARY

The information developed in the course of this investigation into
the condition of the cell wall penetrations of the MSRE facility was of
a preliminary nature, and more work is needed. Completion of the tabu-
lated data should be undertaken with a particular attention to further
on-site inspection to verify conditions. The preparations required for
closure of most penetrations is straightforward since most of the
electrical and gas lines are readily accessible and sealing can be
achieved in the same manner that spare lines are capped. However, spe-—
cial preparations will be required in two cases: (1) any penetration
that will, when cut, have direct contact with cell atmosphere will have
to be identified, and (2) the sampler and sampler off-gas lines will
require considerable effort to remove the lead shielding above the
penetrations in order to obtain access. Most of the penetrations in the
facility are associated with the reactor and drain tank cells. Other
than those penetrations required for monitoring the fuel storage con-
ditions, final sealing of the wall penetrations that exist in the drain
tank cell would achieve the maximum security for in-place fuel storage
for an extended period. The major effort, of course, would be for the
cleanup to gain access to the reactor and fuel processing cells so that

final isolation of the drain tank cell could be achieved.
111

Appendix E. BUILDING 7503 DRAINAGE SYSTEMS

Prepared by

F. J. Peretz

Martin Marietta Energy Systems, Inc., Engineering

August 1985
112

BUILDING 7503 DRAINAGE SYSTEMS

This appendix presents a review of the existing literature and design data
for the Building 7503 drainage systems. Further work to field check the
condition of the drainage systems would be useful, as would studies to

define the flow patterns and volumes of groundwater at the site.

Overall Description

 

Building 7503 is served primarily by four drainage systems. (1) A special
radioactive 1iquid waste system collects water from process and cell floor
drains in a liquid waste tank located in the liquid waste cell. MWaste is
neutralized in this tank if necessary, and then is pumped to the LLW pumping
station at the northwest corner of the site. Waste is then pumped over to
the Bethel Valley facilities for concentration and disposal. (2) Non-
radioactive drainage, including groundwater collected by the building

french drain system and water entering floor drains outside of the cells, is
collected in the building sump. Water is then pumped out of the sump and
discharged from a catch basin southwest of the building out to a drainage
area leading to Melton Hi1l Branch. (3) Storm water is collected in a
series of roof drains, and is piped to another catch basin west of the
building. This water is also discharged to a drainage area. (4) Sanitary
wastewater is collected and drained to a septic tank and drain field also
located west of the building. Fig. E-1 shows a schematic of the storm and
nonradioactive buiding water rainage systems, including the locations of the
catch basins and discharge points. Drawing D-A-AA-40888-C (in pocket inside
back cover) shows the flow diagram for the building radioactive and non-

radioactive collection piping systems. Further details on all of the
Unclassified
ORINL DWG 64-9109

 

From Roof
Drains N I

Building 7503

 

 

A E 32733.29

T —Il//._,B-in. Drain From
/’L'-1 Cell Annulus

 

 

  

 

  

3 in.
L - = '\
4Yein. Drain Tile
Footing Draln
' J
| bscrber Pit
Catch Basin | © i Catch Basin r_. 8 1in.
Invert ' Invert .
N+ = 826 £t 11 in. 843 £t O in. E32665.18
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Fig. E-1.

N18592.83
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Schematic diagram of Building 7503 drainage plan.
114

building drainage systems are given in ORNL-TM-728, Part 1, and in
ORNL/CF-77/391.

Radioactive Liquid Waste Collection

 

The central point for radioactive liquid waste collection is a 11,000

gal. tank located in the liquid waste cell in the north half of the build-
ing. The major sources of waste for which this collection system was
designed are the fioor sumps in each cell, decontamination solutions from
the decontamination tank (which was never installed), and drainage from
"hot" sinks. Radioactive water collected in the building sump room can also
be sent to the 1iquid waste tank. The major lines of the radioactive waste
system are listed in Table E-1. Each cell is provided with a sump to
collect any water which might enter the cell. Steam jets are used to move
water from the sump to the waste tank in all celis except the drain tank
cell and the reactor cell. Those two cells use air jets to avoid
introducing steam into cells containing fuel salts. One other exception is
the sump added to the coolant drain tank cell, which is jetted to a sewer in
the west tunnel of the south electric service area. Under normal condi-
tions, no water is present in the sumps, indicating that the cells do indeed

remain dry.

