Draft Regulatory Guide DG-1263

Draft Reg Guide DG-1263.pdf

10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities

Draft Regulatory Guide DG-1263

OMB: 3150-0011

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U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REGULATORY RESEARCH

March 2014
Division 1

DRAFT REGULATORY GUIDE
Contact: M. Flanagan
(301) 251-7547

DRAFT REGULATORY GUIDE DG-1263
(Proposed New Regulatory Guide)

ESTABLISHING ANALYTICAL LIMITS FOR ZIRCONIUMBASED ALLOY CLADDING
A. INTRODUCTION
Title 10, Section 50.46c, of the Code of Federal Regulations (10 CFR 50.46c) (Ref. 1), calls for
the establishment of analytical limits on peak cladding temperature and integral time at temperature that
correspond to the measured ductile-to-brittle transition for the zirconium-alloy cladding material. This
guide defines an acceptable analytical limit on peak cladding temperature and integral time at temperature
for the zirconium-alloy cladding materials tested in the U.S. Nuclear Regulatory Commission’s (NRC’s)
loss-of-coolant accident (LOCA) research program. This analytical limit is based on the data obtained in
the NRC’s LOCA research program.
The database developed in NRC’s LOCA research program can be supplemented in order to
establish an analytical limit, as called for in 10 CFR 50.46c. Draft Regulatory Guide (DG) -1262,
“Testing for Postquench Ductility” (Ref. 2), provides an experimental technique acceptable to the NRC
for measuring the ductile-to-brittle transition for zirconium-alloy cladding material through ring
compression tests (RCT). This guide describes a method to demonstrate comparable performance with
the established database in order to establish the analytical limit provided in this guide for a particular
cladding alloy not tested in the NRC’s LOCA research program. This guide also describes methods for
establishing analytical limits for zirconium –alloy cladding materials not tested in NRC’s LOCA research
program, or establishing limits for zirconium-alloy cladding materials at conditions other than those used
in NRC’s LOCA research program.
In 10 CFR 50.46c, the NRC calls for measurement of the onset of breakaway oxidation for a
zirconium-alloy cladding material based on an acceptable experimental technique, evaluation of the
measurement relative to ECCS performance, and annual retesting and reporting of values measured
(Ref. 1). DG-1261, “Conducting Periodic Testing for Breakaway Oxidation Behavior” (Ref. 3), provides
an experimental technique acceptable to the NRC for measuring the onset of breakaway oxidation for
zirconium-alloy cladding materials. DG-1261 also describes an acceptable method of meeting the annual
retesting and reporting requirements in 10 CFR 50.46c. This guide describes a methodology for
This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a regulatory
position in this area. It has not received final staff review or approval and does not represent an official NRC final staff position.
Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory
analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may
be submitted to the Rules, Announcements, and Directives Branch, Office of Administration, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001; submitted through the NRC’s interactive rulemaking Web page at
http://www.nrc.gov; or faxed to (301) 492-3446. Copies of comments received may be examined at the NRC’s Public Document
Room, 11555 Rockville Pike, Rockville, MD. Comments will be most helpful if received by June 9, 2014
Electronic copies of this draft regulatory guide are available through the NRC’s interactive rulemaking Web page (see above); the
NRC’s public Web site under Draft Regulatory Guides in the Regulatory Guides document collection of the Library at
http://www.nrc.gov/reading-rm/doc-collections/; and the NRC’s Agencywide Documents Access and Management System
(ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML12284A323. The regulatory analysis may be
found in ADAMS under Accession No. ML12283A188.

establishing a specified and acceptable limit for the total accumulated time that the cladding may remain
above a temperature at which the zirconium alloy has been shown to be susceptible to breakaway
oxidation.
The NRC issues regulatory guides to describe to the public methods that the staff considers
acceptable for use in implementing specific parts of the agency’s regulations, to explain techniques that
the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to
applicants. Regulatory guides are not substitutes for regulations and compliance with them is not
required.
This regulatory guide contains information collection requirements covered by 10 CFR Part 50
that the Office of Management and Budget (OMB) approved under OMB control number 3150-0011.
The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information
collection request or requirement unless the requesting document displays a currently valid OMB control
number. This regulatory guide is a rule as designated in the Congressional Review Act (5 U.S.C. 801–
808). However, OMB has not found it to be a major rule as designated in the Congressional Review Act.

DG 1263; Page 2 of 21

Page
A. INTRODUCTION .................................................................................................................................. 1
B. DISCUSSION ......................................................................................................................................... 4
Background ................................................................................................................................................... 4
Summary of the NRC’s LOCA Research Program ...................................................................................... 4
Existing Embrittlement Database .................................................................................................... 4
An Acceptable Analytical Limit on Peak Cladding Temperature and Time at Elevated
Temperature ..................................................................................................................................... 6
Methodology for Demonstrating Consistency with the Existing Database for New Cladding Alloys ........ 7
Methodology for Establishing a Zirconium-Alloy-Specific Limit .............................................................. 9
Methodology for Establishing Analytical Limits at Peak Oxidation Temperatures Less than 1,204 °C
(2,200 °F) .................................................................................................................................................. 10
Methodology for Establishing Analyticlical Limits for Breakaway Oxidation ......................................... 11
Applying Analytical Limits ...................................................................................................................... 12
Qualification of Hydrogen Pickup Models .................................................................................. 12
Accounting for Uncertainty and Variability in Hydrogen Content .............................................. 12
Postquench Ductility Analytical Limts ......................................................................................... 12
Application in the Rupture Region .............................................................................................. 13
Accounting for Double-Sided Oxidation Due to the Fuel-Cladding Bond Layer ........................ 13
Breakaway Oxidation Analytical Limtis ...................................................................................... 14
C. REGULATORY POSITION................................................................................................................. 14
D. IMPLEMENTATION ........................................................................................................................... 17
GLOSSARY ............................................................................................................................................... 19
REFERENCES ........................................................................................................................................... 20
APPENDIX A: Relationship between Offset Strain and Permanent Strain............................................. A-1
APPENDIX B: Overview of Acceptable Test Matrices .......................................................................... B-1

DG 1263; Page 3 of 21

B. DISCUSSION
Background
In 1996, the NRC initiated a fuel-cladding research program intended to investigate the behavior
of high-exposure fuel cladding under accident conditions. This program included an extensive LOCA
research and testing program at Argonne National Laboratory (ANL), as well as jointly funded programs
at the Kurchatov Institute (Ref. 4) and the Halden Reactor Project (Ref. 5), to develop the body of
technical information needed to evaluate LOCA regulations for high-exposure fuel. The research findings
were summarized in Research Information Letter 0801, “Technical Basis for Revision of Embrittlement
Criteria in 10 CFR 50.46,” dated May 30, 2008 (Ref. 6). Most of the detailed experimental results from
the program at ANL appear in NUREG/CR-6967, “Cladding Embrittlement during Postulated Loss-ofCoolant Accidents,” issued July 2008 (Ref. 7).
The research results revealed that hydrogen, which is absorbed into the cladding during the
burnup-related corrosion process under normal operation, has a significant influence on embrittlement
during a hypothetical LOCA. When that cladding is exposed to high-temperature LOCA conditions, the
elevated hydrogen levels increase the solubility of oxygen in the beta phase and the rate of diffusion of
oxygen into the beta phase. Thus, for cladding exposed to high-temperature LOCA conditions,
embrittlement can occur for times corresponding to less than 17% oxidation in corroded cladding with
significant hydrogen pickup. The research results also revealed that an embrittlement mechanism referred
to as “breakaway oxidation” may occur during prolonged exposure to elevated cladding temperature
during a LOCA.

Summary of the NRC’s LOCA Research Program
Existing Embrittlement Database
The majority of the cladding embrittlement experimental results from the NRC’s LOCA research
program are summarized in NUREG/CR-6967 (Ref. 7). Since the publication of NUREG/CR-6967 in
2008, additional testing was conducted, focusing on cladding materials with hydrogen contents in the
200- to 350-weight parts per million (wppm) range (Refs. 8–9, 11). Additional oxidation and postquench
ductility (PQD) tests were conducted with cladding samples sectioned from high-burnup ZIRLOTM fuel
rods. The two defueled segments used to prepare samples had 25–30 micrometers corrosion-layer
thickness and 300–340 wppm of hydrogen in the cladding metal before oxidation (Ref. 8). Also, the
ductility data for an oxidation sample with ≈600-wppm hydrogen was reassessed (Ref. 9). In addition,
since the publication of NUREG-6967, oxidation and PQD tests were conducted with prehydrided
cladding samples containing 200–300 wppm of hydrogen (Ref. 9).
The tests that were conducted after the publication of NUREG/CR-6967 were combined with the
data reported in NUREG/CR-6967 to generate a more robust and informed description of cladding
embrittlement as a function of hydrogen content.
Before combining the new data with the data reported in NUREG/CR-6967, two refinements
were made in data assessment. The first refinement was to establish and verify the following ductility
criteria: average permanent strain ≥1.0% or, if permanent strain cannot be measured, the average ring
compression test (RCT) offset strain ≥1.41% + 0.1082 Cathcart-Pawel equivalent cladding reacted
(CP-ECR) (Ref. 10). Rounded to the nearest tenth of a percent, this correlation represents the one-sigma
upper bound of offset strain values from 65 RCT data sets with 1.0 to 2.3% permanent strain. (DG-1262
and Appendix A to this regulatory guide provide discussion and details about a ductility criterion based

