[Federal Register Volume 74, Number 155 (Thursday, August 13, 2009)]
[Proposed Rules]
[Pages 40765-40776]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-19423]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 /
Proposed Rules
[[Page 40765]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AH42
[NRC-2008-0332]
Performance-Based Emergency Core Cooling System Acceptance
Criteria
AGENCY: Nuclear Regulatory Commission.
ACTION: Advance notice of proposed rulemaking.
-----------------------------------------------------------------------
SUMMARY: This advance notice of proposed rulemaking (ANPR) presents a
conceptual approach that the Nuclear Regulatory Commission (NRC) is
considering in a rulemaking effort to revise the acceptance criteria
for emergency core cooling systems (ECCSs) for light-water nuclear
power reactors as currently required by NRC regulations that govern
domestic licensing of production and utilization facilities. Revised
ECCS acceptance criteria would reflect recent research findings that
indicate the current criteria should be re-evaluated for all fuel
cladding materials in all potential conditions. Further, the NRC is
considering an approach that would expand the applicability of the rule
to all current and future cladding materials, modify the reporting
requirements, and address the issues raised in a petition for
rulemaking (PRM) regarding crud and oxide deposits and hydrogen content
in fuel cladding. With this ANPR, the NRC seeks comment on specific
questions and issues for consideration related to this proposed
conceptual approach to revising the ECCS acceptance criteria.
DATES: Submit comments by October 27, 2009. Comments received after
this date will be considered if it is practical to do so, but the NRC
is only able to ensure consideration of comments received on or before
this date.
ADDRESSES: You may submit comments by any one of the following methods.
Please include the following number RIN 3150-AH42 in the subject line
of your comments. Comments on rulemakings submitted in writing or
electronic form will be made available for public inspection. Because
your comments will not be edited to remove any identifying or contact
information, the NRC cautions you against including any information in
your submissions that you do not want to be publicly disclosed.
We request that any party soliciting or aggregating comments
received from other persons for submission to the NRC inform those
persons that the NRC will not edit their comments to remove any
identifying or contact information, and therefore they should not
include any information in their comments that they do not want
publicly disclosed. All commenters should ensure that sensitive or
Safeguards Information is not contained in their responses or comments
to this ANPR.
Federal e-Rulemaking Portal: Go to http://www.regulations.gov and
search for documents filed under Docket ID NRC-2008-0332. Address
questions about NRC dockets to Carol Gallagher (301) 492-3668; e-mail
[email protected].
E-mail comments to: [email protected]. If you do not
receive a reply e-mail confirming that we have received your comments,
contact us directly at (301) 415-1677.
Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland
20852, between 7:30 am and 4:15 pm during Federal workdays. (Telephone
(301) 415-1677).
Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at
(301) 415-1101. You can access publicly available documents related to
this document using the following methods:
NRC's Public Document Room (PDR): The public may examine and have
copied for a fee, publicly available documents at the NRC's PDR, Public
File Area Room O1-F21, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland. The PDR reproduction contractor will copy
documents for a fee.
NRC's Agencywide Document Access and Management System (ADAMS):
Publicly available documents created or received at the NRC are
available electronically at the NRC's Electronic Reading Room at http://www.nrc.gov/NRC/reading-rm/adams.html. From this page, the public can
gain entry into ADAMS, which provides text and image files of NRC's
public documents. If you do not have access to ADAMS or if there are
any problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at (800) 397-4209, (301) 415-4737, or by e-mail
to [email protected].
FOR FURTHER INFORMATION CONTACT: Barry Miller, Mail Stop O-9E3, Office
of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; telephone (301) 415-4117, or e-mail
[email protected].
SUPPLEMENTARY INFORMATION:
I. Background
In SECY-98-300, ``Options for Risk-Informed Revisions to 10 CFR
part 50--`Domestic Licensing of Production and Utilization Facilities,'
'' dated December 23, 1998 (ADAMS Accession number ML992870048), the
NRC began to explore approaches to risk-informing its regulations for
nuclear power reactors. One alternative (termed ``Option 3'') involved
making risk-informed changes to the specific requirements in the body
of Title 10 of the Code of Federal Regulations (10 CFR) Part 50. As the
NRC began to develop its approach to risk-informing these requirements,
it sought stakeholder input in public meetings. Two of the regulations
identified by industry as potentially benefitting from risk-informed
changes were 10 CFR 50.44 and 10 CFR 50.46. Section 50.44 specifies the
requirements for combustible gas control inside reactor containment
structures and Sec. 50.46 specifies the requirements for light-water
power reactor emergency core cooling systems. For Sec. 50.46, the
potential was identified for making risk-informed changes to
requirements for both ECCS cooling performance and ECCS analysis
acceptance criteria in Sec. 50.46(b).
Additionally, on March 14, 2000, as amended on April 12, 2000, the
Nuclear Energy Institute (NEI) submitted a PRM requesting that the NRC
amend its regulations in Sec. Sec. 50.44 and 50.46 (PRM-50-71). The
NEI petition noted that these two regulations apply to only two
specific zirconium-based fuel cladding alloys (Zircaloy and ZIRLO
\TM\). NEI
[[Page 40766]]
stated that reactor fuel vendors had subsequently developed new
cladding materials other than Zircaloy and ZIRLO \TM\ and that in order
for licensees to use these new materials under the regulations,
licensees had to request NRC approval of exemptions from Sec. Sec.
50.44 and 50.46. On September 16, 2003, (68 FR 54123), the NRC amended
Sec. 50.44 to include new, risk-informed requirements for combustible
gas control. The regulation was also modified to be applicable to all
boiling or pressurized water reactors regardless of the type of fuel
cladding material utilized.
On March 3, 2003, in response to SECY-02-0057, ``Update to SECY-01-
0133, `Fourth Status Report on Study of Risk-Informed Changes to the
Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations
on Risk-Informed Changes to 10 CFR 50.46 (ECCS Acceptance Criteria)'
'', the Commission issued a staff requirements memorandum (SRM) (ADAMS
Accession number ML030910476) directing the NRC staff to move forward
to risk-inform its regulations in a number of specific areas. Among
other things, this SRM directed the NRC staff to modify the ECCS
acceptance criteria to provide for a more performance-based approach to
meeting the ECCS requirements in Sec. 50.46.
Separately from the Commission's efforts to modify its regulations
to provide a more risk-informed, performance-based regulatory approach,
the NRC had also undertaken a fuel cladding research program intended
to investigate the behavior of high exposure fuel cladding under
accident conditions. This research program included an extensive loss-
of-coolant accident (LOCA) research and testing program at Argonne
National Laboratory (ANL), as well as jointly funded programs at the
Kurchatov Institute and the Halden Reactor project, to develop the body
of technical information needed to support the new regulations.
The effects of both alloy composition and fuel burnup (the extent
to which fuel is used in a reactor) on cladding embrittlement (i.e.,
loss of ductility) under accident conditions were studied in this
research program. The research program identified new cladding
embrittlement mechanisms and expanded the NRC's knowledge of previously
identified mechanisms. The research results revealed that alloy
composition has a minor effect on embrittlement, but the cladding
corrosion which occurs as fuel burnup increases has a substantial
effect on embrittlement. One of the major findings of NRC's research
program was that hydrogen, which is absorbed in the cladding during the
burnup-related corrosion process under normal operation, has a
significant influence on the embrittlement during a hypothetical
accident. Increased hydrogen content increases both the solubility of
oxygen in zirconium and the rate at which it is absorbed, thus
increasing the amount of oxygen in the metal during high temperature
oxidation in LOCA conditions. Oxygen is what ultimately causes
embrittlement in zirconium, but hydrogen content is a good indicator of
burnup embrittlement effects because of its ability to allow this
increased oxygen absorption. Because of hydrogen's effect, the
embrittlement thresholds can be correlated with the pre-accident
hydrogen concentration. Further, the NRC's research program found that
oxygen from the oxide fuel pellets enters the cladding from the inner
surface if a bonding layer exists between the fuel pellet and the
cladding, in addition to the oxygen that enters from the oxide layer on
the outside of the cladding. Moreover, under conditions that might
occur during a small-break LOCA [such as an extended time-at-
temperature below 1000 degrees Centigrade ([deg]C) (1832 degrees
Fahrenheit ([deg]F))], the accumulating oxide on the surface of the
cladding can break up; this can allow large amounts of hydrogen to
diffuse into the cladding, thus exacerbating the embrittlement process.
