ML113050583
| ML113050583 | |
| Person / Time | |
|---|---|
| Issue date: | 12/29/2011 |
| From: | Laura Dudes, Mcginty T Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking |
| To: | |
| Mensah, T M, NRR/DPR, 415-3610 | |
| Shared Package | |
| ML113050591 | List: |
| References | |
| RIS-11-012, Rev 1 | |
| Download: ML113050583 (12) | |
See also: RIS 2011-12
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001
December 29, 2011
NRC REGULATORY ISSUE SUMMARY 2011-12, REVISION 1
ADEQUACY OF STATION ELECTRIC DISTRIBUTION SYSTEM
VOLTAGES
ADDRESSEES
All holders of, or applicants for, a power reactor operating license or construction permit under
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of
Production and Utilization Facilities, except those who have permanently ceased operations
and have certified that fuel has been permanently removed from the reactor vessel.
All holders of, and applicants for design certifications and combined licenses under 10 CFR Part 52, Licenses, Certifications and Approvals for Nuclear Power Plants.
INTENT
The U.S. Nuclear Regulatory Commission (NRC) is issuing this Regulatory Issue Summary
(RIS) to clarify the NRC staffs technical position on existing regulatory requirements.
Specifically, this RIS clarifies voltage studies necessary for Degraded Voltage Relay (second
level undervoltage protection) setting bases and Transmission Network/Offsite/Station electric
power system design bases for meeting the regulatory requirements specified in General
Design Criteria (GDC) 17 to 10 CFR Part 50, Appendix A. For nuclear power plants that were
licensed before GDC 17 applied, the updated final safety analysis report provides the applicable
design criteria. This RIS does not transmit any new requirements or staff positions. No specific
action or written response is required.
BACKGROUND
The events at Millstone and Arkansas Nuclear One (ANO) that led to the NRC staffs position
regarding degraded voltage protection for nuclear power plant Class 1E electrical safety buses
for sustained degraded transmission network (grid) voltage conditions, and expectations for
voltage calculations for the plant offsite/station electric power system design respectively, are
discussed below as a reminder of past operating experience.
Millstone Unit 2
Electrical grid events at the Millstone Station, in July of 1976 demonstrated that when the Class
1E buses are supplied by the offsite power system, sustained degraded voltage conditions on
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the grid can cause adverse effects on the operation of Class 1E loads. These degraded voltage
conditions will not be detected by the Loss-of-Voltage Relays (LVRs) which are designed to
detect loss of power to the bus from the offsite circuit(s). The LVRs low voltage dropout setting
is generally in the range of 0.7 per unit voltage or less, with a time delay of less than 2 seconds.
As a result of further evaluation of the Millstone events, it was determined that improper voltage
protection logic can also cause adverse effects on the Class 1E systems and equipment, such
as spurious load shedding of Class 1E loads from the standby diesel generators and spurious
separation of Class 1E systems from offsite power due to normal motor starting transients. For
more information regarding this event, see Agencywide Documents Access and Management
System (ADAMS) Accession No. ML093521388.
As a result of these Millstone events, the NRC requested that all licensees implement degraded
voltage protection as described in a 1977 Generic Action (Multi-plant Action B-23) to ensure
automatic protection of safety buses and loads. Multi-plant Action B-23 provides guidance
which applies to all operating reactors at that time and plants licensed since, on how to comply
with the requirements in 10 CFR Part 50, Appendix A, GDC 17. Since degradation of the offsite
power system can lead to or cause the failure of redundant Class 1E safety-related electrical
equipment, the NRC requested that licensees install degraded voltage protection schemes
(second level of voltage protection (Degraded Voltage Relays (DVRs)) for the station electric
power system) as described in NRC letters dated June 2 & 3, 1977 (Multi-plant Action B-23),
Statement of Staff Positions Relative to Emergency Power Systems for Operating Reactors,
which were sent to all operating nuclear power plant licensees. As an example, see the NRC
letter dated June 2, 1977, ADAMS Accession No. ML100610489, sent to the licensee for Peach
Bottom Atomic Power Station. In this letter, the NRC requested that these DVR circuits satisfy
the following criteria:
a) The selection of voltage and time delay setpoints shall be determined from an
analysis of the voltage requirements of the safety-related loads at all station electric
power system distribution levels;
Note: Voltage requirements of all safety-related loads should be determined
based on manufacturers design and operating requirements. For
example, safety injection motors have starting and running voltage
requirements. Motor operated valves have minimum operating voltage
requirements. Motor Control Center contactors have minimum pickup
and operating voltages. All voltage requirements for all safety-related
loads need to be preserved by the DVR circuit(s) during all operating and
accident conditions.
