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Organisation de Coopération et de Développement Économiques
Organisation for Economic Co-operation and Development

NEA/CSNI/R(2011)10

19-Jan-2012___________________________________________________________________________________________English text only
NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSNEA/CSNI/R(2011)10
Unclassified
Studsvik Cladding Integrity Project (SCIP)
Executive Summary

English text only

JT03314631
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Complete document available on OLIS in its original format

NEA/CSNI/R(2011)10

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTThe OECD is a unique forum where the governments of 34 democracies work together to address the economic, socialand environmental challenges of globalisation. The OECD is also at the forefront of efforts to understand and to help
governments respond to new developments and concerns, such as corporate governance, the information economy and thechallenges of an ageing population. The Organisation provides a setting where governments can compare policy
experiences, seek answers to common problems, identify good practice and work to co-ordinate domestic and international
policies.
The OECD member countries are: Australia, Austria, Belgium, Canada, Chile, the Czech Republic, Denmark, Estonia,
Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Israel, Italy, Japan, Luxembourg, Mexico, the Netherlands,
New Zealand, Norway, Poland, Portugal, the Republic of Korea, the Slovak Republic, Slovenia, Spain, Sweden,
Switzerland, Turkey, the United Kingdom and the United States. The European Commission takes part in the work of theOECD.
OECD Publishing disseminates widely the results of the Organisation’s statistics gathering and research on economic,
social and environmental issues, as well as the conventions, guidelines and standards agreed by its members.

This work is published on the responsibility of the OECD Secretary-General.
The opinions expressed and arguments employed herein do not necessarily reflect the official
views of the Organisation or of the governments of its member countries.

NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membership consists of
30 OECD member countries: Australia, Austria, Belgium, Canada, the Czech Republic, Denmark, Finland, France,
Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg, Mexico, the Netherlands, Norway, Poland,
Portugal, the Republic of Korea, the Slovak Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, the United Kingdom
and the United States. The European Commission also takes part in the work of the Agency.
The mission of the NEA is:– to assist its member countries in maintaining and further developing, through international co-operation, thescientific, technological and legal bases required for a safe, environmentally friendly and economical use of
nuclear energy for peaceful purposes, as well as– to provide authoritative assessments and to forge common understandings on key issues, as input to government
decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainable
development.
Specific areas of competence of the NEA include the safety and regulation of nuclear activities, radioactive waste
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law
and liability, and public information.
The NEA Data Bank provides nuclear data and computer program services for participating countries. In these andrelated tasks, the NEA works in close collaboration with the International Atomic Energy Agency in Vienna, with which it
has a Co-operation Agreement, as well as with other international organisations in the nuclear field.

This document and any map included herein are without prejudice to the status of or sovereignty over any territory, to the delimitation ofinternational frontiers and boundaries and to the name of any territory, city or area.
Corrigenda to OECD publications may be found online at: www.oecd.org/publishing/corrigenda.
© OECD 2012
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COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

Within the OECD framework, the NEA Committee on the Safety of Nuclear Installations (CSNI) is
an international committee made of senior scientists and engineers, with broad responsibilities for safety
technology and research programmes, as well as representatives from regulatory authorities. It was set up
in 1973 to develop and co-ordinate the activities of the NEA concerning the technical aspects of the design,
construction and operation of nuclear installations insofar as they affect the safety of such installations.
The committee’s purpose is to foster international co-operation in nuclear safety amongst the NEA
member countries. The CSNI’s main tasks are to exchange technical information and to promotecollaboration between research, development, engineering and regulatory organisations; to review
operating experience and the state of knowledge on selected topics of nuclear safety technology and safetyassessment; to initiate and conduct programmes to overcome discrepancies, develop improvements andresearch consensus on technical issues; and to promote the co-ordination of work that serves to maintaincompetence in nuclear safety matters, including the establishment of joint undertakings.
The clear priority of the committee is on the safety of nuclear installations and the design andconstruction of new reactors and installations. For advanced reactor designs the committee provides aforum for improving safety related knowledge and a vehicle for joint research.
In implementing its programme, the CSNI establishes co-operate mechanisms with the NEA’s
Committee on Nuclear Regulatory Activities (CNRA) which is responsible for the programme of the
Agency concerning the regulation, licensing and inspection of nuclear installations with regard to safety. It
also co-operates with the other NEA’s Standing Committees as well as with key international organisations(e.g., the IAEA) on matters of common interest.