The research results also confirmed an older finding that if
cladding rupture occurs during a LOCA, large amounts of hydrogen
produced from the steam-cladding reaction can enter the cladding inside
surface near the rupture location. These research findings have been
summarized in Research Information Letter (RIL) 0801, ``Technical Basis
for Revision of Embrittlement Criteria in 10 CFR 50.46,'' (ADAMS
Accession number ML081350225) and the detailed experimental results
from the program at ANL are contained in NUREG/CR-6967, ``Cladding
Embrittlement during Postulated Loss-of-Coolant Accidents'' (ADAMS
Accession number ML082130389).
In response to the research findings identified in RIL 0801, the
NRC completed a preliminary safety assessment of currently operating
reactors (ADAMS Accession number ML090340073). This assessment found
that due to realistic fuel rod power history, measured cladding
performance under LOCA conditions, and current analytical
conservatisms, sufficient safety margin exists for operating reactors.
Therefore, any changes to the ECCS acceptance criteria to account for
the new findings can reasonably be addressed through rulemaking.
After the NRC publicly released the technical basis information in
RIL 0801 on May 30, 2008, and NUREG/CR-6967, on July 31, 2008, it
published a Federal Register (FR) document on July 31, 2008, (73 FR
44778), requesting that public stakeholders comment on the adequacy of
the technical basis and identify issues that may arise with respect to
experimental data development, regulatory costs, or impacts of
potential new requirements. The comments received in response to this
document can be found at http://www.regulations.gov by searching on
docket ID NRC-2008-0332. On September 24, 2008, the NRC held a public
workshop to discuss stakeholder comments on the adequacy of the
technical basis and to give the public and industry another opportunity
to provide further comment and input. The workshop included
presentations and open discussion between representatives of the NRC,
international regulatory and research agencies, domestic and
international commercial power firms, fuel vendors, and the general
public. The meeting summary, including a list of attendees and
presentations, is available at ADAMS Accession number ML083010496.
Since 2002, the NRC has met with the Advisory Committee on Reactor
Safeguards (ACRS) multiple times to discuss the progress of the LOCA
research program and rulemaking proposals. Provided in the table below
are the dates and ADAMS Accession numbers of the relevant ACRS meetings
and associated correspondence.
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ADAMS accession
Date Meeting/letter number
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October 9, 2002................ Subcommittee ML023030246 *
Meeting.
October 10, 2002............... Full Committee ML022980190 *
Meeting.
October 17, 2002............... Letter from ACRS to ML022960640
NRC staff.
December 9, 2002............... Response letter ML023260357
from NRC staff to
ACRS.
September 29, 2003............. Subcommittee ML032940296 *
Meeting.
[[Page 40767]]
July 27, 2005.................. Subcommittee ML052230093 *
Meeting.
September 8, 2005.............. Full Committee ML052710235 *
Meeting.
January 19, 2007............... Subcommittee ML070390301 *
Meeting.
February 2, 2007............... Full Committee ML070430485 *
Meeting.
May 23, 2007................... Letter from ACRS to ML071430639
NRC staff.
July 11, 2007.................. Response letter ML071640115
from NRC staff to
ACRS.
December 2, 2008............... Subcommittee ML083520501 * &
Meeting. ML083530449
December 4, 2008............... Full Committee ML083540616 *
Meeting.
December 18, 2008.............. Letter from ACRS to ML083460310
NRC staff.
January 23, 2009............... Response letter ML0836640532
from NRC staff to
ACRS.
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* ADAMS file is a transcript of the ACRS meeting.
On March 15, 2007, Mark Leyse submitted to the NRC a PRM (ADAMS
Accession number ML070871368). In the petition, which was docketed as
PRM 50-84, the petitioner requested that all holders of operating
licenses for nuclear power plants be required to operate such plants at
operating conditions (e.g., levels of power production, and light-water
coolant chemistries) necessary to effectively limit the thickness of
crud \1\ and/or oxide layers on fuel rod cladding surfaces. The
petitioner requested the NRC to conduct rulemaking in the following
three specific areas:
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\1\ Crud is a foreign substance which may be deposited on the
surface of fuel cladding which can impede the transfer of heat. Crud
most frequently refers to deposits of iron or nickel metallic
particles eroded from pipe and valve surfaces. These particles of
stable isotopes may become ``activated'' when they are irradiated in
the reactor and transform into radioactive isotopes such as cobalt-
60. The NRC makes a distinction between crud and pure zirconium
oxidation layers. Although both materials contain metal oxides, crud
does not originate at the fuel rod, while zirconium oxide forms on
fuel cladding when the cladding material reacts with oxygen.
---------------------------------------------------------------------------
(1) Establish regulations that require licensees to operate light-
water power reactors under conditions that are effective in limiting
the thickness of crud and/or oxide layers on zirconium-clad fuel in
order to ensure compliance with Sec. 50.46(b) ECCS acceptance
criteria;
(2) Amend Appendix K to 10 CFR part 50 to explicitly require that
the steady-state temperature distribution and stored energy in the
reactor fuel at the onset of a postulated LOCA be calculated by
factoring in the role that the thermal resistance of crud deposits and/
or oxide layers plays in increasing the stored energy in the fuel
(these requirements also need to apply to any NRC-approved, best-
estimate ECCS evaluation models used in lieu of Appendix K to Part 50,
calculations); and
(3) Amend Sec. 50.46 to specify a maximum allowable percentage of
hydrogen content in [fuel rod] cladding.
On May 23, 2007, (72 FR 28902), the NRC published a notice of
receipt for this petition in the FR and requested public comment on the
petition. The public comment period ended on August 6, 2007. After
evaluating the public comments, the NRC resolved the Leyse petition by
deciding that each of the petitioner's issues should be considered in
the rulemaking process. The NRC's determination was published in the FR
on November 25, 2008, (73 FR 71564).
Because the issues raised in PRM-50-84 pertain to ECCS analysis and
acceptance criteria, the need for rulemaking to address the
petitioner's technical concerns will be addressed in this rulemaking.
Technical details associated with the NRC's evaluation of the
rulemaking requests in PRM-50-84 are discussed in Section III.4 of this
document.
II. Rulemaking Objectives
The scope of the rulemaking contemplated by this ANPR includes four
separate rulemaking objectives:
Objective 1: Expand the applicability of Sec. 50.46 to include any
light-water reactor fuel cladding material:
In this rulemaking, the NRC is considering expansion of the rule's
applicability (which currently addresses only Zircaloy and
ZIRLOTM cladding) to include any light-water reactor fuel
cladding material. As used in this ANPR, the term ``fuel cladding'' (or
simply ``cladding'') refers only to the cylindrical material that
surrounds and contains the nuclear fuel, not a fuel/cladding system.
The rulemaking may clarify the general applicability of Sec. 50.46 to
require that all light-water nuclear power reactors must be provided
with an ECCS designed so that after a postulated LOCA, a coolable core
geometry would be maintained, excessive combustible gases would not be
generated, and long-term cooling would be assured. The applicability
expansion would also encompass the request in PRM-50-71, filed by NEI
(see 65 FR 34599; May 31, 2000, and 73 FR 6600; November 6, 2008), to
establish requirements that apply to all zirconium-based cladding
alloys, including current and anticipated alloys. The NRC's high-burnup
fuel research program investigated cladding embrittlement in a number
of different zirconium-based cladding alloys and concluded that the
results were applicable equally to all of the zirconium-based alloys.
Therefore, new zirconium-specific criteria can be formulated in a
performance-based manner that would satisfy the request in PRM-50-71.
Because this applicability expansion may also aim to encompass any
potential new cladding materials developed in the future that are not
zirconium-based, the NRC notes that such materials would still need an
extensive technical foundation to receive NRC approval. However, this
applicability expansion would eliminate the need for licensees to
request, and the NRC to review and approve, exemptions from Sec. 50.46
for these potential new non-zirconium cladding materials.
Objective 2: Establish performance-based requirements and
acceptance criteria specific to zirconium-based cladding materials that
reflect recent research findings:
The second objective of this rulemaking is to enhance the
performance-based features of Sec. 50.46 by replacing the current
Sec. 50.46(b) prescriptive analytical limits with fuel cladding
performance requirements and acceptance criteria. These performance
requirements, based upon the recent findings from the NRC's high burnup
research program, would ensure that an adequate level of cladding
ductility is maintained throughout a postulated LOCA.
