[Federal Register Volume 83, Number 90 (Wednesday, May 9, 2018)]
[Notices]
[Pages 21310-21316]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-09801]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-263; NRC-2018-0090]
Northern States Power Company: Monticello Nuclear Generating
Plant
AGENCY: Nuclear Regulatory Commission.
ACTION: Exemption; issuance.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) has issued an
exemption for the Monticello Nuclear Generating Plant, Docket No. 50-
263, in response to an April 6, 2017, request from Northern States
Power Company. Specifically, the exemption is from the regulation that
requires where redundant trains of systems necessary to achieve and
maintain hot shutdown conditions located within the same fire area
outside of primary containment, one of the redundant trains remains
free of fire damage by one of three methods of physical separation.
DATES: The exemption was issued on May 1, 2018.
ADDRESSES: Please refer to Docket ID NRC-2018-0090 when contacting the
NRC about the availability of information regarding this document. You
may obtain publicly[dash]available information related to this document
using any of the following methods:
Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0090. Address
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
attached exemption. In addition, for the convenience of the reader, the
ADAMS accession numbers are provided in a table in the ``Availability
of Documents'' section of the exemption.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Robert F. Kuntz, Office or Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-3733, email: [email protected].
SUPPLEMENTARY INFORMATION: The text of the exemption is attached.
Dated at Rockville, Maryland, this 3rd day of May 2018.
For the Nuclear Regulatory Commission.
Robert F. Kuntz,
Senior Project Manager, Plant Licensing Branch III, Division of
Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
Attachment--Exemption
NUCLEAR REGULATORY COMMISSION
Docket No. 50-263
Northern States Power Company
Monticello Nuclear Generating Plant
Exemption
I. Background
Northern States Power Company, doing business as Xcel Energy
(the licensee), is the holder of Renewed Facility Operating License
Number 50-263, which authorizes operation of the Monticello Nuclear
Generating Plant (Monticello). The license provides, among other
things, that the facility is subject to all rules, regulations, and
orders of the U.S. Nuclear Regulatory Commission (NRC or the
Commission) now or hereafter in effect.
The facility consists of a boiling water reactor located in
Wright County, Minnesota.
II. Request/Action
In its letter dated April 6, 2017 (Agencywide Documents Access
and Management System (ADAMS) Accession No. ML17096A599), as
supplemented by its letter dated November 20, 2017 (ADAMS Accession
No. ML17324B361), the licensee requested an exemption from Title 10
of the Code of Federal Regulations (10 CFR), Part 50, Appendix R,
Section III.G.2, which requires that where redundant trains of
systems necessary to achieve and maintain hot shutdown conditions
are located within the same fire area outside of primary
containment, that one of the redundant trains remains free of fire
damage by either a 3-hour rated barrier; or 20 feet horizontal
separation, no intervening combustibles, and detection and
suppression system; or a 1-hour barrier, and detection and
suppression systems. The licensee requested NRC approval for
Monticello to use a method to maintain a hot shutdown train free of
fire damage that is not one of the acceptable methods listed in 10
CFR part 50, Appendix R, Section III.G.2. The licensee's exemption
request is intended to justify why the proposed alternative, the use
of a shorting switch, is acceptable in accordance with the
requirements of 10 CFR 50.12, Specific Exemptions.
The regulatory framework that applies to Monticello is contained
in 10 CFR 50.48(b)(1) which requires that plants licensed before
January 1, 1979, to meet Sections III.G, J, and O, of Appendix R to
10 CFR part 50. Monticello began commercial operations in 1971.
Section III.G.2 of 10 CFR part 50, Appendix R, requires, that,
``where cables or equipment, including associated non-safety
circuits that could prevent operation or cause maloperation due to
hot shorts, open circuits, or shorts to ground, of redundant trains
of systems necessary to achieve and maintain hot shutdown conditions
are located within the same fire area outside of primary
containment, one of the following means of ensuring that one of the
redundant trains is free of fire damage shall be provided: a.
Separation of cables and equipment and associated non-safety
circuits of redundant trains by a fire barrier having a 3-hour
rating. Structural steel forming a part of or supporting such fire
barriers shall be protected to provide fire resistance equivalent to
that of the barrier; b. Separation of cables and equipment and
associated non-safety circuits of redundant trains by a horizontal
distance of more than 20 feet with no intervening combustible or
fire hazards. In addition, fire detectors and an automatic fire
suppression system shall be installed in the fire area; or c.
