COMMISSION STAFF WORKING PAPER Accompanying the document Report to the Council and the European Parliament Report from the Commission to the Council and the European Parliament. Operation of the High Flux reactor in the year 2010
Date: 2012-04-16
Institutions: European Commission
Subjects: Belgium, France, Joint Research Centre, Member States' contribution, Netherlands, energy research, nuclear chemistry, nuclear fuel, nuclear reactor, nuclear safety, nuclear technology

EUROPEAN COMMISSION 

Brussels, 16.4.2012  
SWD(2012) 86 final 

COMMISSION STAFF WORKING PAPER 
Accompanying the document 

Report to the Council and the European Parliament 

Report from the Commission to the Council and the European Parliament "Operation 
of the High Flux reactor in the year 2010" 

{COM(2012) 171 final} 

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COMMISSION STAFF WORKING PAPER 
Accompanying the document 

Report to the Council and the European Parliament 

Report from the Commission to the Council and the European Parliament "Operation 
of the High Flux reactor in the year 2010" 

1. 

INTRODUCTION 

This staff working document is a companion document to the Report from the Commission to 
The  Council  and  the  European  Parliament  "Operation  of  the  High  Flux  Reactor  in  the  year 
2010" 

The  High  Flux  Reactor  (HFR),  managed  by  the  Institute  for  Energy  and  Transport  of  the 
European  Commission’s  Joint  Research  Centre  (JRC),  is  one  of  the  most  powerful  multi-
purpose materials testing reactors in the world. 

The  HFR  is  of  the  tank-in-pool  type,  light  water  cooled  and  moderated  and  operated  at  45 
MW. In operation since 1961, and following a new vessel replacement in 1984, the HFR has a 
technical life beyond the year 2015. The reactor provides a variety of irradiation facilities and 
possibilities in the reactor core, in the reflector region and in the poolside. Horizontal beam 
tubes are available for research with neutrons as are gamma irradiation facilities. Furthermore, 
excellently  equipped  hot  cell  laboratories  on  the  Petten  site  provide  virtually  all  envisaged 
Post Irradiation Examinations (PIE) possibilities. 

The  close  co-operation  between  the  JRC  and  the  Nuclear  Research  and  Consultancy  Group 
(NRG) on all aspects of nuclear research and technology has led to a unique HFR structure, in 
which both organisations are involved. The European Atomic Energy Community (Euratom) 
is the owner of the plant (for a lease of 99 years) and the JRC the plant and budget manager. 
The JRC develops a platform around HFR as a tool for European collaborative programmes 
while  NRG  operates  and  maintains  the  plant  and  manages  the  commercial  activities  around 
the  reactor.  Furthermore  each  organisation  provides  complementary  possibilities  around  the 
reactor  activities,  such  as  the  hot  cell  facilities  of  NRG  or  the  experiment  commissioning 
laboratory of the JRC. 

As of February 2005, NRG has become the holder of the operation licence granted under the 
Dutch Nuclear Energy Law.  

During  the  last  three  decades  the  HFR  has  been  operated  from  supplementary  research 
programmes  regularly  discussed  by  the  European  Council.  On  25  May  2009,  the  European 
Council  adopted  a  three-year  (2009-2011)  supplementary  research  programme  to  be 
implemented  by  the  JRC  for  Euratom  concerning  the  operation  of  the  Community’s  High 
Flux Reactor.  

The present document reports the implementation of the scientific and technical objectives for 
the  year  2010.  The  report  also  provides  information  regarding  the  financial  contributions 
received  for  the  execution  of  the  programme  and  the  yearly  contribution  to  the 

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decommissioning  fund  that  the  supplementary  research  programme  provides  to  Euratom, 
owner of the HFR. 

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HFR: REACTOR MANAGEMENT 

The  HFR  reactor  is  operated  by  NRG,  who  has  an  operating  licence  granted  by  the  Dutch 
national regulator, the Kernfysische Dienst. As for nuclear power plants, the HFR is subject to 
legally required 10-year periodic reviews which are performed by NRG. The HFR has been 
also  the  subject  of  an  independent  INSARR  (Integrated  Safety  Assessment  for  Research 
Reactors)  review  performed  by  the  International  Atomic  Energy  Agency  (IAEA)  in  April 
2011. 

HFR Safety, Operation and Related Services 

Operating Schedule 

Due to the repair of the HFR primary cooling system (Bottom Plug Liner – BPL) in 2010, the 
operation cycles of the HFR could not be planned in advance. At the beginning of the year, 
the HFR was in operation for cycle 2009-10 which consisted of a scheduled number of 14.3 
operation  days.  Cycle  2010-01  was  planned  before  the  BPL  repair  and  consisted  of  a 
scheduled number of 31.7 operation days. The BPL repair started on February 20th, 2010 in 
parallel  to  the  periodic  in-service  inspection  and  the  yearly  containment  leak  test.  The 
preparation  and  implementation  of  the  repair  was  monitored  and  inspected  by  the  Dutch 
nuclear safety regulatory authority in accordance with the applicable national regulations and 
requirements and the reactor was restarted on September 9th 2010. A new cycle pattern was 
prepared  consisting  of  a  scheduled  total  number  of  102  operation  days.  The  target  was  to 
ensure  the  availability  of  sufficient  irradiation  capacity  till  the  end  of  2010.  The  HFR  has 
therefore been in operation in 2010 for a total of 146 days (Figure 1). 

Availability

e
g
a
t
n
e
c
r
e
P

120

100

80

60

40

20

0

2000

2001

2002

2003

2004

2005

2006

2007

2008

2009

2010

A1: Number of days
at power / (number of
days at power +
number of days of
unscheduled
shutdowns)

A2: Number of days
at power / 365

A3: Number of days
at power / number of
scheduled operating
days

Year

Figure 1: HFR availability 

Nominal power has been 45 MW (standard operating power), with a total energy production 
of approximately 6577.9 MWday, corresponding to a fuel consumption of about 8.2 kg 235U. 

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Flux  measurement  at  the  reactor  power  of  500  kW  in  support  of  the  HFR  conversion 
programme has been carried out during cycle 2010-04. After the scheduled end of operation 
of each cycle, the shut-downs included activities performed in the framework of the regular 
HFR’s operators training. The detailed operating characteristics for 2010 are given in Table 1. 