DG 1263; Page 3 of 21

on RCT offset strain.) The second refinement was to develop and use a new methodology to determine
the pretest hydrogen content in the cladding metal for corroded cladding (Ref. 8).
Ductility and hydrogen data presented in NUREG/CR-6967 were reassessed to determine
embrittlement oxidation levels versus hydrogen content for prehydrided and high-burnup cladding. When
the tests that were conducted after the publication of NUREG/CR-6967 were combined with the data
reported in NUREG/CR-6967, and the refinements in hydrogen content and the relationship between
offset and permanent strain were made, the resulting behavior description of cladding embrittlement as a
function of hydrogen content could be depicted as shown in Figure 1.
For modern as-fabricated cladding (Zry-2, Zry-4, ZIRLOTM, and M5), embrittlement thresholds
cluster at 19–20% CP-ECR, as compared to 16% CP-ECR for older Zry-4 cladding. However, this
improvement is negated with hydrogen pickup as low as 100 wppm. A bilinear function for CP-ECR
versus hydrogen content was used to fit the embrittlement data for prehydrided and high-burnup cladding.
The embrittlement rate is steep for cladding with ≤400-wppm hydrogen. For higher hydrogen content,
the embrittlement rate is more gradual because embrittlement occurs during the heating ramp at
≤1,180 °C (≤2,156 °F). High-burnup ZIRLOTM with 600-wppm hydrogen is highly ductile at
4% CP-ECR, but the peak oxidation temperature was only 1,130 °C (2,066 °F). Embrittlement is highly
sensitive to both hydrogen content and peak oxidation temperature.

Figure 1. Ductile-to-brittle transition oxidation level (CP-ECR) as a function of pretest hydrogen
content in cladding metal for as-fabricated, prehydrided, and high-burnup cladding
materials. Samples were oxidized at ≤1,200 °C ±10 °C and quenched at 800 °C. For highburnup cladding with about 550-wppm hydrogen, embrittlement occurred during the
heating ramp at 1,160–1,180 °C peak oxidation temperatures (Ref. 8).

DG 1263; Page 4 of 21

An Acceptable Analytical Limit on Peak Cladding Temperature and Integral Time at Temperature
In 10 CFR 50.46c, the NRC calls for the establishing of analytical limits on peak cladding
temperature and integral time at temperature, which correspond to the measured ductile-to-brittle
transition for the zirconium-alloy cladding material (Ref. 1). The ductile-to-brittle threshold defined in
Figure 2 is an acceptable analytical limit on integral time at temperature as calculated in local oxidation
calculations using the Cathcart-Pawel (CP) correlation (Ref. 11). This analytical limit is acceptable for
the zirconium-alloy cladding materials tested in the NRC’s LOCA research program, which were Zry-2,
Zry-4, ZIRLOTM, and M5. This analytical limit is based on the data obtained in the NRC’s LOCA
research program. Since PQD tests above 400-wppm hydrogen were conducted at a peak oxidation
temperature below 1,204 °C (2,200 °F), a separate PCT analytical limit must be defined that is consistent
with test temperature. A limit on peak cladding temperature of 1,204 °C (2,200 °F) below 400-wppm
cladding hydrogen content and 1,121 °C (2,050 °F) at or above 400-wppm cladding hydrogen content is
acceptable.

Embrittlement Oxidation Limit (% ECR)

Demonstrating that ECCS performance is such that local oxidation and peak cladding temperature
are calculated below the analytical limits defined in Figure 2 is acceptable to demonstrate compliance
with 10 CFR 50.46c.
20
18

PCT
≤ 2200°F

16
14
12

PCT
≤ 2050°F

10
8
6
4
2
0
0

100

200

300

400

500

600

700

800

Pre-Transient Hydrogen Content (wppm)

Figure 2. An acceptable analytical limit on peak cladding temperature and integral time at
temperature (as calculated in local oxidation calculations using the CP correlation (Ref. 11))
For zirconium-alloy cladding materials not tested in the NRC’s LOCA research program, a
demonstration of comparable performance with the database established in the NRC’s LOCA research
program would be necessary in order to establish the analytical limit provided in this guide as the limit for
that alloy. Draft Regulatory Guide (DG)-1262, “Testing for Postquench Ductility” (Ref. 2), provides an
experimental technique acceptable to the NRC for measuring the ductile-to-brittle transition for
zirconium-alloy cladding material through ring compression tests (RCT). This guide describes a method
to demonstrate comparable performance with the established database in order to establish the analytical
limit provided in this guide for a particular cladding alloy not tested in the NRC’s LOCA research
program.

DG 1263; Page 5 of 21

The database established in the NRC’s LOCA research program and the resulting analytical limit
described in this regulatory guide are intended to provide a best-estimate limit for the ductile-to-brittle
transition for zirconium alloys. The analytical limit described in this guide is applicable to Zircaloy-2
(Zry-2), Zircaloy-4 (Zry-4), ZIRLOTM, and M5. In some instances, a zirconium-alloy cladding material
may experience the transition from ductile to brittle behavior at a higher level of oxidation than the
established database. This regulatory guide also describes a methodology to establish a zirconium-alloyspecific limit other than the limit provided in this guide.
The database established in the NRC’s LOCA research program and the resulting analytical limit
described in this regulatory guide are intended to bound emergency core cooling system (ECCS)
performance. In the test program, experiments were conducted at maximum oxidation temperatures
permitted by the criteria in 10 CFR 50.46. Some ECCSs may perform such that the maximum oxidation
temperature is significantly below 1,204 degrees Celsius (°C) (2,200 degrees Fahrenheit (°F)). Oxidation
at lower temperatures has been shown to increase the allowable calculated oxidation before
embrittlement. Therefore, conducting tests at lower peak temperatures may provide additional margin for
some zirconium-alloy cladding materials. This regulatory guide describes a methodology to establish
analytical limits at peak oxidation temperatures less than 1,204 °C (2,200 °F).

Methodology for Demonstrating Consistency with the Existing Database for New Cladding
Alloys
For zirconium-alloy cladding materials not tested in the NRC’s LOCA research program, a
demonstration of comparable performance with the established database is necessary. The objective of
PQD testing to demonstrate consistency with the analytical limit provided in Figure 2 of this regulatory
guide is to confirm that the transition to brittle behavior does not take place at a lower equivalent cladding
reacted (ECR) than the provided limit. A range of material conditions can serve to provide a
characterization of PQD behavior through the spectrum of conditions expected during operation and
during a transient. Repeat testing can be used to address expected variability in oxidation behavior. The
methodology outlined in this regulatory guide includes testing of as-received, prehydrided, and irradiated
material. The methodology outlined in this regulatory guide uses the experimental procedure in DG-1262
to generate RCT data to demonstrate consistency with the analytical limit in Figure 2 of this regulatory
guide.
As-received cladding material may be used to characterize an alloy’s oxidation embrittlement
behavior in the as-received condition. The analytical limit provided in Figure 2 of this regulatory guide
can be used to reduce the extent of testing by focusing on specific oxidation levels. One approach would
be to conduct oxidation and quench testing at the transition ECR defined in Figure 2, an ECR 1% above,
and an ECR 1% below this limit. Following the guidance of DG-1262, each oxidation and quench sample
would be segmented into three RCT samples. The average of these three RCT samples would be
compared to the ductility criterion defined in terms of ≥1.0% permanent strain or an offset strain ductility
criterion presented in Appendix A to this regulatory guide. This would generate nine RCT results for
as-received cladding material.
Prehydrided cladding material may be used to characterize the effect of hydrogen on an alloy’s
oxidation embrittlement behavior. The entire range of a cladding material’s anticipated hydrogen level
should be characterized. To characterize the range of a cladding material’s anticipated hydrogen content,
an acceptable approach would be to determine the ductile-to-brittle transition for prehydrided material in
increments not more than every 100 wppm of hydrogen. The analytical limit provided in Figure 2 of this
regulatory guide can be used to reduce the extent of testing by focusing on specific oxidation levels at
each hydrogen level. One approach would be to conduct oxidation and quench testing at the transition
ECR defined in Figure 2 for a given hydrogen content, an ECR 1% above, and an ECR 1% below this
DG 1263; Page 6 of 21

limit. Following the guidance of DG-1262, each oxidation and quench sample would be segmented into
three RCT samples. The average of these three RCT samples would be compared to the ductility criterion
defined in terms of ≥1.0% permanent strain or an offset strain ductility criterion presented in Appendix A.
This would generate nine RCT results at each hydrogen level.
Irradiated cladding material can be used to demonstrate that a cladding alloy’s embrittlement
behavior is accurately characterized by using prehydrided material. To demonstrate this, an acceptable
approach would be to determine the ductile-to-brittle transition for irradiated material with hydrogen
contents within 50 wppm of the anticipated maximum hydrogen content and within 50 wppm of half of
the anticipated maximum hydrogen content. The analytical limit provided in Figure 2 of this regulatory
guide can be used to reduce the extent of testing by focusing on specific oxidation levels relevant for the
hydrogen content of the irradiated material. One approach would be to conduct oxidation and quench
testing at the transition ECR defined in Figure 2 for the irradiated material’s hydrogen content, an ECR
1% above, and an ECR 1% below this limit. Following the guidance of DG-1262, each oxidation and
quench sample would be segmented into three RCT samples. The average of these three RCT samples
would be compared to the ductility criterion defined in terms of ≥1.0% permanent strain or an offset strain
ductility criterion presented in Appendix A. This would generate nine RCT results at each hydrogen
level, and a total of 18 RCT results for irradiated material.
Appendix B to this regulatory guide presents a high-level overview of an acceptable test matrix.
The test matrix overview is intended to provide a clear picture of the range of material and test conditions
that could be used to demonstrate comparable embrittlement behavior with the analytical limit in
Figure 2. It is intended to complement the test matrix guidance in DG-1262.
To demonstrate comparable performance with the existing database and adopt the analytical
limits provided in this guide for a new fuel design, the applicant would submit experimental results as
part of the documentation supporting the NRC staff’s review and approval of the new fuel design
(i.e., license amendment request or vendor topical report). The applicant would provide details of the
experimental technique (unless the experiments were conducted in accordance with DG-1262) and the
results of experiments conducted with as-fabricated, prehydrided, and irradiated cladding material.
Provided that the experimental results for the new fuel design measured the transition from ductile to
brittle behavior to be no lower than the analytical limit defined in Figure 2,1 an acceptable method to
demonstrate that licensees meet the requirements of 10 CFR 50.46c is demonstrating that ECCS
performance is such that local oxidation is calculated below the analytical limit defined in Figure 2.

