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.

Proposed Rules Federal Register

40765 

Vol. 74, No. 155 

Thursday, August 13, 2009 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 50 

RIN 3150–AH42 

[NRC–2008–0332] 

Performance-Based Emergency Core 

Cooling System Acceptance Criteria 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Advance notice of proposed 

rulemaking. 

SUMMARY: This advance notice of 

proposed rulemaking (ANPR) presents a 

conceptual approach that the Nuclear 

Regulatory Commission (NRC) is 

considering in a rulemaking effort to 

revise the acceptance criteria for 

emergency core cooling systems (ECCSs) 

for light-water nuclear power reactors as 

currently required by NRC regulations 

that govern domestic licensing of 

production and utilization facilities. 

Revised ECCS acceptance criteria would 

reflect recent research findings that 

indicate the current criteria should be 

re-evaluated for all fuel cladding 

materials in all potential conditions. 

Further, the NRC is considering an 

approach that would expand the 

applicability of the rule to all current 

and future cladding materials, modify 

the reporting requirements, and address 

the issues raised in a petition for 

rulemaking (PRM) regarding crud and 

oxide deposits and hydrogen content in 

fuel cladding. With this ANPR, the NRC 

seeks comment on specific questions 

and issues for consideration related to 

this proposed conceptual approach to 

revising the ECCS acceptance criteria. 

DATES: Submit comments by October 27, 

2009. Comments received after this date 

will be considered if it is practical to do 

so, but the NRC is only able to ensure 

consideration of comments received on 

or before this date. 

ADDRESSES: You may submit comments 

by any one of the following methods. 

Please include the following number 

RIN 3150–AH42 in the subject line of 

your comments. Comments on 

rulemakings submitted in writing or 

electronic form will be made available 

for public inspection. Because your 

comments will not be edited to remove 

any identifying or contact information, 

the NRC cautions you against including 

any information in your submissions 

that you do not want to be publicly 

disclosed. 

We request that any party soliciting or 

aggregating comments received from 

other persons for submission to the NRC 

inform those persons that the NRC will 

not edit their comments to remove any 

identifying or contact information, and 

therefore they should not include any 

information in their comments that they 

do not want publicly disclosed. All 

commenters should ensure that 

sensitive or Safeguards Information is 

not contained in their responses or 

comments to this ANPR. 

Federal e-Rulemaking Portal: Go to 

http://www.regulations.gov and search 

for documents filed under Docket ID 

NRC–2008–0332. Address questions 

about NRC dockets to Carol Gallagher 

(301) 492–3668; e-mail 

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 http://www.nrc.gov/ 

NRC/reading-rm/adams.html. From this 

page, the public can gain entry into 

ADAMS, which provides text and image 

files of NRC’s public documents. If you 

do not have access to ADAMS or if there 

are any problems in accessing the 

documents located in ADAMS, contact 

the NRC PDR Reference staff at (800) 

397–4209, (301) 415–4737, or by e-mail 

to 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 RiskInformed 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 

§ 50.46 specifies the requirements for 

light-water power reactor emergency 

core cooling systems. For § 50.46, the 

potential was identified for making riskinformed changes to requirements for 

both ECCS cooling performance and 

ECCS analysis acceptance criteria in 

§ 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 §§ 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 


sroberts on DSKGBLS3C1PROD with PROPOSALS
40766 Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 

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 

§§ 50.44 and 50.46. On September 16, 

2003, (68 FR 54123), the NRC amended 

§ 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 RiskInformed 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 § 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-attemperature below 1000 degrees 

Centigrade (°C) (1832 degrees 

Fahrenheit (°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 steamcladding reaction can enter the cladding 

inside surface near the rupture location. 

These research findings have been 

summarized in Research Information 

Letter (RIL) 0801, ‘‘Technical Basis for 

Revision of Embrittlement Criteria in 10 

CFR 50.46,’’ (ADAMS Accession 

number ML081350225) and the detailed 

experimental results from the program 

at ANL are contained in NUREG/CR– 

6967, ‘‘Cladding Embrittlement during 

Postulated Loss-of-Coolant Accidents’’ 

(ADAMS Accession number 

ML082130389). 