Enclosure of cable and equipment and associated non-safety circuits
of one redundant train in a fire barrier having a 1-hour fire
rating. In addition, fire detectors and an automatic fire
suppression system shall be installed in the fire area.''
In its April 29, 2014, triennial fire protection inspection
report 05000263/2014008, (ADAMS Accession No. ML14119A216), the NRC
staff identified two pairs of Drywell Spray (DWS) motor-operated
valve (MOV) control cables that are not
[[Page 21311]]
protected in accordance with an acceptable option provided in 10 CFR
part 50, Appendix R, Section III.G.2. In 2012, the licensee
installed a modification, called a shorting switch, to mitigate the
lack of protection. The shorting switch modification had been
approved for use at some plants that had adopted a risk-informed
(RI), performance-based (PB) fire protection program (FPP) under 10
CFR 50.48(c)(4). Although Monticello had at one time expressed
intent to adopt a 10 CFR 50.48(c)(4) FPP (ADAMS Accession No.
ML053460342), Monticello later withdrew its letter of intent (ADAMS
Accession No. ML102000433).
The requirements at 10 CFR part 50, Appendix R, Section III.G.2,
require that hot shorts and open circuits be considered, and the
licensee's analysis showed that the shorting switch modification
could fail to meet its design purpose if certain hot shorts and open
circuits were to occur due to fire damage. Therefore, on April 6,
2017, the licensee submitted an RI request for an exemption from the
requirements of 10 CFR part 50, Appendix R, Section III.G.2, to
address postulated spurious actuations of the DWS MOVs that could
occur in the event that an open circuit caused the shorting switch
to fail to perform its function.
III. Discussion
Pursuant to 10 CFR 50.12, the Commission may, upon application
by any interested person or upon its own initiative, grant
exemptions from the regulations when: (1) the exemptions are
authorized by law, will not present an undue risk to public health
or safety, and are consistent with the common defense and security;
and (2) when special circumstances are present. The licensee
requested an exemption from 10 CFR part 50, Appendix R, Section
III.G.2, claiming that the special circumstances of 10 CFR
50.12(a)(2)(ii), which states that, ``Application of the regulation
in the particular circumstances would not serve the underlying
purposed of the rule or is not necessary to achieve the underlying
purpose of the rule,'' apply.
The underlying purpose of 10 CFR part 50, Appendix R, Section
III.G.2, is to provide reasonable assurance of fire protection of
safe shutdown capability by providing a means to ensure that one of
the redundant trains of systems necessary to achieve and maintain
hot shutdown conditions is free of fire damage. The licensee's
position is that the safety benefit, when measured using accepted
probabilistic risk assessment (PRA) techniques, is ``virtually'' the
same as if the plant had used one of the three separation options
described in 10 CFR part 50, Appendix R, Section III.G.2.a, b, or c.
The NRC staff's evaluation of the licensee's exemption request
is provided below.
3.1 Deterministic Technical Evaluation
The fire scenario, as described in the licensee's exemption
request, is that there will be spurious operation of two normally
closed DWS MOVs due to a fire. The cables are routed from the
control room and may be subject to a fire in three other rooms. Two
of the rooms are in Fire Area IX, the rooms (called fire zones) are
Fire Zone 13C--Turbine Building East--Engineered Safeguards Feature
Motor Control Center Area, and Fire Zone 19C--Turbine Building
East--Pipe and Cable Tray Penetration Area. The third room is in
Fire Area XII, Fire Zone 19B, Turbine Building East and Engineered
Safeguards Features Motor Control Center Cable Tunnel. It is within
these three rooms that the separation required by 10 CFR part 50,
Appendix R, Section III.G.2, is not provided.
The scenario postulates that a fire in one of these areas could
damage the control cables to the two DWS MOVs and cause the normally
closed valves to spuriously open. If these valves were to open while
the same division's residual heat removal (RHR) pump were operating,
the scenario postulates that the RHR pumps would be damaged and safe
shutdown capability would be impaired.
3.1.1 Explanation of Postulated Scenario and Shorting Switch
Modification
The NRC staff evaluated the licensee's analysis of how the
protection provided by the shorting switch compares to the 10 CFR
part 50, Appendix R, Section III.G.2 requirement that one train be
free of fire damage by comparing the installed shorting switch
configuration to the configuration required by the regulation. This
section includes a discussion of how the installed shorting switch
works to prevent a spurious opening of the DWS MOVs.
