[Federal Register Volume 72, Number 99 (Wednesday, May 23, 2007)]
[Proposed Rules]
[Pages 28902-28906]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-9910]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[Docket No. PRM-50-84]


Mark Edward Leyse; Receipt of Petition for Rulemaking

AGENCY: Nuclear Regulatory Commission.

ACTION: Petition for rulemaking; notice of receipt.

-----------------------------------------------------------------------

SUMMARY: The Nuclear Regulatory Commission (NRC) has received and 
requests public comment on a petition for rulemaking dated March 15, 
2007, filed by Mark Edward Leyse. The petition has been docketed by the 
NRC and has been assigned Docket No. PRM-50-84. The petitioner is 
requesting that the NRC amend the regulations that govern domestic 
licensing of production and utilization facilities to require that 
nuclear power facilities be operated to limit the thickness of crud 
(corrosion products) layers and/or the thickness of oxide layers on 
fuel rod cladding surfaces. The petitioner also requests that the 
requirements pertaining to Emergency Core Cooling System (ECCS) 
evaluation models be amended to require that the steady-state 
temperature distribution and stored energy in reactor fuel at the onset 
of a postulated loss-of-coolant accident (LOCA) be calculated by 
factoring in the role that the thermal resistance of crud and/or oxide 
layers on cladding plays in increasing the stored energy in the fuel. 
Lastly, the petitioner requests that the acceptance criteria for 
emergency core cooling systems for light-water nuclear power reactors 
be amended to stipulate a maximum allowable percentage of hydrogen 
content in cladding of fuel rods.

[[Page 28903]]


DATES: Submit comments by August 6, 2007. Comments received after this 
date will be considered if it is practical to do so, but assurance of 
consideration cannot be given except as to comments received on or 
before this date.

ADDRESSES: You may submit comments by any one of the following methods. 
Please include the following number (PRM-50-84) in the subject line of 
your comments. Comments on petitions submitted in writing or in 
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 personal 
information such as social security numbers and birth dates in your 
submission.
    Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555. Attention: Rulemaking and Adjudications staff.
    E-mail comments to: <a href="/cdn-cgi/l/email-protection#4013050319002e32236e272f36"><span class="__cf_email__" data-cfemail="97c4d2d4ced7f9e5f4b9f0f8e1">[email protected]</span></a>. If you do not receive a reply e-
mail confirming that we have received your comments, contact us 
directly at (301) 415-1966. You may also submit comments via the NRC's 
rulemaking Web site at http://ruleforum.llnl.gov. Address comments 
about our rulemaking Web site to Carol Gallagher, (301) 415-5905; (e-
mail <a href="/cdn-cgi/l/email-protection#187b797f58766a7b367f776e"><span class="__cf_email__" data-cfemail="25464442654b57460b424a53">[email protected]</span></a>). Comments can also be submitted via the Federal 
eRulemaking Portal http:www.regulations.gov.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland, between 7:30 a.m. and 4:15 p.m. on Federal workdays.
    Publicly available documents related to this petition may be viewed 
electronically on the public computers located at the NRC Public 
Document Room (PDR), O1 F21, One White Flint North, 11555 Rockville 
Pike, Rockville, Maryland. The PDR reproduction contractor will copy 
documents for a fee. Selected documents, including comments, may be 
viewed and downloaded electronically via the NRC rulemaking Web site at 
http://ruleforum.llnl.gov.
    Publically available documents created or received at the NRC after 
November 1, 1999 are also available electronically at the NRC's 
Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. 
From this site, the public can gain entry into the NRC's Agencywide 
Documents Access and Management System (ADAMS), which provides text and 
image files of NRC's public documents. If you do not have access to 
ADAMS or if there are problems in accessing the documents located in 
ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-
4737 or by e-mail to <a href="/cdn-cgi/l/email-protection#49392d3b09273b2a672e263f"><span class="__cf_email__" data-cfemail="74041006341a06175a131b02">[email protected]</span></a>.
    For a copy of the petition, write to Michael T. Lesar, Chief, 
Rulemaking, Directives and Editing Branch, Division of Administrative 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.

