TECHNICAL NOTE
ORP/TAD-77-1
Evaluation of
Tritium
Recycle and Buildup
in a
Pressurized Water Reactor
THE UNITED STATES
ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
March 1977
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TECHNICAL NOTE
ORP/TAD-77-1
Evaluation of Tritium Recycle
and Buildup in a,Pressurized water Reactor
C. Bruce Smith
U.S. Environmental Protection Agency
Office of Radiation Programs
Technology Assessment Division
Energy Systems Analysis Branch
March 1977
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PREFACE
Phe Office of Radiation Programs (ORP) of the Environmental
Protection Agency carries out a national program designed to evaluate
public health impact from ionizing and nonionizing radiation, and to
promote development of control necessary to protect the public health
and the environment. This report provides the technical information
necessary for ORP to evaluate the environmental aspects concerning
tritium recycle in a pressurized water reactor (PWR). This
information is important since the impact of tritium recycle is
significantly affected by the design and operation of PWRs. Results
of this paper may be used as input into dose models to analyze the
total environmental impacts of tritium released for various tritium
recycle operations.
David s. Smith
Director
Technology Assessment Division (AW-459)
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TABLE OF CONTENTS
N
Page
SUMMARY i
I. Introduction 1
II. Physical Model of Activity Transport 1
A. Routine Operations 2
1. Compartment 1 - Primary Coolant System 2
2. Compartment 2 - Liquid Radwaste System 3
and RMWST
3. Compartment 3 - Spent Fuel Pit 3
H. Compartment H - Refueling Water Storage 3
Tank
5. Compartments 5 and 6 - Secondary System 4
Steam Generators and the Remaining
Secondary System
B. Shutdown 4
III., Mathematical Model of Activity Buildup 5
IV. Parameter Selection 7
A. Trojan, FSAR Data, Chapter 11 7
B. PWR-Gale Code NUREG-0017 8
C. Other Parameters 8
D. Conclusions of Parameter Values 9
E. Ranges of Parameter 9
V. Analysis and Results 10
A. Annual Tritium Source Term 10
B. Analysis of a Standard Tritium Source Term of 11
700 Ci/year
C. Effects of Secondary System Parameters 13
1. Reactor Losses as a Function of Steam 13
Generator Slowdown
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2. Reactor Losses as a functiort sof Primary 13
to Secondary Leak Rate
D, Effects of Various Limiting Concentrations 14
VI. Significance of Results 14
REFERENCES 16
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TABLES
Table
1. Effects of Various Limiting 17
Primary Coolant Concentrations
2. Parameter Identification 18
3. Base Case Parameters 22
U. PWR Tritium Releases 23
FIGURES
Figure
1. Schematic of Systems 24
2. Block Diagram of System 25
During Routine operation
3. Refueling Volumes 26
U. Mathematical Models of the 27
Compartments
5. Primary Coolant Activities 28
6. Plant Activity versus Time 29
7. Intentional Removal versus 30
Source Term
8. Time Until Removal Begins versus 31
Source Term
9. Activity Buildup in Each Compartment 32
(700 Ci/yr Source Term)
10. Tritium Losses from Base Plant Design 33
(700 Ci/yr Source Term)
11. Liquid and Gaseous Environmental 34
Releases
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12. Secondary System Releases versus 35
Steam Generator Slowdown Rate
(700 Ci/yr Source Term)
13. Effect of Primary to Secondary 36
Leakage (700 Ci/yr Source Term)
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SUMMARY
Recycling radioactive liquid wastes at light-water cooled nuclear
power reactors is becoming a popular method of .limiting release levels
to as low as practicable. As a result of recycling liquid waste,
almost all of the radioactive materials in a liquid waste stream can
be renoved by filters, demineralizers, reverse osmosis units, or
svaporators. However, tritium in the liquid waste streams cannot be
selectively removed by the aforementioned conventional liquid radwaste
treatment equipment. Combined with its relatively long half life
(12., 33 years), recycling results in a buildup of large volumes of
tritium contaminated liquids.
In a boiling water reactor (BWR), the production of tritium is
primarily in the fuel and control rods. The zirconium cladding around
these rods prevents most of the tritium from escaping from the rods to
the reactor liquids. Thus, the buildup of tritium in a BWR is not
considered a significant plant problem. In a pressurized water
reactor (PWR), tritium is also produced in the fuel and control rods,
as well as by production from chemicals (boron, lithium) in the
priaiary coolant liquids. Since the quantities of tritium introduced
into the primary coolant system of a PWR are much higher than in a
BWR, the management and impact of tritium at a PWR must be evaluated.
It stiould also be noted that most of the tritium produced in the fuel
of PWRs or BWRs remains in the clad fuel and control rods and is
eventually transported to the reprocessing plant.
To evaluate the magnitude of the potential tritium handling
problem at PWRs and to develop recommendations for solutions, a
computerized model of tritium buildup in a PWR was developed. The
results from this model will be used to assess the potential
environmental impact of tritium recycle for the entire nuclear power
reactor industry and will be instrumental in developing standards or
guidance concerning tritium recycle. The model incorporates six
compartments - primary coolant system, liquid radwaste system,
condensate storage tank, spent fuel pool, steam generators, and the
remainder of the secondary system; and 4 release pathways
containment building leakage, primary-to-secondary leakage,
evaporation from the spent fuel pool and refueling canal and
intentional removal from the primary system to limit the buildup of
tritium in the coolant. A reactor operator can control the buildup of
tritium in the primary coolant system of a PWR by utilizing several
control options - the intentional removal of liquids from the primary
coolant system with subsequent offsite disposal, the intentional
environmental release of liquids from the liquid radioactive waste
system, and an increase of the evaporation rates from the spent fuel
pit and the refueling water canal. To illustrate the effects of
tritium recycle, a total recycle system with intentional removal of
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primary coolant liquids was selected as the operator control
mechanism. Thus, the planned release of tritium to the environment
from the liquid radioactive waste system was not included in this
analysis. EPA. has not endorsed any of the previously mentioned
control mechanisms, and will reserve judgment until a complete
evaluation has been made. The computer model was used to analyze the
environmental tritium releases, buildup of tritium in the reactor,
volumes and activities of tritiated water that must be removed to
control tritium buildup, the effects of various source terms on
tritium buildup, and the sensitivity of the buildup to various
secondary system parameters. Generally, the following conclusions
were reached for the base case plant (discussed later in this report)
that is presented:
(1) Based on the reactor system analyzed in this paper,
intentional removal of tritium from the primary coolant systen
bacomes necessary during the 40 year life of the reactor if the
production and release of tritium in the primary coolant system is
higher than 625 curies per year. In the remainder of this paper
this production and release is termed the "source" term. For
source terms greater than 1,000 curies/year, removal becomes
necessary within 5 years of initial operation of the reactor. The
removal (termed intentional removal) involves either the removal
of volumes of tritiated primary coolant system liquids for
solidification and offsite disposal or selective removal (and
disposal) of tritium from the primary coolant liquids. This
removal is necessary to limit the concentration of tritium in the
primary coolant to established operating conditions for the
reactor (assumed in this paper to be 2.5 jjCi/ml of H3).
(2) For a source term Df 700 curies/year, about 80% of the losses
from the reactor during the reactor life are equally distributed
between radioactive decay and uncontrollable releases to the
environment. For larger source terms, intentional removal becomes
the dominating plant loss mechanism. This removal is necessary
for the reasons indicated in (1) above.
(3) During the 10 years of operation of a reactor whose source
term is 700 curies per year, about 80% of the uncontrollable
environmental releases are equally distributed between evaporation
and leakage to the containment building with subsequent release to
the environment.
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(4) The rate of steam generator blowdown in the secondary system
of: a PWR has little effect upon losses or buildup of tritium at a
reactor. However, if there is no primary-to-secondary leakage,
there is a significant reduction in environmental losses from the
reactor in conjunction with an increase in the volumes of
tritiated liquids that must be intentionally removed from the
primary coolant system.
