Nuclear Safety
NEA/CSNI/R(2023)4
July 2024
www.oecd-nea.org

Status Report on Fuel Safety

Implications of Extended Enrichment
and High Reactivity/High Suppression
Core Designs
A Report from the Working
Group on Fuel Safety (WGFS)

Nuclear Energy Agency

NEA/CSNI/R(2023)4
Unclassified

English text only
29 July 2024

NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

Status Report on Fuel Safety Implications of Extended Enrichment and High
Reactivity/High Suppression Core Designs
A Report from the Working Group on Fuel Safety (WGFS)This document is available in PDF format only.

JT03547920
OFDE

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STATUS REPORT ON FUEL SAFETY IMPLICATIONS OF EXTENDED ENRICHMENT AND HIGH REACTIVITY/HIGH SUPPRESSION CORE DESIGNSNEA/CSNI/R(2023)4  3

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSThe Committee on the Safety of Nuclear Installations (CSNI) addresses NEA programmes
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4  NEA/CSNI/R(2023)4

Acknowledgements
This status report is based on the statements and opinions provided by the Nuclear Energy
Agency (NEA) Working Group on Fuel Safety (WGFS) members and experts from variousorganisations. In particular, the following individuals gave valuable input to the paper orparticipated in the extensive review process:•Anna Alvestav, Swedish Radiation Safety Authority (SSM), Sweden•

Masaki Amaya, Japan Atomic Energy Agency (JAEA), Japan•

Michelle Bales, NEA•

Alžběta Bednářová, State Office for Nuclear Safety (SÚJB), Czechia•

Felix Boldt, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), Germany•

Simon Bowditch, Office for Nuclear Regulation (ONR), United Kingdom•

James Corson, Nuclear Regulatory Commission (NRC), United States•

Aleksei Gusev, TVEL, Russian Federation•

Branislav Hatala, VUJE, Slovak Republic•Zoltán Hózer, Centre for Energy Research, Hungary•

Hyedong Jeong, Korea Institute of Nuclear Safety (KINS), Korea•

Koji Kitano, Japan Nuclear Regulation Authority (NRA), Japan•

Lucas Kyriazidis, Nuclear Regulatory Commission (NRC), United States•

Bernadett Lente, Hungarian Atomic Energy Authority, Hungary•

Olivier Marchand, Institute for Radiation Protection and Nuclear Safety (IRSN),France•Dostal Martin, UJV, Czechia•

Jose Maria Rey Gayo, Spanish Nuclear Safety Council (CSN), Spain•

Martin Sevecek, Alvel, Czechia

Special thanks go to Michelle Bales (at the time of drafting: Nuclear
Regulatory Commission, United States) and James Corson (NRC, United States)
for leading the task group, drafting the report, and for consolidating the various
opinions, comments and remarks; to Koji Kitano (NRA, Japan) and Zoltán Hózer(Centre for Energy Research, Hungary) for contributing to the safety
implications evaluation table and for providing comments on the report; and to
Masaki Amaya (JAEA, Japan) and Jinzhao Zhang (Tractabel, Belgium) for providing
comments as well.

STATUS REPORT ON FUEL SAFETY IMPLICATIONS OF EXTENDED ENRICHMENT AND HIGH REACTIVITY/HIGH SUPPRESSION CORE DESIGNSNEA/CSNI/R(2023)4  5

Table of contents

List of abbreviations and acronyms ..................................................................................................... 6
Executive summary ............................................................................................................................... 8
1. Introduction ....................................................................................................................................... 9
1.1. Background, objectives and scope ................................................................................................ 9
1.2. Procedure and organisation of the report .................................................................................... 10
2. Summary of international experience related to extended enrichment ..................................... 11
2.1. Explicit regulatory limits on 235U enrichment ............................................................................. 112.2. Operating limits and practical restrictions related to enrichment ............................................... 122.3. Ongoing activity to extend enrichment levels............................................................................. 122.4. Qualification of neutronic calculation tools for enrichment of 5-8% ......................................... 122.5. Ongoing or completed analysis of extended enrichment operation ............................................ 13
2.6. In-pile testing of extended enrichment fuels ............................................................................... 14
2.7. New fuel testing requirements .................................................................................................... 15
2.8. Relevant operating limits ............................................................................................................ 16
2.9. Assembly or fuel rod peaking factor ........................................................................................... 16
2.10. Use of burnable poisons ............................................................................................................ 17
2.11. Moderator temperature coefficient (MTC) ............................................................................... 17
2.12. Cycle length and considerations of instrument calibration ....................................................... 17
2.13. Shutdown margin ...................................................................................................................... 18
2.14. Human error in core loading ..................................................................................................... 18
2.15. Mixed cores and transition cores .............................................................................................. 19
3. Summary of safety implications evaluation .................................................................................. 20
3.1. Impacts of extended enrichment on design limits ....................................................................... 20
3.2. Opportunities for collaborative research ..................................................................................... 21
4. Conclusions and recommendations................................................................................................ 224.1. Conclusions................................................................................................................................. 224.2. Recommendations ....................................................................................................................... 22
References ............................................................................................................................................. 24
Appendix A: Questionnaire in preparation for a status report on fuel safety implications ofextended enrichment and high reactivity/high suppression core designs....................................... 27
Appendix B: Safety implications evaluation table ............................................................................ 29

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List of abbreviations and acronyms
ATF

Accident tolerant fuel

BfS

Bundesamt für Strahlenschutz (Federal Office for Radiation
Protection, Germany)

BWR

Boiling water reactor

CAPS

CSNI Activity Proposal Sheet

CSN

Spanish Nuclear Safety Council

CSNI

Committee on the Safety of Nuclear Installations (NEA)

FAST

Fuel Analysis under Steady-state and Transients

FIDES-II

Second Framework for Irradiation Experiments (NEA joint
project)

FRG

Fission gas release

GRS

Gesellschaft für Anlagen- und Reaktorsicherheit (Germany)

IAEA

International Atomic Energy Agency

ICSBEPInternational Criticality Safety Benchmark Evaluation Project

IFA

Instrumented fuel assembly

IFBA

Integral fuel burnable absorbers

IRSN

L'Institut de radioprotection et de sûreté nucléaire (Institute for
Radiation Protection and Nuclear Safety, France)

IVNET

Innovative and Viable Nuclear Energy Technology

JAEA

Japan Atomic Energy Agency

JAERI

Japan Atomic Energy Research Institute

JMTR

Japan Materials Testing Reactor

KINS

Korea Institute of Nuclear Safety

KUCA

Kyoto University Critical Assembly

LAGER

Laser Ablation of Gadolinium Evolution Radially

LOCA

Loss-of-coolant accident

LWR

Light water reactor

MTC

Moderator temperature coefficient

NEA

Nuclear Energy Agency

NRA

Japan Nuclear Regulation Authority

NRC

Nuclear Regulatory Commission (United States)

