Magurele R&D community results in TRANSAT project

Nita Iulian 1, Fako Raluca1, Meglea Sorin1,
Cristian Postolache2
1RATEN CITON Atomistilor 409 Magurele, Romania
2“Horia Hulubei”, National Institute for Physics and Nuclear Engineering
Reactorului Street no. 30, Magurele, Romania
23rd International Conference “New Cryogenic and Isotope Technologies for Energy and Environment“ - EnergEn 2021
Băile Govora, Romania, October 26 – 29, 2021

Activities developed by Magurele R&D community in the TRANSAT project are briefly
described in this paper with stress on outcomes to be used in future national major
investments in the nuclear field.
Also, considering the analyse of overall results up to date on each work package the
other R&D team is proposing new R&D activities to be considered for development in
a new project proposal after the completion of the current TRANSAT activities.
The paper will show the Magurele community results within TRANSAT project mainly
in the field of the tritium permeation in materials.
The role of design & engineering team for input data mining and for potential further
implementation of experimental activities within TRANSAT (TRANSversal Actions for
Tritium) project is described.

CANDU 6 – tritium generation
Tritium is a by-product of energy production in CANDU reactors. Tritium is generated in CANDU reactors by several mechanisms
such as ternary fission within fuel elements and neutron absorption of deuterium. Tritium is produced at a rate of 2 x 1012
Bq/MW (e) in the heat transport system, but the majority (97%) comes from the moderator where it is produced at a rate of ~7.2
x 1013 Bq/MW (e). Tritium concentrations in the moderator and heat transport systems of the CANDU reactors are expected to
reach plateau levels after several years of operation (about 80 Ci/kg in the moderator systems and from 2 to 3 Ci/kg in the
Primary Heat Transport system (PHT) by the end of the reactor design life).

The radiological impact due to Cernavoda NPP operation is measured in terms of
dose for the population. Dose assessment for the population is based on the results
of the liquid and gaseous effluent monitoring program. The Environmental
Radiological Surveillance Program at Cernavoda NPP has been developed and
implemented in order to verify the effectiveness of the effluent control programs and
to be able to detect any changes in the concentrations of radioactive substances in
the environmental samples on time.
For the most accurate estimation of the impact of the operation of the plant on the
environment, the program of pre-operational environmental monitoring at Cernavoda
NPP was carried out between 1984 and 1996. The measurements carried out within
this program allowed the background radioactivity characterization at Cernavoda
and the possibility of assessing by comparison the impact of the plant on the
environment.
Permanently, the results of the radiological monitoring of the environment are
compared with the results of the preoperative environmental monitoring program
carried out between 1984 and 1996. So far, no relevant changes in the radioactivity
of the environment have been detected in the Cernavoda area compared to the
period prior to the commissioning of the nuclear unit. There is a strong correlation
between moderator tritium concentration and airborne tritium emission.
References: Environmental report 2016, issued by Cernavoda NPP; Andrew Roberts et al., CTRF - Project Status Update, A
presentation to the 20th ICSI Conference “Progress in Cryogenics and Isotopes Separation” Calimanesti-Caciulata, Romania,
October 23-24, 2014; Cernavoda NPP environmental reports etc.

TRANSAT
general objectives
• Assess technologies to minimise tritium permeation at source;
• Assess technologies to capture and store tritium from treatment of
metallic waste and liquid and gaseous effluents;
Also to include:
(i) An assessment of the tritium inventory using state-of-the-art
modelling tools;
(ii) Refinement of the knowledge on outgassing, radiotoxicity,
radioecology, radiobiology, dosimetry and metrology of tritium;
(iii) Engineering solutions for detritiation techniques and waste
management;
(iv) Tritium permeation control.

RATEN CITON & IFIN HH
partners in WP1 – TRANSAT
Work Package general objectives:
• Review of the experimental activities concerning the development of the barriers against
tritium permeation and the validation of the barrier(s) based on coatings.
• Assessment of an active barrier system that prevent tritium migrating through surfaces
will be carry out.
• Treatment of the operational tritiated gases generated in the fission plants such as
plenum gas purification, He purification in gas coolant reactor and the He purification
system in the TBM from the fusion applications will be assessed and further developed.
• Proposal for a route for the separation of lithium isotopes.

RATEN CITON & IFIN HH
partners in WP1 – TRANSAT
Structure of the WP1:
WP1 has two main tasks:
• Task 1: Studies on tritium migration in view of refinement of knowledge on diffusion,
retention and release mechanisms and validation of barriers against tritium permeation
(ENEA, KIT, IIT, RATEN, IFIN-HH).
• Task 2: Treatment of the operational tritiated gases generated in the fission (plenum gas
purification, He purification in gas coolant reactor) and fusion (He purification system in
TBM ) activities and assessment of a viable route for the separation of lithium isotopes
(ENEA, KIT).