In general, it is concluded that for the base case presented and a
source term production of 700 curies per year that (1) uncontrolled
environmental releases can be reduced during the UO year life Of the
reactor by 65% or more by using total recycle of tritiated liquids*;
(2) several thousand curies and several hundred thousand gallons of
tritiated water (with a tritium concentration of approximately 2.5
MCi/nl) must be intentionally removed for disposal or treatment over
the plant life if "total" tritiated liquid recycle is undertaken; and
(3) about one million gallons of water and 5,000 curies of tritium
will be remaining in the reactor systems at the end of its planned 40
year life.
*NOPEJ Removal of tritium from the reactor is necessary to maintain
the primary coolant concentration of tritium to the established
operating conditions assumed in this paper (2.5pCi/ml). The
removed tritium is also assumed not to be available as an
environmental contaminant.
iii
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EVALUATION OF TRITIUM RECYCLE AND BUILDUP IN A PWR
I. INTRODUCTION
The major modes of production of tritium in a PWR is by fissioning
in fuel rods, neutron capture reactions with poison material used in
control rods, and activation of primary coolant additives, such as
boron and lithium. The tritium produced in the fuel rods and control
rods is released to the primary coolant either through defects in the
rod claddings or by diffusion through the cladding. Tritium
eventually contaminates the liquid radwaste system, the secondary
coolant system, the spent fuel pit, and the condensate storage tank
either during routine operations or refueling operations. To model
the buildup and release of tritium at a PWR employing recycle of
tritium contaminated liquids, the interrelationships among systems and
the movement of tritium to and from these systems must be analyzed.
The following sections describe the physical and mathematical models
of tritium recycle and the results of selected parametric studies.
II. PHYSICAL MODEL OF ACTIVITY TRANSPORT
The relationships of the PWR systems, or compartments, are
illustrated in a schematic model in Figure 1. The PWR systems
containing tritium can be separated into the following functional
compartments: (1) primary system; (2) liquid radwaste system and
reactor makeup water storage tank (RMWST) ; (3) spent fuel pit; (U)
refueling water storage tank; (5) steam generators; and (6) the
remainder of the secondary system. A block diagram of the transport
of tritium into and out of these compartments is presented in
Figure 2.
Actually there are more leakage pathways and routing options
available in a PWR than are presented in Figures 1 and 2. However, in
modeling the movement of tritium in a PWR some pathways may be
combined or even eliminated without significantly affecting the
overall modeling. For example, the leakage of liquids from the
primary coolant system into the containment building (L) includes
other leakage pathways. Some of these leakages are from the auxiliary
building, radwaste building, spent fuel pit, and from various tanks
which are collected and routed back to the liquid radwaste system or
which contribute as minor gaseous release pathways. These leakages
can be accounted for by either adjusting parameters, such as the
shinbleed rate (which may also be routed from the primary system
ultimately to the radwaste system), or by increasing the leakage to
the containment building. Also, the components of the liquid radwaste
system may be modeled by using only the reactor makeup water storage
tank (RMWST) volume, which herein is considered to include the liquid
radwaste system volumes and the actual RMWST volume. This assumption
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is realistic since liquid radwaste system components do not
selectively remove tritium from the liquid stream. Another option is
the method of indicating the percentage of tritium recycle employed.
This percentage can be expressed either as the fraction of shimbleed
flow, primary coolant leakage, makeup flow from the RMWST to the
primary coolant system, or any combination of the three. It should
also be noted that the computer program used for the parametric
analysis of tritium buildup includes a provision for limiting the
primary coolant concentration to a predetermined level by increasing
the intentional removal (R) (see table 2) , as necessary. R simply
indicates some type of planned or intentional removal of primary
coolant liquids from the primary coolant system (or other systems with
high tritium concentrations) in order to maintain the tritium
concentration within the technical specifications of the reactor.
This removal may be some type of selective removal of tritium from the
liquids or removal of large volumes of tritiated liquids for disposal.
The limiting primary coolant tritium concentration that was chosen
was 2.5MCi/ml. This value, or one within several tenths of 1 jiCi/ml,
was selected because of concerns with potential occupational exposures
in the reactor building atmosphere and spent fuel pit areas during
operation and refueling*»2,3. Based on assumed values of primary
coolant leakage rates into the closed containment building, this
concentration limits the buildup of tritium in the containment
building atmosphere to levels previously indicated to be acceptable by
utilities. Many of the recent PWR's are being designed with a
continuous purge of the containment building atmosphere, thus the
limiting primary coolant concentration may be increased significantly
(by at the least a factor of 2), since the equilibrium containment
concentrations of tritium would be substantially lower. For these
PWR's, many of the results and conclusions presented in this paper
would be altered in a manner which is somewhat proportional to the
limiting coolant concentration used. The calculated removals and
environmental releases for several higher limiting primary coolant
concentrations are presented in table 1 and the results are discussed
in section IV.D.
The following sections discuss (1) tritium buildup in each of the
compartments during routine operation and shutdown and (2) the
relationship between compartments.
A. Routine Operations
1. Compartment 1 - Primary Coolant System
Tritium in the primary coolant system can be transported from the
system by leakage into the containment building and the auxiliary
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building. The tritium that remains in the water phase of these
leakages is drained to the liquid radwaste system. The tritium that
is in the gaseous phase is eventually released through the containment
building ventilation system. A small side flow of primary coolant
(shitnbleed) may be processed through the liquid radwaste system and
collected in the RMWST. Primary coolant makeup is then returned to
the primary system from the RMWST. Other losses from the primary
system include primary to secondary leakage, evaporation of tritium
from the refueling canal during refueling shutdown, radioactive decay,
and intentional removal (discussed previously) .
2. Compartment 2 - Liquid Radwaste System and RMWST
As indicated previously relative to tritium, the liquid radwaste
system can be modeled as simply a tank which includes all the liquid
radwaste system tanks and components and the reactor makeup water
storage tank (RMWST). The additions to this system come from
shimbleed from the primary coolant system, the collection of
containment building liquids contaminated by primary coolant leakage,
or via routing of steam generator blowdown liquids to this system.
Losses from this system can result from controlled discharges to the
receiving waters of the plant, makeup to the primary coolant system,
radioactive decay, and continuous makeup to the spent fuel pit. As
indicated tritium can be transported to the liquid radwaste system by
the routing of steam generator blowdown liquids; however, in the base
case analyzed in this paper, it is assumed that the blowdown is routed
instead to the main condenser. Another option, which is analyzed in
section IV. C. 1, is to route the blowdown directly to the
environment.
3. Compartment 3 - Sjaent Fuel Pit
During routine operations tritium is added to the spent fuel pit
by makeup from the RMWST. Removal of tritium from the reservoir is
either by radioactive decay or evaporation. During refueling shutdown
a portion of the spent fuel pit water becomes mixed with the primary
coolant system in the refueling canal. As previously indicated,
during refueling shutdown tritium is removed from the refueling canal
by evaporation and decay.
<4. Compartment 4 - Refueling Water Storage Tank
Tritium is transported to the refueling water storage tanks
(RWSI1) (often called condensate storage tanks) mainly as a result of
refueling operations. During refueling the refueling canal is flooded
with a large volume of water from the RWST. At the end of refueling
operations, this water, which is now contaminated with tritium, is
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returned to the RWST. The only loss from the RWST during routine
reactor operation is radioactive decay.
5. Compartment 5 and 6 - Secondary System Steam Generators and
the Remaining Secondary System
During routine operation tritium may leak or diffuse from the
primary coolant system into the steam generators of the secondary
system. From the steam generators a small fraction of the tritium is
carried over in the steam to the turbines and returned to the steam
generators in the feedwater. If the steam generators are blowndown,
tritium can be transported to the environment, the RMWST (via the
radwaste system), or the main condenser. Some of the tritium which
accumulates in the remainder of the secondary system is released to
the environment by steam leakage, steam jet air ejector exhaust
discharges, and condensate leakage to the turbine building.