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TABLE OF CONTENTS

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT ..................................... 2
COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS ....................................................... 3
INTRODUCTION AND SUMMARY ........................................................................................................... 7
THE PROJECT ............................................................................................................................................... 9
Task 0: PCMI - Driving force for fuel failures............................................................................................ 9
Results Task 0: PCMI - Driving force for fuel failures ............................................................................. 10
Task 1: PCI, the classic ISCC failure mechanism ..................................................................................... 11
Results Task 1: PCI, the classic ISCC failure mechanism ........................................................................ 11
Tasks 2 and 3: Hydrogen induced failures ................................................................................................ 12
Results Tasks 2 and 3: Hydrogen induced failures ................................................................................... 13

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INTRODUCTION AND SUMMARYThe SCIP programme1 has generated extensive experimental data on irradiated BWR and PWR
claddings by conducting; ramp testing of 11 rodlets, comprehensive nondestructive examinations (NDE),
destructive examinations (DE), mechanical testing and extensive microstructural characterization by meansof SEM and TEM. With respect to PCI, a modified mandrel testing technique to study iodine induced
stress corrosion cracking (ISCC) under well controlled test conditions, was developed and mechanical test
techniques were modified to better simulate the in-reactor behavior.

Displacement (mm)The extensive amount of data has been used to improve the understanding of the fuel rod behavior andthe fuel cladding failure mechanisms: Pellet Cladding Interaction (PCI), Hydride Embrittlement (HE) and
Delayed Hydride Cracking (DHC), fulfilling the overriding objective of the project, which was to improvethe understanding of the dominant failure mechanisms for LWR fuel cladding that can arise during normal
operation or anticipated transients. Furthermore, a DHC model has been developed, critical DHC
parameters were identified, a PCI screening/simulation test method developed and high burnup (BU) fuelrod behavior during ramp tests studied.

DHC mechanical test study as function of temperature for unirradiated
SRA Zircaloy-4 (200 ppm hydrogen)1
Studsvik Cladding Integrity Project (SCIP) was initiated in 2004 as an OECD/NEA project with Studsvik Nuclear as theoperating agent. Since the start 27 signatories from 13 countries have joined the program representing more than 40 organizations(nuclear safety organizations, vendors, utilities and research laboratories)

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THE PROJECT

The overriding objective of the project was to improve the understanding of the dominant failure
mechanisms for LWR fuel cladding that can arise during normal operation or anticipated transients,
focusing on high burnup claddings.
Three failure mechanisms were studied within three corresponding Project Tasks and the driving force
for the failure mechanism was studied separately in Task 0.
0.
1.

2.
3.

PCMI (Pellet Clad Mechanical Interaction) from power transients, ramp tests, presented as stress andstrain data as function of power, time and fuel properties
Pellet-clad interaction (PCI): stress corrosion cracking (SCC) initiated at the cladding inner surface
under the combined effect of the mechanical loading and chemical environment caused by anincrease in the fuel pellet temperature during a power increase.
Hydrogen embrittlement (HE): time independent fracture of existing hydrides.
Delayed hydride cracking (DHC): time dependent crack initiation and propagation through fractureof hydrides that form ahead of the crack tip.

Task 0: PCMI - Driving force for fuel failures
The objective was to obtain a relationship between cladding residual strain and Linear Heat Rate
(LHR) as a function of burnup level and hold time, based on experimental data. The obtained data can beused to get a rough estimation of maximum stresses and strains in the short ramp tests and residual strain
evolvement as a function of hold time.
In total 11 ramp rodlets were studied, including PWR and BWR materials and burnup levels
between 40 and 75 MWd/kgU. Different ramp types were used for evaluation of strain changes during theholding time. The local diametral residual strain along the cladding was determined by profilometry. The
measured residual strain was plotted as functions of local power. Extensive PIE was performed on thetested rods before and after the ramp tests.
These effects and other fuel parameters were further explored by modeling performed by several
SCIP members. Two separate modeling workshops were arranged, the first was organized by Studsvik
and the second by CIEMAT (Spain).
Three different irradiated cladding materials and one unirradiated were tested at temperatures in therange of 310 ºC to 400 ºC to study the effect of irradiation and strain rate on the hardening relaxation
properties and to obtain a relationship between stress, strain and time for irradiated claddings with highburnup levels,. The results were, among others, used as input data for tuning of the different fuelperformance codes used. In Hardening-Relaxation (HRX) testing a cladding sample is internallypressurized, resulting in a diameter increase during the so called hardening phase. After a pre-set diameterchange (strain) is obtained, it is kept constant by reducing the pressure during the relaxation phase.

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Results Task 0: PCMI - Driving force for fuel failures
The residual strain results from the ramp tests showed that for low to medium BU levels, there was no
measurable residual strain in the BWR fuel tested under typical BWR conditions. On the other hand, athigh BU levels, the residual strain increased rapidly with LHR, indicating larger pellet expansion and thus
likely larger cladding hoop stresses.