In response to the research findings 

identified in RIL 0801, the NRC 

completed a preliminary safety 

assessment of currently operating 

reactors (ADAMS Accession number 

ML090340073). This assessment found 

that due to realistic fuel rod power 

history, measured cladding performance 

under LOCA conditions, and current 

analytical conservatisms, sufficient 

safety margin exists for operating 

reactors. Therefore, any changes to the 

ECCS acceptance criteria to account for 

the new findings can reasonably be 

addressed through rulemaking. 

After the NRC publicly released the 

technical basis information in RIL 0801 

on May 30, 2008, and NUREG/CR–6967, 

on July 31, 2008, it published a Federal 

Register (FR) document on July 31, 

2008, (73 FR 44778), requesting that 

public stakeholders comment on the 

adequacy of the technical basis and 

identify issues that may arise with 

respect to experimental data 

development, regulatory costs, or 

impacts of potential new requirements. 

The comments received in response to 

this document can be found at http:// 

www.regulations.gov by searching on 

docket ID NRC–2008–0332. On 

September 24, 2008, the NRC held a 

public workshop to discuss stakeholder 

comments on the adequacy of the 

technical basis and to give the public 

and industry another opportunity to 

provide further comment and input. The 

workshop included presentations and 

open discussion between 

representatives of the NRC, 

international regulatory and research 

agencies, domestic and international 

commercial power firms, fuel vendors, 

and the general public. The meeting 

summary, including a list of attendees 

and presentations, is available at 

ADAMS Accession number 

ML083010496. 

Since 2002, the NRC has met with the 

Advisory Committee on Reactor 

Safeguards (ACRS) multiple times to 

discuss the progress of the LOCA 

research program and rulemaking 

proposals. Provided in the table below 

are the dates and ADAMS Accession 

numbers of the relevant ACRS meetings 

and associated correspondence. 

Date Meeting/letter ADAMS accession number 

October 9, 2002 .................................. Subcommittee Meeting .................................................................................. ML023030246 * 

October 10, 2002 ................................ Full Committee Meeting ................................................................................ ML022980190 * 

October 17, 2002 ................................ Letter from ACRS to NRC staff ..................................................................... ML022960640 

December 9, 2002 .............................. Response letter from NRC staff to ACRS ..................................................... ML023260357 

September 29, 2003 ........................... Subcommittee Meeting .................................................................................. ML032940296 * 


sroberts on DSKGBLS3C1PROD with PROPOSALS
Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 40767 

1Crud 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 cobalt60. 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. 

Date Meeting/letter ADAMS accession number 

July 27, 2005 ...................................... Subcommittee Meeting .................................................................................. ML052230093 * 

September 8, 2005 ............................. Full Committee Meeting ................................................................................ ML052710235 * 

January 19, 2007 ................................ Subcommittee Meeting .................................................................................. ML070390301 * 

February 2, 2007 ................................ Full Committee Meeting ................................................................................ ML070430485 * 

May 23, 2007 ...................................... Letter from ACRS to NRC staff ..................................................................... ML071430639 

July 11, 2007 ...................................... Response letter from NRC staff to ACRS ..................................................... ML071640115 

December 2, 2008 .............................. Subcommittee Meeting .................................................................................. ML083520501 * & 

ML083530449 

December 4, 2008 .............................. Full Committee Meeting ................................................................................ ML083540616 * 

December 18, 2008 ............................ Letter from ACRS to NRC staff ..................................................................... ML083460310 

January 23, 2009 ................................ Response letter from NRC staff to ACRS .................................................... ML0836640532 

* 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) 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 zirconiumclad fuel in order to ensure compliance 

with § 50.46(b) ECCS acceptance 

criteria; 

(2) Amend Appendix K to 10 CFR part 

50 to explicitly require that the steadystate 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 § 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 § 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 

§ 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 § 50.46 for these 

potential new non-zirconium cladding 

materials. 