B. Shutdown
During refueling shutdown, a fraction of the primary coolant
system is drained to the liquid radwaste system for processing and
then retained in the RMMST for the duration of refueling operations.
The remaining primary coolant water in the refueling canal is then
mixed with a large volume of water from the refueling water storage
tank (RWST) and a small fraction of the water from the spent fuel pit.
Tritium losses from the reactor during refueling operations are from
radiDactive decay, evaporation from the spent fuel pit, and
evaporation from the refueling canal in the containment building.
Figure 3 illustrates the losses and liquid volumes in the compartments
during refueling. (Note that the two compartments of the secondary
system are not disturbed during* a refueling shutdown.)
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111. MATHEMATICAL MODEL DF ACTIVITY BUILDUP
Figure H illustrates mathematically the transport to and from each
of the six compartments previously discussed. The six compartments
can be mathematically described by the following differential
equations.
1) dAl(t)/dt = P * N*Al(t) + Q*A2(t)
2) dA2(t)/dt = T*Al(t) «• V*A2(t) + H*ASG (t)
2)
3)
dA2(t)/dt = T*Al(t)
dA3(t)/dt = U*A2(t) * W*A3 (tj
dAU(t)/dt = -LAMBDA*AU
-3 •• r*^^ « • k «_9*_ . ^^^.^ ^** •*_ m ^ « «
m QAH jt;/at: = -LiannuA'aH |tj
5) dASG(t)/dt = LL*AS*Al(t) + G*ASG(t) * C*ASEC (t)
6) dASEC(t)/dt = (1-LL) *AS*A1 (t) * D*ASG (t) * F*ASEC(t)
The left side of each equation is the time rate of change of
activity in that particular compartment. The right side of each
equation is composed of additions and removals of tritium for each of
the compartments. All terms are in units of Curies/month. The terms
of the right side of these equations are:
Equation 1
N*Al(t)
Equation 2
Q*A2(t)
T*Al(t)
V*A2(t)
Production of tritium in
compartment 1 (primary coolant
system) plus release of tritium into
the primary system from fuel and
control rods
Radioactive decay; intentional
removal; primary system leakage to
the containment building, auxiliary
building, and secondary system; and
shimbleed for compartment 1
Reactor water makeup from
compartment 2 (RMWST) and liquid
radioactive waste system
Addition of tritium to compartment 2
via leakage collected in building
drains and primary system shimbleed
Radioactive decay in compartment 2;
reactor water makeup to compartment
1 from compartment 2; water makeup
to compartment 3 (spent fuel pit)
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Equation 3
Equation <*
Equation 5
Equation 6
H*ASG (t)
U*A2(t)
W*A3(t)
-LAMBDA*A4(t)
LL*AS*Al(t)
G*ASG(t)
Blowdown from compartment ASG (steam
generators) if blowdown is routed to
compartment 2
Water makeup to compartment
compartment 2
3 from
C*ASEC(t)
Radioactive decay and evaporation
from compartment 3
Radioactive decay in compartment U
(refueling water storage tank)
Primary to secondary leakage and
diffusion of tritium into the steam
generator liquids
Steam generator blowdown; carryover
of tritium by the production of
steam in the steam generators;
radioactive decay in the steam
generators
Feedwater to the steam generators
from compartment SEC (the remainder
of the secondary system)
(I-LL)*AS*A1(t)Primary to secondary leakage and
diffusion of tritium to the steam
phase of the steam generators
D*ASG(t) Steam generator carryover and
blowdown of tritium from the steam
generators to the remainder of the
secondary system
F*ASEC(t) Radioactive decay; steam leakage;
condensate leakage; air ejector
partitioning; steam generator
feedwater from the remainder of the
secondary system
The definitions of the components of each term are presented in
Table 2. Each of these equations has been solved for the activity in
each compartment as a function of time*. The tritium releases from
each compartment can be calculated by integrating the activities to
obtain an average activity during an interval of time. The average
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activities are then multiplied by the appropriate removal fractions to
obtain tritium releases for the time interval. Description of the
computerized models of the six compartments is not presented in this
paper,.
IV. PARAMETER SELECTION
The computer program that utilizes the equations in section III
requires the input of 33 parameters. The values of the parameters
selected for the base case reactor system are presented in table 3.
For the computer analysis, these parameters are held constant
throughout the 40 years of operation of a reactor. Naturally, for an
actual reactor system nany of these parameters will vary considerably
fron year-to-year or even month-to-month. Thus, the constant values
chosen for the parameters are considered the expected long-term
average over the life of the reactor. The following paragraphs
indicate the basis of these values, and briefly the significance of
some of them. The symbols and values of the parameters correspond to
the listing presented in table 3.
A- Trojan FSAR Data. Chap_ter 11
The following parameter values were selected from the Trojan
Nuclear Plant FSAR«: V1,V2,V3,V4,V1REF,LIMIT,FSFP,ELSPO,ELSFPR, and
ELRCR. For other PWRs the four volumes selected (V1,V2,V3,V4) may
vary considerably; however, for almost all large PWRs, the total
volume of these systems will be around 1,000,000 gallons. The
selection of the volumes of these systems is not significant to the
overall results presented in this paper. The fraction of mixing of
the spent fuel pit water (FSFP) with refueling water during refueling
shutdown for any reactor will be a small fraction of the spent fuel
pit water available (V3) and is not a critical parameter. The three
evaporation rates, ELSPO, ELRCR, and ELSFPR are extremely critical to
the results of this paper and to the tritium recycling operations of a
PWR. For a sensitivity analysis of these three parameters to be
worthwhile, a better data base is required. It can be expected that
the water temperatures of these volumes are representative of
operating PWRs. However, since evaporation rates are very sensitive
to changes in water temperature, small changes in the water
temperatures can significantly affect the evaporation rate of tritium.
In actual operation we would not expect the tritium evaporation rates
to decrease significantly from the values presented in table 3;
however, the rates can be increased by approximately factors of 2-U by
increasing water temperatures 30-40°F. As illustrated in Figure 10 of
this paper, evaporation rate increases of a factor of 2-4 would
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significantly affect environmental releases and intentional removal
rates during the lifetime of the reactor.
B. PtfR-Gale Code NUREG-0017
The Nuclear Regulatory Commission has issued NUREG-0017* which
discusses the parameter values used for LPS (primary to secondary
leakage), LST (steam leak), LCOND (condensate leak), FCO (Carryover
fraction in the steam generator) and LL (indicator that primary to
secondary leakage is into the water phase of the steam generator). We
assumed that the primary to secondary leak rate includes diffusion of
tritium through the steam generator walls. If a large total transfer
of tritium to the secondary system is assumed, its effect upon tritium
buildup and release at the reactor can be ascertained generally from
figure 10 of this paper. Variations in the parameter are of less
consequence than variations in water evaporation rates that were
discussed in section IV.A. We also assumed that the carryover of
tritium in a steam generator (FCO) is the same as the iodine carryover
presented in NUREG-0017. The moisture carryover of PWR steam
generators is much lower (about 0.1%) and may be more representative
of tritium carryover. This parameter requires some verification;
however, this parameter (FCO) has very little effect upon tritium
buildup or release at a PWR.
C. Other Parameters
A number of parameter values in table 3 indicate the options
chosen for operating the reactor. First it was decided not to include
planned releases of tritium from the liquid radioactive waste system,
i.e., this reactor is considered as a total tritium recycle plant.
Thus FRM, FRC and FS are 1.0. The parameter values for FBV, BLDN,
INDBN, and RR specify the mode of operation of steam generator
blowdown. Variations of steam generator blowdown options are
discussed later in this paper. The secondary system flow rates and
liquid masses (STR,MS,MSG) were selected based on reasonable values
from currently designed large PWRs. Variations in these parameters do
not have a significant effect on the results of this paper. One of
the secondary system parameters, the tritium partition factor for the
air ejector (PAG), was estimated from available information in PWR
safety analysis reports; however, any reasonable value chosen should
not significantly affect the results of this paper. The years of
operation (YROP), the range indicator (RANGE), and the volume of
refueling water used during refueling (V4REF) were chosen arbitrarily.