To reduce the likelihood of a spurious actuation, the licensee
installed a shorting switch on one of the valves in series. There
are two trains of DWS. A shorting switch is installed on MOV MO-2020
(Division I), and installed on MOV MO-2021 (Division II). The other
valves in series, MOVs MO-2022 (Division I) and MO-2023 (Division
II), are not equipped with a shorting switch, and therefore may be
subject to an energized cable fault that could cause a spurious
opening of those valves. Figure 1 of the licensee's exemption
request includes a one-line diagram of the system.
When the control room switch is in the closed position, the
shorting switch creates an electrical circuit that provides a low
impedance path bypassing the valve's ``open'' coil. If an energized
cable fault or hot short were to occur that would energize the
``open'' coil, this low impedance path would divert enough current
away from the ``open'' coil through the shorting switch electrical
circuit to prevent the ``open'' coil from actuating. When the
control room switch is set to the open position, this low impedance
path is removed from the circuit and the valve can be opened
normally. The shorting switch only functions to prevent spurious
actuation of the valve in the event of an energized cable fault. A
simplified shorting switch circuit is shown in Figure 2 of the
licensee's exemption request.
The fire scenario of concern would involve three fire-induced
failures. First, an energized cable fault or hot short would need to
occur on control circuitry for the DWS MOV that does not have a
shorting switch installed, for example MO-2022. Second, the fire
would need to cause a cable to become severed, also called an open
circuit, on one of the conductors for the shorting switch protected
valves, such as MO-2020. Third, the fire would have to cause that
same severed cable to MO-2020 to be exposed to an energized cable
fault or hot short. Essentially the severed cable would remove the
shorting switch from the circuit, thereby, defeating the design
capability of the shorting switch. Similarly, the pair of valves
MO[dash]2021 and MO-2023 would be vulnerable to the same potential
failure mode. Note that both valves in a pair, MO-2020 and MO-2022
or MO-2021 and MO-2023, would need to be impacted to remove the
shorting switch from the circuit. A hot short from one cable in the
first pair and one cable in the second pair would not create a
condition where the RHR pumps could be damaged.
3.2 Risk-Informed Technical Evaluation
The licensee's exemption request includes a risk assessment of
the proposed plant change. The use of risk information in a 10 CFR
part 50, Appendix R, exemption request is in accordance with
Regulatory Position 1.8 of Regulatory Guide (RG) 1.189, ``Fire
Protection for Nuclear Power Plants,'' Revision 2, dated October
2009 (ADAMS Accession No. ML092580550), which says that RI/PB
methodologies may be used to evaluate the acceptability of FPP
changes; however, for this approach, the licensee should use
methodologies and acceptance criteria that the NRC has reviewed and
approved. RG 1.174, ``An Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to
the Licensing Basis,'' Revision 2, (ADAMS Accession No.
ML100910006), includes guidance for RI changes to a plant's current
licensing bases.
Accordingly, the NRC staff reviewed the licensee's exemption
request using the review methodology and criteria contained in RG
1.174, Revision 2, which includes the following elements:
Defining the proposed change,
Performing an engineering analysis, including an
evaluation that the proposed change is consistent with the defense-
in-depth (DID) philosophy and the principle that sufficient safety
margins are maintained,
Assessing the technical adequacy of the PRA analysis,
the methods used to determine the risk impact of the proposed
change, and the results of the risk impact assessment,
Defining the implementation and monitoring program to
ensure that no unexpected adverse safety degradation occurs due to
the proposed change, and
Confirming that an integrated approach was used to
evaluate the proposed change.
3.2.1 Proposed Change to the Appendix R Program
Pursuant to 10 CFR 50.12, the licensee requested an exemption
from 10 CFR part 50, Appendix R, Section III.G.2 requirements with
respect to the protection of the control circuitry for the DWS MOVs.
In lieu of meeting the protection requirements of Section III.G.2,
the licensee has installed a shorting switch modification on the
control circuitry for one MOV in each division of the DWS system to
reduce the risk impact of a fire-induced multiple spurious operation
(MSO) that fails both MOVs. A detailed
[[Page 21312]]
description of the modification is provided in Enclosure 1, Section
3.1, of the licensee's exemption request.
