NRC Generic Letter 2003-01, Control Room Habitability
| ML031620248 | |
| Person / Time | |
|---|---|
| Issue date: | 06/12/2003 |
| From: | Matthews D B NRC/NRR/DRIP/RORP |
| To: | |
| Blumberg M NRR/DSSA 415-1083 | |
| References | |
| OMB 3150-0011, TAC M2788 GL-03-001 | |
| Download: ML031620248 (10) | |
OMB Control No.: 3150-0011 June 12, 2003NRC GENERIC LETTER 2003-01:CONTROL ROOM HABITABILITY
Addressees
All holders of operating licenses for pressurized-water reactors (PWRs) and boiling-waterreactors (BWRs), except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel and more than 1 year has elapsed since fuel was irradiated in the reactor vessel.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to:
(1)alert addressees to findings at U.S. power reactor facilities suggesting that the controlroom licensing and design bases, and applicable regulatory requirements (see section below) may not be met, and that existing technical specification surveillance requirements (SRs) may not be adequate, (2)emphasize the importance of reliable, comprehensive surveillance testing to verifycontrol room habitability, (3)request addressees to submit information that demonstrates that the control room ateach of their respective facilities complies with the current licensing and design bases, and applicable regulatory requirements, and that suitable design, maintenance and testing control measures are in place for maintaining this compliance, and(4)collect the requested information to determine if additional regulatory action is require BackgroundThe control room is the plant area, defined in the facility licensing basis, from which actions aretaken to operate the plant safely under normal conditions and to maintain the reactor in a safe condition during accident situation For most facilities, the habitability criteria of General Design Criterion 19 (GDC 19) in 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," apply to this are The control room envelope (CRE) is the plant area, defined in the facility licensing basis, that encompasses the control room and may encompass other plant area The structures that make up the CRE are designed to limit the inleakage of radioactive and hazardous materials from areas external to the CR Control room habitabilityML031620248 GL 2003-01 systems (CRHSs) typically provide the functions of shielding, isolation, pressurization, heating,ventilation, air conditioning and filtration, monitoring, and the sustenance and sanitation necessary to ensure that the control room operators can remain in the control room and take actions to operate the plant under normal and accident condition The personnel protection features incorporated into the design of a particular plant's CRHSs depend on the nature and scope of the plant-specific challenges to maintaining control room habitabilit In the majority of the CRHS designs, isolation of the normal supply and exhaust flow paths and pressurization of the CRE relative to adjacent areas are fundamental to ensuring a habitable control room.During the design of a nuclear power plant, licensees perform analyses to demonstrate that theCRHSs, as designed, provide a habitable environment during postulated design basis event These design analyses model the transport of potential contaminants into the CRE and their remova The amount of inleakage of assumed contaminants is important to these analyse Unaccounted-for contaminants entering the CRE may impact the ability of the operators to perform plant control function If contaminants impair the response of the operators to an accident, there could be increased consequences to the public health and safet There are two typical CRE design These designs are referred to as positive-pressure andneutral-pressure CRE Both designs focus on limiting the amount of contaminants entering the CR For radiological challenges, the positive-pressure CRE intentionally pressurizes the CRE with air from outside the CR The pressurization air is treated by a high-efficiency particulate air filter and iodine adsorption media to remove contaminant The neutral-pressure CRE does not intentionally pressurize the CRE, but limits inleakage of contaminants by isolating controlled flow paths into the CR Most plants with a positive-pressure CRE have a technical specification SR to verify that those ventilation systems serving the CRE can maintain the CRE at a positive differential pressure relative to adjacent area These surveillance tests (typically referred to as a P surveillance) are generally implemented through a technical specificationSR for the CRHS Plants with a neutral-pressure CRE design typically do not have a CRE integrity testing progra (The term "neutral-pressure" means only that the CRE is not intentionally pressure The actual pressure of the CRE may be positive, neutral, or negative relative to adjacent areas.)In addition to the P surveillance described above, licensees have performed CRE integritytesting at approximately 30 percent of the power reactor facilities using the standard test method described in American Society for Testing and Materials (ASTM) consensus standard E741, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution." Unlike the P surveillance, the ASTM E741 test determines the total CREinleakage from all source It is well suited for assessing the integrity of positive-pressure or neutral-pressure CRE The test basically involves homogeneously dispersing a nontoxic tracer gas throughout the CRE and measuring the dilution of the tracer gas caused by inleakag The results of the ASTM E741 tests indicate that the P surveillance is not a reliable method fordemonstrating CRE integrit For all but one facility tested using the ASTM E741 standard, the measured inleakage was greater than the inleakage assumed in the design basis analyse In some cases, even though the licensees had routinely demonstrated a positive P relative toadjacent areas at their facilities, the measured inleakage was several orders of magnitude greater than the value previously assume Affected facilities were subsequently able to GL 2003-01 achieve compliance with the control room radiation protection regulatory requirements bysealing, adding new ductwork, changing their CRE, or reanalyzing their control room habitability.Use of the P surveillance as an indicator of CRE integrity has two inherent deficiencie First,it does not measure CRE inleakag The P surveillance infers that no contamination canenter the CRE if the CRE is at a higher pressure than adjacent area Second, the Psurveillance cannot determine whether there may be unrecognized sources of pressurization of the CRE that could introduce contaminants into the CRE under accident condition Two possible unrecognized contamination pathways are the CRHS fan suction ductwork that is located outside the CRE, and the pressurized ducts that traverse the lower pressure CRE en route to another plant area.The ASTM E741 testing has helped to identify a spectrum of CRHS deficiencies that affect(1) system design, construction, and quality, (2) system boundary construction and integrity, and (3) technical specification SR Licensees have determined that the performance of the CRHSs can be affected by (1) the gradual degradation in associated equipment such as seals, floor drain traps, fans, ductwork, and other components, (2) the drift of throttled dampers, (3) maintenance on the CRHSs, and (4) inadvertent misalignments of the CRHS Since inleakage is influenced by pressure differentials between the CRE and adjacent areas, changes in ambient pressure in these adjacent areas can affect the CRE inleakag These changes can be the result of a modification, the degradation of the ventilation systems serving these areas, or inadequate preventive and corrective maintenance programs.Licensees and NRC staff have identified other deficiencies in CRHS design, operation, andperformance from the review of license amendments, licensee event reports, and records and reports prepared pursuant to 10 CFR 50.5 These deficiencies showed that the licensees'
CRHSs did not meet their design base Some of these deficiencies are discussed in Regulatory Issue Summary 2001-19, "Deficiencies in the Documentation of Design Basis Radiological Analyses Submitted in Conjunction with License Amendment Requests." For example, some licensees credited the operation of CRHSs based upon actuation of high- radiation signals from instrumentatio Further investigation revealed that for some licensees the system would not be actuated due to incorrect setpoints or placement of the instrumentatio Other CRHS designs appear not to have considered unfiltered or once-filtered inleakage through idle CRHS ventilation train Without adequate consideration of such design issues, design basis radiation exposure limits may be exceeded.Previous to the ASTM E741 testing, a group of licensees had trouble meeting the control roomcriteria in Three Mile Island (TMI) Action Item III.D.3.4, "Control Room HabitabilityRequirements," that the NRC ordered most licensees to implement after the accident at TM At that time, radiological source term research suggested that the distribution of the chemical forms of iodine released during an accident could be different from the distribution in thetraditional source term defined in U.S. Atomic Energy Commission Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites."
