Regulatory Guide 1.196

Control Room Habitability at Light-Water Nuclear Power Reactors.

ML063560144
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Issue date:	01/23/2007
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RG-1.196, Rev 1
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The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staffconsiders acceptable for use in implementing specific parts of the agency's regulations, techniques that the staff uses in evaluating specific problemsor postulated accidents, and data that the staff need in reviewing applications for permits and license Regulatory guides are not substitutesfor regulations, and compliance with them is not require Methods and solutions that differ from those set forth in regulatory guides will be deemedacceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission.The NRC staff encourages and welcomes comments and suggestions in connection with improvements to published regulatory guides, as well as itemsfor inclusion in regulatory guides that are currently being develope The NRC staff will revise existing guides, as appropriate, to accommodatecomments and to reflect new information or experienc Written comments may be submitted to the Rules and Directives Branch,Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.Regulatory guides are issued in 10 broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities;4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review;and 10, General.Requests for single copies of draft or active regulatory guides (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission,Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to Distribution@nrc.gov. Electronic copies of this guide and other recently issued guides are available through the NRC's public Web site under the Regulatory Guides documentcollection of the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRC's Agencywide DocumentsAccess and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML063560144.U.S. NUCLEAR REGULATORY COMMISSIONJanuary 2007Revision 1REGULATORY GUIDEOFFICE OF NUCLEAR REGULATORY RESEARCHREGULATORY GUIDE 1.196CONTROL ROOM HABITABILITYAT LIGHT-WATER NUCLEAR POWER REACTORS INTRODUCTIONThis 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 CFRPart 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 locatedto minimize the effects of fires and explosions.*GDC 4, "Environmental and Dynamic Effects Design Bases," requires SSCs important to safetyto 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 safetynot 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 betaken to operate the nuclear reactor safely under normal conditions and to maintain the reactor in a safe condition under accident conditions, including a LOC Adequate radiation protectionis 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 determinedthat the information presented in Appendix B to that guide did not accurately represent a viable technical specification for CRH at light-water nuclear power reactor In particular, it referred to failure of a particular surveillance as a plant state, rather than having the results of the surveillance factorinto the operability determinatio In addition, it did not provide for a definite time to restore functionalityto the control room envelope, whereas all improved standard technical specifications (iSTS) contain such provision 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 grantedthe technical specification changes represented by Appendix B to the original version of this guid 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 wa Moreover, the owners' group Technical Specification Task Forcehas provided ample opportunity for public comment regarding this revisio Therefore, the staff views the removal of Appendix B as a neutral action, for which further public comments are unnecessar 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 requirementsof 10 CFR Part 50, and that the Office of Management and Budget (OMB) approved under OMB control number 3150-001 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 numbe An example of a changed performance parameter that may require re-analysis is an increase in control room envelope(CRE) inleakage beyond that assumed in previous CRH assessments.2As used in this guide, the licensing basis is the documentation that describes how the plant meets applicable regulations. Design bases are defined in 10 CFR 5 Regulatory Guide 1.186, "Guidance and Examples for Identifying 10 CFR 50.2 Design Bases" (Ref. 2), provides additional guidanc The design bases are a subset of the licensing base Thus, this guide uses "licensing bases" to refer to both.Rev. 1 of RG 1.196, Page 3 DISCUSSIONA licensee may use this guide for assessing CRH following changes to control room habitabilitysystems (CRHSs) or the sources that would lead to consequences to the operato 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,1 alignment, or function of the CRHSs*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 shipmentsThe primary design function of CRHSs is to provide a safe environment in which the operatorcan keep the nuclear reactor and auxiliary systems under control during normal operations and can safelyshut down those systems during abnormal situations to protect the health and safety of the public and plant worker 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 safet It is important for the operators to be confident of their safety in the control room to minimize errors of omission and commissio 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 RoomHabitability 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 CR Only the sections of NEI 99-03 that are specifically stated in the regulatory positions should be considered to be endorsed by the NR 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 99-0 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 basis2 and associated documents will define the termsfor a particular facility.Control Room: The plant area, defined in the facility's licensing basis, in which actions can betaken to operate the plant safely under normal conditions and to maintain the reactor in a safe condition during accident situation 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 acciden Rev. 1 of RG 1.196, Page 4Control Room Envelope (CRE): The plant area, defined in the facility's licensing basis, thatin 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 roo This area encompasses the control room, and may encompass other non-critical areasto 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 situation The CRHSs include the CR Rev. 1 of RG 1.196, Page 5 REGULATORY POSITION1.An Overview of the Process of Demonstrating and Maintaining CRHIn demonstrating that a facility's control room conforms to the GDCs, the following CRH aspectsare typically assessed:*radiological doses*protection from the effects of hazardous materials