Building Water Collection and Dispersal

 

The central point of the nonradioactve building water collection system is
the pump room and building sump (Fig. E-2), located underneath the special
equipment room in the southeast part of the building. Access to the pump
room is gained through a ladder tunnel beginning on the main floor

(elevation 852 ft). Two sump pumps and one "pit pump" are in the pump room,
115

 

 

Table E-1. Radioactive liquid waste lines

Line Description Material Size
No.

Lines entering liquid waste storage tank:

333 Reactor cell sump (air jet) Stainless 3/4"
343 Drain tank cell sump (air jet) Stainless 3/4"
314 Caustic scrubber tank steam jet Steel 3/4"
316 Liquid waste cell sump (steam jet) Stainless 3/4"
318 Equipment storage cell sump (steam jet) Stainless 3/4"
320 Spare cell sump {steam jet) Stainless 3/4"
322 Fuel processing cell sump (steam jet) Stainless 3/4"
326 Pit pump discharge Stainless 2"
339 Caustic addition Stainless 2"
340 Hot sinks Stainless 2"
308 Process water addition {siphon break) Stainless 3/4"
301 Waste pump discharge (recirc. & mixing) Stainless 2"
948 Waste tank vent to blower Stainless 6"
Lines entering waste pump:

300 Liquid waste storage tank discharge Stainless 2-1/2"
303 Decontamination tank discharge Stainless 2-1/2"
Other lines:

305 Waste pump to central pumping station Stainless 2"
302 Waste pump to sand filter Stainless 2"
306 Sand filter to decontamination tank Stainless 2"
303 Decontamination cell sump to waste pump Stainless

 
AV E L TROMAETER

 

 

116

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Section of sump room, looking east. (ART configuration)
117

as seen in the flowsheet D-AA-A-40888-C. The sump and pump room were
constructed as part of the building modifications for installation of the
Aircraft Reactor Test (ART) in the late 1950's, and was modified only
slightly for the MSRE. Thus, most of the design information is found with
the design package for the ART. Drawing D-KP-19040-F shows many of the

construction details of the building drainage system.

The building sump collects water from nonradioactive floor drains throughout
the facility, as listed in Table E-2. Standard designs were used for these
drains. The sump also recieves groundwater from a french drain system which
serves the deeper cells and support areas in the south end of the building,
and the charcoal bed south of the building. Again, this french drain

system was largely constructed as part of the ART modifications. Figs. E-3,
E-4 and E-5 show cross sections of the building foundations, with the
approximate elevations of the french drains indicated. Similar french
drains may exist in the original north section of the building, but specific
design information was not found. The two sump pumps in the pump room were
recently instrumented to determine what fraction of time they operate.
Preliminary indications are that the pump with the deeper float operates
about one hour a day, discharging 2000 gallons in an hour. The other pump,
which would only operate should the first pump not be able to keep up with
the inflow, has not operated during this period of observation. At one time
during the interim period between ART and MSRE construction, the sump pump
failed and was out of operation for some time. During this time, the lower
levels of the building were flooded with several feet of water. This
experience indicates that continued operation of the pumps is necessary for

the assured safe storage of the salts in the MSRE facility.
118

 

 