OECD

Organisation for Economic Co-operation and Development

ONR

Office for Nuclear Regulation (United Kingdom)

STATUS REPORT ON FUEL SAFETY IMPLICATIONS OF EXTENDED ENRICHMENT AND HIGH REACTIVITY/HIGH SUPPRESSION CORE DESIGNSNEA/CSNI/R(2023)4  7

ORNL

Oak Ridge National Laboratory (United States)

PIE

Post irradiation examinations

PWR

Pressurised water reactor

SFR

Sodium cooled fast reactors

SSM

Swedish Radiation Safety Authority

SÚJBState Office for Nuclear Safety (Czechia)

UO2

Uranium dioxide

VVER

Water-cooled water moderated energy reactor (Russian design)

WGFS

Working Group on Fuel Safety (NEA)

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8  NEA/CSNI/R(2023)4

Executive summary
Rising international interest in increasing fuel burnup limits and fuel cycle length may
require fuel enrichment above 5% and high reactivity/high suppression core designs. Basedon this interest in extended fuel enrichment (5-8%), the Nuclear Energy Agency (NEA)
Working Group on Fuel Safety (WGFS): 1) collected information on fuel enrichment limits
for uranium dioxide fuel in light water reactors in NEA member countries through aquestionnaire; 2) consolidated the information in a status report 3) utilised WGFS expertiseto evaluate fuel safety implications of extended enrichment fuel and high reactivity/highsuppression core designs; and 4) provided recommendations for collaborative analyticaland experimental research on this topic. The task group focused on in-reactor behaviour ofhigh reactivity/high suppression core designs. Issues related to the front-end (e.g. fuelfabrication) and back-end (e.g. spent fuel storage) of the fuel cycle are outside the scope of
this report.
The questionnaire responses indicated that while there is limited operating experience in
NEA member countries with fuel enrichment of 5-8%, extensive experimental data fromresearch and test reactors could be used to validate neutronic and fuel performance codes
and methods for extended enrichment fuel. From these responses, the group concluded that
there are no major experimental gaps related to extended enrichment fuel behaviour.
The group also concluded that few design limits are directly related to fuel enrichment.
However, fuel enrichment can impact core radial and axial power profiles, as well as theradial power profile across the fuel pellet. These changes may make it challenging to meetexisting design limits. Furthermore, there is limited experience with extended enrichment
fuel or gadolinia doped fuel behaviour in design basis accident conditions.
Therefore, the group recommends performing experiments on extended enrichment or
gadolinia-doped fuel, particularly for reactivity-initiated accident or power ramp
conditions. Such experiments could be proposed under the NEA Second Framework for
Irradiation Experiments (FIDES-II) joint project. The group also recommends launching
two code benchmark activities. The first benchmark activity would focus on fuel thermal
mechanical behaviour and may involve tests on extended enrichment fuel material. This
activity could be conducted under the auspices of the WGFS. The second benchmark
activity would focus on the neutronic behaviour of a high reactivity/high suppression corewith extended enrichment fuel. Similar efforts are currently underway in the United States;
these core designs could potentially serve as the basis of the recommended benchmark
activity. This activity should be conducted by an NEA group focused on neutronics or core
design. Together, these recommended activities would verify that extended enrichment fuelis well understood and that existing design limits are still valid for fuel of 5-8% enrichment.
The expert opinions collected in this report should be revisited during the activities todetermine if they can be confirmed or not by the work.

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1. Introduction

1.1. Background, objectives and scope
An international interest in operating fuel to higher burnup may require initial fuelenrichment above 5 weight percent (%).1 It is expected that regulatory authorities will soon
be faced with applications for fuel designs utilising enrichments of 5-8%. To prepare forthese expected applications, a Committee on the Safety of Nuclear Installations (CSNI)
activity was proposed for a “Status Report on Fuel Safety Implications of Extended
Enrichment and High Reactivity/High Suppression Core Designs.” The objective of thisactivity is to consolidate information regarding fuel enrichment limits of uranium dioxide(UO2) fuel for light water reactors (LWRs) in NEA member countries and utilise theexpertise of the NEA Working Group on Fuel Safety (WGFS)2 to evaluate possible fuelsafety implications of extended enrichment (5-8%) and high reactivity/high suppression
core designs.3
The scope of this report is limited to UO2 fuel in LWRs. However, there are two areas
outside of this scope that must be acknowledged. First, research reactors in many countries
today operate with fuel at enrichment levels above 5%. This experience demonstrates, tosome extent, the technical feasibility of extended enrichment operation and supporting
analysis requirements. At the same time, it must be recognised that the size, complexityand design of research reactors is significantly different than large LWRs (i.e. significantly
lower pressures and temperatures, limited use of poisons), often allowing for a more
simplified safety basis. In addition, in most cases the fuel used in research reactors is not
UO2. This report does not attempt to capture operating experience of highly enriched, nonUO2 fuel in research reactors and the high enrichment UO2 fuel used in sodium cooled fast
reactors (SFR) is not discussed, for the fuel in SFRs has different loads (e.g. high linear
heat generation rates, higher temperatures, fast neutron spectrum) and the fuel behaviour ischaracterised by specific phenomena (e.g. very high fission gas release, fuel restructuring,
redistribution of fuel constituents). At the same time, there are examples of research andtest reactors that utilise UO2 fuel enriched slightly above 5% (e.g. the Halden Reactor in
Norway) and examples of non-resident test articles of UO2 fuel that have been irradiated inresearch reactors where conditions were established to simulate LWR conditions (e.g. loops
within the Halden Reactor in Norway). In some cases, this experience may be useful to
1

Unless otherwise noted, enrichments are presented in terms of weight percent. For simplicity,
weight percent enrichment will be indicated using the % sign.
2

The NEA WGFS was established to advance the understanding of fuel safety issues by assessing
the technical basis for current safety criteria and their applicability to high burn-up and to new fueldesigns and materials, including fuels with increased enrichment.
3

The proposed activity, presented in a CSNI Activity Proposal Sheet (CAPS), includes highreactivity/high suppression/extended operating cycle core designs because of the relevance of thesecore designs to operation with extended enrichment. One of the biggest implications of operatingwith enrichments above 5% will be that the fuel will have an increased initial reactivity excess,which will need to be absorbed (e.g. by burnable poisons or other means). Some fuel designs, suchas MOX, or countries operating on 24-month cycles, may already be dealing with core designs that
require substantial power suppression early in life and this operating experience has relevance for
operating with extended enrichment.