All details on power interruptions and power disturbances, which occurred in 2010, are given 
in  Table  2.  It  shows  that  nine  automatic  reactor  shut-downs  (scrams),  two  manual  reactor 
shut-downs,  one  manual  power  decrease  and  one  automatic  power  decrease  occurred  (see 
Figure  2).  Two  of  these  scrams  were  due  to  human  intervention  and  the  remaining  scrams 
were due to intervention by the safety systems of the reactor instrumentation devices. 

Figure 2: HFR unscheduled shutdowns 

Maintenance Activities 

In  2010  the  maintenance  activities  consisted  of  the  preventive,  corrective  and  break  down 
maintenance of all systems, structures and components of the HFR as described in the annual 
and  long-term  maintenance  plans. These activities  are executed  with the objective to  enable 
the  safe  and  reliable  operation  of  the  HFR  and  to  prevent  inadvertent  scrams  caused  by 
insufficient maintenance. Also, the periodic leak test as one of the license requirements (0.2 
bars  overpressure  for  24  hours  duration)  and  the  extended  In-Service  Inspection  (ISI) 
including  the  measurements  relative  to  the  reparation  of  the  BPL  were  successfully 
performed. All modifications were implemented after the revision of the plant description and 
operating instructions and following successful commissioning and testing. 

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Table 1: 2010 operational characteristics 

OPERATING TIME 

SHUT-DOWN TIME 

Cycle 

HFR 

Generated  

Planned 

Low 

Nominal 

Other 

Total 

Planned 

Unsche-
duled 

Number 
Interruptions 

of 

Begin-End 

Cycle 

Energy 

Power 

Power 

Use 

2010 

MWd 

hrs 

h.min 

h.min 

h.min 

h.min 

h.min 

h.min 

PD 

Scram 

01.01 – 15.01 

09.10 

689.17 

16.01 – 19.02 

10.01 

1424.77 

344 

760 

00.42 

343.10 

343.52 

16.00 

0.08 

1 

02.10 

759.17 

761.27 

78.33 

20.02 – 08.09 

BPL repair and maintenance period 

20.36 

4803.24 

09.09 – 08.10 

10.02 

1207.93 

09.10 – 12.11 

10.03 

1391.70 

13.11 – 17.12 

10.04 

1351.25 

18.12 – 31.12 

11.01 

513.07 

688 

760 

736 

274 

05.15 

640.06 

645.21 

29.34 

45.05 

10 

02.40 

741.19 

743.59 

79.40 

17.21 

2 

01.55 

719.50 

721.45 

104.00 

14.15 

01.57 

274.01 

275.58 

60.02 

TOTAL : 

6577.89 

3562 

14.39 

3477.43 

20.36 

3512.58 

5171.13 

76.49 

2 

11 

Percentage of total time in 2010 (8760 h) : 

Percentage of planned oper. time (3562 h) : 

0.17 

0.41 

39.70 

97.63 

0.24 

0.58 

40.10 

59.03 

0.88 

98.62 

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DATE  CYCLE  TIME 

OF 

ACTION 

RESTART 
OR 
POWER IN-
CREASE 

NOMINAL/
ORIGINAL
POWER 

2010 

hour 

hour 

hour 

h.min 

Table 2: 2010 full power interruptions of HFR 

ELAPSED TIME TO 

DISTURBANCE CODE 

RESTART  OR 
POWER 
INCREASE 

NOMINAL/ 

1 

2 

3 

ORIGINAL 

MW 

POWER 

h.min 

REACTOR 
SYSTEM 
EXPERIMENT 
CODE 

COMMENTS 

OR 

13 Jan  09.10 

09.17 

09.25 

10.07 

00.08 

00.21 

MS  0 

A  R  Secondary 

system 

10.02 

17.42 

17.46 

17.52 

00.04 

00.10 

AS 

0 

R 

I 

Primary system 

10.02 

20.47 

20.51 

21.02 

00.04 

00.15 

AS 

0 

R 

I 

Primary system 

09 
Sept 

09 
Sept 

10 Sep  10.02 

00.56 

00.59 

01.13 

00.03 

00.17 

AS 

0 

P 

I 

Exp 292-01 

16 Sep  10.02 

16.12 

16.21 

16.43 

00.09 

00.31 

AS 

0 

P  M  Exp. 354-02 

17 Sep  10.02 

09.13 

09.18 

09.48 

00.05 

00.35 

AS 

0 

P  H  Exp. 354-02 

23 Sep   10.02 

16.18 

16.24 

16.57 

00.06 

00.39 

AS 

0 

P  M  Exp. 354-01 

The  reactor  was  manually  shut-down 
preventively; due to ice crust forming in 
the  NH-channel,  not  enough  cooling 
water was available. 

Primary  flow  <90%  caused  a  reactor 
scram. 

Scram  on  primary  flow  <90%.  Reactor 
restarted  after  the  primary  pressure  and 
flow was checked. 

Too 
for 
low  cool  water  pressure 
experiment  TIRO-1  caused  a  reactor 
scram. 

Too  high  cool  water  pressure 
for 
experiment TYCOMO-2 caused a reactor 
scram. 

Incorrect  cool  water  flow  controller 
readjustment leading to a reactor scram. 

The  cool  water  hose  of  experiment 
TYCOMO-1  slipped  off  and  caused  a 
reactor scram. 

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24 Sep  10.02 

08.00 

MS  0 

P  R  Exp.  354-01  and 
354-02 

To  renew  the  cool  water  hoses  of  the 
experiments  TYCOMO  I  and  2  the 
reactor was manually shut-down. 

26 Sep  10.02 

26 Sep  10.02 

19.44 

12.24 

19.48 

13.11 

20.17 

28.24 

00.04 

29.11 

00.33 

AS 

0 

P 

I 

Exp. 292-01 

07 Oct  10.02 

13.32 

13.38 

13.52 

00.06 

00.20 

AS 

0 

R 

I 

Instrumentation 

Due  to  manipulations  with  experiment 
TIRO-1 the cool water pressure dropped 
leading to a reactor scram. 

Instrumentation  fault  caused  a  reactor 
scram. 