Table E-2. Standard building drainage lines.
Line Description Material Size
No.
Drain lines entering sump:
352 Radiator air duct floor drain Cast iron 3"
353 Radiator stack floor drain Cast iron 3"
354 Coolant drain cell floor drain Steel 3"
364 Coolant drain cell (via 354) Cast iron 3"
355 Blower house floor drain Cast iron 3"
356 Blower house ramp drain (via 355) Cast iron 3"
357 West tunnel floor drain (via 355) Cast iron 3"
358 Service room floor drain Cast iron 4"
359 Service tunnel drain Steel 3"
365 Vent house valve pit drain Cast iron 2"
325 55 gal drum drain, pit pump suction Steel 3"
French drain 1ines entering sump:
361 Reactor cell french drains Cast iron 4"
363  Southeast french drains (via 361) Cast iron 4"
362 Southwest building and charcoal bed
cell french drains (via 361) Cast iron 4"
360 Sevice tunnel french drains Cast iron 4"
Drain lines entering 55 gal. monitoring drum:
350 Ventilation stack drain line Cast iron 3"
3561 Filter pit drain lines Stainless 2"
Lines entering outdoor catch basin:
327  Sump pump "A" discharge Steel 3"
337  Sump pump "B" discharge (via 327) Steel 2"
Charcoal bed cell overflow (via 327)
331 Reactor cell annulus overflow Steel 8"
330 Pit pump discharge (via 331) Steel 4"
Lines entering pit pump:
325 55 gal. monitoring drum discharge Steel 3"
328 Charcoal bed cell drainage Cast iron 3"
329  Reactor cell annulus drainage Steel 3"
Other lines:
326 Pit pump discharge to liquid waste
storage tank Stainless 2"
Catch basin to field Concrete 10"

 
119
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122

Several lines entering the pump room flow into a 55 gal. drum rather than
entering the sump directly. The drum is used to collect drainage from areas
which might have contained small levels of activity, and allows for
monitoring prior to release. The "pit pump" could be used to send drainage
colleted in the drum to the radioactive liquid waste tank in the event that

activity is detected.

Water from the sump is pumped up to a catch basin southwest of the building,
and flows out into a field through a 12 in. concrete line. As shown in

Fig. E-1, the 3 in. pump discharge line branches into an 8 in. cell annulus
drain line before circling around the charéoal absorber pit en route to the
catch basin. The reactor cell annulus can be emptied by pumping to the
catch basin via line 331 using the pit pump. This operation would remove a
major reservior of water adjacent to the cell in which the salts are now
stored. Caution in disposal of this water would be dictated by the
presence of chromate rust inhibitors, as well as the possibility of induced

radioactivity.

Storm Water Dispersal

 

Collection of storm water is accomplished with a standard roof drain piping
system. Water flows by gravity to a catch basin, separate from the one
serving the building sump, from which it flows out another 12 in. concrete
line (Fig. E-1). Site grading and ditches are used to carry storm water

from around the perimiter of the building.
123

Sanitary Sewage Collection and Disposal
A standard sanitary sewage collection system was installed in the building.
Sewage flows by gravity to a septic tank located west of the building. The

water leaving the septic tank then flows into a drain field west of the

septic tank.
.y

Appendix F.

125

GEOLOGIC INVESTIGATIONS RELATIVE TO MOLTEN SALT REACTOR
DECOMMISSIONING

Dr. D. W. Byerly, Department of Geology

University of Tennessee, Knoxville, Tennessee

December 1984
126

Geologic Investigations Relative to Molten Salt Reactor Decommissioning
Dr. D. W. Byerly, Department of Geology

University of Tennessee, Knoxville, Tennessee

The Molten Salt Reactor Experiment (MSRE) facility is located in
the Melton Valley area of the Oak Ridge National Laboratory (ORNL). The
valley is bordered by Haw Ridge (elevation 317 m) on the northwest and
Copper Ridge (elevation 400 m) on the southeast. Bulldings housing the
reactor and other facilities related to the reactor are situated on the
northwest side of the valley near the base of Haw Ridge. Topographically,
the valley is not flat, but instead consists of relatively steep-sided,
irregularly-shaped, low hills with summit elevations averaging about
275 m. Relief in the valley ranges up to 43 m.

The buildings are at a ground elevation of about 259 m within a slight
swale (partially resulting from construction back-fill) near the head of a
draw which is tributary to Melton Branch that flows southwesterly through
the main valley. All drainage from Melton Valley ultimately enters the
Clinch River below Melton Hill Dam southwest of the confluence of Melton
Branch at White Oak Creek.