b) The voltage protection shall include coincidence logic to preclude spurious trips of the
offsite power source;
c) The time delay selected shall be based on the following conditions:
(1) The allowable time delay, including margin, shall not exceed the maximum
time delay that is assumed in the final safety analysis report (FSAR) accident
analyses;
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Note: Time delay condition (1) indicates that the DVR circuits should be
designed assuming coincident sustained degraded grid voltage and
accident events. Upon the onset of the coincident accident and
degraded grid event, the time delay for the DVR circuit should allow
for separation of the 1E buses from the offsite circuit(s) and
connection to the 1E onsite supplies in time to support safety
system functions to mitigate the accident in accordance with the
FSAR accident analyses.
(2) The time delay shall override the effect of expected short duration grid
disturbances, preserving availability of the offsite power source(s); and
(3) The allowable time duration of a degraded voltage condition at all distribution
system levels shall not result in failure of safety-related systems or
components;
d) The voltage monitors (or DVRs as defined above) shall automatically initiate the
disconnection of offsite power source(s) whenever the voltage and time delay limits
have been exceeded; and
e) The voltage monitors (DVRs) shall be designed to satisfy the requirements of IEEE
Standard 279-1971, Criteria for Protection Systems for Nuclear Power Generating
Stations; and
f) The Technical Specifications shall include limiting conditions for operation,
surveillance requirements, trip setpoints with minimum and maximum limits, and
allowable values for second-level voltage protection DVRs.
The NRC incorporated the staff positions to meet GDC-17 requirements in Multi-plant Action
B-23 into Branch Technical Position (BTP) of the Standard Review Plan (SRP/NUREG-0800),
PSB-1, Revision 0, Adequacy of Station Electric Distribution System Voltages, dated July 1981
(ADAMS Accession No. ML052350520), which was updated later becoming BTP 8-6 of the
SRP, Revision 3, Adequacy of Station Electric Distribution System Voltages, dated March
2007 (ADAMS Accession No. ML070710478). In addition, the SRP provides a design
approach, consistent with the original Multi-plant Action B-23, with respect to the selection of the
time delay for the DVR circuit.
Arkansas Nuclear One
Another degraded voltage event, in September of 1978, at ANO station demonstrated that
degraded voltage conditions could exist on the Class 1E buses even with normal transmission
network (grid) voltages, due to deficiencies in equipment between the grid and the Class 1E
buses (Offsite/Station electric power system design) or by the starting transients experienced
during certain accident events not originally considered in the sizing (design) of these circuits.
Information Notice No. 79-04, Degradation of Engineered Safety Features, (ADAMS
Accession No. ML031180118) provides additional information regarding this event.
The NRC staff issued Generic Letter 79-36, August 8, 1979, Adequacy of Station Electric
Distribution Systems Voltages (ADAMS Legacy No. 7908230155), expanding its generic review
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of the adequacy of electric power systems for operating nuclear power plants. Specifically, the
NRC requested that all licensees review the electric power systems at each of their nuclear
power plants to determine analytically if, assuming all onsite sources of AC power are not
available, the offsite power system and the station electric power system is of sufficient capacity
and capability to automatically start as well as run all required safety-related loads.
