[Federal Register Volume 67, Number 90 (Thursday, May 9, 2002)]
[Notices]
[Pages 31385-31389]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-11623]


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NUCLEAR REGULATORY COMMISSION


Proposed Generic Communication (TAC No. MB2788); Control Room 
Envelope Habitability

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of opportunity for public comment.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to issue 
a generic letter concerning control room envelope (CRE) habitability 
determination. The purpose of the proposed generic letter is to: (1) 
Alert addressees to findings at U.S. power reactor facilities that 
suggest that CRE licensing and design bases, and applicable regulatory 
requirements may not be met, and that a technical specification 
surveillance requirement may not be adequate to verify CRE operability, 
(2) emphasize the importance of reliable, comprehensive surveillance 
testing to verify CRE habitability, and (3) request addressees to 
submit information that demonstrates that the CRE at each of their 
respective facilities complies with the current licensing and design 
basis and applicable regulatory requirements, and that suitable design, 
maintenance and testing control measures are in place for maintaining 
this compliance. The NRC is seeking comment from interested parties 
regarding both the technical and regulatory aspects of the proposed 
generic letter, presented under the Supplementary Information heading.
    The NRC will consider comments received from interested parties in 
the final evaluation of the proposed generic letter. The NRC's final 
evaluation will include a review of its technical positions and, as 
appropriate, an analysis of the value/impact on licensees. Should this 
generic letter be issued by the NRC, it will become available for 
public inspection in the NRC Public Document Room.
    The NRC maintains an Agencywide Documents Access and Management 
System (ADAMS) which provides text and image files of NRC's public 
documents. These documents may be accessed through the NRC's Public 
Electronic Reading Room (PERR) on the Internet at  http://www.nrc.gov/reading-rm/adams.html >. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) reference staff by phone at 1-800-397-
4209 or 301-415-4737, by e-mail to [email protected]>, or by Fax at 301-415-
3548. The ADAMS Accession No. for the document containing the proposed 
generic letter is ML021090031.

DATES: Comment period expires August 7, 2002. Comments submitted after 
this date will be considered if it is practical to do so, but assurance 
of consideration cannot be given except for comments received on or 
before this date.

[[Page 31386]]


ADDRESSES: Submit written comments to Chief, Rules and Directives 
Branch, Division of Administrative Services, U.S. Nuclear Regulatory 
Commission, Mail Stop T6-D59, Washington, DC 20555-0001. Written 
comments may also be delivered to 11545 Rockville Pike, Rockville, 
Maryland, between 7:45 a.m. to 4:15 p.m., Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, located at One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland.

FOR FURTHER INFORMATION CONTACT: W. Mark Blumberg, (301) 415-1083

SUPPLEMENTARY INFORMATION:

NRC Generic Letter 2002-XX: Control Room Envelope Habitability

Addressees

    All holders of operating licenses for pressurized-water reactors 
(PWRs) and boiling-water reactors (BWRs), except those who have 
permanently ceased operations and have certified that fuel has been 
permanently removed from the reactor vessel and that it has been more 
than one year 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 
that suggest that the control room envelope (CRE) licensing and design 
bases, and applicable regulatory requirements (see section below) may 
not be met, and that a technical specification surveillance requirement 
may not be adequate to verify CRE operability,
    (2) Emphasize the importance of reliable, comprehensive 
surveillance testing to verify CRE habitability, and
    (3) Request addressees to submit information that demonstrates that 
the CRE at each of their respective facilities complies with the 
current licensing and design basis and applicable regulatory 
requirements, and that suitable design, maintenance and testing control 
measures are in place for maintaining this compliance.

