Embed Size (px)
Citation preview
Development of Differential Die-AwayInstrument for Characterization of Swedish
Spent Nuclear Fuel
TOMAS MARTINIK
Licentiate ThesisDivision of Applied Nuclear Physics
Department of Physics and AstronomyUppsala University
2015
AbstractThe Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project was established in 2009 by the U.S. Depart-
ment of Energy with main objective to investigate, and potentially develop and test new technologies for spent nuclear
fuel (SNF) characterization. In Sweden the SNF is currently being considered to be verified and encapsulated in can-
isters and deposited into a geological repository. The need for an independent instrument for SNF verification by the
Swedish operator turned into the collaborative effort with NGSI-SF to develop an instrument for future deployment in
Sweden.
One of the techniques investigated within this project is the differential die-away (DDA) technique, which following
the theoretical investigation by means of high fidelity Monte Carlo simulations indicated the potential to be applied for
determining of various spent fuel assembly (SFA) parameters.
This work introduces the first deployable DDA instrument which was designed to be used for characterizing of
Swedish SFAs currently stored in the Central Interim Storage Facility for Spent Nuclear Fuel (Clab). All the instrument
components relevant for DDA design functionality were evaluated to ensure reliable operation in Clab. Although most
of the components were tuned with special consideration given to concerns from the operator (The Spent Nuclear Fuel
and Waste Management Company) , several post-simulation modification of the design were made. These modifications
are described in this work.
A complementary study of the detector responses to asymmetrically burned SFAs indicated a different detector re-
sponses, depending on which of the four different orientations was used to assay individual SFAs. This study illustrated
the sensitivity of detectors with respect to the SFA orientation if there is a strong burn-up gradient across the SFA and
hence a strong asymmetry in isotopic distribution in the SFA. In addition, the study of asymmetry provided the informa-
tion on different operational scenarios of the DDA instrument. The DDA instrument may provide general information
about the complete SFA as well as give local information about certain parts of the SFA.
List of papers
This thesis is based on the following papers, which are referred to in the text by their Roman
numerals.
I T. Martinik, V. Henzl, S. Grape, S. Jacobsson Svärd, P. Jansson, M. T. Swinhoe, and S. J.
Tobin: "Simulation of Differential Die-away Instrument’s Response to AsymmetricallyBurned Spent Nuclear Fuel", Nuclear Instruments and Methods A 788, 79-85, 2015,
doi:10.1016/j.nima.2015.02.058, Los Alamos National Laboratory Report
LA-UR-14-21574 (2014).
II T. Martinik, V. Henzl, S. Grape, P. Jansson, M.T. Swinhoe, A. V. Goodsell, and S. J. Tobin:
"Design of a Prototype Differential Die-Away Instrument Proposed for Swedish SpentNuclear Fuel Characterization", Submitted for publication in Nuclear Instruments and
Methods A, October 2015. Los Alamos National Laboratory Report LA-UR-15-28331
(2015).
III T. Martinik, V. Henzl, S. Grape, P. Jansson, M.T. Swinhoe, A. V. Goodsell, and S. J. Tobin:
"Development of Differential Die-Away Instrument for Characterization of Swedish SpentNuclear Fuel", Proceedings of 37th ESARDA Symposium, Manchester, United Kingdom,
2015, Los Alamos National Laboratory Report LA-UR-15-23345 (2015). .
IV T. Martinik, V. Henzl, S. Grape, S. Jacobsson Svärd, P. Jansson, M. T. Swinhoe, and S. J.
Tobin: "Characterization of Spent Nuclear Fuel with a Differential Die-Away Instrument",
Proceedings of IAEA Symposium on International Safeguards: Linking Strategy,
Implementation and People, Vienna, Austria, 2014, Los Alamos National Laboratory
Report LA-UR-14-28074 (2014). .
Reprints were made with permission from the publishers.
Contents
Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ix
1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
1.1 Nuclear Safeguards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
1.2 The nuclear fuel cycle in Sweden . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
1.3 NGSI-SF project . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
2 Description of the DDA technique . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
3 DDA design for Clab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
3.1 Models of spent fuel assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
3.2 Limiting conditions for the DDA instrument . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
3.3 Description of individual instrument components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
3.4 MCNP simulations of the first prototype for deployment in Clab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
3.4.1 Neutron flux distribution around the SFA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
3.4.2 Determination of detector count rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
3.4.3 Lead shielding for 3He detectors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
3.4.4 NG shielding and tailoring material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
3.4.5 The effect of a Cd liner around the SFA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
3.5 Additional and/or undecided considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25
3.6 Modifications of the design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25
4 Detector response to asymmetrically burned SFAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
5 "Working concepts" of DDA alteration for different purposes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
6 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
7 Outlook . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
Acknowledgement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37
References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39
Abbreviations
BU Burn-up
BWR Boiling Water Reactor
Clab Swedish Central Interim Storage Facility for Spent Nuclear Fuel
Clink Encapsulation Facility
cps counts per second
CT Cooling Time
DA Destructive Assay
DDA Differential Die-Away
DDSI Differential Die-Away Self Interrogation
DN Delayed Neutron
Euratom European Atomic Energy Community
FFA Fresh Fuel Assembly
IAEA International Atomic Energy Agency
IE Initial Enrichment
KBS-3 Concept for Geological Storage
M Multiplication
MCNP Monte Carlo Transport Neutron Particle
MCNPX Monte Carlo Transport Neutron Particle eXtended
N/A Not Available
NDA Non-Destructive Assay
NG Neutron Generator
NGSI Next Generation Safeguards Initiative
NGSI SF Next Generation Safeguards Initiative Spent Fuel project
NNSA National Nuclear Security Administration
PWR Pressurized Water Reactor
S/B Signal-to-Background
SFA Spent Fuel Assembly
SFL Spent Fuel Library
SKB Spent Nuclear Fuel and Waste Management Company
SNF Spent Nuclear Fuel
SSM Swedish Radiation Safety Authority
U.S. United States
1. Introduction
The fission of nuclear material (uranium, plutonium, thorium, etc.) has been recognized as an
effective source of electricity production and is today widely used in a safe and controlled way
inside the nuclear reactors. However, nuclear material which is used in the form of fabricated
fuel pellets and fuel rods in nuclear reactors must be handled in a controlled way. The framework
which aims to ensure that the nuclear material is not misused is known as "nuclear safeguards"
and is further described in Section 1.1.
Sweden produces a lot of electricity from nuclear power plants as summarized in Section 1.2.
After irradiation, the spent nuclear fuel (SNF) contains a significant amount of nuclear material.
The nuclear material needs to be verified, and a common way to verify spent nuclear fuel is using
non-destructive assay. The Next Generation Safeguards Initiative Spent Fuel project (NGSI-SF),
which aims to develop new non-destructive techniques for assaying SNF, and is further described
in Section 1.3.
The theory of one such non-destructive technique — differential die-away (DDA) — is de-
scribed in Chapter 2. The measurement and analysis capabilities of DDA instrument have been
evaluated. Chapter 3 describes the proposed instrument designed to fit into the Swedish Central
Interim Storage Facility for Spent Nuclear Fuel (Clab). Section 3.6 summarizes changes intro-
duces after the simulation studies were completed.
Results of the investigation about detector responses to asymmetrically burned spent fuel as-
semblies (SFAs) are described in Chapter 4. Additional designs for the DDA instrument are
presented in Chapter 5.
1.1 Nuclear Safeguards
In December 1953, U.S. president Eisenhower launched the Atoms for Peace program which
was supposed to create the first policy related to handling nuclear material and technologies [1].
During the following years, several non-proliferation agreements between several countries were
constructed. Based on these agreements, in 1970, the Non-Proliferation Treaty was created to 1)
prevent the dissemination and nuclear weapons, 2) to promote the nuclear disarmament, and 3)
to promote the peaceful use of nuclear energy [1]. Article 3 of this treaty is related to nuclear
safeguards [1].
Nuclear Safeguards may be described as a framework with the main objective of ensuring
that nuclear material and technologies, which are widely used in many industrial sectors, are
not misused for non-peaceful purposes. To ensure this, several international as well as national
organizations are responsible for carrying out regular controls and verifications. For example,
the International Atomic Energy Agency (IAEA) and the European Atomic Energy Community
(Euratom) operate on international and European level, respectively. On the state level in Sweden,
the Swedish Radiation Safety Authority (Strålsäkerhetsmyndigheten - SSM) is as well responsible
for inspection and verification of nuclear material. Because nuclear material may be subjected
to diversion for non-peaceful applications, its accurate verification is required by independent
regulatory bodies such as IAEA and Euratom. The control and verification of nuclear material
may be carried out using several tools. The techniques used for verification may be categorized,
and the primary methods are destructive assay (DA) and non-destructive assay (NDA) of samples.