Recent Inspection Findings
Despite lessons learned from past events, and the generic communications on degraded
voltage protection and adequate station voltages, NRC inspectors have identified incorrect
implementation of degraded voltage protection schemes by the licensees at various plants
during inspections. Specifically, the existing degraded voltage setpoints at some plants were
not adequate to protect the safety-related components during degraded voltage conditions for
accident and non-accident conditions. In some cases, the voltage conditions were too low to
power the safety-related equipment but high enough to prevent transferring of safety loads to
the standby power source. In addition, the time delays provided for the degraded voltage
protection relays were not consistent with the accident analysis assumptions for those plants.
Although the licensees analyses were site-specific, the NRC staff is concerned that other
licensees might not have adequately implemented the staff positions and guidance issued
previously to address the adequacy of station electrical distribution system voltages. Examples
of inspection findings recently identified by the inspectors include the following:
DC Cook Units 1 and 2
During the safety system design and performance capability, biennial baseline inspection (NRC
Inspection Report No. 50-315/03-07(DRS); 50-316/03-07(DRS)) (ADAMS Accession No.
ML032260201) at the DC Cook Nuclear Power Plant, in July of 2003, NRC inspectors identified
that the degraded voltage protection scheme was bypassed whenever the 4160V buses were
not being supplied through the reserve auxiliary transformers. This resulted in a lack of
automatic degraded voltage protection during normal operation and for the first 30 seconds of
an accident when engineered safety feature loads were being sequenced onto the safety buses.
This condition did not meet the staff position described in BTP PSB-1 and the electrical scheme
is contrary to the design criteria for degraded voltage protection stated in an NRC letter to the
licensee (a version of a letter sent to all licensees) dated June 3, 1977. This issue was
reviewed by the NRR technical staff under Task Interface Agreement (TIA) 2004-02, and the
staff concluded that the degraded voltage protection design at DC Cook was inadequate and as
such should be modified to include degraded voltage protection during normal operation as well
(ADAMS Accession No. ML043480350). Because the NRC staff had approved DC Cooks
degraded voltage protection design in 1980, the staffs 2005 determination that the design was
inadequate constituted a change in position and was subject to a backfit analysis. By letter
dated November 9, 2005 (ADAMS Accession No. ML050680057), the NRC imposed a
facility-specific compliance backfit on DC Cook Nuclear Plant, Units 1 and 2 to bring the facility
into compliance with its license, the rules and orders of the Commission, and the licensees
written commitments. The licensee implemented a plant modification to the degraded voltage
relaying circuit to make it functional during normal operation (see ADAMS Accession No.
ML060530405) addressing the backfit issue.
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Fermi Unit 2
In May of 2008, NRC inspectors determined that the time delay settings of the degraded voltage
relays for both divisions I and II of the Class 1E electrical distribution system were inadequate.
The time delays could impact the emergency core cooling system (ECCS) injection timing
requirements of the licensees 10 CFR 50.46 loss-of-coolant accident (LOCA) analysis during a
degraded voltage condition. The licensees degraded voltage protection scheme could result in
the voltage being too low to adequately power the ECCS equipment but high enough to prevent
the emergency diesel generators from connecting to the safety-related buses in a timely
manner. This issue was reviewed by the NRR technical staff under TIA 2007-03 (ADAMS
Accession No. ML080420435). The staff determined that the current degraded voltage
protection scheme was inadequate as the time delay relay settings for the degraded voltage
relays for both divisions could impact the emergency core cooling system injection timing
requirements. Additionally, for a short period of time under degraded voltage conditions,
voltage could be too low for the proper operation of safety-related motors but high enough to
prevent emergency diesel generator start. Because the NRC staff had approved Fermis
degraded voltage protection design in 1981, the staffs 2008 determination that the design was
inadequate constituted a change in position and was subject to a backfit analysis. The staff
determined that the provisions of 10 CFR 50.109 (a) (4) were applicable, and that a modification
was necessary to bring the facility into compliance with the rules and orders of the Commission.
See NRC Inspection Report 05000341/2008008 (ADAMS Accession No. ML081720585) for
additional details. The NRC approved the plant modification in License Amendment No.183
(ADAMS Accession No. ML102770382).