Objective 2: Establish performancebased 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 § 50.46 by 

replacing the current § 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 § 50.46(a)(3)(i) to emphasize 

the importance of reporting reduction in 

margins to the acceptance criteria and 


sroberts on DSKGBLS3C1PROD with PROPOSALS
40768 Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 

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 § 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 § 50.46 to include any light-water 

reactor fuel cladding material: 

This first conceptual approach 

involves the applicability of the rule as 

defined in § 50.46(a)(1)(i). Currently, 

this provision is limited to fuel rods 

clad in Zircaloy or ZIRLOTM. The recent 

LOCA research program conducted 

testing on a wide range of zirconiumbased 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, § 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 § 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 § 50.46(b)(3) Maximum 

hydrogen generation, § 50.46(b)(4) 

Coolable geometry, and § 50.46(b)(5) 

Long-term cooling. The § 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 nonzirconium 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 performancebased 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 (§ 50.46(b)(1)) and local 

cladding oxidation (§ 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 °F (1204 °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 CathcartPawel (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-attemperature) 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 § 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 °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 


sroberts on DSKGBLS3C1PROD with PROPOSALS
Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 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-attemperature (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-attemperature (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, alloyspecific 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 °F to 1875 °F (649 °C to 1024 

°C), shall remain greater than the 

calculated duration that cladding 

surface temperature anywhere on the 

fuel rod remains above 1200 °F (649 °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 °F (649 °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 °F (649 °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 °F (28 °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 °F (28 °C). 


sroberts on DSKGBLS3C1PROD with PROPOSALS
40770 Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 

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 °F (56 °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 °F (56 °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 °F. 

Calculated PCT Reporting trigger 

< 2090 (i.e., not within 5 percent of 

2200 °F limit).

Any change ≥ 100 °F. 

2090–2099 °F ........... Any change ≥ 11 °F. 

2100–2109 °F ........... Any change ≥ 10 °F. 

2110–2119 °F ........... Any change ≥ 9 °F. 

2120–2129 °F ........... Any change ≥ 8 °F. 

2130–2139 °F ........... Any change ≥ 7 °F. 

2140–2149 °F ........... Any change ≥ 6 °F. 

2150–2159 °F ........... Any change ≥ 5 °F. 

2160–2169 °F ........... Any change ≥ 4 °F. 

2170–2179 °F ........... Any change ≥ 3 °F. 

2180–2189 °F ........... Any change ≥ 2 °F. 

2190–2199 °F ........... Any change ≥ 1 °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 §§ 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 zirconiumclad fuel in order to ensure compliance 

with § 50.46(b) ECCS acceptance 

criteria; 

(2) Amend Appendix K to 10 CFR part 

50 to explicitly require that the steadystate 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 § 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 


sroberts on DSKGBLS3C1PROD with PROPOSALS
Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 40771 

2AOA means Axial Offset Anomaly. 

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 

§ 50.46(b) on peak clad temperature 

(2200 °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 

§ 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 

§ 50.46(b). The NRC reviews and 

approves all licensee evaluation models 

to determine if they are acceptable. 

For best-estimate evaluation models, 

§ 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 lossof-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. 

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 § 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 


sroberts on DSKGBLS3C1PROD with PROPOSALS
40772 Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 

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, § 50.46(a)(3)(i) requires licensees 

to determine if the change or error is 

significant. For significant changes, 

§ 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 § 50.46 requirements. In 

situations when the § 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 § 50.46(b) would be 

reportable events as described in 

§§ 50.55(e), 50.72, and 50.73. In these 

situations, the licensee would be 

required under § 50.46(a)(3)(ii) to 

propose immediate steps to demonstrate 

compliance or bring the plant design or 

operation into compliance with § 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 § 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 

§ 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 § 50.46 to all fuel cladding materials. 

Should the rule be expanded to include 

any cladding material, or only be 

expanded to include all zirconiumbased 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 

§ 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 °F (649 °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 hydrogendependent 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 hotcell 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? 


sroberts on DSKGBLS3C1PROD with PROPOSALS
Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 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 prehydrided 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 °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 § 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 § 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 

§ 50.46(a)(3) to resolve the discrepancy. 

The NRC also requests comment on the 

imposition of a requirement that one or 

more fuel assemblies be inspected at the 

end of each fuel cycle to demonstrate 

the validity of crud levels analyzed in 

the ECCS model. 

11. What information exists to 

facilitate developing an acceptable crud 

deposition model that could correlate 

crud deposition with measured primary 

water coolant chemistry (e.g., iron-oxide 

concentration)? For boiling water 

reactors, it is difficult to perform visual 

inspections or poolside measurements 

of fuel rod crud thickness without first 

removing the channel box. A crud 

deposition model would facilitate the 

confirmation of design crud layers 

assumed in the ECCS evaluations and 

provide an indicator to reactor operators 

when crud levels approach unanalyzed 

conditions. Are there ongoing or 

planned industry efforts to monitor 

water coolant chemistry for comparison 

to observed crud deposition? If so, what 

amount of success has been obtained? 

