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Access and Managem ent System (ADAMS) at http://www.nrc.gov/reading-rm /adam s.htm l, under Accession No. ML063560144.

U.S. NUCLEAR REGULATORY COMMISSION January 2007

Revision 1

REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.196

CONTROL ROOM HABITABILITY

AT LIGHT-WATER NUCLEAR POWER REACTORS

A. INTRODUCTION

This guide provides guidance and criteria that the staff of the U.S. Nuclear Regulatory Commission

(NRC) considers acceptable for implementing the agency’s regulations in Appendix A, “General Design

Criteria for Nuclear Power Plants,” to Title 10, Part 50, of the Code of Federal Regulations (10 CFR

Part 50), “Domestic Licensing of Production and Utilization Facilities,” as they relate to control room

habitability (CRH). Specifically, this guide outlines a process that licensees may apply to control rooms

that are modified, are newly designed, or must have their conformance to the regulations reconfirmed.

In Appendix A to 10 CFR Part 50, General Design Criteria (GDC) 1, 3, 4, 5, and 19 apply to CRH,

as follows:

• GDC 1, “Quality Standards and Records,” requires that structures, systems, and components

(SSCs) important to safety be designed, fabricated, erected, and tested to quality standards

commensurate with the importance of the safety functions performed.

• GDC 3, “Fire Protection,” requires that SSCs important to safety be designed and located

to minimize the effects of fires and explosions.

• GDC 4, “Environmental and Dynamic Effects Design Bases,” requires SSCs important to safety

to be designed to accommodate the effects of, and to be compatible with, the environmental

conditions associated with normal operation, maintenance, testing, and postulated accidents,

including loss-of-coolant accidents (LOCAs).
Rev. 1 of RG 1.196, Page 2

• GDC 5, “Sharing of Structures, Systems, and Components,” requires that SSCs important to safety

not be shared among nuclear power units unless it can be shown that such sharing will not

significantly impair their ability to perform their safety functions, including, in the event

of an accident in one unit, the orderly shutdown and cooldown of the remaining units.

• GDC 19, “Control Room,” requires that a control room be provided from which actions can be

taken to operate the nuclear reactor safely under normal conditions and to maintain the reactor

in a safe condition under accident conditions, including a LOCA. Adequate radiation protection

is to be provided to permit access and occupancy of the control room under accident conditions

without personnel receiving radiation exposures in excess of specified values.

Since the NRC initially issued Regulatory Guide 1.196 in May 2003, the staff determined

that the information presented in Appendix B to that guide did not accurately represent a viable

technical specification for CRH at light-water nuclear power reactors. In particular, it referred to

failure of a particular surveillance as a plant state, rather than having the results of the surveillance factor

into the operability determination. In addition, it did not provide for a definite time to restore functionality

to the control room envelope, whereas all improved standard technical specifications (iSTS) contain

such provisions. Moreover, Appendix B was included as a “strawman,” to be deleted when details

had been more carefully worked out with industry participation, and those technical specifications

placed in the iSTS with all other acceptable technical specifications.

As of the publication date of this Revision 1 of Regulatory Guide 1.196, no utility has been granted

the technical specification changes represented by Appendix B to the original version of this guide.

Consequently, the NRC staff elected to remove Appendix B (and all related references) from this revision.

Removal of Appendix B from this revised guide does not require any stakeholder to take any action

and does not reduce safety in any way. Moreover, the owners’ group Technical Specification Task Force

has provided ample opportunity for public comment regarding this revision. Therefore, the staff views

the removal of Appendix B as a neutral action, for which further public comments are unnecessary. 

For that reason, the staff chose not to issue this revision as a draft guide for public comment

before publishing this Revision 1 of Regulatory Guide 1.196.

This regulatory guide contains information collections that are covered by the requirements

of 10 CFR Part 50, and that the Office of Management and Budget (OMB) approved under OMB control

number 3150-0011. The NRC may neither conduct nor sponsor, and a person is not required to respond to,

an information collection request or requirement unless the requesting document displays a currently

valid OMB control number.
An example of a changed performance parameter that may require re-analysis is an increase in control room envelope 1

(CRE) inleakage beyond that assumed in previous CRH assessments.

As used in this guide, the licensing basis is the documentation that describes how the plant meets applicable regulations. 

2

Design bases are defined in 10 CFR 50.2. Regulatory Guide 1.186, “Guidance and Examples for Identifying

10 CFR 50.2 Design Bases” (Ref. 2), provides additional guidance. The design bases are a subset of the licensing bases.

Thus, this guide uses “licensing bases” to refer to both.

Rev. 1 of RG 1.196, Page 3

B. DISCUSSION

A licensee may use this guide for assessing CRH following changes to control room habitability

systems (CRHSs) or the sources that would lead to consequences to the operator. Changes that may

impact the existing CRH assessments and may result in re-analysis of the licensee’s CRH include

the following examples:

• changes in procedures, operation, performance, alignment, or function of the CRHSs 1

• new hazardous chemicals or radioactive sources introduced onsite or in the vicinity of the plant

• increases in hazardous chemical or radioactive source quantities, concentrations, locations,

or shipments

The primary design function of CRHSs is to provide a safe environment in which the operator

can keep the nuclear reactor and auxiliary systems under control during normal operations and can safely

shut down those systems during abnormal situations to protect the health and safety of the public

and plant workers. If the control room is not habitable or the response of the operator is impaired

during an accident, there could be increased consequences to public health and safety. It is important

for the operators to be confident of their safety in the control room to minimize errors of omission

and commission. The regulatory positions in Section C provide methods that the NRC staff

considers acceptable for ensuring that the public and the control room operator are protected.