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understand the performance of extended enrichment fuel and therefore will be discussed
within the report. Second, there are ongoing research programmes in several countries
related to the development of new fuel forms with extended enrichment for advanced
reactor designs. This research may have relevance to the development of a safety basis for
UO2 fuel in LWRs. However, the design constraints and historical safety basis of large
LWRs may prove far more restrictive than for new advanced reactors. This report does not
attempt to capture ongoing research on extended enrichment fuel designs for advanced
reactor designs, since the operational conditions in advanced reactors could be different
compared to LWRs.
This report focuses on in-reactor fuel behaviour. Issues related to the front-end (e.g. fuelfabrication) and back-end (e.g. spent fuel storage) of the fuel cycle are not covered in thisreport. Such issues (e.g. criticality safety) are important and could be addressed by other
NEA working groups.
While extended enrichment may be paired with increased fuel burnup limits in somemember countries, this report does not attempt to evaluate the fuel safety implications ofincreasing fuel burnup limits. The fuel safety implications of increasing fuel burnup are
considered as an important issue. They were discussed in the NEA report Very High Burnups in Light Water Reactors (NEA, 2006). Based on the latest knowledge and findings, thefuel safety implications of increasing fuel burnup should be addressed again, in anotheractivity.

1.2. Procedure and organisation of the report
This report first presents a summary of international experience related to extendedenrichment (Chapter 2). The summary follows the format and scope of a questionnaire used
to collect information from members of the activity task group. The questionnaire can befound in Appendix A. The report then provides a summary of the evaluation of possible
fuel safety implications of extended enrichment (5-8%) and high reactivity/highsuppression core designs (Chapter 3). The evaluation was facilitated by a comprehensive
table of fuel performance criteria utilised in the NEA CSNI Technical Opinion Paper No.
19: Applicability of Nuclear Fuel Safety Criteria to Accident-Tolerant Fuel Designs (NEA,2022). The evaluation table can be found in Appendix B. The evaluation captures thequalitative determination of the extent of the public database and data gaps as well as
members’ consideration of where there are opportunities for collaborative research.
Finally, the report presents conclusions and recommendations from the task group (Chapter
4).

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2. Summary of international experience related to extended enrichment

To begin the activity, a questionnaire was developed to collect information regarding fuelenrichment limits and high reactivity/high suppression core designs in NEA member
countries. The questionnaire was designed to capture experience generally with core
designs that require high power suppression as a way to gain insight into one of the fuelperformance concerns with enrichment above 5%. In other words, the activity will include
one of the fuel performance aspects of high enrichment, defined broadly, to gain someinsight despite a lack of specific operating or performance experience with extendedenrichment. The questionnaire distributed to task group participants is provided inAppendix A and a summary of the replies is provided below.

2.1. Explicit regulatory limits on 235U enrichment
Respondents were asked if there is an explicit regulatory limit on 235U enrichment in theircountry. If there is a limit, they were asked to specify if the limit is in reference to fuelmanufacturing operation, transportation, storage and/or use. All respondents reported that
there are no explicit limits on 235U enrichment in their country. However, in most countries
existing safety analyses for either manufacturing operation, transportation, storage or use
have assumed an enrichment limit of 5%. These analyses are part of the regulatory footprint
and would have to be re-performed if 235U enrichment exceeded 5%. Some countries haveestablished performance-based criteria that could be used for such re-analysis. Most
notably, in Japan the measures for the following criteria are required to be met for
manufacturing fuel with enrichment greater than 5%:
1. Exposing the public to radiation should be avoided.
2. Effective measures (neutron absorber/shielding, mass/shape/volume control, etc.)
to keep subcriticality for single- and multi-unit plants should be applied.
3. Detection and mitigation measures against criticality accidents should be secured.
Respondents from Japan also explained with regard to fuel manufacturing operation thatwhen the enrichment exceeds 5%, criticality must be considered and accident scenarios andcountermeasures are required. The transportation of fuel enriched beyond 5% may requirea licensing issue for new shipping containers. Respondents from Japan also noted that there
is no regulatory restriction on storage but domestic facilities, with some exceptions, aredesigned and licensed for enrichment of 5% or less. There are no regulatory restrictions on
use but new fuel storage pools are designed to contain fuel of less than 5% enrichment.
(Yamasaki & Unesaki, 2010)
Also notable is the existence of analysis simplifications when enrichment is below 5%. For
example, the International Atomic Energy Agency (IAEA) SSR-6 “Regulations for the Safe
Transport of Radioactive Material 2018 Edition” (IAEA, 2018) includes an allowance todisregard flooding in criticality assessments for packages containing uranium hexafluoride
only, with a maximum uranium enrichment of 5% 235U. Should transport of uranium
hexafluoride with an enrichment > 5% be required, this allowance could not be utilised andpackage flooding would have to be considered in the criticality assessment or else thepackage would need to show it had multiple high standard water barriers.

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2.2. Operating limits and practical restrictions related to enrichment
Respondents were asked if there are established operating limits (e.g. inherent reactivity
feedback), or other practical restrictions that effectively limit 235U enrichment in theircountry. If there are limits, they were asked if the limit was in reference to fuelmanufacturing operation, transportation, storage and/or use.
Most countries reported that an enrichment limit of 5% has been integrated within theexisting safety analysis or technical specifications in various ways. Increasing theenrichment would require revisiting these safety analyses and technical specifications toevaluate the implications of an increase. Multiple respondents speculated that otheroperating limits may interact with an enrichment limit, including:•limits on inherent reactivity feedback operating limits;•use of burnup credit in fuel storage;•

requirements for negative moderator temperature coefficient during power
operation;•

limits on the maximum linear heat generation rate during operation;•requirements for subcriticality during fuel handling;•

limits on local power peaking;•burnup limits;•storage capacity limits.

Respondents from Germany elaborated that the 5% limit is based on a non-proliferation
rationale (Federal Office for the Safety of Nuclear Waste Management, 2022).

2.3. Ongoing activity to extend enrichment levels
Respondents were asked if there was active interest in their country to extend the current
enrichment limit of LWRs. They were asked if there is any indication that an intended useof high enrichment fuel would be combined with any accident tolerant fuel (ATF) design.
At this time, there are active industry efforts to increase enrichment above 5% for LWR
fuel in the United States and for water-water energetic reactor (VVER) designs in Russia.
In both cases, the efforts are being pursued in combination with ATF designs. Respondents
representing Czechia, France and the United Kingdom speculated there may be interested
in the distant future. Respondents from Japan and France reported that national
programmes have considered enrichment above 5% in the past, but both ultimately
abandoned the idea. Respondents from the United Kingdom reported that enrichment above
5% is being considered for advanced reactor designs.