3.2.2 Engineering Analysis
Regulatory Position 2.1 of RG 1.174, Revision 2, indicates that,
for RI changes to the plant licensing basis, the licensee should
evaluate the proposed change to determine whether it is consistent
with the DID philosophy and the principle that sufficient safety
margins are maintained.
Fire Protection DID
Regulatory Position 2.1.1 of RG 1.174, Revision 2, provides
guidance on maintaining the philosophy of nuclear safety DID and
identifies several elements to consider in this evaluation. DID
involves prevention, protection, and mitigation. With respect to
nuclear power plant FPPs, the regulations in 10 CFR part 50,
Appendix R, Section II.A state that the FPP shall extend the concept
of DID to fire protection in fire areas important to safety with the
following objectives:
to prevent fires from starting;
to detect rapidly, control, and extinguish promptly
those fires that do occur; and
to provide protection for structures, systems and
components important to safety so that a fire that is not promptly
extinguished by the fire suppression activities will not prevent the
safe shutdown of the plant.
An engineering analysis that evaluates the impact of a proposed
change to an Appendix R FPP on the balance among these FPP DID
elements is deemed by the NRC staff to satisfy the RG 1.174
guidance. Enclosure 1, Section 3.2, of the exemption request
provides the licensee's evaluation of the FPP DID elements. Fire
protection DID elements consist of administrative controls such as
plant procedures to limit combustible materials or control hot work
activities, plant design features, fire protection inspections,
installed fire detection and suppression systems, and passive fire
protection features such as fire barriers.
The licensee's position is that the use of a shorting switch
meets the underlying purpose of the rule by providing equivalent
protection to one of the separation methods of 10 CFR part 50,
Appendix R, Section III.G.2. The licensee chose to install the
shorting switch in lieu of possibly separating the cables for the
valves in series (MO-2021 from MO-2023, and MO-2020 from MO-2022)
into separate areas. The following sections discuss the fire
protection DID elements of preventing fires, suppressing fires that
do occur, and protecting safe shutdown.
Fire Protection DID Element 1--Preventing Fires
The licensee indicated that each of the three rooms has
administratively controlled restrictions on combustibles. The
licensee described that of the three zones, only Fire Area IX, Fire
Zone 13C, has significant fixed ignition sources, which are motor
control centers. The NRC staff finds that this exemption does not
degrade the preventing fires DID element, because the proposed
change does not introduce additional combustibles or ignition
sources at such a level that necessitates additional controls be put
in place to prevent fires from starting.
Fire Protection DID Element 2--Detecting and Extinguishing Fires
The licensee indicated that all three of the rooms addressed in
this exemption are equipped with full area ionization smoke
detection systems. Only Fire Area IX, Fire Zone 13C, has significant
fixed ignition sources and it is equipped with an automatic water
based suppression system which the licensee indicates is based on
the significance of the fire hazards contained within that room. The
smoke detection system annunciates to the control room which results
in response of the fire brigade.
Each of the three rooms included in this exemption has fire hose
stations and fire extinguishers in the rooms or in adjacent rooms.
Fire Area IX, Fire Zone 13C, and Fire Area XII, Fire Zone 19B, are
900 square feet, are considered large rooms, and have extinguishers
and hose stations within the rooms. Fire Area IX, Fire Zone 19C,
does not have a fire hose station or extinguisher in the room.
Because Fire Area IX, Fire Zone 19C, has a small floor area of 204
square feet, the NRC staff concludes that it is reasonable that
extinguishers and fire hoses could be brought from adjoining areas.
The NRC staff also concludes that this exemption does not degrade
the detecting and extinguishing fires DID element, because the
installation of the switches (1) does not impact the ability of the
installed detection and suppression systems to detect and extinguish
a fire, and (2) does not impact the fire brigades ability to
manually extinguish a fire using the installed extinguishers and
fire hose stations.
Fire Protection DID Element 3--Safe Shutdown
The NRC staff determined that the safe shutdown element of fire
protection DID is impacted by this exemption request. The licensee
proposes to install an engineered feature called a shorting switch,
in lieu of the protection required by 10 CFR part 50, Appendix R,
Section III.G.2. Compliance with the regulation by use of a barrier,
or separation with fire detection and suppression, protects against
possible failure modes, but the shorting switch modification results
in a possible failure mode involving hot shorts and open circuits.