12 Oct  10.03 

12.34 

17 Oct  10.03 

14.32 

12.54 

15.01 

00.20 

MP  30 

E  R  Exp. 333-01 

Unexpected 
experiment HICU 333-01 was checked. 

temperature 

rise 

in 

00.29 

AP 

10 

A  E  Ventilation 

system 

Reactor  hall  ventilation  system  switched 
off  automatically  with  an  automatic 
power decrease as result. 

21 Oct  10.03 

15.53 

15.57 

16.09 

00.04 

00.12 

AS 

0 

R  H  Reactor interlock  Activation  of  incorrect  switch  of  the 
reactor  interlock  with  a  reactor  scram  as 
result. 

1. LEADING TO 

- automatic shut-down 

- manual shut-down 

- automatic power decrease 

- manual power decrease 

AS 

MS 

AP 

MP 

2. RELATED TO 

- reactor 

- experiment 

- auxiliary system 

- production facility 

R 

E 

A 

P 

3. CAUSE 

- scheduled 

- requirements 

- instrumentation 

- mechanical 

- electrical 

- human 

S 

R 

I 

M 

E 

H 

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2. 

REPAIR OF THE HFR PRIMARY COOLANT SYSTEM 

The  HFR  reactor  vessel  is  located  in  a  concrete  pool  with  floor  and  walls  lined  with 
aluminium.  For  cooling  and  radiation  protection  purposes  the  pool  is  filled  with  water 
(approximately 9 meters deep); the top of the pool is open to enable loading and unloading of 
fuel and experimental rigs. Two sections of the reactor primary coolant system are embedded 
in the concrete at the bottom of the reactor pool. These parts are conical in shape and usually 
identified as “reducers”. During the in-service inspection performed in the summer of 2005, 
some deformations were detected in two parts of the section of the reactor’s primary cooling 
water  system  embedded  in  concrete.  Investigations  were  immediately  started  to  analyse  the 
cause of the deformations and to determine their possible impact on the safety and reliability 
of  the  HFR.  The  frequency  of  the  periodic  inspections  of  both  the  reducers  and  the 
deformations was also increased. 

The  deformations  were  thought  to  derive  from  a  galvanic  corrosion  process  (due  to  a 
combination  of  carbon  steel,  moisture  and  aluminium)  causing  a  decrease  of  the  reducer’s 
aluminium  wall  thickness  and  expansion  of  the  corrosion  product  (having  a  volume  2  to  3 
times greater than that of aluminium), pushing the reducer wall inwards. From 2005 onwards, 
additional investigations were executed, allowing for accurate measuring of the deformation 
and of the residual material thickness. In August 2008 a gas bubble jet was detected, released 
in the primary water from one of the deformations. The gas bubble jet appeared to emerge at a 
frequency of once every 10-20 minutes and lasted each time 10-30 seconds. Consequently the 
Dutch regulator granted a temporary license based on three conditions: 

• 

• 

• 

The  HFR  would  only  operate  when  necessary  for  medical  needs  (no  alternatives 
available to ensure supply of medical radioisotopes); 

If  any  water  leakage  from  a  reducer  was  detected,  the  HFR  would  be  immediately 
shut down; 

Works for a final repair of the reducers would start at the latest on March 1st, 2010. 

The  final  repair  required  setting  up  a  dedicated  project  organisation  by  appointing  an 
experienced  project  manager  and  a  project  team  selected  on  the  basis  of  specific  expertise, 
including  external  experts.  The  project  objective  was  to  safely  repair  or  replace  the  two 
reducers  by  gaining  access  through  the  concrete  ceiling  of  the  sub  pile  room,  the  space 
directly under the reactor vessel. 

A major decision in the project was to drain the reactor pool at an early stage of the work for 
the  safety  of  the  workers.  Although  calculations  indicated  that  the  pool  bottom  would  be 
strong  enough  to  contain  the  full  amount  of  water  after  removal  of  the  concrete  from  the 
cavities, the pool was nonetheless drained in order to exclude any risk of flooding in case of 
failure.  This  decision  had  far  reaching  consequences,  since  it  required  the  installation  of 
alternative  radiation  shielding  material  on  the  pool  bottom  (tungsten  alloy  with  a  density  of 
18000 kg/m3). Each step of the repair project was tested on a full scale mock-up. A replica of 
the  entire  section  with  one  of  the  reducers  helped  to  select  the  most  efficient  tools  and 
procedures for the removal of the concrete and its restoration. This training involved workers 
who  later  had  to  perform  the  actual  operation  in  a  limited  space  of  the  sub  pile  room.  The 
material used for replacing the removed concrete was supplied by a Swedish mine. It provided 

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the  same  high  density  concrete  filler  (magnetite)  which  was  used  50  years  ago  for  the 
construction of the HFR.  

The repair started on February 19th, 2010. After removing the reactor internals (fuel elements, 
control  rods,  isotope  production  rigs  and  experimental  facilities)  the  radiation  protection 
equipment  inside  the  reactor  and  shielding  blocks  at  the  bottom  of  the  reactor  pool  were 
installed in order to reduce overall radiation levels within the reactor building. After covering 
the  pool  top  with  the  concrete  slabs  and  draining  the  water,  the  radiation  levels  were 
measured. It became apparent that, although the shielding had reduced the radiation levels by 
the required order of magnitude, at some locations the levels were still too high. Additional 
shielding  had  therefore  to  be  designed,  made  and  fitted,  which  proved  to  be  adequate.  The 
concrete was removed using a hollow-drill-technique. Holes were cut of various lengths in the 
concrete  around  the  reducer  and  the  last  portion  of  concrete  was  removed  by  cutting  with 
hand tools, to avoid damaging the reducers. 

Inspections  carried  out  after  cleaning  the  reducers’  surface  confirmed  that  the  cause  of  the 
deformations  was  galvanic  corrosion,  initiated  by  the  combination  of  steel,  aluminium  and 
moisture. The reactor pool has an aluminium liner which is designed in such a way, that when 
small amounts of pool water are released due to thermal movement, they are transported to a 
drain system through small channels in the concrete. Some of this moisture migrated through 
the  concrete  and  reached  the  reducers.  Inspections  also  showed  that  the  reducer  material 
outside the deformations areas was in a perfect condition. It was therefore decided to proceed 
with  a  local  repair  by  removing  the  damaged  material  (a  larger  hole  at  the  bottom  and  a 
smaller  hole  at  the  top  of  the  reducers).  All  four  holes  were  then  closed  by  welding  pre-
fabricated plates of aluminium. The top plates were welded from within the reducers through 
the  larger  holes  and  then  the  bottom  plates  were  welded  from  outside.  After  welding  was 
completed, the cavities were filled with rebars and concrete. Finally, the reactor systems were 
submitted  to  extensive  commissioning  tests  prior  to  reloading  the  core  and  starting-up  the 
HFR on September 9th, 2010. 