Could a properly correlated crud model 

be sufficiently accurate to preclude the 

need for crud measurements at the end 

of each fuel cycle? 

Cost Considerations 

12. The U.S. commercial nuclear 

power industry claims that 

implementation of the proposed rule 

would be a significant burden in both 

money and resources. The industry has 

discussed an implementation cost of 

approximately $250 million (NRC– 

2008–0332–0008.1 at http:// 

www.regulations.gov). 

a. What options are available to 

reduce this implementation cost? 

b. Are there changes in core operating 

limits, fuel management, or cladding 

material that would reduce the cost and 

burden of implementing the proposed 

hydrogen based PQD criterion without 

negatively impacting operations? 

c. A staged implementation would be 

more manageable for both the NRC and 

industry. One potential approach 

involves characterizing the plants based 

upon safety margin and deferring 

implementation for the licensees with 

the largest safety margin (e.g., lowest 

calculated CP–ECR). The NRC requests 

comment on this implementation 

approach. 

Available Supporting Documents 

The following documents provide 

additional background and supporting 

information regarding this rulemaking 

activity and corresponding technical 

basis. The documents can be found in 

the NRC’s Agencywide Document 

Access and Management System 

(ADAMS). Instructions for accessing 

ADAMS were provided under the 

ADDRESSES section of this document. 


sroberts on DSKGBLS3C1PROD with PROPOSALS
40774 Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 

Date Document ADAMS accession number 

July 31, 2008 ..................................................... NUREG/CR–6967, ‘‘Cladding Embrittlement 

During Postulated Loss-of-Coolant Accidents’’.

ML082130389. 

May 30, 2008 ..................................................... Research Information Letter (RIL) 0801, 

‘‘Technical Basis for Revision of Embrittlement Criteria in 10 CFR 50.46’’.

ML081350225. 

September 24, 2008 .......................................... Public Meeting Summary ................................. ML083010496.

February 23, 2009 ............................................. Plant Safety Assessment of RIL 0801 ............. ML090340073. 

July 31, 2008 ..................................................... Federal Register Notice (73 FR 44778), ‘‘Notice of Availability and Solicitation of Public 

Comments on Documents Under Consideration To Establish the Technical Basis for 

New Performance-Based Emergency Core 

Cooling System Requirements’’.

Reference the Federal Register at 73 FR 

44778. 

March 30, 2009 ................................................. Supplemental research material—additional 

PQD tests.

ML090690711. 

March 30, 2009 ................................................. Supplemental research material—additional 

breakaway testing.

ML090700193. 

March 31, 2009 ................................................. Draft proposed procedure for Conducting Oxidation and Post-Quench Ductility Tests With 

Zirconium-based Cladding Alloys.

ML090900841. 

March 23, 2009 ................................................. Draft proposed procedure for Conducting 

Breakaway Oxidation Tests With ZirconiumBased Cladding Alloys.

ML090840258. 

January 8, 2009 ................................................. Update on Breakaway Oxidation of Westinghouse ZIRLO Cladding.ML091330334. 

May 7, 2009 ....................................................... Impact of Specimen Preparation on Breakaway Oxidation (Non-Proprietary).ML091350581. 

List of Subjects in 10 CFR Part 50 

Antitrust, Classified information, 

Criminal penalties, Fire protection, 

Intergovernmental relations, Nuclear 

power plants and reactors, Radiation 

protection, Reactor siting criteria, 

Reporting and recordkeeping 

requirements. 

The authority citation for this document is 

42 U.S.C. 2201. 

Dated at Rockville, MD, this 29th day of 

July 2009. 

For the Nuclear Regulatory Commission. 

R.W. Borchardt, 

Executive Director for Operations. 