Methodology for Establishing a Zirconium-Alloy-Specific Limit
The existing database and resulting analytical limit described in this regulatory guide are intended
to provide a best-estimate limit for the ductile-to-brittle transition for zirconium alloys. The analytical
limit described in this guide is applicable to Zry-4, Zry-2, ZIRLOTM, and M5. In some instances, a
zirconium-alloy cladding material may experience the transition from ductile to brittle behavior at a
higher level of oxidation than the established database.
The objective of PQD testing to establish an alloy-specific limit is to characterize a cladding
alloy’s embrittlement behavior through the entire spectrum of conditions expected during operation. A
diverse matrix of material conditions can provide a complete characterization, and repeat testing can be
used to address expected variability in oxidation behavior. The methodology outlined in this regulatory

For accurate comparison to the research data, local oxidation calculations must be performed using the CP correlation.

DG 1263; Page 7 of 21

guide includes testing of as-received, prehydrided, and irradiated material. The methodology uses the
experimental procedure in DG-1262 to generate RCT data to establish a zirconium-alloy-specific limit.
As-received cladding material may be used to characterize an alloy’s oxidation embrittlement
behavior in the as-received condition. The methodology outlined in this regulatory guide for establishing
a zirconium-alloy-specific limit other than the analytical limit provided in Figure 2 includes more repeat
testing than was outlined for demonstrating consistency with the established database. One approach
would be to conduct oxidation and quench testing at a wide range of ECRs to scope out a zirconium-alloy
cladding material’s oxidation behavior in the as-received condition. Testing would then focus on the
ECR range between ductile and brittle results and include three repeat oxidation and quench tests at two
ECR levels (a total of six oxidation and quench tests). Following the guidance of DG-1262, each
oxidation and quench sample would be segmented into three RCT samples. The average of these three
RCT samples would be compared to the ductility criterion defined in terms of ≥1.0% permanent strain or
the offset strain ductility criterion presented in Appendix A.
Prehydrided cladding material may be used to characterize the effect of hydrogen on an alloy’s
oxidation embrittlement behavior. The entire range of a cladding material’s anticipated hydrogen level
should be characterized. To characterize the range of a cladding material’s anticipated hydrogen content,
an acceptable approach would be to determine the ductile-to-brittle transition for prehydrided material in
increments not more than every 100 wppm of hydrogen. The methodology outlined in this regulatory
guide for establishing a zirconium-alloy-specific limit other than the analytical limit provided in Figure 2
includes more repeat testing at each hydrogen level than was outlined for demonstrating consistency with
the established database. One approach would be to conduct oxidation and quench testing at the
transition ECR defined in Figure 2 and, based on the result of this test, proceed to an ECR 2% above or an
ECR 2% below this limit (2% above if the initial test was ductile, 2% below if the initial test was brittle).
Testing would then focus on the ECR range between ductile and brittle results and include three repeat
oxidation and quench tests at the transition ECR level. Following the guidance of DG-1262, each
oxidation and quench sample would be segmented into three RCT samples. The average of these three
RCT samples would be compared to the ductility criterion defined in terms of ≥1.0% permanent strain or
the offset strain ductility criterion presented in Appendix A.
Irradiated cladding material can be used to demonstrate that a cladding alloy’s embrittlement
behavior is accurately characterized by using prehydrided material. To demonstrate this, an acceptable
approach would be to determine the ductile-to-brittle transition for irradiated material with hydrogen
contents within 50 wppm of the anticipated maximum hydrogen content and within 50 wppm of half of
the anticipated maximum hydrogen content. The methodology outlined in this regulatory guide for
establishing a zirconium-alloy-specific limit other than the analytical limit provided in Figure 2 includes
more repeat testing for irradiated material than was outlined for demonstrating consistency with the
established database. One approach would be to conduct oxidation and quench testing at the transition
ECR defined by the as-received and prehydrided testing described above. Testing would then focus on
the ECR range between ductile and brittle results and include three repeat oxidation and quench tests at
the transition ECR level. Following the guidance of DG-1262, each oxidation and quench sample would
be segmented into three RCT samples. The average of these three RCT samples would be compared to
the ductility criterion defined in terms of ≥1.0% permanent strain or an offset strain ductility criterion
presented in Appendix A.
Appendix B to this regulatory guide provides a high-level overview of an acceptable test matrix.
The test matrix overview is intended to provide a clear picture of the range of material and test conditions
that could be used to establish an alloy-specific limit other than the analytical limit in Figure 2. It is
intended to complement the test matrix guidance in DG-1262.

DG 1263; Page 8 of 21

To establish a zirconium-alloy-specific limit for a new or existing fuel design, the applicant
would provide experimental results as part of the documentation supporting the NRC staff’s review and
approval of the new or existing fuel design (i.e., license amendment request or vendor topical report).
The applicant would provide details of the experimental technique (unless the experiments were
conducted in accordance with DG-1262) and the results of experiments conducted with as-fabricated,
prehydrided, and irradiated material, as well as a specified analytical limit on peak cladding temperature
and integral time at temperature that corresponds to the measured ductile-to-brittle transition for the
zirconium-alloy cladding material.
Upon review and approval of the fuel design, an acceptable method to demonstrate that licensees
meet the requirements of 10 CFR 50.46c is demonstrating that ECCS performance is such that local
oxidation is calculated below the specified analytical limit provided.

Methodology for Establishing Analytical Limits at Peak Oxidation Temperatures Less than
1,204 °C (2,200 °F)
The existing database and resulting analytical limit described in this regulatory guide is intended
to bound ECCS performance. In the test program, experiments were conducted at maximum oxidation
temperatures ≤1,200 °C ±10 °C and quenched at 800 °C.2 Some ECCS may perform such that the
maximum oxidation temperature is significantly below 1,204 °C (2,200 °F). Oxidation at lower
temperatures has been shown to increase the allowable calculated oxidation before embrittlement.
Therefore, conducting tests at lower peak temperatures may provide additional margin for some
zirconium-alloy cladding materials.
The objective of PQD testing to establish a limit at a peak cladding temperature lower than
1,204 °C (2,200 °F) is to characterize a cladding alloy’s embrittlement behavior through the entire
spectrum of conditions expected during operation. A diverse matrix of material conditions can serve to
provide a complete characterization, and repeat testing can be used to address expected variability in
oxidation behavior. The methodology outlined in this regulatory guide includes testing of as-received,
prehydrided, and irradiated material. The methodology outlined in this regulatory guide uses the
experimental procedure in DG-1262 to generate RCT data to establish a limit at a peak cladding
temperature lower than 1,204 °C (2,200 °F).
As-received cladding material may be used to characterize an alloy’s oxidation embrittlement
behavior in the as-received condition. The methodology outlined in this regulatory guide for establishing
an analytical limit at a peak oxidation temperature less than 1,204 °C (2,200 °F)includes more repeat
testing for irradiated material than was outlined for demonstrating consistency with the established
database. One approach would be to conduct oxidation and quench testing at a wide range of ECRs to
scope out a zirconium-alloy cladding material’s oxidation behavior in the as-received condition. Testing
would then focus on the ECR range between ductile and brittle results and include three repeat oxidation
and quench tests at two ECR levels (a total of six oxidation and quench tests). Following the guidance of
DG-1262, each oxidation and quench sample would be segmented into three RCT samples. The average
of these three RCT samples would be compared to the ductility criterion defined in terms of ≥1.0%
permanent strain or the offset strain ductility criterion presented in Appendix A.
2

These test conditions were selected with the objective of bounding the performance of ECCSs. They are considered
relevant and bounding for current light-water reactor ECCSs. However, it may be necessary to evaluate and possibly
modify the conditions accordingly for ECCSs of advanced reactor designs. In addition, postquench ductility
measurements were made at 135 °C. During the 1973 hearing, investigators suggested considering a test temperature
no higher than the saturation temperature during reflood (i.e., ≈135 °C). This test condition is considered relevant for
current light-water reactor ECCSs. However, it may be necessary to evaluate and possibly modify the conditions
accordingly for ECCSs of advanced reactor designs.

