OMB Control No.: 3150-0011

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, DC 20555-0001

June 12, 2003

NRC GENERIC LETTER 2003-01: CONTROL ROOM HABITABILITY

Addressees

All holders of operating licenses for pressurized-water reactors (PWRs) and boiling-water

reactors (BWRs), except those who have permanently ceased operations and have certified

that fuel has been permanently removed from the reactor vessel and more than 1 year has

elapsed since fuel was irradiated in the reactor vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to:

(1) alert addressees to findings at U.S. power reactor facilities suggesting that the control

room licensing and design bases, and applicable regulatory requirements (see section

below) may not be met, and that existing technical specification surveillance

requirements (SRs) may not be adequate,

(2) emphasize the importance of reliable, comprehensive surveillance testing to verify

control room habitability,

(3) request addressees to submit information that demonstrates that the control room at

each of their respective facilities complies with the current licensing and design bases,

and applicable regulatory requirements, and that suitable design, maintenance and

testing control measures are in place for maintaining this compliance, and

(4) collect the requested information to determine if additional regulatory action is required. 

Background

The control room is the plant area, defined in the facility licensing basis, from which actions are

taken to operate the plant safely under normal conditions and to maintain the reactor in a safe

condition during accident situations. For most facilities, the habitability criteria of General

Design Criterion 19 (GDC 19) in 10 CFR Part 50, Appendix A, “General Design Criteria for

Nuclear Power Plants,” apply to this area. The control room envelope (CRE) is the plant area,

defined in the facility licensing basis, that encompasses the control room and may encompass

other plant areas. The structures that make up the CRE are designed to limit the inleakage of

radioactive and hazardous materials from areas external to the CRE. Control room habitability

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systems (CRHSs) typically provide the functions of shielding, isolation, pressurization, heating,

ventilation, air conditioning and filtration, monitoring, and the sustenance and sanitation

necessary to ensure that the control room operators can remain in the control room and take

actions to operate the plant under normal and accident conditions. The personnel protection

features incorporated into the design of a particular plant’s CRHSs depend on the nature and

scope of the plant-specific challenges to maintaining control room habitability. In the majority of

the CRHS designs, isolation of the normal supply and exhaust flow paths and pressurization of

the CRE relative to adjacent areas are fundamental to ensuring a habitable control room.

During the design of a nuclear power plant, licensees perform analyses to demonstrate that the

CRHSs, as designed, provide a habitable environment during postulated design basis events. 

These design analyses model the transport of potential contaminants into the CRE and their

removal. The amount of inleakage of assumed contaminants is important to these analyses. 

Unaccounted-for contaminants entering the CRE may impact the ability of the operators to

perform plant control functions. If contaminants impair the response of the operators to an

accident, there could be increased consequences to the public health and safety. 

There are two typical CRE designs. These designs are referred to as positive-pressure and

neutral-pressure CREs. Both designs focus on limiting the amount of contaminants entering

the CRE. For radiological challenges, the positive-pressure CRE intentionally pressurizes the

CRE with air from outside the CRE. The pressurization air is treated by a high-efficiency

particulate air filter and iodine adsorption media to remove contaminants. The neutral-pressure

CRE does not intentionally pressurize the CRE, but limits inleakage of contaminants by isolating

controlled flow paths into the CRE. Most plants with a positive-pressure CRE have a technical

specification SR to verify that those ventilation systems serving the CRE can maintain the CRE

at a positive differential pressure relative to adjacent areas. These surveillance tests (typically

referred to as a ∆P surveillance) are generally implemented through a technical specification

SR for the CRHSs. Plants with a neutral-pressure CRE design typically do not have a CRE

integrity testing program. (The term “neutral-pressure” means only that the CRE is not

intentionally pressured. The actual pressure of the CRE may be positive, neutral, or negative

relative to adjacent areas.)

In addition to the ∆P surveillance described above, licensees have performed CRE integrity

testing at approximately 30 percent of the power reactor facilities using the standard test

method described in American Society for Testing and Materials (ASTM) consensus standard

E741, “Standard Test Method for Determining Air Change in a Single Zone by Means of a

Tracer Gas Dilution.” Unlike the ∆P surveillance, the ASTM E741 test determines the total CRE

inleakage from all sources. It is well suited for assessing the integrity of positive-pressure or

neutral-pressure CREs. The test basically involves homogeneously dispersing a nontoxic

tracer gas throughout the CRE and measuring the dilution of the tracer gas caused by

inleakage.