10 CFR part 50, Appendix R, Section III.G.2, specifically states
that a plant licensed before January 1, 1979, must address these
failure modes (i.e., ``maloperation due to hot shorts [and] open
circuits'').
Although the licensee has chosen to use a RI analysis to compare
compliance with the regulation and the proposed alternative using a
shorting switch, the following deterministic features are in place,
in addition to the fire prevention, fire detection, and fire
suppression that are discussed above.
A fire would have to occur in one of the three subject areas and
damage the cables to two of the MOVs. One MOV cable would have to be
subjected to an energized fault or hot short, and the second MOV
cable would have to be subjected to both a hot short and a severed
cable also called an open circuit. For the combination of cable
faults to damage the RHR pumps, the pumps would have to be running
at the time of the cable faults. Although possible in an actual
plant event, the licensee did not assume in its evaluation that
plant operators would turn off the pumps before they became damaged.
The NRC staff considers this assumption to be conservative because
the licensee indicated that operators would initiate a controlled
shutdown to preclude equipment failures.
Additionally, the NRC staff determined that hot shorts would
have to be of sufficient duration to open the MOVs enough to result
in a flow that would cause RHR pump failure due to runout and that
typically, hot shorts are of a very short duration.
These aspects of the scenario, the likelihood of cable faults,
the assumption that the RHR pumps are operating, and the possible
operator actions and timing related to mitigating the potentially
damaging configuration were not explicitly credited in the analysis.
The NRC staff has determined that the DID discussion regarding
prevention, protection, and mitigation satisfies the RG 1.174
guidance for a DID analysis because it discussed multiple means to
accomplish safety functions in accordance with the guidance provided
in Regulatory Position 2.1.1 of RG 1.174.
Safety Margins
In Enclosure 1, Section 3.4.3, of the exemption request, the
licensee provided its assessment of how sufficient safety margins
are maintained. The licensee explained that the design and
installation of the shorting switches was completed using applicable
codes and standards and that the Monticello safety analyses were not
impacted by the installation of the switches or the exemption
request. In its letter dated November 20, 2017, in response to the
NRC's October 18, 2017, request for additional information (RAI)
(ADAMS Accession No. ML17293A091), the licensee indicated that
sufficient safety margins are demonstrated by the design, operation,
and performance monitoring of the shorting switches. The licensee
indicated that the RHR system currently meets all applicable codes
and standards (with the exception of the stated 10 CFR part 50,
Appendix R, Section III.G.2 noncompliance), and also stated that
granting the exemption will not affect Monticello's ability to
demonstrate consistency with all applicable codes and standards.
In its November 20, 2017, letter, the licensee also summarized
some of the PRA bases for ensuring sufficient safety margins. The
summarized bases included maintaining a FPP that meets regulatory
requirements, using a fire PRA (FPRA) that was developed in
accordance with NUREG/CR-6850, ``Fire PRA Methodology for Nuclear
Power Facilities,'' having had formal industry peer reviews of
internal events PRAs (IEPRAs) and FPRAs, and using verified and
validated fire models.
The NRC staff concludes that the licensee's safety margins
assessment is acceptable because it demonstrated that codes and
standards or their alternatives approved by the NRC are met, and
that the safety analysis acceptance criteria described in the
licensing basis are met.
[[Page 21313]]
3.2.3 PRA
The licensee performed a risk impact assessment for installation
of the shorting switches rather than physically separating the
control circuitry for the DWS MOVs in accordance with the 10 CFR
part 50, Appendix R, separation requirements. For the assessment,
the risk was evaluated by estimating the change in risk between an
Appendix R-compliant configuration and the as-installed and as-
operated configuration of the shorting switches. The risk assessment
was provided in Enclosure 1, Section 3.3, of the licensee's
exemption request.
Technical Adequacy of the PRA
The licensee used Revision 4.0 of the Monticello FPRA model to
perform the risk impact assessment. For the development of the FPRA,
the licensee modified its IEPRA model to capture the effects of
fire. Therefore, the NRC staff evaluated both the IEPRA and FPRA
quality information provided by the licensee in the exemption
request to determine whether the plant-specific PRA used in the risk
impact assessment includes sufficient scope, level of detail, and
technical adequacy for this assessment.
