/

GA-A14S5S UC-77

A REVIEW OF FISSION PRODUCT PLATEOUT INVESTIGATIONS

AT GENERAL ATOMIC

a by

D. L HANSON

Prepared under Contract EY-76-C-03-0167

Project Agreement No. 17 for the San Francisco Operations Office

Department of Energy

DATE PUBLISHED: DECEMBER 1977

GEI ERAL ATOMIC COMPANY

NOTICE This report was prepared as an account of work sponsored by the United States Government.

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DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

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GA-A14555 UC-77

A REVIEW OF FISSION PRODUCT PLATEOUT INVESTIGATIONS

AT GENERAL ATOMIC

D. L

by

HANSON

- NOTICE -This report was prepared as an account of work sponsored by the United States Government Neither the United States nor the United States Department of Energy, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liabihty or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represenu thai its uK would not infringe pnvately owned nghts

Prepared under Contract EY-76-C-034)167

Project Agreement No. 17 for ttie San Francisco Operations Office

Department of Energy

GENERAL ATOMIC PROJECT 3224

DATE PUBUSHED: DECEMBER 1977

GENERAL ATOMIC COMPANY If

V

FOREWORD

Investigation of fission product plateout has been suggested as an

area for collaborative work under the DOE/BMFT Umbrella Agreement between

the U. S. A. and Germany. This status report on plateout work at GAC is

intended to provide background for planning the joint program.

iii

ABSTRACT

The status of fission product plateout studies at General Atomic

is reviewed and suggestions are offered for future work. The

deposition, or plateout, of condensible radionuclides in the primary

circuits of gas-cooled reactors affects shielding requirements,

maintenance procedures, and plant availability as well as representing

a significant radiological source and/or sink for certain hypothetical

accidents. Physical models and computer codes used to describe these

plateout phenomena for reactor analysis are presented along with their

limitations and possible refinements. The review includes portions of

the recent AIPA study which sought to quantify the effects of

uncertainties in input parameters on plateout code predictions.

Major emphasis is placed upon the design methods verification program

to assess the validity of plateout predictions by comparison of

calculated behavior with experimental transport data.

V

TABLE OF CONTENTS

FOREWORD iii

ABSTRACT v

1. INTRODUCTION 1

1.1 Plateout Under Normal Conditions 1

1.2 Plateout Under Accident Conditions 2

1.3 Analytical Modeling of Plateout 3

1.4 AIPA Assessment of Plateout 4

1.5 Design Methods Verification 4

2. PHYSICAL MODELS OF PLATEOUT PHENOMENA 7

2.1 PAD Code 8

2.2 PADLOC Code 9

2.3 Material Property Data 11

3. AIPA ASSESSMENT OF PLATEOUT 12

4. DESIGN VERIFICATION PROGRAMS 17

4.1 GA Deposition Loop 17

4.2 BMI Deposition Loop 21

4.3 Peach Bottom EOL Program 26

4.4 CPL-2 Test Program 35

4.5 Fort St. Vrain Surveillance 43

5. CONCLUSIONS 45

REFERENCES 47

APPENDIX A . .• A-1

vii

LIST OF FIGURES

1. Plateout distributions of Sr-90 13

2. Plateout distributions of Cs-137 14

3. Plateout distributions of 1-131 25

4. GA fission product deposition loop 18

5. PAD code simulation of loop No. 1 19

6. PAD code simulation of loop No. 3 20

7. BMI deposition loop schematic 23

8. Distribution of Cs-137 and Te-129 activity in BMI deposition

loop 24

9. Distribution of 1-131 activity in BMI deposition loop 25

10. Isometric of primary coolant system 27

11. Cross section of steam generator 29

12. Cesium profiles in Peach Bottom cold ducts from internal

scans 30

13. Cesium deposition profiles in cold duct 31

14. Axial Cs-137 distribution in steam generator 32

15. Plateout distribution of Cs-137 and Cs-134 in Peach Bottom

HTGR 34

16. Effect of surface sorptivity on cesium plateout distribution . . 36

17. CPL-2 loop schematic 37

18. Heat exchanger - recuperator 40

19. Comparison of calculated and observed Cs-137 plateout profiles

for tube B-31 from CPL 2/1 heat exchanger 41

20. Comparison of calculated and observed 1-131 plateout profiles

for tube B-31 from CPL 2/1 heat exchanger 42

viii

1. INTRODUCTION

The transport behavior of condensible radionuclides in the primary

coolant circuits of HTRs is a vital aspect of reactor analysis under both

normal and accident conditions. A key aspect of these phenomena is the

deposition, or plateout, of condensible species on the structural surfaces

of the circuit; such accumulation of radioactivity affects shielding re­

quirements, maintenance procedures, and plant availability as well as

representing a significant radiological source term for certain hypothe-

tized accidents. Under certain other accident scenarios (e.g., an unre­

stricted core heatup event), plateout can be a mitigating factor: fission

products released from the fuel may deposit on reactor internals thereby

reducing the amount released to the containment building and, ultimately,

to the environment. Despite their apparent importance, these plateout

phenomena are not well understood; consequently, there is an on-going

program at General Atomic to characterize these plateout effects.

Generally, the approach has been empirical with the primary objective

of developing engineering tools which can be used for reactor design and

analysis. Much of this work has been reported in piecemeal fashion over

the past decade in a series of papers, articles, and quarterly progress

reports; it is the purpose of this discussion to highlight past, present,

and future plateout work at GAC with particular emphasis given to on-going

experimental programs to verify current design methods.

1.1 Plateout Under Normal Conditions

The coated-particle ceramic fuel used in modern HTRs is highly reten­

tive of fission products. However, the vigorous efforts of fuel designers

1

notwithstanding, a small fraction of the fission products escapes from the

fuel and is introduced into the helium coolant. Naturally, the primary ob­

jective of fission product transport analysis is an accurate prediction of

the total core release of radioactivity under all credible operating modes.

Given the extreme complexity of the task, there is a tendency to consider

the job done when total core releases have been estimated. Indeed, for the

noble gases krypton and xenon, this is essentially the case; however,

for the condensible species (primarily metals, iodine, and tellurium),

the job has just begun. For the engineer designing steam generator shield­

ing or planning circulator maintenance procedures, knowledge of the total

core release is of little value. What he needs to know is the expected

contamination levels on a particular component, for example, the circulator.

In other words, the plateout distribution of condensible radioactivity is

required as well as the total amount released. While such concerns are

generic to all HTRs, they are especially signficant for the gas turbine

HTR wherein periodic maintenance of the turbomachinery is required. The

plateout levels will largely determine the feasibility of contact mainten­

ance; if remote procedures are necessary, plant operating costs will be

adversely affected.

1.2 Plateout Under Accident Conditions

Likewise, in analysis of hypothesized accident scenarios, there is a

distribution of transient responses (e.g., flow rates, temperature changes,

etc.) throughout the primary circuit which may result in re-distribution

of radioactivity. For example, the design basis depressurization accident

(DBDA) results in transient wall shear stresses in certain locations which

are temporarily higher than those prevailing under normal operation.

Experiments (Ref. 1) have shown that such transients can result in re-

entrainment or "liftoff" of portions of the deposited radioactivity. (While

a detailed discussion of liftoff phenomena is beyond the scope of this

document, plateout and liftoff are intimately related, and both ultimately

must be understood and modeled in reactor analysis.) Since there is

2

a distribution of both shear ratios and specific activity throughout the

circuit, a rigorous assessment of the total liftoff requires integration

over all circuit surfaces. In addition to this mechanical "liftoff" of

plateout activity, molecular desorption of radiologically important fission

products like iodine may also occur because of changing conditions in the

circuit.

In the above discussion, plateout was a radiological source, but in

other accidents plateout could be an inportant sink. Despite redundant

engineered safeguards, there is a remote possibility that a loss-of-forced-

circulation (LOFC) accident could precipitate an unrestricted core heatup.

Unless corrective action were taken, the core would eventually release

large quantities of radioactivity. While the decay heat would result in

very high fuel temperatures, other portions of reactor internals would

remain relatively cool and could serve as an important sink for fission

products volatilized from the core. Plateout within the prestressed

concrete reactor vessel (PCRV) could greatly reduce the radioactivity

that escaped into the containment and, subsequently, into the environment.