Peach Bottom Atomic Power Station Units 2 and 3
Exelon did not use the safety-related degraded grid relay trip setpoint specified in the Technical
Specifications (TS) as a design input in calculations to ensure adequate voltage was available
to all safety-related components required to respond to a design basis LOCA. Instead, Exelon
used the results from a Voltage Regulation Study to establish the voltage level for system
operability. The study credited the use of non-safety-related equipment (load tap changers) to
raise the voltage level. This allowed higher voltages to be used in the design calculations for
components than would be allowed by the TS setpoint. The NRR technical staff reviewed the
issue in TIA 2009-07 (ADAMS Accession No. ML102710178). The staff concluded that the
licensee must demonstrate that the existing degraded voltage trip setpoints, including allowable
values and time delays shown in the licensees TS Table 3.3.8.1, are adequate to protect and
provide the required minimum voltage to all safety-related equipment. Since the load tap
changers are not safety-related and are subject to operational limitations and credible failures,
they cannot be relied on to establish degraded voltage relay setpoints and time delay input for
design basis calculations. For additional details, see NRC Inspection Report 05000277/2010004 and 05000278/2010004 (ADAMS Accession No. ML103140643). The
licensee subsequently issued Licensee Event Report 2-10-04 (ADAMS Accession No.
ML103280505) based on the determination that certain plant equipment could be degraded as a
result of lower voltages that could exist during a postulated design basis loss-of-coolant event
coupled with certain degraded voltage conditions.
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Palo Verde Nuclear Generating Station Units 1, 2, and 3
In July of 2009, an NRC inspection team questioned the calculations that demonstrate adequate
voltage to safety-related loads during worst case loading conditions and the adequacy of a time
delay of 35 seconds for transfer of safety buses to the onsite power supplies should an actual
degraded voltage condition occur. The licensees calculation assumed a voltage above the
degraded bus setpoint to demonstrate adequate voltage at the terminals of the safety-related
loads rather than the degraded voltage dropout setpoint value. The licensee maintains that a
degraded voltage condition concurrent with a design basis accident is not credible. See NRC
Inspection Report 05000528; -529; and -530/2009008, ADAMS Accession No. ML093240524
regarding the inspection finding. The NRR technical staff reviewed the issue in TIA 2010-05
(ADAMS Accession No. ML102800340). The staff concluded that the licensees calculation
must demonstrate that the trip setpoint adequately protects the Class 1E equipment powered by
the safety-related bus from a potentially damaging degraded voltage condition, and the time
delay to transfer from a degraded offsite source to the standby power source to support the
emergency core cooling equipment operation must be consistent with accident analysis time
assumptions, as recommended by BTP PSB-1 (NUREG 0800).
SUMMARY OF ISSUES
Because the NRC continues to identify inspection findings associated with degraded voltage,
the NRC is providing clarifying information on two issues related to the need for two sets of
calculations for the design of the electric power systems of a nuclear power plant and its
interface with the transmission network as defined in GDC 17. The two issues are (1) Degraded
Voltage Relaying Design Calculations, and (2) Offsite/Station Electric Power System Design
Calculations.
(1) The Degraded Voltage Relaying Design Calculations establish the necessary settings of
the DVRs to ensure that all safety-related components are provided adequate voltage
based on the design of the plant power distribution system (and the offsite circuits),
including the design of the Class 1E distribution system in the plant and its most limiting
operating configuration(s).
(2) The Offsite/Station Electric Power System Design Calculations specify the voltage
operating parameters of the plant electrical distribution system based on the
transmission network (grid) operating parameters. This interface calculation establishes
operating voltage bands for all plant electrical buses, which ensures that all plant
safety-related components and systems have proper voltage for starting and running in
all operational configurations (expected operational and accident line-ups and
conditions). Therefore, based on normal grid operation (including (grid)
post-contingency), the degraded voltage relays will not operate, maintaining the offsite
power supply to the plant electrical distribution system.