FOR FURTHER INFORMATION CONTACT: Michael T. Lesar, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555. Telephone: 301-415-7163 or Toll-Free: 1-800-368-5642 or E-mail: 
<a href="/cdn-cgi/l/email-protection#7b362f373b352938553c140d"><span class="__cf_email__" data-cfemail="db968f979b958998f59cb4ad">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION:

Background

    The NRC has received a petition for rulemaking dated March 15, 
2007, submitted by Mark Edward Leyse (petitioner). The petitioner 
requests that the NRC amend 10 CFR part 50, ``Domestic Licensing of 
Production and Utilization Facilities.'' Specifically, the petitioner 
requests 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, fuel cycle lengths, and light-water 
coolant chemistries) necessary to effectively limit the thickness of 
crud (corrosion products) layers and/or oxide layers on fuel rod 
cladding surfaces. The petitioner believes that new regulations are 
needed for reactor-operation parameters, uranium-oxide and mixed-oxide 
fuel and cladding, in order to ensure that cladding is free of unsafe 
thicknesses of crud and/or oxide, which in turn would help ensure that 
nuclear power plants operate in compliance with 10 CFR 50.46(b). 10 CFR 
50.46(b) stipulates that the calculated peak cladding temperature (PCT) 
must not exceed 2200[deg]F in the event of a loss-of-coolant accident 
(LOCA). The petitioner also requests that 10 CFR part 50, Appendix K, 
``ECCS Evaluation Models'' be amended to require that the steady-state 
temperature distribution and stored energy in the fuel at the onset of 
a postulated LOCA be calculated by factoring in the role that the 
thermal resistance of crud and/or oxide layers on cladding plays in 
increasing the stored energy in the fuel. Lastly, the petitioner 
requests that Sec.  50.46 be amended to stipulate a maximum allowable 
percentage of hydrogen content in fuel cladding.
    The NRC has determined that the petition meets the threshold 
sufficiency requirements for a petition for rulemaking under 10 CFR 
2.802. The petition has been docketed as PRM-50-84. The NRC is 
soliciting public comment on the petition for rulemaking.

Discussion of the Petition

    The petitioner states that layers of crud and oxide on cladding 
surfaces of nuclear fuel rods could cause the temperature of fuel rods 
to increase up to 300 [deg]F to 600 [deg]F during power plant 
operations. The petitioner also states that during a LOCA, the thermal 
resistance of insulating layers of crud and oxide on cladding, and 
increased fuel temperatures will cause the PCT to be higher than if the 
cladding were clean. The petitioner believes that if a large break (LB) 
LOCA occurred at a nuclear power plant that operated with heavy crud 
and oxide layers, there is a high probability that the PCT would exceed 
the 2200 [deg]F limit in Sec.  50.46(b)(1). The petitioner states that 
increased hydrogen content in cladding contributes to cladding 
embrittlement. The petitioner believes that Sec.  50.46 should also be 
amended to specify a maximum allowable percentage of hydrogen in 
cladding.
    The petitioner states that in 2001, the Indian Point Unit 2 
facility had a PCT of 2188 [deg]F during a computer simulated LB LOCA. 
The petitioner believes that if heavy crud and oxide layers were 
present and included in the calculation, it is ``highly probable'' the 
calculated PCT would have exceeded the 2200 [deg]F limit, perhaps by 
hundreds of degrees Fahrenheit. The petitioner states that if the 2200 
[deg]F limit was exceeded during actual operation, the cladding could 
lose its physical integrity and result in a core meltdown that would 
release radioactive material and contaminate the environment. The 
petitioner states that in 1995, the Three Mile Island Unit 1 facility 
(TMI-1) operated with crud deposits on the surface of fuel rods that 
raised the cladding temperature by 180 to 270 [deg]F or greater over 
the typical operating temperature of 346 [deg]C during Cycle 10. The 
petitioner believes that if an actual LB LOCA had occurred at TMI-1, 
the crud and oxide layers on the cladding would have caused the PCT to 
exceed 2200 [deg]F.
    The petitioner states that because corrosion is not detected during 
plant operation, a significant length of time passes before corrosion 
progresses enough to perforate cladding and cause an increase in 
``offgas'' activity meaning that heavily corroded fuel rods are often 
operated at full power for significant periods of time and could cause 
the cladding to fracture during the reflood period and lose structural 
integrity. The petitioner concludes that this could compromise the 
structural soundness