Background

    The control room is the plant area where actions are taken to 
operate the plant safely under normal conditions, maintain the reactor 
in a safe condition, or mitigate the consequences of an accident. The 
CRE encompasses the control room and other rooms and areas that 
personnel must access to accomplish plant control functions in the 
event of an accident. The structures that make up the CRE are designed 
to limit the inleakage of contaminants such as radioactive materials, 
hazardous chemicals, and smoke from areas outside the CRE. CRE 
habitability systems (CREHSs) 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 
safely remain in the CRE. The personnel protection features 
incorporated into the design of a plant's CREHSs depend on the nature 
and scope of the plant-specific challenges to maintaining CRE 
habitability. Isolation of the CRE atmosphere from the atmosphere of 
adjacent areas is fundamental to ensuring a habitable environment.
    During the design of a nuclear power plant, licensees perform 
analyses to demonstrate that the CRE and the CREHSs, as designed, 
provide a habitable environment during postulated design basis events. 
These design analyses model the transport of potential contaminants 
into the CRE and their removal. The amount of inleakage of contaminants 
assumed is important to these analyses. Unaccounted-for contaminants 
entering the CRE may impact the ability of the operators to perform 
plant control functions. If contaminants impair the response of the 
operators to an accident, there could be increased consequences to the 
public health and safety.
    Typically, there are two CRE designs. These designs are referred to 
as positive-pressure and neutral-pressure CREs. Both designs focus on 
limiting the amount of contaminant entering the CRE. The positive-
pressure CRE intentionally pressurizes the CRE with air from outside 
the CRE. The pressurization air is treated by a high-efficiency 
particulate air filter and iodine absorption media to remove 
contaminants. The neutral-pressure CRE does not intentionally 
pressurize the CRE, but limits inleakage of contaminants by isolating 
controlled flow paths into the CRE. Plants with a positive-pressure CRE 
have generally implemented testing programs. These programs verify 
those ventilation systems serving the CRE can maintain the CRE at a 
positive differential pressure relative to adjacent areas. These 
testing programs are generally implemented through a technical 
specification surveillance requirement for the CREHSs. The tests are 
typically referred to as a P test. Plants with a neutral-
pressure CRE design typically do not have a CRE integrity testing 
program. (The term neutral-pressure means only that the CRE is not 
intentionally pressured. The actual pressure of the CRE may be 
positive, neutral, or negative relative to adjacent areas.)
    In addition to the P surveillance testing described above, 
approximately 30 percent of all addressees have performed CRE integrity 
testing 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 test, the E741 test 
measures the total CRE inleakage from all sources. It is well suited 
for assessing the integrity of positive-pressure or neutral-pressure 
CREs. The test basically involves homogeneously dispersing a nontoxic 
tracer gas throughout the CRE and measuring the dilution of the tracer 
gas caused by inleakage.
    The results of the E741 tests indicate that the P testing 
is not a reliable method for demonstrating CRE integrity. For all but 
one facility tested using the E741 standard, the measured inleakage was 
greater than the inleakage assumed in the design basis analyses. In 
some cases the measured inleakage was several orders of magnitude 
greater than the value previously assumed even though some licensees 
had routinely demonstrated a positive P relative to adjacent 
areas at their facilities. Affected facilities were subsequently able 
to achieve compliance with the CRE radiation protection regulatory 
requirements by sealing, adding new duct work, changing their CRE or by 
re-analysis of their CRE habitability.
    The P surveillance test has two deficiencies. First, it 
does not measure CRE inleakage. The P surveillance test infers 
that contamination cannot enter the CRE if the CRE is at a higher 
pressure than adjacent areas. Second, the P test cannot 
determine whether there may be unrecognized sources of pressurization 
of the CRE that could introduce contaminants into the CRE under 
accident conditions. Two possible contamination pathways are the CREHS 
fan suction duct work that is located outside the CRE, and the 
pressurized ducts that traverse the lower pressure CRE en route to 
another plant area.
    The E741 testing has helped to identify a spectrum of CREHS 
deficiencies that affect system design, construction, and quality; 
system boundary construction and integrity; and technical specification 
surveillance requirements. Licensees have determined that the 
performance of the