• control of the reactor from either the control room or the alternate shutdown panelThe 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 areasidentified in action (a) are consistent with the licensing bases(c)determination of the performance characteristics for operating modes associated withradiological 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 inthe 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 CRHSs2.Demonstrating and Maintaining CRHRegulatory Positions 2.1 through 2.7 provide guidance on the process of demonstratingand maintaining CRH.2.1Identification of the Licensing Bases for the CRHSs2.1.1Identification of the Control Room and the CREConfirmation of a facility's ability to meet CRH requirements begins with the identificationof the control room and the CR A description of the control room and CRE may be contained in a number of plant document 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 stage Rev. 1 of RG 1.196, Page 62.1.2Determination of the Licensing BasesIn demonstrating the habitability of a facility's control room, it is essential that the licenseeknow the facility's licensing bases for its CRHS The sources of the licensing bases of the CRHSs should be identifie 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 CRHS 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 chang The staff may have reviewed and approvedthe licensing bases of facilities licensed before the issuance of this guid The original licensing bases may have been submitted as part of the construction permit applicatio Licensees may have modified themin response to NRC question 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. Amendments to the OL may have resulted in changes to the licensing bases of the CRHS 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 commen These proposed GDCs addressed CR Although facilitiesmay 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 licensin 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 descriptionof the licensing bases will be the documents filed in support of the licensing application (under 10 CFR Parts 50 and 52).2.2Determination of Whether the CRHSs Are Consistent with the Licensing Bases2.2.1Comparison of System Design, Configuration, and Operation with the Licensing BasesLicensees should compare the design, configuration, and operation of their CRHSs and the systemsthat are in adjacent areas and could interact with the CRE to their licensing bases to ensure consistenc The review of the configuration of the CRHSs should include the construction and the alignment of the systems and structures that make up the CRHS For new reactors and existing CRHSs undergoingredesign, this comparison should be made upon completion of constructio 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 clarificatio 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 base Licensees should employ methods similar to those provided in Section 5 of NEI 99-03 when they perform these comparisons for other CRHS Rev. 1 of RG 1.196, Page 72.2.2Interactions Between the CRHSs and Adjacent AreasThe 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 operat Potential interactions between the CRHSs and adjacent areas that may increase the transfer of contaminants into the CRE should be identifie 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 thattraverses the CRE or areas adjacent to the CRE.2.3Determination of Performance Characteristics2.3.1Performance of CRHSsThe licensee should determine the performance characteristics of the CRE, its ventilation systems, andsystems that serve or traverse areas within or located adjacent to the CR 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 pressure Performance characteristics are neededto achieve the following objectives:*Establish the operating parameters for incorporation into the licensing bases (for new reactorsor 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 severalparameters important to maintaining CR These parameters may include system flow rates, carbon filter efficiencies, actuation signals, and CRE integrity test 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 Unitsof 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 licensedfacilities, all but one facility measured greater inleakage than that assumed in the design analyse In some cases, the measured inleakage exceeded the amount assumed in the design analyses by several orders of magnitud 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 staffconsiders acceptable to determine CRE integrit As discussed in Regulatory Position 2.2.2, systems outside the CRE may impact CRE integrit Testing may also be needed to understand the influence of these systems on CR Consistent with Regulatory Position 2.2.1, licensees should ensurethat 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 1.19 Rev. 1 of RG 1.196, Page 82.3.2Identification of the Limiting ConditionThe limiting condition for CRH is the configuration that results in the maximum consequencesto the control room operator Sometimes the limiting condition will arise from the configuration that produces the greatest inleakage and sometimes it will no 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 operator As an example, CRE inleakage may be greatest for a radiological accident that does not have a LOO 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 bepresumed that the LOCA is the limiting accident because it has the largest initial source of activit 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 operator 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 radiologicalaccidents, 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 varywith the accident.*The location of the release points for the various accidents relative to the control room intakesmay result in less favorable atmospheric dispersion and higher magnitude intake concentrations.Licensees should factor all the potential differences in accidents and the CRHS's performancein order to determine the limiting condition.A few plants are within the exposure range for an accidental release from a nearby nuclear plantor have separate control rooms for multiple units on the same sit An accident in an adjacent unit should not prevent the safe shutdown of an operating uni Regulatory Position 2.6, "Reactor Control,"