If YROP is chosen to be less than 40 years, results of interest can be
extracted from the tables and figures presented. It is assumed that
almost all the refueling water available (V4REF) would be used during
a refueling shut down. The shimbleed (S) rate that was chosen is a
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reasonable value for large PWRs. It could vary by a factor of two or
more; however, its value does not significantly affect the results of
this paper. The containment building leak rate (L) and fraction of
this leakage that remains in the liquid phase (FL) (and thus, is
returned to the liquid radioactive waste system) were chosen
somewhat arbitrarily. Values of FL were not available; however, in
choosing the value of 0.5 it is assumed that about half of the leakage
will be hot (thus, being in a vapor phase) and about half will be
relatively cold (thus, remaining as a liquid). The value selected for
L is higher than the value presented in NUREG-0017. The primary
coolant leakages from systems in the containment building and
auxiliary buildings were assumed to be included. Any leakage
remaining as a liquid was assumed to be eventually returned to the
liquid radioactive waste system. More information concerning these
two parameters (L and FL) is necessary, since they can affect the
results of this paper significantly.
D. Conclusions of Parameter Values
The major conclusions concerning parameter selection are that more
information is necessary for the parameter values of (1) evaporation
rates (ELSPO,ELSFPR,ELRCR), (2) the transfer of tritium to the
secondary system (LPS), and (3) the primary system leakage to the
containment and auxiliary building (L) and the fraction of this
leakage remaining in the liquid phase (FL).
The values of these parameters significantly affect the results of
this paper. The values chosen for this paper are reasonable, and the
results of this paper may be generally applied to a large PWR.
However, for a tritium recycle analysis of a specific PWR, these
parameter values should be derived for the specific system of that
PWR.
E. Ranges of Parameter
The expected ranges for the parameters in table 3 are:
1) V1+V2+V3+V4 800,000 - 1,250,000 gallons
2) V1REF 30,000 - 40,000 gallons
3) V4REF 200,000 - 350,000 gallons
U) YROP 30 - UO years
5) L UNKNOWN
6) LIMIT 1.0 - 3.0MCi/ml applicable NRC Regulatory
Guide for the majority of PWRs. 5.0
-7.5pCi/ml is applicable for continuous
purge systems in containment buildings.
7) S 0.7 - 3.0 gpm
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8) FSFP 0.10 - .20
9) ELSPO 25 - 150 gallons/day (estimate)
10) ELSFPR 75 - 300 gallons/day (estimate)
11) ELRCR 300.- 1500 gallons/day (estimate)
12) LPS 0-50 gallons/day (estimate of leakage
rate which also compensates for
diffusion)
13) FL 0.5 - 1.0
14) BLDN 0 - 350,000 Ib/hr
15) LST 340 - 1700 Ib/hr
16) LCOND UNKNOWN
17) MS 1,500,000 - 5,000,000 Ib
18) MSG 400,000 - 500,000 Ib. total for all stean
generators
19) FBV 0 - .33
20) PAJ UNKNOWN
21) STR 1,300,000 - 1,700,000 Ib/hr
22) FCO 0.0008 - 0.002 or (moisture carryover)
0.01 - 0.10 (iodine carryover)
V. ANALYSIS AND RESULTS
The Trojan reactor components and operational characteristics were
chosen as the base case for a parametric analysis of the buildup and
release of tritium at a large PWR. Table 3 presents the values of the
parameters (which are defined in table 2) for the base case. Each
analysis is based on the values presented in table 3 unless a
different parameter is specifically indicated in the discussion of the
analysis. The parametric study can be divided into three groups which
are discussed in the following sections.
A. Annual Tritium Source Term
Four annual tritium source terms (350; 700; 1,000; 1,400 curies)
were arbitrarily chosen for analysis using the base case parameters
presented in table 3. Figures 5 and 6 indicate the tritium buildup in
the primary coolant system and the total reactor plant over a 40 year
time period. In figure 5 the primary coolant activities represented
by curves 2, 3, and 4 reach their limiting concentration (which
corresponds to about 2.5 jiCi/ml concentration) within approximately 12
years. This limiting concentration (about 850 curies in the primary
coolant system for the base case) for the primary coolant system is
controlled by the intentional removal of primary coolant liquids.
These removed liquids can either be solidified and disposed of at a
waste burial ground or treated for selective removal of tritium, if a
system is available, and the tritium then disposed of as a solid. The
increase in tritium activity in the primary coolant system is a slow
10
-------
procass and is also greatly influenced by the activity levels in
compartment 2 (RMWST and liquid radioactive waste system). Thus,
intentional removal of tritium can be accomplished effectively from
either the primary system, the RMWST, or the liquid radioactive waste
system on a continuous or batch basis. This flexibility allows
selective removal techniques to be utilized that have low flow rates
but which can operate continuously for long periods of time. Curve 1
indicates that a production term of 350 Ci/yr would not result in the
limiting concentration during the normal lifetime of a reactor.
Figure 7 indicates the tritium removal rates (in curies or gallons of
tritiated liquids) that will be required for various tritium source
terns. Removal becomes necessary when the annual tritium source term
in the primary coolant system reaches 625 curies/year. Also for each
incremental increase (above 625 curies) of 100 curies in the annual
source term ,. the total volume of tritiated liquids that must be
removed (over the lifetime of the reactor) from the primary coolant
system increases approximately 400,000 gallons.
As indicated in figure 8, during the first few years of reactor
operation, it is not necessary to intentionally remove tritium from
the primary coolant system. However, it should be noted that for
source terms larger than about 1,000 curies per year, removal does
become necessary within five years after initial operation of the
reactor.
B. Analysis of a Standard Annual Tritium Source Term of 700 Cj/year
For an analysis of various aspects of tritium in a PWR, a constant
annual source term of 700 Ci/yr was chosen. The selection of this
source term was somewhat arbitrary but appears reasonable for the
following reasons:
1) 700 Ci/year is consistent with tritium source terms of
operating reactors based on available data, as indicated in
table 4.
2) 700 Ci/year is a high enough source term so that removal from
the primary coolant system may be included in the analysis.
The activity buildup in each compartment and in the total reactor
plant is presented in figure 9. It should be noted that the primary
coolant system, secondary system, and reactor makeup water storage
tank (RMWST) reach a stable activity level much sooner than the spent
fuel pit or the refueling water storage tank. The secondary system
retains very little tritium and is not a concern at the end of the
plant's useful lifetime. The equilibrium level of tritium of this
reactor case is approximately 5,100 - 5,500 curies.
11
-------
Figure 10 presents the tritium losses and retention for the base
plant design for a 700 Ci/yr source term. The losses are divided into
three groups - radioactive decay, uncontrolled environmental losses,
and intentional removal. Decay and environmental losses combined
constitute more than 80X of the total tritium losses from the plant.
Each of these loss mechanisms are approximately equal in magnitude.
Note, the intentional removal could be discharged directly to the
environment thus significantly increasing the fraction of liquids
discharged. However, in this paper it is assumed that the liquids
that are intentionally removed from the primary coolant system are not
discharged to the environment.
The environmental losses are divided into three major categories -
evaporation (during normal operation from the spent fuel pit and
during shutdown from the spent fuel pit and the refueling canal),
leakage from the containment building, and secondary system losses
(primarily from steam leakage and condensate leakage). About 80% of
the environmental losses are from evaporation and leakages to the
containment building with both pathways being approximately the same
over the 40 year life of the plant. It should also be noted that
intentional removal is assumed to be necessary after 11 years 9 months
of operation. The hatched area in figure 10 indicates the annual
increase in activity that is retained in the plant. After about 30
years, the plant activity is stabilized. Thus, block 3 in figure 10
is indicative of the losses of the system not only for the 40th year
but also for each year from the 30th year through the 40th year.