Activities developed by RATEN specialists in the first half of the project planned period were
focused on the assessment of term sources relevant for fusion and fission reactors and
potential applications for engineering activities to improve tritium management in the
nuclear open-cycle. A review of the potential tritium sources from fission operational
reactors in EU is developed considering different reactors types.

Romania: capital – Bucharest; EU member country since 1 January 2007
Cernavoda NPP

* * * - Nuclear Power Reactors in the World,
Reference data series no. 2, International
Atomic Energy Agency, 2017 Edition
* * * - OPERATING EXPERIENCE WITH
NUCLEAR POWER STATIONS IN MEMBER
STATES (2017 EDITION), Produced by the
IAEA in Austria June 2017
2016

At the end of 2007, 44 PHWR reactors (from 90 ÷ 880 MWe power output) with a
net capacity of 22.4 GWe were commissioned in the world.
The overall generation of tritium in the moderator and primary heat transport
systems in the PHWR is estimated in the literature at a maximum value of
90000TBq.Gwe-1 and that the accumulation of tritium in the heavy water, at the
end of the service life of the reactor, is foreseen to be about 7.4.105TBq.Gwe-1.
Tritium (T) is produced in the heavy-water moderator and coolant (about twenty
times more in the former) through activation of deuterium (D) by neutrons:
The average production rate of tritium in the moderator of a CANDU power reactor is about 7.5 X 1010 Bq
per kg of heavy water per year. A representative moderator volume is 281 cubic meters; a CANDU power
reactor produces approximately 2.3 X 1016 Bq/year of tritium in the moderator system.
According to referenced source: “members of the public living in the vicinity of nuclear facilities licensed by
the CNSC received very low doses of radiation from tritium exposures. Doses are typically less than 3 µSv
per year for nuclear power stations and the AECL CRL site, and less than 67 µSv per year for tritium
processing facilities. These doses are well below the public dose limit of 1000 µSv per year (1 mSv per
year).”
[source: Tritium Releases and Dose Consequences in Canada in 2006, Minister of Public Works
and Government Services Canada 2009 Catalogue number CC172-52/2009E-PDF
ISBN 978-1-100-13930-2, Published by the Canadian Nuclear Safety Commission (CNSC)

It is forecasted that trittium concentrations in CANDU reactors could rise and
reach at maturity about 80 Ci/kg in Moderator System and 2 ÷ 3 Ci/kg in the
Primary Heat Transport System. Also, it is known that Moderator System
contribute mode than 50% to tritium emissions.
In the following figures are shown tritium activity growth in the moderator
systems (left) and Primary Heat Transport (PHT) systems (right) of Embalse
(Argentina), Pt. Lepreau, Gentilly-2 (Canada) and Wolsung-1 (Korea).
[source: S. K. Sood, C. Fong, K. M. Kalyanam, K. B. Woodall – A compact, low cost, tritium
removal plant for CANDU-6 reactors, CA9900076, 2004

Moderator tritium concentrations have effect on tritium emission. Based on monitoring
data from Ontario Power Generation and Wolsong reactors there is a correlation
between moderator tritium concentration and airborne tritium emission.
The correlation for Cernavoda NPP emissions measurements and moderator tritium
level evolution is shown below.
[source: Environmental Report 2016, Cernavoda NPP]
[source: Andrew Roberts and al., CTRF – Project Status Update, A
presentation to the 20th ICSI Conference “Progress in Cryogenics and
Isotopes Separation”, Calimanesti – Caciulata, Romania, October 23-24,
2014]

Example: Reported tritium emissions to the environment due to
CANDU reactors in Canada
Tritiated water inventories and tritium management are considered
important RD directions, not only by fission community but, also, by the
fusion one. Many studies focused on tritium management aspects were
performed worldwide based on extensive RD programs, for:
- Technologies for heavy water separation, technical analysis and
tritium analysis;
- Tritium handling facility, devoted to the investigation of problems
related to the operational safety and fusion power plants;
- Hydrogen isotope distillation;
- Tritiated waste management etc.
[source: Tritium Releases
and Dose Consequences in
Canada in 2006, Minister of
Public Works and
Government Services
Canada 2009 Catalogue
number CC172-52/2009E-
PDF ISBN 978-1-100-
13930-2, Published by the
Canadian Nuclear Safety
Commission (CNSC)