When possible, this guide incorporates guidance contained in NEI 99-03, “Control Room

Habitability Assessment Guidance” (Ref. 1). The staff has reviewed this document and concluded

that portions of NEI 99-03 can serve as a valuable resource on CRH. Only the sections of NEI 99-03

that are specifically stated in the regulatory positions should be considered to be endorsed by the NRC.

The staff’s endorsement of these sections should not be considered an endorsement of the remainder

of NEI 99-03 or of any other document referenced in NEI 99-03. Appendix A to this guide summarizes

the staff’s endorsement.

Definitions of key terms used within the context of this regulatory guide are given below. 

However, in most cases a facility’s licensing basis and associated documents will define the terms 2

for a particular facility.

Control Room: The plant area, defined in the facility’s licensing basis, in which actions can be

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

during accident situations. It encompasses the instrumentation and controls necessary for safe shutdown

of the plant and typically includes the critical document reference file, computer room (if used as

an integral part of the emergency response plan), shift supervisor’s office, operator wash room and kitchen,

and other critical areas to which frequent personnel access or continuous occupancy may be necessary

in the event of an accident.
Rev. 1 of RG 1.196, Page 4

Control Room Envelope (CRE): The plant area, defined in the facility’s licensing basis, that

in the event of an emergency, can be isolated from the plant areas and the environment external to the CRE. 

This area is served by an emergency ventilation system, with the intent of maintaining the habitability

of the control room. This area encompasses the control room, and may encompass other non-critical areas

to which frequent personnel access or continuous occupancy is not necessary in the event of an accident.

Control Room Habitability Systems (CRHSs): The systems, defined in the facility’s licensing basis,

that typically provide the functions of shielding, isolation, pressurization, heating, ventilation,

air conditioning and filtration, monitoring, and sustenance and sanitation necessary to ensure that

operators can remain in the control room, take actions to operate the plant under normal conditions,

and maintain it in a safe condition during accident situations. The CRHSs include the CRE.
Rev. 1 of RG 1.196, Page 5

C. REGULATORY POSITION

1. An Overview of the Process of Demonstrating and Maintaining CRH

In demonstrating that a facility’s control room conforms to the GDCs, the following CRH aspects

are typically assessed:

• radiological doses

• protection from the effects of hazardous materials

• control of the reactor from either the control room or the alternate shutdown panel

The process of demonstrating the above aspects includes the following actions:

(a) identification of the licensing bases for the (i) CRHSs, (ii) areas adjacent to the CRE,

and (iii) ventilation systems that serve or traverse the CRE and those located in areas adjacent to

the CRE

(b) determinations of whether the design, configuration, and operation of the systems and areas

identified in action (a) are consistent with the licensing bases

(c) determination of the performance characteristics for operating modes associated with

radiological and hazardous chemical accidents

(d) calculation of the radiological dose consequences to control room operators

(e) evaluation of the habitability of the control room during a postulated hazardous chemical release

(f) assessment of whether a radiological, hazardous chemical, or smoke challenge could result in

the inability of the control room operators to control the reactor from either the control room

or, in the event of smoke, from the alternate shutdown panel

(g) maintenance and monitoring of the CRHSs

2. Demonstrating and Maintaining CRH

Regulatory Positions 2.1 through 2.7 provide guidance on the process of demonstrating

and maintaining CRH.

2.1 Identification of the Licensing Bases for the CRHSs

2.1.1 Identification of the Control Room and the CRE

Confirmation of a facility’s ability to meet CRH requirements begins with the identification

of the control room and the CRE. A description of the control room and CRE may be contained

in a number of plant documents. These documents might include the updated final safety analysis report

(UFSAR), the original final safety analysis report (FSAR), the safety evaluation for the operating license

(OL), system descriptions, plant drawings, operating procedures, plant amendment requests,

NRC safety evaluations, Three Mile Island (TMI) Action Plan Item III.D.3.4 submittals, and responses

to staff questions at the construction permit and OL stages.
Rev. 1 of RG 1.196, Page 6

2.1.2 Determination of the Licensing Bases

In demonstrating the habitability of a facility’s control room, it is essential that the licensee

know the facility’s licensing bases for its CRHSs. The sources of the licensing bases of the CRHSs

should be identified. Licensees should consider the documents identified in Section 4.3 of NEI 99-03

(Ref. 1) as potential sources that define the licensing bases for their CRHSs. Focusing on the events

that may have established or changed these bases may help narrow this search.

Over the facility’s lifetime, the licensing bases change. The staff may have reviewed and approved

the licensing bases of facilities licensed before the issuance of this guide. The original licensing bases

may have been submitted as part of the construction permit application. Licensees may have modified them

in response to NRC questions. In addition, the licensing bases were part of the application for the OL

(FSAR). Depending on the plant vintage, licensees may have modified their licensing bases

in response to TMI Action Plan Item III.D.3.4. Amendments to the OL may have resulted in changes

to the licensing bases of the CRHSs. Licensees should review the applicable plant changes to their

licensing bases to determine the current bases.

A group of reactors received their construction permits or OLs before the GDCs were promulgated. 

During that time, proposed GDCs (sometimes called “Principal Design Criteria”) were published

in the Federal Register for comment. These proposed GDCs addressed CRH. Although facilities

may have been licensed before the promulgation of the GDCs, licensees may have committed to

the form of the GDCs as they existed at the time of licensing. A review of the record associated with

the construction permit and OL proceedings should confirm whether licensees made such a commitment. 

Therefore, licensees that received their construction permits or OLs before the GDCs were promulgated

should review their commitments to the draft form of the GDC to understand their CRH licensing bases.