[[Page 28904]]

and the ability to keep the core cooled and illustrates the impact that 
the thermal resistance of heavy layers of oxide and crud on cladding 
would have during a LOCA.
    The petitioner cites a September 30, 2003, Advisory Committee on 
Reactor Safeguards (ACRS) meeting transcript stating that the thickness 
of crud deposited on the cladding during pressurized water reactor 
(PWR) plant operation is often not known because ``a great deal of PWR 
crud comes off the cladding during reactor shutdown.'' The petitioner 
notes that crud deposits on cladding in PWRs have been measured at up 
to 125[mu]m thick. The ACRS found that a crud layer with steam 
blanketing would have ``extremely poor conductivity'' and that crud is 
``difficult to characterize'' because its thermal conductivities 
``depend on [its] morpholog[ies].''
    The petitioner also cites an Electric Power Research Institute 
(EPRI) study to be completed in 2008 that will attempt ``to determine 
the effect of tenacious crud on fuel surface heat transfer.'' The 
petitioner notes that this study is for crud in a boiling water reactor 
(BWR) but believes the results can also be applied to PWRs.
    The petitioner also notes that the EPRI study found that Zirconium 
dioxide (ZrO<INF>2</INF>) has a low thermal conductivity and is used 
industrially as an insulating material. The petitioner cites an EPRI-
sponsored study entitled, ``Taming the Crud Problem: The Evolution,'' 
presented at the Advances in Nuclear Fuel Management II Conference in 
October 2003 which states that: ``Oxide can form, with or without the 
benefit of crud, in the presence of sustained cladding temperatures. 
Like crud, formation of an oxide layer inhibits heat transfer causing 
accelerated corrosion which can potentially lead to fuel failure.''
    The petitioner describes a fuel rod that failed at TMI-1 during 
Cycle 10 that may have had about a 200[mu]m-thick layer of crud and 
oxide. The petitioner believes that if a LB LOCA occurred during this 
fuel cycle, the layer of crud and oxide would have inhibited effective 
heat transfer and likely caused the PCT to exceed 2200 [deg]F, possibly 
causing a meltdown. The petitioner also reiterates that in 2001, the 
Indian Point Unit 2 facility had a PCT of 2188 [deg]F in a computer 
simulated LB LOCA. The petitioner believes that if Indian Point Unit 2 
had cladding conditions similar to those of TMI-1 Cycle 10, it is 
highly probable the PCT would also have been greater than 2200 [deg]F. 
The petitioner states that TMI-1 is not the only PWR to experience 
crud-induced corrosion failures. In 1997, the Palo Verde Unit 2 and 
Seabrook facilities both had the same problem.
    The petitioner cites NUREG-1230, ``Compendium of ECCS Research for 
Realistic LOCA Analysis'' and states that the stored energy in the fuel 
increases because cladding encased in heavy amounts of crud and oxide 
cannot transfer heat efficiently to coolant during the blowdown phase 
of the event. The petitioner states that the increased stored energy 
caused by heavy crud and oxide layers on fuel cladding and the delay in 
the transfer of the heat to the coolant cause the cladding to be 
subjected to extremely high temperatures for much longer than if the 
cladding was clean (free of crud and oxidation) at the onset of the 
LOCA. The petitioner believes that this would result in more 
degradation of the fuel and embrittlement of cladding. The petitioner 
also states that when the cladding reacts with steam, an exothermic 
reaction occurs that generates additional heat on the cladding.
    The petitioner cites an ACRS meeting transcript from February 2, 
2007, in which an NRC staff member explained that a basic LOCA 
transient calculation includes an oxidation limit and involves time and 
temperature. The petitioner also notes that NUREG-1230 states that 