DG 1263; Page 9 of 21

Prehydrided cladding material may be used to characterize the effect of hydrogen on an alloy’s
oxidation embrittlement behavior. The entire range of a cladding material’s anticipated hydrogen level
should be characterized. To characterize the range of a cladding material’s anticipated hydrogen content,
an acceptable approach would be to determine the ductile-to-brittle transition for prehydrided material in
increments not more than every 100 wppm of hydrogen. One approach would be to conduct oxidation
and quench testing at the transition ECR defined in Figure 2, and, based on the result of this test, proceed
to an ECR 2% above or an ECR 2% below this limit (2% above if the initial test was ductile, 2% below if
the initial test was brittle). Testing would then focus on the ECR range between ductile and brittle results
and include three repeat oxidation and quench tests at the transition ECR level. Following the guidance
of DG-1262, each oxidation and quench sample would be segmented into three RCT samples. The
average of these three RCT samples would be compared to the ductility criterion defined in terms of
≥1.0% permanent strain or the offset strain ductility criterion presented in Appendix A.
Irradiated cladding material can be used to demonstrate that a cladding alloy’s embrittlement
behavior is accurately characterized by using prehydrided material. To demonstrate this, an acceptable
approach would be to determine the ductile-to-brittle transition for irradiated material with hydrogen
contents within 50 wppm of the anticipated maximum hydrogen content and within 50 wppm of half of
the anticipated maximum hydrogen content. The methodology outlined in this regulatory guide for
establishing an analytical limit at a peak oxidation temperature less than 2,200 °F includes more repeat
testing for irradiated material than was outlined for demonstrating consistency with the established
database. One approach would be to conduct oxidation and quench testing at the transition ECR defined
by the as-received and prehydrided testing conducted as described above. Testing would then focus on
the ECR range between ductile and brittle results and include three repeat oxidation and quench tests at
the transition ECR level. Following the guidance of DG-1262, each oxidation and quench sample would
be segmented into three RCT samples. The average of these three RCT samples would be compared to
the ductility criterion defined in terms of ≥1.0% permanent strain or the offset strain ductility criterion
presented in Appendix A.
Appendix B to this guide presents a high-level overview of an acceptable test matrix. The test
matrix overview is intended to provide a clear picture of the range of material and test conditions that
could be used to establish a limit at a peak cladding temperature lower than 1,204 °C (2,200 °F). It is
intended to complement the test matrix guidance in DG-1262.
To establish analytical limits at peak oxidation temperatures less than 1,204 °C (2,200 °F), the
applicant would provide experimental results as part of the documentation supporting the NRC staff’s
review and approval of the new fuel design or existing fuel design (i.e., license amendment request or
vendor topical report). The applicant would provide details of the experimental technique (unless the
experiments were conducted in accordance with DG-1262) and the results of experiments conducted with
as-fabricated, prehydrided, and irradiated material, as well as a specified analytical limit on peak cladding
temperature and integral time at temperature that corresponds to the measured ductile-to-brittle transition
for the zirconium-alloy cladding material.
For a given zirconium alloy, an applicant would be able to define analytical limits on integral
time at temperature (CP-ECR as a function of cladding hydrogen) corresponding to different peak
cladding temperature analytical limits. This approach may provide margin for high-burnup, highcorrosion, low-power fuel rods that experience a relatively benign temperature transient.
Upon review and approval of the fuel design, an acceptable method to demonstrate that licensees
meet the requirements of 10 CFR 50.46c is demonstrating that ECCS performance is such that local
oxidation is calculated below the specified analytical limit provided.
DG 1263; Page 10 of 21

Methodology for Establishing Analytical Limits for Breakaway Oxidation
The purpose of the requirements in 10 CFR 50.46, “Acceptance Criteria for Emergency Core
Cooling Systems for Light-Water Nuclear Power Reactors,” is to ensure core coolability during and
following a LOCA. If breakaway oxidation occurs, the embrittlement process is accelerated. Therefore,
the PQD analytical limits established in accordance with 10 CFR 50.46 are no longer effective to preclude
embrittlement, and core coolability may not be maintained even if the analytical limits on peak cladding
temperature and local oxidation (surrogate for time at temperature) are not exceeded.
In 10 CFR 50.46c, the NRC calls for measurement of the onset of breakaway oxidation for a
zirconium cladding alloy based on an acceptable experimental technique, evaluation of the measurement
relative to ECCS performance, and reporting of values measured (Ref. 1).
Based on data reported by Leistikow and Schanz (Ref. 12), zirconium alloys have been shown to
be susceptible to the breakaway oxidation phenomenon at temperatures as low as 650 °C. At 650 °C, it
took more than 4 hours (beyond LOCA-relevant times) for Zry-4 to accumulate 200-wppm hydrogen,
while at 800 °C, the time to accumulate 200-wppm hydrogen was only 1 hour (within LOCA-relevant
times). Thus, time spent in steam at ≤650 °C was benign with regard to breakaway oxidation and
hydrogen accumulation because of the very low oxidation rate. Because NUREG/IA-0211,
“Experimental Study of Embrittlement of Zr-1%Nb VVER Cladding under LOCA-Relevant Conditions,”
issued March 2005 (Ref. 4), did not present hydrogen-accumulation data for temperatures between
650 °C and 800 °C, there is no basis for not including time spent at temperatures >650 °C in establishing
the analytical limit for transient time.
To establish a zirconium-alloy-specific limit for a new or existing fuel design, the applicant
would provide experimental results for testing for breakaway oxidation behavior as part of the
documentation supporting the NRC staff’s review and approval of the new or existing fuel design
(i.e., license amendment request or vendor topical report). DG-1261 provides an experimental technique
to measure the onset of breakaway oxidation in order to establish a specified and acceptable limit on the
total accumulated time that a cladding may remain at high temperature. The applicant would provide
details of the experimental technique (unless the experiments were conducted in accordance with
DG-1261) and the results of experiments conducted. Applicants would establish the time limit for the
total accumulated time that the cladding may remain above 650 °C as part of the documentation
supporting the NRC staff’s review and approval of the new or existing fuel design (i.e., license
amendment request or vendor topical report).
Applicants may elect to establish the analytical limit for breakaway oxidation with conservatism
relative to the measured minimum time (i.e., reduce the time) to the onset of breakaway oxidation. This
approach may reduce the likelihood of reassessing small-break LOCA cladding temperature histories in
the event of a minor change in measured time to breakaway oxidation. For example, the minimum time
to breakaway oxidation may be demonstrated to occur at 975 °C at a time of 4,000 seconds. An applicant
may elect to establish an analytical limit of 3,000 seconds for the total accumulated time that the cladding
may remain above 650 °C.
Upon review and approval of the fuel design, an acceptable method to demonstrate that licensees
meet the requirements of 10 CFR 50.46c is demonstrating that ECCS performance is such that the total
accumulated time that the cladding is predicted to remain above a temperature at which the zirconium
alloy has been shown to be susceptible to this phenomenon is not greater than the proposed limit.

DG 1263; Page 11 of 21

Applying Analytical Limits
Qualification of Hydrogen Pickup Models
An alloy-specific cladding hydrogen uptake model will be required if a licensee chooses to use
the hydrogen-dependent embrittlement threshold provided in this regulatory guide. To establish an alloyspecific cladding hydrogen uptake model for a new or existing fuel design, the applicant would provide
postirradiation examination hydrogen measurement data and a hydrogen uptake model as part of the
documentation supporting the NRC staff’s review and approval of the new or existing fuel design
(i.e., license amendment request or vendor topical report). The documentation should include a claddingspecific plot of predicted versus measured cladding hydrogen content. The post-irradiation examination
data supporting the hydrogen uptake model should include values for multiple burnup levels, encompass
all applicable operating conditions and reactor coolant chemistry, and should quantify axial, radial, and
circumferential variability. (See the next section for further details.)
Accounting for Uncertainty and Variability in Hydrogen Content
Variation of hydrogen content across the radius of the cladding (hydride rim effect) and over
short axial distances (pellet-pellet interface effect) has been observed by many investigators. Studies
using prehydrided Zry-4 with dense hydride rims have demonstrated that the homogenization of hydrogen
across the radius of the cladding is very rapid at >900 °C due to the affinity of the beta phase for
hydrogen, as well as the high solubility of hydrogen in this phase. In the NRC’s LOCA research
program, significant circumferential variation ( ±100–140 wppm) in hydrogen content was measured (by
the LECO inert gas fusion thermal conductivity method) and observed (by optical microscopy) for highburnup cladding alloys. For oxidation test times at 1,200 °C up to the embrittlement CP-ECR level, no
significant diffusion of hydrogen in the circumferential direction was observed. Hydrogen-concentration
variations of 450 to 750 wppm measured for cladding quarter segments before LOCA testing remained
after LOCA testing.
The uncertainty in the model can be characterized and quantified by supporting the model with
postirradiation examination that include values for multiple burnup levels, encompasses all applicable
operating conditions and reactor coolant chemistry, and quantifies axial, radial, and circumferential
variability.
To apply the analytical limit in Figure 2 to an individual fuel rod (or fuel rod grouping), the
allowable CP-ECR should be based on predicted peak circumferential average hydrogen content for the
individual rod (or fuel rod grouping).
Postquench Ductility Analytical Limits
Based on the approved ECCS evaluation models and methods, the applicant should identify the
limiting combination of break size, break location, and initial conditions and assumptions that maximize
predicted peak cladding temperature and local oxidation (surrogate for time at temperature).
Combinations of initial conditions and uncertainties will vary between Appendix K, “ECCS Evaluation
Models,” to 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” and bestestimate methods. Separate cases may be necessary to identify the limiting scenario for peak cladding
temperature relative to local oxidation and vice versa. The applicant should demonstrate that predicted
peak cladding temperature remains below the lesser of the regulatory limit of 2,200 °F and the maximum
oxidation PQD temperature. The applicant will also need to demonstrate that the maximum predicted
local oxidation remains below the established PQD analytical limits.

