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

4346 

Vol. 74, No. 15 

Monday, January 26, 2009 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 50 

[NRC–2007–0016; PRM–50–87] 

Raymond A. Crandall; Denial of 

Petition for Rulemaking 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Denial of petition for 

rulemaking. 

SUMMARY: The Nuclear Regulatory 

Commission (NRC) is denying the 

petition for rulemaking (PRM) filed by 

Mr. Raymond A. Crandall on May 17, 

2007, and docketed on June 22, 2007 

(Docket No. PRM–50–87). In his 

petition, the petitioner requested that 

the NRC amend the regulations that 

govern domestic licensing of production 

and utilization facilities to eliminate the 

specific criteria related to the 

radiological doses for control room 

habitability at nuclear power plants. 

The petitioner stated that the current 

deterministic radiological dose 

requirements for control room 

habitability have resulted in several 

negative safety consequences, including 

an increased risk to public safety. He 

requested that the NRC delete the 5 rem 

whole body dose limit and the 0.05 

sievert (Sv) (5 rem) total effective dose 

equivalent (TEDE) limit specified in the 

current regulations. 

DATES: The docket for PRM–50–87 is 

closed as of January 26, 2009. 

ADDRESSES: Publicly available 

documents related to this petition, 

including the PRM and the NRC’s letter 

of denial to the petitioner may be 

viewed using the following methods: 

Federal e-Rulemaking Portal: Go to 

http://www.regulations.gov and search 

for documents related to this PRM filed 

under docket ID NRC–2007–0016. 

NRC’s Public Document Room (PDR): 

The public may examine publicly 

available documents and have them 

copied for a fee at the NRC’s PDR, 

Public File Area O–1 F21, One White 

Flint North, 11555 Rockville Pike, 

Rockville, Maryland. 

NRC’s Agencywide Document Access 

and Management System (ADAMS): 

Publicly available documents created or 

received at the NRC are available 

electronically via 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 the NRC’s public documents. If 

you do not have access to ADAMS or 

have any problems in accessing the 

documents located in ADAMS, contact 

the NRC PDR Reference staff at 1–800– 

397–4209, or 301–415–4737, or by email to PDR.resource@nrc.gov. 

FOR FURTHER INFORMATION CONTACT: A. 

Jason Lising, Office of Nuclear Reactor 

Regulation, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, telephone: (301) 415–3220, or tollfree: 800–368–5642; e-mail: 

Jason.Lising@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Background 

II. Petitioner’s Requests 

III. Reasons for Denial 

IV. Public Comments 

V. Denial of Petitions 

I. Background 

On May 17, 2007, the NRC received 

a PRM from Raymond A. Crandall 

(ADAMS Accession No. ML071490250); 

the PRM was docketed by the NRC as 

PRM–50–87. The petitioner requested 

that the NRC amend Title 10 of the Code 

of Federal Regulations Part 50 (10 CFR 

Part 50), ‘‘Domestic Licensing of 

Production and Utilization Facilities’’ to 

remove the specific criteria related to 

the radiological doses for control room 

habitability at nuclear power plants 

from 10 CFR 50.67, ‘‘Accident source 

term,’’ and General Design Criterion 

(GDC) 19, ‘‘Control room,’’ in Appendix 

A, ‘‘General Design Criteria for Nuclear 

Power Plants,’’ to 10 CFR Part 50. The 

NRC published a notice of receipt and 

request for public comment in the 

Federal Register on July 12, 2007 (72 FR 

38030). The 75-day public comment 

period ended on September 25, 2007. 

The petitioner noted that the current 

regulations provide specific dose 

criteria for demonstrating the 

acceptability of the control room design 

during radiological release events. 

These criteria are based on deterministic 

radiological dose analyses performed by 

the licensee and reviewed by the NRC. 

NRC regulatory guides and standard 

review plans provide acceptable 

methodologies that can be used by 

licensees to perform dose analyses, 

which are then incorporated, as 

appropriate, into the licensing basis for 

the licensee’s facility. The petitioner 

stated that the deterministic dose 

analysis methodology and associated 

regulatory process result in several 

negative safety consequences: 

(1) Current Designs Not Optimum 

‘‘Control room designs that are not 

optimum for ensuring continued control 

room habitability. Current designs 

required in order to meet the current 

dose methodology criteria may actually 

increase the probability of having to 

evacuate the control room compared to 

establishing the design based on good 

engineering principles.’’ 

(2) Procedures Not Optimized 

‘‘Site procedures for mitigation of the 

dose consequences to control room 

personnel that are not optimum for 

ensuring control room habitability. The 

procedures designed to ensure 

consistency with the dose analysis 

assumptions are inconsistent with more 

effective mitigation strategies.’’ 

(3) Challenges to Safety Systems 

‘‘Unnecessary challenges to safety 

systems, such as increased challenges to 

the Emergency Diesel Generators if 

control room ventilation system fans are 

loaded on the diesels early in the 

accident to meet analysis assumptions.’’ 

(4) Inappropriate Technical 

Specification (TS) Action Statements 

‘‘Technical Specifications Action 

Statement requirements that result in a 

net increase in the risk to the public. 

This specifically refers to Technical 

Specifications that require a plant 

shutdown for failure to meet a control 

room dose analysis input assumption.’’ 

(5) Unjustified Technical 

Specification Surveillances 

‘‘Technical Specifications 

Surveillance requirements that cannot 

be cost-justified based on the risksignificance. This results in the required 

expenditure of resources that could be 

used on risk-significant improvements.’’ 

The petitioner suggested amendments 

that would eliminate the specific 

radiological dose acceptance criteria 

and, thereby, the need for deterministic 

dose analyses and the associated 

regulatory processes, including the need 

for applicable TSs. He stated that the 

Federal Register / Vol. 74, No. 15 / Monday, January 26, 2009 / Proposed Rules 4347 

proposed changes would not eliminate 

the requirement for the control room to 

be designed to ensure safe conditions 

under accident conditions, but it would 

address his safety concerns with the 

current regulations. 