embrittled cladding can fragment upon contact with emergency cooling 
water in a severe accident. The embrittlement is a function of 
temperature, time, the supply of steam and zircaloy, and can lead to 
the loss of effective cooling, making it relevant to fuel rod safety. 
The petitioner also notes that NUREG-1230 also states, ``[the] amount 
of residual thermal energy [in the fuel rod] influences the time 
required to quench the reactor core with emergency cooling water 
[emphasis added].''
    The petitioner states that absorption of hydrogen would also 
contribute to a loss of cladding ductility during a LOCA along with 
cladding degradation and massive oxidation. He cites a failed fuel rod 
from the TMI-1, Cycle 10 event when hydrogen absorption caused hydrided 
material to break away from the outer portions of the cladding. The 
petitioner believes that the effects of increased stored energy due to 
a heavy crud layer in the fuel and the severity of cladding oxidation, 
embrittlement, and resulting fuel degradation during an actual event 
would be substantially greater than in an ECCS calculation based on 
clean cladding.
    The petitioner also states that little or no evidence exists that 
crud has ever been properly factored into PCT calculations for 
postulated LOCAs. He cites a June 17, 2003, Idaho National Engineering 
and Environmental Laboratory document that he believes stated that crud 
has not been applied to severe accident codes because it has not been 
demonstrated to be necessary and that users have not chosen to consider 
effects of crud. The petitioner also cites the 2002 annual report on 
ECCS evaluation from the Callaway facility that he believes proves that 
``little attention'' was placed on effects of heavy crud on thermal 
resistance. The petitioner states that most cladding that experienced 
crud-induced corrosion failures recently at PWRs involved high-power, 
one-cycle fuel. He cites the TMI-1, Cycle 10 and Callaway, Cycle 6 
events as examples and notes that the effects of crud can occur 
quickly. The cladding perforation at TMI-1 was detected only 121 days 
into the cycle.
    The petitioner states that the values of the stored energy in BOL 
fuel or fuel with burnups between 30 to 35 GWd/MTU are used to 
calculate PCTs during postulated LOCAs. The petitioner also believes it 
is significant that the stored energy of fuel sheathed in cladding with 
heavy crud and oxide layers is substantially higher than fuel of the 
same burnup rate sheathed in clean cladding that he states is used in 
PCT calculations performed for postulated LOCAs and during safety 
evaluations of the certification process of newer designs such as the 
Westinghouse AP 1000 reactor. The petitioner believes that the AP 1000 
PCTs were not calculated for the maximum stored energy that fuel can 
reach during operation and that recent experiences with fuel at TMI-1, 
Palo Verde Unit 2, and Seabrook were not considered in PCT calculations 
performed during recent power ``uprates'' at other nuclear power 
facilities.
    The petitioner states that axial offset anomaly (AOA) or crud-
induced power shift (CIPS) are phenomena caused by crud on cladding and 
can indicate how frequently crud affects nuclear power plant operation. 
The petitioner also states that AOA occurs in PWRs when crud deposits 
on cladding contain enough boron to reduce the rate of fission in the 
vicinity of the crud. He cites NRC Information Notice 97-85, ``Effects 
of Crud Buildup and Boron Deposition on Power Distribution and Shutdown 
Margin'' that describes how AOA causes power distribution shifts toward 
the bottom of the reactor core as a result of reduced fission in the 
upper reactor core. The petitioner states that although crud deposits 
must be at least 35[mu]m thick for AOA to occur, it is possible that 
crud deposits thicker than 35[mu]m do not cause AOA because not all