For facilities licensed following the issuance of this regulatory guide, the sources for the description

of the licensing bases will be the documents filed in support of the licensing application (under 10 CFR

Parts 50 and 52).

2.2 Determination of Whether the CRHSs Are Consistent with the Licensing Bases

2.2.1 Comparison of System Design, Configuration, and Operation with the Licensing Bases

Licensees should compare the design, configuration, and operation of their CRHSs and the systems

that are in adjacent areas and could interact with the CRE to their licensing bases to ensure consistency. 

The review of the configuration of the CRHSs should include the construction and the alignment

of the systems and structures that make up the CRHSs. For new reactors and existing CRHSs undergoing

redesign, this comparison should be made upon completion of construction. Section 5 of NEI 99-03 (Ref. 1)

provides a method of comparing the plant’s configuration and operation of ventilation systems with

the licensing bases that is acceptable to the NRC staff with one clarification. Licensees should also

establish the performance characteristics discussed in Regulatory Position 2.3.1 to ensure consistency

between the operation of the control room ventilation systems and the licensing bases. Licensees should

employ methods similar to those provided in Section 5 of NEI 99-03 when they perform these

comparisons for other CRHSs.
Rev. 1 of RG 1.196, Page 7

2.2.2 Interactions Between the CRHSs and Adjacent Areas

The conditions that exist in the areas adjacent to the CRE influence the performance of the CRHSs.

Although these systems might not be expected to operate during an emergency, during a loss of offsite

power (LOOP), or with a single failure, inleakage may be increased if they do operate. Potential

interactions between the CRHSs and adjacent areas that may increase the transfer of contaminants into

the CRE should be identified. These interactions may be caused by ventilation systems that supply or

exhaust air from areas adjacent to the CRE, are located in areas adjacent to the CRE, or have ductwork that

traverses the CRE or areas adjacent to the CRE.

2.3 Determination of Performance Characteristics

2.3.1 Performance of CRHSs

The licensee should determine the performance characteristics of the CRE, its ventilation systems, and

systems that serve or traverse areas within or located adjacent to the CRE. These parameters include,

but are not limited to, differential pressures, system flow measurements (i.e., makeup and recirculation

flow rates), duct static pressures, and filter differential pressures. Performance characteristics are needed

to achieve the following objectives:

• Establish the operating parameters for incorporation into the licensing bases (for new reactors

or those that have modified their CRE or associated ventilation systems).

• Determine the impact on systems caused by changes in the operation, design, alignment,

or procedures.

• Define the limiting condition for the applicable design basis events.

• Determine new limiting conditions or perform new analyses.

Technical specifications require licensees to periodically perform measurements of several

parameters important to maintaining CRH. These parameters may include system flow rates, carbon filter

efficiencies, actuation signals, and CRE integrity tests. Engineered-safety-feature atmospheric cleanup

systems in light-water-cooled nuclear power plants should be tested and evaluated in accordance with

Regulatory Guide 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” (Ref. 3).

In CRE integrity tests performed in 1991 – 2001 by approximately 30 percent of the licensed

facilities, all but one facility measured greater inleakage than that assumed in the design analyses. 

In some cases, the measured inleakage exceeded the amount assumed in the design analyses by several

orders of magnitude. Regulatory Guide 1.197, “Demonstrating Control Room Envelope Integrity

at Nuclear Power Reactors” (Ref. 4), provides guidance on this issue and an approach that the NRC staff

considers acceptable to determine CRE integrity. As discussed in Regulatory Position 2.2.2,

systems outside the CRE may impact CRE integrity. Testing may also be needed to understand

the influence of these systems on CRH. Consistent with Regulatory Position 2.2.1, licensees should ensure

that their assumed control room inleakage input value used in any accident calculations or evaluations

(Regulatory Positions 2.4 and 2.5) are validated by the test methods provided in Regulatory Guide 1.197.
Rev. 1 of RG 1.196, Page 8

2.3.2 Identification of the Limiting Condition

The limiting condition for CRH is the configuration that results in the maximum consequences

to the control room operators. Sometimes the limiting condition will arise from the configuration

that produces the greatest inleakage and sometimes it will not. The latter situation can occur because

the configuration that results in the largest inleakage may have mitigative features that result in smaller

consequences to the control room operators. As an example, CRE inleakage may be greatest

for a radiological accident that does not have a LOOP. However, the absence of a LOOP could provide

mitigative features that reduce the overall consequences to the control room operators.

In determining the limiting condition for potential radiological accidents, it should not be

presumed that the LOCA is the limiting accident because it has the largest initial source of activity. 

Other accidents (e.g., fuel handling accidents) may produce larger control room operator doses because

the manner in which the CRHSs respond may provide less protection to the operators. Therefore,

licensees should perform an analysis of the consequences of each applicable radiological accident

as discussed in Regulatory Position 2.4 to ensure that they have identified the limiting accident.

Unless a facility relies on a common control room isolation process for all types of radiological

accidents, the limiting accident may not be obvious for the following reasons:

• The inleakage characteristics of the envelope may vary with the CRHS’s response to an accident.

• The mitigative equipment used to reduce the radioactivity released to the environment may vary

with the accident.

• The location of the release points for the various accidents relative to the control room intakes

may result in less favorable atmospheric dispersion and higher magnitude intake concentrations.

Licensees should factor all the potential differences in accidents and the CRHS’s performance

in order to determine the limiting condition.

A few plants are within the exposure range for an accidental release from a nearby nuclear plant

or have separate control rooms for multiple units on the same site. An accident in an adjacent unit

should not prevent the safe shutdown of an operating unit. Regulatory Position 2.6, “Reactor Control,”

describes criteria used for determining a safe shutdown of the reactor. The release point, atmospheric

dispersion, and postulated source term from the adjacent unit should be reviewed to assess the impact

on the operating unit’s control room.