II. Petitioner’s Request 

In PRM–50–87 the petitioner 

requested that the NRC take the 

following actions: 

1. Revise the regulations related to 

control room habitability at nuclear 

power plants by deleting the following 

sentences from GDC 19: 

‘‘Adequate radiation protection shall 

be provided to permit access and 

occupancy of the control room under 

accident conditions without personnel 

receiving radiation exposures in excess 

of 5 rem whole body, or its equivalent 

to any part of the body, for the duration 

of the accident. Applicants for and 

holders of construction permits and 

operating licenses under this part who 

apply on or after January 10, 1997, 

applicants for design certifications 

under part 52 of this chapter who apply 

on or after January 10, 1997, applicants 

for and holders of combined licenses 

under part 52 of this chapter who do not 

reference a standard design certification, 

or holders of operating licenses using an 

alternative source term under § 50.67, 

shall meet the requirements of this 

criterion, except that with regard to 

control room access and occupancy, 

adequate radiation protection shall be 

provided to ensure that radiation 

exposures shall not exceed 0.05 Sv (5 

rem) total effective dose equivalent 

(TEDE) as defined in § 50.2 for the 

duration of the accident.’’ 

2. Revise the regulations related to 

control room habitability at nuclear 

power plants to delete from paragraph 

(b)(2)(iii) in 10 CFR 50.67 this language: 

‘‘Adequate radiation protection is 

provided to permit access to and 

occupancy of the control room under 

accident conditions without personnel 

receiving radiation exposures in excess 

of 0.05 Sv (5 rem) total effective dose 

equivalent (TEDE) for the duration of 

the accident.’’ 

III. Reasons for Denial 

1. General 

The NRC has reviewed Mr. Raymond 

Crandall’s petition and has determined 

that it does not provide adequate 

justification to remove the control room 

radiological dose acceptance criteria 

from NRC regulations. The NRC does 

not agree with the petitioner’s assertion 

that the control room radiological dose 

acceptance criteria have resulted in 

negative safety consequences. 

Performance-based regulations, such 

as § 50.67 and Appendix A to 10 CFR 

Part 50, do not provide prescriptive 

requirements and, therefore, do not 

require licensees to use specific designs 

or methodologies to comply with the 

regulations. The NRC, however, does 

provide regulatory guidance to licensees 

that includes acceptable designs and 

methodologies for demonstrating 

compliance with the regulations. The 

use of the guidance is optional, and 

licensees are free to propose alternative 

means of complying with the NRC’s 

regulations. 

Design-basis dose consequence 

analyses are intentionally based upon 

conservative assumptions and are 

intended to model the potential hazards 

that would result from any credible 

accident, not necessarily the most 

probable accident. As stated in footnotes 

to 10 CFR 100.11, ‘‘Determination of 

exclusion area, low population zone, 

and population center distance,’’ and 10 

CFR 50.67, ‘‘Accident source term,’’ 

‘‘[t]he fission product release assumed 

for these calculations should be based 

upon a major accident, hypothesized for 

purposes of site analysis or postulated 

from considerations of possible 

accidental events, that would result in 

potential hazards not exceeded by those 

from any accident considered credible. 

Such accidents have generally been 

assumed to result in substantial 

meltdown of the core with subsequent 

release of appreciable quantities of 

fission products.’’ 

The performance-based control room 

dose criterion is designed to maintain 

an acceptable level of control room 

habitability even under the maximum 

credible accident scenario. The NRC has 

determined that providing an acceptable 

level of control room habitability for 

design-basis events is necessary to 

provide reasonable assurance that the 

control room will continue to be 

effectively manned and operated to 

mitigate the effects of the accident and 

protect public health and safety. 

Meeting or exceeding the design-basis 

control room dose limit would not 

impose an immediate evacuation 

requirement on the control room 

operators. Moreover, by removing the 5 

rem acceptance criterion, a regulatory 

basis for the acceptance of the 

radiological protection aspects of 

control room designs would no longer 

exist and would not support the 

Commission’s policy regarding 

performance-based regulations. 

The conservative assumptions used in 

design-basis dose consequence analyses 

need not and should not form the basis 

for restricting actions described in 

emergency operating procedures. These 

procedures are designed to ensure that 

during an accident all available means 

are used to assess actual radiological 

conditions and to maintain emergency 

worker doses As Low As Reasonably 

Achievable (ALARA), as required by 10 

CFR Part 20, ‘‘Standards For Protection 

Against Radiation.’’ Additionally, no 

NRC regulations, including 10 CFR Part 

20, ‘‘Standards for Protection Against 

Radiation,’’ require evacuation of the 

control room when the design-basis 

control room dose limit is exceeded. 

Emergency operating procedures 

include guidance for controlling doses 

to workers under emergency conditions. 

This guidance would be applicable in 

the unlikely event that control room 

doses were projected to exceed the 

design-basis dose limit during an actual 

emergency. 

2. NRC Staff Responses to the 

Petitioner’s Assertions 

A. Current Designs Are Not Optimum 

1. The petitioner stated that because 

the primary objective of control room 

habitability is to ensure continuous 

occupancy, the primary focus should be 

on minimizing whole body doses from 

noble gases. He stated that some 

common control room designs, such as 

the filtered air intake pressurization 

design, focus on compliance with 

existing dose criteria. He concluded that 

the current requirements and 

operational criteria focus on minimizing 

the thyroid dose at the expense of 

increasing the whole body dose from 

noble gases which increases the 

probability that the control room will 

require evacuation. 

The NRC reviewed the petitioner’s 

concern regarding the increase in whole 

body dose from noble gases, which he 

believes results from the intentional 

intake of filtered air into the control 

room under design-basis accident (DBA) 

conditions. The NRC agrees that a 

relatively small increase in whole body 

dose due to noble gases may result from 

the intake of filtered air into the control 

room. However, this small increase in 

dose would not increase the probability 

of a control room evacuation. Therefore, 

operators would be able to monitor 

plant indications and take appropriate 

accident mitigating actions from the 

control room, and there would be no 

increase in risk to public health and 

safety. The NRC’s conclusion is based 

on a review of several existing DBA 

control room dose analyses that 

determined the impact on whole body 

dose resulting from filtered air intake 

pressurization to the control room. The 

NRC performed parametric evaluations 

and determined that while filtered air 

4348 Federal Register / Vol. 74, No. 15 / Monday, January 26, 2009 / Proposed Rules 

intake pressurization may result in a 

small addition to the control room 

whole body dose from noble gases, the 

increase is more than offset by the 

reduction in thyroid dose and TEDE 

from inhalation of radioactive 

particulates, such as iodine. 

Based upon its analyses, the NRC 

does not agree with the petitioner’s 

assertion regarding the negative safety 

impact of providing filtered intake flow 

into the control room. The NRC’s 

performance-based criterion in GDC 19 

requires that an applicant provide a 

control room habitability design that 

meets the specified dose criterion. 