Objective 3: Revise the LOCA reporting requirements:
The third objective of this rulemaking is to amend Sec.
50.46(a)(3)(i) to emphasize the importance of reporting reduction in
margins to the acceptance criteria and
[[Page 40768]]
the periodic reporting of susceptibility to breakaway oxidation.
Objective 4: Address the issues raised in PRM-50-84, which relate
to crud deposits and hydrogen content in fuel cladding:
The fourth objective of this rulemaking is to amend Sec. 50.46 as
necessary to address the technical issues on which the PRM-50-84
petitioner's three requests for rulemaking are based. The need for and
extent of any changes that may be needed to address these issues will
be determined during this rulemaking.
III. Specific Proposals
The NRC presents the following conceptual approach to revising 10
CFR 50.46 under the outlined objectives:
Objective 1: Expand the applicability of Sec. 50.46 to include any
light-water reactor fuel cladding material:
This first conceptual approach involves the applicability of the
rule as defined in Sec. 50.46(a)(1)(i). Currently, this provision is
limited to fuel rods clad in Zircaloy or ZIRLO\TM\. The recent LOCA
research program conducted testing on a wide range of zirconium-based
alloys such that research findings and future testing requirements are
believed to be applicable to all zirconium-based alloys. Therefore, the
NRC intends to expand the applicability of the rule to all zirconium-
based alloys. This would allow the introduction of future, advanced
zirconium-based alloys without the need for exemption requests.
However, NRC approval would still be required.
In addition, the NRC is considering further expansion of the rule's
applicability to include all light-water reactors (LWRs) without regard
to the type of fuel cladding material utilized in the design.
Currently, Sec. 50.46 states that the ECCS must be designed so that
its calculated cooling performance following postulated LOCAs conforms
to the five criteria set forth in Sec. 50.46(b). To accomplish such a
change, the NRC is considering an approach where the proposed revision
would specify that all fuel cladding material used in LWRs, without
regard to its composition, must satisfy the three general conditions
which currently exist as the criteria specified in Sec. 50.46(b)(3)
Maximum hydrogen generation, Sec. 50.46(b)(4) Coolable geometry, and
Sec. 50.46(b)(5) Long-term cooling. The Sec. 50.46(b)(3) criterion
would be modified to limit generation of any combustible gas, rather
than just hydrogen, with recognition that different cladding materials
could potentially react to produce different combustible gases. Because
the NRC's recent research findings are only applicable to LWRs with
zirconium-based cladding alloys, detailed ECCS acceptance criteria for
different cladding materials could not now be specified in the
regulations. Therefore, the NRC is considering a cladding-specific
regulatory approach that would require applicants with non-zirconium
cladding materials to propose specific detailed criteria to demonstrate
how coolable core geometry, long-term cooling, and minimal generation
of combustible gases would be ensured. In order to develop such
cladding-specific criteria, applicants would need to fully develop and
understand all of the material's degradation mechanisms, chemical and
physical properties, and any other characteristics that may affect its
behavior in the core during normal operation and under LOCA conditions.
The NRC would review the applicant's proposed criteria and issue its
approval only if the criteria ensure that the three general conditions
are met, that the cladding-specific criteria can be demonstrated to be
met during all credible LOCA scenarios, and that they are sufficient to
ensure adequate protection of public health and safety. Section IV of
this document requests comment on this conceptual approach to expanding
the rule's applicability.
For LWRs using zirconium-based alloys, cladding-specific criteria
can and will be specified in the regulations based on the results of
the NRC's LOCA research program. These criteria will ensure adequate
cladding ductility is maintained via specified performance
requirements. A general discussion on the nature of these criteria is
provided below under Objective 2.
Objective 2: Establish performance-based requirements and
acceptance criteria specific to zirconium-based cladding materials that
reflect recent research findings:
Cladding Ductility
In the current rule, the preservation of cladding ductility, via
compliance with regulatory criteria on peak cladding temperature (Sec.
50.46(b)(1)) and local cladding oxidation (Sec. 50.46(b)(2)), ensures
that the core remains amenable to cooling. The recent LOCA research
program identified new cladding embrittlement mechanisms which
demonstrated that the current combination of peak cladding temperature
(2200 [deg]F (1204 [deg]C)) and local cladding oxidation (17 percent
equivalent cladding reacted (ECR)) criteria do not always ensure post
quench ductility (PQD). It is important to recognize that the loss of
cladding ductility is the result of oxygen diffusion into the base
metal and not directly related to the growth of a zirconium dioxide
layer on the cladding outside diameter. In the current provision, the
peak local oxidation limit is used as a surrogate to limit time at
elevated temperature and associated oxygen diffusion. This surrogate
approach is possible because both oxidation and diffusion share a
strong temperature dependence. In the recent LOCA research program, the
Cathcart-Pawel (CP) weight gain correlation was used to quantify the
time at elevated temperature at which ductility was lost (nil
ductility). For this reason, the proposed amendment would include a
requirement that local cladding oxidation (which is being used as a
surrogate for limiting time-at-temperature) be calculated using the
same Cathcart-Pawel correlation (see Regulatory Guide 1.157 regarding
use of the Cathcart-Pawel oxidation correlation rather than the Baker-
Just correlation cited in 10 CFR part 50, Appendix K, Part I.A.5).
To enhance the performance-based aspects of Sec. 50.46 (and
achieve an objective of this rulemaking), the limits on peak cladding
temperature and local oxidation would be replaced with specific
cladding performance requirements and acceptance criteria which ensure
that an adequate level of cladding ductility is maintained throughout
the postulated LOCA. For example, the rule may specify that retention
of cladding ductility is defined as the accumulation of >= 1.00 percent
permanent strain prior to failure during ring-compression loading at a
temperature of 135 [deg]C and a displacement rate of 0.033 millimeters
per second (mm/sec). Section IV of this document requests comment on
alternative ways to define an acceptable measure of ductility. This
acceptance criterion would be used to define analytical limits for peak
cladding temperature and local oxidation based on cladding performance
during tests in which cladding specimens are exposed to double-sided
steam oxidation up to a specified peak oxidation temperature and CP-
ECR. Analytical limits would be calculated as a function of initial
cladding hydrogen content (weight parts per million (wppm) in metal).
The NRC intends to issue a regulatory guide detailing an acceptable
experimental test methodology for defining analytical limits in
accordance with these performance requirements. Included in this test
methodology would be guidance for treating ring-compression test
results which fail in such a way that permanent strain cannot be
measured. The guidance would provide a
[[Page 40769]]
relationship of permanent strain to offset-displacement.
This ANPR also provides two possible approaches for determining the
acceptability of current and future cladding alloys in accordance with
the proposed performance requirements. Two approaches are described as
follows, however the NRC recognizes there may be other alternatives.
Approach A--Analytical Limits Defined Within Regulatory Guidance:
The focal point of this approach would be a future regulatory
guidance document which defines an acceptable, generically-applicable
set of analytical limits for peak cladding temperature and maximum
allowable time-at-temperature (expressed as calculated local oxidation,
CP-ECR) as a function of pre-transient hydrogen content in the cladding
metal, excluding hydrogen in the cladding oxide layer. These acceptable
analytical limits would be based on the results of NRC's LOCA research
program. Appendix A of this document outlines the conceptual path for
approving both current and future cladding alloys using this approach.
Approach B--Cladding-Specific Analytical Limits Defined by an
Applicant:
The second approach involves establishing cladding-specific and/or
temperature-specific analytical limits for peak cladding temperature
and maximum allowable time-at-temperature (expressed as calculated
local oxidation, CP-ECR) as a function of pre-transient hydrogen
content in the cladding metal, excluding hydrogen in the cladding oxide
layer. This approach would provide optimum flexibility for defining
more specific analytical limits to gain margin to the ECCS performance
criteria. However, unlike citing analytical limits within a regulatory
guide, this approach places the burden of proof on the applicant to
validate their analytical limits and address experimental variability
and repeatability. As a result, this approach would necessitate a
larger number of PQD tests (relative to confirming the applicability of
the regulatory guide). Analytical limits, along with the experimental
procedures, protocols, and specimen test results used in their
development, would be subject to NRC review and approval. Appendix B of
this document includes further discussion to illustrate the possible
implementation of this approach.