 

The results of the ASTM E741 tests indicate that the ∆P surveillance is not a reliable method for

demonstrating CRE integrity. For all but one facility tested using the ASTM E741 standard, the

measured inleakage was greater than the inleakage assumed in the design basis analyses. In

some cases, even though the licensees had routinely demonstrated a positive ∆P relative to

adjacent areas at their facilities, the measured inleakage was several orders of magnitude

greater than the value previously assumed. Affected facilities were subsequently able to
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achieve compliance with the control room radiation protection regulatory requirements by

sealing, adding new ductwork, changing their CRE, or reanalyzing their control room

habitability.

Use of the ∆P surveillance as an indicator of CRE integrity has two inherent deficiencies. First,

it does not measure CRE inleakage. The ∆P surveillance infers that no contamination can

enter the CRE if the CRE is at a higher pressure than adjacent areas. Second, the ∆P

surveillance cannot determine whether there may be unrecognized sources of pressurization of

the CRE that could introduce contaminants into the CRE under accident conditions. Two

possible unrecognized contamination pathways are the CRHS fan suction ductwork that is

located outside the CRE, and the pressurized ducts that traverse the lower pressure CRE en

route to another plant area.

The ASTM E741 testing has helped to identify a spectrum of CRHS deficiencies that affect

(1) system design, construction, and quality, (2) system boundary construction and integrity,

and (3) technical specification SRs. Licensees have determined that the performance of the

CRHSs can be affected by (1) the gradual degradation in associated equipment such as seals,

floor drain traps, fans, ductwork, and other components, (2) the drift of throttled dampers, 

(3) maintenance on the CRHSs, and (4) inadvertent misalignments of the CRHSs. Since

inleakage is influenced by pressure differentials between the CRE and adjacent areas, changes

in ambient pressure in these adjacent areas can affect the CRE inleakage. These changes can

be the result of a modification, the degradation of the ventilation systems serving these areas,

or inadequate preventive and corrective maintenance programs.

Licensees and NRC staff have identified other deficiencies in CRHS design, operation, and

performance from the review of license amendments, licensee event reports, and records and

reports prepared pursuant to 10 CFR 50.59. These deficiencies showed that the licensees’

CRHSs did not meet their design bases. Some of these deficiencies are discussed in

Regulatory Issue Summary 2001-19, “Deficiencies in the Documentation of Design Basis

Radiological Analyses Submitted in Conjunction with License Amendment Requests.” 

For example, some licensees credited the operation of CRHSs based upon actuation of highradiation signals from instrumentation. Further investigation revealed that for some licensees

the system would not be actuated due to incorrect setpoints or placement of the

instrumentation. Other CRHS designs appear not to have considered unfiltered or once-filtered

inleakage through idle CRHS ventilation trains. Without adequate consideration of such design

issues, design basis radiation exposure limits may be exceeded.

Previous to the ASTM E741 testing, a group of licensees had trouble meeting the control room

criteria in Three Mile Island (TMI) Action Item III.D.3.4, “Control Room Habitability

Requirements,” that the NRC ordered most licensees to implement after the accident at TMI. 

At that time, radiological source term research suggested that the distribution of the chemical

forms of iodine released during an accident could be different from the distribution in the

traditional source term defined in U.S. Atomic Energy Commission Technical Information

Document (TID) 14844, “Calculation of Distance Factors for Power and Test Reactor Sites.” 

Because of the possible differences, the staff allowed licensees to postpone changing their

control rooms until the ongoing source term research was completed or until a generic letter on

control room habitability was issued. The staff believed that postponing changes was

reasonable since the source term research or improved methods of analyses might prove that
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the changes were unnecessary. Many of these licensees that postponed changes 

incorporated compensatory actions into their operating procedures to assure that the control

room operators would be protected in case of an accident. Since then, some licensees have

found that they could not meet the thyroid dose limits for habitability without using

compensatory actions. The NRC also allowed these facilities to use compensatory actions until

completion of the source term research. In August 2000, the NRC staff incorporated the results

of the source term research into Regulatory Guide 1.183, “Alternative Radiological Source

Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” which is now

available for use by licensees.