APPENDIX A 

An Approach for Determining the 

Acceptability of Zirconium-Based Cladding 

Alloys: Analytical Limits Defined Within 

Regulatory Guidance 

This approach would include a future 

regulatory guidance document that defines 

an acceptable, generically-applicable set of 

analytical limits for peak cladding 

temperature and maximum allowable timeat-temperature (expressed as calculated local 

oxidation, CP–ECR) as a function of pretransient hydrogen content in the cladding 

metal (excluding hydrogen in the cladding 

oxide layer). These acceptable analytical 

limits would be developed using NRC’s 

empirical database with consideration of 

experimental variability and repeatability. 

Figure A shows the results of ringcompression tests conducted on asfabricated, hydrogen charged, and irradiated 

specimens of Zircaloy-2, Zircaloy-4, 

ZIRLOTM and M5 cladding material 

(documented in NUREG/CR–6967). Note that 

hydrogen concentrations were slightly 

adjusted (± 5 wppm) to illustrate results of 

multiple ring-compression tests run at the 

same CP–ECR and hydrogen concentration. 

Peak oxidation temperature is identified for 

samples tested below 2200 °F (1204 °C). 


sroberts on DSKGBLS3C1PROD with PROPOSALS
Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 40775 

The analytical limit on PCT would be 

restricted to the peak oxidation temperature 

during testing of the cladding specimens 

used in the development of this limit. 

Furthermore, caveats on the applicability of 

the analytical limits may be required to 

capture limiting aspects of the steam 

oxidation temperature profile used during 

the testing. For example, if the calculated 

time at the specified PCT is less than the time 

at peak oxidation temperature of the 

supporting empirical database (for a given 

CP–ECR), or the calculated quench 

temperature is lower than 800 °C, then an 

applicability caveat may be required. 

Existing Cladding Alloys 

No PQD testing would be required to 

approve cladding alloys included in the 

NRC’s LOCA research program. Under this 

approach, a fuel vendor would submit a 

topical report (TR) seeking NRC approval of 

each zirconium-based cladding alloy’s 

analytical limits on PCT and time-attemperature (CP–ECR, as a function of 

cladding hydrogen content). The TR would 

reference the acceptable analytical limits 

within the Regulatory Guide. 

New Cladding Alloys 

Under this approach, a fuel vendor would 

submit a TR which demonstrates that the 

results of PQD tests on a specific new alloy 

are applicable to the acceptable analytical 

limits defined within the Regulatory Guide. 

A TR would need to include the results of 

testing, conducted in accordance with NRC’s 

acceptable experimental methodology, which 

demonstrates that the embrittlement behavior 

of the new cladding alloy is consistent with 

the embrittlement behavior of the cladding 

alloys tested in NRC’s LOCA research 

program by comparing test results to the 

defined analytical limit. This would likely 

require testing of the new cladding alloy with 

varying hydrogen contents, which are 

oxidized to calculated oxidation levels (CP– 

ECR) at or near the analytical limit for that 

hydrogen level as provided in regulatory 

guidance. Demonstrating ductile behavior in 

cladding samples with calculated oxidation 

levels at or near the analytical limit may 

serve to confirm the applicability of the 

analytical limit to a new cladding alloy. The 

range of hydrogen contents in test samples 

required may be limited by proposing 

cladding hydrogen design limits based on hot 

cell examinations of irradiated samples of the 

new cladding alloy following lead test 

assembly campaigns. Regulatory guidance 

would be provided to address the variability 

in measured offset strain of ring-compression 

test results. Section IV of this ANPR 

specifically seeks comment on the treatment 

of variability in ductility measurements of 

ring-compression tests. 

For this description, it is assumed that 

sufficient justification for the use of hydrogen 

charged cladding specimens has been 

accepted as a surrogate for testing on 

irradiated cladding segments. If sufficient 

justification for the use of hydrogen charged 

cladding specimens has not been accepted as 

a surrogate for testing on irradiated cladding 

segments, approving new cladding alloys 

would require PQD testing of irradiated 

material. Section IV of this ANPR requests 

information on any ongoing or planned 

testing aimed at validating this pre-hydrided 

surrogate. 