1.3 Analytical Modeling of Plateout

Clearly then, it is important to model plateout phenomena. Plateout

distribution calculations at GAC are currently performed with the PAD

code, a transient, one-dimensional mass transfer code (Ref. 2). PAD is

a useful tool for modeling normal plant operation, but it is unable to

handle the complex geometry and rapidly changing conditions of a postulated

LOFC event. Furthermore, the numerical solution techniques employed in

PAD are very inefficient. Therefore, a new plateout code PADLOC has

recently been developed at GAC (Ref. 3). Although the physical models

are analogous, PADLOC is able to treat much more complex geometries and

event histories projected for LOFC conditions. Furthermore, through use

of more sophisticated numerical techniques, PADLOC is orders of magnitude

more efficient than PAD. Both codes will be discussed in more detail in

Section 2.

3

In parallel with this code development work, laboratory studies have

been performed to generate the associated material property data which

serve as input to the codes. Typically, these differential experiments

are sorption measurements to quantify the ability of structural materials

to sorb volatile fission products. These data will also be reviewed in

Section 2.

1.4 AIPA Assessment of Plateout

For the sake of argument, assume that the physical models incorporated

in PAD and PADLOC are correct. There still persists large uncertainties

in the calculated plateout behavior because of large uncertainties in the

associated input data (core releases, material properties, plant perfor­

mance, etc.); the consequence of these uncertainties is increased risk

and/or excessive design conservatism with attendant economic penalties.

This general question regarding the impact of uncertainties has been

addressed probabilistically as part of the Accident Initiation and Progres­

sion Analysis (AIPA) program. The primary objective of the AIPA study

is to provide guidance for planning and funding HTGR safety research and

development. Another objective is to consider alternative design options

through use of risk assessment. While AIPA was far-ranging in scope

(attempting to include all credible events having significant radiological

consequences), some valuable insight into the impact of plateout phenomena

was developed (Refs. 4, 5, 6). Pertinent results will be reviewed in

Section 3.

1.5 Design Methods Verification

While the above analytical investigations are useful, they must be

supplemented by experimental verification programs to confirm the correct­

ness of the analytical predictions. In fact, such verification is legally re­

quired by 10CFR50 (Ref. 7) and ANSI N45.2.11 (Ref. 8). Efforts at GAC to veri

plateout distribution predictions began almost a decade ago with the con­

struction of a small out-of-pile deposition loop (Ref. 1). Five loop

experiments were performed to determine the deposition behavior of iodine.

4

strontium, and cesium; attempts to correlate these results with PAD code

predictions met with mixed success. Reference 1 also reports on related work

to model plateout distributions in the Battelle Memorial Institute out-of-

pile loop (Ref. 9) and in the Dragon HTR (Ref. 10). In 1975, a comprehen­

sive program was initiated to verify all aspects of fission product trans­

port under normal plant operation (Ref. 11); the major plateout work under

this DOE-sponsored program is analysis of the deposition profiles measured

in the four CPL-2 in-pile loop tests (performed by CEA in the Pegase

reactor in Cadarache, France). The CPL-2 tests were similar to other

fission product transport experiments in the Pegase facility; namely,

the Idylle 03 test (Ref. 12) sponsored by Dragon and the Saphir tests

(Ref. 13) sponsored by KFA.

Concurrent with this effort, extensive plateout studies are being

performed under the Peach Bottom End-of-Life Program which is jointly

sponsored by DOE and EPRI. After Peach Bottom was shut down for decom­

missioning, the distribution of plateout in the primary circuit was deter­

mined by a combination of in situ gamma scanning (Ref. 14) and radiochemi­

cal analysis of samples destructively removed from the steam generator

tube bundle and coolant ducts (Ref. 15). The measured cesium profiles

were in excellent agreement with those predicted with the PAD code (Ref. 16).

In addition to these on-going programs, a comprehensive fission product

surveillance program is under way at Fort St. Vrain. These operating data

will provide valuable insight into all facets of radionuclide transport.

The most important plateout data are likely to come from the plateout probes

designed to sample the primary coolant both up- and downstream of the steam

generators (Ref. 17). An iodine monitoring system has also been installed

to permit the determination of the circulating iodine inventory. Additional

plateout information will be obtained by gamma scanning and radioassay of

accessible components.

The above work related primarily to plateout under normal operating

conditions. Complementary tests have also been performed to study plateout

5

under LOFC conditions. The earliest work involved small-scale, integral

tests to mock-up the reactor internals under LOFC conditions; the results

are reported in Ref. 18 and reviewed in Ref. 5. In one type of test,

irradiated fuel was annealed in an induction furnace to temperatures up

to 2500 C. The fractional release from the mock-up as well as the plateout

distribution within it were determined by radiochemical analysis. In other

tests, a 1/12-scale model of the Fort St. Vrain PCRV was utilized. Since

the mock-ups included simulation of the liner cooling system, considerable

surface area operated at relatively low temperatures; hence, the measured

fractional releases were very low. Recently, a comprehensive test program

has been initiated under the DOE-sponsored HTGR Safety Program to extend

these early results (Ref. 19). The approach will be similar, but the

important variables (flow rates, temperatures, etc.) will be more system­

atically investigated. These tests will be used to verify the PADLOC code

for LOFC analysis.

The integral experiments described in the preceeding paragraphs are

in various stages of completion. Some of the more significant tests and

corresponding results will be highlighted in Section 4.

6

2. PHYSICAL MODELS OF PLATEOUT PHENOMENA

The transport and deposition of condensible activity from flowing helium

to fixed surfaces is essentially a convective mass transfer problem, and many

mathematical models of the process have been offered. Reference 1 gives

a partial summary. Usually, the deposition process is conceived as a two-

step process: gaseous diffusion to the wall and a wall effect, typically

an adsorption process. The latter step is necessary because experiments

have shown that under some circumstances, surfaces can have a limited capa­

city for certain fission product species. In general, either the gaseous

diffusion or the wall effect can be dominant.

Most theoreticians choose to approximate the gaseous diffusion process

by a film resistance analogous to that used for convective heat transfer;

however, a variety of models have been offered for the wall effect. The

earliest work (e.g., Refs. 9, 20, 21) favored a linearized wall coefficient

because it allowed for an analytical solution of the governing equations.

In the simplest of models, molecules are transported from the bulk stream

concentration through a resistance 1/k, where k is a mass transfer coeffi­

cient (analogous to a film heat transfer coefficient), to the interface con­

centration. From the interface, the molecules go to the surface through a

resistance 1/k , where k is a wall coefficient. This wall coefficient w w

describes the effects of imperfect retention at the wall: k^ equal to

infinity corresponds to a perfect sink condition, and k„ equal to zero

corresponds to no retention at the surface.

The more sophisticated models, including PAD, couple the gaseous dif­

fusion process as described by a mass transfer coefficient with an equili­

brium sorption process. The assumption is that the kinetics of adsorption

are rapid compared to the gaseous diffusion process. If other than Henrian

7

or Langmuir behavior is considered, the equations become non-linear thereby

requiring numerical solutions. Recently, Iniotakis (Ref. 22) has developed

an even more sophisticated model which allows for in-diffusion into the

bulk of the substrate as well as surface adsorption; in his model, the

adsorption model is Henrian, and the adsorption and in-diffusion processes

are not coupled so that an analytic solution is still possible.

Further refinements in modeling plateout phenomena are anticipated in

the future. An obvious extension of the present models would be to couple

the adsorption and in-diffusion processes. In addition, chemical reaction

between the depositing fission product and the surface may well have to be

included to model LOFC plateout, for it has been shown that volatile metal

iodides can form under these conditions (Ref. 23). The physical modeling

done by investigators in the field of heterogeneous catalysis (e.g., Ref. 24)

might be readily extended to account for simultaneous adsorption, in-diffusion,

and chemical reaction. A more difficult task would be to develop models

accounting for the effects of particulate matter (carbon, metallic oxide

scale, etc.) which may be present in the primary coolant of HTRs (Ref. 25)

on fission product transport. A suitable aerosol transport has been

developed at GAC (Ref. 26), but simultaneous molecular and aerosol transport

models have yet to be formulated.

Even more complex transport models might be conceived, but the approach

taken at GAC has been an empirical one. Theoretical model development has

evolved more rapidly than the experimental characterization of the physics

of plateout. Hence, for the present, the obviously simple PAD model will

be retained and tested against experimental data as they become available; if

PAD is unable to correlate these data, then additional sophistication will

be introduced as necessary,

2.1 PAD Code

The PAD code is described in detail in Ref. 2, and its essential fea­

tures are reviewed in Appendix A. Briefly, PAD is a transient, one-

dimensional mass transfer codej an explicit finite difference solution is

8

obtained for the coupled, non-linear differential equations describing the

conservation of mass with a convective boundary condition. The coolant and

surface concentrations are coupled by an equilibrium adsorption process;

either Langmuir or Freundlich (concentration-dependent) isotherms may be

employed to describe the sorption process. The model allows for production

by precursor decay and treats recirculation in a closed loop. The code

has three options for treating deposition: (1) no sorption (e.g., a

non-adsorbing noble gas); (2) no desorption (the surface is a perfect sink

or, more precisely, the vapor pressure over the surface is zero for all

surface concentrations); and (3) desorption (an adsorption isotherm is

employed such that at a given surface temperature and partial pressure, an

equilibrium surface concentration exists beyond which no further accumula­

tion occurs).