Figure 11 indicates the fraction of the environmental releases
that are gases or liquids. Over the 40 year lifetime of the reactor
almost 90% of the environmental tritium release is expected to be in
the gaseous phase. All of the tritium which is intentionally removed
during the 40 years of operation (2,780 curies) and the tritium
retained at the end of the 40 year period (5,460 curies) is in the
liquid phase. A determination of the ultimate fate of these liquids
is outside the scope of this paper.
12
-------
c' Effects of Secondary System parameters
1. Reactor Losses as a Function of Steam Generator Slowdown
Figure 12 indicates the effects of recycling steam generator
blowdovm liquids versus routing them directly to the environment.
The increase in the total environmental releases from the reactor
is less than 3% for direct release of blowdown. However, the
gaseous releases now decreased to about 80% of the total
environmental release for the cases with blowdown discharged to
the environment, versus 90% for the case where blowdown is
recycled. This results from the liquid releases almost doubling
while the gaseous releases are decreased by about 8%. For the
case of the blowdown routed to the environment, the rate of
blowdown has very little effect on the magnitude of the
environmental releases. Thus, blowdown rates of 5 gpm or even
several hundred gpm (average annual blowdown rates) will produce
similar results as those presented in Figure 12.
2- Reactor Losses as a Function of Primary to secondary Leak Rate
Figure 13 indicates the environmental losses and intentional
removcil necessary for a reactor with no primary to secondary leakage
and one with an 18.7 gpm leakage rate. The annual source term used
was 700 curies. The radioactive decay losses and the retention of
tritium in the reactor are similar for both primary-to-secondary
leakage cases (see figure 10). Several conclusions may be reached
concerning the results of zero primary-to-secondary leakage in
comparison to a reactor with 18.7 gpm leakage.
1) There are no liquid environmental releases; the gaseous
releases are 9% lower; and total environmental releases are
19% lower.
2]i Intentional removal becomes necessary almost 3 years sooner.
3)i The number of curies of tritium that must be intentionally
removed is 61% higher (1,480 curies versus 2,780 curies) and
the volumes of liquids that are required to be removed are
67% higher (500,000 gallons versus 300,000 gallons).
13
-------
D. Effects of Various Limiting Concentrations
Table 1 indicates the impacts of limiting the tritium
concentrations in the primary coolant system of a PWR. Two source
terms (700 Ci/yr and 1,400 Ci/yr) were chosen and three limiting
primary coolant concentrations (2.5, 5.0, and 7.5 pCi/ml) were
analyzed for these source terms. As discussed previously, for the
base case parameters chosen, intentional removal is necessary if the
sourca term is greater than 625 Ci/yr and the limiting concentration
is about 2.5 pCi/ml. For the two higher limiting concentrations
selected, intentional removal would not become necessary until a much
higher source term is reached. For these two cases the environmental
releases, primary coolant concentration, and total curies built up in
the reactor would stabilize primarily as a result of radioactive decay
in the plant and uncontrolled environmental losses. The significant
environmental losses are from the refueling canal evaporation during
refueling, evaporation from the spent fuel pit during refueling and
routine operation, primary system leakage into the containment
atmosphere with subsequent release to the environment, and release via
secondary system leakages.
For the 1,400 Ci/yr source term, intentional removal is necessary
for the 5 jiCi/ml limit but not for the 7.5 pCi/ml limit. Doubling the
source term from 700 Ci/yr to 1,400 Ci/yr and doubling the
concahtration limit from 2.5pCi/ml to 5.0MCi/ml results in
approximately the same quantity of primary coolant that must be
intentionally removed (300,000 gallons versus 250,000 gallons);
however, the curies of tritium intentionally removed from the primary
coolant system increases by almost a factor of two (2,785 curies to
4,525 curies). Also the initial time period after reactor startup for
commencing removal is fairly close (12 years 9 months versus 13 years
10 months).
A major conclusion from this overall analysis is that as the
limiting primary coolant concentration is increased there is a nearly
proportional decrease in intentional removal required, an increase in
uncontrolled environmental releases, and an increase in the buildup of
tritium in the reactor.
VI. SIGNIFICANCE OF RESULTS
The results presented in Section IV will provide the information
to make an assessment of the environmental and inplant impacts of
decisions related to the recycle of tritiated liquids in pressurized
water reactors. The data provided in this paper presents an estimate
of the reduction in discharges of tritium to the environment under
various conditions; the rate of buildup and the equilibrium level of
14
-------
tritium in the reactor plant liquids; and the amount of tritium which
will be discharged to the environment through plant ventilation
systems and from the secondary system leakages which may be considered
unavoidable.
Further, this paper indicates the time at which the limiting
conditions of operation will be attained and the amount of tritium
which would have to be intentionally removed in order to meet the
coolant specification limits for a given tritium source term. This
information will enable an assessment to be made of the rate at which
tritium must be removed from the coolant system and subsequently the
capacity of the tritium removal systems which must be employed to
achieve the desired ends. Because of the slow buildup of tritium in
the coolant system, the tritium may be controlled on a periodic, batch
basis during plant shutdowns. This will be particularly advantageous
since, if it is necessary to add a tritium control system, it will be
possible to provide this treatment capability independent of plant
operation. Thus, the tritium control system will not impact on the
plant reliability and the plant safety during operations.
In summary this paper, while it has not reached conclusions as to
the desirability of recycle of tritiated liquids nor the necessity for
provision of a tritium control system in a pressurized water reactor,
it has documented a realistic estimate of the inplant and
environmental release source terms associated with tritium recycle.
Froai this information the impact on plant personnel, plant operation,
and the population doses may be directly evaluated, and decisions may
be made regarding tritium control options.
15
-------
REFERENCES
1. tfilson, Robert L. and Jay Y. Lee. "Radioactive Liquid Waste
System for a Dry Site," for presentation at the ANS Annual
Meeting, Las Vegas, Nevada, June 19-22 (1972)
.
2. "Technical Paper Oconee Radiochemistry Survey Program," Summary of
four presentations presented to the American Nuclear Society, New
Orleans, Louisiana, June 8-13 (1975).
3. Final Safety Analysis Report for Sequoyah Nuclear Power Plant,
Units 1 and 2, NRC Docket Nos. 50-387 and 388, Appendix 11A.
4. Korn, G rani no A. Ph.D., and Theresa M. Korn, Mathematical Handbook
for Scientists and Engineers, Page 416, Sylvester's Theorem,
Second edition, McGraw Hill, 1968 (Library of Congress Catalog
Card Number 67-16304) .
5. Final Safety Analysis Report of the Trojan Nuclear Power Plant,
Docket No. 50-344.
6. Calculation of Releases of Radioactive Materials in Gaseous and
Liquid Effluents from Pressurized Water Reactors (PWR-Gale Code) ,
NUREG-0017, Office of Standards Development, U.S. Nuclear
Regulatory Commission, April, 1976.
7. Gruhlke, James M. , "Pressurized Water Reactor Effluent Discharge
Trends in the United States," U.S. Environmental Protection
Agency, Office of Radiation Programs, Technology Assessment
Division, Energy Systems Analysis Branch, Washington, D. C. ,
Presented at the ANS Annual Meeting, June 13-18, 1976, in Toronto,
Canada.