Although NRC regulatory guidance 

provides examples of acceptable design 

approaches, the approach used to meet 

the criterion is largely under the control 

of an applicant. In order to meet this 

requirement, many licensees have 

chosen to incorporate filtered air intake 

pressurization into their control room 

emergency ventilation designs to reduce 

the cumulative dose to operators during 

a DBA. The purpose of providing 

filtered air intake pressurization flow is 

to establish positive pressure in the 

control room relative to the adjacent 

areas, thereby reducing the quantity of 

unfiltered air inleakage. Limiting 

unfiltered inleakage significantly 

reduces the thyroid dose from 

inhalation. 

2. The petitioner also stated that the 

current regulation is inconsistent with 

the goal of allowing operators to remain 

in the control room in order to mitigate 

accident consequences. He stated that 

common designs, such as a filtered air 

intake pressurization system, which 

focus on compliance with existing 

criteria, increase the probability that the 

control room will have to be evacuated. 

The 5 rem control room design 

criterion is not a maximum integrated 

dose above which control room 

evacuation is mandated during an 

accident. Rather, the criterion provides 

a design basis to ensure that the control 

room will maintain a habitable 

environment for operators to control the 

plant during a DBA. 

The petitioner based his assertion on 

the assumption that filterable activity is 

not likely to be a significant contributor 

to dose in a reactor accident. As an 

example, the petitioner used the March 

1979 Three Mile Island Unit 2 accident. 

Since the accident, the NRC has 

expended considerable resources to 

better define the expected quantity and 

distribution of activity that could be 

released during a major reactor accident. 

As a result of this research, the NRC 

promulgated 10 CFR 50.67 on December 

23, 1999 (64 FR 72001). Under 10 CFR 

50.67, a licensee can apply for a license 

amendment to adopt an alternative 

source term (AST) that reflects a more 

realistic assessment of the timing of the 

release and the quantity and 

distribution of activity that could be 

released during a major accident 

hypothesized for purposes of design 

analyses. Many licensees have used this 

approach to comply with NRC 

regulations governing control room 

dose. 

In addition, 10 CFR 50.67 revised the 

control room dose criterion from a 5 rem 

whole body dose, or its equivalent to 

any organ, to a 5 rem TEDE. The 

relatively low thyroid organ weighting 

factor, as defined in 10 CFR 20.1003, 

‘‘Definitions,’’ and used in the 

calculation of TEDE, allows for a 

significant reduction in the controlling 

aspects of the thyroid dose, which 

normally governed compliance with 

control room dose guidelines. The NRC 

has significantly improved the accuracy 

of the source term and dose 

methodology used in design-basis dose 

consequence analyses. The updated 

source term and dose methodology 

address the petitioner’s concerns 

regarding the emphasis on thyroid dose 

in control room habitability analyses. 

3. The petitioner noted that the dose 

from increased iodine concentration can 

be mitigated by use of potassium iodide 

(KI) or respiratory protection, but the 

current regulations do not permit these 

mitigation measures to be used in 

design analyses. 

The NRC agrees that KI or SelfContained Breathing Apparatuses 

(SCBAs) do have merit as short-term 

compensatory measures. However, the 

potential medical complications of KI 

and the potential adverse impacts to 

human performance of SCBAs make 

these measures unsuitable for long-term 

use. Further, the NRC’s policy of 

ensuring that process or other 

engineering controls are in place instead 

of relying on the use of personal 

protective equipment is clearly set forth 

in 10 CFR 20.1701, ‘‘Use of process or 

other engineering controls’’ and 10 CFR 

20.1702, ‘‘Use of other controls.’’ This 

policy is consistent with the 

recommendations of international and 

national radiation protection 

committees as described in Paragraph 

167 of the International Commission on 

Radiological Protection (ICRP) 

Publication 26. 

Paragraph 167 of ICRP Publication 26 

recommends that ‘‘[a]s far as is 

reasonably practicable, the 

arrangements for restricting 

occupational exposure should be 

applied to the source of radiation and to 

features of the workplace. The use of 

personal protective equipment should 

in general be supplementary to these 

more fundamental provisions. The 

emphasis should thus be on intrinsic 

safety in the workplace and only 

secondarily on protection that depends 

on the worker’s own actions,’’ such as 

the ingestion of KI or use of respiratory 

equipment. Further, the use of 

respiratory equipment by control room 

personnel during an emergency 

condition would impede the 

performance of functions necessary for 

the protection of public health and 

safety. Therefore, the NRC has not 

permitted licensees to rely on either KI 

or respiratory protection as a permanent 

solution to demonstrate compliance 

with the control room radiological dose 

guidelines, although such measures are 

available if the fundamental dose design 

provisions are less effective than 

anticipated. 

4. The petitioner stated that it is 

inconsistent to provide credit for 

respiratory protection in control room 

habitability toxic gas release 

evaluations, but not for design analyses. 

The NRC does not agree with the 

petitioner. In the case of toxic gas 

releases, continued plant operation or a 

normal plant shutdown would be 

required. In the case of a major reactor 

accident involving radiological releases, 

control room personnel must implement 

extensive emergency operation 

procedures to ensure public health and 

safety. Wearing respiratory protection 

during normal operations or even 

during an orderly shutdown, should it 

be necessary as a result of a toxic gas 

release, would not be expected to 

present significant challenges to control 

room personnel equivalent to those 

present during a reactor accident. The 

NRC is reluctant to place any more of 

a burden than is absolutely necessary on 

control room personnel, who would 

already be significantly tasked ensuring 

that all emergency procedures are 

carried out without error. 

B. Procedures Are Not Optimized 

The petitioner stated that control 

room dose mitigation procedures must 

be consistent with the licensing basis 

and may not be the optimum mitigation 

strategy for more likely conditions. For 

example, he stated that control room 

dose models do not model dispersion as 

a period during the day with higher 

concentrations while the plume is 

blowing towards the control room and 

then a period of zero concentration for 

the rest of the day. Instead, analysis 

methods simplify this effect by 

assuming that a lower concentration is 

present continuously. The petitioner 

claimed that if procedures were revised 

to include a control room purge mode 

Federal Register / Vol. 74, No. 15 / Monday, January 26, 2009 / Proposed Rules 4349 

strategy, a ‘‘calculated increase in 

consequences in the simplistic design 

basis analysis’’ would result. 

The NRC disagrees with the 

petitioner. The NRC’s regulations do not 

require that procedures be limited to the 

most limiting licensing-basis 

assumptions. Further, the NRC expects 

licensees to develop procedures that 

address the full-scope review of designbasis events and conditions. 

With respect to the petitioner’s 

example, procedures to operate the 

control room in its design-basis mode 

must be provided. These procedures do 

not preclude licensees from creating 

additional procedures to purge the 

control room if warranted by plant 

conditions. Licensees are permitted to 

develop and implement such 

procedures under existing NRC 

regulations. 