The original version of PAD (Ref. 2) did not consider variable cool­

ant pressures which occur in a gas turbine HTR. This original version

also encountered numerical stability problems when large desorption pres­

sures were calculated. Both of these limitations were removed in the course

of a theoretical investigation of plateout in a gas turbine HTGR (Ref. 27).

The modified code is referred to as PADGT.

2.2 PADLOC Code

The PADLOC code (Ref. 3) is essentially an extended and improved

version of the PAD code. The motivating force behind its development was

the need for an engineering tool to model plateout under LOFC conditions.

Virtually any geometry can be simulated by an arbitrary network of branches

and nodes; PAD is restricted to a network of 12 branches in series. The

other major restriction with PAD is its numerical inefficiency because of

the explicit time integration method employed. PADLOC develops a set of

coupled, non-linear partial differential equations, a pair for each branch

(one for coolant concentration, the other for surface concentration); the

equations are linearized and solved by matrix techniques, and then the

non-linearities are accounted for by iteration. The resultant code is

orders of magnitude more efficient than PAD.

9

PADLOC can model very complicated time histories. The major input

variables (sources, flow rates, temperatures, pressure, etc.) are all time-1

and space-dependent. Transport under both forced and free convection are

modeled; correlations for predicting mass transfer coefficients for a

variety of geometries are automatically selected depending upon the flow

regime. Closed branches of the network can be broken at some specified

point in time (e.g., to simulate a pipe rupture or the opening of a relief

valve). Mass flow rates in any branch can not only vary in magnitude but

may also undergo multiple changes in direction. The cross-sectional area

along any branch may vary linearly between specified values at the branch

endpoints.

Sorption behavior is also specified by branch. As with PAD, options

include no sorption, perfect sink, and isotherm-controlled sorption. Three

types of sorption isotherms can be specified: generalized linear, Langmuir,

and Freundlich. The generalized linear isotherm is to be used for conser­

vative estimates of sorptive capabilities in the absence of experimental

data: it is assumed that the desorption pressure of the sorbed species is

equal to the fractional coverage (the actual concentration divided by the

saturation, or monolayer, concentration) times the vapor pressure of the

pure species at that temperature; the monolayer concentration will be esti­

mated from the molecular dimensions of the adsorbate. This simple model is

analogous to Raoult's law for ideal solutions.

The present version of PADLOC does have several limitations. First,

it only treats a single species; PAD can treat simultaneously five species

including decay chains. The code also does not consider in-diffusion

and/or chemical reaction. The present numerical technique is sufficiently

versatile that it could be extended to the solution of more complicated

differential equations incorporating these additional effects; however,

considerable reprograming would be involved. Further code developments

will await the findings of the experimental programs to characterize fission

product transport. It is emphasized that the objective is development of

engineering tools with demonstrated capabilities for reactor analysis;

empirical models that are physically simple but that correlate the data may

fulfill the need. i

10

2.3 Material Property Data

The material property data that serve as input to the present plateout

codes are gaseous diffusivities and sorption isotherms. The gaseous diffu­

sivities are used in the estimation of mass transfer coefficients; the

reference diffusivities used for reactor design are derived primarily from

analysis of diffusion tubes from the GAIL IV in-pile loop (Ref. 28).

Gaseous diffusivities along with fluid properties are employed in empirical

correlations to calculate convective mass transfer coefficients. PAD

uses the correlation recommended by Treybal (Ref. 29); PADLOC contains a

library of correlations for mass transfer under forced and free convection

for a variety of geometries. Sorption isotherms are generated in differ­

ential laboratory experiments using pseudo-isopiestic and Knudsen cell

techniques (Ref. 30).

The sorption data base is limited. Reference 4 reviewed and synthe­

sized the available data for the sorption of iodine, cesium, and strontium

on graphite and of iodine and cesium on metals; other condensible species

are assumed to exhibit "perfect sink" behavior when calculating their plate­

out distributions. The sorptivities of metal are especially ill-defined;

however, recent publications have added to the understanding: ORNL studies

of iodine on chromium oxide and magnetite (Ref. 31), and GAC studies of

cesium on Incoloy 800 (Ref. 32). Further iodine work at GAC is planned

under the ERDA Safety Program (Ref. 19), but extensive additional efforts

will be required before the sorptive capacities of pertinent metals for

important fission products are characterized.

11

3. AIPA ASSESSMENT OF PLATEOUT

The uncertainties in the plateout distributions of key radionuclides

under normal operating conditions were evaluated in AIPA, Vol. 5 (Ref. 4);

the impact of these uncertainties on occupational exposure rates was ad­

dressed in AIPA, Vol. 7 (Ref. 6).

The 40-year plateout distributions of Sr-90, Ag-llOm, Te-131, 1-131,

Cs-137, La-140, and Ce-144 were considered. The study was performed spe­

cifically for the 3000-MW(t) Fulton-generation HTGRs (Ref. 33), but the

qualitative results are considered broadly applicable to all thermal HTRs.

The effects of uncertainties in core gaseous release, core metallic release,

gaseous diffusivity, mass transfer coefficient, graphite sorptivity, and

metal sorptivity were determined. Parameter studies were performed with

the PAD code to determine the sensitivity of the plateout distributions

to changes in the above variables. Following a logarithmic transform,

the total variance in specific activity (Ci/cm^) was calculated by the pro­

pagation of errors method. Finally, one-sided upper 95% confidence state­

ments were made about the specific activity (Ci/cm^) at local points through­

out the primary circuit. Typical results are shown in Figs. 1 - 3 : median

and upper 95% distributions as well as the "expected" and "design" distribu­

tions are given for Sr-90, Cs-137, and 1-131.

The largest uncertainties are in the distributions of cesium and iodine.

For cesium, the large uncertainty arises from approximately equal contri­

butions from core metallic release and metal sorptivityj for iodine, the

metal sorptivity dominates. The uncertainty in strontium and silver plate­

out distributions results from uncertainties in the core metallic release

and in the mass transfer process. Cerium and lanthanum uncertainty can

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be attributed to core release and to mass transfer while the latter domi­

nates for tellurium.

While these results are considered at least qualitatively correct,

i.e., most of the important variables have been identified, they do

suffer a number of limitations. Probably, the most insidious flaw is

that the results are biased by the models and assumptions used to simulate

the transport process. The mass transport model in PAD assumes transport

of atomic species only; HTRs may contain large quantities of circulating

particulate matter. If fission products sorb on this dust, their transport

characteristics will be altered. Since aerosol transport is still an

emerging science, there is little hope of accurately modeling the phenomena

from first principles. All that can be said presently is that the limited

experimental plateout profiles generally indicate molecular transport.

Other areas of modeling difficulties involve the extent to which sorbed

fission products penetrate in the bulk of structural materials. Also

the precise nature of the sorption of iodine on steels is unclear. These

questions will only be solved by additional research -- probably involving

large-scale integral testing.

16

4, DESIGN VERIFICATION PROGRAMS

4.1 GA Deposition Loop

The GA deposition loop (Ref. 1) was constructed to study plateout

under conditions approaching those in the evaporator-economizer sections of

an ITTGR steam generator. High pressure helium ('v,20 atm) circulated in

fully developed turbulent flow (Re = '\'12,000) in a closed loop. The loop

was constructed from a 0.93 in. i.d. tube of low alloy steel (T2 - 1/2% Cr,

1/2% Mo). As shown schematically in Fig. 4, the entire loop assembly was

inserted into a high-pressure, high-temperature autoclave. Four large

autoclave heaters were the primary heat source for the assembly; loop sur­

face temperatures ranged from a low of about 50 C when a water-cooled

chill block was clamped to portions of the loop tubing to a high of about

500 C when the chill block was omitted.

Deposition of cesium, strontium, and iodine was investigated. Cesium

tagged with Cs-137 and strontium tagged with Sr-85 impregnated on graphite

powder and Pdl2 tagged with 1-131 were used as source materials; the sources

were loaded in porous graphite crucibles. These charged crucibles (typi­

cally, five in number) were then placed in an electrical resistance source

heater located centrally in the helium flow path. Evaporation of the source

material from the graphite surfaces into the flowing helium provided a

realistic source of fission product nuclides.