[[Page 31387]]

CRE and the CREHSs can be affected by (1) the gradual degradation in 
associated equipment such as seals, floor drain traps, fans, duct work, 
and other components; (2) the drift of throttled dampers; (3) 
maintenance on the CRE boundary or the CREHSs; and (4) inadvertent 
misalignments of the CREHSs. Since inleakage is influenced by pressure 
differentials between the CRE and adjacent areas, changes in ambient 
pressure in these adjacent areas can affect the CRE inleakage. These 
changes can be the result of a modification, the degradation of the 
ventilation systems serving these areas, or inadequate preventive and 
corrective maintenance programs.
    Licensees and NRC staff have identified other deficiencies in CREHS 
design, operation, and performance from the review of license 
amendments, Licensee Event Reports, and records and reports prepared 
pursuant to 10 CFR 50.59. These deficiencies showed that the licensees' 
CREs did not meet their design bases. Some of these deficiencies are 
discussed in Regulatory Issue Summary 2001-19, ``Deficiencies in the 
Documentation of Design Basis Radiological Analyses Submitted in 
Conjunction With License Amendment Requests.'' For example, some 
licensees credited the operation of CREHSs based upon actuation of 
high-radiation signals from instrumentation. Further investigation 
revealed that the system would not be actuated due to incorrect 
setpoints or placement of the instrumentation. Other CRE designs appear 
not to have considered unfiltered or once-filtered inleakage through 
idle CREHS ventilation trains. Without adequate consideration of such 
design deficiencies, design basis radiation exposure limits may be 
exceeded.
    Previous to the E741 testing, a group of licensees had trouble 
meeting the CRE criteria in Three Mile Island (TMI) Action Item 
III.D.3.4, ``Control Room Habitability Requirements,'' that the NRC 
ordered most licensees to implement after the accident at TMI. At that 
time, radiological source term research suggested that the distribution 
of the chemical forms of iodine released during an accident could be 
different from the distribution in the traditional source term defined 
in U.S. Atomic Energy Commission Technical Information Document (TID) 
14844, ``Calculation of Distance Factors for Power and Test Reactor 
Sites.'' Because of the possible differences, the staff allowed 
licensees to postpone changing their CREs until the ongoing source term 
research was completed or until a generic letter on CRE habitability 
was issued. The staff believed that postponing changes were reasonable 
since the source term research or improved methods of analyses might 
prove that they were unnecessary. Many of these licensees incorporated 
compensatory actions into their operating procedures to assure that the 
control room operators would be protected in case of an accident. Since 
then, other licensees have found that they could not meet the thyroid 
dose limits for habitability without using compensatory actions. The 
NRC also allowed these facilities to use compensatory actions until 
completion of the source term research. In August 2000, the NRC staff 
incorporated the results of the source term into Regulatory Guide 
1.183, ``Alternative Radiological Source Terms for Evaluating Design 
Basis Accidents at Nuclear Power Reactors,'' and it is now available 
for use by licensees.
    Although many CRE integrity testing programs focus on radiological 
concerns, radiation is only one potential design basis challenge to the 
protection of the operators. The inleakage of other contaminants may 
have a greater impact on CRE habitability. An inleakage rate that is 
tolerable for one contaminant may not be tolerable for another. The CRE 
licensing basis describes the hazardous chemical releases considered in 
the CRE design, the design features, and the administrative controls 
implemented to mitigate the consequences of these releases to the 
control room operators. Smoke and other byproducts of fire within the 
CRE or in adjacent areas are among the contaminants that can have an 
adverse impact on CRE habitability.