Tritium monitoring is a crucial aspect and main monitoring requirements refer to the surface contamination of solids as well as of gas
and liquid media. In case of CANDU reactors several streams of waste, contaminated with tritium, are generated during operation and
they are as follows:
- Primary waste streams consisting from:
o Spent components from reactor, contaminated by heavy water;
o Various solid materials and organic liquids from maintenance and repairs (e. g. contaminated
equipment, tubes, tools, hydraulic oil and/or lubricants etc.);
o Various solid materials from decontamination and contamination control, laboratory and cleaning
operations (e.g. clothes, masks, spent filters and resins, concrete and wood scrap, towels, gloves etc.).
- Secondary waste streams consisting from:
o Tritiated water concentrates generated as residual effluents from clean-up and recovery
systems of heavy water moderated reactors;
o Tritiated solid waste generated during the detritiation treatment;
o Tritiated liquid waste generated during the decontamination of primary waste fluxes.

The permeation stand, accomplished by IS System Romania,
consists of three concentric tubular chambers made 316 SS and a
thermostatic thermal mantle.
• The inner chamber was designed for Tritium storage.
• The intermediate chamber is dedicated for determination of the
tritium permeated through the stainless steel wall.
• The outer chamber, designed for evaluation of secondary
permeation phenomena.
The volume of the inner and intermediate chambers was determined
by the picnometric method.
The obtained values were:
• 19.689 g (ml) for inner chamber and
• 5.202 g (ml) for intermediate chambers
The volume of outer chamber was not determined, being ignored in
the experiments performed.
Designing and building of the
permeation stand.

Accomplishment of the
experimental assembly
Schematic view of the tritium permeation facility
PS - Permeation stand with 3 chambers, HM- Heating mantle, TC-
Temperature controller, V1-Valve for He:T mixture, V2- Valve for D2 H2O
mixture, V3 and V4 – Valve for air recirculation in outer chamber, TGM-
Tritium gas monitor RS400-HTO, Overhoff, USA with 4 ionising chambers in
cross with 200 ml, measurement range: 1 ... 19999 µCi/m3
Picture of tritium
permeation facility

LOADING THE INTERMEDIATE CHAMBER WITH
DEUTERIUM SATURATED WITH WATER VAPOR
The intermediate chamber was filled with D2 gas saturated with water
vapor.
The filling protocol
• 5 ml of purified water were introduced into the saturator and
cooled with liq. N
• The entire installation was vacuumed using vacuum pump.
• The connections with D2 supply and vacuum pump was closed
and saturator was heated at 80° C. The pressure of water vapors,
determined at manometer has grown to 2 kPa.
• The D2 cylinder was open. The D2 pressure was regulated by the
Hg valve at 102.66 kPa.
• By opening the connecting valve, D2 was pumped into the
permeation stand.
DS- D2 supply (cylinder with
pressure regulator), HgV-
Valve with Hg, V- Valve, S-
Saturator, QW-Quartz wool,
EH – Electrical heater, M-
Manometer, VP- Dry vacuum
pump Anest Iwata Japan
(ultimate pressure 10 Pa),
PS- Permeation system

In the experiment was used a glass ampoule with T2 gas with 740 GBq, purchased from
the USSR in 1977. Theoretical composition of the gas mixture after 43 years of storage is
T2 3 % and 3He 97%. The theoretical T2 activities is 62 GBq.
OBTAINING OF T2 :3He MIXTURE AMPOULES
TA- USSR Tritium ampoule, TP- Toepler pump, V- Valves,
3WV- 3 way valve, M - Breaking magnet, VP –Dry Scroll
Vacuum Pump SH 110 Varian France, ultimate pressure 6
Pa, AC- Compressed air, T:HeA- T2:3He mixture ampoules
with two
The T2 :3He mixture was transferred in 4 ampoules using a
pump with Hg piston, Toepler type and sealed at flame.
The volume of ampoules and connecting lines were
determined and used in computing of theoretical activity per
each ampoule (~3 GBq).
The residual gas mixture was pumped in initial ampoule
and sealed at flame.
3 ampoules will be used in experiment and one is
dedicated for experimental determination of T activity.

LOADING THE INER CHAMBER
WITH T2:He MIXTURE
HeS- He supply (cylinder with pressure
regulator), HgV- Valve with Hg, V- Valve, Dry
Scroll Vacuum Pump SH 110 Varian France, TA-
T:He ampoule, M- Breaking magnet, PS-
Permeation system
The inner chamber was filled with T2:3He:4He mixture.
The filling protocol
• The two routes of installation was vacuumed using
SH 110 vacuum pump.
• The 4He cylinder was open. The 4He pressure was
regulated by the Hg valve at 102.66 kPa.
• The connections with vacuum pump was closed
and section to the inner chamber of the T:He
ampoule was broken using the magnet.
• By broking of the section to the He supply, 4He and
T2:3He mixture was pumped into the inner chamber
of permeation stand.