Cladding embrittlement is highly sensitive to both hydrogen content
and peak oxidation temperature, and this relationship is applicable to
both approaches. The discussion in the Appendices to this document
describes an approach that would demonstrate compliance with the
proposed change and illustrate this relationship.
Implementing any hydrogen based analytical limits, similar to the
descriptions contained in the Appendices, requires an accurate, alloy-
specific hydrogen uptake model. Section IV of this document seeks
comment on the development of these models and how best to deal with
the axial, radial, and circumferential variability in hydrogen
concentration.
Two-Sided Oxidation
Prompted by research which found that oxygen from the inside
diameter fuel bonding layer present in high burnup fuel rods may
diffuse into the base metal of the cladding, the NRC is proposing a new
analytical requirement to specifically account for the potential
diffusion of oxygen from the cladding inside diameter. Because the
formation of a fuel bonding layer may depend on fuel rod design and
power history, licensees would be required to develop and justify a
burnup threshold above which this phenomenon would be specifically
accounted for within local cladding oxidation calculations.
Breakaway Oxidation
The NRC may also propose new requirements addressing breakaway
oxidation. The recent LOCA research program discovered that the
protective cladding oxide layer will undergo a phase transformation,
become unstable, and allow for the uptake of hydrogen into the base
metal. The timing of this transformation is sensitive to many
parameters including the cladding manufacturing process. Licensees
would be responsible for ensuring that the timing of the oxide phase
transformation is measured for each cladding alloy utilized in their
core to determine susceptibility to early breakaway oxidation. The
proposed rule would specify the required testing method, along with an
acceptable measure of breakaway oxidation behavior. The NRC intends to
issue a regulatory guide detailing an acceptable experimental
methodology for defining new criteria under these requirements. For
example, the proposed rule may specify that the minimum measured time
until the onset of breakaway oxidation, defined as when hydrogen uptake
reaches 200 wppm anywhere on a cladding segment subjected to high
temperature steam oxidation ranging from 1200 [deg]F to 1875 [deg]F
(649 [deg]C to 1024 [deg]C), shall remain greater than the calculated
duration that cladding surface temperature anywhere on the fuel rod
remains above 1200 [deg]F (649 [deg]C).
The measured timing of the oxide phase transformation for each
cladding alloy, along with the experimental procedures and protocols
used in their development, would be subject to NRC review and approval.
Section IV of this document seeks public comment on a draft
experimental methodology for conducting breakaway oxidation testing
with zirconium-based cladding alloys.
Application of the proposed breakaway oxidation criterion would
involve new analytical requirements, including an additional break
spectrum analysis to identify the limiting combination of inputs that
maximize the time above elevated temperatures which are susceptible to
breakaway oxidation for the given cladding alloy (e.g., 1200 [deg]F
(649 [deg]C)). Each licensee would be required to demonstrate that this
calculated duration remained below the measured minimum time to
breakaway oxidation. As an alternative, the NRC is considering tying
breakaway oxidation to the rule's applicability statement. For example,
the proposed revision would only be applicable to zirconium-based
alloys which do not experience the breakaway phenomena within a
specified time period. This approach would eliminate the need for each
licensee to perform and maintain a current updated final safety
analysis report (UFSAR) break spectrum analysis for breakaway
oxidation. To set the specified time period within the proposed rule's
applicability statement, the NRC is seeking information related to the
maximum time span with cladding surface temperature above 1200 [deg]F
(649 [deg]C) for the full range of piping break sizes and nuclear steam
supply system (NSSS)/ECCS design combinations. If successful, this
alternative approach would include a simpler pass/fail breakaway
testing requirement up to this specified time period (as opposed to
searching for and quantifying the limiting time to breakaway). Section
IV of this document seeks to obtain this input.
Objective 3: Revise the LOCA reporting requirements.
Redefining a Significant Change or Error:
The reporting requirement in 10 CFR 50.46(a)(3)(i) currently
defines a significant change or error as one that results in a
calculated peak cladding temperature (PCT) different by more than 50
[deg]F (28 [deg]C) from the temperature calculated for the limiting
transient using the last acceptable model, or is a cumulation of
changes and errors such that the sum of the absolute magnitudes of the
respective temperature changes is greater than 50 [deg]F (28 [deg]C).
[[Page 40770]]
The NRC is considering revising the reporting requirements by
redefining what constitutes a significant change or error in such a
manner as to make the reporting requirements dependent upon the margin
between the acceptance criteria limits and the calculated values of the
respective parameters (i.e., PCT or CP-ECR). The redefinition would aim
to capture the importance of being close to the limits by making
reporting of a change dependent upon the margin to the acceptance
criteria. The NRC believes this redefinition should also expand the
current reporting scope to include CP-ECR, in addition to PCT, as a
parameter required for reporting. The timeliness requirements for
reporting would remain the same (i.e., 30 days for a significant change
or error). The following definitions exemplify a specific approach the
NRC is considering:
If the calculated parameter (PCT or CP-ECR) has margin greater than
5 percent of its acceptance criterion limit, then a significant change
or error is one that results in:
(i) A PCT change of 100 [deg]F (56 [deg]C) or greater,
(ii) A CP-ECR change of 2 percent or greater, or
(iii) An accumulation of changes and errors such that the sum of
the absolute magnitudes of the changes and errors is greater than 100
[deg]F (56 [deg]C) or 2 percent, respectively.
If the calculated parameter (PCT or CP-ECR) is within 5 percent of
its acceptance criterion limit, then a significant change or error is
one that results in a calculated 10 percent or greater reduction in the
remaining margin.
The following table gives an example for how the PCT criterion
reporting would be ``triggered'' for a plant with a PCT limit of 2200
[deg]F.
------------------------------------------------------------------------
Calculated PCT Reporting trigger
------------------------------------------------------------------------
< 2090 (i.e., not within 5 percent of 2200 Any change >= 100 [deg]F.
[deg]F limit).
2090-2099 [deg]F.......................... Any change >= 11 [deg]F.
2100-2109 [deg]F.......................... Any change >= 10 [deg]F.
2110-2119 [deg]F.......................... Any change >= 9 [deg]F.
2120-2129 [deg]F.......................... Any change >= 8 [deg]F.
2130-2139 [deg]F.......................... Any change >= 7 [deg]F.
2140-2149 [deg]F.......................... Any change >= 6 [deg]F.
2150-2159 [deg]F.......................... Any change >= 5 [deg]F.
2160-2169 [deg]F.......................... Any change >= 4 [deg]F.
2170-2179 [deg]F.......................... Any change >= 3 [deg]F.
2180-2189 [deg]F.......................... Any change >= 2 [deg]F.
2190-2199 [deg]F.......................... Any change >= 1 [deg]F.
------------------------------------------------------------------------
The NRC recognizes that there are other possible approaches for
implementing the concept that the reporting obligation depends upon the
margin to the relevant acceptance criteria. Section IV of this document
seeks specific comment on this approach to modifying the reporting
requirements.
Breakaway Oxidation Susceptibility Reporting
The NRC is also considering reporting requirements related to
breakaway oxidation. Different zirconium-based alloys have varying
susceptibility to breakaway oxidation that is dependent on factors such
as alloy content, manufacturing process, and surface preparation, among
others. The NRC is concerned that during the life-cycle of an alloy
used by a fuel vendor, both intentional and unintentional changes may
be made in the aforementioned conditions. The effect of the changes can
only be determined by testing samples throughout the life-cycle of an
alloy of the current cladding material for breakaway oxidation
potential. The NRC plans to propose to include periodic testing of
cladding samples as part of the annual licensee report pertaining to
the LOCA licensing basis. The new requirement would be consistent with
the following concept: licensees would report to the NRC at least
annually as specified in Sec. Sec. 50.4 or 52.3, as applicable,
results of testing of each type of zirconium-based cladding alloy
employed in their reactor core for susceptibility to breakaway
oxidation. If a cladding alloy is found to have greater susceptibility
to breakaway oxidation than would be acceptable for the corresponding
time-at-temperature of the ECCS performance analysis, the affected
licensee would be required to propose immediate steps to reduce the
impact of breakaway oxidation on their ECCS performance analysis.
Section IV of this document seeks specific comment on this approach to
modifying the reporting requirements.