describes criteria used for determining a safe shutdown of the reacto 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 accidentsshould be used to determine the limiting condition.2.4Radiological ConsequencesLicensees should calculate control room operator doses for the methodology and accidentsidentified 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 value 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 92.5Hazardous ChemicalsLicensees should evaluate the impact of hazardous chemicals on control room operatorsusing 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 chemicalsin the vicinity of their plant site 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 usesof lan 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 applicabl 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 expectedperformance value When the envelope and associated ventilation systems are operational, the calculationshould be based on an inleakage value determined according to Regulatory Guide 1.197 (Ref. 4).2.6Reactor ControlThis Regulatory Position provides guidance for assessing whether a radiological, hazardouschemical, or smoke challenge could result in the inability of the control room operators to control the reactorfrom the control room, or in the event of a smoke challenge, from the control room or the alternate shutdown pane 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 thatan accident arising from a radiological event, hazardous chemicals, or a smoke challenge will not preventthe control room operators from controlling the reacto 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 methodolog The specific acceptancecriterion 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 envelopeor if a fire develops in the control roo In these situations, smoke could challenge the operator's abilityto shut down the reactor from within the control room or remotel No regulatory limit exists on the amountof smoke allowed in the control roo Therefore, the plant's ability to manage smoke infiltration is assessed qualitativel 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 even The staff endorses Appendix E of NEI 99-03 (Ref. 1) as an acceptable method for performing this qualitative assessment with the following exception The second sentence of Section 1should 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 delete The last bullet in Section 3 should be the last bullet in Section Rev. 1 of RG 1.196, Page 102.7Maintaining and Monitoring CRHSsCRH 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 trainin 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 basi The following methods of maintaining and monitoring CRHSs should be used.2.7.1Periodic Evaluations and MaintenancePeriodic 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 performe CRHS programs should assess the systemand 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 CR Routine performancemeasurements are described in Regulatory Position 2. The complexity of testing following modifications should depend on the effect of the modification on CR Regulatory Guide 1.197 (Ref. 4)provides a testing method for verification of CRE integrit A frequency for CRE integrity testing is provided in Regulatory Guide 1.19 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 CRHS Table H-1 of NEI 99-03 (Ref. 1)should be used to identify systems and components to be included in a maintenance progra 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),