The NRC agrees that control room 

purging may be a reasonable action 

during a reactor accident when the level 

of outside airborne concentration of 

radioactive material is less than the 

level inside the control room. However, 

the conditions favorable for control 

room purging cannot be predicted, and 

the NRC cannot credit control room 

purging in the DBA analysis unless the 

timing of the release can be accurately 

established. For accidents where NRC 

regulatory guidance has established the 

release duration, the NRC has accepted 

credit for control room purging after the 

release has ended. As a design criterion, 

GDC 19 does not supplant the radiation 

protection standards of 10 CFR Part 20, 

which treat the radiation exposure of 

control room operators as occupational 

exposure. Therefore, the NRC expects 

licensees to maintain the accumulated 

dose of their radiation workers ALARA. 

During an accident, health physics 

personnel would monitor the 

radiological conditions in the control 

room and other emergency response 

facilities. These health physicists are 

responsible for making appropriate 

recommendations to plant personnel on 

actions that can be taken to maintain the 

dose to emergency responders ALARA. 

C. Challenges to Safety Systems 

The petitioner stated that the current 

design requirements, which are usually 

imposed to ensure the assumptions of 

the dose analysis are met, may not be 

optimum from an overall risk 

perspective. As an example, he stated 

that a common design requirement 

specifies that the normal control room 

ventilation must isolate on receipt of a 

safety injection or containment isolation 

signal during an assumed loss-ofcoolant accident. The petitioner stated 

that it is more logical to delay control 

room isolation until radioactivity is 

detected in the control room or it is 

known that a radioactive plume is 

blowing towards the control room. The 

petitioner suggested that mitigating 

design strategies should be based on 

overall risk reduction designed for more 

likely conditions, not on one unlikely 

set of fixed hypothetical conditions. 

The NRC does not agree with the 

petitioner. Contrary to the petitioner’s 

assertion, the NRC’s regulations do not 

require immediate control room 

isolation or immediate appearance at 

the control room intake of the 

radioactive plume assumed in designbasis dose consequence analyses. The 

NRC has approved, in accordance with 

its regulations, plant designs that do not 

immediately isolate the control room 

ventilation system. Further, design 

bases that include the immediate startup 

of control room ventilation systems and 

loading of electrical buses and diesel 

generators with this equipment do not 

require operation of plant systems 

beyond their design capabilities; the 

diesels are specifically designed and 

sized to accommodate these safety 

loads. Therefore, the performance of 

these systems should not be impacted, 

and there is no increased risk to public 

health and safety. 

D. Inappropriate Technical 

Specification Action Statements 

The petitioner stated that the 

conservative nature of the current 

radiological dose mitigation analyses 

also results in inappropriate TS action 

statements. He stated that ‘‘there is 

insignificant safety significance to the 

TS associated with control room 

habitability and yet there are shutdown 

requirements.’’ The petitioner believes 

that in order to evaluate the net public 

safety risk associated with these TS 

shutdown requirements, small but 

quantifiable public risks associated with 

the shutdown of a nuclear power plant 

must be considered, including but not 

limited to the following: 

1. Risk associated with bringing the 

plant through a transient and another 

thermal cycle; 

2. Airborne pollutants released by the 

fossil units required to operate to make 

up for lost power; and 

3. Potential for challenging electric 

power grid stability with the public risk 

associated with the possibility of rolling 

blackouts or brownouts or, under the 

worst conditions of grid instability, the 

potential for a loss of offsite power at 

multiple nuclear power facilities. 

The petitioner claimed that the 

shutdown requirement increases the net 

public risk and should be eliminated 

because it is only imposed as a ‘‘matter 

of compliance.’’ 

The NRC disagrees with the 

petitioner. The NRC has approved 

license amendments to replace TS 

requirements for an immediate 

shutdown for an inoperable control 

room envelope boundary with 

requirements for immediate mitigating 

actions and restoration of the control 

room envelope to operable status within 

90 days. 

The NRC has determined that none of 

the regulations proposed to be changed 

by the petitioner directly require a plant 

shutdown in response to control room 

habitability issues. Existing NRC 

regulations permit a licensee to propose 

alternative TS action requirements to its 

plant shutdown requirements. The NRC 

notes that even if the petitioner’s 

proposed regulatory changes were 

made, licensees would still need to 

submit a license amendment to justify 

changes to their TSs for NRC approval. 

A controlled shutdown and cooldown 

of a plant is a safe evolution within the 

design capability of the plant and would 

not result in undue risk to public safety. 

In the event of unusual circumstances 

associated with adverse electrical power 

grid instability or other complicating 

issues that would be associated with a 

plant shutdown, there are processes 

available for a licensee to obtain 

regulatory relief to safely continue plant 

operation (e.g., emergency/exigent 

technical specification change, 

enforcement discretion). 

E. Unjustified Technical Specification 

Surveillances 

The petitioner stated that ‘‘individual 

input assumptions for radiological dose 

analyses have no significance in 

predicting reality or the acceptability of 

results. Even if actual conditions were 

such that one of the assumptions was 

non-conservative by a couple orders of 

magnitude, the ultimate result (in this 

case habitability of the control room) 

would still be acceptable due to the 

significant conservatisms in the other 

assumptions and the simplicity of 

effective mitigating actions such as the 

use of KI.’’ He stated that although most 

control room habitability surveillances 

can be performed with minimal 

resources, licensees have been required 

to demonstrate the accuracy of the 

assumption regarding unfiltered 

inleakage using an unjustified tracer gas 

testing method that costs approximately 

$100,000 per test. The petitioner stated 

these tests have demonstrated that 

although inleakage values assumed in 

the analyses were nonconservative, 

there was no safety significance and 

continued operation was justified. The 

4350 Federal Register / Vol. 74, No. 15 / Monday, January 26, 2009 / Proposed Rules 

petitioner concluded that the 

expenditure for tracer gas testing could 

be better used for improvements that 

would likely be more beneficial to plant 

safety; therefore, the required 

performance of this test could have a net 

negative safety consequence. The 

petitioner stated that previous 

surveillances, such as a pressurization 

test, combined with lessons learned 

from tracer gas testing result in an 

effective preventative maintenance 

program. 

The NRC does not agree with the 

petitioner’s assertion that individual 

input assumptions for radiological dose 

analyses have no significance in 

predicting reality or the acceptability of 

results. The NRC places a high priority 

on operator safety; the requirements 

contained in GDC 19 should be retained 

because they provide physical and 

psychological protection for operators 

and ultimately for the general public. 