While the DA techniques modify, change or in some way impact the item under investigation,
NDA techniques do not alter the assayed item.
11
In terms of safeguards, two observables are routinely determined for nuclear material samples:
the isotopic composition of sample (qualitative analysis) and the mass of nuclear material (quan-
titative analysis). These two properties are used to properly characterize and verify both unknown
as well as well-defined samples of radioactive and nuclear material.
1.2 The nuclear fuel cycle in SwedenThe nuclear fuel cycle consists of three main parts. The front end of the nuclear cycle typically
deals with fuel fabrication and preparation for use in a nuclear reactor. During the second part of
the cycle, reactor irradiation, the isotopic composition of the fuel changes rapidly. One parameter
describing a fuel property before irradiation in a reactor core is the initial enrichment (IE). After
several irradiation cycles in the reactor core, the fuel parameter used to characterize the fuel is
the burn-up (BU). This parameter is related to the power utilization of the fuel. Typical values
for fully burned SFAs are 30-45 GWd/tU. After the reactor operation, the fuel is transferred to
wet storage pools to cool. The parameter which defines the time from the extraction of the fuel
from reactor core until present time is referred to as cooling time (CT). While the fuel in some
countries resides in water pools next to the nuclear reactor before its transport to another storage
facility, the fuel from the Swedish nuclear power plants, i.e. 7 Boiling Water Reactor (BWR) and
3 Pressurized Water Reactors (PWR) currently in operation1, is transported to the Clab. In Clab
all the SNF is stored in water with an active cooling system.
There are several options on how to proceed with future SNF handling: 1) reprocess the SNF
and use the nuclear material for production of new SFAs for irradiation in nuclear reactors, 2)
deposit the fuel into a final repository, or 3) postpone the decision. In Sweden, a decision by the
company responsible for the management of spent nuclear fuel and nuclear waste, SKB, was made
to aim to deposit the fuel in a geological repository. The location around the Forsmark nuclear
power plant has been chosen as the most suitable option for the geological repository among all
those considered.
The Spent Nuclear Fuel and Waste Management Company, SKB, has suggested a concept for
geological storage called KBS-3. This concept is based on three barriers separating the nuclear
fuel from the biosphere: 1) a copper canister, 2) bentonite clay, and 3) the granite bedrock. Before
the fuel can be loaded in the copper canisters and later deposited in the geological repository, it is
important to verify the SNF parameters.
Given the desire for independent verification of all SFAs before encapsulation, the SFAs are
expected to be assayed before insertion into the specially designed canisters. These canisters
are designed to accommodate either 12 BWR or 4 PWR SFAs (Fig. 1.1). The inner mechanical
supporting grid of the canister consists of cast iron. The copper was primarily selected as a
corrosion resistant material, but it also has the benefit of being a good conductor, so the SFAs can
cool more rapidly. These canisters will be deposited 500 meters below into the granite bedrock,
and the spaces between the canisters and the tunnels between canister locations will be filled with
bentonite clay.
In order to encapsulate the SFAs before their shipment to the site for the geological repository,
a dedicated encapsulation facility called Clink will be constructed and start its operation around
the year 2025. The encapsulation and canister loading will be performed in Clink, where the final
verification of the SFA parameters will also be performed. The Clink facility will be built in con-
nection to the currently existing Clab facility in Oskarshamn. Fig. 1.2 is a schematic illustration
of the existing and planned nuclear waste facilities.
The operation of Clink is expected to end approximately 25 years after the termination of the
last nuclear power plant in Sweden, i.e. around the year 2070. Eventually, it is expected that
more then 40 000 BWR SFAs and around 8000 PWR SFAs will be encapsulated and deposited
into the geological repository. The SFAs will vary both in terms of fuel type and dimensions, as
well as in terms of fuel parameters (IE, BU and CT). The majority of the SFAs stored in Clab in
1In addition, two boiling water reactors have been already shut down.
12
fuel assembly
PWRBWR
copper canister
iron
Figure 1.1. Schematic cross-sectional (top view) of the KBS-3 canister. Two different inserts are shown forboth BWR and PWR [2, 3].
nuclear power plant
fuel assembly
repository
temporary storage
VERIFICATIONOF PARAMETERSNEEDED
Figure 1.2. Schematic illustration of the back-end of the Swedish nuclear fuel cycle [2, 4].
2007 were fully burned SFAs. Particularly, the average burn-up of stored SFAs dated in 2007 was
34 GWd/tU and 41 GWd/tU for BWR and PWR SFAs, respectively. More detailed information
about the SFAs currently stored in Clab can be found for example in [5].
1.3 NGSI-SF project
In 2009, the National Nuclear Security Administration (NNSA) within the U.S. Department of En-
ergy established a project called The Next Generation Safeguards Initiative (NGSI). The NNSA
"concluded that a comprehensive initiative to revitalize the international safeguards technologyand human resource base by leveraging U.S. technical assets and partnerships was urgentlyneeded to keep pace with demands and emerging safeguards challenges" [6]. The NGSI is built on
five pillars: 1) Policy and Outreach, 2) Concepts and Approaches, 3) Technology Development,
4) Human Resources, and 5) Infrastructure Development.
13
Under pillar 3), The Next Generation Safeguards Spent Fuel Project (NGSI-SF) began its ef-
fort in 2009. The main objective of the project was to develop and test integrated technologies
designed to meet the safeguards challenges in the next 25 years. Upcoming NDA technologies
may strengthen the field of nuclear safeguards and provide the technical toolkit for safeguards
inspectors and others to determine the following technical goals more effectively and accurately:
1. Detect replaced or missing pins from a given SFA to confirm its integrity and to deter diver-
sions,
2. Estimate plutonium mass and related plutonium and uranium fissile mass parameters in
given SFAs, and
3. Verify the IE, BU, and CT of SNF.
Initially, 14 different non-destructive techniques were selected and scrutinized in terms of their
applicability for SNF parameters verification and determination [7]. Following the simulation-
based evaluation, the most promising techniques were selected for further detailed investigations.
Two additional technical goals that enables access to the Clab facility in order to meet facility-
specific needs have been added since 2009: direct measurement and indirect estimation of heat
content and estimation of reactivity (in terms of multiplication determination).
14
2. Description of the DDA technique
The methodology of the differential die-away (DDA) technique is well known and was introduced
several decades ago. Traditionally, the DDA technique was applied to characterize nuclear waste
drums and the analysis was performed in a different way as compared to the one within the NGSI-
SF project. Our new approach is based on the traditional DDA features; however, we shifted the
interest in the neutron population evolution and signal detection to earlier time scales before the
neutron thermalize, rather than the traditional method of after the injected neutrons thermalize.
The DDA technique is one of the NDA techniques which is being investigated within the NGSI-
SF project for possible applications in SNF analysis. The DDA technique is based on active inter-
rogation of nuclear material items by external neutrons. The external pulse source, e.g. neutron
generator (NG), injects neutrons which interrogate the content of an assayed item. The DDA
technique can be used to assay both fresh fuel assemblies (FFAs) and SFAs.
The highly energetic neutrons that are delivered by the NG into the SFA penetrate deep the
SFA and create secondary neutrons primarily by means of induced fission. These secondary neu-
trons are then detected by a set of detectors that surround the SFA. The detection time distribution
is analyzed to deduce information about SFA properties. For example, a typical distribution of
neutron detection times, following a delivery of a single neutron pulse of 10 μs is illustrated in
Fig. 2.1 [Paper I]. It can be seen that the time distribution peaks approximately around 15-20 μsafter the interrogation starting point. The detected neutrons in an early time domain (<100 μs)
primarily consist of "burst neutrons", i.e. neutrons which do not cause any fission and reach neu-
tron detectors straight from the NG or neutrons have a minimal number of scattering interactions.
In general, the contribution of "burst neutrons" is almost independent of the characteristics pa-
rameters of the assayed SFA as described in the dedicated study by Henzl, et al. [8]. Based on
the findings of [8], the burst neutrons can be treated as constant background and may hence be
subtracted from the overall detected signal. The fission part of the neutron signal typically peaks
between 50 μs and 70 μs after the end of the interrogation pulse. This part of the signal then
dies away exponentially with a constant referred to as the die-away time (τ), which is driven in
the early time domain by the amount of neutron absorbers, while in the later time domain by the
fission chain reactions.