For hazardous chemicals, a logic process similar to that employed for radiological accidents

should be used to determine the limiting condition.

2.4 Radiological Consequences

Licensees should calculate control room operator doses for the methodology and accidents

identified in Regulatory Guide 1.195 (Ref. 5) or Regulatory Guide 1.183 (Ref. 6). For CREs under

construction, the control room operators’ doses should be based on expected CRHS performance values.

When the envelope and associated ventilation systems are operational, the inleakage value should be

determined using Regulatory Guide 1.197 (Ref. 4).
Rev. 1 of RG 1.196, Page 9

2.5 Hazardous Chemicals

Licensees should evaluate the impact of hazardous chemicals on control room operators

using the methodology of Regulatory Guide 1.78, “Evaluating the Habitability of a Nuclear Power Plant

Control Room During a Postulated Hazardous Chemical Release” (Ref. 7). Regulatory Guide 1.78

encourages licensees to conduct periodic surveys of stationary and mobile sources of hazardous chemicals

in the vicinity of their plant sites. The periodicity should be based on the number, size, and type

of industrial and transportation activities in the vicinity of the plant and regional and local changes in uses

of land. The staff recommends conducting a survey of the location, types, and quantities of the mobile

and stationary hazardous chemical sources at least once every 3 years, or more frequently as applicable. 

The staff also recommends annual performance of an onsite survey of hazardous chemical sources.

For CREs under construction, the hazardous chemical analysis should be based on the expected

performance values. When the envelope and associated ventilation systems are operational, the calculation

should be based on an inleakage value determined according to Regulatory Guide 1.197 (Ref. 4).

2.6 Reactor Control

This Regulatory Position provides guidance for assessing whether a radiological, hazardous

chemical, or smoke challenge could result in the inability of the control room operators to control the reactor

from the control room, or in the event of a smoke challenge, from the control room or the alternate

shutdown panel. This regulatory position does not address the performance of the reactor controls

and instrumentation systems that are affected by environmental conditions caused by a radiological,

hazardous chemical, or smoke event, nor does it address human engineering (i.e., temperature, vibration,

sound, or lighting).

Demonstrating the facility’s ability to maintain a habitable control room includes ensuring that

an accident arising from a radiological event, hazardous chemicals, or a smoke challenge will not prevent

the control room operators from controlling the reactor. Facilities should demonstrate that they meet

the reactor control aspects of their licensing bases (typically GDC-19). The specific acceptance criteria

for radiological events are summarized in Regulatory Position 4.5 of Regulatory Guide 1.195 (Ref. 5)

for plants employing TID-14844 source term methodology and Regulatory Position 4.4 of Regulatory

Guide 1.183 (Ref. 6) for plants employing alternative source term methodology. The specific acceptance

criterion for chemical events is given in Regulatory Position 3.1 of Regulatory Guide 1.78 (Ref. 7).

Smoke may be a CRH concern if there is significant inleakage from outside the envelope

or if a fire develops in the control room. In these situations, smoke could challenge the operator’s ability

to shut down the reactor from within the control room or remotely. No regulatory limit exists on the amount

of smoke allowed in the control room. Therefore, the plant’s ability to manage smoke infiltration

is assessed qualitatively. Licensees should perform a qualitative assessment to ensure that the plant

can be safely shut down from either the control room or the alternate shutdown panel during an internal

or external smoke event. The staff endorses Appendix E of NEI 99-03 (Ref. 1) as an acceptable method

for performing this qualitative assessment with the following exceptions. The second sentence of Section 1

should read, “The guidance ensures that the operator maintains an ability to safely shut down the plant

during a smoke event originating inside or outside the control room.” Replace the words “fire/smoke event”

in the first sentence of Section 2 with “smoke event originating from either inside or outside the control room.” 

The title of Section 3 should be “Contingency Logic Evaluation,” and the third bullet should be deleted.

The last bullet in Section 3 should be the last bullet in Section 2.
Rev. 1 of RG 1.196, Page 10

2.7 Maintaining and Monitoring CRHSs

CRH is maintained and monitored during the operating life of the plant by a CRHS program. 

A CRHS program includes periodic evaluations, maintenance, configuration control, and training. 

This Regulatory Position covers CRHS programs, and it provides methods to mitigate degraded

and nonconforming conditions when the plant does not meet the specific acceptance criteria given

in Regulatory Position 2.6 or is outside its licensing basis. The following methods of maintaining

and monitoring CRHSs should be used.

2.7.1 Periodic Evaluations and Maintenance

Periodic evaluations of CRH demonstrate that the CRHSs meet their functional criteria. 

These evaluations include periodic assessments and tests.

Periodic assessments of the CRHS’s material condition, configuration controls, safety analyses,

and operating and maintenance procedures should be performed. CRHS programs should assess the system

and material conditions as described in Section 9.3.1, “System Material Condition,” of NEI 99-03 (Ref. 1).

Licensees should perform testing to ensure they maintain CRH. Routine performance

measurements are described in Regulatory Position 2.3.1. The complexity of testing following

modifications should depend on the effect of the modification on CRH. Regulatory Guide 1.197 (Ref. 4)

provides a testing method for verification of CRE integrity. A frequency for CRE integrity testing

is provided in Regulatory Guide 1.197. Regulatory Position 2.5 above provides a method and a suggested

frequency for evaluating the impact of hazardous chemicals on control room operators.