(1) Degraded Voltage Relaying Design Calculations
Proper design of a degraded voltage relaying scheme is needed to ensure that safety-related
systems are supplied with adequate voltages. The purpose of the NRC-developed BTP PSB-1
(revised later to become BTP 8-6) is to provide additional guidance to supplement the 1977
Generic Action (Multi-plant Action B-23) and the SRP and to provide some design details of a
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DVR circuit that satisfies the regulatory requirements (there may be other designs that satisfy
the requirements). The DVR design should protect (ensure voltage requirements are met)
Class 1E safety-related buses and components from sustained degraded voltage conditions on
the offsite power system coincident with an accident as well as during non-accident conditions.
The Class 1E buses should separate from the offsite power system within a few seconds (or
immediately if the design philosophy recommended in BTP PSB-1 is followed) if an accident
occurs coincident with sustained degraded voltage conditions. During normal plant operation,
the Class 1E safety-related buses should automatically separate from the power supply within a
short interval if sustained degraded voltage conditions are detected. The time delay chosen
should be optimized to ensure that permanently connected Class 1E loads are not damaged
under sustained degraded voltage conditions (such as a sustained degraded voltage below the
DVR voltage setting(s) for the duration of the time delay setting).
DVR Setting Design Calculations
Licensee voltage calculations should provide the basis for their DVR settings, ensuring
safety-related equipment is supplied with adequate voltage (dependent on equipment
manufacturers design requirements), based on bounding conditions for the most limiting
safety-related load (in terms of voltage) in the plant.
Note: All voltage requirements for the safety-related equipment must be
preserved by the DVR circuit(s). For example, safety injection motors have
starting and running voltage requirements. Motor operated valves have minimum
operating voltage requirements. Motor Control Center contactors have minimum
pickup and operating voltages. All voltage requirements for all safety-related
loads need to be preserved during all operating and accident conditions.
These voltage calculations should model offsite circuits and the plant electrical
distribution system, including the plant safety-related electrical distribution system, such
that the limiting voltage at the bus monitored by the DVR can be calculated in terms of
the voltage at the terminals of the most limiting safety-related component in the plant in
all required operating conditions (such as starting and running). These models should
include all plant equipment (including non-safety-related) that can affect voltage supplied
to the safety-related equipment. As a minimum, the model should utilize loads on the
plant distribution system consistent with the specific transient or accident being
analyzed. These models would allow calculation of voltages at terminals of all
safety-related equipment with the voltage at the DVR monitored bus at the DVR dropout
setting, providing the necessary design basis for the DVR voltage settings. In this
manner, the DVR circuit ensures adequate voltage (starting and running) to all
safety-related equipment. Voltage-time settings for DVRs should be selected so as to
avoid inadvertent separation of safety buses from the offsite power system during unit
startup, normal operation (including motor starting), and shutdown.
These DVRs should disconnect the Class 1E buses from any power source other than
the emergency diesel generators (onsite sources) if the degraded voltage condition
exists for a time interval that could prevent the Class 1E safety-related loads from
achieving their safety function.
RIS 2011-12, Rev. 1
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Note: Upon the onset of the coincident accident and degraded grid event, the
time delay for the DVR circuit must allow for separation of the 1E buses from the
offsite circuit(s) and connection to the 1E onsite supplies in time to support safety
system functions to mitigate the accident in accordance with the FSAR accident
analyses.
The DVRs should also prevent prolonged operation of Class 1E safety-related loads at
degraded voltage, which could result in equipment damage.
The operation of voltage correcting equipment, external to the 1E distribution system,
should not be assumed for DVR setpoint analyses.