3. 

HFR AS A TOOL FOR RESEARCH 

Network on Neutron Techniques Standardization for Structural Integrity (NeT) 

NeT, the European Network on Neutron Techniques Standardization for Structural Integrity, 
supports progress towards improved performance and safety of European energy production 
systems.  The  JRC,  next  to  its  role  as  manager  of  NeT,  contributes  to  the  scientific  work 
through neutron scattering for residual stress measurement and assessment of thermal material 
ageing effects, using its beam tube facilities at the HFR. About 35 organizations are actively 
participating  in  the  work  of  NeT,  including  eight  organizations  from  the  new  EU  member 
states and three organisations from candidate countries. 

In  2010,  because  of  the  extended  shutdown  of  the  HFR,  the  NeT-related  experimental 
activities  were  restricted  to  small  angle  scattering  studies  of  materials  ageing  processes, 
mainly  in  candidate  materials  for  Generation  IV  nuclear  reactors.  The  NeT  Steering 
Committee  met  twice  in  2010  to  discuss  the  progress  in  the  different  Task  Groups  and  to 
prepare  the  work  of  NeT  for  the  future.  Among  the  Technical  Groups,  TG1  is  setting  a 
benchmark for numerical and experimental work that will continue to be available to nuclear 
engineers  for  testing  their  methods  in  the  future.  With  TG1,  the  first  Task  Group  of  NeT 
coming to an end, a final report is in preparation. 

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3.1. 

Residual stress measurements by neutron diffraction at the HFR 

Despite the extended shut-down of the HFR two series of investigations have been undertaken 
at the residual stress measurement facilities: 

• 

• 

Using the Large Component Neutron Diffraction Facility of Horizontal Beam tube 4 
(HB4),  a  series  of  measurements  was  executed  on  a  slice  cut  from  a  bi-metallic 
piping weld representative of a connection from a nuclear Reactor Pressure Vessel to 
its primary piping. These investigations were addressing measurement methodology 
with the aim to study the possibilities for obtaining improved measurement quality in 
inhomogeneous  materials  such  as  multi-pass  fusion  welds.  The 
instrument 
capabilities  at  HB4,  in  particular  the  ability  to  rotate  the  specimen  during 
measurement (rocking) were exploited in this context. 

At  HB5,  a  series  of  feasibility  measurements  was  performed  on  a  thick  walled 
stainless  steel  piping  weld  with  a  small  inner  diameter,  representative  of  an 
application in the sodium cooled fast reactor. Such measurements were in principle 
found  to  be  feasible,  but  the  possibilities  for  the  necessary  specimen  modifications 
remain to be established. 

3.2.  Microstructure  evolution 

in  high  Cromium  content  oxide  dispersion 

strengthened steels after thermal ageing using Small Angle Neutron Scattering  

Ferritic  Oxide  Dispersion  Strengthened  (ODS)  Steels  are  among  the  candidate  materials  for 
Generation  IV  nuclear  reactors.  While  offering  good  high  temperature  creep  strength, 
radiation resistance and oxidation resistance, they may suffer from hardening embrittlement, 
which  is  attributed  to  the  phase  separation  into  Iron-rich  α  and  Cromium-rich  α'  phases 
forming  during  thermal  ageing  of  the  material  at  intermediate  temperatures.  In  order  to 
promote  their  use,  it  is  necessary  to  understand  the  evolution  of  their  microstructure  during 
thermal  ageing.  The  effects  of  thermal  aging  at  475°C  for  100,  500  and  1000  hours  on  the 
microstructure of the selected ODS steels have been investigated with Small Angle Neutron 
Scattering  (SANS).  The  hardening  during  thermal  ageing  has  been  monitored  by  Vickers 
microhardness measurements. 

All materials showed a strong SANS signal in both the as-recrystallised and the aged states. 
The SANS measurements revealed that thermal treatments result in substantial changes in the 
microstructure at the nanometer size scale. The analysis of the experimental results indicates 
changes  in  the  size  distribution  of  the  scattering  centres.  During  the  heat  treatment  new 
scatterers  with  mean  size  of  2.5-3  nm  were  formed.  Hardness  was  found  to  significantly 
increase as a function of ageing time. The hardening is related mainly to the formation of a 
Cromium-rich α' phase. 

3.3. 

CRYO Experiment: Alpha-Emitters Radiotoxicity Reduction 

The JRC is conducting exploratory research dealing with the supposed reduction of the half-
life  of  alpha  decay  isotopes  embedded  in  a  metallic  matrix  at  cryogenic  temperatures  (4 
Kelvin). Eight discs of pure copper with some Bismuth-209 implanted, in order to obtain by 
transmutation Polonium-210 (210Po) which is a pure alpha emitter with a relatively low half-
life (about 138 days) have been irradiated to study the behaviour of the alpha emitter under 
cryogenic temperature. The possible of the acceleration of alpha decay emitter, embedded in a 
metallic  matrix,  at  cryogenic  temperature  may  ultimately  provide  an  option  for  a  faster 

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reduction of the amount of nuclear waste and therefore, a new irradiation experiment called 
CRYO  has  been  carried  out  in  the  HFR.  The  CRYO  experiment  was  irradiated  for  1  cycle 
(~29 full power days). The irradiation of CRYO took place between December 22nd, 2009 and 
January 15th, 2010. The samples have been shipped to the JRC-ITU Elements for a series of 
PIE.  

4. 