A maintenance program should be established for the CRHSs. Table H-1 of NEI 99-03 (Ref. 1)

should be used to identify systems and components to be included in a maintenance program. Guidance

on air filtration and adsorption units of post-accident engineered-safety-feature atmosphere cleanup

and normal atmospheric cleanup system maintenance is provided in Regulatory Position 5 of Regulatory

Guides 1.52 (Ref. 3) and 1.140, “Design, Inspection, and Testing Criteria for Air Filtration and Adsorption

Units of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants” (Ref. 13),

respectively. ASHRAE Guideline 1-1996 (Ref. 14) may be used to establish a maintenance program

for systems that handle hazardous chemical and smoke challenges.

2.7.2 Configuration Control and Training

Configuration control and training are effective tools that can minimize the impact that changes

to CRHSs can have on CRH. Section 9.4 of NEI 99-03 (Ref. 1) provides configuration controls

that include CRE boundary and breach control, procedure control, toxic gas control, design change,

and safety analysis controls.

The staff endorses the controls discussed in Sections 9.4.1 through 9.4.5 of NEI 99-03 (Ref. 1)

with two exceptions. The staff does not endorse Appendix K, “Control Room Envelope Boundary

Control Program,” referenced in Section 9.4.1, “CRE Boundary/Breach Control.” Instead of endorsing

the method of equating a breach size to an inleakage flow rate, the staff endorses the method of breach

control contained in the STSs (Refs. 8, 9, 10, 11, 12), which allows the control room boundary

to be opened intermittently under administrative controls.
Rev. 1 of RG 1.196, Page 11

For entry and exit through doors, the administrative control of the opening is performed

by the persons entering or exiting the area. For other openings, these controls consist of stationing

a dedicated individual at the opening who is in continuous communication with the control room. 

This individual will have a method to rapidly close the opening when a need for control room isolation

is indicated. Regulatory Guides 1.183 or 1.195 should be used instead of Appendix C (referenced in

Section 9.4.5); Regulatory Guide 1.194 should be used instead of Appendix D (referenced in Section 9.4.5);

Regulatory Guide 1.78 should be used instead of Appendix G (referenced in Section 9.4.5) (Refs. 6, 5,

15, and 7).

Furthermore, the staff endorses Section 9.5, “Training,” of NEI 99-03 with one exception. 

Section 9.5 recommends training using NEI 99-03. Instead, the NRC staff endorses training using only

the sections of NEI 99-03 that the staff has endorsed.

2.7.3 Degraded and Nonconforming Conditions

Methods available to address short term degraded or nonconforming conditions are provided

in Section 8.4, “Methods Available to Address Degraded or Nonconforming Conditions” of NEI 99-03

(Ref. 1). Section 8.4 includes guidance on compensatory measures such as self-contained breathing

apparatus (SCBA) and potassium iodide (KI) tablets. These methods are acceptable with the following

exceptions. Appendices C and D are not endorsed; instead, Regulatory Guides 1.194, “Atmospheric

Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power

Plants” (Ref. 15), and 1.195 (Ref. 5), or Regulatory Guide 1.183 (Ref. 6) should be used. The staff

endorses the use of the guidance in Appendix F of NEI 99-03 while corrective actions are being taken

to resolve CRHSs that do not meet their licensing bases, subject to the following:

(1) Section 2.2 of Appendix F addresses the training and qualification of control room operators

for SCBA. If SCBA units will be used as an interim compensatory measure for greater than

90 days while the plant is in Operating Condition or Mode 1, simulator crew training accident

scenarios in which operators wear SCBAs should be performed. These scenarios should last

about 2 hours and include a simulated watch turnover.

(2) Section 2.6 of Appendix F addresses the availability of adequate methods to refill depleted

SCBA cylinders. The impact of a LOOP or airborne contamination at the refill compressor

stations should be considered.

(3) Section 2 of Appendix F addresses additional guidance for evaluating the habitability of a control

room during a chemical release. Replace the sentence beginning with “Additional guidance”

with “Additional guidance is provided in Regulatory Guide 1.78, ‘Evaluating the Habitability

of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release’” (Ref. 7).

(4) Section 2.5 and 2.6 of Appendix F should be replaced with the following to correct an editorial error:

“2.5 Persons Using Tight Fitting (Facepiece) Respirators Should Not Have Any Facial Hair That

Interferes With the Sealing Surfaces of the Respirator. The required minimum staffing of control

room operators qualified in SCBA use should be clean shaven.

2.6 Adequate Method(s) to Refill SCBA Air Cylinders Should Be Available.”

Some licensees were allowed to leave TMI Action Item III.D.3.4 actions open until the alternative

source term rulemaking and regulatory guidance were published. These actions were completed

with the issuance of 10 CFR 50.67 and Regulatory Guide 1.183 (Ref. 6). The regulatory positions

in this regulatory guide on control room habitability provide methods that the NRC staff considers

acceptable for closing open TMI Action Plan Item III.D.3.4 actions.
Draft Regulatory Guide DG-1114 is available electronically under Accession #ML020790125 in the NRC’s 3

Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html.

Copies are also available for inspection or copying for a fee from the NRC’s Public Document Room (PDR), which is

located at 11555 Rockville Pike, Rockville, Maryland; the PDR’s mailing address is USNRC PDR, Washington, DC

20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548,

and by email to PDR@nrc.gov.

Rev. 1 of RG 1.196, Page 12

D. IMPLEMENTATION

The purpose of this section is to provide information to applicants and licensees regarding

the NRC staff’s plans for using this regulatory guide. No backfitting is intended or approved

in connection with the issuance of this guide.