Therefore, the data used in the analyses 

to determine operator safety should be 

accurate, and when data are uncertain, 

appropriate conservatisms are applied. 

The NRC does not agree with the 

petitioner’s statement that the 

expenditure for tracer gas testing could 

be better used for improvements that 

would likely be more beneficial to plant 

safety nor does the NRC agree that the 

performance of tracer gas testing could 

have a net negative safety consequence. 

The potential dose to the operator must 

be quantified in order to ensure that the 

requirements of GDC 19 are met; the 

specific measurement of inleakage is 

one of the inputs to the analyses used 

to quantify the potential dose to the 

operator. Prior to the use of tracer gas 

to measure inleakage, the quantity of 

inleakage was assumed rather than 

measured and subsequently found to be 

nonconservative. Tracer gas testing is 

justified because it ensures operator 

safety. Other methods of measuring 

inleakage have not been successfully 

demonstrated. 

F. Petitioner’s Proposed Alternatives to 

Current NRC Guidance 

The NRC has decided to deny this 

petition for rulemaking and would 

normally not discuss the petitioner’s 

proposed guidance in this document. 

However, in order to clarify the NRC’s 

decision to maintain the current 

radiological dose requirements, the 

following discussion is provided. 

Under Commission policy, the NRC’s 

regulations for control room habitability 

provide performance-based 

requirements to ensure that plant 

personnel are adequately protected. The 

NRC has concluded that prescriptive 

requirements or guidance, such as that 

proposed by the petitioner, may 

unnecessarily restrict a licensee’s 

options for complying with the NRC’s 

regulations. 

The petitioner proposed revisions to 

the NRC’s regulatory guidance to help 

implement his proposed rule change. 

NRC regulatory guidance is not an 

appropriate subject for a PRM and the 

NRC will not generally consider such 

requests through this process. Further, 

current NRC regulatory guidance 

provides one acceptable mechanism for 

licensees and applicants to meet the 

requirements of the NRC’s regulations. 

Applicants and licensees may propose 

alternative means of complying with the 

NRC’s regulations, which will be 

evaluated by the NRC staff on a case-bycase basis. 

1. The petitioner recommended that 

the control room ventilation system 

should isolate on the detection of high 

radiation or toxic intake. The NRC 

disagrees with the petitioner. All control 

rooms are required by TSs to take 

appropriate action upon detection of 

radiation or toxic gas. Appropriate 

action may differ from plant to plant 

depending on location, design, and TSs. 

Because plants are unique, licensees can 

demonstrate compliance with the 

control room design criteria by taking 

different approaches. The petitioner’s 

suggestion does not address the longterm release situations that would be 

expected under a worst case accident 

scenario. Control room isolation alone 

would not be an acceptable solution 

because it does not adequately consider 

the long term breathing air requirements 

necessary to provide a safe working 

environment in the control room. After 

a relatively short period of time, an 

intake of air into the control room 

would be necessary. Licensees include 

these considerations in their sitespecific control room habitability 

analyses. Therefore, the NRC concludes 

that changing guidance to recommend 

control room isolation on detection of 

high radiation or toxic gas is an 

unnecessarily prescriptive 

recommendation in comparison to the 

existing performance-based dose 

criterion. 

2. The petitioner recommended that 

the control room have a minimum of 

one foot of concrete shielding (or 

equivalent) on all surfaces. The NRC 

disagrees with the petitioner. The NRC 

believes that control rooms are 

adequately protected from the effects of 

direct radiation because current 

regulations require that either a 5 rem 

whole body or a 5 rem TEDE acceptance 

criterion be met under DBA conditions. 

Licensees include the effects of direct 

radiation from all potential sources in 

their control room dose consequence 

analyses. Typically these sources 

include the following: 

• Contamination of the control room 

atmosphere by the intake and 

infiltration of the radioactive material 

contained in the radioactive plume 

released from the facility; 

• Direct shine from the external 

radioactive plume released from the 

facility with credit for control room 

structural shielding; 

• Direct shine from radioactive 

material in the containment with credit 

for both the containment and control 

room structural shielding; and 

• Radiation shine from radioactive 

material in systems and components 

inside or external to the control room 

envelope, including radioactive material 

buildup on the control room ventilation 

filters. 

Many control rooms already have one 

foot or more of concrete shielding on all 

surfaces. One foot of concrete shielding 

does not guarantee adequate protection 

from radiation. For example, surfaces 

with 1 foot of concrete with 

penetrations for various equipment, 

such as electrical wiring and ventilation 

ducts, may not provide any more 

protection than non-concrete surfaces or 

surfaces with less than 1 foot of 

concrete. To show compliance with the 

current control room dose criterion, 

licensees provide detailed radiological 

calculations to ensure that under DBA 

conditions control room personnel will 

be adequately protected. Licensees have 

demonstrated compliance with the 

regulations crediting many different 

design approaches. The NRC concludes 

that recommending that the control 

rooms have one foot of concrete 

shielding is an unnecessarily 

prescriptive recommendation. 

3. The petitioner recommended that 

because of the low risk significance of 

being outside the control room 

habitability program guidelines, a plant 

shutdown should not be required in this 

condition. Rather, the petitioner 

recommended that the program could 

specify that timely actions should be 

taken to return the plant to within the 

guidelines. If not complete within 30 

days, the petitioner suggested that a 

special report would be sent to the NRC 

with a justification for continued 

operations and a proposed schedule for 

meeting the guidelines. The NRC 

disagrees with the petitioner that a 

regulatory change is required to permit 

these changes to plant TSs. The NRC 

allows deviations from the integrity of 

the control room envelope without 

requiring an immediate plant shutdown. 

4. The petitioner recommended that 

as an alternative to the total removal of 

Federal Register / Vol. 74, No. 15 / Monday, January 26, 2009 / Proposed Rules 4351 

1As defined in 10 CFR 20.1003, ‘‘Total Effective 

Dose Equivalent (TEDE) means the sum of the 

effective dose equivalent (for external exposures) 

and the committed effective dose equivalent (for 

internal exposures).’’ The effective dose equivalent 

for external exposures includes the whole body 

dose from noble gases. The committed effective 

dose equivalent for internal exposure includes the 

thyroid dose from inhalation of iodine. 

dose guidelines from the regulations, 

most of his concerns could be resolved 

if the dose criteria were based solely on 

the whole body dose from noble gases. 