Five experiments were performed; typical results for cesium and iodine

are shown in Figs. 5 and 6, respectively. Because of the chill block,

surface temperatures were low in both experiments (-260 C). Also shown in

the figures are the profiles predicted with the PAD code using reference

17

TC NO.

MAIN PRESSURE VESSEL

AUTOCLAVE

HEATERS

1 - I N . TUBING {l-\/h C r - 1 MOLY OR MILD STEEL)

SOURCE

HEATER

LOOP BLOWER

HELIUM

CHILL BLOCK

100 CM

Fig . 4 GA f i s s i o n product depos i t i on loop (Ref. 1)

18

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sorption isotherms; because of the low temperatures and surface concentra­

tions, perfect sink behavior is predicted in both cases -- hence the mono-

tonic decrease in specific activity with increasing distance from source.

Inspection of Fig. 5 shows large discrepancies between the predicted

and measured cesium profiles in Loop 1. The slope of the experimental pro­

file over the first 50 cm of tubing is much steeper than predicted, yet

at other locations (see dotted lines) the slope is about as expected. Far

more distressing, however, is the large but unpredicted increase of plate­

out levels in the chill section. The cause of this increase is unknown

as is the reason for the abrupt decrease in activity, which occurs about

half-way through the chill. Loop 2, a strontium experiment, gave qualita­

tively similar results. As shown in Fig. 6, the 1-131 plateout profile

obtained in Loop 3 is in good agreement with the predicted profile. The

iodine profile is most interesting when compared to the cesium profile

from Loop 1. The operating conditions of these two loops were quite

similar, although Loop 3 did have a 20% higher mass flow rate and slightly

lower surface temperatures ( 20 to 60 C lower). The effect of flow dis­

turbers (particularly at the entrance contraction and 90-deg bend at the

apex) seems much less pronounced on the iodine profile compared to the cesium

profile. But more surprising, the chill section seems to have had much

less effect on iodine than on cesium and strontium --a strange effect

indeed since iodine is the most volatile. In Loop 4, another cesium experi­

ment, the chill block was eliminated in order to attain higher surface

temperatures (350 to 450 C)• In this case, the measured profile exhibited

the expected perfect sink behavior. The pronounced effect of the chill block

on the deposition of cesium and strontium, but not iodine, has not been

explained.

4.2 BMI Deposition Loop

In the early 1960's, researchers at Battelle Memorial Institute per­

formed a series of plateout experiments in an out-of-pile loop; since

21

their work was reported in considerable detail (Ref. 9), it provided a

natural validation exercise for the PAD code. Helium at 20 atm was circu­

lated in the loop shown schematically in Fig. 7; the basic material of

construction was 0.37 in. i.d. type 316 stainless steel. The source of

radioactivity was a mildly irradiated specimen of coated-particle fuel

which could be heated to 1000 C. Leaving the source chamber, the helium

at about 650 C first traversed a high-temperature isothermal zone, then a

water-cooled section, and finally a low-temperature isothermal section.

The surface temperatures in the chill section were not measured but have

been estimated to range from 575 to 100 C (Ref. 1). The flow was in the

transitional regime with Reynold's numbers ranging from 5000 to 7000 within

the loop. After a period of several days' operation, the loop was disected,

and the plateout distribution determined by gamma counting 'vl/2 in. long

sections of pipe. The radionuclides identified in the pipe scans were

Ce-141 + Ce-144, Ba-La-140, Zr-Nb-95, Ru-103, Cs-137, Te-129, and 1-131.

Typical results for Cs-137, Te-129, and 1-131 are given in Figs^ 8 and

9. In general, the Ce-141 + Ce-144, Ba-La-140, Zr-Nb-95, and Ru-103

profiles resembled the Te-129 profiles. The thin solid lines in the

figures represent the BMI researchers' attempt to correlate the data (Ref.

9), and the other lines represent Hanson's attempt (Ref. 1).

Preliminary parameter studies were performed with the PAD code using ref­

erence physical data as a base. Good qualitative agreement was realized

when treating the surface as a perfect sink for Ce-144, Zr-95, Ru-103,

and Te-129 and when employing extrapolated adsorption isotherms for 1-131

and Cs-137. To obtain the proper cesium distribution, it was necessary to

presume that the SS316 surfaces of the BMI loop had a cesium sorptivity

1/333 that of oxidized SS304 for which laboratory sorption data are avail­

able (Ref. 34); the assumption is not unreasonable since oxidized SS304

was 80-100 times more sorptive than as-received SS304, and SS316 is more

oxidation resistant than is SS304. The only iodine sorption data available

at the time of analysis (1974) were taken on 1% Cr, 1/4% Mo chromaloy

22

369 TO 371-72

373 74

404 Swaqelok

Specimen heater

Fig. 7 BMI deposition loop schematic (Ref. 9)

23

Fig. 8 Distribution of Cs-137 and Te-129 activity in BMI deposition loop (Ref. 9)

D|QQ - 0.125 cm /sec, '(.8 x 10'5 Atoms of I

TC

1 1 1 1 High - Temperctue Isothermal Zone

Perfect Sink ''^ ^

20

03

- ^

o o O 00

o -oO-

.131

Sb O O °o QCQ o

— I 1 1 1 Low-Temperature Isottiermol Zone

60 80 100 120 '160 180 2 0 0 2 2 0

Section Number 240 V 280 300 320 340 360 380

Fig. 9. Distribution of 1-131 activity in BMI deposition loop (Ref. 9)

steel (Ref. 35). Shown in Fig. 9 is a PAD calculation assuming that SS316

had an iodine sorptivity 1/10 of that of chromaloy steel; inspection shows

that the actual SS316 sorptivity must be even less than that assumed, and

that the heat of adsorption (i.e., the temperature dependence) was also

less than assumed. The mass transfer coefficient was also overpredicted

by about 70% using the reference gaseous diffusivities and correlation.

The PAD calculations shown in Figs. 8 and 9 used gaseous diffusivities

1/2.5 times the reference values.

4,3 Peach Bottom EOL Program

Peach Bottom Unit 1 was a 40-MW(e) prototype HTGR owned and operated

by Philadelphia Electric. The two-loop primary coolant circuit, shown sche­

matically in Fig. 10, was comprised of a steel reactor vessel, vertical U-

tube steam generators, motor-driven compressors, and interconnecting piping.

After seven years of commercial operation, the reactor was shut down for de­

commissioning on October 31, 1974, because of its uneconomically small size.

An End-of-Life Program has been under way since then to evaluate the core

and plant performance in order to verify reference HTGR design methods;

particular attention is being given to fission product transport on the

fuel side and to performance of the Incoloy 800 on the plant side. Consi­

derable plateout data have been obtained by a combination of in-situ gamma

scanning (Refs. 14, 16) and radiochemical analysis of destructively removed

samples (Ref. 15).

The plateout distribution of gamma-emitting nuclides in the primary

circuit at end-of-life was determined by in-situ scanning; the work was

performed by IRT Corporation, San Diego, California, under subcontract to

General Atomic. The specific activity was mapped by scanning the accessible

ducting at 11 locations with a Ge(Li) detector and by axially traversing

80 steam generator tubes with travelling CdTe detectors from the water side.

Following destructive removal of trepan samples, a traveling intrinsic

germanium detector was inserted sequentially into two vertical ducts and

26

A ORNLXSCANS (EXTERNAL) • IRT/SCANS (EXTERNAL) • SUNTAC TREPANNED SAMPLE

— IRT /SCANS (INTERNAL)

MAIN MOTOR HYDRAULIC COUPLING

PONY MOTOR COOLING BLOWER PONY MOTOR

MAIN HELIUM COMPRESSOR

LOOP 2

L00P1

BURN HOLES FOR SLECTED TUBE SAMPLING

REACTORS VESSEL

COMPRESSOR OUTLET

COMPRESSOR INLET

Fig . 10 I sometr ic of primary coolant system (Ref. 16)

the plateout mapped along 6-m runs of ducting. Calibration measurements

on mock-ups allowed reduction of the spectra to specific activity.

The external duct scans were straightforward; ORNL had made similar

measurements throughout Core 2 operation (Ref. 36); scan locations are

shown in Fig. 10. Mapping the plateout activity on the tube bundle was

more formidable in that the steam generator tubes, some with a 9.6-mm inside

diameter, had to be traversed to a depth of 6 m (the steam generator cross

section is shown in Fig. 11). A miniaturized, tantalum-shielded, cadmium

telluride semiconductor detector was found acceptable. With access at the

tubesheet, the detector affixed to a coaxial cable was inserted to the bottom

of the U-tube, and spectra were acquired at 0.15-m increments as the detector

was withdrawn remotely. A similar configuration for transporting the ger­

manium detector was used for the internal duct scans.