2.4. Qualification of neutronic calculation tools for enrichment of 5-8%
Respondents were asked if there are any efforts ongoing in their country to qualify existing
calculations tools for fuel in the range of 5-8% enrichment.
Most respondents indicated that there were no active efforts to qualify existing tools forfuel in the range of 5-8% enrichment. However, respondents from Japan pointed to a past
R&D project. Criticality experiments simulating 235U enrichment of the level of 5-10%
were conducted at KUCA (Kyoto University Critical Assembly) by using uranium of both
93% and natural enrichment (Shiroya et. al., 1988; Yamamoto et. al., 2007). Through these
STATUS REPORT ON FUEL SAFETY IMPLICATIONS OF EXTENDED ENRICHMENT AND HIGH REACTIVITY/HIGH SUPPRESSION CORE DESIGNSNEA/CSNI/R(2023)4  13

criticality experiments, the accuracy of code (Continuous-Energy Monte Carlo anddeterministic) and nuclear data library was validated (Yamasaki et al., 2007; Nakajima et.
al., 2012). The criticality experiment data were registered in the International Criticality
Safety Benchmark Evaluation Project (ICSBEP) database (case name LEU-METTHERM-005) developed under the framework of the NEA Nuclear Science Committee anddistributed by the NEA Data Bank. Respondents from the United States pointed to ongoing
efforts to qualify the SCALE code for increased enrichment and high burnup applications.
Some lattice physics level studies were performed via SCALE and no unexpected or
anomalous trends were found when increased enrichments (up to 8 wt.%) were analysed.
Respondents from Russia reported that codes being used for neutronic design calculations
for VVERs are certified for the enrichment range up to 6% 235U and that validation was
carried out for enrichment up to 6.5%, taking into account the increase in the burnable
absorber content. Finally, respondents from Czechia (UJV) noted that research reactors
today run with the fuel enriched up to 20% and therefore some calculational tools are
qualified for higher enrichments already. They note that qualification up to 8% might needjustification, but 5% is not perceived as a “hard limit” for current core design tools.

2.5. Ongoing or completed analysis of extended enrichment operation
Respondents were asked if there was any ongoing or completed analysis on the use of fuelwith extended enrichment in their country.
Respondents from Hungary noted that reactor physics calculations were carried out for fast
reactor fuel (ALLEGRO gas-cooled fast reactor design, UO2 fuel with up to 20%
enrichment). Respondents from Czechia (UJV) noted that analysis was completed for
implementation of a 6-year fuel cycle in the VVER-440 reactor (Heraltova, 2015).
Respondents from Japan noted that criticality experiments were conducted at KUCA using
erbia-doped fuel and it was evaluated that the erbia-doped fuel with an enrichment of morethan 5% could be equivalent in criticality safety to uranium fuel of 5wt% or less (Kuroishi& Yamasaki, 2008). The analysis conditions were a 4-loop pressurised water reactor(PWR) and enrichment of 6% (Innovative and Viable Nuclear Energy Technology[IVNET] Development Project, 2008). An analysis of the benchmark problem of reactor
physics for the next generation fuels with a burnup of 70 GWd/t and an enrichment of about
6% was performed (Yamamoto et al., 2002). Criticality experiments were conducted at
Toshiba NCA using low gadolinia concentration. In this experiment, no above-5%
enrichment fuels were used. The aim of the experiment was to measure the reactivity of
low-concentration gadolinia. This low-concentration gadolinia was studied to suppress theinitial reactivity of above-5% enrichment fuel. Usually, gadolinia is used to control thechange of reactivity by fuel burnup. On the other hand, this low-concentration gadolinia
would be used to control criticality safety during the process of fuel fabrication so thatabove-5% enrichment fuels can be treated in a current fabrication facility without largescale modification. Although this experiment was conducted for the study of above-5%
enrichment fuel, only “under-5%” enrichment fuels (maximum 4.9%) were used (Kikuchi
et al., 2009). Respondents from the United States noted that SCALE4 is currently
undergoing development and assessment work for increased enrichments. Four technicalreports were issued from NRC-sponsored work at Oak Ridge National Laboratory (ORNL)
that investigated the effects of high burnup and extended enrichment fuel. These reports
looked at front end transportation canisters (Hall, et al., 2020), ATF (Hall, et al., 2021),
PWR (Hall, et al., 2021) and boiling water reactor (BWR) (Cumberland, et al., 2021) fuelassembly designs. The PWR volume looked at a conventional Westinghouse 17x17 design4

www.ornl.gov/scale/references

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and the BWR volume looked at a convectional GNF-2 10x10 BWR design. Respondents
from Russia explained that an analysis of the use of fuel with an enrichment of more than
5% in 235U for the implementation of 24-month fuel cycles of VVER-1000 and VVER1200 was performed. The impact on the neutronic characteristics of the core, nuclear safety,
accumulation and yield of fission products, activity of corrosion products, discharge andrelease of tritium, and decay heat was studied. Further, an analysis of nuclear safety during
fuel fabrication was performed.

2.6. In-pile testing of extended enrichment fuels
Respondents were asked if there was any intention to carry out in-pile tests with extendedenrichment fuel samples in their country and/or if there are any data available from pastexperiments characterising fuel behaviour in the range of 5-8% enrichment.
Respondents from Japan noted that there were some bilateral projects between the Halden
Reactor Project and Japanese organisations, including the Japan Atomic Energy ResearchInstitute (JAERI, predecessor to Japan Atomic Energy Agency or JAEA), as Instrumented
Fuel Assembly (IFA)-208, 209, 224, 225, in which 5-8% enriched fuels were irradiated in
the Halden Boiling Water Reactor (Yanagisawa, et al., 1978; Uchida, et al., 1976; Uchida& Ichikawa, 1975). But achieved burnup levels were low, less than 10 GWd/tU. Also,
JAERI performed some in-pile tests at the Japan Materials Testing Reactor (JMTR) in the
1970s (Special Committee on Fuel Safety of Nuclear Safety Research Association, 1978;
Special Committee on Fuel Safety of Nuclear Safety Research Association, 1980), but the
cladding used for these tests was thinner than that for commercial reactors, and the achieved
burnup levels were quite low (<1GWd/tU). Respondents from the United States also
referenced data available from Halden, stating they expect historic tests to be sufficient for
the validation of their fuel performance code (FAST) and neutronics code (SCALE) for
enrichment in the range of 5-8%. Respondents from Russia reported that a series of tests
was performed in 2020-2021 at the critical test facility of the Kurchatov Institute todetermine the characteristics of compositions containing fuel with 235U enrichment of 6.5%
and erbium oxide absorber. The test compositions simulate the uranium-erbium fuel withrespect to enrichment, average erbium content and neutron spectrum. Critical tests areplanned with uranium-erbium fuel enriched to 6.5% of 235U.
In addition to the JAERI irradiations referenced above, many highly instrumented
irradiations with >5% enriched UO2 fuel were conducted in the Halden Reactor over theyears. Typical enrichments for these tests were in the 7-13% range, with fuel rod lengths
normally in the 400-600 mm range. Operating conditions were typically 300-600 W/cm,
with the obtained burnup range starting from fresh fuel up to beyond 90 MWd/kgU in somecases.5 These tests were regularly highly instrumented to measure, for example, the fuelcentre temperature, fuel rod pressure and fission gas release (FGR), and fuel stack or
cladding length change. They will have contributed in-pile data for model evaluation andverification on thermal and mechanical analysis, fission gas release and pellet/cladmechanical interactions for fuel modelling codes used by industry.
For most of these tests, post irradiation examinations (PIE) were also conducted on aselection of test rods. The extent of PIE data available varied somewhat between theexperiments, but examinations of note include axial gamma scanning (supplemented with

5

Note that the higher enrichments used in these tests were needed to compensate for the low neutron
flux of the Halden reactor. Thus, the experience with higher enrichment fuel in Halden may not bedirectly applicable to fuel with enrichment greater than 5% in a commercial light water reactor.

