Performance-Based Emergency Core Cooling System Acceptance Criteria, 40765-40776 [E9-19423]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 155 (Thursday, August 13, 2009)]
[Proposed Rules]
[Pages 40765-40776]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-19423]


========================================================================
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 https://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 
Carol.Gallagher@nrc.gov.
    E-mail comments to: Rulemaking.Comments@nrc.gov. 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 https://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 PDR.resource@nrc.gov.

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 
Barry.Miller@nrc.gov.

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 https://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.

------------------------------------------------------------------------
                                                        ADAMS accession
              Date                  Meeting/letter          number
------------------------------------------------------------------------
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.
------------------------------------------------------------------------
* 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:
---------------------------------------------------------------------------

    \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 
mea