Because of the possible differences, the staff allowed licensees to postpone changing their control rooms until the ongoing source term research was completed or until a generic letter on control room habitability was issue The staff believed that postponing changes was reasonable since the source term research or improved methods of analyses might prove that GL 2003-01 the changes were unnecessar Many of these licensees that postponed changes incorporated compensatory actions into their operating procedures to assure that the control room operators would be protected in case of an acciden Since then, some licensees have found that they could not meet the thyroid dose limits for habitability without using compensatory action The NRC also allowed these facilities to use compensatory actions until completion of the source term researc In August 2000, the NRC staff incorporated the results of the source term research into Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," which is now available for use by licensees.Although many CRE integrity testing programs focus on radiological concerns, radiation is onlyone potential design basis challenge to the protection of the operator The inleakage of other contaminants may have a greater impact on control room habitabilit An inleakage rate that is tolerable for one contaminant may not be tolerable for anothe The control room licensing basis describes the hazardous chemical releases considered in the CRE design, the designfeatures, and the administrative controls implemented to mitigate the consequences of these releases to the control room operator Smoke and other byproducts of fire within the CRE or in adjacent areas are among the contaminants that can have an adverse impact on control room habitability.DiscussionInformation obtained by the NRC indicates that some licensees have not maintained adequateconfiguration control over their CREs and have not corrected identified design and performance deficiencie The primary design function of CRHSs is to provide a safe environment in which the operator can control the nuclear reactor and auxiliary systems during normal operations and can safely shut down these systems during abnormal situations to protect the health and safety of the publi It is important for the operators to be confident of their safety in the control room to minimize errors of omission and commissio Errors of omission and commission are morelikely if CRHSs do not properly perform as intended in response to challenges from off-normal or accident situation The control room must be safe so that operators can remain in the control room to monitor plant performance and take appropriate mitigative action This is an underlying assumption in both the design basis and severe accident risk analyse It is, therefore, imperative to the health and safety of the public that operators are safe in the control room at all times.The scope and magnitude of the problems that NRC staff and certain licensees have identifiedraise concerns about whether similar design, configuration, and operability problems exist at other reactor facilitie The NRC staff is particularly concerned about whether licensees'
programs to maintain configuration control of CRHSs are sufficient to demonstrate that the physical and functional characteristics of CRHSs are consistent with and are being maintained according to their design base It is emphasized that the NRC's position has been, and continues to be, that it is the responsibility of individual licensees to know the licensing basis forthe CRHS Licensees should also have appropriate documentation of the design basis and procedures in place, in accordance with NRC regulations, for performing necessary assessments of plant or procedure changes that may affect the performance of the CRHS GL 2003-01 The technical specifications for about 75 percent of the control rooms (mostly positive-pressureCREs) have an SR to measure the P from the CRE to adjacent area The bases of theImproved Standard Technical Specifications state that this SR demonstrates control room integrity with respect to unfiltered inleakag The ASTM E741 integrated testing proves that it does no Because 10 CFR 50.36 requires technical specifications to be derived from thesafety analyses, the staff believes that the existing deficiency should be correcte This correction is consistent with NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient To Assure Plant Safety," which describes the staff's expectation that licensees correct technical specifications that are found to "contain non- conservative values or specify incorrect actions." Because of the importance of ensuring habitable control rooms under all normal and off-normalplant conditions, the addressees are requested to provide certain information that will enable the NRC staff to verify whether addressees can demonstrate and maintain the current design bases for the CRHSs at their facilitie
Addressees
are encouraged, but not required, to work closely with industry groups on the coordination of their response Coordinating the responsespromotes efficiency since it leads to a uniform approach to demonstrating compliance with the design bases of their CREs.NEI 99-03, "Control Room Habitability Assessment Guidance," provides industry genericguidance on control room habitabilit The NRC staff reviewed NEI 99-03, but rather than fully endorse NEI 99-03, the NRC staff developed its own guidanc Regulatory Guide 1.196 (formerly DG-1114), "Control Room Habitability at Light-Water Nuclear Power Reactors,"
endorses NEI 99-03 to the extent possible and provides additional guidanc Licensees are notrequired to comply with Regulatory Guide 1.196, but may find it useful in responding to this generic lette Licensees that are unable to confirm item 1 under the Requested Information section may use Regulatory Guide 1.196 to develop and implement corrective actions.Requested Information
Addressees