The dominant gamma-emitting nuclides were Cs-137 and Cs-134; their

distributions were similar. Radioassay of the destructively removed samples

confirmed the specific cesium activities determined by the in-situ scanning;

Sr-90 was also measured but the specific strontium activity was about

1/1000 that of cesium. Neutron activation analysis of leach solutions

failed to detect any 1-129 or Te-126. The internal scans (Fig. 12) revealed

little local structure to the plateout over the 6-m length of ducting

traversed, only a gradual decrease in specific activity in the direction of

coolant flow. Along the 110-m run of ducting from the steam generator

exit back to the reactor vessel, the specific activity was more variable

but the trend was the same (Fig. 13). In the steam generator, a signifi­

cant entrance effect was observed in the superheater section; the activity

was highest where the inlet jet impinged and lowest at the ends of the bundle

despite the presence of a flow baffle; the effect damped out with penetra­

tion into the bundle, and the axial profile was uniform at the economizer

exit (Fig. 14). When the axial profiles are averaged, the specific activity

decreased monotonically across the tube bundle.

28

72 IN. LD.

ZS FT. -4 IN. ECONOMIZER INLET TUBES

SHELL DRAIN - 9 0 :N. LO.

Fig. 11 Cross section of steam generator

4.0

1.0 0.8

^ 0

CM E u

G

Coolant Flow

—A-—r- — ^ - - i i

^ • • * . *

<

5= 4.0 o a

2.0

1.0 0.8

A 134 cs

• '37 cs

0.6

H0.4

— Predicted ^lope

1.0 2.0 3.0 4.0 5.0

Distance below Location 1-4 (m) 6.0

A «34c3

• '37c,

— Predicted Slope

Coolant Flow 1.0

0.8

1.0 2.0 3.0 4.0 5.0

Distance below Location I-10 (m)

Fig, 12 Cesium profiles in Peach Bottom cold ducts from internal scans (Ref. 16)

30

10 -5

Cs-137 Profiles

Predicted — Least Squares

^ <»

10

10 -7

®

•s. •9

A ^

^O

>

o

~^A ®

o Source Method

IRT external

134 Ql

137

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external sample internal

A

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®

Loop I: Filled Symbols

Loop 2: Open Symbols

u

o

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1^

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<

C «> u c o o

10'

10'

20 40 60 80 100

Distance from S.G. Exit (m)

120

Fig. 13 Cesium deposition profiles in cold duct (Ref. 16)

31

\J

1

2

3

4

5

6

•

o

A

•

• • • •

• •

Superheater Inlet 110 Superheater Outlet 13 Evaporator Outlet 102 Economizer Inlet 107

• • •

• •

• •

0

0 • 0 •

• 0

• o

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o 0

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o • o •

0 •

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=_ Coolant Flow

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A

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Specific Activity (/* Ci/cm^)

3.0 2.0

Fig. 14 Axial Cs-137 distribution in steam generator (Ref. 16)

32

The complete experimental and predicted cesium plateout distributions

are compared in Fig. 15, the format of which is the PAD code representation

of the Peach Bottom primary circuit; all the IRT data are displayed therein --

the collapsed steam generator data along with the duct scans. The specific

activity is plotted as a function of fractional cumulative surface area.

(Note that the abscissa is drawn to scale within a given section but differs

from one section to another.) Two PAD calculations are shown: (1) mass

transfer control (i.e., the surfaces are perfect sinks for cesium), and (2)

sorptivity control. In both cases, the time-average core release rate of

cesium was adjusted so that the predicted specific activity at the evaporator

inlet (shell side) was approximately equal to the measured value (^ yCi/cm^).

Since the decay of Xe-137 produced negligible amounts of Cs-137 compared to

the directly released component, the relative distributions shown in Fig. 15

apply equally well to Cs-134 which has no gaseous precursor.

Inspection of Fig. 15 indicates that the mass transfer control, or

perfect sink, case (solid lines) resulted in good agreement everywhere ex­

cept in the hot duct leading from the reactor vessel to the steam generator.

Here the specific cesium activity is overpredicted by an order of magnitude.

Since the flow geometry is simple (a circular duct), prediction of the mass

transfer coefficient should be reasonably accurate. Thus, the logical con­

clusion is that the deposition process in the hot duct is not limited by

mass transfer effects but rather by the high surface temperature.

Once again, the major difficulty is choice of appropriate sorption iso­

therms to describe the cesium sorptive capacity of the surface. The hot

duct cladding was constructed of SS304 for which sorption data are available

(Ref. 34). For SS304, the oxidation state of the surface has a profound

effect with preoxidation favoring increased sorption. Another complication

was that all exposed surfaces in the primary circuit were covered with a car­

bonaceous deposit produced by cracking of lubricating oil which had leaked

into the primary circuit (Ref. 37). This carbon deposit was possibly a

significant sink for cesium. Conceivably then, the plateout surfaces may

33

10 -4

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10 -5

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K) -7

o. o. 3

IRT DATA AS OF 7/1/76

Perfect Sink

Sorptivity Control

Loop I: Filled

Loop 2: open

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Fraction of Accumulated Surface Area

Fig . 15 P la teou t d i s t r i b u t i o n of Cs-137 and Cs-134 in Peach Bottom HTGR (Ref. 16)

34

be more appropriately characterized as being carbonaceous rather than

metallic. The physical and chemical characteristics of this carbon deposit

are still being determined at GA and ORNL. Presently, it is known to range

from about 6 to 12 pm in thickness and to be 80 to 90% carbon with up to

10% iron and detectable amounts of sulfur and silicon (Ref. 36). Little

is known about the structure; this is unfortunate because the cesium sorp­

tivity of carbon substances is strongly structure-dependent. Given these

uncertainties, the surface sorptivity was investigated parametrically (Fig.

16); details are given in Ref. 16,

In summary, the experimentally observed cesium plateout distribution

in Peach Bottom can be predicted almost exactly with the PAD code providing

appropriate sorption isotherms are employed. However, the observed sorption

behavior is consistent with either assuming that the primary cesium sink is

a relatively oxide-free SS304 surface or assuming that the carbon deposit has

a cesium sorptivity intermediate to that of graphite and matrix. These assump­

tions are equally feasible; in reality, both probably contributed to the

total sorptive capacity of the surface. Cesium deposition throughout the

circuit was apparently mass transfer controlled with the exception of the

hot duct. The superheater entrance effect probably resulted from the mal­

distribution of coolant flow. The profiles suggest that cesium was trans­

ported primarily in atomic form despite the presence of carbonaceous dust.

4.4 CPL-2 Test Program

The CPL-2 test program is a series of four fission product transport

experiments that were performed in the Pegase test reactor. The tests

were conducted by the Commissariat k I'Energie Atomique (CEA) as a part

of a private cooperative program between CEA and GAC; however, the data

analysis at GAC is being conducted under the DOE-sponsored Code Valida­

tion Program.

The CPL-2 loop is shown schematically in Fig. 17; the essential features

of the test section are a fuel element (representative of GAC prismatic

35

10

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Assumed Sorptivity

Perfect Sinkj . 137 c e 3 0 4 It) { CS •

Matrix (1) ) cs'^^ A Graphite (1) . , ^.,. ^ -

Loop 1: Fi ed SS 304 1/20 . „ «

Loop 2: Open SS 304 (l/IO Matrix (1/20)

\ f

(X) Sorptivity Relative to Base Material

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Fraction of Accumulated Surface Area

ig . 16 Effect of surface sorpt iv i ty on cesium plateout d i s t r ibu t ion (Ref. 16)

36

6

1 2 3 4 5 6 7 8 9 10 11 12 13 14

FUEL REFLECTOR BASKET HEAT EXCHANGER FILTERS AND GAS SAMPLING FILTER OUTLETS GAS SAMPLING TO Ge-Li DETECTOR DUST SAMPLER TOOUT-OF-PILE LOOP HEAT EXCHANGER PRESSURE TUBE REGULATION VALVE BLOWER FROMOUT-OF-PILE TROLLEY

362 CM

Fig . 17 CPL-2 loop schematic (Ref. 38)

37

block design), reflector element, and a counter-current, shell-and-tube i

heat exchanger-recuperator. The other main components are the water-gas

heat exchanger (to regulate the inlet temperature to the test section),

the blower, and the interconnecting piping; the loop was highly instru­

mented to monitor operation. A purification/gas-treatment loop allowed

regulation and measurement of impurities in the helium. The release of

fission products from the fuel was monitored by a gas-sampling system as

well as diffusion probes up- and downstream of the heat exchanger-recuperator.