Given the traditional definition of the DDA, the die-away time is referred to as a neutron mean
lifetime in a certain environment [9]. The DDA technique in our research is characterized by high
multiplication and a mixture of multiple neutron generations; hence the die-away time reflects the
mean time of the entire neutron population. Neutrons are simultaneously created and absorbed
as the neutron fission chain population develops and progresses through the SFA. Thus, introduc-
ing the mean life time of entire generation may be a reasonable approach for indistinguishable
neutrons originating from various generations of the fission chain reactions.
Should the interrogated fuel assembly contain no fissile material, the injected neutron popula-
tion will die away quickly, reflected by the fast decrease in neutron count rate registered by the
detectors. If fissile material is present inside the assayed item, the neutron population remains in
SFA for longer time as additional fission neutrons support the chain reactions, and consequently
prolong the die-away time.
In contrast to fissile isotopes, neutron poisons with high cross section for neutron absorption
shorten the life time of neutron population, i.e. its die-away time. The die-away time is thus
an implicit indicator of the balance between the neutron-producing fissile material and neutron-
absorbing fission products and actinides.
In traditional passive neutron counting based techniques, the neutron population evolution in
time can be approximated by a single exponential, with the decay constant being the die-away time
15
s]µtime [0 100 200 300 400 500 600 700 800 900 1000
dete
cted
neu
tron/
sour
ce n
eutro
n
-710
-610
-510
-410neutron generator
fission neutrons
total DDA signal
neutron generator
fission neutrons
total DDA signal
sµ100-200
sµ200-500
sµ500-1000
Figure 2.1. Simulated distribution of neutron detection times as registered by all DDA detectors for a 4% IE, 15 GWd/tU BU, 5 y CT PWR SFA. The evolution of the total neutron signal (dotted line), fissionneutron signal (solid line) and burst neutron signal (dashed line) is shown up to 1000 μs after the start ofthe neutron interrogation pulse. Three different time windows (100-200 μs, 200-500 μs and 500-1000 μs)typically used in the analysis are also demarcated.
[9]. In the case of DDA, the neutron population evolution is more complex, because the spatial
as well as energy distribution undergo dramatic changes along the way from the NG through the
SFA to the individual detectors. Because it cannot be approximated by a single exponential, the
analysis of the DDA signal is performed in limited time domains, where a single exponential
approximation can be justified. Our attention was primarily focused on the time domain of 100-
200 μs, which in previous studies [8, 10] has been identified as the most promising time domain to
extract unique information about SFA multiplication, IE, and BU. This time window is dependent
on the instrument design and may therefore change for a different design.
Another important quantity that is characteristic for each SFA and can be measured by the DDA
technique-based instrument is the multiplication (M) [8]. Multiplication can be either "passive"
or "active" in case of the DDA technique. When defined as the number of neutrons produced in
the SFA and the surrounding setup per incoming neutron from the NG, it is referred to as activemultiplication. This is however linearly related to passive multiplication that is defined as the
number of neutrons produced in the SFA per neutron originating inside the SFA, primarily in
spontaneous fission.
In general, the multiplication, be it passive or active, is closely related to the BU of the SFA.
Fresh fuel or fuel with low BU typically has a high fissile content and is thus characterized by
a high multiplication in contrary to the highly burned fuel with a low multiplication factor. In
addition, the multiplication reflects the isotopic composition of the SFA, hence it is dependent on
IE and CT as well. In general, the average multiplication of a SFA is implicitly defined by all these
parameters and tends to increase with higher IE, while decreases with higher BU and longer CT.
In this study, a net multiplication calculated in Monte Carlo Transport Neutron Particle eXtended
(MCNPX) code [11] corresponds to the "active multiplication" described above.
The principal advantage (and also challenge) of the DDA technique for SNF analysis is its
variety in determination of fuel parameters. The DDA instrument is expected to measure multi-
plication and total fissile mass (239Pue f f ); and determine the total Pu mass, IE, BU and potentially
identify the illicit replacement (or missing) fuel pins.
The studies illustrate that the DDA instrument may be designed to deploy with other NDA
techniques [12]. An advantage of the DDA instrument is also the high maturity of the hardware
[7], whereas the the high acquisition costs can be considered a disadvantage [7].
16
3. DDA design for Clab
In this chapter, the DDA design as simulated in MCNPX and suggested for deployment in Clab
is described in terms of its individual components. As the design, and potentially also the final
instrument, is subject to combined deployment with a number of other NDA instruments, several
limitations have been considered during the investigation. Section 3.2 lists the main limitations
taken into account in the simulations. The design itself as simulated and suggested is described
in Section 3.3, where the crucial components are further described. More information on how the
selection of the individual components was made is described in Section 3.3. This topic is also
covered in Paper II.
Although the MCNPX simulations were performed for a design specifically tuned for deploy-
ment in Clab, the design had to be further modified due to several mechanical and other con-
straints. Thus, the differences between the simulated design illustrated in Section 3.3 and the one
which was actually submitted for approval by SKB have been summarized in Section 3.6 and also
in Paper III. All the main changes implemented into the proposed design are discussed in that pa-
per together with possible positive and/or negative impacts on the DDA instrument functionality.
3.1 Models of spent fuel assemblies
The simulations using the MCNPX calculation code are done with so-called "MCNP models" of
SFAs. These models reflect a collection of SFAs with varying fuel parameters (IE, BU and CT)
and are defined by the isotopic composition given by the material cards of the MCNP input deck.
Models vary according to which procedure was applied in the fuel depletion step. Each set of
models created with the same pattern or with the same purpose is represented by a separate spent
fuel library (SFL). More details about the SFLs can be found in [13, 14].
While some of the fuel models represent realistic fuel irradiation scenarios, the others were
created to test particular features of the DDA performance or other techniques being investigated
under the NGSI-SF project. The distinguishing characteristics of a number of different SFLs are
summarized below. If not otherwise stated, the models were created for standardized Westing-
house 17 x 17 PWR SFAs [15].
• SFL1: The first spent fuel library which was developed. It consists of 64 SFAs with IE 2-
5%, BU between 15-60 GWd/tU and CT of 1, 5, 20, and 80 years. SFAs were modeled with
infinite reflective boundaries which reflects the position of the SFA inside the reactor core.
• SFL2a: The first "realistic" SFL. It has been simulated reflecting a more real word shuffling1
scheme using 1/8 of the reactor core. It consists of 77 SFAs in total created by shuffling
scheme #12 with IE of 2-5%, BU of 15-60 GWd/tU and CT of 14 days, 1, 5, 20, 40 and 80
years. Unlike SFL1, SFL2a does not contain the unrealistic SFAs such as 45 GWd/tU and
IE of 2 or 3%, or SFAs with 60 GWd/tU BU and IE less than 5%.
• SFL2c: This library represents an extension of SFL2a with additional shuffling schemes3
for fuel depletion. It covers the same span of SFAs as the SFL2a, but only for an IE of 4%.
• SFL2e: This library was created to explore the effects of calibration errors in individual
NDA instruments. In total, 9 SFAs from SFL2a were arbitrary modified by increasing or
1A pattern under which the SFAs are loaded and restructured during their time inside the nuclear reactor core.2SFAs which were created according the realistic shuffling scheme #1 are part of SFL2a. This Shuffling scheme was
later complemented by additional shuffling schemes which are part of SFL2c.3SFL2c contains the SFAs generated by using shuffling schemes #2 and #3. These two shuffling schemes are currently
considered "unrealistic"; however, some of them were used in the nuclear power plants decades ago.
17
decreasing the uranium and plutonium content by 3% or by changing the amount of neutron
absorbers ± 30%.
• SFL3: It consists of 12 SFAs with 4% IE; 15, 30, 45 GWd/tU BU; and 1,5,20 and 80 y CT
which were depleted similarly as SFL1, but under different environmental conditions, i.e.
temperature, pressure, moderator density, boron concentration and the presence of burnable
absorber rods.
• SFL4: It is called the "mystery library" and was developed in order to test the instruments
performance against (to the user) unknown SFAs.
• SFL5: This library is dedicated to BWR SFAs. The purpose of this library was to test basic
irradiation parameters of BWR spent fuel. Vertically, the SFAs are divided into 25 segments
representing regions of different burn-ups. SFL5 consists of 75 different BWR SFAs.
Within the scope of this study, we limited our use of SFLs to SFL2a and SFL2c. More SFLs
were developed with different purposes in addition to the ones listed in this work.
3.2 Limiting conditions for the DDA instrumentEven though the DDA instrument was studied independently of other techniques, in reality some
components need to be shared with other NGSI-SF techniques because some of them are planned
to be deployed and tested in the same facility.