FUEL IRRADIATIONS IN THE HFR 

HELIOS Fuel Experiment: Americium Transmutation 

Americium  is  one  of  the  radioactive  elements  that  contribute  to  a  large  part  of  the 
radiotoxicity  of  spent  fuels.  Transmutation  by  irradiation  in  nuclear  reactors  of  long-lived 
nuclides (like 241Am) is, therefore, an option for the reduction of the mass and radiotoxicity of 
nuclear waste. The Helios experiment, as part of the FP6 EUROTRANS Integrated Project on 
Partitioning  and  Transmutation,  deals  with  irradiation  of  Uranium-free  fuels  containing 
americium.  The  main  objective  of  the  HELIOS  irradiation  is  to  study  in-pile  behaviour  of 
Uranium-free fuel target such as CerCer (Pu, Am, Zr)O2 and Am2Zr2O7+MgO or CerMet (Pu, 
Am)O2 +Mo, in order to gain knowledge on the role of microstructure and temperature on gas 
release and on fuel swelling. During the irradiation of such kind of fuel, a significant amount 
of helium is produced due to the nuclear transmutation of americium. The study of gas release 
is of vital importance to allow better performance of the Uranium \-free fuels. Two different 
approaches are followed to reach early helium release: 

1.  Provide  release  paths  by  creating  open  porosity,  i.e.  release  paths  to  the  plenum  gas. 
Therefore,  in  the  HELIOS  test  matrix  a  composite  target  with  a  MgO  matrix  containing  a 
network of open porosity has also been included. 

2.  Increase  target  temperature  in  order  to  promote  the  release  of  helium  from  the  matrix. 
Americium or americium/plutonium zirconia based solid solutions along with CerMet targets 
have been included in the test matrix to study the effects of the temperature. The role of the 
plutonium  in  association  with  americium  is  to  increase  the  temperature  of  the  target  at  the 
beginning or during irradiation. 

HELIOS,  started  on  April  29,  2009  and  has  been  successfully  concluded  on  February  19, 
2010 after ≈ 242 Equivalent Full Power Days (EFPD). The irradiation was flawless although 
it  had  to  finish  one  cycle  earlier  due  to  the  HFR  repair.  During  irradiation,  cladding 
temperatures were monitored, as well as the central temperatures in two HELIOS pins. After 
irradiation, the experiment underwent dismantling and PIE in the NRG Hot Cell Laboratory. 
In  2010  all  non-destructive  measurements  have  been  performed  on  the  5  pins  consisting  of 
visual inspection, profilometry to measure dimensional changes of the cladding, and gamma 
scanning to look at the behaviour of fission products inside the pins. 

Clear  differences  in  behaviour  were  detected  between  the  cold  and  warm  pins  consisting  of 
CerMet  fuel  with  molybdenum  matrix  fuel.  The  colder  pins  showed  relatively  stable 
behaviour  whereas  in  the  profile  of  the  warm  pins  the  individual  pellets  indicated  a  clear 
swelling  of  the  fuel  pellets  against  the  cladding.  This  result  was  not  predicted  by  the 
modelling since the higher temperatures were expected to release the helium from the matrix 
more easily, reducing swelling at higher temperatures. 

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4.1.  MARIOS Fuel Irradiation: Minor Actinide Recycling 

In the frame of the EURATOM 7th Research Framework Programme (FP7), the 4-year project 
FAIRFUELS  (Fabrication,  Irradiation  and  Reprocessing  of  FUELS  and  targets  for 
transmutation)  aims  at  a  more  efficient  use  of  fissile  material  in  nuclear  reactors  by 
implementing transmutation. Transmutation provides a way to reduce the volume and hazard 
of high level radioactive waste by recycling the most long-lived components. In this way, the 
nuclear  fuel  cycle  can  be  closed  in  a  sustainable  manner.  The  FAIRFUELS  consortium 
consists  of ten European  research institutes, universities and industry. The project started in 
2009  and  is  coordinated  by  NRG.  NRG  and  the  JRC  work  closely  on  the  HFR  irradiations 
scheduled  in  FAIRFUELS.  The  MARIOS  irradiation  programme  deals  with  heterogeneous 
recycling  of  minor  actinides  in  sodium-cooled  fast  reactors  (i.e.  the  minor  actinides  bearing 
blanket concept). Minor actinides, such as americium and curium, are long-lived elements in 
the high level waste, which are currently not recycled. The aim of MARIOS irradiation test is 
to investigate more closely the behaviour of minor actinide targets in an uranium oxide matrix 
carrier.  In  these  targets,  large  amounts  of  helium  are  produced,  which  causes  significant 
damage to the material under irradiation. For the first time americium (241Am) is included in a 
(natural) uranium oxide matrix Am0.15U0.85O1.94 to conduct an experiment in order to study the 
behaviour in terms of helium production and swelling. 

The  design  of  MARIOS  was  finalised  in  2010  after  the  nuclear  analyses  and  the  thermo-
mechanical analyses had been concluded. The fuel pellets, made by CEA in France, arrived in 
April  2010  and  the  capsules,  filled  with  the  fuel  pellets,  were  prepared  in  Petten.  The 
americium-containing samples were squeezed inbetween metal disks to ensure a flat and well-
defined temperature profile. Four separate capsules were prepared. The sample holder for the 
experiment and the assembly of all elements were finalised in December 2010 and received 
the required approvals. The MARIOS irradiation will start the first quarter of 2011 and will 
last for approximately 300 full power days. 

4.2. 

SPHERE Fuel Irradiation: Safer Fuels 

Within  the  FP7  FAIRFUELS  project  the  irradiation  test  SPHERE  was  planned  for  2011. 
SPHERE has been designed to compare conventional pellet-type fuels with so-called Sphere-
Pac fuels. The latter have the advantage of an easier, dust-free fabrication process. Especially 
when  dealing  with  highly  radioactive  minor  actinides,  dust-free  fabrication  processes  are 
essential to reduce the risk of contamination.  

To  assess  the  irradiation  performance  of  Sphere-Pac  fuels  compared  to  conventional  pellet 
fuel,  a  dedicated  SPHERE  irradiation  experiment  will  be  performed.  For  this  purpose, 
americium containing fuel, both pellet-type and Sphere-Pac-type, will be fabricated  at JRC-
ITU in Germany. These fuels will be irradiated at HFR in a dedicated test-facility. It will be 
the  first  irradiation  test  of  this  kind,  as  Minor  Actinides  bearing  Sphere-Pac  fuel  has  never 
been irradiated before. The SPHERE irradiation should start in at the end of 2011 / beginning 
of 2012 and will last for approximately 300 full power days.  

In  2010  the  preliminary  design  for  the  SPHERE  irradiation  test  was  decided,  while  the 
fabrication of the fuel will start in 2011. 