APPENDIX B 

An Approach for Determining the 

Acceptability of Zirconium-Based Cladding 

Alloys: Cladding-Specific Analytical Limits 

Defined by an Applicant 

This approach involves establishing 

cladding-specific and/or temperature-specific 

analytical limits for peak cladding 

temperature and maximum allowable timeat-temperature (expressed as calculated local 

oxidation, CP–ECR) as a function of pretransient hydrogen content in the cladding 

metal (excludes hydrogen in the cladding 

oxide layer). This approach would provide 

optimum flexibility for defining more 

specific analytical limits to gain margin to 

the ECCS performance criteria. However, 

unlike citing analytical limits within a 

regulatory guide, this approach places the 

burden of proof on the applicant to validate 

their analytical limits and address 


sroberts on DSKGBLS3C1PROD with PROPOSALSEP13AU09.025</GPH>
40776 Federal Register / Vol. 74, No. 155 / Thursday, August 13, 2009 / Proposed Rules 

experimental variability and repeatability. As 

a result, this approach would necessitate a 

larger number of PQD tests (relative to 

confirming the applicability of the regulatory 

guide). Analytical limits, along with the 

experimental procedures, protocols, and 

specimen test results used in their 

development, would be subject to NRC 

review and approval. 

This approach would require that the PQD 

test results on irradiated cladding segments 

documented in NUREG/CR–6967 be 

considered in the development of analytical 

limits. Deviations in cladding performance 

relative to this empirical database must be 

identified and dispositioned. 

Existing Cladding Alloys 

In the case of existing cladding alloys, the 

rule may specify the following performance 

requirement to ensure an adequate retention 

of cladding ductility: 

Accumulation of ≥ 1.00 percent permanent 

strain prior to failure during ringcompression loading at a temperature of 135 

°C and a displacement rate of 0.033 mm/sec 

on a cladding specimen exposed to doublesided steam oxidation up to a specified peak 

oxidation temperature and CP–ECR. 

Analytical limits on allowable time-attemperature (CP–ECR) and peak cladding 

temperature would need to be defined as a 

function of initial cladding hydrogen content 

(wppm in metal) to demonstrate this 

performance requirement is met. A topical 

report (TR) would be generated to document 

the basis for the new analytical limits. 

Existing alloys which were included in the 

NRC high-burnup research program may 

reference the test results documented in 

NUREG/CR–6967 in the development of new 

analytical limits. This data was generated 

following experimental protocols acceptable 

to the NRC, so no further justification related 

to its validity would be required. 

Using an approved hydrogen uptake model 

for an existing cladding alloy, the TR would 

provide the methodology to convert the 

hydrogen-based analytical limits to some unit 

of measure more readily applied within 

reload safety analyses (e.g., fuel rod burnup 

or fuel duty). Uncertainties related to 

hydrogen uniformity and uncertainties 

introduced by the conversion from hydrogen 

to another unit of measure would need to be 

addressed. 

New Cladding Alloys 

In the case of new cladding alloys, the rule 

may specify the following performance 

requirement to ensure an adequate retention 

of cladding ductility: 

Accumulation of ≥ 1.00 percent permanent 

strain prior to failure during ringcompression loading at a temperature of 135 

°C and a displacement rate of 0.033 mm/sec 

on a cladding specimen exposed to doublesided steam oxidation up to a specified peak 

oxidation temperature and CP–ECR. 

Analytical limits on allowable time-attemperature (CP–ECR) and peak cladding 

temperature would need to be defined as a 

function of initial cladding hydrogen content 

(wppm in metal) to demonstrate this 

performance requirement is met. A TR would 

be generated to document the basis for the 

new analytical limits. The PQD test results 

on irradiated cladding segments documented 

in NUREG/CR–6967 would need to be 

considered in the development of analytical 

limits. PQD testing would be required to (1) 

establish analytical limits in accordance with 

the performance requirements that would be 

specified within the rule, and (2) 

demonstrate the applicability of the NUREG/ 

CR–6967 empirical database. A TR could 

document that the PQD testing had been 

conducted to strictly adhere to the accepted 

experimental protocols documented in 

regulatory guidance documents, or if 

alternative testing procedures were used, 

then NRC review and approval of those 

laboratory procedures would be required. 