Given the operator’s demands and limitations [16], not all the components could be selected
as the best available option, but as a pragmatic solution in the given situation. The detailed de-
scription of the limitations can be found in Paper II. The main limitations are also summarized
here:
• dimensions of the external frame (illustrated as red in Fig. 3.1) for combined deployment
with the Differential Die-away Self Interrogation (DDSI) instrument [17],
• dimensions of the inner insert (illustrated as yellow in Fig. 3.1 [17]),
• use of 3He rather than fission chambers.
External Frame
Insert
Figure 3.1. An illustration of the external frame (red color) and insert (yellow color) used for the DDSIinstrument [17] which are expected to be shared for both DDA and DDSI instruments.
On the top of these limitations, SKB also preferred to select the lower intensity NG (108 n/s)
rather than the stronger intensity one (1010 n/s) due to fewer necessary requirements to accommo-
date4 it in the water pool at Clab.
4The lower intensity NG does not need an active cooling system, unlike the stronger intensity NG.
18
3.3 Description of individual instrument componentsThis section introduces simulations performed using the MCNPX [11] code in order to design
individual components of the instrument as proposed for deployment in Clab. Taking into account
the complexity of the instrument, different types of simulations were done to design a functioning
and deployable instrument. The Clab design is described in more detail in Paper II.
A B C D
E
F
G
P
O
H
I
J
K
L
M
N
Q
R
S
x
y
H2O
#1 #2 #3 #4
#5
#6
#7
#8
#9#10#11#12
Figure 3.2. Schematic drawing of the horizontal cross-sectional view of the DDA instrument design pro-posed for test deployment in Clab. Labels correspond to the components listed in Table 3.1.
The DDA instrument can be used in two different modes. The first mode determines the fuel
parameters while the second mode can be used to determine partial defects (i.e. whether a fraction
of the fuel pins have been diverted or substituted).
The design as simulated in MCNPX and proposed for deployment in Clab is illustrated in
Fig. 3.2 (horizontal view) and Fig. 3.3 (vertical view). The dimensions of the main components
(which are discussed in later subsections) are listed and explained in Table 3.1
3.4 MCNP simulations of the first prototype for deployment in Clab
3.4.1 Neutron flux distribution around the SFA
The neutron flux serves a first indication of the expected neutron count rates on the detectors,
which is important for the design of the detectors and electronics operating in the expected neutron
field.
The NG is foreseen to be operated with the lowest achievable duty cycle [19], which in the
case of the P385 NG is 5%. The pulse length and pulse frequency are then the key parameters
which require consideration when determining the expected count rates. As described in [14],
for a given NG intensity and fixed duty cycle, the pulse length and interrogation frequency are
the parameters which dictate the number of neutrons released in a single pulse. Consequently,
the longer the pulse, the higher the number of neutrons released during one pulse. Therefore, the
19
JA
BC L
T
O
U
V
E
F
G
S
R
z
x
H2O
Figure 3.3. Schematic drawing of the vertical cross-sectional view of the DDA instrument design proposedfor test deployment in Clab. Labels correspond to the components listed in Table 3.1.
absolute magnitude of the active signal is proportional to the length of the interrogating pulse as
well as the signal to background ratio.
The neutron flux distribution and signal-to-background (S/B) ratio were simulated for two se-
lected SFAs that represent the extreme cases in terms of both active and passive neutron emission.
The first case is characterized by the relatively high active neutron emission together with a low
passive neutron emission (background), while the second case represents the opposite extreme.
The two selected SFAs for simulation were a 5 % IE, 60 GWd/tU BU, 5 y CT SFA; and a 5 % IE,
15 GWd/tU BU, 5 y CT SFA. The active neutron flux was simulated for different interrogation
pulse lengths of 10-90 μs in order to determine how that affects the S/B ratio. The passive neutron
emission is considered constant in time as the amount of spontaneously fissioning isotopes do not
significantly change with the time scale relevant to the SFA assay. Details of this simulation can
be found in Paper II.
The results illustrated in Fig. 3.4 indicate that the maximum neutron flux for one interrogation
pulse is not expected to exceed 106 n · cm−2s−1. In reality, the count rates (cps) on the detectors
are expected to be lower as a result of imperfect efficiency of 3He detector to capture all neutrons
entering its volume.
While the active contribution for SFAs displayed in the left and right panels of Fig. 3.4 appears
to stay within the same ranges, the passive neutron flux is significantly higher in the high BU case
(left panel of Fig. 3.4). In fact, the passive emission is approximately a factor of 150 larger for
the 60 GWd/tU SFA as compared to the 15 GWd/tU SFA. The 60 GWd/tU was therefore chosen
as the limiting case SFA for further investigation in order to obtain a conservative estimate of the
S/B ratio within the expected time intervals for actual measurements.
20
Table 3.1. Specifications of individual components of the DDA instrument proposed for test deployment inClab (Paper II).
Component Dimensions Additional parametersand properties
A 3He Detector Diameter =1.25 cm; 7.5 atm pressure
Length =2.54 cm
B Polyethylene Annular thick.= 1.4 cm Density =0.96 g/cm3
C Cd liner Thickness = 0.1 cm Density =8.65 g/cm3
D Detector pitch 1 cm Filled by air
E Assembly 21.4 x 21.4 cm PWR model 17x17
F Steel Insert Thick. = 0.23 cm Density = 8.03 g/cm3
G Water gap 0.39 cm Density = 1.00 g/cm3
H Lead shielding 7.74 cm Density =11.34 g/cm3
I Lead shielding 26.96 cm Density =11.34 g/cm3
J Lead shielding 5 cm Density =11.34 g/cm3
K Tungsten 22.66 cm Density =19.25 g/cm3
L Tungsten Thick. = 3 cm Density =19.25 g/cm3
M Lead shielding 26.96 cm Density =11.34 g/cm3
N Steel enclosure 24.96 cm Density = 7.92 g/cm3
O Steel basket Thick. = 0.97 cm Density = 8.03 g/cm3
P Water gap Thick. = 0.97 cm Density = 1.00 g/cm3
Q Steel enclosure 6.74 cm Density=7.92 g/cm3
R NG target N/A [18] N/A [18]
S NG enclosure N/A [18] N/A [18]
T Lead shielding 50.72 cm Density=11.34 g/cm3
U Steel enclosure 11.28 cm Density=7.92 g/cm3
V Lead shielding 13.28 cm Density=11.34 g/cm3
s]µtime [0 50 100 150 200 250 300 350 400
]-1
.s-2
neut
ron
flux
[n.c
m
210
310
410
510
610
s]µtime [0 50 100 150 200 250 300 350 400
]-1
.s-2
neut
ron
flux
[n.c
m
210
310
410
510
610s pulseµdet #1 : 90 s pulseµdet #6 : 90 s pulseµdet #1 : 10 s pulseµdet #6 : 10
det #1 : passive det #6 : passive
Figure 3.4. Time dependent distribution of the neutron flux passing through the surface corresponding to theposition of the front (#1) or back (#6) detectors, illustrated for one interrogation pulse. Both figures displayresults for active neutron emission induced by interrogation pulses of 10 μs and 90 μs length as well as thepassive neutron emission from spontaneous fission in the SFA. Simulated data of detection probabilities areillustrated for a 5% IE, 15 GWd/tU, 5 y CT SFA (left panel) and a 5% IE, 60 GWd/tU, 5 y CT SFA (rightpanel), respectively, and are multiplied by the total number of neutrons released during one pulse given aNG duty cycle of 5%. Uncertainties of data points are smaller than the markers used (less than 0.01%).
Due to the thin layer of Cd covering the detectors, an efficient absorption of thermal neutrons
is expected. For the neutron flux simulations, the data were analyzed separately for thermal and
epithermal neutrons below 0.5 eV, and for epithermal and fast neutrons with an energies above
0.5 eV. The thermal/fast ratio has been determined for both the front (1.4 ± 0.1 %) and back
21
(9.1 ± 0.1 %) detectors in order to illustrate that the thermal fraction is higher compared to the
epithermal.
Based on the results given by the neutron flux simulation, the decision was made to keep the
detectors rather small to be able to: 1) limit the count rates which the electronics can handle, and
2) limit the vertical region of the SFA, particularly for BWR SFAs, which is assayed.
3.4.2 Determination of detector count rates
The neutron flux simulation results described in the previous section provided only approximate
estimates of the number of neutrons that enter the detector volume. The maximum count rate is
the key factor which we shoot for in order to ensure for adequate selection of electronics.