The geological conditions of portions of Melton Valley have been
interpreted as favorable for storage/disposal of low-level radioactive
wastes. The facilities in Melton Valley used for hosting the ORNL-
generated low—-level radiocactive wastes include: burial grounds, waste
pits, treatment operations, and hydraulic fracturing.

Bedrock forming the foundation for the MSRE consists of the lower
portion of a "package"” of rocks referred to as the Conasauga Group.

Rocks of this group grade compositionally downward into the Rome Forma-~
tion underlying Haw Ridge and upward into the Knox Group which forms
Copper Ridge. Variegated shales, commonly calcareous, comprise most of
the lower Conasauga; however, intercalations of silty to pure limestone
may be present as lenses. The upper portion of the Conasauga, on the
southeastern side of the valley where gradation with the Knox Group

occurs, generally contains more limestone.
127

The knobby topography in Melton Valley is typical of weathered and
eroded shale. The knobs or hills are aligned in rows paralleling the
regional structure, forming above the shaly bedrock, whereas the interven-
ing low areas are ordinarily underlain by the carbonate rocks. Regolith
above the bedrock is variable in thickness. Generally, it is thickest
(up to 9 m) on hill crests and thinnest (less than 1.5 m) in the flat,
low-lying areas. The total thickness of the Conasauga Group is about
600 m.

The trace of the Copper Creek fault, a major thrust fault in the
Ridge and Valley province of the southern Appalachian orogen, strikes NE-
SW along the northwest slope of Haw Ridge. The fault dips southeast
projecting below Melton Valley and Copper Ridge at considerable depth.

Sedimentary strata of the Conasauga Group forming Melton Valley also
strike northeast and dip to the southeast, forming the hanging wall of
the Copper Creek fault. The fault and related deformation of the rocks
in the valley resulted from tectonism that occurred over 200 million
years ago. Subsequent deformation of these rocks is not evident.

Considering the present tectonic setting of North America and the
fact that the fault as well as structural elements have not been active,
at least during the most recent 70 to 80 million years, the area can be
considered to have a low seismic risk (seismic risk zone 2, Modified
Mercalli intensity less than V or VI). The nearest significant earthquake
epicenters historically are Charleston, South Carolina, and New Madrid,
Missouri; however, both are considered too distant to have or have had
serious detrimental impacts on this area.

Joints or fractures within the Conasauga that are associated with
past tectonism have not appreciably affected the strength of the bedrock,
but have resulted in a secondary effective porosity which influences the
geohydrology of the valley.

In Melton Valley, the zone of ground water saturation generally
occurs at greater depths beneath summits than in the low-lying bottoms.
In low areas such as adjacent to perennial streams like Melton Branch or

to ephemeral streams in the gullies dissecting the low hills in Melton
128

Valley, the water table may be found within l.5 m below the land surface.
On hill tops, the depth to the water table may exceed 6 m. The general
pattern of ground water movement is laterally along bedding strike
(NE-SW), but jointing often modifies this pattern. An average rate of
ground water movement within the shales of the Conasauga has been deter-
mined to be 15 cm/day.

Ground water behavior at the site is a key factor which must be
addressed prior to deciding the disposition of the radioactive materials
contained in the MSRE facilities. A hydrologic monitoring program should
be developed to establish baseline qualitative and quantitative water
data. Water is an important factor for at least two reasons. One, it can
potentially deteriorate the waste containers, especially if the geochem-
istry of the water is reactive, and two, water is the most likely vehicle
for transmitting the radioactivity to the biosphere should containment
fail,

Al though the MSRE facility is apparently sound structurally, it must
be kept in mind that the facility was not designed for long—-term entomb-
ment of radioactive wastes. Multilevels of barriers, including engineered
as well as the natural geologic setting, are necessary for safe, long-term
storage/disposal of radioactive wastes. Therefore, several geological
aspects of the facility site need more assessment of a site-specific
nature before on-site storage/disposal is considered a feasible alterna-
tive. To be acceptable, the site should be capable of safely containing
the wastes without institutional controls (i.e., sump pumping of
infiltrated water, etc.). Among the geologic aspects that should be
included in further evaluation of the site include:

® Characterization of all earth materials on the site (anthropogeneous
material, bedrock, etc.)
® Site design — slopes, placement of fill soils, underdrains, founda-
tions, etc.
® Site hydrology — water inventory including quantity and quality,
ground water flow nets, piezometric levels, etc. '
Site-specific investigation of the above should be considered as the very
minimum of any program plan for the evaluation of the site for on-site

entombment of the MSRE radioactive materials.
44,

45"49.