Consistent with the information provided in NRC Regulatory Issue
Summary 2007[dash]06, ``Regulatory Guide 1.200 Implementation,''
March 22, 2007 (ADAMS Accession No. ML070650428), the NRC staff uses
RG 1.200, ``An Approach for Determining the Technical Adequacy of
Probabilistic Risk Assessment Results for Risk-Informed
Activities,'' Revision 2 (ADAMS Accession No. ML090410014).
The licensee stated that a full-scope peer review was performed
in April 2013, for the IEPRA model (Revision 3.2). The peer review
was performed using Nuclear Energy Institute (NEI) 05-04, Revision
2, ``Process for Performing Internal Events PRA Peer Reviews Using
the ASME/ANS [American Society of Mechanical Engineers/American
Nuclear Society] PRA Standard'' (ADAMS Accession No. ML083430462),
as clarified by RG 1.200, Revision 2. The PRA standard provides
supporting requirements for the PRA against capability categories
(CC) CC-I, CC-II, or CC-Ill. The peer review resulted in
identification of PRA standard supporting requirements that did not
meet CC-II, or that were met and had related findings (Reference:
Evaluation of Risk Significance of Permanent Integrated Leak Rate
Testing Extension--ML16047A273). In Enclosure 2 of the exemption
request, the licensee provided the peer review finding-level facts
and observations (F&Os) against the PRA standard supporting
requirements and the licensee's resolution to each of the F&Os. The
licensee stated that all of the finding-level F&Os have been
resolved and that none were determined to affect the exemption
request.
The licensee stated that a full-scope peer review of the FPRA
model (Revision 1a) was performed in March 2015, using NEI 07-12,
Revision 1, ``Fire Probabilistic Risk Assessment (FPRA) Peer Review
Process Guidelines,'' June 2010 (ADAMS Accession No. ML102230070),
and RG 1.200, Revision 2. The peer review resulted in identification
of PRA standard supporting requirements that did not meet CC-II, or
CC-III for one supporting requirement (Reference: Monticello ILRT
license amendment--ML16047A273). In Enclosure 3 of the exemption
request, the licensee provided the peer review finding-level F&Os
against the PRA standard supporting requirements and its resolution
to each of the F&Os. The licensee stated that all of the finding-
level F&Os have been resolved and that none were determined to
affect the exemption request.
The licensee stated that a focused-scope peer review of Revision
4.0 of the FPRA model was performed in December 2016, of a subset of
high-level requirements impacted by the use of enhanced fire
modeling methods that were implemented subsequent to the March 2015,
peer review. The licensee provided the two peer review finding-level
F&Os from this focused-scope peer review in Enclosure 4 of the
exemption request. The licensee stated that the two finding-level
F&Os have been resolved and that neither was determined to affect
the exemption request. The licensee also stated that the PRA used in
the risk impact assessment represents the current as-installed and
as-operated configuration of Monticello.
The NRC staff reviewed the exemption request to determine the
technical adequacy of the Monticello IEPRA and FPRA models used for
this exemption request. The licensee stated that it evaluated its
PRA against Revision 2 of RG 1.200 and the ASME/ANS PRA standard.
The licensee stated that it had resolved all peer review and
focused-scope peer review finding-level F&Os and concluded that they
had no impact on the exemption request. Based on the information
provided by the licensee, the NRC staff found that the licensee's
PRA represents the current as-installed and as-operated plant, and
the margin between the reported risk values and the guidance
recommended values is acceptable.
The NRC staff concludes that the IEPRA is adequate and can be
used to support the FPRA because the licensee demonstrated that the
resolution of the F&Os did not affect the technical adequacy of the
licensee's PRA analysis submitted to support the licensee's risk
evaluation of the proposed exemption request.
The NRC staff concludes that the IEPRA is adequate and can be
used to support the FPRA because the licensee demonstrated that the
resolution of the F&Os support the technical adequacy of the
licensee's PRA analysis submitted for the licensee's risk evaluation
of the proposed exemption request.
The NRC staff also concludes that the FPRA is of sufficient
technical adequacy and that its quantitative results can be used to
demonstrate that the change in risk due to the lack of physical
separation between the DWS division meets the acceptance guidelines
in RG 1.174 because the licensee demonstrated that the resolution of
the relevant F&Os supports the determination that the quantitative
results are adequate and have no significant impact on the FPRA. For
several F&Os, the NRC staff determined that the resolutions could
impact the delta risk results reported in the exemption request, but
that their resolution is unlikely to change the delta risk results
reported by the licensee in the exemption request enough to increase
the delta core damage frequency (CDF) and the delta large early
release frequency (LERF) by an amount necessary to exceed the RG
1.174 risk guidelines for very small changes.