DG 1263; Page 12 of 21

Because of the strong function of allowable local oxidation with cladding hydrogen content (see
Figure 1), the applicant may elect to subdivide the fuel rods within the core based on cladding hydrogen
content, burnup, fuel rod power, or a combination. For example, peak cladding temperature and local
oxidation calculations would be performed on three representative sets of fuel rods (e.g., 0–30 gigawattdays per metric ton of uranium (GWd/MTU), 30–45 GWd/MTU, and 45–62 GWd/MTU) using bounding
power histories for each fuel rod grouping. The predicted peak cladding temperature and local oxidation
would then be compared to the analytical limits for that range of burnup/hydrogen.
Application in the Rupture Region
During a postulated LOCA, fuel rods may be predicted to balloon and rupture as a result of
elevated cladding temperature and differential pressure (between rod internal pressure and system
pressure, which is decreasing because of a break in pressure boundary). The regions of the fuel rod near
the ballooned and ruptured location will thus be exposed to oxidation from the inside surface of the
cladding. Combined with oxygen diffusion from the cladding outside diameter (OD), oxygen diffusion
from the cladding inside diameter (ID) would further limit integral time at temperature to reach the
analytical limit in Figure 2. In addition, local regions above and below the rupture opening will absorb
significant hydrogen due to the steam oxidation on the ID, which may result in locally brittle regions
above and below the rupture. Finally, the balloon region will experience wall thinning, which impacts the
calculation of ECR because the value is taken to be a percentage of the preoxidation cladding thickness.
The LOCA acceptance criteria that limit peak oxidation temperature and maximum oxidation
level versus hydrogen content are based on retention of ductility. As discussed above, ductility will not
be retained everywhere in the balloon region.
To investigate the mechanical behavior of ruptured fuel rods, the NRC conducted integral LOCA
testing, designed to experience ballooning and rupture, on as-fabricated and hydrogen charged cladding
specimens and high burnup fuel rod segments exposed to high temperature steam oxidation followed by
quench (Ref. 13). The integral LOCA testing confirms that continued exposure to high temperature steam
environment weakens the already flawed region of the fuel rod surrounding the cladding rupture. Hence,
limitations on integral time at temperature are necessary to preserve an acceptable amount of mechanical
strength and fracture toughness. In addition, this research demonstrated that the degradation in strength
and fracture toughness with prolonged exposure to steam oxidation was enhanced with pre-existing
cladding hydrogen content.
Therefore, in regions of the fuel rod where the calculated conditions of transient pressure and
temperature lead to a prediction of cladding swelling, an acceptable approach would be to define the
cladding thickness as the cladding cross-sectional area divided by the cladding circumference, taken at a
horizontal plane at the elevation of the rupture, and to calculate two-sided oxidation using the CP
correlation and apply the analytical limit in Figure 2 (or an alternative specified and acceptable analytical
limit).
Accounting for Double-Sided Oxidation Due to the Fuel-Cladding Bond Layer
The NRC’s LOCA research program identified that, for high-burnup fuel, oxygen can diffuse into
the cladding metal during a LOCA from the ID as well as from the OD, even when no steam oxidation is
occurring on the ID (Refs. 5 and 6). The ID oxygen diffusion phenomenon was discovered in the United
States in 1977, confirmed by tests in Germany in 1979, and is seen in the present results.
Combined with oxygen diffusion from the cladding OD, oxygen diffusion from the cladding ID
would further limit integral time at temperature to nil ductility. To account for the observation that
oxygen can diffuse into the cladding metal during a LOCA from the ID, one acceptable approach would
DG 1263; Page 13 of 21

be to calculate two-sided local oxidation for fuel rods with a local (nodal) exposure beyond
30 GWd/MTU. It should be noted that there would be no metal-water-reaction heat associated with this
process on the ID, in contrast to the situation in a rupture node. A threshold for the onset of this inside
surface oxidation source other than 30 GWd/MTU may be proposed by a licensee and provided as part of
the documentation supporting the NRC staff’s review and approval of the new or existing fuel design
(i.e., license amendment request or vendor topical report). A threshold other than 30 GWd/MTU could be
supported by metallographic images of bonding layers as a function of burnup.
Breakaway Oxidation Analytical Limits
Based on the approved ECCS evaluation models and methods, the applicant should identify the
limiting combination of break size, break location, and initial conditions and assumptions that maximize
the total accumulated time that the cladding is predicted to remain above a temperature at which the
zirconium alloy has been shown to be susceptible to this phenomenon. The applicant should demonstrate
that this time interval remains below the established alloy-specific breakaway oxidation analytical limit.
The applicant may credit operator actions to limit the duration at elevated temperatures provided
these actions are consistent with existing emergency operating procedures and the timing of such actions
is validated by operator training on the plant simulator.

C. REGULATORY POSITION
Regulatory Positions 1 through 4 provide acceptable methods for establishing an analytical limit on peak
cladding temperature and integral time at temperature for zirconium-alloy cladding materials. Applicants
should use one of the four methods provided. Regulatory Position 5 provides an acceptable method for
establishing an analytical limit for breakaway oxidation.
1.

Apply the specified and acceptable limit defined in Figure 2 of this regulatory guide for cladding
materials tested in the NRC’s LOCA research program.

Demonstrate comparable behavior of cladding alloys not tested in the NRC’s LOCA research
program, with the database established in the NRC’s LOCA research program, in order to apply
the limit defined in Figure 2 of this regulatory guide.
a. Conduct oxidation and quench testing on (1) as-fabricated material, (2) prehydrided
material for the entire range of a cladding material’s anticipated hydrogen level (testing
pre-hydrided material in increments not more than every 100 wppm hydrogen), and (3)
irradiated material for the entire range of a cladding material’s anticipated hydrogen level
(testing irradiated material with hydrogen contents within 50 wppm of the anticipated
maximum hydrogen content and within 50 wppm of half of the anticipated maximum
hydrogen content) at the transition ECR defined in Figure 2 for the each sample’s
hydrogen level, an ECR 1% above and an ECR 1% below this limit.
b. Determine the ECR at which the material transitions from ductile-to-brittle behavior
using the results of ring compression testing conducted using the experimental procedure

DG 1263; Page 14 of 21

and the guidance provided in DG-1262 for each material condition called for in
Regulatory Position 2.a 3. Compare to the limit defined in Figure 2 of this guide.
c. If the experimental results for the new fuel design measured the transition from ductile to
brittle behavior to be no lower than the analytical limit defined in Figure 2, the analytical
limit defined in Figure 2 may be established for the cladding alloy not tested in NRC’s
LOCA research program.
d. Provide details of experimental techniques, unless conducted using the guidance in DG1262, as part of the documentation supporting the staff’s review and approval of the new
or existing fuel design (i.e., license amendment request or vendor topical report).
e. Provide results of experiments conducted with as-fabricated, irradiated material and
identify the specific analytical limit on peak cladding temperature and integral time at
temperature as part of the documentation supporting the staff’s review and approval of
the new or existing fuel design (i.e., license amendment request or vendor topical report).
The limit should correspond the ductile-to-brittle transition for the zirconium-alloy
cladding material and the oxidation temperature of the oxidation and quench experiments.
3. Establish a zirconium alloy specific analytical limit on peak cladding temperature and integral
time at temperature at a peak cladding oxidation temperature of 2200°F.
a. Conduct oxidation and quench testing on (1) as-fabricated material, (2) prehydrided
material for the entire range of a cladding material’s anticipated hydrogen level (testing
pre-hydrided material in increments not more than every 100 wppm hydrogen), and (3)
irradiated material for the entire range of a cladding material’s anticipated hydrogen level
(testing irradiated material with hydrogen contents within 50 wppm of the anticipated
maximum hydrogen content and within 50 wppm of half of the anticipated maximum
hydrogen content) at four oxidation levels for each material condition3, in increments not
greater than 2% ECR.
b. With the results of four oxidation levels, for each material condition called for in (a)
above, determine the ECR range in which the transition from ductile-to-brittle behavior
occurs and conduct three repeat oxidation and quench tests at each ECR level within this
range using the guidance provided in DG-1262.
c. Determine the ECR at which the material transitions from ductile to brittle behavior,
using the results of ring compression testing conducted using the experimental procedure
and the guidance provided in DG-1262 for each material condition called for in
Regulatory Position 3.a.
d. Provide details of experimental techniques, unless conducted using the guidance in
DG-1262, as part of the documentation supporting the NRC staff’s review and approval

“each material condition” refers to the range of as-fabricated, prehydrided and irradiated material called for within the
discussion section of this regulatory guide. For a zirconium alloy with an anticipated, end of life hydrogen content, the
range of material conditions called for within the discussion section of this regulatory guide would include (1) asfabricated, (2) pre-hydrided material at 100 wppm H, (3) pre-hydrided material at 200 wppm H, (4) pre-hydrided
material at 300 wppm H, (5) pre-hydrided material at 400 wppm H, (6) irradiated material with a hydrogen content of
200±50 wppm H, and (7) irradiated material with a hydrogen content of 400±50 wppm H. See also Appendix B of this
guide for a high-level overview of an acceptable test matrix.

DG 1263; Page 15 of 21

of the new or existing fuel design (i.e., license amendment request or vendor topical
report).
e. Provide the results of experiments conducted with as-fabricated, irradiated material and
identify the specific analytical limit on peak cladding temperature and integral time at
temperature as part of the documentation supporting the NRC staff’s review and approval
of the new or existing fuel design (i.e., license amendment request or vendor topical
report). The limit should correspond the ductile-to-brittle transition for the zirconiumalloy cladding material and the oxidation temperature of the oxidation and quench
experiments.
4.