Discussion

    The NRC is concerned that some licensees have not maintained 
adequate configuration control over their CREs and have not corrected 
identified design and performance deficiencies. Errors of omission and 
commission are more likely if CREHSs and CREs do not properly perform 
as intended in response to challenges from off-normal or accident 
situations. The CRE must be safe so that operators remain in the CRE to 
monitor plant performance and take appropriate mitigative actions. This 
is an underlying assumption in both the design basis and severe 
accident risk analyses. It is, therefore, imperative to the health and 
safety of the public that operators are confident of their safety in 
the CRE at all times.
    The scope and magnitude of the problems that NRC staff and 
licensees have identified raise concerns about whether similar design, 
configuration, and operability problems exist at other reactor 
facilities. The NRC staff is particularly concerned about whether 
licensees' programs to maintain configuration control of CREHSs are 
sufficient to demonstrate that the physical and functional 
characteristics of CREHSs are consistent with and are being maintained 
according to their design bases. It is emphasized that the NRC's 
position has been, and continues to be, that it is the responsibility 
of individual licensees to know the licensing basis for the CRE and 
associated CREHSs. 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 CRE and CREHSs.
    The technical specifications for about 75 percent of the CREs 
(comprised mostly of positive-pressure CREs) have a Surveillance 
Requirement (SR) to measure the P from the CRE to adjacent 
areas. The bases of the Improved Standard Technical Specifications say 
that this SR demonstrates CRE integrity with respect to unfiltered 
inleakage. The E741 integrated testing proves that it does not. Because 
10 CFR 50.36 requires technical specifications to be derived from the 
safety analyses, the staff feels that the existing deficiency should be 
corrected. This correction is consistent with the 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 CREs under all 
normal and off-normal plant conditions, the addressees are requested to 
provide certain information that will enable the NRC staff to verify 
whether addressees can demonstrate and maintain the current design 
bases for the CRE at their facilities. Addressees are encouraged, but 
not required, to work closely with industry groups on the coordination 
of their responses. Coordinating the responses is more efficient and 
public confidence may ensue from a uniform approach to demonstrating 
compliance with the design bases of their CREs.
    NEI 99-03, ``Control Room Habitability Assessment Guidance,'' 
provides industry generic guidance on CRE habitability. The NRC staff 
reviewed NEI 99-03, but rather than fully endorse NEI 99-03, the NRC 
staff developed its own guidance. Draft Regulatory Guide DG-1114, 
``Control

[[Page 31388]]

Room Habitability at Nuclear Power Reactors,'' endorses NEI 99-03 to 
the extent possible and provides additional guidance. Licensees are not 
required to comply with DG-1114, but may find it useful in responding 
to this generic letter. Licensees unable to confirm item 1 under the 
Required Information section may also use DG-1114 to develop and 
implement corrective actions.

Requested Information

    Addressees are requested to provide the following information 
within 180 days of the date of this generic letter.
    1. Confirmation that your facility's CRE meets its applicable 
habitability regulatory requirements (e.g., GDC 1, 3, 4, 5, and 19) and 
that the CRE and CREHSs are designed, constructed, configured, 
operated, and maintained in accordance with the facility's design and 
licensing basis. Emphasis should be placed on confirming:
    (a) That the most limiting unfiltered inleakage into your CRE (and 
the filtered inleakage if applicable) is no more than the value assumed 
in your design basis radiological analyses for CRE habitability. 
Describe how and when you performed the analyses, tests, and 
measurements for this confirmation.
    (b) That the most limiting unfiltered inleakge into your CRE (and 
filtered inleakage if applicable) is incorporated into your fire and 
hazardous chemical assessment, and the CRE integrity preserves the 
reactor control capability from either the CRE or the alternate 
shutdown panel in the event of a fire.
    (c) That your technical specifications are adequate to demonstrate 
the OPERABILITY of your CRE (where OPERABILITY is defined by your 
technical specifications). If you currently have a P 
surveillance requirement to demonstrate CRE integrity, provide the 
basis for your conclusion that it remains adequate to demonstrate CRE 
integrity in light of the E741 testing results. If your facility does 
not currently have a technical specification surveillance requirement 
for your CRE, explain how and on what frequency you confirm your CRE 
integrity.
    (2) If you currently use compensatory measures to demonstrate CRE 
habitability, describe the compensatory measures at your facility and 
the corrective actions to retire these compensatory measures in 
accordance with your related commitments.
    (3) If you believe that your facility is not required to meet 
either the GDC, draft GDC, or ``Principle Design Criteria'' regarding 
CRE habitability, provide documentation (e.g. PSAR, FSAR sections etc.) 
of the basis for this conclusion and identify your actual requirements.