Objective 4: Address the issues raised in PRM-50-84, which relate
to crud deposits and hydrogen content in fuel cladding:
In this ANPR, the NRC addresses the three requests for rulemaking
in PRM-50-84:
(1) Establish regulations that require licensees to operate light-
water power reactors under conditions that are effective in limiting
the thickness of crud and/or oxide layers on zirconium-clad fuel in
order to ensure compliance with Sec. 50.46(b) ECCS acceptance
criteria;
(2) Amend Appendix K to 10 CFR part 50 to explicitly require that
the steady-state temperature distribution and stored energy in the
reactor fuel at the onset of a postulated LOCA be calculated by
factoring in the role that the thermal resistance of crud deposits and/
or oxide layers plays in increasing the stored energy in the fuel
(these requirements also need to apply to any NRC-approved, best-
estimate ECCS evaluation models used in lieu of Appendix K to part 50,
calculations); and
(3) Amend Sec. 50.46 to specify a maximum allowable percentage of
hydrogen content in [fuel rod] cladding.
PRM-50-84 Rulemaking Requests 1 and 2
Because the petitioner's first two requests for rulemaking are
technically related, they are addressed together in the following
discussion. When evaluating PRM-50-84, the NRC reviewed the technical
information provided by the petitioner and by all public commenters.
The NRC's detailed analysis of all public comments was published in the
FR on November 25, 2008 (73 FR 71564). A summary of key comments that
influenced the NRC's conclusions follows.
The NEI opposed granting PRM-50-84 because the petition relies
heavily on atypical operating experiences at four plants: River Bend
(1998-1999 and 2001-2003), Three Mile Island Unit 1 (1995), Palo Verde
Unit 2 (1997), and Seabrook (1997), where thick crud layers developed
during normal operation. NEI stated that the incidents cited by the
petitioner were isolated operational events and would not have been
prevented by imposing specific regulatory limits on crud thickness. NEI
noted that the industry is actively pursuing root cause evaluations and
has developed corrective actions to mitigate further cases of excessive
crud formation.
NEI also stated that reactor licensees use approved fuel
performance models to determine fuel rod conditions at the start of a
LOCA. NEI stated that the impact of crud and oxidation on fuel
temperatures and pressures may be determined explicitly or implicitly
in the system of models used. NEI referenced the NRC review guidance in
the Standard Review Plan (SRP) (NUREG-0800) noting that SRP Section 4.2
states that the impact of corrosion on thermal and mechanical
performance should be considered in the fuel design analysis, when
comparing to the design stress and strain limits. NEI and industry
commenters in general opposed issuing new regulations related to crud,
stating that the existing regulations and voluntary guidance regarding
crud are sufficient.
The NRC agrees with NEI that new requirements imposing specific
[[Page 40771]]
regulatory limits on crud thickness would not necessarily have
prevented the occurrences of heavy crud deposits that were the
unexpected consequences of the operational events cited in PRM-50-84.
Nevertheless, formation of cladding crud and oxide layers is an
expected condition at nuclear power plants. Although the thickness of
these layers is usually limited, the amount of accumulated crud and
oxidation varies from plant to plant and from one fuel cycle to
another. Intended or inadvertent changes to plant operational practices
may result in unanticipated levels of crud deposition. The NRC agrees
with the petitioner that crud and/or oxide layers may directly increase
the stored energy in reactor fuel by increasing the thermal resistance
of cladding-to-coolant heat transfer, and may also indirectly increase
the stored energy through an increase in the fuel rod internal
pressure.
As previously discussed, NEI commented that reactor licensees use
approved fuel performance models to determine fuel rod conditions at
the start of a LOCA and that the impact of crud and oxidation on fuel
temperatures and pressures may be determined explicitly or implicitly
by the system of models used. The NRC believes that to accurately model
fuel performance during normal and postulated accident conditions, it
is essential that fuel performance and LOCA evaluation models include
the thermal effects of both crud and oxidation whenever their
accumulation changes the calculated results. Recently, power reactor
licensees have been submitting an increased number of license amendment
applications requesting significant increases in licensed power levels.
In some cases, these increases have reduced the margin between
calculated ECCS performance and current ECCS acceptance criteria. This
trend further supports the need to ensure that the effects of both crud
and oxidation are properly accounted for in ECCS analyses. The
technical concerns related to the thermal effects of oxidation and crud
raised by the petitioner's rulemaking requests are addressed separately
below.
Oxidation. The accumulation of cladding oxidation and its
associated effects on fuel cladding acceptance criteria are being
addressed by the ongoing work to revise the ECCS acceptance criteria.
Thus, the concerns related to oxidation raised by the petitioner's
rulemaking requests are encompassed by Objective 2 of this section.
Crud. 10 CFR 50.46 requires the licensee of a facility to perform
LOCA accident analyses to demonstrate that a nuclear reactor has an
ECCS that is designed so its calculated performance meets the
acceptance criteria in Sec. 50.46(b) on peak clad temperature (2200
[deg]F) and maximum local oxidation (17 percent). Licensees must
evaluate a plant's ECCS by calculating its performance with an
acceptable evaluation model. An acceptable model is one that either
complies with the required and acceptable features in Appendix K to
Part 50--ECCS Evaluation Models; or, for best-estimate models, complies
with the Sec. 50.46(a)(1)(i) requirement that there is a high level of
probability that the calculated cooling performance will not exceed the
acceptance criteria in Sec. 50.46(b). The NRC reviews and approves all
licensee evaluation models to determine if they are acceptable.
For best-estimate evaluation models, Sec. 50.46(a)(1)(i) requires
that ``The evaluation model must include sufficient supporting
justification to show that the analytical technique realistically
describes the behavior of the reactor coolant system during a loss-of-
coolant accident.'' For Appendix K models, section I.B. of Appendix K
to Part 50 states, ``The calculations of fuel and cladding temperatures
as a function on time shall use values for gap conductance and other
thermal parameters as functions of temperature and other applicable
time-dependent variables.'' Crud accumulation and its effects are not
explicitly identified as required parameters to be included in best-
estimate or Appendix K to Part 50 models.
However, based on these requirements, the NRC has prepared
regulatory review guidance that addresses the accumulation of crud and
oxidation deposits on fuel cladding surfaces. This guidance is in the
format of review criteria in NUREG-0800, ``Standard Review Plan (SRP)''
which are used by the NRC staff to review licensees' evaluation models.
SRP Section 4.2, ``Fuel System Design,'' Section 4.3, ``Nuclear
Design,'' and Section 4.4, ``Thermal and Hydraulic Design'' all contain
specific criteria related to the accumulation of crud and oxidation on
fuel cladding surfaces. For example, on page 4.2-6 of SRP Section
4.2.2, fuel system damage acceptance criterion iv. states:
iv. Oxidation, hydriding, and the buildup of corrosion products
(crud) should be limited, with a limit specified for each fuel
system component. These limits should be established based on
mechanical testing to demonstrate that each component maintains
acceptable strength and ductility. The safety analysis report should
discuss allowable oxidation, hydriding, and crud levels and
demonstrate their acceptability. These levels should be presumed to
exist in items (i) and (ii) above. The effect of crud on thermal
hydraulic considerations and neutronic (AOA) \2\ considerations are
reviewed as described in SRP Sections 4.3 and 4.4.
---------------------------------------------------------------------------
\2\ AOA means Axial Offset Anomaly.
---------------------------------------------------------------------------
Page 4.2-15 of SRP Section 4.2 also states that the calculational
models used to determine fuel temperature and stored energy should
include phenomenological models addressing ``Thermal conductivity of
the fuel, cladding, cladding crud and oxidation layers'' and ``Cladding
oxide and crud layer thickness.'' Review criteria in SRP Section 4.4
specifically note that the thickness of oxidation layers and crud
deposits must be accounted for in critical heat flux calculations and
when determining the pressure drop throughout the reactor coolant
system.
The NRC review guidance in the SRP supports interpreting Sec.
50.46(a) and Appendix K to Part 50 to include crud as a required
parameter in these analyses. However, because crud is not explicitly
identified in the regulations and the regulatory guidance in the SRP is
not an enforceable requirement, there is ambiguity in the current
requirements. The NRC is considering amending its regulations to
explicitly identify crud as one of the parameters that must be
addressed in ECCS analysis models. This change would eliminate any
ambiguity between the current rule language and the current SRP review
guidance. Licensee evaluation models could be formulated to calculate
the accumulation of crud or assume an expected maximum thickness. The
resulting effects on fuel temperatures would be determined based on the
predicted or assumed thickness of deposits.