(2) Offsite/Station Electric Power System Design Calculations
The offsite power source is the preferred source of power to safely shut down the plant during
design basis accidents, abnormal operational occurrence, and reactor trips. The licensees
voltage calculations should provide the basis for proper operation of the plant safety-related
electrical distribution system, when supplied from the offsite circuit(s) (from the transmission
network). These calculations should demonstrate that the voltage requirements (both starting
and running voltages) of all plant safety-related systems and components are satisfied based on
operation of the transmission system (including the bounding transmission system single
contingency in terms of voltage drop) and the plant onsite electric power system during all
operating configurations of transmission network and plant systems. In addition, during
accident conditions, the nuclear unit generator trip (transmission system single contingency)
and associated transmission system voltage drop should be factored into the accident case
voltage calculations since unit trip occurs as a result of the accident. In this way, all
safety-related systems and components will function as designed with proper starting and
running voltages during all plant conditions and the DVRs will not actuate (separating the
transmission network supply). The following are guidelines for voltage drop calculations derived
from Generic Letter 79-36, which have been supplemented to add clarifying information. They
do not represent new NRC staff positions.
Guidelines for voltage drop calculations
a) The plant voltage analysis, while supplied from the transmission network, should be
based on the operating voltage range of the transmission network connection. This
transmission owner/operator supplied voltage range should address all transmission
network and plant system operating configurations and should also include voltage
drop due to the bounding worst case transmission system contingency (transmission
system contingencies include trip of the nuclear power plant). The unit trip grid
contingency voltage drop value should be used in the accident cases in accordance
with the plant accident analyses since a unit trip occurs with an accident.
b) Separate analyses should be performed assuming the power source to the safety
buses is (1) the unit auxiliary transformer; (2) the startup transformer; and (3) other
available connections (e.g., from all available connections) to the offsite network one
by one assuming the need for electric power is initiated by (1) an anticipated transient
such as a unit trip (e.g., anticipated operational occurrence), or (2) an accident,
whichever presents the bounding load demand on the power source.
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c) For multi-unit stations, a separate analysis should be performed for each unit
assuming (1) an accident in the unit being analyzed and simultaneous shutdown and
cooldown of all other units at the station in accordance with the plants licensing basis;
or (2) an anticipated transient (anticipated operational occurrence/GDC 17) in the unit
being analyzed (e.g., unit trip) and simultaneous shutdown and cooldown of all other
units at that station, whichever presents the largest load situation.
d) All actions that the electric power system is designed to automatically initiate or
control should be assumed to occur as designed (e.g., automatic bulk or sequential
loading or automatic transfers of bulk loads from one transformer to another,
automatic starts of components, operation of automatic voltage controlling equipment
such as capacitor bank switching or load tap changers). All non-safety-related plant
auxiliary loads should be included, as applicable, in the plant loading studies since
their operation can affect voltage to safety-related equipment.
e) Manual load shedding should not be assumed.
f) For each event analyzed, the maximum load necessitated by the event and the mode
of operation of the unit at the time of the event should be assumed in addition to all
loads caused by expected automatic actions and manual actions permitted by
administrative procedures.
g) The voltage analysis should include documentation for each condition analyzed, of
the voltage at the input and output of each transformer and at each intermediate bus
between the connection of the offsite circuit(s) and the terminals of each
safety-related load.
h) The calculated voltages at the terminals of each safety-related load should be
compared with the required voltage range for normal operation and starting of that
load calculated in Item a) above. Any identified inadequacies of calculated voltage
should require immediate remedial action.
i) For each case evaluated, the calculated voltages on each safety bus should
demonstrate adequate voltage at the safety bus and down to the component level.
j) To provide assurance that actions taken to assure adequate voltage levels for
safety-related loads do not result in excessive voltages, assuming the maximum
expected value of voltage at the connection to the offsite circuit(s), a determination
should be made of the maximum voltage expected at the terminals of all safety-related
equipment and their starting circuits (if applicable). If this voltage exceeds the
maximum voltage rating of any safety-related equipment, immediate remedial action
should be taken.
k) Analysis documentation should include a statement of the assumptions for each case
analyzed.
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BACKFIT DISCUSSION
The NRC has evaluated this RIS against the criteria of 10 CFR 50.109, 10 CFR Part 50, Appendix A, GDC 17, NRC Letter dated June 2, 1977, Statement of Staff Positions Relative to
Emergency Power Systems for Operating Reactors, BTP-1 and later BTP 8-6 (both of
NUREG 0800) and Generic Letter 79-36, and has determined that it does not represent a
backfit. Specifically, NRC staff technical positions outlined in this RIS are consistent with the
aforementioned regulations and generic communications, while providing more detailed
discussion concerning the necessary voltage calculations supporting DVR settings based only
on voltage requirements of Class 1E components and the Class 1E distribution system design.