The time distributions of the estimated active and passive count rates are depicted in Fig. 3.5
for one back (#6) and one front (#1) detector. Fig. 3.5 displays the results for the limiting SFA
with a 5% IE, 5y CT and 60 GWd/tU BU.
s]µtime [0 50 100 150 200 250 300 350 400
coun
t rat
e on
det
ecto
rs [c
ps]
310
410
510
610s pulseµdet #1 : 90 s pulseµdet #6 : 90 s pulseµdet #1 : 10 s pulseµdet #6 : 10 s pulseµdet #1 : 50 s pulseµdet #6 : 50
det #1 : passive det #6 : passive
Figure 3.5. Simulated time dependent distribution of count rates following a single interrogation pulse onthe front (#1) and back (#6) detectors plotted with the constant passive background. The data representsdifferent pulse lengths (10, 50 and 90 μs) for a total NG strength of 108 n/s operated with a 5% duty cycleon a 5% IE, 60 GWd/tU BU, 5 y CT SFA (results were obtained for design depicted in Fig. 3.2 and Fig. 3.3,with the modification L=0 cm). Uncertainties for data points are smaller than markers used (less than0.01%).
It can be noticed in Fig. 3.5 that for a 10 μs pulse the active count rate on the front detector
(green line) dominates over the background only for an approximate 90 μs time interval, and both
the active (blue line) and passive (dark yellow) count rates on the back detector are the same order
of magnitude.
The results for both the front (brown line) and back (grey line) detectors illustrate that the
induced active signal dominates, after peaking, over the passive emission for at least a 100 μstime interval (available for signal analysis) approximately for interrogation pulse lengths longer
than 40-50 μs.
When the 90 μs pulse is used, the induced signal is dominant for approximately 250 μs for
both front (black line) and back (red line) detectors.
These detection rate distributions create a basis for selecting the interrogation pulse length
needed to provide a sufficiently long time interval that can be used in the analysis. In other
words, only a certain time domain can be used to obtain particular observables necessary for
characterizing a given SFA [8, 10].
22
The integrals of simulated count rates on individual detectors within the time domain 75-175
μs are presented in Table 3.2 for the two (5% IE, 15 GWd/tU and 60 GWd/tU BU, and 5 y CT)
selected SFAs and the design illustrated in Fig. 3.2 (with the modification L=0 cm). The calculated
S/B ratios and the die-away times are included in Table 3.2 for all detectors as well as for only the
back detectors.
The results are calculated for a total measurement time of 5 minutes and a proposed 50 μs NG
pulse length. The 50 μs pulse was selected as a compromise between sufficiently high S/B ratio.
Shorter pulse lengths (~10 μs) give a better S/B ration, while the longer pulse lengths (~90 μs)
result in better separation of fission neutrons and interrogation neutrons.
Table 3.2. Sum of simulated active count rates on different groups of detectors in the time domain 75-175μs with a NG pulse length of 50 μs. Results are illustrated for two SFAs having 5% IE, 5 y CT and 15 or60 GWd/tU BU. The proposed measurement time is 300 s with a 5% NG duty cycle and NG strength 108 n/s(results were obtained for the design illustrated in Fig. 3.2 and Fig. 3.3 with the modification of L=0 cm).
60 GWd/tU BU 15 GWd/tU BUBACK DETECTORS
Active signal [counts] 4.68×106 ± 0.05% 1.34×107 ± 0.03%
Die-Away time [μs] 86.90 ± 1.1% 131.30 ± 1.40%
S/B ratio 2.05 ± 0.14% 937.0 ± 1.70%
ALL DETECTORS
Active signal [counts] 1.81×107 ± 0.02% 4.54×107 ± 0.01%
Die-Away time [μs] 71.30 ± 0.60% 103.10 ± 0.70%
S/B ratio 2.70 ± 0.08% 1056.85 ± 0.98%
The simulated maximal count rates for individual detectors illustrated in Fig. 3.2 (but L=0 cm)
are summarized in Paper II for three different pulse lengths. The maximal count rates (cps) on
front detector (#1 in Fig. 3.2 with the modification of L=0 cm) observed for 10, 50 and 90 μspulse lengths were 2.1×105 ± 0.2 %, 6.2×105 ± 0.1 % and 7.7×105 ± 0.1 %, respectively. This
values were found to be acceptable for adequate electronics functionality5.
3.4.3 Lead shielding for 3He detectors
The SFAs also emit gamma rays, which cannot be ignored in the DDA performance evaluation.
The particular isotopes in our SFAs models which are most responsible for gamma ray emissions
are 134Cs, 137Cs and 154Eu. As long as the 3He detectors and their electronics are sensitive also to
gamma rays, the detectors need to be shielded from these gamma rays similarly as in the case of
the DDSI instrument [17]. The DDSI instrument was found to function reliably with a 5 cm thick
lead shield between the 3He detectors and the assayed SFA.
Given the experience from DDSI, the 5 cm thick lead shield considered to be sufficient also for
the DDA instrument. To verify this, simulation of the deposited dose rate to the detector volume
was performed using the SFA with the highest gamma-ray emission (5% IE, 60 GWd/tU BU,
and 5y CT SFA). The results, as summarized in Paper II, indicated that the maximum expected
gamma dose rate on individual 3He detectors ranges from 15.3 to 26.6 ± 1.5 rad/h. These gamma-
ray exposures are considered low enough for operating with 3He because similar dose rates were
observed also for the DDSI instrument.
5The value on the back detectors is important; however, since the count rate is significantly lower on back detectors
compared to front ones, the limiting information for electronics was determined by front detectors.
23
3.4.4 NG shielding and tailoring material
The NG shield is another crucial component needed to ensure that the NG target and the associ-
ated electronics are not damaged by the gamma radiation coming from the SFAs. The deposited
gamma dose inside the NG target was simulated for the design as illustrated in Fig. 3.2 with the
dimension L=0 cm. The gamma dose for the position without any shield between the NG and the
SFA was found to be too high (2570.1 ± 0.1 rad/h) to ensure the safe operation of the NG. As
described in Paper II, the NG needed to be shifted farther away from the SFA, with tungsten filling
the gap between the NG and the SFA. The results illustrate that 3 cm of tungsten between the NG
and the SFA was sufficient to reduce the dose rates to acceptable levels corresponding to 58.7 ±0.5 rad/h as given by the NG manufacturer (approximately around 100 rad/h for the Sodern type
of NG [20]). The results of a dedicated study done by Goodsell, et al. [21] furthermore showed
that no other material to tailor highly energetic neutrons is needed.
3.4.5 The effect of a Cd liner around the SFA
The question of whether to use a Cd liner around the SFA was first raised by Lundkvist, et al
[22]. While a Cd liner turned out to be important to protect from oversampling the SFA edges by
thermalized neutrons [22], the question was whether it should be used in the Clab design or not.
Since the Clab design is very similar to the Blanc and Menlove design in Fig. 4.2, this design was
selected for a dedicated study of the value of a Cd liner in the Clab design6.
In the study, a 0.1 cm thick cadmium cover was inserted around the SFA in the Blanc and
Menlove design (Fig. 4.2). The results described in Paper II do not show any significant difference
in DDA functionality with or without the Cd liner. The evaluation was done in terms of DDA
parameter determination, i.e. whether this capability is affected with the Cd present around SFA
(right panel of Fig. 3.6) or not (left panel of Fig. 3.6). The largest difference was found to be
in the magnitude of the DDA signal and the die-away time, which both naturally decreased (by
~16-20 %) due to the presence of additional absorption material. It is also relevant that the span
of individual data points is lower with the Cd liner present (left panel of Fig. 3.6) as compared
to the situation without Cd (right panel of Fig. 3.6). The distinguishing between individual data
points may complicate further analysis for the real time measurements.
Figure 3.6. Simulated die-away time dependence on DDA signal in the time domain 100-200 μs for thedesign illustrated in Fig. 4.2 for a set of SFAs selected from SFL 2a [13] The left panel illustrates the designwith the Cd liner present around SFA, whereas the right panel demonstrates the case without the Cd linerand is equivalent to [8]. Data are collected from back detectors from Fig. 4.2 and for fuels with 5 and 20years of CT. Uncertainties for data points are smaller than the markers used.
6The primary motivation for using this design was because fewer simulations were needed in order to evaluate this
matter. The simulations without the Cd liner have been already done for the Blanc and Menlove design.
24
3.5 Additional and/or undecided considerationsWhile the major components of the DDA design described in previous sections have been thor-
oughly scrutinized and evaluated for deployment in Clab, several components and design features
still remain undecided. The reasons for this vary from computational demands and order of prior-
ity to unknown requirements by SKB.