[[Page 28905]]

crud deposits contain enough boron to cause this phenomenon. The 
petitioner also states that according to a 2002 Department of Energy 
report on nuclear energy plant optimization, the thickest layer of crud 
measured in a PWR was 125[mu]m thick that caused AOA but not cladding 
perforation and that as of 2003, more than 30 fuel cycles in 16 PWRs 
had exhibited AOA.
    The petitioner also cites a 2006 EPRI report that acknowledged that 
crud has contributed to AOA at many power plants since the 1980s 
because fuel cycle operation and power up rates have increased 
appreciably and that excessive crud deposits create operational 
difficulties. According to a Westinghouse Electric Company official, 
AOAs are detectable and closely monitored to ensure that adequate 
shutdown margins can be maintained. Also, a plant can be operated at a 
lower power level if necessary. The petitioner cites the TMI-1, Cycle 
10 event as an example that illustrates how low levels of boron can 
result in a slight AOA even though enough crud was present to induce 
fuel failure from corrosion. The petitioner states that if a heavy crud 
layer was detected during plant operation that did not cause an AOA, it 
is unlikely that the operation power level would be reduced because the 
thermal resistance of the crud and how it would raise the PCT during a 
LOCA would likely not be considered problematic.
    The petitioner describes what he believes was a crud-induced 
cladding corrosion failure of fuel in a BWR at the River Bend facility 
during Cycle 8 from 1998-99. The petitioner states that the fuel 
failure occurred when crud nearly bridged the gap between adjacent rods 
and believes it is significant that most of the failed rods were high-
power, one-cycle rods (much like the recent corrosion-induced PWR fuel 
failures during the TMI-1 Cycle 10, Palo Verde Unit 2 Cycle 9, and 
Seabrook Cycle 5 events). The River Bend Cycle 8 fuel failure resulted 
from thick layers of crud, augmented with copper that accelerated the 
oxidation process to produce a local steam blanketing and high heat 
transfer resistance that created perforations in the fuel cladding 
according to the ``Recent GE BWR Fuel Experience'' report published in 
2000 by the American Nuclear Society and the NRC inspection report 
pertaining to this event. The petitioner concludes that the combined 
effects of crud and oxide layers increased the cladding temperatures 
from around 560 [deg]F to temperatures approaching 1200 [deg]F.
    The petitioner states that if a LOCA had occurred during this 
event, the PCT could likely have exceeded the 2200 [deg]F limit 
specified in Sec.  50.46. The petitioner acknowledges that the NRC 
Licensee Event Report (LER) 50-458/99-016-00 states that the PCT was 
calculated to have been 1700 [deg]F or less and demonstrates a 
substantial margin to the 2200 [deg]F limit. However, the petitioner 
states that the LER ignores NRC guidelines for calculating the 
equivalent cladding reacted (ECR) and believes that the PCT would have 
exceeded 1700 [deg]F during a LB LOCA. The petitioner states that in 
2000 when this LER was filed there was not much knowledge about values 
for the thermal conductivity of crud and how crud layers should be 
modeled in severe accident codes and believes this lack of knowledge 
still exists in 2007. The petitioner reiterates there is little or no 
evidence that crud has ever been properly factored into PCT 
calculations for simulated LOCAs at nuclear power plants.
    The petitioner states that essentially the same cladding condition 
occurred again at the River Bend facility between October 2001 to March 
2003 during the Cycle 11 refueling event after a GE Nuclear Energy 
official had stated that heavy crud buildup during the Cycle 8 event 
was unique and had occurred only once in over 1000 reactor years of 
operation. He cites a paper presented at the 2004 International Meeting 