16
-------
Table 1. Effects of Various Limiting Primary Coolant Concentrations
Primary
Coolant
Limit
(uCI/ml)
2.5
5.0
7.5
2.5
5.0
7.5
Source
Term
Tritium
Buildup
After 40
years of
Operation
(Ci/yr) (Curies)
700
700
700
1,400
1,400
1,400
5,500
6,400
6,400
5,700
11,300
12,900
Intentional
Removal
Gallons for
40 years
300,000
0
0
2,900,000
250,000
0
Curies for
40 years
2,7850 )
0
0
28,220(2)
4,525(3)
0
After
Intentional
Removal
105
0
0
770
140
0
After 40 Years of Operation
(Curies/year)
Environmental
Releases
290
335
335
310
595
670
Radioactive
Decay
305
355
355
320
625
710
Intentional Removal Started After:
(1) 12 years 9 months
(2) 3 years 7 months
(3) 13 years 10 months
-------
Table 2. Parameter Identification
Parameter Definition
Al(t),A2(t),A3(t) Activity of each compartment as a
AU(t),ASG(t),ASEC(t) function of time, Ci
P Production and release of tritium into
compartment 1, Ci/month
N -LAMBDA-30.U*(L+R+S+LPS)/V1, fractional
removal per month from compartment 1 by
radioactive decay, leakage, planned
removal, shimbleed, and primary ' to
secondary leakage, month"1
Q 30.4*(L+R+S+LPS) /V2, fractional addition
per month to compartment 1 as a result of
makeup from compartment 2 to maintain a
proper liquid volume balance in
compartment 1, month-1
T 30.U*(L*FL*FRC + S*FS)/V1, fractional
addition per month to compartment 2 from
compartment 1 via shimbleed and primary
coolant leakage returned to RMWST,
month-*
V -LAMBDA-30.4*(L+R+S+LPS)/FRM*V2-
30.4*ELSPO/V2, fractional removal per
month from compartment 2 by radioactive
decay, makeup to compartment 1, and
makeup to compartment 3, month"1
U 30.U*ELSPO/V2, fractional makeup per
month to compartment 3 from compartment
2, month"1
W -LAMBDA-30.4*ELSPO/V3, fractional removal
per month from compartment 3 be
radioactive decay and evaporation,
month— J
18
-------
Table 2. (Continued)
Parameter Definition
H 30.4*2U*INDBDN*BLDN*(1-FBV)/MSG,
fractional addition per month to
compartment 2 from blowdown of the steam
generators, month-*
AS 30.U*LPS/V1, fractional addition per
month from compartment 1 to the steam
generator liquids or secondary system
Steam via primary to secondary leakage
and diffusion, month"1
G -30.4*24*BLDN/MSG-30.4*24*FCO*STR/MSG-
LAMBDA, fractional removal per month
from the steam generators by blowdown,
moisture carryover in the steam, and
radioactive decay, month-1
C 30.4*24*STR/MS, fractional addition per
month to the steam generators from the
remainder of the secondary system via
feedwater, month-*
D 30.4*24*FCO*STR/MSG + 30.4*24*RR*BLDN*(1-
FBVJMSG, fractional addition per month
from the steam generators to the
secondary system by moisture carryover
and blowdown routed to the main
condensers, month-1
F -LAMBDA-(LST+LCOND+STR/PAJ)*30.4*24/MS
-30.4*24*STR/MS, fractional removal per
month from the secondary system by
radioactive decay, steam leakage,
condensate leakage, air ejector
partitioning, and feedwater makeup to the
steam generators, month-1
V1,V2,V3,VU Volumes of compartments 1 through 4, gal.
19
-------
Table 2. (Continued)
Parameter
Definition
MSG,MS
S
LPS
ELSPO
FL
FS
FRM
FRC
FBV
FCD
BLDN
Mass of liquids in the steam
generators and the remainder of the
secondary system, Ibs.
Primary system leakage to the containment
building, gal/day
Shimbleed, gal/day
Primary to secondary leakage and
compensation for diffusion through steam
generator tubes, gal/day
Removal of tritium from the primary
coolant system in order to control the
primary coolant tritium concentration,
gal/day
Evaporative losses from the spent fuel
pit, gal/day
Fraction of L that remains in the liquid
phase
Fraction recycle of shimbleed
Fraction recycle of RMWST flow used as
makeup to the primary coolant system
Fraction of L*FL liquids that are
recycled
Fraction of steam generator
vented in the gaseous phase
blowdown
Fraction of tritium carried over from the
steam generators to the secondary system
steam
Steam generator blowdown rate, Ib/hr
20
-------
Table 2. (Continued)
Parameter
Definition
INDBDN
LST
LCDND
LL
PSJ
STR
RR
LAMBHft.
Indicator, equals 1 if BLDN is routed to
the; RMWST else it equals 0
Steam leakage from the secondary system,
Ib/hr
COndensate leakage from the secondary
system, Ib/hr
Indicator, equals 1 if LPS is into the
steam generator liquids, equals 0 if LPS
is: into the steam phase in the steam
generator
Partition factor of the steam jet air
ejector
Steaming rate of the steam generators,
Ib/hr
Indicator,, equals 1 if BLDN is routed to
the main condenser else it equals 0
Radioactive decay constant, month"1
21
-------
Table 3 Base Case Parameters
ro
ro
90000
210000
394000
350000
35000
340000
40
1
68
0
2.5
0.1
2920
0.15
48.4
131
464
18.7
1
0
1
8400
1700
2400
1
2000000
300000
1
0
200000
17000000
0.01
1
VI VOLUME PRIMARY SYSTEM, GAL
V2 VOLUME WATER RADWASTE AND COND STR, GAL
V3 VOLUME WATER SPENT FUEL PIT, GAL
V4 VOLUME WATER REFUELING STR POOL, GAL
V1REF VOLUME PRIMARY WATER IN CORE AT SHUTDN, GAL
V4REF VOLUME REFUELING WATER USED DURING REFUELING
YROP YEARS OF OPERATION
IND2 INDICATOR 0=MULTI-RUNS 1=TABLE, 2=PLOT, 3=ANN
L PRIMARY SYSTEM LEAKAGE GAL/DAY
INDBDN INDICATORS IF BLDN ROUTED TO RMWST ELSE=0
LIMIT TRITIUM CONC LIMIT IN PRIMARY SYS, UCI/ML
RANGE INDICATOR FOR CONTROLLING H-3 CONC, UCI/ML
S SHIMBLEED FLOW, GAL/DAY
FSFP FRACTION MIXING SPENT FUEL WATER W/REFUEL H20
ELSPO EVAP LOSS FROM SPENT FUEL PIT OPERATING, GAL/DAY
ELSFPR EVAP LOSS FROM SPENT FUEL SHUTDN, GAL/DAY
ELRCR EVAP LOSS FROM REFUEL CANAL REFUEL, GAL/DAY
LPS PRIMARY TO SECONDARY LEAKAGE, GAL/DAY
FS FRACTION RECYCLE SHIMBLEED
FL FRACTION PRIM COOL LEAK IN LIOUID
FRM FRACTION RMWST RECYCLED
BLDN SLOWDOWN RATE, LB/HR
LST STEAM LEAK, LB/HR
LCOND CONDENSATE LEAK SEC SYS, LB/HR
FRC FRACTION REACTOR DRAIN RECYCLED
MS MASS SECONDARY SYSTEM, tR
MSG MASS OF ALL STEAM GENERATORS, LB
LL INDICATORS , LPS INTO STEAM GEN LIQ
FBV FRACTION BLDN VENTED AS GAS
PAJ PARTITION FACTOR AIR EJECTOR
STR STEAMING RATE, LB/HR
FCO CARRYOVER FRACTION IN STEAM GENERATOR
RR INDICATORS , ROUTED TO CONDENSER ELSE=0
-------
Table 4. PWR Tritium Releases7
Year Tritium Releases (Ci/year)
1972 10U2
1973 715
1974 685
Note: 1) PWRs (zircaloy Clad Fuel) with at least one prior year of
commercial operation.
2) All release data is normalized to a 3400 MWth reactor
which operates at full power 80% of the year (292 days) .
(Note that the operating cases in this paper are for a
reactor operating at full power for 11 months of the year).
3) Releases per year are assumed to be the same as the source
term for the year since none of the reactors employ tritium
recycle.