2,5

KKL-1 (BU 63; 40 min)
KKL-2 (BU 65; 30 sec)
KKL-4 (BU 40; 5 sec)

2
1,5

Residual strain [%]Residual strain [%]

2,5

1
0,5

KKL-3 (BU 56; 7*1+12 h)
O2 (BU 55; 30 sec)

2
1,5
1
0,5
0

0
0

10

20

30

40

0

50

20

30

40

50

40

50

Z-4 (BU 75; 6+6 h)
Z-2 (BU 76; 6+6 h)
Z-3 (BU 76; 5 sec)

2,5

M5-H2 (BU 68; 12 h)

2

Residual strain [%]Residual strain [%]

2,5

10

Local LHR [kW/m]

Local LHR [kW/m]

M5-H1 (BU 67; 5 sec)

1,5
1
0,5
0

2
1,5
1
0,5
0

0

10

20

30

40

50

Local LHR [kW/m]

0

10

20

30

Local LHR [kW/m]

Residual strain vs. local test rod power from the ramp tests, BWR
test results at the top and PWR test results at the bottom.
A comparison of short versus long hold times showed that the residual strain increased slightly withholding time in BWR claddings. For the PWR claddings studied, all the rods had high BU. A comparison
between short and long hold times in these rods showed that the residual strain increased markedly when
the hold time was increased.
The post-irradiation examination (PIE) showed that long hold times lead to higher center porosity andmore sintering between the pellets across the interfaces. For the dish filling there is a general trend of lower
remaining dish values towards the high power part of the rodlets, reflecting stronger pellet swelling at thehigher LHR. The measurements also indicate a clear correlation between the compressed gap, the residualstrain and the LHR, while the relocated gap doesn´t have any power dependence.
The Hardening-Relaxation mechanical testing showed that relaxation increases with temperature andthe unirradiated material has the largest relaxation. Those data was used in the modeling and also used toverify constitutive equations.

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Relaxation as function of strain rate

The average hydrogen concentration in the 11 rodlets ramp tested was measured and it is clear that the
average hydrogen concentration alone is not a sufficient parameter to describe the sensitivity to hydrogen
induced failure during a power ramp.
The modeling efforts were attracting large attention from the participants, resulting in several
voluntary contributions and the results showed good agreement overall. There are areas that may be looked
into when looking for potential improvements, such as the gaseous swelling in models and the need for
actual material data was noted, both these parameters are important when discussing the stress and strain
histories during the ramp tests.
Task 1: PCI, the classic ISCC failure mechanism
The objectives of Subtask 1 were to establish a reliable out-of-pile method for testing the sensitivityof irradiated claddings to PCI, determine and, if possible, quantify the critical parameters for PCI failures.
An expanding mandrel type testing test technique that can assess the sensitivity to iodine induced
PCI (ISCC) of irradiated cladding has been developed. The design was based on a seminar that was held at
the beginning of SCIP to present the past experience using different test methods.
The pellet is simulated by a ceramic insert and the iodine is introduced into the test chamber by theusage of a carrier gas, allowing a controlled partial pressure of iodine. The work focused primarily onknowledge transfer from previous PCI studies and building and qualification of a mandrel type test
technique to study ISCC of irradiated fuel cladding. The equipment was used to perform parameter studies
on irradiated fuel cladding and to compare PIE of ramp tested rods that failed by PCI and mandrel tests ofidentical fuel cladding In addition to the subtasks included in the SCIP programme, CEA has performed
numerical simulations of the mandrel tests as a voluntary contribution.
Results Task 1: PCI, the classic ISCC failure mechanism
2

Total diametral strain to failure, %The results from the tests
technique showed that ISCC
simulated successfully in cold
irradiated Zry-2 without liner as
and irradiated ZIRLO. The
tests confirm the earlier findings
pressure of iodine works as an
for ISCC and an increase invalue does not influence anymore

Irradiated non-liner RX Zircaloy-2
1,5

1

0,5

0

Effect of iodine partial pressure,showing a trend change at 60 Pa
0

50

100

150

I2 partial pressure in the system, Pa

11

200

250

using
the
mandrelfailurescanbeworked unirradiated andwell as in unirradiatedresults of the parameter
that
60 Pa
partialinfinite supply of iodinepressure beyond thisthe ISCC behavior.