respectivel ASHRAE Guideline 1-1996 (Ref. 14) may be used to establish a maintenance program for systems that handle hazardous chemical and smoke challenges.2.7.2Configuration Control and TrainingConfiguration control and training are effective tools that can minimize the impact that changesto CRHSs can have on CR 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 exception 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 control Rev. 1 of RG 1.196, Page 11For entry and exit through doors, the administrative control of the opening is performedby the persons entering or exiting the are For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control roo This individual will have a method to rapidly close the opening when a need for control room isolation is indicate 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 99-0 Instead, the NRC staff endorses training using only the sections of NEI 99-03 that the staff has endorsed.2.7.3Degraded and Nonconforming ConditionsMethods available to address short term degraded or nonconforming conditions are providedin 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) tablet These methods are acceptable with the following exception 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 use 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 operatorsfor SCB 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 performe These scenarios should last about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and include a simulated watch turnover.(2)Section 2.6 of Appendix F addresses the availability of adequate methods to refill depletedSCBA cylinder 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 controlroom during a chemical releas 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 ThatInterferes With the Sealing Surfaces of the Respirato The required minimum staffing of controlroom 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 alternativesource term rulemaking and regulatory guidance were publishe 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 action Draft Regulatory Guide DG-1114 is available electronically under Accession #ML020790125 in the NRC'sAgencywide 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 20555-000 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 IMPLEMENTATIONThe purpose of this section is to provide information to applicants and licensees regardingthe NRC staff's plans for using this regulatory guid 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 establishedan 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 ANALYSISThe regulatory analysis for this regulatory guide is available in Draft Regulatory Guide DG-1114,"Control Room Habitability at Light-Water Nuclear Power Reactors."3 The NRC issued DG-1114in March 2002 to solicit public comment on the initial draft of Regulatory Guide 1.196.BACKFIT ANALYSISThe 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 positio 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 facilit 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 standar Copies are also available for inspection or copying for a fee from the NRC's Public Document Room (PDR), which islocated at 11555 Rockville Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-000 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.2Requests for single copies of draft or active regulatory guides (which may be reproduced) or for placement on anautomatic 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.go Copies are also available for inspection orcopying 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 20555-000 The PDR can also be reachedby telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email to PDR@nrc.gov.3Available through NRC's Public Web site, 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 13REFERENCES1.NEI 99-03, "Control Room Habitability Assessment Guidance," Revision 0,Nuclear Energy Institute, Washington, DC, June 2001.12.Regulatory Guide 1.186, "Guidance and Examples for Identifying 10 CFR 50.2 Design Bases,"U.S. Nuclear Regulatory Commission, Washington, DC, December 2000.23.Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtrationand Adsorption Units of Post Accident Engineered-Safety-Feature Atmosphere Cleanup Systemsin Light-Water-Cooled Nuclear Power Plants," Revision 3, U.S. Nuclear Regulatory Commission,Washington, DC, June 2001.24.Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear PowerReactors," U.S. Nuclear Regulatory Commission, Washington, DC, May 2003.25.Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequencesof Design-Basis Accidents at Light-Water Nuclear Power Reactors," U.S. Nuclear Regulatory Commission, Washington, DC, May 2003.26.Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors," U.S. Nuclear Regulatory Commission, Washington, DC, July 2000.27.Regulatory Guide 1.78, "Evaluating the Habitability of a Nuclear Power Plant Control RoomDuring a Postulated Hazardous Chemical Release," Revision 1, U.S. Nuclear Regulatory Commission, Washington, DC, December 2001.28.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 1410.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 AdsorptionUnits of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants,"

Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, June 2001.214.ASHRAE Guideline 1-1996, "The HVAC Commissioning Process," American Societyof Heating, Refrigerating and Air Conditioning Engineers, Atlanta, Georgia, June 1996.15.Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room RadiologicalHabitability Assessments at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, Washington, DC, May 200 Appendix A to Rev. 1 of RG 1.196, Page A-1APPENDIX AREGULATORY GUIDE ENDORSEMENTOF NEI 99-03, REV. 0, JUNE 2001, BY SECTIONNEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/Remarks1IntroductionNot addressedN/AN/A1.1

Purpose

And ScopeNot addressedN/AN/A 1.2HistoryNot addressedN/AN/A 1.3Document OrganizationNot addressedN/AN/A 2Regulatory Requirements and GuidanceNot addressedN/AN/A 2.1

Purpose

And ScopeNot addressedN/AN/A 2.2Regulatory Requirement: General Design Criterion 19Not addressedN/AN/A2.3Regulatory GuidanceNot addressedN/AN/A2.3.1Regulatory GuidesNot addressedN/AN/A 2.3.2NUREGsNot addressedN/AN/A 2.3.3Information NoticesNot addressedN/AN/A 2.4Generic IssuesNot addressedN/AN/A NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-23Industry Issues Associated withControl Room HabitabilityNot addressedN/AN/A3.1

Purpose

And ScopeNot addressedN/AN/A3.2Licensing Basis Different From As-BuiltPlantNot addressedN/AN/A3.3Analyses Different From As-built orAs-Operated PlantNot addressedN/AN/A3.4DBA Analyzed Not Most LimitingNot addressedN/AN/A3.4.1Adjacent Unit Accident (a Special Case)Not addressedN/AN/A 3.5Smoke InfiltrationNot addressedN/AN/A 3.6Toxic Gas EvaluationNot addressedN/AN/A 3.7Control Room Air In-LeakageGreater Than AssumedNot addressedN/AN/A3.7.1Radiological ConsiderationsNot addressedN/AN/A3.7.2Toxic Gas ConsiderationsNot addressedN/AN/A NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-34Determining CRH Licensing BasisNot addressedN/AN/A4.1