Requested Response

    If an addressee cannot provide the information or cannot meet the 
requested completion date, the addressee should submit a written 
response indicating this within 60 days of the date of this generic 
letter. The response should address any alternative course of action 
the addressee proposes to take, including the basis for the 
acceptability of the proposed alternative course of action.
    The written response should be addressed to the U.S. Nuclear 
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 
20555-0001. A copy of the response should be sent to the appropriate 
regional administrator.
    NRC staff will review the responses to this generic letter and, if 
concerns are identified, will notify affected addressees. The staff may 
conduct inspections to determine licensees' effectiveness in addressing 
this generic letter.

Applicable Regulatory Requirements

    Several provisions of the NRC regulations and plant operating 
licenses (technical specifications) pertain to the issue of CRE 
habitability. The general design criteria (GDC) for nuclear power 
plants (appendix A to 10 CFR part 50), or, as appropriate, quality 
assurance requirements in the licensing basis for a reactor facility, 
stated in appendix B of 10 CFR part 50, and the technical 
specifications, are the bases for the NRC staff's assessment of CRE 
habitability.
    Appendix A to 10 CFR part 50, ``General Design Criteria (GDC) for 
Nuclear Power Plants,'' and the plant safety analyses require or commit 
licensees to design and test safety-related structures, systems, and 
components (SSCs) to provide adequate assurance that they can perform 
their safety functions. The NRC staff applies these criteria to plants 
with construction permits issued on or after May 21, 1971, and to those 
plants whose licensees have committed to them. The applicable GDC are 
GDC 1, 3, 4, 5, and 19. GDC 1 requires quality standards commensurate 
with the importance of the safety functions performed. GDC 3 requires 
SSCs to be designed and located to minimize the effects of fires. GDC 4 
requires SSCs to be designed to accommodate the effects of accidents. 
GDC 5 requires that an accident in one unit will not significantly 
impair orderly shutdown and cooldown of the remaining unit.
    GDC 19 specifies that a control room be provided from which actions 
can be taken to operate the nuclear reactor safely under normal 
conditions and maintain the reactor in a safe condition under accident 
conditions, including a loss-of-coolant accident (LOCA). There must be 
adequate radiation protection to permit personnel to access and occupy 
the control room under accident conditions without receiving radiation 
exposures in excess of specified values.
    Before the issuance of the GDC, proposed GDC (sometimes called 
``principal design criteria'') were published in the Federal Register 
for comment. As they evolved, several of the proposed GDC addressed CRE 
habitability. A facility may have been licensed before the issuance of 
the GDC, but licensees may have committed to the proposed GDC as they 
existed at the time of licensing.
    Following the accident at Three Mile Island (TMI), TMI Action Plan 
Item III.D.3.4, ``Control Room Habitability Requirements,'' as 
clarified in NUREG-0737, ``Clarification of TMI Action Plan 
Requirements,'' required all licensees to assure that control room 
operators would be adequately protected against the effects of 
accidental releases of toxic and radioactive gases and that the nuclear 
power plant could be safely operated or shut down under design basis 
accident conditions. When licensees proposed modifications, the NRC 
issued orders confirming licensee commitments. As a result, most plants 
licensed before the GDC were formally adopted were then required to 
meet the TMI Action Plan III.D.3.4 requirements.
    Appendix B to 10 CFR part 50, ``Quality Assurance Criteria for 
Nuclear Power Plants and Fuel Reprocessing Plants,'' establishes 
quality assurance requirements for the design, construction, and 
operation of those SSCs that prevent or mitigate the consequences of 
postulated accidents that could cause undue risk to the health and 
safety of the public. Criterion III of appendix B, ``Design Control,'' 
requires that design control measures be provided for verifying or 
checking the adequacy of design. A suitable testing program is 
identified as one method of accomplishing this verification.
    Section 36 of 10 CFR part 50, ``Technical Specifications,'' 
requires technical specifications to be derived from the safety 
analyses.
    If, in the course of preparing a response to the requested 
information, an addressee determines that its facility is not in 
compliance with the Commission's requirements, the addressee is 
expected to take appropriate action in accordance with