are requested to provide the following information within 180 days of the date ofthis generic letter.1.Provide confirmation that your facility's control room meets the applicable habitabilityregulatory requirements (e.g., GDC 1, 3, 4, 5, and 19) and that the CRHSs are designed, constructed, configured, operated, and maintained in accordance with the facility's design and licensing base Emphasis should be placed on confirming:(a)That the most limiting unfiltered inleakage into your CRE (and the filteredinleakage if applicable) is no more than the value assumed in your design basis radiological analyses for control room habitabilit Describe how and when you performed the analyses, tests, and measurements for this confirmation.(b)That the most limiting unfiltered inleakage into your CRE is incorporated intoyour hazardous chemical assessment This inleakage may differ from the value assumed in your design basis radiological analyse Also, confirm that the reactor control capability is maintained from either the control room or the alternate shutdown panel in the event of smok GL 2003-01 (c)That your technical specifications verify the integrity of the CRE, and theassumed inleakage rates of potentially contaminated ai If you currently have a P surveillance requirement to demonstrate CRE integrity, provide the basis foryour conclusion that it remains adequate to demonstrate CRE integrity in light of the ASTM E741 testing result If you conclude that your P surveillancerequirement is no longer adequate, provide a schedule for: 1) revising the surveillance requirement in your technical specification to reference an acceptable surveillance methodology (e.g., ASTM E741), and 2) making any necessary modifications to your CRE so that compliance with your new surveillance requirement can be demonstrated. If your facility does not currently have a technical specification surveillancerequirement for your CRE integrity, explain how and at what frequency you confirm your CRE integrity and why this is adequate to demonstrate CRE integrity.2.If you currently use compensatory measures to demonstrate control room habitability,describe the compensatory measures at your facility and the corrective actions needed to retire these compensatory measures.3.If you believe that your facility is not required to meet either the GDC, the draft GDC, orthe "Principal Design Criteria" regarding control room habitability, in addition to responding to 1 and 2 above, provide documentation (e.g., Preliminary Safety Analysis Report, Final Safety Analysis Report sections, or correspondence) of the basis for this conclusion and identify your actual requirements.Requested ResponseIf an addressee cannot provide the information or cannot meet the requested completion date,the addressee should submit a written response indicating this within 60 days of the date of this generic lette The response should address any alternative course of action the addressee proposes to take, including the basis for the acceptability of the proposed alternative course of action and the schedule for completing the alternative course of actio The written response should be addressed to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-000 A copy of the response should be sent to the appropriate regional administrator.NRC staff will review the responses to this generic letter and, if concerns are identified, willnotify affected addressee The staff may conduct inspections to determine licensees'
effectiveness in addressing this generic letter.Applicable Regulatory RequirementsSeveral provisions of the NRC regulations and plant operating licenses (technicalspecifications) pertain to the issue of control room habitabilit The general design criteria for nuclear power plants (10 CFR Part 50, Appendix A), or, as appropriate, the quality assurance requirements in the licensing basis for a reactor facility (stated in 10 CFR Part 50, Appendix B, GL 2003-01 "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants"), and thetechnical specifications, are the bases for the NRC staff's assessment of control room habitabilit Appendix A to 10 CFR Part 50 and the plant safety analyses require or commit licensees todesign and test safety-related structures, systems, and components (SSCs) to provide adequate assurance that they can perform their safety function The NRC staff applies these criteria to plants with construction permits issued on or after May 21, 1971, and to those plants whose licensees have committed to the The applicable GDC are GDC 1, 3, 4, 5, and 1 GDC 1 requires quality standards commensurate with the importance of the safety functions performe GDC 3 requires SSCs to be designed and located to minimize the effects of fire GDC 4 requires SSCs to be designed to accommodate the effects of accidents. GDC 5 requires that an accident in one unit will not significantly impair orderly shutdown and cooldown of the remaining unit.GDC 19 specifies that a control room be provided from which actions can be taken to operatethe nuclear reactor safely under normal conditions and maintain the reactor in a safe condition under accident conditions, including a loss-of-coolant acciden There must be adequate radiation protection to permit personnel to access and occupy the control room under accident conditions without receiving radiation exposures in excess of specified values.Before the issuance of the GDC, proposed GDC (sometimes called "principal design criteria")were published in the Federal Register for commen As they evolved, several of the proposedGDC addressed control room habitabilit A facility may have been licensed before the issuance of the GDC, but the licensee may have committed to the proposed GDC as they existed at the time of licensing.Following the accident at TMI, TMI Action Plan Item III.D.3.4, "Control Room HabitabilityRequirements," as clarified in NUREG-0737, "Clarification of TMI Action Plan Requirements,"
required all licensees to assure that control room operators would be adequately protected against the effects of accidental releases of toxic and radioactive gases and that the nuclear power plant could be safely operated or shut down under design basis accident condition When licensees proposed modifications, the NRC issued orders confirming the licensees'commitment As a result, most plants licensed before the GDC were formally adopted were then subsequently required to meet the TMI Action Plan III.D.3.4 requirements.Appendix B to 10 CFR Part 50 establishes quality assurance requirements for the design,construction, and operation of those SSCs that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the publi Appendix B, Criterion III, "Design Control," requires that design control measures be providedfor verifying or checking the adequacy of desig A suitable testing program is identified as one method of accomplishing this verificatio Appendix B, Criterion XVI, "Corrective Action,"
requires measures to be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, defective material and equipment, and nonconformances are promptly identified and corrected.The regulations in 10 CFR 50.36, "Technical Specifications," require plant technicalspecifications to be derived from the safety analyse GL 2003-01