EXPERIMENTAL DETERMINATION
OF TRITIUM PERMEATION
The parameters used in the experiments
were:
• Work temperature: 2500 C
• Dynamics of sampling: at 48 h with
exception of weekend (72 h)
PS - Permeation stand, HM- Heating mantle, TC-
Temperature controller, V1-Valve for He:T mixture, V2- Valve
for D2 H2O mixture, V3 and V4 – Valve for air recirculation in
outer chamber, TGM- Tritium gas monitor RS400-HTO, TRV-
Tritium Retention vial (volume 1.413 ml), DV – Dead volume
(1.848 ml), VP- Vacuum pump Anest Iwata Japan)

The T permeated in the outer chamber was determined by recirculating the air through the ionization
chamber of the TGM.
The experimentally determined activity was corrected taking into account the total volume analysed
(Ionising Chamber, pump and connection route = 0.25 l).
The T permeated in the intermediate chamber was determined by sample extraction using a pre-
vacuumed tritium retention vial.
Activities of extracted tritium were determined by oxidation in O2 atmosphere in presence of CuO catalyst,
quantitative retention of resulted HTO and activities determination at Liquid Scintillation Spectrometer
TRICARB TR2800 PE.
The extracted activities were used in computed permeated tritium activities.
SAMPLING

FACILITY FOR DETERMINATION OF TRITIUM ACTIVITIES
OS- Oxygen supply, FM-Fluo-meter, TRF- Tritium
retention vial, QT- Quartz tube which contain CuO
catalyst wire, TF- Tubular furnace RT 50-250/11 with
B130 temperature controller Nabertherm Germany,
THORV- HTO retention vials
Combustion protocol
• T Oven / CuO catalytic bed: 8000C
• Oxygen flow rate: 4 l / min
• Oxidation time:1 h
• HTO retention: Fresh distillate water (4 vials with
5 ml each)

RESULTS
Permeation of T in intermediate chamber
P
V
P
DV
V
V
Sf
Bq
A
Bq
A
TRF
TRF
IC
LSC
IC







)
(
]
[
]
[
Determination of T permeated activity
were AIC is activity of T in intermediate chamber, ALSC
is activity determined at LSC, Pf is the sampling
factor = 5 (4 ml/20 ml), VIC is the volume of the
intermediate chamber, VTRF is the volume of the
Tritium retention vial, DV is the dead volume and P is
the computed pressure.
After each extraction, the pressure in the intermediate chamber was recomputed in accordance with the gas
volume extracted ( (VTRF + DV)x P) in the previous stage

RESULTS
Permeation of T in outer chamber
Determination of T permeated activity
6
3
10
]
[
]
/
[
]
[
ml
V
m
Ci
A
Bq
A TGM
OC



were AOC is activity of T in outer chamber, ATGM is
radioactive concentration showed by TGM, V is the total
volume of the recirculation area (250 ml) and 106
represent the correction factor between ml and m3.

ENDING OF THE EXPERIMENT
After 9 days, the heating was stopped, the tritium from the inner chamber
was extracted into the vial with two valves (volume 40.08 ml) using the
Toepler pump (extraction yield > 95%).
The activity of tritium from the inner chamber and from the virgin vial will be
determined using an oxidation system similar to the one presented above.
The three chambers were preliminary decontaminated at 2500 C by vacuum.
In the next step, the decontamination of permeation stand will be realized
using water stream and hydrogen and high temperature.
The decontamination efficiency and background will be determined by using
the permeation protocols previously described (using hydrogen)

PRELIMINARY CONCLUSIONS AND PROPOSED NEXT WORKS
The preliminary experiment shown a exponential accumulation of T in intermediate and outer chambers.
The determination of accumulation of tritium in outer chamber do not present problems.
De determination of tritium accumulation in intermediate chamber present difficulty because the volume is
small and extraction protocol is associated with representative gas loss because the dead volume is high.
The next experiment will be started using following modified parameters:
• Temperature 3500 C
• Pressure of T:He and D:water vapors mixtures more than 300 kPa (3 atm).
• The dead volume will be reduced by modifying the installation and the sampling vial
• Time between two determination 24 h.

Magurele R&D community results in TRANSAT project

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