[[Page 31389]]

requirements of appendix B to 10 CFR part 50 and the plant technical 
specifications to restore the facility to compliance.

Reasons for Information Request

    This generic letter transmits an information request that is 
necessary to permit the assessment of plant-specific compliance with 
applicable regulatory requirements. Specifically, this information will 
enable the NRC staff to determine whether the CREs at power reactor 
facilities comply with the current licensing bases.
    The habitability of the CRE and the operability of the CREHS in the 
event adverse environmental conditions prevail external to the CRE have 
a direct nexus to maintaining public health and safety. Plant design 
bases and severe accident risk analyses both assume that the control 
room operators remain safely within the CRE to monitor plant 
performance and take appropriate mitigative actions. It is essential 
that operators be confident of their safety within the CRE at all 
times.

Backfit Discussion

    This generic letter transmits an information request for the 
purpose of verifying compliance with existing applicable regulatory 
requirements (see the applicable regulatory requirements section of 
this generic letter). This generic letter does not constitute a backfit 
as defined in 10 CFR 50.109(a)(1) since it does not impose 
modifications or additions to structures, systems, and components or to 
the design or operation of an addressee's facility. Nor does it impose 
an interpretation of the Commission's rules that is either new or 
different from a previous staff position. Therefore, no backfit is 
either intended or approved by this generic letter, and the staff has 
not performed a backfit analysis.

Small Business Regulatory Enforcement Fairness Act

    The NRC has determined that this action (a generic letter) is not 
subject to the Small Business Regulatory Enforcement Fairness Act of 
1996.

Federal Register Notification

(To be completed after the public comment period.)

Paperwork Reduction Act Statement

    This generic letter contains an information collection that is 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). This information collection was approved by the Office of 
Management and Budget, clearance number 3150-0011, which expires July 
31, 2003.
    The burden to the public for this information collection is 
estimated to average 200 hours per response, including the time for 
reviewing instructions, searching existing data sources, gathering and 
maintaining the data needed, and completing and reviewing the 
information collection. The NRC is seeking public comment on the 
potential impact of the information collection contained in the generic 
letter and on the following issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC? Will the information have 
practical use?
    2. Is the burden estimate accurate?
    3. Can the quality, utility, or clarity of the information to be 
collected be improved?
    4. How can the burden of the information collection be minimized? 
Can automated collection techniques be used?
    Comments on any aspect of this information collection, including 
suggestions for reducing the burden, should be sent to Records 
Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001 or by Internet electronic mail to 
[email protected]; and to the Desk Officer, Office of Information 
and Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management 
and Budget, Washington, DC 20503.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, an information collection unless the requesting document 
displays a currently valid OMB control number.
    Questions about this matter should be addressed to the technical 
contact or the Office of Nuclear Reactor Regulation project manager for 
your facility.

    Dated at Rockville, Maryland, this 3rd day of May 2002.
    For the Nuclear Regulatory Commission.
William D. Beckner,
Program Director, Operating Reactor Improvements Program, Division of 
Regulatory Improvement Programs, Office of Nuclear Reactor Regulation.
[FR Doc. 02-11623 Filed 5-8-02; 8:45 am]
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