An isokinetic sample of gas was removed downstream of the recuperator and

passed through a cascade impactor to characterize circulating particulate

matter. The loop design and operation were described in greater detail

previously (Ref. 38).

A series of four tests were performed in the CPL-2 program. CPL 2/1

was intended to represent nominal HTGR operating conditions and, hence,

to serve as a benchmark; the total oxidants were to be maintained at llO

ppm. Near the end of the CPL 2/1 irradiation (typical test lasted 60 days),

a relatively large amount of water was inadvertently injected, and the oxidant

level rose to 136 ppm. Consequently, the benchmark test was repeated; in

this test, identified as CPL 2/1-Bis, the oxidant level was maintained within

specifications. In CPL 2/3, the oxidant level was intentionally increased

in stages up to 100 ppm where it was maintained for most of the irradiation.

CPL 2/4 was an in-situ depressurization test: the loop was operated under nominal conditions for 60 days and then rapidly blown down with the effluent

passing through a series of traps to recover the radionuclides released from

the loop. The data acquisition phase of the program (irradiation and post-

irradiation examination) is essentially complete at this writing; analysis

and interpretation of CPL 2/1 results are progressing, but the entire pro­

gram is not scheduled for completion until October, 1978.

The CPL-2 test data promise new insights into all facets of fission

product transport, but for understanding plateout phenomena, the most

important data are the deposition profiles along the tubes of the heat

38

exchanger-recuperator. Shown schematically in Fig. 18, the heat exchanger

consisted of 186 tubes, 4-mm inside diameter and 1.25-m long, arranged in

a close-packed array inside an 86-mm-diameter cylindrical shell. Several

different types of steels, both in an as-received and a pre-oxidized state,

were used to fabricate the tubes; included were the French equivalents of

Incoloy 800, Hastelloy B, SS347, SS410, and T22 (2-1/4% Cr, 1% Mo chromaloy

steel). Since the T22 and stainless steels could not withstand the tem­

perature at the inlet of the heat exchanger, half-length tubes of these

alloys were joined with a half-length tube of 1800 which was oriented in

the upstream (hotter) half of the exchanger.

During operation, helium at '\'750 C entered the tube side of the ex­

changer where it was cooled to - 350 C by a counterflow of cold helium re­

turning from the loop blower. While a series of thermocouples measured

average inlet and outlet gas temperatures, both on the tube- and shell-side,

no measurements of surface temperatures were possible; however, detailed

calculations (Ref. 39) have shown that the temperature and flow fields

within the heat exchanger were very complex. This occurrance introduces

an additional complexity to the interpretation of the deposition profiles.

Upon removal from the loop, the plateout distribution was determined

by a combination of global scanning of the assembled exchanger, scanning

of individual tubes, and leaching followed by the radiochemical analysis.

Radionuclides identified included 1-131, Te-127m, Cs-134, Cs-137, Sr-89,

Sr-90, and Sb-125 as well as the activation products Cr-51, Co-58, Co-60,

Fe-59, and Mn-54. The activation products generally tracked the neutron

flux distribution in the exchanger; most of the fission products exhibited

apparent perfect sink deposition profiles, but the cesium and iodine iso­

topes concentrated in the colder part of the exchanger.

Typical results for Cs-137 and 1-131 are shown in Figs. 19 and 20,

respectively; also shown are PAD code calculations using sorption isotherms

determined by CEA (Ref. 40). These sorption measurements were made on

39

ta

o

A-A

,dO«^

OnO§OoOoObOo9o°/ Kom

o<

=Dr

Fig. 18 Heat exchanger - recuperator

O OBSERVED

CALCULATED (CEA SORPTION DATA),

INCOLOY800 —

> t -

>

TUBE-SIDE INLET

TUBE-SIDE OUTLET

400 800

AXIAL DISTANCE (MM)

1200

Fig. 19 Comparison of calculated and observed Cs-137 plateout profiles for tube B-31 from CPL 2/1 heat exchanger (Ref. 32)

41

r

>• > <

O OBSERVED

INCOLOY 800

CALCULATED (CEA SORPTION DATA)

- I TUBE-SIDE INLET

TUBE-SIDE OUTLET

>

<

400 800 1200

AXIAL DISTANCE (MM)

Fig. 20 Comparison of calculated and observed 1-131 plateout profiles for tube B-31 from CPL 2/1 heat exchanger (Ref. 32)

42

the actual materials of construction and in the actual temperature range,

but they were taken in a partial pressure range some three to four orders

of magnitude higher than that which prevailed in the CPL-2 loops. (It is vir­

tually impossible to perform pseudoisopiestic sorption experiments at pres­

sures much less than 10"^^ atm.) The data were fit with a Freundlich iso­

therm and extrapolated to lower pressures. Inspection of the figures

indicates good agreement between the measured and predicted cesium profile;

the shape of the iodine profile (i.e., the temperature dependence) is correct,

but the absolute magnitude is grossly underpredicted. This pattern was con­

sistently seen among tubes of varying material type, initial oxidation state,

and temperature distribution. The explanation is unclear, but it is sus­

pected that the extrapolation of the sorption data to lower partial pressures

was inappropriate for iodine. Since the analysis of the CPL-2 plateout data

is not yet complete, these results should be considered tentative.

4.5 Fort St. Vrain Surveillance

The most valuable fission product transport data recorded thus far by

the Fort St. Vrain Surveillance Program (Ref. 17) relates to fission gas

release during the initial rise-to-power phase. Up to 28% power, the meas­

ured releases of krypton and xenon isotopes have been about a factor of two

less than predicted (Ref. 17). Although the surveillance program has

yielded little plateout data thus far, its potential over the next several

years is large. The most valuable source of information is expected to be

the two plateout probes, one in each coolant loop. The design of the probe

is such that each one samples both the core inlet and outlet coolant passing

these samples through individual diffusion tubes and filters. Upon removal

during the first refueling shutdown, the initial set of probes will be

radioassayed. Analysis of these data will provide information about the

amount, distribution, and chemical form of condensible radionuclides re­

leased into the coolant during the sampling period. In addition to the

probes, an iodine monitor has also been installed to determine the amount

of circulating iodine in the primary coolant. The device consists of a

43

tube-within-a-tube arrangement through which a sample of primary coolant

is drawn. A small charcoal trap at the inlet orifice of the inner tube

collects 6.7-hr 1-135 and 21-hr 1-133. The outer tube permits the xenon

daughters of the sorbed iodine to be transported to a fast-acting gas

sampling station with a purge flow of purified helium; the design is such

that during the measurement period only xenons produced from the deposited

iodines will be collected for gamma counting. Other plateout information

will be obtained by gamma-scanning accessible components (the PCRV design

limits this approach) and by radioassay of components changed out of the

primary circuit (e.g., spent reflector blocks, filters and charcoal from

the purification train, etc.).

44

5. CONCLUSIONS

From the above discussions, it is obvious that plateout phenomena

are not yet sufficiently characterized to meet the needs of reactor de­

signers; this is true in an empirical sense, not to mention a mechanistic

sense. Hence for the present at least, the huge uncertainties associated

with plateout behavior in AIPA must remain; to reduce them will require

substantial additonal work. This effort will necessarily be long term

and expensive both in terms of manpower and money, and since plateout

is generic to all HTRs, it would seem a natural area for collaborative

work among the various HTR researchers.

Given the limited data base, any conclusions are naturally tentative,

but a few trends are apparent. The PAD code (and, by induction, its

progeny PADLOC) appears to be an adequate empirical tool for prediction

of plateout distributions under steady-state conditions provided appro­

priate material property data are available, the most important of which

are sorption isotherms. Whether these codes will be adequate for the

analysis of LOFC events, with their more extreme conditions, remains

to be seen.

Extensive additional work will be required. Further differential

laboratory experiments will be needed as will large-scale integral tests.

The laboratory tests should focus upon defining the sorptivities of per­

tinent structural materials for the key radionuclides: iodine, strontium,

cesium, tellurium, and silver. New laboratory techniques should be sought

which would allow measurements in the partial pressure ranges expected

in HTRs. The engineering-scale integral tests must include two areas

which have had only limited attention to date. A comprehensive series

45

of tests is urgently needed to establish the extent to which particulate

matter in the primary circuit would alter the transport behavior of con­

densible species. Likewise, a rigorous assessment is lacking of plateout

phenomena under LOFC conditions (large sources, high temperatures, and

free convection). The preceding was not to imply that molecular transport

under steady-state conditions has been sufficiently characterized for it

has not been; it is simply a matter of recommending a more balanced program

for future research.