50_540

55-56.

129

ORNL/TM-9756

INTERNAL DISTRIBUTION

Je. F. Alexander 26, F. J. Peretz

T. W. Burwinkle 27. L. P. Pugh

K. W. Cook 28-32. T. H. Row

B. L. Corbett 33. J. H. Swanks

Ne. W. Durfee 34, R. E. Thoma

Je R. Engel 35. W. W. Thompson

D. E. Ferguson 36. J. R. Trabalka

P. N. Haubenreich 37. V. C. A, Vaughen

F. J. Homan - 38, R. G. Wymer

J. M. Kennerly 39. Central Research Library
L. E. McNeese 40. Document Reference Section
T. E. Myrick 41. Laboratory Records

K. J. Notz 42, Laboratory Records - RC
P. T. Owen 43, ORNL Patent Section

EXTERNAL DISTRIBUTION

Office of Assistant Manager for Energy Research and Development,
DOE-ORO, P.0O. Box E, 0Oak Ridge, TN 37831

R. N. Coy, Office of Surplus Facilities Management, UNC Nuclear
Industries, Post Office Box 490, Richland, WA 99352

C. E. Miller, Program Manager, Surplus Facilities Management
Program, Richland Operations Office, U.S. Department of Energy,
Post Office Box 550, Richland, WA 99352

Technical Information Center, Post Office Box 62, Oak Ridge, TN
37831
 

 

 

 

    

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 
  
 
  
 
  

 

 

 

  

 

 

 

 

 

 

 

 

 

 

   
 

 

 

   

 
  

  

   
  
     
 

 

 

 

 

 

 
 
 
    
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 
 
 

 

 

 

 

 

 

      
  
      
     
  
        

 

 

    
 
     

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

  

 

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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max. Ap s PSt TYPE CANNEgP :QTOR
i o .
o5 CAPACITY |4 ) _
= HE AD 80
V307 (g :
— 2
e :
»
(m) WASTE FILTER J‘f
Wl FUEL PROCESSING SYSTEM D-AA-A-30887 *
COOLING WATER SYSTEM D-AA-A-40889 g4 4
3_ 40-S5 LIqUiC WASTE SYSTEM INST. APPLICATION DIA. |D-Aa-B-4050e |® i
7 | FUEL SYSTEM DAA-A-40880. i
@ > [CRAIN TANK SYSTEM. D AA-A-40882 : *
of GAS SYSTEM B CONTAINMENT VENT. = 1D-AA-A-40883 | %pr
2% - 40-55— .t
' © REFERESCE DRAWINGS DWG. NO. R '
| ) pren. gort: e e St 1o, OAK RIDGE NATIONAL LABORATORY : i
.‘.’fi /V/'(““ At L 258 CTHERWIST . %
3418 6 |SEE DCN 3418 SiG. %é&&% i ?5L 63—"'%{’ FROJICT INSR, BEEiCH TRGINIER ToLEmARCE uaL LlQUlD WASTE SYSTEM . ' '
3306 5 | SEE DCN 3306 T 58 - acnon e s
3102 D [SEE OCN 3102 B-63pim £ e 7 rescrons PROCESS FLOW SHEET
2930 | C |SEE DCN 2930 ikl /‘L,éé..u 77 e = M.S.R.E.
266! B [SEE DCN 2861 ”,;g.‘ (ff.’ T : AAA-40888
249} A |SEE OCN 249i L : . #, : NONE JoB 433-1.0 D-AA-A~ ’
: ~ EArAE: | | Butle 8Ly [Py SEALE? MO

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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