Based on the above, NRC staff concludes that the FPRA model is
of sufficient technical adequacy to support the risk impact
assessment of the proposed change.
Risk Impact Assessment
The licensee stated that the evaluation of the risk for the
proposed change was done using Revision 4.0 of the Monticello FPRA
model to estimate the change in risk between an Appendix R-compliant
configuration and the as-installed and as-operated configuration of
the shorting switches.
In Enclosure 1, Section 3.3.3, of the exemption request, the
licensee described how it developed the risk of the as-installed and
as-operated configuration of the plant with shorting switches
installed. For this plant configuration, the licensee modified the
FPRA model to include new basic events to fail the DWS MOVs due to
fire-induced MSOs (referred to as the ``variant model''). The model
modification included identifying the cables that could cause a DWS
MOV MSO, identifying the plant locations (fire zones) where these
cables are located in the plant, and linking these cables to
specific fire scenarios modeled in the FPRA. The exemption request
also described the revised fault tree logic that incorporated the
new basic events.
Each of the two DWS trains includes two-normally-closed in-
series MOVs that could fail open due to a fire-induced MSO and
result in core damage. Each in-series pair of DWS MOVs were added
together in the fault tree and assigned a hot short probability. The
MOVs without a shorting switch have a hot short probability of 0.39,
which is taken from Volume 2 of NUREG/CR-7150, ``Joint Assessment of
Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE)''
(ADAMS Accession No. ML14141A129). The MOVs with a shorting switch
are assumed to have a failure probability of 1.0E-03, which is the
assumed failure probability of the shorting switch. In enclosure 1,
section 3.3.5 of the exemption request, the licensee justified its
use of the 1.0E-03 failure probability by explaining that it was
found acceptable by the NRC staff in the safety evaluations related
to National Fire Protection Association 805 license amendment
requests by other licensees (see ADAMS Accession Nos. ML15212A796
and ML16223A481). The licensee stated that the control circuitry
configuration for the shorting switch application at these plants is
substantially similar to that for the Monticello DWS MOVs.
The NRC staff finds that the use of a hot short probability of
0.39 is acceptable because it is the most bounding of the MOV hot
short probabilities for grounded and ungrounded alternate current
control circuits as described in Table 8-1 of NUREG/CR-7150, Volume
2. The NRC staff also finds that the licensee's use of the 1.0E-3
failure probability for the shorting switches is acceptable because
the conditions that would
[[Page 21314]]
have to occur to fail a shorting switch are considered extremely
unlikely.
The exemption request further explained that the flow diversion
created by the failure of just one train of DWS MOVs (i.e., spurious
opening of both in-series MOVs) was assumed to result in damage of
the RHR pumps because activation of the Drywell Sprays would result
in lowering the drywell pressure. This in turn could result in the
potential loss of containment accident pressure, which leads to a
loss of net positive suction head and which, in turn, would fail the
RHR pumps. All RHR pumps are potentially damaged because the RHR
removal cross[dash]tie valves are normally kept open. The failure of
the RHR pumps and loss of net positive suction head result in
failure of all associated functions modeled in the PRA (except DWS),
specifically:
Shutdown cooling,
Low pressure coolant injection (LPCI),
Torus cooling (which fails high pressure coolant
injection and reactor coolant isolation cooling when suction is from
the torus),
Core spray,
Alternate injection with condensate service water, the
fire protection system, or RHR service water, and
Primary containment.
Because of the failure of RHR pumps, the torus sprays would also
fail, which is not modeled in the PRA.
In Enclosure 1, Section 3.3.2, of the exemption request, the
licensee described how it developed the risk of the Appendix R-
compliant configuration. For this plant configuration, the licensee
revised the FPRA model to assume the DWS MOVs do not fail due to a
fire-induced MSO (referred to as the ``compliant model''). The
licensee explained that its assumption is conservative because it
assumes a failure probability of zero for the DWS MOVs due to a
fire[dash]induced MSO. The NRC staff concludes that this assumption
is conservative because, although unlikely, there is a greater-than-
zero probability of a large enough fire that could defeat the
Appendix R protection requirements and produce a MSO that would fail
the MOVs.