The NRC also notes that licensees are required to operate their
facilities within the boundaries of the calculated ECCS performance.
During or immediately after plant operation, if actual crud layers on
reactor fuel are implicitly determined or visually observed after
shutdown to be greater than the levels predicted by or assumed in the
evaluation model, licensees would be required to determine the effects
of the increased crud on the calculated ECCS results. In many cases,
engineering judgment or simple calculations could be used to evaluate
the effects of increased crud levels; therefore, detailed LOCA
reanalysis may not be required. In other cases, new analyses would be
performed to determine the effect the new crud
[[Page 40772]]
conditions have on the final calculated results.
The NRC would consider the deposition of a previously unanalyzed
amount of crud to be the same as making a change to or finding an error
in an approved evaluation model or in the application of such a model.
In these cases, Sec. 50.46(a)(3)(i) requires licensees to determine if
the change or error is significant. For significant changes, Sec.
50.46(a)(3)(ii) requires licensees to provide, within 30 days, a report
to the NRC including a schedule for providing a reanalysis or taking
other action as may be needed to show compliance with the Sec. 50.46
requirements. In situations when the Sec. 50.46(b) acceptance criteria
are not exceeded, the licensee could either change the ECCS analysis of
record to conform to the new crud level or make changes to plant design
or operation (e.g., adjust water coolant chemistry) to reduce crud
deposits to the level assumed in the original analysis. Situations
where a model change or error correction results in calculated ECCS
performance that does not conform to the acceptance criteria in Sec.
50.46(b) would be reportable events as described in Sec. Sec.
50.55(e), 50.72, and 50.73. In these situations, the licensee would be
required under Sec. 50.46(a)(3)(ii) to propose immediate steps to
demonstrate compliance or bring the plant design or operation into
compliance with Sec. 50.46 requirements.
In summary, to address the technical concerns related to crud in
the PRM-50-84 petitioner's requests for rulemaking, the NRC is
considering amending Sec. 50.46(a) to specifically identify crud as a
parameter to be considered in best-estimate and Appendix K to Part 50
ECCS evaluation models. Compliance with this requirement during plant
operation would be determined by the process outlined in the scenarios
above.
Under this approach, the NRC would propose new rule language
defining crud as a foreign substance (other than zirconium oxide) which
may be deposited on the surface of fuel cladding and which impedes the
transfer of heat due to thermal resistance and/or flow area reduction.
A requirement would be added stating that ECCS evaluation models must
consider the effects of crud deposition on fuel cladding at the highest
level of buildup expected during a fuel cycle. In addition, to ensure
that plant-specific crud levels are bounded by the levels analyzed in
the ECCS model, the NRC is considering adding a requirement that
licensees inspect one or more fuel assemblies every fuel cycle to
determine the actual thickness of crud on the fuel. Section IV of this
document requests comment on the potential addition of such a
requirement.
PRM-50-84 Rulemaking Request 3
The petitioner's third request for rulemaking--that the NRC amend
Sec. 50.46 to specify a maximum allowable percentage of hydrogen
content in cladding--pertains to the effects on fuel cladding
embrittlement caused by hydrogen in the cladding. The cladding
embrittlement issue will be technically resolved by revising the ECCS
analysis embrittlement acceptance criteria under rulemaking Objective
2. These new acceptance criteria will address the embrittlement effects
of cladding hydrogen content and other pertinent variables.
IV. Issues for Consideration
Based on the specific proposals and discussion above, the NRC
requests comment on the following questions and issues. In submitting
comments, the NRC asks that each comment be referenced to its
corresponding question or issue number, as indicated below.
Applicability Considerations
1. Objective 1 describes a conceptual approach to expanding the
applicability of Sec. 50.46 to all fuel cladding materials. Should the
rule be expanded to include any cladding material, or only be expanded
to include all zirconium-based cladding alloys? The NRC also requests
comment on the potential advantages and disadvantages of the specific
approach described that would expand the applicability beyond
zirconium-based alloys. Is there a better approach that could achieve
the same objective?
2. The rulemaking objectives do not include expanding the
applicability of Sec. 50.46 to include fuel other than uranium oxide
fuel (UO2). Is there any need for, or available information
to justify, expanding the applicability of this rule to mixed oxide
fuel rods?
New Embrittlement Criteria Considerations
3. The NRC requests information related to the maximum time span
with cladding surface temperature above 1200 [deg]F (649 [deg]C) for
the full range of piping break sizes and NSSS/ECCS design combinations.
This information may be used to set a specified minimum time to
breakaway in the proposed rule's applicability statement.
4. The NRC requests comment on the two approaches to establishing
analytical limits for cladding alloys, as described in Section III.2 of
this document and expanded upon in the Appendices, where limits on peak
cladding temperature and local oxidation would be replaced with
specific cladding performance requirements that define an adequate
level of ductility which must be maintained throughout a postulated
LOCA. In addition to general comments on these approaches, the NRC also
seeks specific comment on the following related items:
a. The NRC requests any further PQD ring-compression test data that
may be available to expand the empirical database as shown in Appendix
A of this document.
b. Because no cladding segments tested in the NRC's LOCA research
program exhibited an acceptable level of ductility beyond a hydrogen
concentration of 550 wppm (metal), analytical limits may be restricted
to terminate at this point. Are any further PQD ring-compression test
data available at hydrogen concentrations beyond 550 wppm which
exhibited an acceptable level of ductility?
c. Ring-compression tests conducted on cladding segments with
identical hydrogen concentrations oxidized to the same CP-ECR often
exhibited a range of measured offset displacement. The variability,
repeatability, and statistical treatment of these test results must be
evaluated for defining generic PQD analytical limits. The NRC requests
comments on the variability, repeatability, and statistical treatment
of ductility measurements from samples exposed to high-temperature
steam oxidation.
5. Implementation of a hydrogen-dependent PQD criterion requires an
NRC-approved hydrogen uptake model. The sensitivity of hydrogen pickup
fraction to external factors (e.g., manufacturing process, proximity to
dissimilar metals, plant coolant chemistry, oxide thickness, crud,
burnup, etc.) must be properly calibrated in the development and
validation of this model.
a. The NRC requests information on the size and depth of the
current hot-cell hydrogen database(s) and the industry's ability to
segregate the sensitivity of each cladding alloy to each external
factor and to quantify the level of uncertainty.
b. Pre-test characterization of some irradiated cladding segments
revealed significant variability in axial, radial, and circumferential
hydrogen concentrations.
i. What information exists that could quantify this asymmetric
distribution in the development of a hydrogen uptake model?
[[Page 40773]]
ii. What information exists that could inform the treatment of this
asymmetric hydrogen distribution as a function of fuel rod burnup?
iii. This asymmetric hydrogen distribution could be addressed in
future PQD ring compression tests on irradiated material by such
requirements as orienting ring samples such that the maximum asymmetric
hydrogen concentration is aligned with the maximum stress point or in
pre-hydrided material by introducing asymmetric distribution during
hydriding. The NRC requests comment on these or other methods to treat
asymmetric hydrogen distribution.
Testing Considerations
6. A draft proposed cladding oxidation and PQD testing methodology
is provided at ADAMS Accession number ML090900841.
a. The NRC requests comment on the details of the draft
experimental methodology, including sample preparation and
characterization, experimental protocols, laboratory techniques, sample
size, statistical treatment, and data reporting.
b. The NRC requests information on any ongoing or planned testing
programs that could exercise the draft experimental methodology to
independently confirm its adequacy.
c. Unirradiated cladding specimens pre-charged with hydrogen appear
to be viable surrogates for testing on irradiated cladding segments.
However, the NRC's position remains that future testing to support
cladding approval reviews include irradiated material without further
confirmatory work to directly compare the embrittlement behavior of
irradiated material to hydrogen pre-charged material at the same
hydrogen level. The NRC's LOCA research program reports PQD test
results on twenty irradiated fuel cladding segments of varying
zirconium alloys and hydrogen concentrations that underwent quench
cooling. The NRC requests information on any ongoing or planned testing
aimed at replicating these twenty PQD tests for the purpose of
validating a pre-hydrided surrogate.
d. The NRC is considering defining an acceptable measure of
cladding ductility as the accumulation of =1.00 percent
permanent strain prior to failure during ring-compression loading at a
temperature of 135 [deg]C and a displacement rate of 0.033 mm/sec.