Except in those cases in which an applicant or licensee proposes or has previously established

an acceptable alternative method for complying with specified portions of the NRC’s regulations,

the NRC staff will use the methods described in this guide to evaluate CRH for nuclear power plants

for which the construction permit or license application is docketed after the issue date of this guide

and plants for which the licensees voluntarily commit to the provisions of this guide.

REGULATORY ANALYSIS

The regulatory analysis for this regulatory guide is available in Draft Regulatory Guide DG-1114,

“Control Room Habitability at Light-Water Nuclear Power Reactors.” The NRC issued DG-1114 3

in March 2002 to solicit public comment on the initial draft of Regulatory Guide 1.196.

BACKFIT ANALYSIS

The regulatory guide does not require a backfit analysis as described in 10 CFR 50.109(c)

because it does not impose a new or amended provision in the NRC’s rules or a regulatory staff position

interpreting the NRC’s rules that is either new or different from a previous applicable staff position. 

In addition, this regulatory guide does not require the modification or addition to systems, structures,

components, or design of a facility or the procedures or organization required to design, construct,

or operate a facility. Rather, a licensee or applicant may select a preferred method for achieving

compliance with a license or the rules or orders of the Commission as described in 10 CFR 50.109(a)(7). 

This regulatory guide provides an opportunity to use part of an industry-developed standard.
Copies are also available for inspection or copying for a fee from the NRC’s Public Document Room (PDR), which is 1

located at 11555 Rockville Pike, Rockville, Maryland; the PDR’s mailing address is USNRC PDR, Washington, DC

20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548,

and by email to PDR@nrc.gov.

Requests for single copies of draft or active regulatory guides (which may be reproduced) or for placement on an 2

automatic distribution list for single copies of future draft guides in specific divisions should be made in writing to the

U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services

Section, or by fax to (301)415-2289; email DISTRIBUTION@nrc.gov. Copies are also available for inspection or

copying for a fee from the NRC’s Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville,

Maryland; the PDR’s mailing address is USNRC PDR, Washington, DC 20555-0001. The PDR can also be reached

by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email to PDR@nrc.gov.

Available through NRC’s Public Web site, 

3

http://www.nrc.gov, in the Electronic Reading Room through ADAMS

(using the accession number) and online at http://www.nrc.gov/reactors/operating/licensing/techspecs.html.

Rev. 1 of RG 1.196, Page 13

REFERENCES

1. NEI 99-03, “Control Room Habitability Assessment Guidance,” Revision 0,

Nuclear Energy Institute, Washington, DC, June 2001.

1

2. Regulatory Guide 1.186, “Guidance and Examples for Identifying 10 CFR 50.2 Design Bases,”

U.S. Nuclear Regulatory Commission, Washington, DC, December 2000.2

3. Regulatory Guide 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,” Revision 3, U.S. Nuclear Regulatory Commission,

Washington, DC, June 2001.

2

4. Regulatory Guide 1.197, “Demonstrating Control Room Envelope Integrity at Nuclear Power

Reactors,” U.S. Nuclear Regulatory Commission, Washington, DC, May 2003.2

5. Regulatory Guide 1.195, “Methods and Assumptions for Evaluating Radiological Consequences

of Design-Basis Accidents at Light-Water Nuclear Power Reactors,” U.S. Nuclear Regulatory

Commission, Washington, DC, May 2003.2

6. Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis

Accidents at Nuclear Power Reactors,” U.S. Nuclear Regulatory Commission, Washington, DC,

July 2000.2

7. Regulatory Guide 1.78, “Evaluating the Habitability of a Nuclear Power Plant Control Room

During a Postulated Hazardous Chemical Release,” Revision 1, U.S. Nuclear Regulatory

Commission, Washington, DC, December 2001.2

8. NUREG-1431, “Standard Technical Specifications Westinghouse Plants: Specifications,”

Volume 1, Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, June 2001.

3

(ADAMS Accession Number ML011840223)

9. NUREG-1430, “Standard Technical Specifications Babcock and Wilcox Plants,” Volume 1,

Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, June 2001.

3

(ADAMS Accession Number ML011770186)
Rev. 1 of RG 1.196, Page 14

10. NUREG-1432, “Standard Technical Specifications Combustion Engineering Plants,” Volume 1,

Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, June 2001.

3

(ADAMS Accession Number ML011930335)

11. NUREG-1433, “Standard Technical Specifications General Electric Plants, BWR/4,” Volume 1,

Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, June 2001.

3

(ADAMS Accession Number ML011780639)

12. NUREG-1434, “Standard Technical Specifications General Electric Plants, BWR/6,” Volume 1,

Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, June 2001.

3

(ADAMS Accession Number ML011780537)

13. Regulatory Guide 1.140, “Design, Inspection, and Testing Criteria for Air Filtration and Adsorption

Units of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants,”

Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, June 2001.

2

14. ASHRAE Guideline 1-1996, “The HVAC Commissioning Process,” American Society

of Heating, Refrigerating and Air Conditioning Engineers, Atlanta, Georgia, June 1996.