on LWR Fuel presented by the American Nuclear Society, ``Fuel Failures 
During Cycle 11 at River Bend.'' This paper stated that this fuel rod 
failure was caused by accelerated oxidation of the cladding resulting 
from unusually heavy deposits of tenacious crud that diminished heat 
transfer in local areas of the cladding surface. The petitioner notes 
that the failures occurred in high power, one-cycle rods where heavy 
crud and oxide layers were present. The petitioner believes that the 
PCT during a LB LOCA would have exceeded the 2200 [deg]F limit 
specified in Sec.  50.46 and means that the ECCS design basis for River 
Bend is non-conservative for calculating the PCT for a postulated LOCA 
when heavy crud and oxide layers exist on cladding.
    The petitioner disputes GE Nuclear Energy's conclusion that because 
the heavy crud deposits on fuel rods at the River Bend facility 
occurred at the lower elevations of the fuel assembly and the more 
limiting axial elevations during a postulated LOCA occur at the upper 
elevations of a fuel assembly where at River Bend the crud 
characteristics were normal, the heavy crud deposits would have no 
significant effect on the fuel response to a postulated LOCA. The 
petitioner states that the cladding surface temperatures during the 
River Bend events reached 1200 [deg]F, far above the specified 
licensing basis of about 578 [deg]F. The petitioner believes that the 
higher temperatures due to the heavy crud and oxide layers would result 
in less coolant flow than for clean cladding, would cause the cladding 
to be subjected to extremely high temperatures for a substantially 
longer duration than used in the licensing basis, and result in more 
fuel degradation. The petitioner also believes that the degradation of 
fuel and cladding would further obstruct reflood coolant flow, delay 
transfer of stored energy to the coolant during quench, and that during 
a LOCA there would already be severe cladding degradation, massive 
oxidation, and absorption of hydrogen that would contribute to a loss 
of cladding ductility. The petitioner has concluded that these factors 
mean that the River Bend facility operated in violation of Sec.  
50.46(b) during cycles 8 and 11 of refueling. The petitioner also 
states that the Browns Ferry facility operated from April 2001 to March 
2003 with thick oxide layers at the upper elevations of the fuel rods 
and believes it is significant that the heavy crud and oxide layers 
that caused overheating and cladding perforations at TMI-1 during cycle 
10 were located at upper elevations of fuel assemblies.
    The petitioner cites a 2004 paper, ``An Integrated Approach to 
Maximizing Fuel Reliability'' stating that a lack of understanding 
exists about the interplay of materials, fuel duty, and water chemistry 
variables and reports that crud or corrosion related fuel failures 
occurred at BWRs in six of the years between 1997 to 2004. The 
petitioner also cites an EPRI document, ``2006 Portfolio 41.002 Fuel 
Reliability'' which states that the fuel failure rate has increased in 
both BWRs and PWRs during the last couple of years due to extended and 
more aggressive fuel cycle operation. The petitioner states that 
although the nuclear industry observed that it appeared that nodular 
corrosion had been eliminated from BWR fuel cladding in 2000, by 2004 
it had reemerged at several BWRs. The petitioner believes this is a 
result of increasing fuel duty by extending the length of fuel cycles 
and that problems with crud and oxide will continue unless the NRC 
implements regulations to ensure that BWRs and PWRs do not operate with 
high levels of crud and oxidation on cladding that cause violations of 
Sec.  50.46(b).
    The petitioner states that Appendix K to 10 CFR part 50, ``ECCS 
Evaluation Models'' requires stored energy in nuclear fuel to be 
calculated to yield the highest PCT. The petitioner believes that 
Appendix K should require thermal