4.3. 

High Temperature Reactor Fuel Irradiations 

Background 

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The  High  Temperature  Reactor  (HTR)  is  a  gas-cooled  graphite  moderated  nuclear  reactor 
concept. The high temperature coolant output, effective fuel use, large R&D experience and 
its robust passive safety concept make it an interesting heat and power generating system. The 
HTR is specifically intended for deployment in an industrial environment. HTR fuel consists 
of TRISO-particles which are uranium oxide kernels coated by a porous graphite buffer layer 
and a pyrocarbon-siliconcarbide-pyrocarbon coating. 10,000-15,000 of these particles (1 mm 
diameter)  are  contained  in  a  graphite  matrix  in  the  form  of  a  6  cm  diameter  fuel  sphere 
(“pebble”)  or  in  the  form  of  finger-thick  cylinders  (“compacts”).  In  the  past,  several 
irradiation tests of low-enriched uranium fuel types in the HFR were carried out to determine 
their  limits  with  respect  to  radioactive  fission  product  release  with  increasing  burn-up 
(enhanced fuel use) and at increased fuel temperature (enhanced efficiency). 

Preparation  of  HFR-EU1  fuel  (high  burn-up  fuel  to  test  fission  gas  release)  has  begun  in 
summer  2002  as  a  collaborative  effort  between  the  European  High  Temperature  Reactor 
Technology  Network  (HTR-TN)  led  by  JRC  and  the  Tsinghua  University  Beijing,  China, 
Institute of Nuclear and New Energy Technology (INET) which runs the HTR-10 test reactor 
and which is involved in the construction of the demonstration plant HTR-PM. Historical and 
newly produced fuel should be irradiated to high burn-up and be compared under irradiation 
and  during  safety  tests  (KÜFA).  KÜFA  tests  verify  fission  product  retention  of  irradiated 
pebbles beyond 1600°C under simulated accident conditions. The combination of irradiation 
and the KÜFA tests is a fuel safety qualification and licensing requirement for new fuel and 
enhances  understanding  of  fission  product  release  and  failure  mechanisms  should  coating 
failure occur. 

The  experiment  was  dismantled  in  the  course  of  2010.  As  part  of  the  FP7  ARCHER 
Collaborative Project extensive post irradiation examination (at NRG Petten and JRC-ITU in 
Karlsruhe) and KÜFA tests will be performed starting in 2011. 

5. 

MATERIALS IRRADIATIONS 

HTR Core Structures Graphites 

Graphite  is  a  suitable  material  to  be  used  as  a  neutron  moderator  and  reflector  in  nuclear 
reactors.  Due  to  its  excellent  high  temperature  performance  graphite  is  used  as  a  structural 
material in the High Temperature Reactor (HTR) design. The qualification of nuclear graphite 
is  a  key  element  as  the  properties  of  graphite  are  changing  significantly  and  non-linearly 
under neutron irradiation. Therefore, the graphite properties need to be obtained at different 
neutron dose levels. NRG plays a leading role in the research of graphite irradiation behaviour 
for HTR technology. A large irradiation programme of various nuclear graphites took place in 
the HFR within the RAPHAEL project. The property curves at two irradiation temperatures, 
750ºC and 950ºC, are produced in four irradiation experiments. Re-irradiation of the graphite 
was a crucial part of the irradiation programme and has shown to be very successful. 

In 2010 the full irradiation programme was completed, including the higher dose irradiation 
experiment  (up  to  23  dpa  and  the  experiment  targeting  13  dpa).  The  dismantling  of  the 
experiments took place at the NRG hot cells. The agreed PIE programme was performed and 
the  results  of  the  graphite  property  changes  were  published  at  the  international  nuclear 
graphite specialist meeting. 

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5.1.  Materials for Fission & Fusion (ExtreMat) 

A  large  number  of  materials  developed  for  use  in  extreme  environments  within  the  various 
subprojects of the ExtreMat Integrated Project are tested on their stability under neutron load 
in the High Flux Reactor. Two irradiation experiments have been designed to be irradiated at 
different temperatures and doses (600°C and 900°C, 5 dpa and 300°C and 550°C, 0.7 dpa) in 
order  to  simulate  the  expected  real  conditions.  The  ExtreMat  I  irradiation  was  performed 
successfully,  while  the  ExtreMat  II  irradiation  suffered  from  a  severe  heat  leak  due  to  the 
swelling of ceramic specimens and was ended prematurely. About half of the specimen drums 
from  ExtreMat  II  can  nonetheless  be  considered  as  having  been  irradiated  at  a  single 
irradiation  temperature.  Specimens  from  those  drums  produce  meaningful  data.  A  large 
variety  of  materials  were  irradiated  including  (doped)  graphites,  SiC  and  C  composites, 
tungsten  alloys,  14Cr  ODS  steels,  Rhenium,  fibre-reinforced  Cu  and  CuCrZr.  The  PIE, 
performed  in  2010,  included  measurements  of  physical  properties  such  as  thermal 
conductivity,  thermal  expansion  and  dynamic  Young’s  modulus  and  mechanical  properties 
such  as  tensile  and  flexural  strength.  The  outcome  of  the  PIE  will  result  in  further 
development of the various materials and finally in the qualification for the real application. 

5.2. 

BLACKSTONE Irradiations: Investigation of AGR Lifetime Extension 

The UK has a fleet of Advanced Gas Cooled Reactors (AGRs) operated by British Energy. In 
order  to  extend  the  lifetime  of  the  AGRs,  graphite  data  at  high  dose  and  weight  loss  is 
required, to allow prediction and assessment of the behaviour of AGR graphite cores beyond 
their  currently  estimated  lifetimes.  Graphite  degradation  is  considered  to  be  one  of  the  key 
issues  that  will  determine  the  remaining  life  of  the  AGRs,  thus  materials  property  data  at 
extended weight loss and dose is essential for continued safe operation and lifetime extension. 
The  BLACKSTONE  irradiations  use  samples  from  AGR  core  graphite  and  subject  them  to 
accelerated  degradation  in  the  high-flux  test  reactor.  The  results  are  designed  to  enable  the 
future condition of the AGR graphite to be predicted with confidence. 