46

REFERENCES

1. Hanson, D. L., "Results of the General Atomic Deposition Loop Program,"

USERDA Report GA-A13140, General Atomic Company, April 1, 1976.

2. Vanslager, F. E., and L. D. Mears, "PAD: A Computer Code for Calculating

the Plateout Activity Distribution in a Reactor Circuit," Gulf General

Atomic Report GA-10460, January 1971 .

3. Hudritsch, W. W., and P. D. Smith, "PADLOC: A One-Dimensional Computer

Program for Calculating Coolant and Plateout Fission Product Concentra­

tions," USERDA Report GA-A14401, General Atomic Company, to be published.

4. "HTGR Accident Initiation and Progression Analysis Status Report,

Vol. 5, AIPA Fission Product Source Terms," USERDA Report GA-A13617,

General Atomic Company, February 1976.

5. Ibid., "Vol. 6, Event Consequences and Uncertainties Demonstrating

Safety R^D Importance of Fission Product Transport Mechanisms."

6. Ibid., "Vol. 7, Occupational Radiation Exposures from Gas-Borne and

Plateout Activity."

7. Code of Federal Regulations 10CFR50, Appendix B: "Quality Assurance

Criteria for Nuclear Plant and Fuel Reprocessing Plants."

8. American National Standard Institute ANSI N45.2.11-1974: "Quality

Assurance Requirements for the Design of Nuclear Power Plants."

9 Raines, G. E., et al.. "Experimental and Theoretical Studies of Fission

Product Deposition in Flowing Helium," USAEC Report BMI-1688, Battelle

Memorial Institute, August 21, 1964.

10. Brown, P. E.. et al., "Measurements of Fission Product Deposition on

the Heat Exchangers and Circulators of the Dragon Reactor," Dragon

Project Report DP-564, United Kingdom Atomic Energy Authority, July 1968.

47

11. Jensen, D. D. , M. J. Haire, J. E. Baldassare, and D. L. Hanson, "Plan- ^ ^

ning Guide for Validation of Fission Product Transport Codes," USERDA ^^F

Report GA-A13386, General Atomic Company, April 15, 1975.

12. Ingrao, G., H. F. Mollet, and P. R. Rowland, "The Idylle 03 Fission

Product Migration Experiment," Dragon Project Report DPTN/800.

13. Von der Decken, C , H. Gottaut, J. Malinowski, K. MUnchow, and W. Essler,

"Das Bestrahlungsexperiment Saphir im Reaktor P6gase in Cadarache,

Reaktortagung des DA^F," KFA, Karlsruhe, 1973.

14. Selph, W. E., and D. E. Bryan, "Measurement of Fission Product Activity

in the Peach Bottom Reactor Primary Coolant Loop," USERDA Report GA-A14059,

General Atomic Company, August 1976.

15. Baldwin, N. L., B. L. Norman, and W. E. Bell, "Radiochemical Examination

of Peach Bottom Components," General Atomic Report GA-A14495, to be published.

16. Hanson, D. L., N. L. Baldwin, and W. E. Selph, "Gamma Scanning the

Primary Circuit of the Peach Bottom HTGR," General Atomic Report GA-A14161,

October 31, 1976.

17. "HTGR Fuels and Core Development Program Quarterly Progress Report for

the Period Ending August 31, 1976," USERDA Report GA-A14046, General

Atomic Company, September 24, 1976, p. 4-74 ff.

18. "Fort St. Vrain Nuclear Generating Station Final Safety Analysis Report,"

1967, Appendix D.3, Section D.3-1, Public Service Company of Colorado,

(USAEC Docket 50-267).

19. "GCR Safety Program Quarterly Progress Report for the Period Ending

March 31, 1977," USERDA Report GA-A14382, General Atomic Company,

April 1977, Section 5.

20. Ozisik, M. N., "An Analytical Model for Fission Product Transport and Deposi­

tion from Gas Streams," USAEC Report ORNL-3370, Oak Ridge National

Laboratory, July 1963.

21. Ozisik, M. N., and F. H. Neill, "An Analysis of Fission Product

Deposition and Correlation with Experiment," CONF-650407 (Vol. 2),

International Symposium on Fission Product Release and Transport Under

Accident Conditions, Oak Ridge National Laboratory, April 5-7, 1965. ^ ^

48

Iniotakis, N., J. Malinowski, and K. MUnChow, "Initial Results of

Investigations into Fission Product Deposition in In-Pile Experiments,"

Nuclear Engineering and Design 34, 169 (1975) .

Hoinkis, E., "A Review of the Adsorption of Iodine on Metal and Its

Behavior in Loops," USAEC Report ORNL-TM-2916, Oak Ridge National

Laboratory, May 1970.

Satterfield, C. N., Mass Transfer in Heterogeneous Catalysis, MIT

Press, Cambridge, Massachusetts, 1970.

Bumette, R. D., et al., "Evaluation of Carbon Transport Phenomena

in HTGR Systems," USAEC Report GA-8624, Gulf General Atomic, October 5,

1968.

Craig, G. T., "The Behavior of Particulate Matter in High-Temperature,

Gas-Cooled, Graphite Reactor Primary Coolant Systems," General Atomic

Report GA-A13402, July 1, 1975.

Chmielewski, R. D,, and C. G, Hoot, "Fission Product Plateout Cal­

culations for a Gas Turbine HTGR," USERDA Report GA-A13213, General

Atomic Company, November 1, 1974.

Busch, D. D., "The Nature of Condensable Fission Products in an HTGR

Environment," General Atomic Division of General Dynamics Report

GA-6957, April 15, 1966.

Treybal, R. E., Mass Transfer Operations, 2nd Ed., McGraw-Hill,

New York, 1968, p. 62.

"HTGR Base Program Quarterly Progress Report for the Period Ending

February, 28, 1967," USAEC Report GA-7801, General Dynamics, General

Atomic Division, April 20, 1967.

Osborne, M. P., E. L. Compere, and H, J. de Nordwall, "Studies of

Iodine Adsorption and Desorption on HTGR Coolant Circuit Materials,"

USERDA Report ORNL/TM-5094, Oak Ridge National Laboratory, April 1976.

"HTGR Fuels and Core Development Program Quarterly Progress Report

for Period Ending May 31, 1976," USERDA Report GA-A13941, General

Atomic Company, June 30, 1976.

49

"GASSAR-6," General Atomic Standard Safety Analysis Report, February 5,

1975 (NRC Docket STN 50-535).

Milstead, C. E., and L. R. Ziunwalt, "Cesiiun Deposition on Stainless

Steel," Nucl. Appl. 3, 495 (1967).

Milstead, C. E., W. E. Bell, and J. H. Norman, "Deposition of Iodine

on Low-Chromium-Alloy Steel," Nucl. Appl. Tech. 1_, October 1969.

Dyer, F. F., R. P. Wichner, W. J. Martin, and H. J. de Nordwall,

"Distribution of Radionuclides in the Peach Bottom HTGR Primary Cir­

cuit during Core 2 Operation," USERDA Report ORNL-5188, Oak Ridge

National Laboratory, March 1977.

Scheffel, W. J., N. L. Baldwin, and R. W. Tomlin, "Operating History

Report for the Peach Bottom HTGR," Vol. 1, USERDA Report GA-A13907-1,

General Atomic Company, August 31, 1976.

"HTGR Fuels and Core Development Program Quarterly Progress Report

for the Period Ending August 31, 1975," USERDA Report GA-A13592,

General Atomic Company, September 30, 1975.

"HTGR Fuels and Core Development Program Quarterly Progress Report

for the Period Ending February 29, 1976," USERDA Report GA-A13804,

General Atomic Company, March 31, 1976.

Blanchard, R., CEA, unpublished data.

50

APPENDIX A

REVIEW OF THE MATHEMATICAL MODEL

EMPLOYED BY THE PAD CODE

The mathematical model employed by the FAD code Is thoroughly

documented In Ref. 2; however, for convenience, certain sections of that

document are excerpted below.

A.1. CWSERVATION EQUATIONS

The conservation equations presented here describe the mass balance

for the coolant and surface fission product concentrations In sections of

the reactor circuit with constant cross-sectional areas. The treatment Is

a simplified analysis giving surface concentrations and average coolant

concentrations as a function of axial distance and time. It Is assumed

that the mass-transport process can be described to be between an average

coolant concentration and a surface concentration separated by a thin

gaseous boundary layer film. The quantity of fission products contained In

the thin film Is assumed to be negligible. Figure A-1 Illustrates the

basic mass-transport volume element.