Purpose

And ScopeNot addressedN/AN/A 4.2Understanding the Conceptof Licensing BasisNot addressedN/AN/A4.2.1Design BasisNot addressedN/AN/A4.2.2Supporting Design InformationNot addressedN/AN/A 4.2.3Licensing BasisNot addressedN/AN/A 4.3Licensing Basis SourcesFull endorsementRegulatory Guide1.196, RegulatoryPosition (RP) 2.1.2N/A4.4Performing the Licensing Basis ReviewNot addressedN/AN/A4.5Assembling the CRH AnalysisNot addressedN/AN/A 4.6Documentation of the ExistingPlant CRH Licensing and Design BasisNot addressedN/AN/A NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-45Comparing Existing Plant Configurationand Operations with Licensing Bases for CRHFull endorsementRG 1.196, RP 2.2.1This section provides a method of comparing theplant's configuration and operation of ventilation systems with the licensing bases that is acceptable to the NRC staff with one clarificatio 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 ScopeFull endorsementRG 1.196, RP 2.2.1N/A 5.2Review the As-Built Control RoomEnvelope and Control Room Ventilation SystemsFull endorsementRG 1.196, RP 2.2.1N/A5.3Review the Normal and EmergencyOperating Procedures Affecting the Control Room Ventilation SystemsFull endorsementRG 1.196, RP 2.2.1N/A5.4Review the Testing ProceduresAffecting Control Room Ventilation Systems and the Associated EnvelopeFull endorsementRG 1.196, RP 2.2.1N/A5.5Review the Maintenance Practicesand Procedures for Effect on CRH RequirementsFull endorsementRG 1.196, RP 2.2.1N/A5.6Review the Plant ModificationProcedures for Consideration of the CRH RequirementsFull endorsementRG 1.196, RP 2.2.1 N/A5.7Review the CRH AnalysesFull endorsementRG 1.196, RP 2.2.1 N/A5.8Identified InconsistenciesFull endorsementRG 1.196, RP 2.2.1 N/A NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-56Assessing Industry Issue ApplicabilityNot addressedN/AN/A6.1

Purpose

and Scope Not addressedN/AN/A 6.2Limiting DBANot addressedN/AN/A 6.2.1Recommended Actions to EvaluateLimiting DBANot addressedN/AN/A6.2.2Adjacent Unit AccidentsNot addressedN/AN/A6.3Smoke InfiltrationNot endorsedN/AN/A 6.3.1Recommended Licensee Actionto Address Smoke InfiltrationNot endorsedN/AN/A6.4Toxic Gas EvaluationNot addressedN/AN/A6.4.1Recommended Licensee Actionto Address Toxic Gas EvaluationNot AddressedN/AN/A7Measuring Air In-Leakage(Baseline Test)Not addressedN/AN/A7.1

Purpose

and ScopeNot addressedN/AN/A7.2Preparation for TestingNot addressedN/AN/A 7.3Test PerformanceNot addressedN/AN/A 7.4Resolution of Identified IssuesNot addressedN/AN/A NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-68Dispositioning and ManagingDiscrepanciesNot addressedN/AN/A8.1

Purpose

and ScopeNot addressedN/AN/A8.2Generic Letter 91-18Not addressedN/AN/A 8.3Determining Operabilityand ReportabiltyNot addressedN/AN/A8.4Methods Available to Address Degradedor Nonconforming ConditionsPartial endorsementRG 1.196,RP 2.7.3Appendices C and D are not endorsed.Appendix F exceptions related to:

1.Training and qualification of control roomoperators for SCBA2.Availability of adequate methods to refillSCBA3.Two typographical errors as described in thetext of this guide.8.4.1Compensatory MeasuresPartial endorsementRG 1.196,RP 2.7.3N/A8.4.2Dose Analysis Revision OptionPartial endorsementRG 1.196,RP 2.7.3N/A8.4.3Repairing or Modifying the PlantPartial endorsementRG 1.196,RP 2.7.3N/A8.4.4Technical Specification ChangesPartial endorsementRG 1.196,RP 2.7.3N/A NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-79Long-term CRH Integrity ProgramNot addressedN/AN/A9.1