Establish an analytical limit on peak cladding temperature and integral time at temperature at a
peak oxidation temperature less than 2,200 °F.
a. Conduct oxidation and quench testing on (1) as-fabricated material, (2) prehydrided
material for the entire range of a cladding material’s anticipated hydrogen level (testing
prehydrided material in increments not more than every 100 wppm of hydrogen), and (3)
irradiated material for the entire range of a cladding material’s anticipated hydrogen level
(testing irradiated material with hydrogen contents within 50 wppm of the anticipated
maximum hydrogen content and within 50 wppm of half of the anticipated maximum
hydrogen content) at four oxidation levels for each material condition3, in increments not
greater than 2% ECR.
b. With the results of four oxidation levels for each material condition called for in
Regulatory Position 4.a, determine the ECR range in which the transition from ductile to
brittle behavior occurs and conduct three repeat oxidation and quench tests at each ECR
level within this range using the guidance provided in DG-1262.
c. Determine the ECR at which the material transitions from ductile-to-brittle behavior,
using the results of ring compression testing conducted using the experimental procedure
and the guidance provided in DG-1262 for each material condition called for in
Regulatory Position 4.a.
d. Provide details of experimental techniques, unless conducted using the guidance in DG1262, as part of the documentation supporting the NRC staff’s review and approval of the
new or existing fuel design (i.e., license amendment request or vendor topical report).
e. Provide the results of experiments conducted with as-fabricated, irradiated material and
identify the specific analytical limit on peak cladding temperature and integral time at
temperature as part of the documentation supporting the NRC staff’s review and approval
of the new or existing fuel design (i.e., license amendment request or vendor topical
report). The limit should correspond the ductile-to-brittle transition for the zirconiumalloy cladding material and the oxidation temperature of the oxidation and quench
experiments.
5. Establish an analytical limit for breakaway oxidation.
a. Follow the procedures in DG-1261 to establish the shortest time observed to lead to
breakaway oxidation for a zirconium cladding alloy.

DG 1263; Page 16 of 21

b. Provide the results of the testing as part of the documentation supporting the NRC staff’s
review and approval of the new or existing fuel design (i.e., license amendment request
or vendor topical report).
c. Establish an analytical limit for the total accumulated time the cladding may remain
above 650°C, which is less than or equal to the shortest time observed to lead to
breakaway oxidation.
d. Provide the analytical limit for breakaway oxidation as part of the documentation
supporting the NRC staff’s review and approval of the new or existing fuel design (i.e.,
license amendment request or vendor topical report).

D. IMPLEMENTATION
The purpose of this section is to provide information on how applicants and licensees4 may use
this guide and information regarding the NRC’s plans for using this regulatory guide. In addition, it
describes how the NRC staff complies with the Backfit Rule (10 CFR 50.109) and any applicable finality
provisions in 10 CFR Part 52.
Use by Licensees
Licensees may voluntarily5 use the guidance in this document to demonstrate compliance with the
underlying NRC regulations. Methods or solutions that differ from those described in this regulatory
guide may be deemed acceptable if they provide sufficient basis and information for the NRC staff to
verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations.
Licensees may use the information in this regulatory guide for actions which do not require NRC
review and approval such as changes to a facility design under 10 CFR 50.59 that do not require prior
NRC review and approval. Licensees may use the information in this regulatory guide or applicable parts
to resolve regulatory or inspection issues.
Use by NRC Staff
During regulatory discussions on plant specific operational issues, the staff may discuss with
licensees various actions consistent with staff positions in this regulatory guide, as one acceptable means
of meeting the underlying NRC regulatory requirement. Such discussions would not ordinarily be
considered backfitting even if prior versions of this regulatory guide are part of the licensing basis of the
facility. However, unless this regulatory guide is part of the licensing basis for a facility, the staff may
not represent to the licensee that the licensee’s failure to comply with the positions in this regulatory
guide constitutes a violation.
If an existing licensee voluntarily seeks a license amendment or change and (1) the NRC staff’s
consideration of the request involves a regulatory issue directly relevant to this new or revised regulatory
guide and (2) the specific subject matter of this regulatory guide is an essential consideration in the staff’s
determination of the acceptability of the licensee’s request, then the staff may request that the licensee
4

In this section, “licensees” refers to licensees of nuclear power plants under 10 CFR Parts 50 and 52; and the term
“applicants,” refers to applicants for licenses and permits for (or relating to) nuclear power plants under 10 CFR Parts
50 and 52, and applicants for standard design approvals and standard design certifications under 10 CFR Part 52.

In this section, “voluntary” and “voluntarily” means that the licensee is seeking the action of its own accord, without
the force of a legally binding requirement or an NRC representation of further licensing or enforcement action.

DG 1263; Page 17 of 21

either follow the guidance in this regulatory guide or provide an equivalent alternative process that
demonstrates compliance with the underlying NRC regulatory requirements. This is not considered
backfitting as defined in 10 CFR 50.109(a)(1) or a violation of any of the issue finality provisions in 10
CFR Part 52.
The NRC staff does not intend or approve any imposition or backfitting of the guidance in this
regulatory guide. The NRC staff does not expect any existing licensee to use or commit to using the
guidance in this regulatory guide, unless the licensee makes a change to its licensing basis. The NRC
staff does not expect or plan to request licensees to voluntarily adopt this regulatory guide to resolve a
generic regulatory issue. The NRC staff does not expect or plan to initiate NRC regulatory action which
would require the use of this regulatory guide. Examples of such unplanned NRC regulatory actions
include issuance of an order requiring the use of the regulatory guide, requests for information under
10 CFR 50.54(f) as to whether a licensee intends to commit to use of this regulatory guide, generic
communication, or promulgation of a rule requiring the use of this regulatory guide without further
backfit consideration.
If a licensee believes that the NRC is either using this regulatory guide or requesting or requiring
the licensee to implement the methods or processes in this regulatory guide in a manner inconsistent with
the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in
accordance with the guidance in NUREG-1409 and NRC Management Directive 8.4.

DG 1263; Page 18 of 21

GLOSSARY
breakaway oxidation—For the purposes of this regulatory guide, the fuel-cladding oxidation
phenomenon in which weight gain rate deviates from normal kinetics. This change occurs with a
rapid increase of hydrogen pickup during prolonged exposure to a high-temperature steam
environment, which promotes loss of cladding ductility.
corrosion—For the purposes of this regulatory guide, the formation of a zirconium oxide layer resulting
from the reaction of zirconium with coolant water during normal operation.
loss-of-coolant accident (LOCA)—A hypothetical accident that would result from the loss of reactor
coolant at a rate in excess of the capability of the reactor coolant makeup system, from breaks in
pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to
the double-ended rupture of the largest pipe in the reactor coolant system.
offset strain—For the purposes of this regulatory guide, the value determined from a load-displacement
curve by the following procedure: (1) linearize the initial loading curve, (2) use the slope of the
initial loading curve to mathematically unload the sample at the peak load before a significant
load drop (≈30–50%) indicating a through-wall crack along the length of the sample, and
(3) determine the offset displacement (distance along the displacement axis between loading and
unloading lines). This offset displacement is normalized to the outer diameter of the preoxidized
cladding to determine a relative plastic strain.
oxidation—For the purposes of this regulatory guide, the formation of a zirconium oxide layer resulting
from the reaction of zirconium with high-temperature steam during LOCA conditions.
permanent strain—For the purposes of this regulatory guide, the difference between the posttest outer
diameter (after the sample is unloaded) and the pretest outer diameter of a cladding ring,
normalized to the initial diameter of the cladding ring.

DG 1263; Page 19 of 21

REFERENCES6
1.

“Proposed Rule FRN,” ADAMS Accession No. ML12283A174

DG-1262, “Testing for Postquench Ductility,” U.S. Nuclear Regulatory Commission,
Washington, DC.

DG-1261, “Conducting Periodic Testing for Breakaway Oxidation Behavior,” U.S. Nuclear
Regulatory Commission, Washington, DC.

NUREG/IA-0211, “Experimental Study of Embrittlement of Zr-1%Nb VVER Cladding under
LOCA-Relevant Conditions,” U.S. Nuclear Regulatory Commission, Washington, DC,
March 2005. (ADAMS Accession No. ML051100343)

IFE/KR/E-2008/004, “LOCA Testing of High Burnup PWR Fuel in the HBWR. Additional PIE
on the Cladding of the Segment 650-5,” Institute for Energy Technology, Kjeller, Norway,
April 2008. (ADAMS Accession No. ML081750715)

Research Information Letter 0801, “Technical Basis for Revision of Embrittlement Criteria in
10 CFR 50.46,” U.S. Nuclear Regulatory Commission, Washington, DC, May 30, 2008.
(ADAMS Accession No. ML081350225)

NUREG/CR-6967, “Cladding Embrittlement during Postulated Loss-of-Coolant Accidents,”
U.S. Nuclear Regulatory Commission, Washington, DC, July 2008. (ADAMS Accession
No. ML082130389)

Yan, Y., T.A. Burtseva, and M.C. Billone, “Post-Quench Ductility Results for North Anna HighBurnup 17×17 ZIRLOTM Cladding with Intermediate Hydrogen Content,” ANL letter report to
NRC, April 17, 2009. (ADAMS Accession No. ML091200702)

May 15th email correspondence

10.