The Blackstone irradiations were completed in 2010 after achieving the required weight loss 
and dose levels. In 2010 the dismantling of both BLACKSTONE experiments took place and 
the PIE started. In the meantime two new irradiations are being prepared to achieve an even 
higher dose and weight loss. The objective of these experiments is to consolidate the database 
produced in phase I and to provide an extensive properties database for HRA/HYA graphite.  

6. 

IRRADIATION FOR FUSION REACTOR TECHNOLOGY 

The  HFR  capabilities  are  very  much  used  for  screening  and  qualifying  fusion  materials, 
components and technology. The HFR contributes to the fusion technology development by 
simulating ITER and DEMO conditions in terms of irradiation temperature and neutron load 
and subsequently providing experimental results on heavily neutron irradiated materials using 
the hot cell laboratories to perform the post-irradiation testing. The main areas of interest are 
the  ITER  vacuum  vessel,  the  development  of  high  heat  flux  components  and  blanket 
structures, along with the development of the reduced activation materials such as 9Cr steels 
and  innovative  materials  such  as  fibre  reinforced  composites.  In  addition,  irradiation 
behaviour  of  ITER  diagnostic  instrumentation  and  the  in-vessel  parts  of  heating  systems, 
which require dedicated assessments and testing programmes, are of great interest. As part of 
the  qualification  of  materials  supporting  the  licensing  of  a  future  reactor,  the  design  of  the 
International  Fusion  Materials  Irradiation  Facility  (IFMIF)  is  under  development.  The  HFR 

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provides  ample  opportunity  to  qualify  specific  materials  for  the  IFMIF  target  section, 
instrumentation  and  mock-ups.  Presentations  on  ITER  and  DEMO  development  and 
qualification activities and the role of HFR in these activities have been delivered at regular 
fusion symposia and conferences. An example of experiments prepared in 2010 is POSITIFE, 
which tests first wall mock-ups under simultaneous thermal cycling and neutron irradiation.  

HIDOBE Experiments: Beryllium for Fusion 

In  the  HIDOBE  (HIgh  DOse  irradiation  of  BEryllium)  experiment,  various  grades  of 
constrained and unconstrained beryllium pebbles and titanium-beryllide samples are irradiated 
at  high  temperatures  in  the  High  Flux  Reactor.  High  dose  irradiation  testing  is  necessary 
because  the  beryllium  irradiation  behaviour  in  fusion  reactors,  in  particular  its  swelling  and 
creep,  has  an  impact  on  e.g.  structural  integrity  of  the  breeder  blanket  and  the  tritium 
inventory  and  it  is  considered  a  critical  issue  for  the  Helium  Cooled  Pebble  bed  (HCPB) 
concept for performance, safety and waste management. 

The HIDOBE experiment consists of two separate rigs (HIDOBE-01 and HIDOBE-02), each 
loaded with 19 grades of beryllium and titanium beryllides, and have target fluences that are 
aimed to produce about 15% and 30% of the DEMO end-of life conditions (3000 and 6000 
appm helium in beryllium, respectively). 

HIDOBE-01 PIE 

HIDOBE-01  has  reached  its  irradiation  target  after  two  years  of  irradiation  and  has  been 
dismantled  in  2008  and  2009.  The  sample  specimens  have  been  retrieved  and  an  extensive 
PIE campaign has started in April 2010 on a selection of materials. 

HIDOBE-02 irradiation 

Irradiation of HIDOBE-02 is ongoing and will be completed at the end of May 2011. It has 
then been irradiated for roughly four years and will have achieved ~90% of its initial target.  

6.1. 

HICU 

HICU  (High  neutron  fluence  Irradiation  of  pebble  staCks  for  fUsion)  has  the  objective  to 
investigate the impact of a fusion reactor neutron spectrum on ceramic breeder pebble beds, as 
used  in  a  HCPB.  The  influence  of  constraining  conditions  on  the  thermo-mechanical 
behaviour on the pebble beds is studied. The irradiation rig features a cadmium shield to tailor 
the neutron spectrum in order to achieve fusion reactor (DEMO) conditions (relevant dpa/He 
ratio). 

Irradiation of HICU started in 2008 and continued in 2009 and 2010. In 2010 four cycles of 
irradiation  (typically  28  days  each)  were  performed,  interrupted  by  the  HFR  BPL  repair.  In 
the fourth cycle in November 2010, rising temperatures in the central, hottest section of HICU 
indicate that HICU’s cadmium shield had been spent (or was burned up). This means that the 
desired  neutron  spectrum  cannot  be  achieved  anymore  in  the  centre  of  the  experiment.  The 
experiment  was  taken  out  of  irradiation  and  additional  calculations  on  the  burn-up  of  the 
shield and the irradiation target of the ceramic breeder samples in the experiment have been 
started. 

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6.2. 

Accelerator Driven System (ADS) Material Development 

An experimental ADS for the transmutation of actinides is under development in Europe. It 
features Liquid Lead Bismuth Eutectic (LBE) as reactor coolant. LBE has a low melting point 
(135  ˚C),  but  has  corrosive  properties  for  structural  materials  and  welds.  In  addition, 
transmutation  of  Bismuth  to  the  high  radiotoxic  210Po  is  a  safety  issue  in  the  design  of  the 
ADS. Materials R&D is needed to test the corrosion behaviour of T91, 316L steels and weld 
specimens during irradiation in contact with LBE, and to examine the deposition of  210Po in 
the irradiation containers and on the specimens after irradiation. 

The irradiation of the two capsules was completed after the first three cycles of 2009. Due to 
the HFR core loading,  the IBIS  experiment was moved from position G7  to  H6  for  the  last 
cycle.  Also  there,  the  target  temperatures  of  300  and  500oC  were  achieved.  IBIS  has  been 
irradiated in the HFR for 250 Full Power Days in total, to an irradiation dose ranging from 1 
(lower temperature) to 2 dpa (high temperature capsule). 

During 2009, the facility for specimen retrieval was commissioned and built in the Hot Cell 
Laboratory.  The  Fuel  Cell  line  of  the  Hot  Cell  Laboratory  was  selected  because  of  the 
presence  of  the  alpha  emitting  radionuclide  210Po.  To  assess  the  risk  of  handling  210Po,  a 
Hazard  and  Operability  (HAZOP)  study  was  performed,  which  formed  the  basis  of  the 
workplan for the retrieval of specimens. The procedure on the retrieval was performed on a 
container loaded with specimens and LBE that was not irradiated. Almost no wetting of the 
LBE on the specimens was observed. A dedicated tensile machine was also installed in this 
cell  to  test  the  irradiated  specimens.  The  tensile  specimens  have  showed  no  effects  of  the 
cyclic heating to 300oC in LBE on the elongation and the ultimate tensile strength. 