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radial gamma scans in a few cases), rod puncturing and FGR analysis and basic
ceramography (i.e. light optical microscopy).
It is worth noting that in addition to the bilateral research sponsored by JAERI at Halden
mentioned above, the Halden Reactor operated with highly enriched fuel material as its“driver” fuel rods for decades. The Halden Reactor is powered by so called “driver” fuelassemblies that operate in heavy water boiling water reactor conditions. Driver fuel rods
are UO2 and resemble commercial fuel rods in their dimensions except that the active fuellength of a driver fuel rod is approximately one metre. The initial enrichment of driver fuelrods included values between 6 and 10%. Due to the nature of the experiments performedin the Halden Boiling Water Reactor, driver fuel assemblies generally operated at highlinear heat rates (Holcombe et al., 2014). The operating experience gained by the Halden
driver fuel may provide useful insight into fuel performance considerations for extendedenrichment fuel.

2.7. New fuel testing requirements
Respondents were asked if there are any requirements for testing new types of highenrichment fuel including: a) material testing; b) testing during manufacturing; c) testingunder conditions as close as possible to operating conditions in a reactor; and/or d)
preliminary operation at research facilities.
The responses all generally described a similar situation, which is that there are no specified
requirements outlined for testing to qualify high enrichment fuel. Rather, most countries
have general requirements that apply to all new fuel types, whether ATF, evolutions ofcurrent fuel types or high enrichment fuel. In general, the requirement to demonstrate that
fuel performance models reflect as-built performance dictates subsequent testing needs.
Several respondents offered additional discussion on the definition of test requirements for
new fuel designs. Respondents from Czechia explained that tests for new fuel types are
specified once the regulator receives the application describing what innovations/changes
are proposed. They explained that in the past, reference in-pile operation of the same fueltype was required before a new design was approved. Respondents from the United
Kingdom said that:
The general expectation is that fuel must be supported by a robust program of pilot
loadings and sufficient relevant operating experience and testing. It is expected thatsuitable operating limits have been defined which have sufficient safety margin toallow for both uncertainty in manufacturing parameters and transient events. It is
expected that licensees have a systematic program of post-irradiation component
examination and testing to ensure that the arguments made in the safety case remain
valid.
Respondents from the Slovak Republic reported that there are no specific requirements formaterial testing or preliminary operation at research facilities. However, there are
requirements related to testing during manufacturing (covered by the nuclear power plant
operator) and testing under conditions as close as possible to operating conditions (includedin the requirements of Slovak legislation). Respondents from Russia noted that when using
new burnable absorbers, the supervisory body of the Russian Federation requires materials
science research. They also note that there are requirements for trial operation of individual
fuel assemblies with increased enrichment in operating reactors.

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2.8. Relevant operating limits
Respondents were asked about specific operating limits that may be relevant to highreactivity/high suppression/extended operating cycle core designs. The specific operation
limits included in the questionnaire were:•assembly or fuel rod peaking factor;•

limits on burnable poisons;•moderator temperature coefficient;•cycle length with respect to instrument calibration (drift);•shutdown margin.

In response to this question, respondents from France explained that enrichment was an
input to neutronic calculations used to demonstrate that, when the reactor core is critical,
the inherent reactivity feedbacks are able to guaranty the core stability. The neutronic
conception and the safety assessment lead to the following limits:
1. A higher discharge burnup has consequences for fuel in the safety studies(fragmentation, relocation and dispersion during postulated design basis accidents,
more frequent and ampler fuel assembly or rod bowing);
2. A higher boron concentration or supplementary rod cluster control assemblies
might be required to ensure subcriticality in shutdown states.
Respondents from the United Kingdom also noted generally that all the operation limits
identified by the question are considered as part of the reload design and safety case. For
either existing or new plants, these parameters are usually limited as part of the safety caseto demonstrate fault tolerance with the desired cycle length and load factor, but limits on
individual parameters are not prescribed by the regulator.
Finally, respondents from Spain noted generally that there are no explicit requirements inany of the operating limits inquired, except those stated in the respective safety analysis as
an input or assumption for the safety analysis. They noted that as a general requirement, it
is legally stated that: “The analysis of accidents shall be conducted with safety margins
such that the maintenance of the safety functions is guaranteed with due considerations tothe uncertainties inherent to the processes involved.”

2.9. Assembly or fuel rod peaking factor
Respondents from Czechia, Korea, Japan, the Slovak Republic and Russia reported that
there are limits for assembly or fuel rod peaking factor. Respondents from Japan specified
that for PWRs, a power peaking factor is evaluated for each reload core (Japan Electric
Association Code, 2018). For BWRs, a maximum linear heat generation rate or a minimum
critical power ratio is evaluated for each reload core (Japan Electric Association Code,2018). Respondents from Germany explained that requirements are dependent on the safetyassessment of every specific nuclear power plant and core loading. Respondents from
Hungary, France, Sweden, the United Kingdom and the United States explained that there
are no generic limits defined, though limits for each reload are established to demonstratethat safety criteria for operation and accidents are met.

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2.10. Use of burnable poisons
Many countries authorise gadolinia-doped fuel and some respondents cited practicalboundaries for gadolinia doping, though there are no established limits. With that said, insome countries fuel suppliers limit the enrichment of the gadolinia-doped rods to prevent
them from becoming limiting in the safety analysis.
Respondents from Japan noted that fuel is generally designed to be within a gadolinia
concentration of 10% or less. Respondents from Sweden explained that there are limits on
different combinations of nodal enrichment and burnable absorber (uranium enrichment in
combination with the amount of gadolinia within a rod). In Sweden there are no establishedmaximal limits regarding the amount of burnable poison (number of rods or gadolinia
concentration) since having annual cycles reduces the need for large amounts of burnablepoison. Respondents from France reported that any limits on burnable poisons are practicaland imposed by the code qualifications. Respondents from Hungary reported that VVER440 fuel assemblies, which have a total of 126 fuel rods, are in use with zero, three or six
gadolinia-containing rods and varying levels of enrichment. Examples can be found in(Végh et al., 2015).
Respondents from the United States reported that fuel designs utilising gadolinia andintegral fuel burnable absorbers (IFBA) have been approved. As in other countries, fuelvendors may have established limits for the number of rods with gadolinia or burnablepoisons, or the concentration of those materials within a fuel rod, in their approved
methods. Nevertheless, there are no regulatory limits established that limit the use ofburnable poisons in the United States.