Recognizing the difficulty of measuring permanent strain, the NRC
requests comment on alternative regulatory criteria defining an
acceptable measure of cladding ductility.
7. The proposed revisions to Sec. 50.46 include a new testing
requirement related to breakaway oxidation. Due to the observed effects
of manufacturing controlled parameters (e.g., surface roughness, minor
alloying, etc.) on the breakaway phenomena, the proposed approach would
include periodic testing requirements to ensure that both planned and
unplanned changes in manufacturing processes do not adversely affect
the performance of the cladding under LOCA conditions.
a. The NRC requests comment on the testing frequency and sample
size provided in the breakaway oxidation testing methodology (ADAMS
Accession number ML090840258) and technical basis for the proposed
breakaway oxidation testing requirement.
b. Is there any ongoing or planned testing to further understand
the sensitivity of breakaway oxidation to parameters controlled during
the manufacturing process?
Revised Reporting Requirements Considerations
8. The NRC requests comment on the proposed concept that the
reporting obligation in Sec. 50.46 depend upon the margin to the
relevant acceptance criteria. Please also comment on the specific
approach to implement this objective as described under Objective 3 in
Section III of this document.
9. The NRC requests comment on the proposed concept of adding the
results of breakaway oxidation susceptibility testing to the annual
reporting requirement. Are there other implementation approaches that
could help ensure that a zirconium-based alloy does not become more
susceptible to breakaway during its manufacturing and production life-
cycle?
Crud Analysis Considerations
10. The NRC requests comment on the proposed regulatory approach in
which crud is required to be considered in ECCS evaluation models. If
actual crud levels should exceed the levels considered in the
evaluation model, the situation would be considered equivalent to
discovering an error in the ECCS model. The licensee would then be
subject to the reporting and corrective action process specified in
Sec. 50.46(a)(3) to resolve the discrepancy. The NRC also requests
comment on the imposition of a requirement that one or more fuel
assemblies be inspected at the end of each fuel cycle to demonstrate
the validity of crud levels analyzed in the ECCS model.
11. What information exists to facilitate developing an acceptable
crud deposition model that could correlate crud deposition with
measured primary water coolant chemistry (e.g., iron-oxide
concentration)? For boiling water reactors, it is difficult to perform
visual inspections or poolside measurements of fuel rod crud thickness
without first removing the channel box. A crud deposition model would
facilitate the confirmation of design crud layers assumed in the ECCS
evaluations and provide an indicator to reactor operators when crud
levels approach unanalyzed conditions. Are there ongoing or planned
industry efforts to monitor water coolant chemistry for comparison to
observed crud deposition? If so, what amount of success has been
obtained? Could a properly correlated crud model be sufficiently
accurate to preclude the need for crud measurements at the end of each
fuel cycle?
Cost Considerations
12. The U.S. commercial nuclear power industry claims that
implementation of the proposed rule would be a significant burden in
both money and resources. The industry has discussed an implementation
cost of approximately $250 million (NRC-2008-0332-0008.1 at http://www.regulations.gov).
a. What options are available to reduce this implementation cost?
b. Are there changes in core operating limits, fuel management, or
cladding material that would reduce the cost and burden of implementing
the proposed hydrogen based PQD criterion without negatively impacting
operations?
c. A staged implementation would be more manageable for both the
NRC and industry. One potential approach involves characterizing the
plants based upon safety margin and deferring implementation for the
licensees with the largest safety margin (e.g., lowest calculated CP-
ECR). The NRC requests comment on this implementation approach.
Available Supporting Documents
The following documents provide additional background and
supporting information regarding this rulemaking activity and
corresponding technical basis. The documents can be found in the NRC's
Agencywide Document Access and Management System (ADAMS). Instructions
for accessing ADAMS were provided under the ADDRESSES section of this
document.
[[Page 40774]]
------------------------------------------------------------------------
ADAMS accession
Date Document number
------------------------------------------------------------------------
July 31, 2008................... NUREG/CR-6967, ML082130389.
``Cladding
Embrittlement
During Postulated
Loss-of-Coolant
Accidents''.
May 30, 2008.................... Research ML081350225.
Information
Letter (RIL)
0801, ``Technical
Basis for
Revision of
Embrittlement
Criteria in 10
CFR 50.46''.
September 24, 2008.............. Public Meeting ML083010496.
Summary.
February 23, 2009............... Plant Safety ML090340073.
Assessment of RIL
0801.
July 31, 2008................... Federal Register Reference the
Notice (73 FR Federal Register
44778), ``Notice at 73 FR 44778.
of Availability
and Solicitation
of Public
Comments on
Documents Under
Consideration To
Establish the
Technical Basis
for New
Performance-Based
Emergency Core
Cooling System
Requirements''.
March 30, 2009.................. Supplemental ML090690711.
research
material--additio
nal PQD tests.
March 30, 2009.................. Supplemental ML090700193.
research
material--additio
nal breakaway
testing.
March 31, 2009.................. Draft proposed ML090900841.
procedure for
Conducting
Oxidation and
Post-Quench
Ductility Tests
With Zirconium-
based Cladding
Alloys.
March 23, 2009.................. Draft proposed ML090840258.
procedure for
Conducting
Breakaway
Oxidation Tests
With Zirconium-
Based Cladding
Alloys.
January 8, 2009................. Update on ML091330334.
Breakaway
Oxidation of
Westinghouse
ZIRLO Cladding.
May 7, 2009..................... Impact of Specimen ML091350581.
Preparation on
Breakaway
Oxidation (Non-
Proprietary).
------------------------------------------------------------------------
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
The authority citation for this document is 42 U.S.C. 2201.
Dated at Rockville, MD, this 29th day of July 2009.
For the Nuclear Regulatory Commission.
R.W. Borchardt,
Executive Director for Operations.
APPENDIX A
An Approach for Determining the Acceptability of Zirconium-Based
Cladding Alloys: Analytical Limits Defined Within Regulatory Guidance
This approach would include a future regulatory guidance
document that defines an acceptable, generically-applicable set of
analytical limits for peak cladding temperature and maximum
allowable time-at-temperature (expressed as calculated local
oxidation, CP-ECR) as a function of pre-transient hydrogen content
in the cladding metal (excluding hydrogen in the cladding oxide
layer). These acceptable analytical limits would be developed using
NRC's empirical database with consideration of experimental
variability and repeatability. Figure A shows the results of ring-
compression tests conducted on as-fabricated, hydrogen charged, and
irradiated specimens of Zircaloy-2, Zircaloy-4, ZIRLOTM
and M5 cladding material (documented in NUREG/CR-6967). Note that
hydrogen concentrations were slightly adjusted ( 5 wppm)
to illustrate results of multiple ring-compression tests run at the
same CP-ECR and hydrogen concentration. Peak oxidation temperature
is identified for samples tested below 2200 [deg]F (1204 [deg]C).
[[Page 40775]]
[GRAPHIC] [TIFF OMITTED] TP13AU09.025
The analytical limit on PCT would be restricted to the peak
oxidation temperature during testing of the cladding specimens used
in the development of this limit. Furthermore, caveats on the
applicability of the analytical limits may be required to capture
limiting aspects of the steam oxidation temperature profile used
during the testing. For example, if the calculated time at the
specified PCT is less than the time at peak oxidation temperature of
the supporting empirical database (for a given CP-ECR), or the
calculated quench temperature is lower than 800 [deg]C, then an
applicability caveat may be required.
Existing Cladding Alloys
No PQD testing would be required to approve cladding alloys
included in the NRC's LOCA research program. Under this approach, a
fuel vendor would submit a topical report (TR) seeking NRC approval
of each zirconium-based cladding alloy's analytical limits on PCT
and time-at-temperature (CP-ECR, as a function of cladding hydrogen
content). The TR would reference the acceptable analytical limits
within the Regulatory Guide.