15. Regulatory Guide 1.194, “Atmospheric Relative Concentrations for Control Room Radiological

Habitability Assessments at Nuclear Power Plants,” U.S. Nuclear Regulatory Commission,

Washington, DC, May 2003.2
Appendix A to Rev. 1 of RG 1.196, Page A-1

APPENDIX A

REGULATORY GUIDE ENDORSEMENT

OF NEI 99-03, REV. 0, JUNE 2001, BY SECTION

NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

1 Introduction Not addressed N/A N/A

1.1 Purpose And Scope Not addressed N/A N/A

1.2 History Not addressed N/A N/A

1.3 Document Organization Not addressed N/A N/A

2 Regulatory Requirements and Guidance Not addressed N/A N/A

2.1 Purpose And Scope Not addressed N/A N/A

2.2 Regulatory Requirement: 

General Design Criterion 19

Not addressed N/A N/A

2.3 Regulatory Guidance Not addressed N/A N/A

2.3.1 Regulatory Guides Not addressed N/A N/A

2.3.2 NUREGs Not addressed N/A N/A

2.3.3 Information Notices Not addressed N/A N/A

2.4 Generic Issues Not addressed N/A N/A
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-2

3 Industry Issues Associated with

Control Room Habitability

Not addressed N/A N/A

3.1 Purpose And Scope Not addressed N/A N/A

3.2 Licensing Basis Different From As-Built

Plant

Not addressed N/A N/A

3.3 Analyses Different From As-built or

As-Operated Plant

Not addressed N/A N/A

3.4 DBA Analyzed Not Most Limiting Not addressed N/A N/A

3.4.1 Adjacent Unit Accident (a Special Case) Not addressed N/A N/A

3.5 Smoke Infiltration Not addressed N/A N/A

3.6 Toxic Gas Evaluation Not addressed N/A N/A

3.7 Control Room Air In-Leakage

Greater Than Assumed

Not addressed N/A N/A

3.7.1 Radiological Considerations Not addressed N/A N/A

3.7.2 Toxic Gas Considerations Not addressed N/A N/A
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-3

4 Determining CRH Licensing Basis Not addressed N/A N/A

4.1 Purpose And Scope Not addressed N/A N/A

4.2 Understanding the Concept

of Licensing Basis

Not addressed N/A N/A

4.2.1 Design Basis Not addressed N/A N/A

4.2.2 Supporting Design Information Not addressed N/A N/A

4.2.3 Licensing Basis Not addressed N/A N/A

4.3 Licensing Basis Sources Full endorsement Regulatory Guide

1.196, Regulatory

Position (RP) 2.1.2

N/A

4.4 Performing the Licensing Basis Review Not addressed N/A N/A

4.5 Assembling the CRH Analysis Not addressed N/A N/A

4.6 Documentation of the Existing

Plant CRH Licensing and Design Basis

Not addressed N/A N/A
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-4

5 Comparing Existing Plant Configuration

and Operations with Licensing Bases

for CRH

Full endorsement RG 1.196, RP 2.2.1 This section provides a method of comparing the

plant’s configuration and operation of ventilation

systems with the licensing bases that is acceptable

to the NRC staff with one clarification. 

Licensees should also establish the performance

characteristics discussed in Regulatory Position

2.3.1 to ensure consistency between the operation

of the control room ventilation systems and the

licensing bases.

5.1 Purpose and Scope Full endorsement RG 1.196, RP 2.2.1 N/A

5.2 Review the As-Built Control Room

Envelope and Control Room Ventilation

Systems

Full endorsement RG 1.196, RP 2.2.1 N/A

5.3 Review the Normal and Emergency

Operating Procedures Affecting

the Control Room Ventilation Systems

Full endorsement RG 1.196, RP 2.2.1 N/A

5.4 Review the Testing Procedures

Affecting Control Room Ventilation

Systems and the Associated Envelope

Full endorsement RG 1.196, RP 2.2.1 N/A

5.5 Review the Maintenance Practices

and Procedures for Effect on CRH

Requirements

Full endorsement RG 1.196, RP 2.2.1 N/A

5.6 Review the Plant Modification

Procedures for Consideration

of the CRH Requirements

Full endorsement RG 1.196, RP 2.2.1 N/A

5.7 Review the CRH Analyses Full endorsement RG 1.196, RP 2.2.1 N/A

5.8 Identified Inconsistencies Full endorsement RG 1.196, RP 2.2.1 N/A
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-5

6 Assessing Industry Issue Applicability Not addressed N/A N/A

6.1 Purpose and Scope Not addressed N/A N/A

6.2 Limiting DBA Not addressed N/A N/A

6.2.1 Recommended Actions to Evaluate

Limiting DBA

Not addressed N/A N/A

6.2.2 Adjacent Unit Accidents Not addressed N/A N/A

6.3 Smoke Infiltration Not endorsed N/A N/A

6.3.1 Recommended Licensee Action

to Address Smoke Infiltration

Not endorsed N/A N/A

6.4 Toxic Gas Evaluation Not addressed N/A N/A

6.4.1 Recommended Licensee Action

to Address Toxic Gas Evaluation

Not Addressed N/A N/A

7 Measuring Air In-Leakage

(Baseline Test)

Not addressed N/A N/A

7.1 Purpose and Scope Not addressed N/A N/A

7.2 Preparation for Testing Not addressed N/A N/A

7.3 Test Performance Not addressed N/A N/A

7.4 Resolution of Identified Issues Not addressed N/A N/A
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-6

8 Dispositioning and Managing

Discrepancies

Not addressed N/A N/A

8.1 Purpose and Scope Not addressed N/A N/A

8.2 Generic Letter 91-18 Not addressed N/A N/A

8.3 Determining Operability

and Reportabilty

Not addressed N/A N/A

8.4 Methods Available to Address Degraded

or Nonconforming Conditions

Partial endorsement RG 1.196,

RP 2.7.3

Appendices C and D are not endorsed.

Appendix F exceptions related to:

1. Training and qualification of control room

operators for SCBA

2. Availability of adequate methods to refill

SCBA

3. Two typographical errors as described in the

text of this guide.