2.11. Moderator temperature coefficient (MTC)
All respondents reported that there are limits in place in their country for MTC, specificallythat it must be negative. Respondents from the United States elaborated that while there are
no explicit limits for MTC, licensees must demonstrate that they meet safety criteria in theiroperational and accident analyses. NRC staff guidance states that “MTC should be nonpositive over the entire fuel cycle when the reactor is at a significant power level” (USNRC, 2007).

2.12. Cycle length and considerations of instrument calibration
Currently, many countries operate with cycle lengths of approximately one year, but mostare moving to longer cycle lengths. Respondents from Japan noted a common cycle length
to be 13 months and respondents from Hungary reported that cycles lengths have recently
increased to 15 months. Respondents from Czechia also reported that cycle lengths arestarting to shift to 15-18 months. Respondents from Russia reported that a typical cycle
length is already 18 months, specifying that all VVER-1000 plants are operating in the
18-month cycle and that VVER-1200 plants are in a process of transition to the 18-monthfuel cycle (to be finished by 2025). Respondents from France note that 900 megawatt
electric reactors (3-loops) remain on 12 months cycles; however, for 1 300 and 1 450
megawatt electric reactors (4-loops) the cycle lengths are around 18 months.
There are no explicit limits on cycle length based solely on instrument calibration
requirements, but many respondents reported that there are requirements for instrument
calibration on established frequencies. Respondents from France added that “most of themandatory periodic tests can be performed during the cycle (with reactor at zero power if

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18  NEA/CSNI/R(2023)4
necessary). Hence, this aspect does not limit the cycle length in France.” Respondents from
Russia said:
When an 18-month cycle was being introduced, there were some problems withjustifying the duration between equipment tests. Regulations strictly required
testing some equipment at least every 18 months. Current regulations state that thetesting frequency shall be justified in the design. Thus, currently there are no strict
requirements in Russian regulations regarding the instrument calibration frequency,
system design shall contain all necessary justifications. During transition of the
VVER plants to the 18-month cycle, we did not face any limitations originatingfrom the instrumentation calibration and we do not expect such limitations for the
24-month cycle. Of course, implementation of the 24-month fuel cycle will require
additional justification of the instrument calibration frequency.
Considering the link between cycle length and enrichment, respondents from Korea notedthat “although the cycle length can be technically increased by increasing the fuelenrichment, the maximum cycle length can be determined by the excess reactivity, MTC(Moderator Temperature Coefficient) at the beginning of the cycle. The operating cycle
length is currently set according to the periodic inspection, but the amount of the length is
not restricted by excess reactivity or MTC.”

2.13. Shutdown margin
Respondents reported that there are constraints on shutdown margin, and in many cases
these constraints are established in the safety analysis. Respondents from Japan reported
that under operational conditions, the shutdown margin must be designed to allow the core
to go subcritical even when the single control rod with the highest worth is completely
pulled out of the core and cannot be inserted (Japan Electric Association Code, 2018).
Respondents from Sweden explained that the constraints are reactor-specific andestablished in safety analyses, where in controlled shutdown states, subcriticality should be
at least 1%. Respondents from France said that shutdown margin must be shown to be
sufficiently high in the safety demonstration of accident scenarios, such as a steam line
break at zero power. Respondents from Hungary noted that shutdown margin cannot beless than 2% according to the design provided by the fuel supplier. Respondents from theUnited States explained that while there is no explicit limit on shutdown margin, General
Design Criterion 27 requires that the reactor coolant systems have a combined capability,
in conjunction with poison addition by the emergency core cooling system, of reliably
controlling reactivity changes under postulated accident conditions, with appropriate
margin for stuck rods (General Design Criteria for Nuclear Power Plants, 2022).
Respondents from Germany said that shutdown reactivity for boron injection systems is
5% for PWRs (Nuclear Safety Standards Commission, 2012). They explained that the point
in time with the highest reactivity is to be selected. This does not always have to be at the
beginning of the cycle (e.g. for BWRs with burnable neutron poisons the reactivity canincrease at the beginning before it decreases again) (see Sections 3.2 (3), 3.2 (4) and 3.2(7) of the Safety Requirements for Nuclear Power Plants (Federal Ministry for theEnvironment, Nature Conservation and Nuclear Society, 2015a), and Interpretation I-1 3(1)(Federal Ministry for the Environment, Nature Conservation and Nuclear Safety, 2015b).

2.14. Human error in core loading
Respondents were asked if there are requirements in place to specifically address thepotential for “human error” in core loading. If so, respondents were asked if the limits relate

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in any way with the possibility that high reactivity/high suppression/extended operating
cycle core designs may be more sensitive to core loading errors.
Respondents from Russia confirmed that the potential for human error in core loading must
be addressed in the core loading process (see NP-006-16 of [Scientific and Engineering
Centre for Nuclear and Radiation Safety, 2018]), explaining that typically the fuel design
justification includes analysis of such an error aimed at demonstrating that the error is
detected by monitoring systems and no negative consequences occur. Respondents from
Germany explained that misloading is postulated in events on safety levels 2 (AnticipatedOperational Occurrence) and 3 (Design Basis Accident) (see Annex 2 of SiAnf events D225, D3-17, S2-20 and S3-15 of [Federal Ministry for the Environment, Nature Conservation
and Nuclear Society, 2015a]). In this way, human errors can be intercepted. Otherwise
SiAnf 3.1 (13) (Federal Ministry for the Environment, Nature Conservation and Nuclear
Society, 2015a) is still generally valid. All other respondents reported that there are no
requirements specifically addressing the potential of human error in core loading. However,the potential of an error in core loading could be prevented by ensuring limits are met inreactor physics tests before operating (e.g. ANSI/ANS 19.6.1 “Reload Startup Physics
Tests for Pressurized Water Reactor”) as well as adherence to operating procedures
intended to prevent human error in core loading.

2.15. Mixed cores and transition cores
Respondents were asked if there are particular restrictions or operating requirements formixed cores and transition cores in their country.
Respondents reported that there are no specific restrictions or operating requirements formixed cores and transition cores. Respondents explained that the requirements for mixed
cores are the same as for homogeneous cores; the safety criteria must be fulfilled. The
differences in flowrate, neutronic power (if any), etc. must be taken into account.
Respondents from France noted that for transition cores, besides the “classical” safety
demonstration, the main issue is often related to pellet-cladding interaction risk, as theburnup pattern within the core is often quite different from the equilibrium cycle (for which
the demonstration related to the absence of pellet-cladding interaction risk is performed).
Respondents from the Slovak Republic noted that if operating requirements are needed,
they are established as necessary in co-operation with the fuel producer. Respondents from
Spain noted that mixed cores are allowed but must be licensed specifically by the regulatory
body.