For this approach, defining analytical 

limits for new cladding alloys would likely 

require testing at a range of hydrogen 

contents, with ring-compression test results 

at multiple calculated oxidation levels. Test 

samples with calculated oxidation levels 

sufficient to display brittle behavior, as well 

as test samples with calculated oxidation 

levels which display ductile behavior, would 

be necessary to define the transition from 

ductile to brittle behavior. Regulatory 

guidance would be provided to address the 

variability in measured offset strain of ringcompression test results. Section IV of this 

ANPR specifically seeks comment on the 

treatment of variability in ductility 

measurements of ring-compression tests. The 

range of hydrogen contents in test samples 

required may be limited by proposing 

cladding hydrogen design limits based on hot 

cell examinations of irradiated samples of the 

new cladding alloy following lead test 

assembly campaigns. 

Multifaceted Analytical Limits 

Recognizing that higher burnup fuel rods 

(with higher hydrogen concentrations) 

operate at a reduced power level (relative to 

lower burnup fuel rods), defining analytical 

limits for maximum allowable ECR at 

multiple peak oxidation temperatures would 

also be possible. For example, a TR could 

document the results of testing conducted at 

peak oxidation temperatures of 2200 °F (1204 

°C), 2000 °F (1093 °C), and 1800 °F (982 °C), 

which are targeted at low burnup (low 

corrosion), medium burnup (medium 

corrosion), and high burnup (high corrosion) 

fuel rods, respectively. Testing to support 

these new limits would require testing at a 

range of hydrogen contents, with ringcompression test results at multiple 

calculated oxidation levels to define the 

transition from ductile to brittle behavior. In 

this case, it may be necessary to elect to 

strictly adhere to the accepted experimental 

protocols documented in regulatory guidance 

documents, thereby limiting regulatory 

exposure related to testing procedures and 

the validity of the data. 

Implementation of the multifaceted 

analytical limits would require separating all 

of the fuel rods in the core into three 

categories and then ensuring that all fuel rods 

within each category satisfies their respective 

analytical limits on both CP–ECR and PCT. 

While it is anticipated that this approach 

would provide flexibility, it would also 

necessitate a more complex LOCA analysis 

and reload-by-reload confirmation. This 

approach also relies on tacit assumptions 

regarding the currently approved LOCA 

model’s ability to accurately simulate the 

thermal-hydraulic conditions in every region 

of the reactor core (as opposed to simulating 

a core average response or pseudo hot 

channel location). Modeling uncertainties 

with respect to predicting local conditions 

throughout the reactor core would need to be 

addressed. 

Using an approved hydrogen uptake model 

for a new cladding alloy, the TR would need 

to provide the methodology to convert the 

hydrogen-based analytical limits to some unit 

of measure more readily applied within 

reload safety analyses (e.g., fuel rod burnup 

or fuel duty). Uncertainties related to 

hydrogen uniformity and uncertainties 

introduced by the conversion from hydrogen 

to another unit of measure would need to be 

addressed. 

For this description, it is assumed that 

sufficient justification for the use of hydrogen 

charged cladding specimens has been 

accepted as a surrogate for testing on 

irradiated cladding segments. If sufficient 

justification for the use of hydrogen charged 

cladding specimens has not been accepted as 

a surrogate for testing on irradiated cladding 

segments, approving new cladding alloys 

would require PQD testing of irradiated 

material. Section IV of this ANPR requests 

information on any ongoing or planned 

testing aimed at validating this pre-hydrided 

surrogate. 

[FR Doc. E9–19423 Filed 8–12–09; 8:45 am] 

BILLING CODE 7590–01–P 

DEPARTMENT OF TRANSPORTATION 

Federal Aviation Administration 

14 CFR Part 39 

[Docket No. FAA–2009–0713; Directorate 

Identifier 2007–NM–303–AD] 

RIN 2120–AA64 

Airworthiness Directives; Airbus Model 

A318 Series Airplanes 

AGENCY: Federal Aviation 

Administration (FAA), DOT. 

ACTION: Notice of Proposed Rulemaking 

(NPRM). 

SUMMARY: We propose to adopt a new 

airworthiness directive (AD) for the 

products listed above. This proposed 

AD results from mandatory continuing 

airworthiness information (MCAI) 

originated by an aviation authority of 

another country to identify and correct 

an unsafe condition on an aviation 

product. The MCAI describes the unsafe 

condition as: 

Some operators have reported airframe 

vibration under specific flight conditions 

including gusts. 

Investigations have revealed that under 

such conditions, vibrations may occur when 


sroberts on DSKGBLS3C1PROD with PROPOSALS