8.4.1 Compensatory Measures Partial endorsement RG 1.196,

RP 2.7.3

N/A

8.4.2 Dose Analysis Revision Option Partial endorsement RG 1.196,

RP 2.7.3

N/A

8.4.3 Repairing or Modifying the Plant Partial endorsement RG 1.196,

RP 2.7.3

N/A

8.4.4 Technical Specification Changes Partial endorsement RG 1.196,

RP 2.7.3

N/A
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-7

9 Long-term CRH Integrity Program Not addressed N/A N/A

9.1 Purpose and Scope Not addressed N/A N/A

9.2 CRH Integrity Program Not addressed N/A N/A

9.3 Periodic Evaluations Not addressed N/A N/A

9.3.1 System Material Condition Full endorsement RG 1.196,

RP 2.7.1

N/A

9.3.2 Post-maintenance Activities Not addressed N/A N/A

9.3.3 In-leakage Assessments Not addressed N/A N/A

9.3.3.1 Assessment Scope Not addressed N/A N/A

9.3.3.2 Assessment Frequency Not addressed N/A N/A

9.3.3.3 Determine Need to Test Not addressed N/A N/A

9.3.4 Toxic Gas Evaluation Not addressed N/A N/A

9.4 Configuration Control Partial Endorsement RG 1.196,

RP 2.7.2

The NRC staff references the configuration

controls in Section 9.4. These include CRE

boundary and breach control, procedure control,

toxic gas control, design change, and safety

analysis controls.

9.4.1 CRE Boundary / Breach Control Partial endorsement RG 1.196,

RP 2.7.2

The NRC staff does not endorse

Appendix K.

9.4.2 Procedure Control Full endorsement RG 1.196,

RP 2.7.2

N/A

9.4.3 Toxic Chemical Control Full endorsement RG 1.196,

RP 2.7.2

N/A

9.4.4 Design Change Control Full endorsement RG 1.196,

RP 2.7.2

N/A
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-8

9.4.5 Safety Analyses Control Partial endorsement RG 1.196,

RP 2.7.2

Rather than endorse Appendices C, D, and G

referenced in Section 9.4.5, Regulatory Guides

1.183, 1.194, 1.195, and 1.78 should be used to

provide analysis assumptions used in safety

analyses.

9.5 Training Partial endorsement RG 1.196,

RP 2.7.2

The NRC staff endorses training

using only the sections of NEI

99-03 that the staff has

endorsed.

9.6 Testing Not addressed N/A N/A

10 References Not addressed N/A N/A

Appendix A Licensing Basis History Not addressed N/A N/A

Appendix B Control Room Habitability

Regulatory Information

Not addressed N/A N/A

Appendix C CRH Dose Analysis: 

Regulatory Enhancements

Not endorsed RG 1.196,

RP 2.7.3 

Regulatory Guide 1.195 should be used.

Appendix D Atmospheric Dispersion Not endorsed RG 1.196,

RP 2.7.3

Regulatory Guide 1.194 should be used.

Appendix E Smoke Infiltration Impact

on Safe Shutdown

Partial endorsement RG 1.196, RP 2.6 The NRC staff endorses Appendix E as an

acceptable method for performing this qualitative

assessment with exceptions stated in Regulatory

Position 2.6. However, the reference to Section 6

is not endorsed. Remove the words “as described

in Section 6” in the first sentence.
NEI 99-03 

Section

NEI 99-03 Section Title Endorsement

Status

Regulatory Guide

Section

Exceptions/Remarks

Appendix A to Rev. 1 of RG 1.196, Page A-9

Appendix F Compensatory Measures

Allowable on an Interim Basis

Partial endorsement RG 1.196, RP 2.7.3 Appendix F exceptions relate to:

1. Training and qualification of control room

operators for SCBA

2. The impact of a loss of offsite power or

airborne contamination at the refill

compressor stations.

3. Two typographical errors as described in the

text of this guide.

Appendix G Toxic Gas Assessments Not Addressed N/A Regulatory Guide 1.78 should be used.

Appendix H System Assessment Partial endorsement RG 1.196, RP 2.7.1 The NRC staff endorses Table H-1 as guidance for

developing a maintenance program.

Appendix I Testing Program Not Addressed N/A N/A

Appendix J Control Room Envelope

Sealing Program 

Not addressed N/A N/A

Appendix K Control Room Envelope

Boundary Control Program

Not endorsed RG 1.196, RP 2.7.2 The staff does not endorse Appendix K. Instead

of endorsing the method of equating a breach size

to an inleakage flow rate, the staff endorses the

method of breach control contained in the STSs

(NUREG-1431, NUREG 1430, NUREG-1432,

NUREG-1433, and NUREG-1434) (Refs. 8–12),

which allows the control room boundary to be

opened intermittently under administrative

controls.

Appendix L Glossary of Terms Not addressed N/A N/A
Appendix B to Rev. 1 of RG 1.196, Page B-1

APPENDIX B

ACRONYMS

ASHRAE American Society for Heating, Refrigeration and Air-Conditioning Engineers

ASME American Society of Mechanical Engineers

B&W Babcock and Wilcox

BWR Boiling Water Reactor

CE Combustion Engineering

CRE Control Room Envelope

CRH Control Room Habitability

CRHS Control Room Habitability System

ASTM American Society for Testing and Materials

ESF Engineered Safety Feature

FSAR Final Safety Analysis Report

GE General Electric

GDC General Design Criteria

LOCA Loss-of-Coolant Accident

LOOP Loss of Offsite Power

NEI Nuclear Energy Institute

NRC Nuclear Regulatory Commission

OL Operating License

OMB Office of Management and Budget

SCBA Self-Contained Breathing Apparatus

SRP Standard Review Plan

SSC Structures, Systems, and Components

STS Standard Technical Specification

TMI Three Mile Island

TSTF Technical Specification Task Force

TSC Technical Support Center

UFSAR Updated Final Safety Analysis Report