The NRC does not agree with the 

proposition that the dose criteria should 

be based solely on the whole body dose 

from noble gases. The control room dose 

criterion of 5 rem whole body or its 

equivalent to any organ imposes two 

requirements on licensees: Satisfaction 

of the whole body dose criterion, which 

is generally dominated by the dose from 

noble gases; and satisfaction of the 

organ-specific dose guidelines, which 

are generally dominated by the thyroid 

dose from the inhalation of iodine. In 

most cases, demonstrating compliance 

with thyroid dose guidelines poses a 

significantly greater challenge to 

licensees than does compliance with the 

whole body dose criterion. 

The 1999 amendment to 10 CFR 50.67 

(64 FR 12117), revised the control room 

dose limit to allow licensees to show 

compliance with either the existing 

limits, using the traditional Technical 

Information Document (TID)–14844 

source term assumptions, or a revised 

single control room dose criterion of 5 

rem TEDE,1 if the licensee adopts the 

AST. With the ability to reassess a 

maximum credible radiological release 

using the AST, many licensees have 

shown compliance with the § 50.67 

single control room dose criterion of 5 

rem TEDE. Licensees have 

accomplished this while achieving an 

enhanced degree of operational 

flexibility not realized using the 

traditional TID–14844 source term with 

the associated whole body dose 

criterion and organ dose guidelines. 

Because compliance with § 50.67 is 

demonstrated by calculating the TEDE, 

the relative contribution of the thyroid 

dose to the demonstration of 

compliance with the control room 

criterion has been substantially and 

appropriately reduced. In addition, 

many licensees that continue to use the 

traditional TID–14844 source term have 

incorporated the guidance in Regulatory 

Guide (RG) 1.195, ‘‘Methods and 

Assumptions for Evaluating 

Radiological Consequences for DesignBasis Accidents at Light-Water Nuclear 

Power Reactors’’ (ML031490640) to 

achieve operational flexibility. 

Following the guidance in RG 1.195, 

licensees are able to evaluate control 

room habitability using a 50 rem thyroid 

dose guideline. This represents a 

significant relaxation from the 30 rem 

thyroid dose guideline that was 

incorporated into previous guidance 

documents. 

The petitioner also stated that the 

whole body dose from noble gases is 

likely to be the only possible dose 

impact that may result in control room 

evacuation. The NRC does not accept 

the premise that any maximum credible 

radiological release would result in the 

necessity for a control room evacuation. 

As stated previously, the 5 rem control 

room design criterion is not intended to 

be a maximum integrated dose level at 

which control room evacuation would 

be mandated during an accident. Rather, 

the criterion is used as a design basis to 

ensure that the control room, by design, 

will provide a habitable environment for 

the control of the plant under the 

maximum credible radiological release 

conditions, and as such will provide 

reasonable assurance of adequate 

protection. 

The petitioner stated that most of his 

concerns would be resolved if credit for 

SCBAs or KI was allowed in the analysis 

of the dose from iodines and 

particulates. The NRC does not agree 

with the option of replacing engineering 

controls for radiological protection with 

credit for personal protective 

equipment. As discussed previously, the 

option of allowing credit for SCBAs or 

KI to show compliance with the control 

room performance-based design 

criterion is inimical to the NRC design 

philosophy incorporated into 10 CFR 

Part 20, as well as international 

standards for radiological protection as 

set forth in ICRP Publication 26. 

IV. Public Comments 

1. Overview of Public Comments 

The NRC’s notice of receipt and 

request for public comment invited 

interested persons to submit comments. 

The comment period for PRM–50–87 

closed on September 25, 2007. The NRC 

reviewed and considered the comments 

in its decision to deny the petition. The 

NRC received two public comments, 

one from Mr. Walston Chubb 

(ML072681072), and one from Mr. 

James H. Riley on behalf of the Nuclear 

Energy Institute (NEI) (ML072690232). 

2. Mr. Walston Chubb Comment 

Comment: Mr. Chubb recommended 

that operators be required to remain on 

duty until they are relieved or their 

short-time doses are between 100 and 

200 rem. 

NRC Response: The primary objective 

of GDC 19 is to ensure that the design 

of the control room and its habitability 

systems provide a ‘‘shirt-sleeved’’ 

environment for operators during both 

normal and accident conditions. This 

environment facilitates operator 

response to normal and accident 

conditions while minimizing errors of 

omission or commission. Another 

objective is to ensure that the radiation 

dose levels in the control room would 

make it the safest location on site, 

thereby allowing the operators to remain 

in the control room. Any reduction in 

operator accident response capabilities 

may negatively impact public health 

and safety. 

The NRC’s decision to apply the 5 

rem whole body dose criterion was 

based on the following: 

• A whole body radiation exposure of 

5 rem is considered unlikely to cause 

increased anxiety that would result in 

operator impairment, since the criterion 

is comparable to the occupational dose 

limits. 

• A whole body radiation exposure of 

5 rem would not result in any somatic 

response that could result in operator 

impairment. Generally, the onset of 

clinically observable somatic effects 

occurs between 25 and 50 rem. 

• GDC 19, as a design criterion, does 

not supplant the radiation protection 

standards of 10 CFR Part 20. The 

radiation exposure of control room 

operators is controlled, as for any 

radiation worker at the facility, as 

occupational exposure under 10 CFR 

Part 20. In the statements of 

consideration for the 10 CFR Part 20 

rulemaking (56 FR 23365; May 21, 

1991), the NRC stated that the dose 

limits for normal operation should 

remain the primary guidelines for an 

emergency. 

The statement of considerations in the 

proposed and final rule amending 10 

CFR 50.67 and GDC 19 (64 FR 12117, 

March 31, 1999; and 64 FR 71990, 

December 23, 1999, respectively) 

included the NRC’s basis for 

establishing the 5 rem TEDE as the GDC 

19 numeric criterion for licensees 

applying for amendment under 10 CFR 

50.67. It also reaffirmed the position 

that the criteria in GDC 19 and the final 

rule are based on occupational exposure 

limits. 

The 5 rem control room design 

criterion is not intended to be a 

maximum integrated dose above which 

control room evacuation would be 

mandated during an accident. Rather, 

the 5 rem design criterion ensures that 

the control room, by design, will 

provide a habitable environment for the 

4352 Federal Register / Vol. 74, No. 15 / Monday, January 26, 2009 / Proposed Rules 

control of the plant under all DBA 

conditions. 

Providing a safe working environment 

for the highly skilled professionals 

needed to operate a nuclear power plant 

is a primary objective of NRC 

regulations related to occupational and 

accident dose, and it is a paramount 

goal throughout the entire nuclear 

power industry. The NRC concludes 

that the proposal to set the control room 

design criterion at 100 rem, which is 

well above the level at which the onset 

of clinically observable somatic effects 

would occur, is antithetical to the 

fundamental principle of protecting 

public health and safety and is not 

acceptable. 