Under 10 CFR 50.109, a backfit can be defined as a proposed action that is a modification of
the procedures required to operate a facility and may result from the imposition of a regulatory
staff position that is either new or different from a previously applicable staff position.
FEDERAL REGISTER NOTIFICATION
Although this RIS is informational and does not represent a departure from the current
regulatory requirements, a notice of opportunity for public comment on this RIS was published in
the Federal Register (76 FR 2924) on January 18, 2011, for 30 days. On February 23, 2011, a
Notice was published in the Federal Register extending the comment period for additional 30
days to March 19, 2011, based on the request from Nuclear Energy Institute (ADAMS
Accession No. ML110330025). There were fourteen organizations/individuals that provided
comments, which were considered before issuance of this RIS. Each of the comments were
documented and responded to by NRC staff and are available in ADAMS at Accession No.
ML113050588. This response supersedes the information provided earlier in ADAMS at
Accession Nos. ML111600659 and ML112371830, which were incorrectly released as final
documents when in fact they were drafts. Changes between the draft and final public comment
resolution documents can be viewed in ADAMS at Accession No. ML11357A142.
This RIS does not represent a departure from current regulatory requirements.
CONGRESSIONAL REVIEW ACT
This RIS is not a rule as designated by the Congressional Review Act (5 U.S.C. §§ 801-886)
and, therefore, is not subject to the Act.
PAPERWORK REDUCTION ACT STATEMENT
This RIS does not contain any information collections and, therefore, is not subject to the
requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to respond to, a request for
information or an information collection requirement unless the requesting document displays a
currently valid Office of Management and Budget control number.
RIS 2011-12, Rev. 1
Page 11 of 11
CONTACT
This RIS requires no specific action or written response. If you have any questions, please
contact the technical contact listed below or the appropriate regional office.
/RA/ /RA/by RNelson for
Laura A. Dudes, Director Timothy J. McGinty, Director
Division of Construction Inspection Division of Policy and Rulemaking
and Operational Programs Office of Nuclear Reactor Regulation
Office of New Reactor
Technical Contacts: Roy Mathew, NRR/DE/EEEB
301-415-2965
E-mail: roy.mathew@nrc.gov
Gurcharan Matharu, NRR/DE/EEEB
301-415-4057
E-mail: gurcharan.matharu@nrc.gov
Kenn A Miller, RES/DE/MEEB
301-251-7458
E-mail: kenn.miller@nrc.gov
RIS 2011-12, Rev. 1
Page 11 of 11
CONTACT
This RIS requires no specific action or written response. If you have any questions, please
contact the technical contact listed below or the appropriate regional office.
/RA/ /RA/ by RNelson for
Laura A. Dudes, Director Timothy J. McGinty, Director
Division of Construction Inspection Division of Policy and Rulemaking
and Operational Programs Office of Nuclear Reactor Regulation
Office of New Reactor
Technical Contacts: Roy Mathew, NRR/DE/EEEB
301-415-2965
E-mail: roy.mathew@nrc.gov
Gurcharan Matharu, NRR/DE/EEEB
301-415-4057
E-mail: gurcharan.matharu@nrc.gov
Kenn A Miller, RES/DE/MEEB
301-251-7458
E-mail: kenn.miller@nrc.gov
Note: NRC generic communications may be found on the NRC public website,
http://www.nrc.gov, under Electronic Reading Room/Document Collections.
DISTRIBUTION:
ADAMS Accession No.: ML113050583 * Email attached
OFFICE NRR/DE/EEEB Tech Editor NRR/DE/EEEB NRR/DE NRR/DORL OE
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OFFICE NRO
NAME LDudes
DATE 12/13/11
OFFICAL RECORD COPY