Purpose

and ScopeNot addressedN/AN/A 9.2CRH Integrity ProgramNot addressedN/AN/A 9.3Periodic EvaluationsNot addressedN/AN/A 9.3.1System Material ConditionFull endorsementRG 1.196,RP 2.7.1N/A9.3.2Post-maintenance ActivitiesNot addressedN/AN/A9.3.3In-leakage AssessmentsNot addressedN/AN/A 9.3.3.1Assessment ScopeNot addressedN/AN/A 9.3.3.2Assessment FrequencyNot addressedN/AN/A 9.3.3.3Determine Need to TestNot addressedN/AN/A 9.3.4Toxic Gas EvaluationNot addressedN/AN/A 9.4Configuration ControlPartial EndorsementRG 1.196,RP 2.7.2The NRC staff references the configurationcontrols in Section These include CRE boundary and breach control, procedure control, toxic gas control, design change, and safety analysis controls.9.4.1CRE Boundary / Breach ControlPartial endorsementRG 1.196,RP 2.7.2The NRC staff does not endorseAppendix K.9.4.2Procedure Control Full endorsementRG 1.196,RP 2.7.2N/A9.4.3Toxic Chemical ControlFull endorsementRG 1.196,RP 2.7.2N/A9.4.4Design Change ControlFull endorsementRG 1.196,RP 2.7.2N/A NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-89.4.5Safety Analyses ControlPartial endorsementRG 1.196,RP 2.7.2Rather than endorse Appendices C, D, and Greferenced 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.9.5TrainingPartial endorsementRG 1.196,RP 2.7.2The NRC staff endorses trainingusing only the sections of NEI 99-03 that the staff has endorsed.9.6TestingNot addressedN/AN/A 10ReferencesNot addressedN/AN/A Appendix ALicensing Basis HistoryNot addressedN/AN/A Appendix BControl Room HabitabilityRegulatory InformationNot addressedN/AN/AAppendix CCRH Dose Analysis: Regulatory EnhancementsNot endorsedRG 1.196,RP 2.7.3 Regulatory Guide 1.195 should be used.Appendix DAtmospheric DispersionNot endorsedRG 1.196,RP 2.7.3Regulatory Guide 1.194 should be used.Appendix ESmoke Infiltration Impacton Safe ShutdownPartial endorsementRG 1.196, RP 2.6The NRC staff endorses Appendix E as anacceptable method for performing this qualitative assessment with exceptions stated in Regulatory Position However, the reference to Section 6 is not endorse Remove the words "as described in Section 6" in the first sentenc NEI 99-03 SectionNEI 99-03 Section TitleEndorsementStatusRegulatory GuideSectionExceptions/RemarksAppendix A to Rev. 1 of RG 1.196, Page A-9Appendix FCompensatory MeasuresAllowable on an Interim BasisPartial endorsementRG 1.196, RP 2.7.3Appendix F exceptions relate to:1.Training and qualification of control roomoperators for SCBA2.The impact of a loss of offsite power orairborne contamination at the refill compressor stations.3.Two typographical errors as described in thetext of this guide.Appendix GToxic Gas AssessmentsNot AddressedN/ARegulatory Guide 1.78 should be used.

Appendix HSystem AssessmentPartial endorsement RG 1.196, RP 2.7.1The NRC staff endorses Table H-1 as guidance fordeveloping a maintenance program.Appendix ITesting ProgramNot AddressedN/AN/AAppendix JControl Room EnvelopeSealing Program Not addressedN/AN/AAppendix KControl Room EnvelopeBoundary Control ProgramNot endorsedRG 1.196, RP 2.7.2The staff does not endorse Appendix Insteadof 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 LGlossary of TermsNot addressedN/AN/A Appendix B to Rev. 1 of RG 1.196, Page B-1APPENDIX BACRONYMSASHRAEAmerican Society for Heating, Refrigeration and Air-Conditioning EngineersASMEAmerican Society of Mechanical Engineers B&WBabcock and Wilcox BWRBoiling Water Reactor CECombustion Engineering CREControl Room Envelope CRHControl Room Habitability CRHSControl Room Habitability System ASTMAmerican Society for Testing and Materials ESFEngineered Safety Feature FSARFinal Safety Analysis Report GEGeneral Electric GDCGeneral Design Criteria LOCALoss-of-Coolant Accident LOOPLoss of Offsite Power NEINuclear Energy Institute NRCNuclear Regulatory Commission OLOperating License OMBOffice of Management and Budget SCBASelf-Contained Breathing Apparatus SRPStandard Review Plan SSCStructures, Systems, and Components STSStandard Technical Specification TMIThree Mile Island TSTFTechnical Specification Task Force TSCTechnical Support Center UFSARUpdated Final Safety Analysis Report