3. NEI Comments 

NEI provided the following 

comments: 

Comment: ‘‘It is not so much the 

value of the exposure limits that is the 

problem. The NRC should be more open 

to other methods of analysis proposed 

by licensees. Every Regulatory Guide 

states that the guidance is one method 

acceptable to the staff and that other 

methods proposed by licensees will be 

evaluated on a case-by-case basis. 

However, in practice it is often difficult 

to justify different approaches.’’ 

NRC Response: To the extent that the 

comment implicitly criticizes the NRC 

for allegedly failing to consider 

alternatives for compliance with GDC 19 

and 10 CFR 50.67 in a manner other 

than that suggested in a regulatory 

guide, that concern is beyond the scope 

of this petition for rulemaking. Further, 

the commenter presented no basis for 

this implicit criticism—the NRC 

routinely considers licensee and 

applicant-proposed alternatives to 

methods set forth in a Regulatory Guide. 

However, the NRC expects licensees and 

applicants to provide technically 

sufficient basis for the use of an 

alternative for compliance with an NRC 

regulation, which is also consistent with 

the regulatory policies of the NRC. That 

a licensee or applicant may find it 

difficult to provide sufficient basis 

justifying the use of an alternative 

approach, however, would not appear to 

present a valid regulatory concern. 

Comment: Existing emergency 

filtration systems should be maintained 

to practical performance criteria. NEI 

stated that this area has a lot of potential 

for improvement and gave the following 

examples: 

• The current practice (i.e., RG 1.52, 

‘‘Design, Inspection, and Testing 

Criteria for Air Filtration and 

Adsorption Units of Post-Accident 

Engineered-Safety-Feature Atmosphere 

Cleanup Systems in Light-Water-Cooled 

Nuclear Power Plants’’) (ML011710176) 

is to apply a safety factor of 2 for 

laboratory testing of charcoal beds. The 

actual efficiencies are typically much 

higher than those allowed by RGs. 

• Some plants have an 8-inch 

charcoal bed, for which only 4 inches is 

allowed to be credited. 

• Other plants have filtration systems 

in series, for which only one composite 

filter can be credited. 

NRC Response: The NRC’s position on 

existing emergency filtration systems is 

outlined in RG 1.52, Revision 3, issued 

June 2001. The previous revision of the 

RG included a safety factor as great as 

7 whereas Revision 3 includes a safety 

factor of 2 to account for degradation of 

the system between test periods. A 

safety factor represents margin in the 

capability of the adsorbent (carbon) 

installed in the system to perform the 

required safety function. Because carbon 

can degrade between test periods, a 

safety factor provides confidence that 

the anticipated degradation will not be 

beyond the minimum level necessary to 

perform its required safety function. 

RG 1.52, Revision 3, indicates that a 

4-inch carbon bed in U.S. nuclear power 

plants is 99 percent efficient, with a 

safety factor of 2 and a penetration (as 

defined in American Society for Testing 

and Materials D 3803–89) of less than or 

equal to 0.5 percent. The NRC believes 

that a 4-inch carbon bed thickness is 

sufficient to provide adequate 

protection, and that the 4 inches, as 

reflected in the RG, is not intended to 

be an upper limit on bed thickness. It is 

acceptable to provide additional carbon 

that may include 6 inches, 8 inches, or 

even greater bed thickness. The NRC 

also believes there are benefits provided 

by carbon bed thicknesses greater than 

4 inches that are not reflected in the RG. 

The benefits may include longer bed life 

contributing to lower overall cost. 

With respect to filtration systems in 

series, they are treated as a composite 

(i.e., the sum of individual filters in 

series). For example, the efficiency of 

two 2-inch beds in series is the same as 

one 4-inch bed. 

Comment: In response to the 

petitioner’s statement that current TS 

for system performance should be 

eliminated and that the administrative 

portion of the TS could include a 

requirement to have a control room 

habitability program, NEI commented, 

‘‘This recommendation is covered by 

TSTF–448 and GL 2003–01.’’ 

Response: NRC agrees with the 

comment. NRC prepared and made 

available a model safety evaluation (SE) 

and a model no-significant-hazardsconsideration (NSHC) determination 

relating to the modification of technical 

specification (TS) requirements 

regarding the habitability of the control 

room envelope (CRE) for referencing in 

license amendment requests (LARs). 

NRC also made available an associated 

model LAR for use by licensees to 

prepare such LARs. The TS 

modification is based on NRC staff 

approved changes to the improved 

standard technical specifications (STS) 

(NUREGs 1430–1434, available on 

NRC’s public Web site at www.nrc.gov/ 

reactors/operating/licensing/techspec/ 

current-approved-sts.html) that were 

proposed by the pressurized and boiling 

water reactor owners groups’ Technical 

Specifications Task Force (TSTF) on 

behalf of the commercial nuclear 

electrical power generation industry, in 

STS change traveler TSTF–448, 

Revision 3 (ML063460558). NRC 

published a Notice of Availability of the 

SER in the Federal Register on January 

17, 2007 (72 FR 2022). Generic Letter 

(GL) 2003–01, dated June 12, 2003, is 

available on ADAMS (ML031620248). 

Comment: In response to the 

petitioner’s proposed guidance, NEI 

provided the following comments: 

• The control room ventilation 

system should isolate on the detection 

of high radiation or toxic gas intake. NEI 

commented, ‘‘A good many control 

rooms in the industry already operate in 

this manner. Conversely, there are some 

plants that do not have automatic 

initiation of the emergency mode. 

Making this a requirement could result 

in an undue (and expensive) 

modification/backfit. For those plants 

susceptible to toxic gas intrusion, 

automatic initiation is typically the case 

(although not specifically implemented 

in all cases). If required, this also could 

result in undue (and expensive) 

modifications.’’ 

• The control room should have a 

minimum of one foot of concrete 

shielding (or equivalent) on all surfaces. 

NEI commented, ‘‘It is unlikely that all 

control rooms have one foot of concrete 

shielding on all surfaces. This 

requirement could result in undue (and 

expensive) modifications. A similar 

concern applies to the technical support 

center, which may also be affected by 

this requirement.’’ 

• SCBAs and KI tablets should be 

readily available for operator use. 

Operators should maintain training in 

SCBAs. NEI commented, ‘‘The use of 

these methods has merit, but additional 

evaluation of their effects is necessary. 