ORNL/NUREG-17, “Zirconium Metal-Water Oxidation Kinetics IV. Reaction Rate Studies,”
U.S. Nuclear Regulatory Commission, Washington, DC, August 1977.

11.

Billone, M.C., T.A. Burtseva, and Y. Yan, “Cladding Tests for Conditions, Monthly Letter Status
Report,” ANL letter report to the NRC, October 22, 2009, ANL Response to ANPR

12.

Leistikow, S., and G. Schanz, “Oxidation Kinetics and Related Phenomena of Zircaloy-4 Fuel
Cladding Exposed to High Temperature Steam and Hydrogen-Steam Mixtures under PWR
Accident Conditions,” Nuclear Engineering and Design, 103: 65–84.

Publicly available NRC published documents are available electronically through the NRC Library at:
http://www.nrc.gov/reading-rm/doc-collections/. The documents can also be viewed on-line or printed for a fee in the
NRC’s Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR,
Washington, DC 20555; telephone 301-415-4737 or (800) 397-4209; fax (301) 415-3548; and e-mail
pdr.resource@nrc.gov.

DG 1263; Page 20 of 21

13.

USNRC staff report, “Mechanical Behavior of Ballooned and Ruptured Cladding” May 2011,
Draft available as ADAMS Public Accession Number ML111370032

DG 1263; Page 21 of 21

APPENDIX A
RELATIONSHIP BETWEEN OFFSET STRAIN AND PERMANENT
STRAIN
For as-fabricated cladding compressed at room temperature (RT) or 135 degrees Celsius (°C) and
at 0.033 millimeter per second (mm/s) to a total displacement of 2 millimeters (mm), the difference
between offset displacement and permanent displacement is ≤0.2 mm, which corresponds to a strain
difference of ≈2%. As the applied displacement is decreased, the plastic deformation decreases and the
deviation between offset and permanent strain also decreases. This was demonstrated by conducting a set
of ring compression tests designed to result in low permanent strains of 1.0 to 2.3%. Table A-1 shows the
results of these tests.
Table A-1. Results of Ring Compression Tests Conducted with As-Fabricated Cladding Samples at
RT and 2 mm/minute Displacement Rate. Total Applied Displacements Were Chosen to
Give Low Permanent Strains (dd/Do) in the Range of 1.0 to 2.3% and Corresponding Low
Offset Strains
Material
(Do, mm)
15×15 Zry-4
(10.91 mm)
17×17
ZIRLOTM
(9.48 mm)
17×17 M5
(9.48 mm)

Sample ID
IPS or
AG No.
101B7
101B8
101B9
101B10
109D7
109D8
109D9
109D10
636B2
636B3
636B4

Offset
Displacement
δd, mm
0.24
0.20
0.20
0.16
0.25
0.17
0.14
0.14
0.18
0.14
0.15

Permanent
Displacement
dd, mm
0.21
0.17
0.18
0.14
0.22
0.16
0.12
0.12
0.19
0.14
0.15

Permanent
Strain
dd/Do, %
1.9
1.6
1.6
1.3
2.3
1.7
1.3
1.3
2.0
1.5
1.6

Strain
Difference
(δd – dd)/Do, %
0.3
0.3
0.2
0.2
0.3
0.1
0.2
0.2
0.0
0.0
0.0

For as-fabricated and prehydrided cladding oxidized at ≤1,200 °C, the difference between offset
and permanent displacement depends on both the oxidation level and the magnitude of the permanent
displacement. For material with high ductility, the difference in displacements can be as high as 0.5 mm.
For material with essentially no ductility, both offset and the permanent displacement values are in the
"noise of uncertainty" and their difference can be as low as 0.01 mm.
However, what is relevant to the determination of the ductile-to-brittle transition oxidation level
is the error in offset strain as determined by the difference between offset (δp/Do in %) and permanent
(dp/Do in %) strains for permanent strains in the range of 1.0 to 2.3%. Figure A-1 summarizes the data
reported in Refs. 1 and 15, in Figures 1 and 2 of this procedure, and in Table A-1. The data are plotted as
a function of Cathcart-Pawel equivalent cladding reacted (CP-ECR). Low values of permanent strain at
low CP-ECR levels (e.g., 5–10%) are from prehydrided Zircaloy-4 (Zry-4), high-burnup Zry-4, and
ZIRLOTM samples. Low values of permanent strain at intermediate CP-ECR levels (10–18%) are from
high-burnup ZIRLOTM and M5 samples. Low values of permanent strain at high CP-ECR values (15–
20%) are from as-fabricated cladding materials. The best linear fit to the data is given by:

Appendix A to DG-1263, Page A-1

δp/Do - dp/Do = 0.25 + 0.0863 CP-ECR

(A1)

The one-sigma upper bound to the data is given by:
δp/Do - dp/Do = 0.41 + 0.1082 CP-ECR

(A2)

Because of the large data scatter in Figure A-1, the one-sigma upper bound is used to establish the offsetstrain ductility criterion. It is derived by setting the permanent strain (dp/Do) in Equation (A2) to 1.0%:
δp/Do ≥ 1.41 + 0. 1082 CP-ECR

(A3)

For multiple offset-strain data points at the same CP-ECR level, the average value for the data set,
rounded to the nearest tenth of a percent, should be used for δp/Do in Equation (A3). Similarly, the limit
calculated from the right-hand side of Equation (A3) should also be rounded to the nearest tenth of a
percent.

Appendix A to DG-1263, Page A-2

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

10 12 14 16 18 20 22

CP-Predicted ECR (%)

1-σ Upper Bound

Best Fit

ZIRLO

Zircaloy

Figure A-1. Difference in offset strain and permanent strain as a function of calculated oxidation level (CP-ECR) for permanent strains
near the embrittlement threshold (1.0% to 2.3%) for as-fabricated, prehydrided, and high-burnup cladding alloys oxidized at
1,200 °C and ring-compressed at RT and 135 °C and at 0.033 mm/s.

Offset Minus Permanent Strain (%)

Overview of Sample Test Matrix To Generate Ring Compression Test Data To Establish an Alloy-Specific Limit or
To Establish a Limit at a Peak Cladding Temperature Lower than 2,200 °F

Appendix B to DG-1263, Page B-1

As-received cladding material may be used to conduct scoping tests to identify the oxidation equivalent cladding reacted (ECR) where
transition behavior likely occurs. Table B-1 provides a sample test matrix for this scoping test.

The objective of postquench ductility testing to establish an alloy-specific limit or to establish a limit at a peak cladding temperature lower
than 2,200 °F is to characterize a cladding alloy’s embrittlement behavior through the entire spectrum of conditions expected during operation. A
diverse matrix of material conditions can provide a complete characterization, and repeat testing can be used to address expected variability in
oxidation behavior. The test matrix to generate ring compression test data to establish an alloy-specific limit or to establish a limit at a peak
cladding temperature lower than 2,200 °F provided here includes testing of as-received, prehydrided, and irradiated material.

B-1.

This appendix will provide two examples. The first series of test matrices could be used to generate ring compression test data to establish
an alloy-specific limit or to establish a limit at a peak cladding temperature lower than 2,200 degrees Fahrenheit (°F). The second series of test
matrices could be used to generate ring compression test data to demonstrate consistency with the analytical limit provided in Figure 2 of this
regulatory guide.

Draft Regulatory Guide (DG)-1262, “Testing for Postquench Ductility,” provides a detailed test procedure that is acceptable for generating
postquench ductility data through ring compression tests. This appendix is intended to provide a simple, straightforward overview of acceptable
test matrices. The test matrix overviews provided in this appendix are consistent with the guidance in DG-1262.

OVERVIEW OF ACCEPTABLE TEST MATRICES

APPENDIX B

Average ≥
Offs et s tra i n
Cri teri on ?

Average of 3
RTC samples

Ring
compression
Sample
Offset Strain
Measurement

Oxidation Level
(ECR)

___ %

Offset
strain
Criterion of
10% ECR =

2.5

Yes /
No

_________ %

10%

___ %

2
___ %

Offset
strain
Criterion of
13% ECR =

2.8

Yes /
No

_________ %

___ %

13%

___ %

2
___ %

Offset
strain
Criterion of
17% ECR =

3.2

Yes /
No

_________ %

___ %

17%

___ %

Offset
strain
Criterion of
20% ECR =

3.6

Yes /
No

_________ %

___ %

20%

Table B-1. Sample Test Matrix for Scoping Tests for As-Received Cladding Material

Appendix B to DG-1263, Page B-2

From the scoping test, a brittle result and a ductile result will likely be identified. For example, the average of three samples at 17% ECR
may be determined to be ductile using the ductility criterion ≥1.0% permanent strain or the offset strain criterion defined in Appendix A to this
regulatory guide, while the average of three samples at 20% ECR may be determined to be brittle. Following the evaluation of results from the
scoping tests, the next set of tests with as-received cladding material should be conducted within the range where the brittle and ductile results
were observed, to identify the ECR at which the transition occurs. For example, if the average of three samples at 17% ECR was determined to be
ductile, while the average of three samples at 20% ECR was determined to be brittle, the next set of tests should be conducted at 18% and 19%
ECR. In the transition region, repeat tests provide improved characterization because of variability in oxidation behavior. Therefore, a sample test
matrix for testing in this region includes multiple oxidation and quench tests at each oxidation level, as shown in Table B-2. The transition from
ductile to brittle behavior should be identified to occur at the highest Cathcart-Pawel equivalent cladding reacted (CP-ECR) at which the
permanent strain is ≥1.0%.