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3. Summary of safety implications evaluation

One objective of this activity was to utilise the expertise of the WGFS to evaluate possible
fuel safety implications of extended enrichment (5-8%) and high reactivity/highsuppression core designs. The evaluation was facilitated by a comprehensive table of fuelperformance criteria utilised in the NEA CSNI Technical Opinion Paper No. 19 Applicability of Nuclear Fuel Safety Criteria to Accident-Tolerant Fuel Designs (NEA,2022). The evaluation table can be found in Appendix B. The key findings of the evaluation
are captured below.

3.1. Impacts of extended enrichment on design limits
In the evaluation of extended enrichment, there were few instances where enrichment itself
was expected to have a clear impact on an existing design limit. However, increasedenrichment and additional burnable absorbers may change core-wide and individual fuelrod axial and radial power profiles and the excess reactivity in the core. These changes
could impact the margin to applicable safety limits.
As one example, increasing the enrichment may impact fuel rod power peaking factors
assumed in plant safety analyses, particularly for high burnup rods. Under existingenrichment limits, there is little residual fissile material in the fuel at high burnup;
increasing the enrichment may allow these higher burnup rods to reach the failure limits incertain accident scenarios (e.g. during a reactivity-initiated accident).
As a second example, changing the enrichment would impact the fuel rod radial power
profile. This would impact various fuel performance characteristics, including the fuel rod
radial temperature profile, fission gas production and release, and fuel-cladding gap
thermal conductivity. Furthermore, the addition of gadolinia to UO2 fuel reduces its thermal
conductivity,6 while neutron absorption by zirconium diboride added to fuel produces
helium gas that impacts the rod internal pressure. All these effects could impact the margins
to fuel rod safety criteria (e.g. by reducing the margin to fuel centerline melt or bychallenging rod internal pressure limits).
Overall, the task members noted that the impact of enrichment on the various fuel safety
criteria may be minor, but the evaluation concludes that design-specific analysis andconfirmation of the safety basis for various limits should be performed. In some cases, thismay mean analysis to quantify the impact of enrichment and the analysis will show that the
impact is insignificant. In these cases, the existing design limits would be applicable in thecase of extended enrichment. In other cases, an analysis may reveal the impact ofenrichment is not insignificant, though the impact could be bounded by the basis for the
existing limits. In these cases, too, the existing analytical limit (e.g. deposited enthalpy
limits during reactivity accidents, shutdown margin requirements) would be applicable inthe case of extended enrichment. Even if there are many instances where the conclusion ofsuch an analysis could be seen to confirm the adequacy of an existing analytical limit, theanalysis must still be performed to reach that conclusion.
6

At the same time, gadolinia rods produce less power than rods without gadolinia. In current core
designs, the power reduction in gadolinia rods more than compensates for the reduced thermal
conductivity, such that gadolinia rods are typically not limiting in terms of meeting fuel rod safety
criteria. However, this may not be true for future high suppression core designs with increasedenrichment fuel.

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3.2. Opportunities for collaborative research
During the course of this review, the task group did not identify any major experimental
data gaps for validation of the computer codes. The task group noted that numerous
experiments have been conducted to study the neutronic and thermal mechanical behaviourof UO2 fuel with enrichment above 5% and of gadolinia-doped fuel. Much of this data isavailable to NEA member countries. There is also extensive operating experience with fuelcontaining burnable absorbers such as gadolinia or zirconium diboride; however, much ofthis operating experience is not publicly available. It would be beneficial to provide a morethorough description of publicly available experiments involving enrichment above 5% or
gadolinia-doped fuel. This task could be accomplished as part of a future NEA
collaborative research activity.
Additionally, the task group has noted that there may be some value in performing
additional experiments on extended enrichment fuel or gadolinia-doped fuel, particularly
for high burnup fuel under design basis accident conditions. The group noted that two NEA
joint projects are currently studying the behaviour of gadolinia-doped fuel during loss-ofcoolant accident (LOCA) conditions: the Studsvik Cladding Integrity Project and theLOCA-MIR Joint Experimental Programme under the Framework for Irradiation
Experiments. Furthermore, the proposed Laser Ablation of Gadolinium Evolution Radially(LAGER) project would look at the behaviour of gadolinia-doped fuel early in life tovalidate neutronics and fuel performance codes. Additional experiments could be proposed
to study the behaviour of gadolinia-doped fuel or standard UO2 fuel with enrichment greaterthan 5% in power ramp or reactivity-initiated accident conditions, especially for fuel athigh burnup or with new cladding types. The SPARE project is currently harvestingmaterial from the Halden Reactor and could potentially provide higher enrichment material
for power ramp or design basis accident testing.
There are also opportunities to conduct neutronic code-to-code benchmark activities toassess the impact of enrichment on parameters like reactivity coefficients, shutdown
margin, power peaking factors or decay heat. Such activities would require some designspecific information, which may be difficult to obtain from fuel vendors because of
proprietary information concerns. However, there have been some recent activities in the
United States related to high burnup and extended enrichment core designs that could serve
as a starting point for collaborative research efforts (Hall, et al., 2020; Hall, et al., 2021;
Hall, et al., 2021; Cumberland, et al., 2021; Capps, et al., 2021).
Further benchmarks could be conducted to assess the impact of enrichment on fuel thermal
mechanical behaviour, including its effects on the radial power profile, fuel temperature
profile and fission gas release. Adverse impacts on any of these parameters could reducethe margin to fuel safety limits (e.g. rod internal pressure limits, restrictions on fuelcenterline melting).

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4. Conclusions and recommendations

4.1. Conclusions
This report discussed past experience with extended enrichment and high reactivity/highsuppression core designs in NEA member counties, as well as past experimental
programmes dedicated to the behaviour of uranium dioxide fuel with enrichment above
5%. While there is little operating experience with extended enrichment fuel in commercial
nuclear power plants, there have been several experimental campaigns that focused on the
behaviour of extended enrichment uranium dioxide fuels. Furthermore, the Halden Boiling
Water Reactor operated for years with driver fuel above 5% enrichment. Thus, there issome data available to qualify fuel performance and neutronics codes used to perform
safety analysis. Based on this experience, the task group did not identify any major
experimental gaps related to extended enrichment fuel or high reactivity/high suppression
core designs; however, several potential areas of research have been identified, as discussed
below.
The report also provided an evaluation of the fuel safety implications of extendedenrichment (5-8%) fuel and high reactivity/high suppression core designs. The task group
noted that there are few instances where enrichment has a direct impact and fuel safety
limits; however, enrichment and high reactivity/high suppression core designs will impact
core-wide axial and radial power profiles, which may pose a challenge to meeting existing
limits. Furthermore, enrichment will impact the fuel radial power profile, which will in turnaffect the pellet temperature profile, fission gas release and various other parameters that
may impact the margin to design limits under normal and accident conditions.