The medical complications of ingesting 

KI would have to be evaluated for all CR 

personnel. The use of SCBA credit 

would require specific training for 

which operators will need to 

demonstrate the ability to conduct their 

Federal Register / Vol. 74, No. 15 / Monday, January 26, 2009 / Proposed Rules 4353 

safety-related functions while wearing a 

SCBA for several hours.’’ 

• Procedures should be developed to 

ensure control room purging is 

considered when the outside 

concentration is less than the inside 

concentration. NEI commented, 

‘‘Although this appears to be a good 

practice, it can’t be credited in the 

operator dose analysis. The timing of 

purging could be critical based on the 

timing of the release and the release 

pathway. Therefore, this 

recommendation may not have any 

practical merit.’’ 

The petitioner stated that because of 

the low risk significance of being 

outside the control room habitability 

program guidelines, a plant shutdown 

would not be required in this condition; 

rather, the program could specify that 

timely actions should be taken to return 

the plant within the guidelines. If not 

complete within 30 days, a special 

report would be sent to the NRC with 

a justification for continued operation 

and a proposed schedule for meeting the 

guidelines. NEI commented, ‘‘This is a 

valid point that the industry supports.’’ 

The petitioner stated that as an 

alternative to total removal of dose 

guidelines from the regulations, most of 

his concerns could be resolved if the 

dose criteria were based solely on the 

whole body dose from noble gases that 

he believes is the only possible dose 

impact that may result in control room 

evacuation. NEI commented, ‘‘It is not 

clear that the noble gas contribution 

would be limiting in all cases. However, 

this may be the case if KI were allowed 

to be credited.’’ 

Response: These comments have been 

addressed in Section III of this 

document. 

V. Denial of Petition 

Based upon review of the petition and 

comments received, the NRC has 

determined that the conclusions upon 

which the petitioner relies do not 

substantiate a basis to eliminate the 

control room radiological dose 

acceptance criteria from current 

regulations as requested. For the reasons 

discussed previously, the Commission 

denies PRM–50–87. 

Dated at Rockville, Maryland, this 14th day 

of January 2009. 

For the Nuclear Regulatory Commission. 

Annette L. Vietti-Cook, 

Secretary of the Commission. 

[FR Doc. E9–1211 Filed 1–23–09; 8:45 am] 

BILLING CODE 7590–01–P 

DEPARTMENT OF TRANSPORTATION 

Federal Aviation Administration 

14 CFR Part 25 

[Docket No. NM398; Notice No. 25–09–01– 

SC] 

Special Conditions: Model C–27J 

Airplane; Interaction of Systems and 

Structures 

AGENCY: Federal Aviation 

Administration (FAA), DOT. 

ACTION: Notice of proposed special 

conditions. 

SUMMARY: This action proposes special 

conditions for the Alenia Model C–27J 

airplane. This airplane has novel or 

unusual design features when compared 

to the state of technology described in 

the airworthiness standards for 

transport-category airplanes. These 

design features include electronic flightcontrol systems. These special 

conditions pertain to the effects of novel 

or unusual design features such as 

effects on the structural performance of 

the airplane. We have issued additional 

special conditions for other novel or 

unusual design features of the C–27J. 

The applicable airworthiness 

regulations do not contain adequate or 

appropriate safety standards for this 

design feature. These proposed special 

conditions contain the additional safety 

standards that the Administrator 

considers necessary to establish a level 

of safety equivalent to that established 

by the existing airworthiness standards. 

DATES: We must receive your comments 

by February 25, 2009. 

ADDRESSES: You must mail two copies 

of your comments to: Federal Aviation 

Administration, Transport Airplane 

Directorate, Attn: Rules Docket (ANM– 

113), Docket No. NM398, 1601 Lind 

Avenue SW., Renton, Washington 

98057–3356. You may deliver two 

copies to the Transport Airplane 

Directorate at the above address. You 

must mark your comments: Docket No. 

NM398. You can inspect comments in 

the Rules Docket weekdays, except 

Federal holidays, between 7:30 a.m. and 

4 p.m. 

FOR FURTHER INFORMATION CONTACT: 

Holly Thorson, FAA, International 

Branch, ANM–116, Transport Airplane 

Directorate, Aircraft Certification 

Service, 1601 Lind Avenue SW., 

Renton, Washington 98057–3356; 

telephone (425) 227–1357, facsimile 

(425) 227–1149. 

SUPPLEMENTARY INFORMATION: 

Comments Invited 

We invite interested people to take 

part in this rulemaking by sending 

written comments, data, or views. The 

most helpful comments reference a 

specific portion of the special 

conditions, explain the reason for any 

recommended change, and include 

supporting data. We ask that you send 

us two copies of written comments. 

We will file in the docket all 

comments we receive, as well as a 

report summarizing each substantive 

public contact with FAA personnel 

concerning these special conditions. 

You can inspect the docket before and 

after the comment closing date. If you 

wish to review the docket in person, go 

to the address in the ADDRESSES section 

of this preamble between 7:30 a.m. and 

4 p.m., Monday through Friday, except 

Federal holidays. 

We will consider all comments we 

receive on or before the closing date for 

comments. We will consider comments 

filed late if it is possible to do so 

without incurring expense or delay. We 

may change these special conditions 

based on the comments we receive. 

If you want the FAA to acknowledge 

receipt of your comments on this 

proposal, include with your comments 

a self-addressed, stamped postcard on 

which the docket number appears. We 

will stamp the date on the postcard and 

mail it back to you. 

Background 

On March 27, 2006, the European 

Aviation Safety Agency (EASA) 

forwarded to the FAA an application 

from Alenia Aeronautica of Torino, 

Italy, for U.S. type certification of a 

twin-engine commercial transport 

designated as the Model C–27J. The 

C–27J is a twin-turbopropeller, cargotransport aircraft with a maximum 

takeoff weight of 30,500 kilograms. 

Type Certification Basis 

Under the provisions of Section 21.17 

of Title 14 Code of Federal Regulations 

(14 CFR) and the bilateral agreement 

between the U.S. and Italy, Alenia 

Aeronautica must show that the C–27J 

meets the applicable provisions of 14 

CFR part 25, as amended by 

Amendments 25–1 through 25–87. 

Alenia also elects to comply with 

Amendment 25–122, effective 

September 5, 2007, for 14 CFR 25.1317. 

If the Administrator finds that 

existing airworthiness regulations do 

not adequately or appropriately address 

safety standards for the C–27J due to a 

novel or unusual design feature, we 

prescribe special conditions under 

provisions of 14 CFR 21.16.