The coolant and surface concentrations of each fission product are

dependent on a source term, such as the release of the fission product from

the reactor core, on the decay of the parent fission product In the primary

circuit, on the decay of the fission product Itself, on mass transfer

between the coolant and surface, and on axial convection around the reactor

circuit. Hence, the time rate of change of the amount of fission product j

In an Increment of volume of length dx, cross-sectional area A, and wetted

perimeter P is given by

A-1

C, (x.t)

Fig. A-1. Basic mass-transport volume element

A-2

3C,0c,t) — J Adsc - B (x)Adx + X^C^(x,t)Adx - X C (x,t)Adx

Source Decay from Decay term parent

\ l 9[V(x)C.(x,t)]\ 1 - lv(x)Cj (x.t) + ^ ydx A - V(x) Cj (x.t) Adx

Flow out of Increment - flow in » loss rate due to axial convection

- k(x)[C.(x,t) - C (x,t)]Pdx . (A-1)

Mass transfer from average coolant concentra­tion to coolant concentration at the surface

or.

3C,(x,t) 9[V(x)C,(x,t)] • ^ B (x) + X^C^(x,t) - X C (x,t) - ±

'j"~' 'i'l""*"' j"j" ' ' 8x

_ l c ^ [C,(x,t)-C (x.t)] . (A-2) A J Sj

where C.(x.t), C.(x,t) • coolant concentration of fission product j and

Its precursor 1 averaged across the coolant 3

channel, yg/cm ,

C (x,t) •* coolant concentration of fission product J at the ^4 2

" channel surface, yg/cm ,

A-3

X " axial distance coordinate, cm.

t • time coordinate, sec.

B. (x) •• source of fission product j In the coolant, 3

yg/cm -sec.

X.. X > decay constant of fission product J and ^ -1

Its precursor 1, sec ,

V(x) • velocity of coolant, cm/sec,

P/A - ratio of wetted perimeter to cross-sectional area

of coolant channel, cm" .

k(x) - mass-transfer coefficient, cm/sec.

Similarly, the time rate of change of the amount of fission product J

on a surface element of length dx and wetted perimeter P Is given by

3S.(x.t) — J Pdx - b (x)Pdx + X^S^(x.t)Pdx - X S (x.t)Pdx

-' \-Source Decay from Decay term parent

+ k(x)[C (x.t) - C (x,t)]Pdx . (A-3)

Mass transfer from average coolant con­centration to coolant concentration at the surface

A-4

or.

as (x.t) - ^ - b^(x) + X^S^(x,t) - XjSj(x,t)

+ k(x)[C (x.t) - C (x,t)] , (A-4)

where S.(x,t), S.(x.t) - surface concentration of fission product j and J 2

its precursor 1. yg/cm .

b,(x) - source of fission product j on the surface,

yg/cm -sec.

and the remainder of the terms are as previously defined. Equations A-2

and A-4 control the behavior of fission product j within one section of the

reactor circuit. A similar set of equations must be written for each

reactor section in order to completely describe the behavior of fission

project j throughout the reactor circuit. The initial and boundary

conditions are: (1) the initial concentrations are specified and (2) the

cool£mt concentration is continuous around the reactor circuit.

A.2. MASS-TRANSFER COEFFICIENT

The mass-transfer coefficient k(x) used in Eqs. A-2 and A-4 should be

for mass transfer in a turbulently flowing gas stream through a channel of

constant cross-sectional area. A formulation for channels of circular

cross-section is given by

k(x) - [0.023 D(x)/d] [Re(x)]°*^^ [Sc(x)]°-^^ . (A-5)

2 where D(x) - diffusion coefficient of species in gas stream, cm /sec.

d - diameter of circular conduit, cm.

A-5

Re(x) = Reynolds number of gas stream.

Sc(x) = Schmidt number of gas stream.

Although this expression is primarily for channels of circular cross-

section, other shapes can be accommodated by using the hydraulic diameter.

A.3. SURFACE SORPTION

In order to relate the coolant concentration at the surface. C (x.t). s

to the surface concentration, S(x,t), an algebraic expression for the

surface sorption equilibrium is needed. Several equations have been found

to approximate a large amount of the experimental isotherm data on

equilibrium partial pressures versus surface loading. One of these, the

Freundlich equation, can be written as

Cg(x,t) = K^(x) S(x.t)'' '' . (A-6)

where K-(x) = the temperature-dependent Freundlich sorption constant relating

desorption and adsorption coefficients,

n(x) = the temperature-dependent constant which has been observed to

be greater than one.

Although this expression is an empirical equation, it is sufficient for

describing much of the experimental data.

Another expression, the Langmuir equation, can be written as

where K (x) = the temperature-dependent Langmuir sorption constant relating

desorption and adsorption coefficients,

S ^ - saturated surface concentration, sat

A-6

This expression was deduced from a definite hypothesis on the mechanism of

the sorption process, i.e.. that the rate of sorption is dependent on the

number of unoccupied surface sites. As the surface concentration

approaches the saturated surface concentration, the number of sorption

sites becomes vanlshlngly small, resulting in a rapid Increase of the

coolant concentration above the surface. Note that at low coverage values

the coolant concentration becomes linearly proportional to the surface

concentration.

A general expression that results in either equation, depending on the

values of the constants, can be written as

C^(x,t) K(x) S(x.t)

n(x)

- 6„iS(x,t)/S^^^ (A-8)

where 6 n1 Kronecker delta function.

6 , - 1 If n n1 1,

n1 0 if n ^ 1.

Since there may be many Isotopes of the same chemical species sorbed

on the surface, one must consider the surface concentration of each Isotope

with relation to the total surface concentration of all the isotopes of the

same chemical species. Therefore, the surface concentration terms in Eq.

A-8 must be summed over all Isotopes of the same chemical species and the

expression multiplied by the mole fraction of isotope J to obtain the

proper concentration for Isotope j. Hence, Eq. A-8 becomes

C (x.t) - K(x) 4 gS(x.t)1 n(x)

1 - «„i"Cx,t)/S sat.

S^(x,t)

J:s(x,t) (A-9)

Generally, different chemical species can be Included In the summation by

using weighted values of the constants and concentrations in Eq. A-9.

A-7

In order to keep the theoretical treatment relatively simple and

straightforward, the explicit temperature dependence of the sorption

constants in Eqs. A-6, A-7, and A-8 has been suppressed. However, most

of the experimental sorption data are reported in a form that explicitly

shows this temperature dependence. The code Input has therefore been

designed to accept these experimental constants. This is accomplished by

calculating the coolant concentration at the surface for either the

Freundlich or the Langmuir type of sorption kinetics from an equation of

the form

^ . . exp [ACP^BCP.103/T(x)] s(x.t)^^^^^"^^'^Q'/^^^)) o vx.t; „ , (A-10) ^ [1-S(x,t)/DCP'10^-*]

where T(x) " the absolute temperature of the surface, "K,

S(x,t) •• the surface concentration of the sorbed species (units vary

with experiment),

and ACP, BCP, CCP, DCP are constants. Hence, the expressions for K(x) and

N(X) in Eq. A-8 are, respectively,

K(x) - exp [ACP+BCP • 10"'/T(x)] ,

n(x) - CCP+DCP • 10^/T(x) . (A-H)

For Freundlich-type behavior, all the constants are determined from the

experimental data (the denominator will be approximately equal to one).

For Langmulr-type behavior, the constant CCP is set equal to one, the -23

constant DCP is set equal to the product of 10 and the saturated surface

concentration S ^, and the constants ACP and BCP are determined from the sat

experimental data. Note that this procedure is numerically equivalent to

the Kronecker delta formulation given in Eq. A-8. Of course, the surface

concentration tetnns in Eq. A-10 should be summed over all Isotopes of the

same chemical species and the result multiplied by the mole fraction of the

Isotope under consideration (see Eq. A-9).

A-8

In addition to simple Freundlich or simple Langmuir sorption kinetics,

certain systems exhibit Freundlich-type behavior at the higher surface con­

centrations and Langmulr-type behavior at lower concentrations; that is,

the gaseous concentration becomes proportional to the surface concentration

at the lower surface loadings. The proportionality constant Is determined

by evaluating the gaseous concentration at the critical surface loading,

CRITSC, and then dividing by the critical loading. The program will

automatically make this change in sorption behavior if a critical surface

loading, CRITSC, is supplied.

A-9