4.2. Recommendations
The task group identified several potential areas for collaborative research. Among thoseareas, the following research activities should be prioritised. The expert opinions collected
within this report should be revisited during the activities to determine whether they can be
confirmed by the work.
1. Given the limited test data on gadolinia-doped fuel and uranium dioxide fuelbetween 5-8% enrichment, the group recommends performing additional
experiments on gadolinia-doped and extended enrichment fuel under design basis
accident conditions, especially for fuel at high burnup or with new cladding types.
In particular, the group recommends performing power ramp or reactivity-initiatedaccident (RIA) tests to confirm that existing design limits are applicable to thesefuel types. Such experiments could be proposed as part of the Second Framework
for Irradiation Experiments (FIDES-II) NEA joint project. Fuel rods from lead test
assemblies irradiated in commercial nuclear power plants in NEA member states
or fuel from the Halden Boiling Water Reactor could potentially be used in thesepower ramp or RIA tests.
2. Further, the group recommends performing two benchmark studies regarding
extended enrichment fuel.
a. The first benchmark would focus on fuel performance calculations of asingle rod with enrichment of 5-8%. The goal of the benchmark is toevaluate the impact of enrichment on significant figures such as fission gas
release and fuel centerline temperature. The benchmark could be based on
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existing experimental data, or it could be a purely computational exercise
designed to evaluate the impact of enrichment on predictions of fuelperformance codes used in NEA member countries. This benchmark could
be conducted as a WGFS activity.
b. The second benchmark would focus on neutronic calculations for a highreactivity/high suppression core design with fuel enrichment above 5%.
The goal of the benchmark would be to evaluate important core design
parameters, including power peaking factors, shutdown margins andreactivity coefficients. As mentioned, there are related efforts underway insome NEA member countries (e.g. the United States); however, it would
be beneficial to establish a benchmark problem to evaluate neutronics
codes used by the NEA member countries and to identify potential
challenges associated with meeting existing design limits for highreactivity/high suppression core designs with enrichment above 5%.

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Appendix A: Questionnaire in preparation for a status report on fuel safety
implications of extended enrichment and high reactivity/high suppression
core designs
Background:
The objective of this NEA Working Group on Fuel Safety (WGFS) activity is to consolidate information
regarding fuel enrichment limits of UO2 fuel for light water reactors (LWRs) in NEA member countries
and utilise the expertise of the WGFS to evaluate possible fuel safety implications of extended enrichment(5-8%) and high reactivity/high suppression core designs. The questionnaire below supports the first step:
to collect information regarding fuel enrichment limits and high reactivity/high suppression core designs
in NEA member countries.
Discussion:
The CAPS includes high reactivity/high suppression/extended operating cycle core designs because of therelevancy of these core designs to operation with extended enrichment. One of the biggest implications ofoperating with enrichments above 5% will be that the fuel will have an increased initial reactivity excess,which will obviously need to be absorbed (e.g. by burnable poisons or other means). In fact, some fueldesigns, such as MOX, or countries operating on 24-month cycles, may already be dealing with core
designs that require substantial power suppression early in life. This CAPS will capture experiencegenerally with core designs that require high power suppression as a way to gain insight into one of thefuel performance concerns with enrichment above 5%. In other words, the CAPS will include one of thefuel performance aspects of high enrichment, defined broadly, to gain some insight despite a lack of
specific operating or performance experience with high enrichment.
Questionnaire:
Question
1. Is there an explicit limit on 235U enrichment
established by a regulatory limit in your country? If
so, is the limit in reference to fuel manufacturing
operation, transportation, storage and/or use?
2. Are there established operating limits (e.g. inherent
reactivity feedback), or other practical restrictions,which effectively limit 235U enrichment in yourcountry? If so, is the limit in reference to fuelmanufacturing operation, transportation, storage
and/or use?
3. Is there interest in your country to extend the current
limit of LWRs? If so, is there any indication that an
intended use of high enrichment fuel would be
combined with any accident tolerant fuel (ATF)
design?
4. Are there any efforts ongoing in your country toqualify existing calculations tools for fuel in therange of 5-8% enrichment? If so, can you state thelevel of qualification or any determination of whereenhancements would be needed?
5. Is there any ongoing or completed analysis on theuse of fuel with extended enrichment in yourcountry? If yes, please specify the reactor type, fuelenrichment and analysed scenarios.

Response

STATUS REPORT ON FUEL SAFETY IMPLICATIONS OF EXTENDED ENRICHMENT AND HIGH REACTIVITY/HIGH SUPPRESSION CORE DESIGNS28  NEA/CSNI/R(2023)4
6. Is there any intention to carry out in-pile tests withextended enrichment fuel samples in your country?
Is there any data available from past experimentscharacterising fuel behaviour in the range of 5-8%
enrichment?
7. Are there any requirements for testing new types ofhigh enrichment fuel including:a. material testingb. testing during manufacturingc. testing under conditions as close as possible
to operating conditions in a reactor
d. preliminary operation at research facilities
As discussed above, the implications of fuel with 5-8% 235U enrichment shall be considered broadly in thisstatus report. The questions below focus on high reactivity/high suppression/extended operating cycle core
designs.
Question
8. Are there established limits, or any practicalconstraints, in your country for included, but notlimited to:a. Assembly or fuel rod peaking factor
b. Use of burnable poisonsc. Moderator temperature coefficientd. Cycle length with respect to instrument
calibration (drift)
e. Shutdown margin
9. Are there requirements in place to specificallyaddress the potential for “human error” in coreloading? If so, do these limits relate in any way withthe possibility that high reactivity/highsuppression/extended operating cycle core designs
may be more sensitive to core loading errors?
10. Are there particular restrictions or operating
requirements for mixed cores and transition cores inyour country?

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Appendix B: Safety implications evaluation table

The safety implications evaluation table for extended enrichment fuel and high reactivity/highsuppression core designs can be found at: www.oecd-nea.org/r2023table. The table is based on theevaluation tables developed for the NEA CSNI Technical Opinion Paper No. 19 - Applicability of
Nuclear Fuel Safety Criteria to Accident-Tolerant Fuel Designs (NEA, 2022). Note that some ofthe parameters evaluated in the 2022 publication tables were not relevant to fuel enrichment
because they focus on cladding phenomena; these boxes have been greyed out in the extendedenrichment table.

STATUS REPORT ON FUEL SAFETY IMPLICATIONS OF EXTENDED ENRICHMENT AND HIGH REACTIVITY/HIGH SUPPRESSION CORE DESIGNS