Although many CRE integrity testing programs focus on radiological concerns, radiation is only

one potential design basis challenge to the protection of the operators. The inleakage of other

contaminants may have a greater impact on control room habitability. An inleakage rate that is

tolerable for one contaminant may not be tolerable for another. The control room licensing

basis describes the hazardous chemical releases considered in the CRE design, the design

features, and the administrative controls implemented to mitigate the consequences of these

releases to the control room operators. Smoke and other byproducts of fire within the CRE or

in adjacent areas are among the contaminants that can have an adverse impact on control

room habitability.

Discussion

Information obtained by the NRC indicates that some licensees have not maintained adequate

configuration control over their CREs and have not corrected identified design and performance

deficiencies. The primary design function of CRHSs is to provide a safe environment in which

the operator can control the nuclear reactor and auxiliary systems during normal operations and

can safely shut down these systems during abnormal situations to protect the health and safety

of the public. It is important for the operators to be confident of their safety in the control room

to minimize errors of omission and commission. Errors of omission and commission are more

likely if CRHSs do not properly perform as intended in response to challenges from off-normal

or accident situations. The control room must be safe so that operators can remain in the

control room to monitor plant performance and take appropriate mitigative actions. This is an

underlying assumption in both the design basis and severe accident risk analyses. It is,

therefore, imperative to the health and safety of the public that operators are safe in the control

room at all times.

The scope and magnitude of the problems that NRC staff and certain licensees have identified

raise concerns about whether similar design, configuration, and operability problems exist at

other reactor facilities. The NRC staff is particularly concerned about whether licensees’

programs to maintain configuration control of CRHSs are sufficient to demonstrate that the

physical and functional characteristics of CRHSs are consistent with and are being maintained

according to their design bases. It is emphasized that the NRC’s position has been, and

continues to be, that it is the responsibility of individual licensees to know the licensing basis for

the CRHSs. Licensees should also have appropriate documentation of the design basis and

procedures in place, in accordance with NRC regulations, for performing necessary

assessments of plant or procedure changes that may affect the performance of the CRHSs. 
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The technical specifications for about 75 percent of the control rooms (mostly positive-pressure

CREs) have an SR to measure the ∆P from the CRE to adjacent areas. The bases of the

Improved Standard Technical Specifications state that this SR demonstrates control room

integrity with respect to unfiltered inleakage. The ASTM E741 integrated testing proves that it

does not. Because 10 CFR 50.36 requires technical specifications to be derived from the

safety analyses, the staff believes that the existing deficiency should be corrected. This

correction is consistent with NRC Administrative Letter 98-10, “Dispositioning of Technical

Specifications That Are Insufficient To Assure Plant Safety,” which describes the staff’s

expectation that licensees correct technical specifications that are found to “contain nonconservative values or specify incorrect actions.” 

Because of the importance of ensuring habitable control rooms under all normal and off-normal

plant conditions, the addressees are requested to provide certain information that will enable

the NRC staff to verify whether addressees can demonstrate and maintain the current design

bases for the CRHSs at their facilities. Addressees are encouraged, but not required, to work

closely with industry groups on the coordination of their responses. Coordinating the responses

promotes efficiency since it leads to a uniform approach to demonstrating compliance with the

design bases of their CREs.

NEI 99-03, “Control Room Habitability Assessment Guidance,” provides industry generic

guidance on control room habitability. The NRC staff reviewed NEI 99-03, but rather than fully

endorse NEI 99-03, the NRC staff developed its own guidance. Regulatory Guide 1.196 

(formerly DG-1114), “Control Room Habitability at Light-Water Nuclear Power Reactors,”

endorses NEI 99-03 to the extent possible and provides additional guidance. Licensees are not

required to comply with Regulatory Guide 1.196, but may find it useful in responding to this

generic letter. Licensees that are unable to confirm item 1 under the Requested Information

section may use Regulatory Guide 1.196 to develop and implement corrective actions.

Requested Information

Addressees are requested to provide the following information within 180 days of the date of

this generic letter.

1. Provide confirmation that your facility’s control room meets the applicable habitability

regulatory requirements (e.g., GDC 1, 3, 4, 5, and 19) and that the CRHSs are

designed, constructed, configured, operated, and maintained in accordance with the

facility’s design and licensing bases. Emphasis should be placed on confirming:

(a) That the most limiting unfiltered inleakage into your CRE (and the filtered

inleakage if applicable) is no more than the value assumed in your design basis

radiological analyses for control room habitability. Describe how and when you

performed the analyses, tests, and measurements for this confirmation.