7. 

ISOTOPE PRODUCTION 

The year 2010 was unusual for the HFR regarding isotope production. For the first weeks of 
the  year  until  mid-February  the  HFR  continued  to  work  at  maximum  medical  isotope 
production capacity. Isotope production was then stopped during the BPL repair period (at a 
time  of  continuing  international  medical  isotope  shortage)  and  then  restarted  with  a  more 
normal operating pattern in September 2010.  

Until  the  HFR  repair,  the  medical  isotope  production  continued  to  be  given  the  highest 
possible priority. The HFR configuration was set to ensure the absolute maximum production 
levels of key medical isotopes and in particular the production of Molybdenum-99 (99Mo) for 
99mTc  generators.  This  configuration  allowed  as  many  as  11  parallel  99Mo  production 
irradiations  to  be  performed.  At  times,  the  HFR  production  exceeded  the  radiochemical 
processing  capacity  available  within  the  European  supply  network.  It  was  estimated  that 
during this period the HFR produced enough material to allow >50,000 patient scans per day 
to be performed worldwide; this represents around 60% of the global demand. The HFR then 
stopped  isotope  production  between  February  and  September  to  allow  the  essential  BPL 
repair work to be carried out.  

During the year 2010, NRG continued to work closely with the stakeholders in the medical 
isotope  supply  network,  (medical  community,  governments,  the  European  Commission,  the 
OECD/NEA  and  the  IAEA)  coordinating  efforts  to  minimise  the  effects  of  the  supply 
problems  and  to  working  together  towards  establishing  a  future  with  a  greater  certainty  of 
security of supply. A great deal of effort was also given to keeping the isotope customers and 

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in particular the nuclear medicine community informed about the progress of the repair and 
the projected HFR return to service date.  

The  year  2010  once  again  underlined  the  critical  role  performed  by  the  HFR  and  the 
supporting NRG infrastructure in the supply of isotopes for essential medical services. 

8. 

FINANCIAL CONTRIBUTIONS FOR THE EXECUTION OF THE PROGRAMME. 

In  2010,  the  following  financial  contributions  were  received  from  Member  States  for  the 
execution of the programme: 

• 

• 

• 

Belgium: 400,000 € 

France: 300,000 € 

The Netherlands: 8,223,000 € 

It  should  be  noted  that  these  contributions  cover  the  expenses  according  to  Annex  II  to 
Council Decision 2009/410/Euratom. These amounts have been calculated in order to balance 
the forecasted costs of the reactor for the period 2010 taking into account an expected level of 
commercial income. In no case does the Commission cover any operational deficit, including 
potential costs for maintenance or repair. 

From this amount the Commission received 800,000 € as provisions for the Decommissioning 
fund1. 

Other  expenditures  incurred  by  the  JRC  and  paid  from  the  supplementary  research 
programme budget:  

• 

• 

• 

• 

Direct  Personnel  (e.g.  for  HFR  supplementary  research  program  Management): 
290,000 € 

Support HFR (e.g. Legal Advice): 104,000 € 

Utilities (e.g., electricity, water, heating): 345,000 € 

Spent Fuel Management: 935,000 €  

Due to the extended repair period, the net result for NRG for the operation of the HFR in 2010 
was a deficit of 4,692,000 €. 

1 

The yearly contribution to the decommissioning fund has passed from 400,000 €/year to 800,000 €/year 
since 2004 due to a re-evaluation of decommissioning costs. This amount is taken from both the regular 
budget of the supplementary research programme and by the gained interest on the bank account of the 
supplementary  research  programme  (the  amount  of  the  interest  over  2010  was  €  371K  and  therefore, 
€429K was added from the regular supplementary research programme budget of 2010). 

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Glossary and Acronyms 

ADS 

appm 

BPL 

CEA 

Accelerator Driven System 

atomic parts per million 

Bottom Plug Liner 

Commissariat à l’Energie Atomique 

DEMO 

Demonstration Fusion Reactor 

dpa 

EFPD 

EU 

displacements per atom  

Equivalent Full Power Days  

European Union 

Euratom 

European Atomic Energy Community  

EUROTRANS 

European Transmutation 

FAIRFUELS 

Fabrication, Irradiation and Reprocessing of FUELS and target for transmutation 

FIMA 

FP 

Fission per Initial Metal Atoms 

Framework programme 

HAZOP 

Hazard and Operability 

HB 

HCPB 

Horizontal Beam Tube 

Helium Cooled Pebble bed 

HELIOS 

Helium in Oxide Structure 

HFR 

HICU 

High Flux Reactor 

High-fluence Irradiation of breeder Ceramics 

HIDOBE 

High Dose Beryllium Irradiation Rig 

HTR 

IAEA 

IFMIF 

ISI 

ITER 

High Temperature Reactor 

International Atomic Energy Agency 

International Fusion Materials Irradiation Facility 

In-Service Inspection 

International Thermonuclear Experimental Reactor 

JRC-IET 

Joint Research Centre - Institute for Energy and Transport, Petten (NL) 

JRC-ITU 

Joint Research Centre - Institute for Transuranium, Karlsruhe (D) 

LBE 

Lead Bismuth Eutectic 

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LEU 

Low Enriched Uranium 

MARIOS 

Minor Actinides in Sodium-cooled Fast Reactors 

NeT 

NRG 

ODS 

EU Network on Neutron Techniques Standardization for Structural Integrity 

Nuclear Research and consultancy Group 

Oxide Dispersion Strengthened  

OECD/NEA 

Organization for Economic Cooperation and Development / Nuclear Energy Agency 

PIE 

Post Irradiation Examination  

POSITIFE 

Project Pool Side facility Thermally Induced Fatigue 

R&D 

Research and Development 

RAPHAEL 

Reactor for Process Heat and Electricity 

RTD 

SANS 

Research and Technological Development 

Small Angle Neutron Scattering 

TRISO 

TRistructural ISOtopic fuel 

TYCOMO 

TYCO Molybdenum 

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