An independent NG flux monitor is one of the undecided components. This detector should
independently monitor the NG output during the measurement campaign as well as during the
entire lifetime of the NG. As mentioned in Paper II, different options were considered and the
best candidate seems to be a fission chamber positioned right behind the NG.
Another undecided component is how to accommodate BWR SFAs into the instrument. Be-
cause of high computational demands, BWR SFAs have not been simulated specifically with the
Clab design, despite BWR fuel comprising the majority of the fuel stored in Clab. However, as
described in [14], the DDA instrument functionality is expected to be transferable to BWR SFAs
if the water gap around the SFA is filled by a material affecting the NG transport minimally. This
issue is further described in Section 3.6.
3.6 Modifications of the designAlthough the design of the DDA instrument for deployment in Clab was simulated with a focus on
application to Swedish SFAs, several modifications primarily due to the mechanical constraints
have been introduced after the completion of the simulations.
Figure 3.7. Schematic drawing of the DDA detector layout and position of the NG dictated by the use ofdifferent shielding material and other design modifications. The left panel illustrates the situation for theinitially modelled position in Paper II, the right panel illustrates later modifications made due to mechanicalconstraints [Paper III].
Besides the minor changes related to detector pods extensions in order to fit additional electron-
ics or practical arrangements of fitting the instrument in Clab, the main changes are summarized
in Paper III and also listed below:
1. The distance between the NG and SFA position increased due to the selection of 5 cm thick
lead as a shielding material instead the suggested 3 cm thick tungsten shield. In addition, a
2.5 cm thick aluminum block was introduced (Fig. 3.7) in order to accommodate the funnel7
at the top of the DDA insert.
2. A removable Cd liner on the external wall of the DDA insert was selected in case a thicker
water gap (more than 1 cm) appears around the SFA.
7A funnel is used on the top of insert(s) in order to guide the SFA into the instrument.
25
3. A different DDA insert was designed for BWR SFAs because of the smaller BWR dimen-
sions compared to the PWR dimensions. The emerging water gap around the BWR SFA is
expected to be filled by an aluminum insert as illustrated in Fig. 3.8. BWR SFAs are planned
to be positioned in the front part of the cavity to maximize the solid angle for neutron assay,
as illustrated schematically in Fig. 3.8 (right panel).
Figure 3.8. The comparison of the DDA instrument with the inserted PWR SFA (left) and BWR SFA togetherwith the aluminum insert (right)[Paper III].
The listed modifications of the DDA design may have an impact on the DDA functionality if
they are not considered carefully. Apart from these major modifications, several minor modifi-
cations of almost negligible impact have been applied. A summary of minor changes is listed
here:
• Addition of a 2.5 cm thick shielding house, protecting the NG from external gamma rays
from all directions, not only from those coming directly from the SFA.
• A NG monitoring detector is expected to be positioned behind the NG, away from the SFA.
To accommodate this, a nominal cavity of 2.5 cm is considered. Currently, an 18 cm long
fission chamber with 2 g of depleted uranium (DU) is under consideration for this purpose.
• A 1.4 cm thick polyethylene sleeve around the 3He detectors is considered in order to re-
duce the count rate on the detectors and ensure that the electronics will function reliably.
This reduced sleeve is considered primarily on the 3He detectors closest to the NG (front
detectors), but may be added to the remaining detectors.
26
4. Detector response to asymmetrically burned SFAs
The investigation of the neutron population evolution follows the the previous research done by
Lundkvist, et al. [22].
During our studies of the neutron population movement through the SFAs, which are fully de-
scribed in Paper I, the instrument response to asymmetrically burned SFAs was evaluated. Models
of SFAs from SFL2c were used for this purpose, because they feature significant asymmetry in
the isotopic distribution across the SFA fuel pins created by special shuffling patterns [13].
The asymmetry is illustrated for certain isotopes in Fig. 4.1. In the case of 235U, the side-to-side
difference in isotopic mass may be around 15%, but in the cases of 239Pu and 155Gd differences
may be up to 40% and 20%, respectively.
fuel pin
2 4 6 8 10 12 14 16 fuel pin2 4
6 8 10 12 1416
mas
s [g
]
0.81
1.21.41.61.8
22.2
-310×
1
1.2
1.4
1.6
1.8
2
-310×Gd155
1
1.2
1.4
1.6
1.8
2
-310×
fuel pin
2 4 6 8 10 12 14 16 fuel pin2 4
6 8 10 12 1416
mas
s [g
]
5.56
6.57
7.58
8.59
5.5
6
6.5
7
7.5
8
8.5
Pu239
5.5
6
6.5
7
7.5
8
8.5
fuel pin
2 4 6 8 10 12 14 16 fuel pin2 4
6 8 10 12 1416
mas
s [g
]
40424446485052
40
42
44
46
48
50
52
U235
40
42
44
46
48
50
52
Figure 4.1. Asymmetric distribution of a 155Gd (top left panel), 239Pu (top right panel), and 235U (bottompanel) created by shuffling sequence #3 for a 4 % IE, 15 GWd/tU BU, 5 y CT SFA.(Note: 2D extrapolationof masses inside individual fuel pins was used to create the continuous surface plot).
This simulations were performed with a combined DDA and Delayed Neutron (DN) instrument
design by Blanc and Menlove [23], as illustrated in Fig. 4.2. Selected SFAs were rotated and
assayed four times for different orientations and the change in detector response was studied. The
results of the study illustrates that if a strong asymmetry in isotopic composition is present across
the SFA, the front detector responses vary with SFA orientation. While the front detectors vary in
response by approximately ± 4 %, the back detectors are independent of fuel orientation with a
change in response of only ± 1.5 %. These results are illustrated in Fig. 4.3 [Paper I].
27
NG
W
Poly
Cd
Pb
H2OSFA
Steel
3He
Stainles
Figure 4.2. An illustration of design by Blanc and Menlove [23] for combined deployment of DDA anddelayed neutron (DN) techniques.
multiplication1.9 2 2.1 2.2 2.3 2.4 2.5 2.6 2.7
DD
A s
igna
l rel
ativ
e to
SFA
mea
n [%
]
-5-4
-3-2-1
01
2
34
5scheme #1scheme #2scheme #3
scheme #1scheme #2scheme #3
orientation #1 orientation #2
orientation #3 orientation #4
orientation #1 orientation #2
orientation #3 orientation #4
45 30 15tU
GWdBU
20 5 20 5 20 5 CT y
multiplication1.9 2 2.1 2.2 2.3 2.4 2.5 2.6 2.7
DD
A s
igna
l rel
ativ
e to
SFA
mea
n [%
]
-5-4
-3-2-1
01
2
34
5scheme #1scheme #2scheme #3
scheme #1scheme #2scheme #3
orientation #1 orientation #2
orientation #3 orientation #4
orientation #1 orientation #2
orientation #3 orientation #4
45 30 15tU
GWdBU
20 5 20 5 20 5 CT y
Figure 4.3. Relative differences of DDA signals from the four orientation mean value of the simulatedDDA instrument response for each SFA depicted as a function of multiplication. Data points are illustratedseparately for the front (left panel) and back (right panel) detectors for 3 different shuffling schemes (#1:triangles, #2: dots, #3: squares). The approximate correlation between CT, BU and multiplication (M) isindicated by top, bottom secondary x axis, respectively. Different colors represent different orientations ofthe SFA.
multiplication1.9 2 2.1 2.2 2.3 2.4 2.5 2.6 2.7di
e-aw
ay ti
me
rela
tive
to S
FA m
ean
[%]
-5-4
-3-2-1
01
2
34
5scheme #1scheme #2scheme #3
scheme #1scheme #2scheme #3
orientation #1 orientation #2
orientation #3 orientation #4
orientation #1 orientation #2
orientation #3 orientation #4
45 30 15tU
GWdBU
20 5 20 5 20 5 CT y
multiplication1.9 2 2.1 2.2 2.3 2.4 2.5 2.6 2.7di
e-aw
ay ti
me
rela
tive
to S
FA m
ean
[%]
-5-4
-3-2-1
01
2
34
5scheme #1scheme #2scheme #3
scheme #1scheme #2scheme #3
orientation #1 orientation #2
orientation #3 orientation #4
orientation #1 orientation #2
orientation #3 orientation #4
45 30 15tU
GWdBU
20 5 20 5 20 5 CT y
Figure 4.4. Relative differences of die-away times from the four orientation mean value of the simulatedDDA instrument response for each SFA depicted as a function of multiplication. Results are presentedseparately for the front (left panel) and back (right panel) detectors for three different shuffling schemes(#1: triangles, #2: dots, #3: squares). The approximate correlation between CT, BU and multiplication (M)are indicated by top, bottom secondary x axis, respectively. Different colors represent different orientationsof the SFA.