(b) That the most limiting unfiltered inleakage into your CRE is incorporated into

your hazardous chemical assessments. This inleakage may differ from the value

assumed in your design basis radiological analyses. Also, confirm that the

reactor control capability is maintained from either the control room or the

alternate shutdown panel in the event of smoke.
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(c) That your technical specifications verify the integrity of the CRE, and the

assumed inleakage rates of potentially contaminated air. If you currently have a

∆P surveillance requirement to demonstrate CRE integrity, provide the basis for

your conclusion that it remains adequate to demonstrate CRE integrity in light of

the ASTM E741 testing results. If you conclude that your ∆P surveillance

requirement is no longer adequate, provide a schedule for: 1) revising the

surveillance requirement in your technical specification to reference an

acceptable surveillance methodology (e.g., ASTM E741), and 2) making any

necessary modifications to your CRE so that compliance with your new

surveillance requirement can be demonstrated. 

If your facility does not currently have a technical specification surveillance

requirement for your CRE integrity, explain how and at what frequency you

confirm your CRE integrity and why this is adequate to demonstrate CRE

integrity.

2. If you currently use compensatory measures to demonstrate control room habitability,

describe the compensatory measures at your facility and the corrective actions needed

to retire these compensatory measures.

3. If you believe that your facility is not required to meet either the GDC, the draft GDC, or

the “Principal Design Criteria” regarding control room habitability, in addition to

responding to 1 and 2 above, provide documentation (e.g., Preliminary Safety Analysis

Report, Final Safety Analysis Report sections, or correspondence) of the basis for this

conclusion and identify your actual requirements.

Requested Response

If an addressee cannot provide the information or cannot meet the requested completion date,

the addressee should submit a written response indicating this within 60 days of the date of this

generic letter. The response should address any alternative course of action the addressee

proposes to take, including the basis for the acceptability of the proposed alternative course of

action and the schedule for completing the alternative course of action. 

The written response should be addressed to the U.S. Nuclear Regulatory Commission, 

ATTN: Document Control Desk, Washington, DC 20555-0001. A copy of the response should

be sent to the appropriate regional administrator.

NRC staff will review the responses to this generic letter and, if concerns are identified, will

notify affected addressees. The staff may conduct inspections to determine licensees’

effectiveness in addressing this generic letter.

Applicable Regulatory Requirements

Several provisions of the NRC regulations and plant operating licenses (technical

specifications) pertain to the issue of control room habitability. The general design criteria for

nuclear power plants (10 CFR Part 50, Appendix A), or, as appropriate, the quality assurance

requirements in the licensing basis for a reactor facility (stated in 10 CFR Part 50, Appendix B,
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“Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants”), and the

technical specifications, are the bases for the NRC staff’s assessment of control room

habitability. 

Appendix A to 10 CFR Part 50 and the plant safety analyses require or commit licensees to

design and test safety-related structures, systems, and components (SSCs) to provide

adequate assurance that they can perform their safety functions. The NRC staff applies these

criteria to plants with construction permits issued on or after May 21, 1971, and to those plants

whose licensees have committed to them. The applicable GDC are GDC 1, 3, 4, 5, and 19. 

GDC 1 requires quality standards commensurate with the importance of the safety functions

performed. GDC 3 requires SSCs to be designed and located to minimize the effects of fires. 

GDC 4 requires SSCs to be designed to accommodate the effects of accidents. GDC 5

requires that an accident in one unit will not significantly impair orderly shutdown and cooldown

of the remaining unit.

GDC 19 specifies that a control room be provided from which actions can be taken to operate

the nuclear reactor safely under normal conditions and maintain the reactor in a safe condition

under accident conditions, including a loss-of-coolant accident. There must be adequate

radiation protection to permit personnel to access and occupy the control room under accident

conditions without receiving radiation exposures in excess of specified values.

Before the issuance of the GDC, proposed GDC (sometimes called “principal design criteria”)

were published in the Federal Register for comment. As they evolved, several of the proposed

GDC addressed control room habitability. A facility may have been licensed before the

issuance of the GDC, but the licensee may have committed to the proposed GDC as they

existed at the time of licensing.