23
-------
EQUIPMENT:
U
1
D
4^^
t
i
^^^
1
: i
ro
FIGURE 1 SCHEMATIC OF SYSTEMS
A. REACTOR & PRIMARY
COOLANT SYSTEM
B. STEAM GENERATOR
C. LETDOWN DEMIN-
ERALIZER
D. BORON THERMAL
REGENERATION SYSTEM
(WESTINGHOUSE DESIGN)
E. VOLUME CONTROL
TANK
F. SHIM BLEED HOLDUP
TANK
G. DEMINERALIZER(S)
H. BORIC ACID EVAPORATOR
I. REACTOR MAKEUP
WATER STORAGE TANK
(RMWST)
J. REACTOR COOLANT
DRAIN TANK(S) (RCDT)
K. SPENT FUEL PIT
L. REFUELING WATER
STORAGE TANK
M. TURBINE
N. CONDENSER
O. EVAPORATION
P. PRODUCTION, Cl/yr.
R. REMOVAL FOR TRITIUM
CONTROL
S, SHIM BLEED
F. SLOWDOWN (NOTE:
OPTIONS TO ROUTE: TO
ENVIRONMENT; MAIN CONDENSER
RADWASTE SYSTEM, OR RMWST
U. STEAM LEAK
V. CONDENSATE LEAK
W. PRIMARY TO SECONDARY LEAK
X. PRIMARY COOLANT LEAK
Y.RACTION OF PRIMARY COOLANT
LEAK LEAVING AS A GAS
Z. RADIOACTIVE DECAY
AA. DISCHARGE FROM RCDT
BB. DISCHARGE OF SHIMBLEED
CC. DISCHARGE FROM RMWST
-------
AND 6
PRODUCTION
IP)
DECAY
(LAMBDA)
PRIMARY COOLANT
SYSTEM
A1.V1
PRIMARY TO
SECONDARY
LEAKAGE
ILPS)
REMOVAL FOR
TRITIUM
CONTROL
(R)
EVAPORATION
(IFL'L)
ro
in
OPTION 1
DISCHARGE
(01)
DECAY
(LAMBDA)
C (SEE NOTE)
OPTION 2
DISCHARGE
(D2)
OPTION 3
DISCHARGE
(D3)
O
Ul
>2°-
ui
STEAM
GENERATORS
AND REMAINING
SECONDARY SYSTEM
r
REACTOR
MAKEUP
WATER
STORAGE
TANK (RMWST)
A2. V2
I
MAKEUP FOR
SPENT FUEL
PIT EVAPORATION
(ELSPO)
DECAY (LAMBDA)
STEAM (LST)
LEAK
CONDENSATE
LEAK
(LCOND)
j_ DISCHARGE
REFUELING
WATER
STORAGE
TANK
A4. V4
DECAY
(LAMBDA)
SPENT
FUEL
PIT
A3. V3
DECAY
(LAMBDA)
EVAPORATION
(ELSPO)
NOTE: IN MODELING THE LEAKAGE AND SHIM&LEED
THIS ROUTING OPTION IS ELIMINATED SINCE A
SIMILAR CONFIGURATION IS POSSIBLE VIA
MANIPULATION OF DISCHARGE OPTIONS.
FIGURE 2 BLOCK DIAGRAM OF SYSTEMS DURING ROUTINE OPERATION
-------
ro
cr
>•
<
O
en
2
5
UJ
p
REFUELING
WATER
STORAGE
E TANK
o
UJ
o
REACTOR
MAKEUP
WATER
STORAGE
TANK (RMWST)
V2
o
OL
O
o.
<
UJ
U
UJ
O
cc.
O
a.
SPENT
FUEL
PIT
V3
V3 -FSFP'V3
IUVUU VOLUME IN PRIMARY COOLANT
SYSTEM AND REFUELING CANAL
DURING REFUELING
YS/SA VOLUME STORED IN RMWST
DURING REFUELING
» \ VOLUME OF SPENT FUEL PIT
WATER NOT MIXED WITH PRIMARY
COOLANT
I } VOLUME REMAINING
IN REFUELING WATER
STORAGE TANK
DISCHARGE FROM THE REACTOR
DURING REFUELING
© LAMBDA'AI, A2, A3. A4 (RADIOACTIVE DECAY)
(2) EVAPORATION FROM REFUELING
CANAL (ELRCR) AND FROM
SPENT FUEL PIT (ELSFPR)
FSFP FRACTION OF
SPENT FUEL PIT
WATER MIXING
WITH PRIMARY
COOLANT SYSTEM
WATER
EVAPORATION IN SPENT
FUEL PIT
30.4*ELSFPR''A3*n-FSFP)
V3FSFP«V3
EVAPORATION IN REFUELING CANAL
SCU'ELRCRMAI'VIREF/Vl + A3"FSFP + A4*V4REF/V4)
VI +FSFP*V3 + V4REF
FIGURE 3 REFUELING VOLUMES
-------
COMPARTMENT i
COMPARTMENT 2
COMPARTMENT 3
' N'AII
P
t)
PRIMARY COOLANT
SYSTEM
Q
VA2(t)
A2(t)
REACTOR MAKEUP
WATER STORAGE
TANK (RMWST)
A2 M. V2
T*
VI (t)
\SG(t»
SPENT FUEL
PIT
A3 (1). V3
WA310
U*A2(t)
N—LAMBDA-p0.4*JL+R+S+LPS)/V1
O"30.4'(L+R+S+LPS)/V2
COMPARTMENT 4
REFUELING WATER
STORAGE TANK
A4(t). V4
-LAMBDA*A4(t)
V— LAMBDA-30.4*(L+R+S+LPS)/FRM*V2
-30.4'ELSPO/V2
T-30.4*«L'FL'FRC+S'FS)/V1
H-30.4*24*INOBDN*BLDN>(t-FBV)/MSG
COMPARTMENT 6 (SGI
W— LAMBDA-30.4* ELSPO/V3
U-30.4*ELSPO/>V2
COMPARTMENT 6 (SEC)
LL'AS'A
G'ASG(t)
STEAM GENERATORS
ASG(t). MSG
(t)
C
'ASEC(t)
SECONDARY SYSTEM
ASEC(t). MS
(1-LLI
D-A
AS'AI(t)
SG(t)
* \
F*ASEC(t>
1
AS-30.4*LPS/V1
G--30.4*24*BLON/MSG
-30.4*24'FCO'STR/MSG-LAMBDA
C-30.4'24'STR/MS
0-30.4*24*FCO'STR/MSG
+30.4*24*RR*BLON'(1-FBVI/MSG
F—LAMBDA-(LST+LCOND+STR/PAJ)*30.4'24/MS
-3a4*24*STR/MS
FIGURE 4 MATHEMATICAL MODELS OF THE COMPARTMENTS
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ro
oo
CO
UJ
oc
=>
o
to
UJ
o
o
o
o
>-
cc
tc
a.
700 Ci/yr SOURCE TERM
z.
1400 Ci/yr SOURCE TERM
1000 Ci/yr SOURCE TE RM
350 Ci/yr SOURCE TERM
NOTE: PRIMARY COOLANT CONCENTRATION LIMITED TO 2.5 pC1/ml
I I I I I I I I III
I I I I I I I i
l I I I I I I 1
15 17 19 21 23
WAR OF OPERATION
35 37
FIGURE 5 PRIMARY COOLANT ACTIVITIES (PRIOR TO REFUELING)
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ro —
ANNUAL TRITIUM PRODUCTION
1400 Ci/yr
ANNUAL TRITIUM PRODUCTION
700 Cl/yr
ANNUAL TRITIUM PRODUCTION
350 Cl/yr
ANNUAL TRITIUM PRODUCTION
1000 Ci/yr
NOTE: PRIMARY COOLANT CONCENTRATION LIMITFO TO 2.5
I i i i i i
16 18 20 22
YEAR OF OPERATION
FIGURE 6 PLANT ACTIVITY VERSUS TIME
-------
5x10*
4x10*
3x10°
M
O
o
2X106
1x10*
5x10*
4x10*
3x10*
o
3
2x10*
1x10*
MOTE: PRIMARY COOLANT CONCENTRATION
LIMITED TO 2.5 yCi/ml
500
750
1000
1250
1500
1750
2000
2250
SOURCE TERM (P) CURIES/YEAR
FIGURE 7 INTENTIONAL REMOVAL VS. SOURCE TERM
30
-------
50
40
30
tn
cc
<
at
20
10
PRIMARY COOLANT CONCENTRATION
LIMITED TO ?.5
I
I
I
500
750
1000
1250
1500
1750
2000
2250
SOURCE TERM (P) CURIES/YEAR
FIGURE 8 TIME UNTIL REMOVAL BEGINS VS. SOURCE TERM
-------
CO
s
z
Ul
13
ii
Ul
1C
ST.