Scoping test

18%

18%
2

19%
2

_________ %
Yes / No

Average of
RTC samples

Average ≥
Offset strain
Criterion?

Yes / No

_________ %

Offset Strain
___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ %
Measurement

Ring
compression
Sample

Oxidation
Level (ECR)

Table B-2. Sample Test Matrix for Testing As-Received Cladding Material in the Identified Transition Region

Appendix B to DG-1263, Page B-3

Table B-3. Tabulated Values for the Embrittlement Threshold in DG-1263
Hydrogen Content (wppm)
Embrittlement ECR
0
18%
100
15.5%
200
12%
300
9%
400
6%
500
5%
600
4%
700
3%
800
2%

Transition Region

Ductile-tobrittle
transition
identified?

___ %

Yes / No

_________ %

___ %

Yes - continue tests at ECR betw een ductile and
brittle level;
No - conduct test additional scoping tests

Yes / No

Average ≥
Offset strain
Criterion?

___ %

_________ %

___ %

Average of
RTC samples

Offset Strain
___ %
Measurement

Ring
compression
Sample

If 15.5% was brittle,
2nd test at 13.5%
If 15.5% was ductile,
2nd test at 17.5%

1st test
at 15.5%

100

___ %

Yes / No

_________ %

___ %

If 12% was brittle,
2nd test at 10%
If 12% was ductile,
2nd test at 14%

Yes - continue tests at ECR betw een ductile and
brittle level;
No - conduct test additional scoping tests

Yes / No

_________ %

___ %

1st test
at 12%

200

___ %

Yes / No

_________ %

___ %

If 9% was brittle,
2nd test at 7%
If 9% was ductile,
2nd test at 11%

Yes - continue tests at ECR betw een ductile and
brittle level;
No - conduct test additional scoping tests

Yes / No

_________ %

___ %

1st test
at 9%

300

___ %

Yes / No

Yes - continue tests at ECR betw een ductile and
brittle level;
No - conduct test additional scoping tests

Yes / No

___ %

_________ %

___ %

If 6% was brittle,
2nd test at 4%
If 6% was ductile,
2nd test at 8%

400

___ %

_________ %

___ %

1st test
at 6%

Table B-4. Sample Test Matrix for Scoping Tests for Prehydrided Cladding Material

Oxidation
Level (ECR)

Hydrogen
Level
(wppm)

Appendix B to DG-1263, Page B-4

The objective of the scoping tests for prehydrided material is to identify an ECR level at which ductile behavior is observed, and an ECR
level at which brittle behavior is observed, and thus identify the range in which the ductile-to-brittle behavior is observed. Once this range of ECR
levels is identified, the test matrix continues with testing at an ECR level between the ECR level at which ductile behavior is observed and the
ECR level at which brittle behavior is observed. Table B-5 provides a test matrix that can be used at each hydrogen level to characterize
embrittlement behavior at the ECR level at which the ductile-to-brittle transition is expected to occur.

Scoping Tests

Ring
compression
Sample

Transition
ECR

Repeat for each hydrogen level

Yes / No

Average ≥
Offset strain
Criterion?

Appendix B to DG-1263, Page B-5

Irradiated cladding material can be used to demonstrate that a cladding alloy’s embrittlement behavior is accurately characterized by
using prehydrided material. To demonstrate this, an acceptable approach would be to determine the ductile-to-brittle transition for irradiated
material with hydrogen contents within 50 wppm of the anticipated maximum hydrogen content and within 100 wppm of half of the anticipated
maximum hydrogen content. The test matrix below illustrates an acceptable test matrix for a cladding material that is anticipated to have a
maximum hydrogen content of 400-wppm hydrogen at end of life. Table B-6 provides a sample test matrix for demonstrating that a cladding
alloy’s embrittlement behavior is accurately characterized by using prehydrided material.

_________ %

Average of
RTC samples

Offset Strain
___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ %
Measurement

Transition
ECR

Oxidation
Level (ECR)

Hydrogen
Level

Table B-5. Sample Test Matrix for Testing Prehydrided Cladding Material in the Identified Transition Region

Transition Region

Transition
ECR - 1%

Transition
ECR

Transition
ECR + 1%

Transition
ECR - 1%

Half of Licensed Hydrogen Limit ± 50 w ppm

Yes / No

_________ %

Yes / No

_________ %
Yes / No

_________ %

Yes / No

_________ %

Yes / No

_________ %

___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ %

Transition
ECR + 1%

Transition
ECR

Licensed Hydrogen Limit ± 50 w ppm

Table B-6. Sample Test Matrix for Testing Irradiated Cladding Material

Overview of Sample Test Matrix To Generate Ring Compression Test Data To Demonstrate Consistency with the
Analytical Limit Provided in DG-1263

Average ≥ Offset
strain Criterion?
Ductile-to-brittle
transition
comparable to prehydrided material?

Average of RTC
samples

Offset Strain
Measurement

Ring compression
Sample

Hydrogen Level
(
)
Oxidation Level
(ECR)

Appendix B to DG-1263, Page B-6

The objective of postquench ductility testing to demonstrate consistency with the analytical limit provided in Figure 2 of this regulatory
guide is to confirm that the transition to brittle behavior does not take place at a lower ECR than the provided limit. Because of this, the matrix of
material conditions and oxidation levels can be significantly reduced from the matrix outlined in the previous section. A range of material
conditions can serve to provide a characterization through the spectrum of conditions expected during operation, and repeat testing can be used to
address expected variability in oxidation behavior. The test matrix provided here to generate ring compression test data to demonstrate
consistency with the analytical limit provided in Figure 2 of this regulatory guide includes testing of as-received, prehydrided, and irradiated
material. The transition from ductile to brittle behavior should be identified to occur at the highest CP-ECR at which the permanent strain is
≥1.0%. Consistency with the analytical limit provided in Figure 2 of this regulatory guide would be considered demonstrated when the transition
from ductile to brittle behavior is not lower than the provided limit.

B-2.

Irradiated Testing

Transition
ECR from
Fig.2
Transition
ECR + 1%

As-Received
Transition
ECR - 1%

Transition
ECR from
Fig.2

Transition
ECR + 1%

100 w ppm
Transition
ECR - 1%

Appendix B to DG-1263, Page B-7

Ring
1
2
3
1
2
3
1
2
3
1
2
3
1
2
3
1
2
3
compression
Sample
Offset Strain
___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ %
Measurement
Average of RTC
_________ %
_________ %
_________ %
_________ %
_________ %
_________ %
samples
Average ≥ Offset
Yes / No
Yes / No
Yes / No
Yes / No
Yes / No
Yes / No
strain Criterion?
Ductile-to-brittle
transition at or
Yes / No
Yes / No
above limit of
Fig.2?

Oxidation Level
(ECR)

Hydrogen Level
(wppm)

Table B-7. Sample Test Matrix for Testing As-Received, Prehydrided, and Irradiated
Cladding Material To Demonstrate Consistency with the Analytical Limit Provided in Figure 2 of DG-1263

Table B-7 provides a complete test matrix, including as-received, prehydrided, and irradiated material, acceptable to the NRC for using in
postquench ductility testing to demonstrate consistency with the analytical limit provided in Figure 2 of this regulatory guide.

Irradiated cladding material can be used to demonstrate that a cladding alloy’s embrittlement behavior is accurately characterized by
using prehydrided material. To demonstrate this, an acceptable approach would be to determine the ductile-to-brittle transition for irradiated
material with hydrogen contents within 50 wppm of the anticipated maximum hydrogen content and within 100 wppm of half of the anticipated
maximum hydrogen content. Table B-7 illustrates an acceptable test matrix for a cladding material that is anticipated to have a maximum
hydrogen content of 400 wppm hydrogen at end of life.

not more than every 100 wppm of hydrogen. The test matrix in Table B-7 illustrates an acceptable test matrix for a cladding material that is
anticipated to have a maximum hydrogen content of 400-wppm hydrogen at end of life. The analytical limit provided in Figure 2 of this
regulatory guide as a function of hydrogen content may be used as a guide in selecting the range of oxidation levels to be included in the test
matrix. Table 3 of this regulatory guide provides the embrittlement threshold in Figure 2 in tabular form for clarity.

As-Received and Pre-hydrided Testing

Transition
ECR from
Fig.2
Transition
ECR + 1%

200
Transition
ECR - 1%

Transition
ECR from
Fig.2
Transition
ECR + 1%

300
Transition
ECR - 1%

Transition
ECR from
Fig.2
Transition
ECR + 1%

400
Transition
ECR - 1%

Appendix B to DG-1263, Page B-8

Ring
1
2
3
1
2
3
1
2
3
compression
Sample
Offset Strain
___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ % ___ %
Measurement
Average of RTC
_________ %
_________ %
_________ %
samples
Average ≥ Offset
Yes / No
Yes / No
Yes / No
strain Criterion?
Ductile-to-brittle
transition at or
Yes / No
above limit of
Fig.2?

Oxidation Level
(ECR)

Hydrogen Level
(wppm)

Oxidation Level
(ECR)

Hydrogen Level
(wppm)

As-Received and Pre-hydrided Testing

Irradiated Material Testing

Transition
ECR from
Fig.2
Transition
ECR + 1%

Transition
ECR - 1%

Licensed Hydrogen Limit ± 50 w ppm
Transition
ECR from
Fig.2
Transition
ECR + 1%

Transition
ECR - 1%

Half of Licensed Hydrogen Limit ± 50 w ppm

Appendix B to DG-1263, Page B-9

Oxidation Level
(ECR)

Hydrogen Level
(wppm)

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