NEA/CSNI/R(2011)10

Characteristic fracture surface with“fluting”

The mandrel tested samples have low strain to failure as well as typical ISCC features on the fracturesurface such as fluting. This is consistent with known characteristics of PCI failures and indicates that the
mandrel test method developed in SCIP seems to be representative of PCI cladding failures. A generaltrend of increased strain to failure occurs as the test temperature is increased from 320 to 400 °C.
The new expanding mandrel technique is a promising method presenting measurements on-line,
which makes attempts to quantify stress/strain time histories possible and it will be used also in SCIP II.
Tasks 2 and 3: Hydrogen induced failures
The main objective of tasks 2 and 3 was to identify and quantify the critical conditions under whichhydrogen induced failures, Hydrogen Embrittlement (HE) and Delayed Hydride Cracking (DHC), may
occur in BWR and PWR claddings during normal operation and anticipated transients. A secondary
objective was to develop a combined approach to study hydrogen induced failures that includes mechanicaltesting, ramp testing and modeling. A long-term goal is to improve simulations of hydrogen induced
failures during power ramps, to such an extent that the sensitivity to this type of failures can be predictedbased on operational conditions and material properties and thus can be modeled.
The objective of the HE studies was to establish an understanding of hydrogen assisted crack
initiation in zirconium based cladding materials and to model/establish critical hydride parameters.
The objective of the DHC studies was to establish an understanding of hydrogen assisted crack
propagation in zirconium based cladding materials and to establish the critical parameters under which
DHC can be the failure mechanism. A second objective was to develop a model of the DHC failure
mechanism and study/establish some critical input parameters, such as hydrogen diffusion coefficient andstress fields in clad tubes with hydrides.
A workshop was held to wrap up tasks 2 and 3 of the SCIP programme by summarizing key findings
and compare the results with other research programs.
The role of hydrogen induced failure mechanisms on the cladding integrity was studied by threedifferent methods; a) in-pile ramp testing, b) out-of-pile mechanical testing and c) modeling of hydrogen
induced failures. These methods complement each other and were accordingly all used. While the actual
operating conditions in the reactor are simulated in the ramp tests, mechanical testing allows for more
controlled parametric studies. Although mechanical testing in combination with ramp testing to some

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extent can predict in-pile failures they cannot simulate all scenarios. In order to gain a more comprehensive
foundation for predicting cladding failures, modeling of the hydrogen induced failure mechanisms wasused as a complement to mechanical testing.
The subtasks within these two tasks can be divided into the following areas: Parametric studies on
crack initiation in hydrides (HE). Parametric studies on crack propagation by DHC. Development of new
test techniques. Modeling of hydrogen induced failures. PIE of selected rodlets ramp tested within Task 0.
Basic studies to quantify less known parameters. Knowledge transfer of previous studies on H induced
failures (extensive workshop with world-leading experts)
Results Tasks 2 and 3: Hydrogen induced failures
It was shown that in order for HE to occur as a primary failure under operating conditions, a very highdensity of hydrides is required. These hydride density levels may be reached locally at the outer surface ofhigh burnup rods. In these rods incipient cracks may form in the dense hydride rim at the outer surface atstresses well below the yield stress of the cladding alloy, even at operating temperatures.
A PCMI caused failure may thus be triggered by hydride rim cracking, as a high burnup rod hydride
layer may crack, leading to local high stress at the tip of the crack, causing propagation of the crack
through the cladding.

300 ºC, 330 MPa left, 410 MPa right (giving
incipient in the hydride rim)
The critical parameters for DHC were quantified and the results confirm that there is an upper
temperature limit above which DHC does not operate in the zirconium alloys tested. This is most probably
one important contributing factor to why DHC failures have not been documented in PWR reactors.

Tests performed on medium to high burnup Zry-2 claddings showed that DHC can operate in thesealloys at BWR temperatures. Parametric studies showed that DHC may operate at hydrogen concentrations
down to the solubility limit of hydride formation. However, the studies also indicate that in order for DHC
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NEA/CSNI/R(2011)10
to initiate as a primary failure in claddings with low hydrogen concentrations, high hoop stresses have tobe applied.
The mechanical parameter studies confirm that different loading (or power) history triggers different
H induced failures; DHC is suppressed during short ramp tests, but can occur at quite low load levels if thehold-time in a loading step is sufficient long (30-60 min). Incipient crack formation by HE occurs
instantaneouslyatlow load levels.
However,stresses
well above the yieldstress are required forthe incipient cracks
to lead to throughwall cladding failureby shear instability,

The out-of-pile mechanical testing
showed very good agreement with the in pile
testing. In identical fuel rods the same failure
mechanisms were triggered under similar test
conditions.
At BWR operating temperatures andbelow, it was observed that several hydrogenenhanced failure mechanisms can coexist. A
model has been developed in order tocombine different types of hydrogen induced
failures. The model was used to simulate H
induced failures of samples tested by meansof Ring Tensile Testing and Pin Loading
Testing. The results show fairly good
agreement between the model and themechanical testing.
SCIP finished in July 2009 and all SCIP members participate also in the ongoing continuation, theSCIP II project, focusing on the source of the PCMI driving force, the pellet and its behavior, including
doped pellets. The SCIP II budget (slightly larger than SCIP) is about 7.500.000 Euro.

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