New Cladding Alloys
Under this approach, a fuel vendor would submit a TR which
demonstrates that the results of PQD tests on a specific new alloy
are applicable to the acceptable analytical limits defined within
the Regulatory Guide. A TR would need to include the results of
testing, conducted in accordance with NRC's acceptable experimental
methodology, which demonstrates that the embrittlement behavior of
the new cladding alloy is consistent with the embrittlement behavior
of the cladding alloys tested in NRC's LOCA research program by
comparing test results to the defined analytical limit. This would
likely require testing of the new cladding alloy with varying
hydrogen contents, which are oxidized to calculated oxidation levels
(CP-ECR) at or near the analytical limit for that hydrogen level as
provided in regulatory guidance. Demonstrating ductile behavior in
cladding samples with calculated oxidation levels at or near the
analytical limit may serve to confirm the applicability of the
analytical limit to a new cladding alloy. The range of hydrogen
contents in test samples required may be limited by proposing
cladding hydrogen design limits based on hot cell examinations of
irradiated samples of the new cladding alloy following lead test
assembly campaigns. Regulatory guidance would be provided to address
the variability in measured offset strain of ring-compression test
results. Section IV of this ANPR specifically seeks comment on the
treatment of variability in ductility measurements of ring-
compression tests.
For this description, it is assumed that sufficient
justification for the use of hydrogen charged cladding specimens has
been accepted as a surrogate for testing on irradiated cladding
segments. If sufficient justification for the use of hydrogen
charged cladding specimens has not been accepted as a surrogate for
testing on irradiated cladding segments, approving new cladding
alloys would require PQD testing of irradiated material. Section IV
of this ANPR requests information on any ongoing or planned testing
aimed at validating this pre-hydrided surrogate.
APPENDIX B
An Approach for Determining the Acceptability of Zirconium-Based
Cladding Alloys: Cladding-Specific Analytical Limits Defined by an
Applicant
This approach involves establishing cladding-specific and/or
temperature-specific analytical limits for peak cladding temperature
and maximum allowable time-at-temperature (expressed as calculated
local oxidation, CP-ECR) as a function of pre-transient hydrogen
content in the cladding metal (excludes hydrogen in the cladding
oxide layer). This approach would provide optimum flexibility for
defining more specific analytical limits to gain margin to the ECCS
performance criteria. However, unlike citing analytical limits
within a regulatory guide, this approach places the burden of proof
on the applicant to validate their analytical limits and address
[[Page 40776]]
experimental variability and repeatability. As a result, this
approach would necessitate a larger number of PQD tests (relative to
confirming the applicability of the regulatory guide). Analytical
limits, along with the experimental procedures, protocols, and
specimen test results used in their development, would be subject to
NRC review and approval.
This approach would require that the PQD test results on
irradiated cladding segments documented in NUREG/CR-6967 be
considered in the development of analytical limits. Deviations in
cladding performance relative to this empirical database must be
identified and dispositioned.
Existing Cladding Alloys
In the case of existing cladding alloys, the rule may specify
the following performance requirement to ensure an adequate
retention of cladding ductility:
Accumulation of >= 1.00 percent permanent strain prior to
failure during ring-compression loading at a temperature of 135
[deg]C and a displacement rate of 0.033 mm/sec on a cladding
specimen exposed to double-sided steam oxidation up to a specified
peak oxidation temperature and CP-ECR.
Analytical limits on allowable time-at-temperature (CP-ECR) and
peak cladding temperature would need to be defined as a function of
initial cladding hydrogen content (wppm in metal) to demonstrate
this performance requirement is met. A topical report (TR) would be
generated to document the basis for the new analytical limits.
Existing alloys which were included in the NRC high-burnup research
program may reference the test results documented in NUREG/CR-6967
in the development of new analytical limits. This data was generated
following experimental protocols acceptable to the NRC, so no
further justification related to its validity would be required.
Using an approved hydrogen uptake model for an existing cladding
alloy, the TR would provide the methodology to convert the hydrogen-
based analytical limits to some unit of measure more readily applied
within reload safety analyses (e.g., fuel rod burnup or fuel duty).
Uncertainties related to hydrogen uniformity and uncertainties
introduced by the conversion from hydrogen to another unit of
measure would need to be addressed.
New Cladding Alloys
In the case of new cladding alloys, the rule may specify the
following performance requirement to ensure an adequate retention of
cladding ductility:
Accumulation of >= 1.00 percent permanent strain prior to
failure during ring-compression loading at a temperature of 135
[deg]C and a displacement rate of 0.033 mm/sec on a cladding
specimen exposed to double-sided steam oxidation up to a specified
peak oxidation temperature and CP-ECR.
Analytical limits on allowable time-at-temperature (CP-ECR) and
peak cladding temperature would need to be defined as a function of
initial cladding hydrogen content (wppm in metal) to demonstrate
this performance requirement is met. A TR would be generated to
document the basis for the new analytical limits. The PQD test
results on irradiated cladding segments documented in NUREG/CR-6967
would need to be considered in the development of analytical limits.
PQD testing would be required to (1) establish analytical limits in
accordance with the performance requirements that would be specified
within the rule, and (2) demonstrate the applicability of the NUREG/
CR-6967 empirical database. A TR could document that the PQD testing
had been conducted to strictly adhere to the accepted experimental
protocols documented in regulatory guidance documents, or if
alternative testing procedures were used, then NRC review and
approval of those laboratory procedures would be required.
For this approach, defining analytical limits for new cladding
alloys would likely require testing at a range of hydrogen contents,
with ring-compression test results at multiple calculated oxidation
levels. Test samples with calculated oxidation levels sufficient to
display brittle behavior, as well as test samples with calculated
oxidation levels which display ductile behavior, would be necessary
to define the transition from ductile to brittle behavior.
Regulatory guidance would be provided to address the variability in
measured offset strain of ring-compression test results. Section IV
of this ANPR specifically seeks comment on the treatment of
variability in ductility measurements of ring-compression tests. The
range of hydrogen contents in test samples required may be limited
by proposing cladding hydrogen design limits based on hot cell
examinations of irradiated samples of the new cladding alloy
following lead test assembly campaigns.
Multifaceted Analytical Limits
Recognizing that higher burnup fuel rods (with higher hydrogen
concentrations) operate at a reduced power level (relative to lower
burnup fuel rods), defining analytical limits for maximum allowable
ECR at multiple peak oxidation temperatures would also be possible.
For example, a TR could document the results of testing conducted at
peak oxidation temperatures of 2200 [deg]F (1204 [deg]C), 2000
[deg]F (1093 [deg]C), and 1800 [deg]F (982 [deg]C), which are
targeted at low burnup (low corrosion), medium burnup (medium
corrosion), and high burnup (high corrosion) fuel rods,
respectively. Testing to support these new limits would require
testing at a range of hydrogen contents, with ring-compression test
results at multiple calculated oxidation levels to define the
transition from ductile to brittle behavior. In this case, it may be
necessary to elect to strictly adhere to the accepted experimental
protocols documented in regulatory guidance documents, thereby
limiting regulatory exposure related to testing procedures and the
validity of the data.
Implementation of the multifaceted analytical limits would
require separating all of the fuel rods in the core into three
categories and then ensuring that all fuel rods within each category
satisfies their respective analytical limits on both CP-ECR and PCT.
While it is anticipated that this approach would provide
flexibility, it would also necessitate a more complex LOCA analysis
and reload-by-reload confirmation. This approach also relies on
tacit assumptions regarding the currently approved LOCA model's
ability to accurately simulate the thermal-hydraulic conditions in
every region of the reactor core (as opposed to simulating a core
average response or pseudo hot channel location). Modeling
uncertainties with respect to predicting local conditions throughout
the reactor core would need to be addressed.
Using an approved hydrogen uptake model for a new cladding
alloy, the TR would need to provide the methodology to convert the
hydrogen-based analytical limits to some unit of measure more
readily applied within reload safety analyses (e.g., fuel rod burnup
or fuel duty). Uncertainties related to hydrogen uniformity and
uncertainties introduced by the conversion from hydrogen to another
unit of measure would need to be addressed.
For this description, it is assumed that sufficient
justification for the use of hydrogen charged cladding specimens has
been accepted as a surrogate for testing on irradiated cladding
segments. If sufficient justification for the use of hydrogen
charged cladding specimens has not been accepted as a surrogate for
testing on irradiated cladding segments, approving new cladding
alloys would require PQD testing of irradiated material. Section IV
of this ANPR requests information on any ongoing or planned testing
aimed at validating this pre-hydrided surrogate.
[FR Doc. E9-19423 Filed 8-12-09; 8:45 am]
BILLING CODE 7590-01-P