g
OC
Jb
to
Ul
QC
u
z
X
>
Li
<
5000
4500
4000
3500
3000
_ja
IM
1800
1700
1600
1500
1400
1300
1200
1100
1000
900
800
700
600
500
400
300
200
100
OTAL ACTIVITY IN REACTOR
REACTOR MAKEUP WATER STORAGE TANK (RMWST
& LIQUID RADWASTE SYSTEM
SPENT FUEL PIT
REFUELING WATER
PRIMARY COOLANT SYSTEM
40TH YEAR OF OPERATION
SECONDARY SYSTEM
(-94% IN STEAM GENERATORS)
6
12
14
24 26
16 18 20 22
YEAR OF OPE RATION
FIGURE 9 ACTIVITY BUILDUP IN EACH COMPARTMENT (700 Ci/yr SOURCE TERM)
30 32 34 36
38
NOTE: PRIMARY COOLANT CONCENTRATION LIMITED TO 2.5 uCi/ml
-------
BLOCK 1
ANNUAL AVERAGES OVER
A 40 YEAR PERIOD
BLOCK 2
40 YEAR TOTALS
DECAY
ENVIRONMENTAL
EVAPORATION
CONTAINMENT
SEC. SYSTEM
REMOVAL
RETENTION
Ci/yr
253
241
(101)
(91)
(49)
70*
137
GAL/YR
7500*
DECAY
ENVIRONMENTAL
EVAPORATION
CONTAINMENT
SEC. SYSTEM
REMOVAL
RETENTION
CURIES
10.120
9.640
(4.000)
(3,600)
(2.000)
2,780
5.460
FRACTION OF
GALLONS SOURCE TERM, %
300,000
1.044,000"
36
34
M4)
(13)
(7)
10
20
BLOCK 3
40TK YEAR OF OPERATION
Ci/yr %
DECAY
ENVIRONMENTAL
EVAPORATION
CONTAINMENT
SEC. SYSTEM
REMOVAL
RETENTION
305
136
99
55
105
0
44
19
14
8
15
0
•DURING THE YEARS OF
ACTUAL REMOVAL (12 THROUGH 40)
THE ANNUAL AVERAGES
ARE 100 Ci/yr AND 11,000 GAL/DAY
AND THE CONCENTRATION IS 2.46 p Ci/ml
•EXCLUDES SECONDARY
SYSTEM VOLUMES
WHICH CONTAIN VERY LITTLE TRITIUM
800
800
TOTAL SYSTEM LOSSES
ENVIRONMENTAL LOSSES + INTENTIONAL REMOVAL
RADIOACTIVE DECAY
95% OF TOTAL ENVIRONMENTAL LOSSES (UNCONTROLLED)
40TH YEAR EVAPORATION LOSSES
REMOVAL BEGINS ocVcAcr \ CONTAINMENT BUILDING LEAKAGES
AFTER 11 YEARS 9 MONTHS
INTENTIONAL REMOVAL OR DISCHARGE
. SECONDARY SYSTEM RELEASES
I I i i i i i i i i i i
16 18 20 22
YEARS OF OPERATION
FIGURE 10 TRITIUM LOSSES FROM BASE PLANT DESIGN (700 Ci/yr SOURCE TERM)
NOTE: PRIMARY COOLANT CONCENTRATION LIMITED TO 2.5 jj Ci/ml
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co
^>-
o
Ol
111
lit
cc
300
280
260
240
220
210
200
180
160
140
120
100
80
60
40
20
0
GASES
ENVIRONMENTAL RELEASES
GASES 8600 CURIES
LIQUIDS 1070 CURIES
GASES 215 Ci/yr
LIQUIDS 27 Ci/yr
PRIMARY COOLANT CONCENTRATION LIMITED TO 2.5 pCi/ml
LIQUIDS
I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
10 12 14 16 18 20 22
YEARS OF OPERATION
24
26
28 30
32
34
36 38
FIGURE 11 LIQUID AND GASEOUS ENVIRONMENTAL RELEASES (700 Ci/yr SOURCE TERM)
-------
0.0 SLOWDOWN
RECYCLE TO MAIN CONDENSER
40 YEAR RELEASES (TOTAL PLANT)
LIQUIDS = 1070 CURIES
GASES = 8600 CURIES
TOTAL = 9670 CURIES
8400 LB/HR SLOWDOWN
DIRECT RELEASE TO THE ENVIRONMENT
84.000 LB/MR SLOWDOWN
DIRECT RELEASE TO THE ENVIRONMENT
40 YEAR RELEASES (TOTAL PLANT)
LIQUIDS = 1996. CURIES
GASES = 7911 CURIES
TOTAL = 9907 CURIES
40 YEAR RELEASES (TOTAL PLANT)
LIQUIDS = 2000. CURIES
GASES = 7907. CURIES
TOTAL = 9907. CURIES
CO
in
O
in
ui
t/j
<
UJ
_J
UJ
cc
55.0
50.0
45.0
40.0
35.0
30.0
25.0
20.0
15.0
10.0
5.0
0
0.0 SLOWDOWN RECYCLE TO MAIN CONDENSER
-8400 LB/HR SLOWDOWN
DIRECT RELEASE TO ENVIRONMENT
84.000 LB/HR SLOWDOWN
IS ESSENTIALLY THE SAME
NOTF: PRIMARY COOLANT CONCENTRATION LIMITED TO 2.5 yd/ml
I I I I I I I I I I I I I I I I
I I I I I
I I I I I I i I I I I I
10
12 14
16
18
20
22
24
26 28
30
32 34
36
38
YEARS OF OPERATION
FIGURE 12 SECONDARY SYSTEM RELEASES VERSUS STEAM GENERATOR SLOWDOWN RATE (700 Ci/yr SOURCE TERM)
-------
40 YEAR TOTALS
LIQUID RELEASES
GASEOUS RE LEASES
TOTAL ENVIRONMENTAL
RELEASES
INTENTIONAL REMOVAL
RADIOACTIVE DECAY
ACTIVITY RETAINED
REMOVAL (LIQUID VOLUME)
REMOVAL BECAME
NECESSARY AT
0.0 LEAK
RATE
O.O CURIES
7800 CURIES
7800 CURIES
4480 CURIES
10,300 CURIES
5400 CURIES
500,000 GALLONS
8 YEARS, 11 MONTHS
18.7 gpm
LEAK RATE
1070 CURIES
8600 CURIES
9670 CURIES
2780 CURIES
10,120 CURIES
5460 CURIES
300,000 GALLONS
11 YEARS, 9 MONTHS
KEY
18.7 PRIMARY TO SECONDARY LEAK
0.0 gpm PRIMARY TO SECONDARY LEAK
300
en
250
TOTAL ENVIRONMENTAL RELEASES
o
cc.
O
o
LU
DC
5
O
cc
"- 100
CO
LU
200
150
O
50
INTENTIONAL
REMOVAL
NOTE: PRIMARY COOLANT CONCENTRATION LIMITED TO 2.5 pC1/ml
300
250
200
150
100
50
I I .'l I I I I
8 10 12 14 16 18 20 22 24
YEARS OF OPERATION
I I I I I I I I I I I I I I I I I I I I
26 28
30 32
34
36 38
FIGURE 13 EFFECT OF PRIMARY-TO-SECONDARY LEAKAGE (700 Ci/yr SOURCE TERM)
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