28
Besides the DDA detector responses, the change in die-away time was also evaluated. Although
the fluctuation of data points for different die-away times is in the range ± 2.5 % (the majority of
the points within ± 1.5 %) for both front and back detectors, the behavior found for the detector
responses is not repeated here since the differences are indistinguishable from the statistical errors
of the die-away time determination from the exponential fit.
The differences between die-away times and the mean values from the assay of four orientations
are illustrated in Fig. 4.4 [Paper I]. Based on this study, it has been concluded that the back
detectors can be used for global SFA assay (i.e. determining general SFA parameters such as IE,
BU, etc.), and the front detectors for local assay (e.g. partial defects detection).
29
5. "Working concepts" of DDA alteration for differentpurposes
This chapter summarizes several concepts of potential modification or customization of the DDA
instruments proposed for different applications. These concepts have been suggested in Paper IV,
as alternatives for the "heavy" Clab design, if the customer requirements are limited to a specific
function of the DDA instrument or for operation under specific conditions. Based on our findings
during the Clab design finalization [Paper I, Paper II] as well as other investigations [22], the
following concepts are considered:
1. Light-weight/portable DDA instrument (Fig. 5.1),
2. Minimalist DDA instrument (Fig. 5.2), and
3. Partical defect DDA instrument called "Defectoscope" [24] (Fig. 5.3).
Figure 5.1. Schematic cross-sectional view of a proposed "light-weight/portable" DDA instrument. Instru-ment designed by by N.Lundkvist.
The light-weight/portable DDA instrument is characterized by removing the heavy shielding
materials for both the NG and the detectors. The detectors are then moved closer to the SFA which
increases the detection count rates. Fission chambers or other detectors insensitive to gamma rays
are recommended instead of 3He detectors. More details about this design can be found in [22]
and in Paper IV.
The minimalist DDA design consists of a limited number of components. As illustrated in
Paper I, the back detectors are sufficient for global assay and SFA parameter determination. If
local assay is not required, a few back detectors, be it either fission chambers or 3He tubes, are
expected to be fully sufficient for this purpose [Paper IV].
In contrast, if a local assay is requested and/or fuel pin diversion detection capability is required,
the Defectoscope DDA instrument is expected to be suitable for this task [Paper IV]. In this design,
the detectors (preferably a tiny detector with as minimal diameter as realistically achievable) may
31
Figure 5.2. Schematic cross-sectional view of a proposed "minimalist" DDA instrument. Instrument de-signed by Henzl and Martinik.
Figure 5.3. Schematic cross-sectional view of a proposed "defectoscope" DDA instrument. Instrumentdesigned by Henzl and Martinik.
be adjusted using a polyethylene collimator to make it sensitive to only a selection of the SFA fuel
pin(s), as schematically illustrated in Fig. 5.3.
All these designs are considered as working concepts based on experience gained through re-
search on the Clab DDA design. Thus, except for some preliminary simulations done by Lund-
kvist et al. [22], proper evaluation of the performance of these designs has not yet been done.
32
6. Conclusions
The DDA technique, following the high fidelity simulations in MCNPX, demonstrates its capa-
bility for SFAs verification as described in [8, 10, 14]. Given initial research investigations with
the conceptual instrument design by Blanc and Menlove [23], our focus aimed at designing the
instrumentation for characterizing Swedish SFAs. In Sweden, the concept for future handling of
SFAs is motivated by a decision to shut down the nuclear power plants and an intent to deposit
the SFAs in a geological repository.
Within the scope of this work, we adjusted the initial conceptual design [23] for application
in the Central Interim Storage Facility for Spent Nuclear Fuel (Clab). In other words, we have
tuned the conceptual design to be a deployable instrument. We have specifically investigated the
dimensions of detectors, gamma-ray shielding, NG operation, PWR and BWR insert solution and
independent neutron flux monitor position determination. All the main components relevant for
the DDA instrument functionality are summarized in Chapter 3. Section 3.6 describes the main
modifications that were introduced: different shielding material of NG; Cd liner utilization, and a
different BWR insert solution.
The separate study on the detector response to asymmetrically burned SFAs illustrated the
sensitivity of the front detectors to SFAs with large asymmetry in isotopic distribution across the
SFA. In contrast, the back detectors were observed to be insensitive to the asymmetry in spatial
isotopic distribution. The front detectors are hence suitable for local assay (studying limited parts
of the SFA), while the back detectors are better suited for global assay (determining multiplication,
IE, BU, CT, etc.).
Based on the conclusions from Paper I, [22] and 3.4.5, three possible modifications of the
DDA design are introduced: 1) the "Light-weight DDA instrument", 2) the "Minimalistic DDA
instrument", and 3) the "Defectoscope" [24].
In conclusion, the DDA design, as simulated in MCNPX for future deployment in Clab, is
expected to be fully working for the different type of SFAs (BWR and PWR) as well as for
different types of quantitative measurements.
The DDA instrument is expected to measure multiplication and total fissile mass (239Pue f f );
determine the total Pu mass, IE, BU; and potentially identify the illicit replacement (or removal)
of fuel pins.
The projected measurement times of one SFA depends on whether a BWR or a PWR SFA
is assayed, and also on the parameters requested to be verified. Tentative measurements times
for different scenarios scenarios were elaborated on in Paper III. In general, the functionality of
the DDA instrument was evaluated for 5 and 10 minutes for one SFA orientation, both of which
indicated indicated a statistical uncertainty below 1 %. If the SFA is measured in four orientations,
the time is expected to be prolonged accordingly. In addition, the time required for maneuvering
of the fuel assembly will need to be considered as a significant part of the time needed to assay
one SFA.
33
7. Outlook
The plan is to manufacture and perform benchmarking measurements with the DDA instrument on
fresh fuel in Los Alamos National Laboratory, and later to test the performance of the instrument
in Clab on a set of 25 BWR and 25 PWR SFAs. The fresh fuel measurement campaign is planned
for 2016 and possibly diversion scenarios and calibration measurements will be performed in this
campaign. The plans for the spent fuel measurements have been postponed several times and it
is still unclear when they will take place. If the fabrication of the instrument is further delayed,
a possibility to do other experimental measurements with a DDA technique-based instrument is
foreseen.
35
Acknowledgement
I would like to acknowledge the support of Next Generation Safeguards Initiative (NGSI), Of-
fice of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration
(NNSA); and Uppsala University for funding this Ph.D. project.
I hereby appreciate all the help given by my Swedish supervisors: Sophie Grape, Peter Jans-
son and Staffan Jacobsson Svärd; their guidance through my Ph.D. studies, and especially, their
patience and more or less understanding of all my difficulties related to both working as well as
normal life. Many of the particularities related to the Ph.D. studies and academia are still chal-
lenging for me. Thank you for trying to understand!
I would also like to thank Steve Tobin and Holly Trellue for the opportunity to be a part of the
NGSI-SF research project and having the chance to work in such high level research environment
at Los Alamos National Laboratory (LANL) and on definitely exciting research tasks.
My deep gratitude goes to Vlad Henzl, my main advisor from LANL. Working with you on
daily basis was a big inspiration for me and the days I spent in LANL would always be the high-
light of my life. The time I worked with you I enjoyed a lot and every morning, I was looking
forward to the work ahead, while in the evening, regretting that I had to go back home. It wasn’t
only about the high-quality research what I was enjoying, but also lots of fun in discovering new
features of DDA technique and discussing with you all the challenges involved in research. More-
over, I appreciate a lot all your help in my personal life (from you and your whole family)! Thank
you very much!
Many thanks also to:
Martyn Swinhoe, for always being opened to discuss deeply different problematics of my work,
your patience in trying to understand what I were explaining to you, and especially for your com-
ments and advices that were straight to the point. Working with a scientist like you with so many
experiences and big expertise in safeguards field was wonderful inspiration for me.
John Hendricks, for teaching me how to use MCNP code in advance level, for introducing me
to many new features of the MCNP code, answering every question I had about the code (some
of them repeatedly for many times), willing to help always with every problem I encountered in
connection to MCNP simulations! Moreover, thank you John also for all your help with many
other issues! I am really grateful to get the chance being your MCNP student and I will never
forget the time I spent with you with MCNP. Your enthusiasm and always positive attitude which
you apply for solving different challenges always manage to calm me naturally and ensure me
"that everything would be OK". Just sitting next to you and listening to you talking about MCNP
or other different things was really motivating and gave me a lot of new energy to further work!