[[Page 28906]]

conductivity of layers of crud and oxide to be factored into 
calculations of the stored energy in the fuel. The petitioner states 
that because a heavy crud layer would increase the quantity of stored 
energy in the fuel, the PCT would also increase above that of fuel with 
the same burnup sheathed in clean cladding. The petitioner also states 
that instructions specified in Appendix K for calculating the quantity 
of stored energy that contains heavy layers of crud and oxide are non-
conservative.
    The petitioner notes that values of stored energy in BOL fuel or 
fuel with burnups between 30 to 35 Gwd/MTU are used to calculate PCTs 
during postulated LOCAs. However, the petitioner cites a January 2007 
ACRS Subcommittee on Materials, Metallurgy, and Reactor Fuels during 
which a Westinghouse official cited data from LOCA calculations showing 
that single cycle fuel with burnups from zero to approximately 20 or 25 
GWd/MTU yielded the highest PCTs. Westinghouse also stated that at 
burnups of about 30 GWd/MTU, there is approximately a ten percent 
reduction in achievable power, which yields PCTs approximately 100 
[deg]C lower than those of fresh fuel. The petitioner concludes it is 
significant that an ECCS design based on Appendix K requirements is 
non-conservative and hazardous for calculating the quantity of stored 
energy in one-cycle fuel that has heavy crud on the cladding.
    The petitioner states that an increase in hydrogen content in 
cladding contributes to cladding embrittlement. The petitioner cites an 
April 4, 2001, ACRS Reactor Fuels Subcommittee meeting during which an 
expert from Argonne National Laboratory stated that a reduction of 
ductility occurs when hydrogen levels reach about 600 to 700 parts-per-
million (ppm) in Zircaloy cladding. According to the petitioner, 
another expert from the Atomic Energy Research Institute stated that a 
threshold for a reduction of ductility in Zircaloy cladding occurs at 
even a lower hydrogen level of about 150 to 200 ppm. The petitioner 
also cites the TMI-1 Cycle 10 event that included massive hydrogen 
absorption in fuel cladding. The petitioner notes that hydrogen content 
in the cladding of a rod that did not fail measured 700 ppm at TMI-1 
and that this level of hydrogen content in one-cycle cladding is 
similar to the 800 ppm level measured in fuel cladding at the H.B. 
Robinson, Unit 2 facility, a PWR. The petitioner states that some of 
the cladding at TMI-1 Cycle 10 contained levels of hydrogen that 
Argonne National Laboratory found would have caused a loss of cladding 
ductility in addition to the embrittlement resulting from excessive 
oxide levels.

The Petitioner's Proposed Actions

    The petitioner states that new regulations are needed for reactor 
operation parameters, uranium-oxide and mixed-oxide fuel, and fuel 
cladding to ensure that cladding does not contain unsafe amounts of 
crud and oxide to help ensure that nuclear power plants operate in 
compliance with 10 CFR 50.46(b). The petitioner also states that 
nuclear power plant licensees should be required to factor the thermal 
resistance effects of crud and oxide layers on cladding into 
calculations of PCTs for postulated LOCAs at their facilities. Also, 
the NRC needs to consider effects of crud and oxide when reviewing 
power plant operations reports under 10 CFR 50.46, and before approving 
power uprates at existing facilities and new nuclear power plant 
designs, such as the recently certified Westinghouse AP1000 design.
    The petitioner requests that Appendix K to Part 50 be amended to 
require that the steady state temperature distribution and stored 
energy in the fuel at the onset of a postulated LOCA be calculated by 
factoring in the role that the thermal resistance of crud and oxide 
layers on cladding plays in increasing the stored energy in nuclear 
fuel. The petitioner also states that Appendix K should specify 
instructions to more accurately calculate the role that thermal 
resistance of crud and oxide layers on cladding plays in determining 
the stored energy in the fuel and the PCT during a postulated LOCA.
    Lastly, the petitioner requests that Sec.  50.46 be amended to 
include a requirement that stipulates a maximum allowable percentage of 
hydrogen content in cladding because there is extensive evidence that 
excessive hydrogen levels and oxidation on cladding contributes to 
cladding embrittlement. The petitioner concludes that the requested 
amendments should also apply to any NRC-approved, best-estimate ECCS 
evaluations used instead of Appendix K calculations. The petitioner 
believes its requested amendments would ensure that nuclear power 
facilities prevent unsafe amounts of crud and oxide layers on cladding 
from occurring during operation to reduce risks to plant workers and 
the public, and help nuclear power facility operations to comply with 
10 CFR 50.46(b).

    Dated at Rockville, Maryland, this 15th day of May 2007.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. E7-9910 Filed 5-22-07; 8:45 am]
BILLING CODE 7590-01-P


</pre><script data-cfasync="false" src="/cdn-cgi/scripts/5c5dd728/cloudflare-static/email-decode.min.js"></script></body>
</html>