Following the accident at TMI, TMI Action Plan Item III.D.3.4, “Control Room Habitability

Requirements,” as clarified in NUREG-0737, “Clarification of TMI Action Plan Requirements,”

required all licensees to assure that control room operators would be adequately protected

against the effects of accidental releases of toxic and radioactive gases and that the nuclear

power plant could be safely operated or shut down under design basis accident conditions. 

When licensees proposed modifications, the NRC issued orders confirming the licensees’

commitments. As a result, most plants licensed before the GDC were formally adopted were

then subsequently required to meet the TMI Action Plan III.D.3.4 requirements.

Appendix B to 10 CFR Part 50 establishes quality assurance requirements for the design,

construction, and operation of those SSCs that prevent or mitigate the consequences of

postulated accidents that could cause undue risk to the health and safety of the public. 

Appendix B, Criterion III, “Design Control,” requires that design control measures be provided

for verifying or checking the adequacy of design. A suitable testing program is identified as one

method of accomplishing this verification. Appendix B, Criterion XVI, “Corrective Action,”

requires measures to be established to assure that conditions adverse to quality, such as

failures, malfunctions, deficiencies, defective material and equipment, and nonconformances

are promptly identified and corrected.

The regulations in 10 CFR 50.36, “Technical Specifications,” require plant technical

specifications to be derived from the safety analyses.
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If, in the course of preparing a response to the requested information, an addressee determines

that its facility is not in compliance with the Commission’s requirements, the addressee is

expected to take appropriate action in accordance with requirements of Appendix B to

10 CFR Part 50 and the plant technical specifications to restore the facility to compliance.

Reasons for Information Request

This generic letter transmits an information request that is necessary to permit the assessment

of plant-specific compliance with applicable regulatory requirements. Specifically, this

information will enable the NRC staff to determine whether the control rooms at power reactor

facilities comply with the current licensing bases and whether additional regulatory actions are

required.

The habitability of the control room and the operability of the CRHSs in the event of adverse

environmental conditions external to the CRE have a direct link to maintaining public health and

safety. Plant design bases and severe accident risk analyses both assume that the control

room operators can remain safely within the control room to monitor plant performance and

take appropriate mitigative actions. It is essential that operators be confident of their safety

within the control room at all times.

Backfit Discussion

This generic letter transmits an information request for the purpose of verifying compliance with

existing applicable regulatory requirements (see the Applicable Regulatory Requirements

section of this generic letter). This generic letter does not constitute a backfit as defined in

10 CFR 50.109(a)(1) since it does not impose modifications or additions to structures, systems,

and components or to the design or operation of an addressee’s facility. Nor does it impose an

interpretation of the Commission’s rules that is either new or different from a previous staff

position. Therefore, no backfit is either intended or approved by this generic letter, and the staff

has not performed a backfit analysis.

Small Business Regulatory Enforcement Fairness Act

The NRC has determined that this action is not subject to the Small Business Regulatory

Enforcement Fairness Act of 1996. 

Federal Register Notification

A notice of opportunity for public comment was published in the Federal Register on

May 9, 2002 (67 FR 31385). Comments were received from three licensees, three industry

organizations, and one individual. The staff considered all comments that were received. 

The staff evaluation of these comments is accessible electronically from the Agencywide

Documents Access and Management System (ADAMS) at ML030780493. 

Paperwork Reduction Act Statement

This generic letter contains information collections that are subject to the Paperwork Reduction

Act of 1995 (44 U.S.C. 3501 et seq.) These information collections were approved by the

Office of Management and Budget (OMB), approval number 3150-0011, which expires 

January 31, 2004.
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The burden to the public for these information collections is estimated to average 200 hours per

response, including the time for reviewing instructions, searching existing data sources, 

gathering and maintaining the data needed, and completing and reviewing the information

collection. Send comments regarding this burden estimate or on any other aspect of these

information collections, including suggestions for reducing the burden, to the Records

Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, or by Internet electronic mail to INFOCOLLECTS@NRC.GOV; and to the Desk

Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0011), Office of

Management and Budget, Washington, DC 20503. 

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, an

information collection unless the requesting document displays a currently valid OMB control

number.

/RA/

David B. Matthews, Director

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Technical Contact: Mark Blumberg, NRR

301-415-1083

E-mail: wmb1@nrc.gov

Lead Project Manager: Michael Webb

301-415-1347

E-mal: mkw@nrc.gov