Thank you very much John!
Laura O., for your cheerful attitude, understanding and willingness to always help me any time
I faced any challenges. I am really grateful for all your help and I especially appreciate that you
were always willing to talk with me about anything when I needed that! I am glad to have met you!
37
Johnna M., Stacey E. and Rolin L. for understanding my difficulties I experienced during the sec-
ond year of my Ph.D. and for all your help related to this matter.
Katrina K., for your help with final corrections of my writing in several publications, including
this thesis.
The rest of people I’ve met in Los Alamos: Alison, Alexis, Daniela, Estevan, Andrea, Katrina,
Katie, Peter S., Anthony B., Bethany, Metodi, Marian, Michal, Veronika, and everyone else.
My colleagues and friends from Uppsala University: Ali, Augusto, Andrea, Bill, Erik, Erwin,
Matilda, Peter, Vasudha, and everyone else.
Finally, I wish to say thank you for all the support from my family and friends from The Czech
Republic. You were always supporting me in the hard times and I am really grateful for it. Thank
you Mum, Dad, Ada, Jan, Kuba C., Kuba B., Mira K., Karl B., "Expert", Johny, and everyone else.
38
References
[1] G. JANSSENS-MAENHOUT editor. Nuclear Safeguards and Non-Proliferation: syllabus of theESARDA Course. Luxembourg: Office for Official Publications of the European Communities, 2009.
ISBN 978-92-79-09741-6.
[2] SKB-Swedish Nuclear Fuel and Waste Management Company. Spent Nuclear Fuel for Disposal in
the KBS-3 Repository. Technical report, 2010. Technical report (TR-10-13).
[3] SKB-Swedish Nuclear Fuel and Waste Management Company. Pressure Tests of Two KBS-3
Canister Mock-ups. Technical report, 2005. Technical report (TR-05-18).
[4] A. Nyström. Encapsulation Plant at Forsmark. Technical report, SKB-Swedish Nuclear Fuel and
Waste Management Company, 2007. Technical report (R-07-36).
[5] N. Lundkvist. Investigation of Possible Non-Destructive Assay (NDA) Techniques for the Future
swedish Encapsulation Facility. Master thesis, Uppsala University, 2012.
[6] National Nuclear Security Administration (NNSA). Next Generation Safeguards Initiative. URL
�������������� ������������������������������������������
[2015-10-10].
[7] M.A. Humprey, S.J. Tobin, and K.D. Veal. The Next Generation Safeguards Initiative’s Spent Fuel
Nondesctructive Assay Project. Journal of Nuclear Materials Management, 40(3):6, 2012.
[8] V. Henzl, M.T. Swinhoe, S.J. Tobin, and H.O. Menlove. Measurement of the Multiplication of a
Spent Fuel Assembly with the Differential Die-Away Method within the Scope of the Next
Generation Safeguards Initiative Spent Fuel Project. Journal of Nuclear Materials Management,40(3):61, 2012.
[9] D. Reilly, N. Ensslin, H. Smith Jr., and S. Kreiner. Passive Nondestructive Assay of NuclearMaterials. Springfield, VA: US Department of Commerce, National Technical Information Service,
1991. ISBN 0-16-032724-5.
[10] V. Henzl, M.T. Swinhoe, and S.J. Tobin. Direct Measurement of Initial Enrichment and Burn-up of
Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument. Proceedings ofInstitute of Nuclear Materials Management conference, Orlando, Florida, 2012.
[11] J.F. Pelowitz. MCNPXT M User’s Manual Version 2.7.0. April 2011. Los Alamos National
Laboratory Report LA-CP-11-00438.
[12] S.J. Tobin and P. Jansson. Nondestructive Assay Options for Spent Fuel Encapsulation. Technical
report, Los Alamos National Laboratory, 2013. LA-UR-13-22050 (2013).
[13] H.R. Trellue, J.D. Galloway, N.A. Fisher, and S.J. Tobin. Avances in Spent Fuel Libraries.
Proceedings of Institure of Nuclear Materials Management conference, Palm Desert, California,
2013. Los Alamos National Laboratory Report LA-CP-13-24074.
[14] V Henzl. Evaluation of Differential Die-Away Technique Potential in Context of Non-Destructive
Assay of Spent Nuclear Fuel. 2014. Los Alamos National Laboratory report LA-UR-14-29224.
[15] Westinghouse. Westinghouse rfa-2 design. URL
�������������������������������������������������������������������
��������������� �!�"!���#���$�%!�&�'$��� [2015-10-10].
[16] S.J. Tobin. Private Communication, January-March 2014. Affiliated with The Spent Nuclear Fuel
and Waste Management Company, SKB.
[17] A.C. Kaplan, V. Henzl, H.O. Menlove, M.T Swinhoe, A.P. Belian, M. Flaska, and S.A. Pozzi.
Determination of Spent Nuclear Fuel Assembly Multiplication with the Differential Die-Away
Self-Interrogation Instrument. Nuclear Intruments and Methods in Physics Research A, 757:20–27,
2014. http://dx.doi.org/10.1016/j.nima.2014.04.023.
[18] J. Simpson. Export Control - Private Pommunication, April-May 2015. Sales Manager, Thermo
Fisher Scientific, Colorado Springs, USA.
[19] Wikipedia. Duty cycle definition. URL �����������(�����������(��)�� &� ��
[2015-10-10].
[20] P. Le Tourneur. Private Communication. 07/2014-08/2014. EADS-Sodern, 20 Av. Descartes 94451
Limeil-Brevannes Cedex, France.
39
[21] A.V Goodsell, V. Henzl, and M.T. Swinhoe. Differential Die-Away Instrument: Report on Initial
Simulations of Spent Fuel Experiment. 2014. Los Alamos National Laboratory report
LA-UR-14-22162.
[22] N. Lundkvist, V.A. Goodsell, S. Grape, J.S. Hendricks, V. Henzl, M.T. Swinhoe, and S.J. Tobin. A
Qualitative Analysis of the Neutron Population in Fresh and Spent Fuel Assemblies during Simulated
Interrogation using the Differential Die-Away Technique. Submitted for publication to ESARDAbulletin. Los Alamos National Laboratory report LA-UR-15-23441 (2015).
[23] P.C. Blanc, S.J. Tobin, S. Croft, and H.O. Menlove. Plutonium Mass Determination in Spent
Fuel-Delayed Neutron Detection in an Integrated Delayed-Neutron and Differential Die-Away
Instrument with 3He Detectors and a DT Generator. 2011. Los Alamos National Laboratory report
LA-UR-11-01912 (2011).
[24] V. Henzl, T Martinik, and M.T. Swinhoe. Private Pommunication, August-September 2014. Los
Alamos National Laboratory, USA.
[25] T. Martinik, V. Henzl, S. Grape, S.J. Svärd, P. Jansson, M.T. Swinhoe, and S.J. Tobin. Simulation of
Differential Die-Away Instrument’s Response to Asymmetrically Burned Spent Nuclear Fuel.
Nuclear Intruments and Methods in Physics Research A, 2015. Los Alamos National Laboratory
Report LA-UR-14-21574 (2014), Available online at http://dx.doi.org/10.1016/j.nima.2015.02.058.
[26] T. Martinik, V. Henzl, S. Grape, P. Jansson, M.T. Swinhoe, A.V. Goodsell, and S.J. Tobin. Design of
a Prototype Differential Die-Away Instrument Proposed for Swedish Spent Nuclear Fuel
Characterization. Submitted for publication in Intruments and Methods in Physics Research A,
October,2015.
[27] T. Martinik, V. Henzl, S. Grape, P. Jansson, M.T. Swinhoe, A.V. Goodsell, and S.J. Tobin.
Development of a Differential Die-Away Instrument for Characterization of Swedish Spent Nuclear
Fuel. Proceedings of 37th ESARDA Symposium, Manchester, United Kingdom, 2015. Los Alamos
National Laboratory Report LA-UR-15-23345 (2015).
[28] T. Martinik, V. Henzl, S. Grape, S.J. Svärd, P. Jansson, M.T. Swinhoe, and S.J. Tobin.
Characterization of Spent Nuclear Fuel with a Differential Die-Away Instrument. Proceedings ofIAEA Symposium on International Safeguards: Linking Strategy, Implementation and People, Vienna,Austria, 2014. Los Alamos National Laboratory Report LA-UR-14-28074 (2014).