In Enclosure 1, Section 3.3.4, of the exemption request, the
license explained that it calculated the change in risk for the
proposed change by subtracting the calculated risk (CDF) and LERF)
for the compliant model from the calculated risk for the variant
model.
Furthermore, in Enclosure 1, Section 3.3.5, of the exemption
request, the licensee identified several conservatisms in the PRA
model that would overestimate the calculated change in risk. These
conservatisms include: the assumption that all postulated control
room fires fail the shorting switches, assumption that the RHR pumps
are running at the time of the MSO event, and the assumption that
the loss of containment accident pressure and net positive suction
head is instantaneous with the MSO event. The NRC staff finds that
these conservatisms make the model overestimate the calculated
change in risk because not all control room fires fail the shorting
switches, because the RHR pumps may not be running at the time of
the MSO event, and because loss of containment accident pressure and
net positive suction head may not be instantaneous with the MSO
event.
Based on the licensee's description of the fault tree modeling
of the MSO event in the compliant and variant models, the NRC staff
concludes that the hot short probability and shorting switch failure
probability are acceptable, and that the calculated change in risk
is likely conservative. The NRC staff further concludes that the
licensee's method for calculating the change in risk is acceptable.
PRA Results and Comparison with Risk Guidelines
In Enclosure 1, Section 3.3.4, of the exemption request, the
licensee reported the results of its risk impact assessment. The
licensee reported the calculated change in risk (variant model risk
minus compliant model risk) for the proposed plant change to be
1.8E-08 per year for CDF and 1.4E-08 per year for LERF, which are
below the RG 1.174, Revision 2, risk guidelines for a ``very small''
change.
Based on its review of the risk impact assessment results, and
the margin between the reported risk values and the risk guidelines,
the NRC staff concludes that the increase in CDF and LERF from the
proposed change is very small per the definition in RG 1.174,
Revision 2. Also, while the licensee did not provide the total plant
risk from all hazards, the NRC staff finds this acceptable and
consistent with RG 1.174, Revision 2, because there is no indication
that the total CDF and LERF is considerably higher than 1.0E-04 and
1.0E-05 per reactor year, respectively.
3.2.4 Implementation and Monitoring
In Enclosure 1, Section 3.4.5, of the exemption request, the
licensee described the implementation and the monitoring program for
the shorting switches and the DWS MOVs. The licensee explained that
the shorting switches were installed in 2012 and that post-
maintenance testing was conducted to ensure that the switches were
installed in accordance with the approved design and that the MOVs
continued to operate as expected. The DWS MOVs will continue to be
regularly exercised in accordance with the Monticello MOV program,
which has been accepted by the NRC staff, as providing an acceptable
level of quality and safety, and are monitored under the Monticello
Maintenance Rule Program.
In its November 20, 2017, letter, the licensee indicated that
Monticello will generate a preventive maintenance task for the
shorting switches to ensure acceptable resistance, and that this
task will be completed within 180 days of the date of the exemption
is issued. The licensee will introduce performance monitoring of the
shorting switches into the Monticello, Appendix R, program, with the
objective to ensure the shorting switches provide a low impedance
path to ground in the event of a fire-induced hot short. The program
will include acceptance criteria, which if exceeded, will cause the
licensee to enter the issue into its corrective action program.
The NRC staff concludes that the proposed monitoring program for
the shorting switches meets the guidelines of RG 1.174, Revision 2,
and that RI applications include performance monitoring and feedback
provisions.
3.2.5 Integrated Decision-making
As described in the previous sections, the licensee's exemption
request and responses to NRC staff RAIs provided an integrated
approach to evaluating the proposed change. Specifically, the
licensee's assessment of the proposed change included:
Performing a traditional engineering analysis,
including an evaluation that the proposed change is consistent with
the DID philosophy and the principle that sufficient safety margins
are maintained,
Assessing the technical adequacy of the PRA analysis,
evaluating the risk impact of the proposed change, and comparing the
results of the risk impact assessment to the
RG 1.174, Revision 2, risk guidelines, and
Defining the implementation of the proposed change and
of a monitoring program to ensure that no unexpected adverse safety
degradation occurs due to the proposed change.
Based on the NRC staff's review of each of these elements of the
licensee's exemption request, the NRC staff concludes that the
licensee's evaluations are acceptable and in accordance with RG
1.174, Revision 2, and that the risk increase of the proposed change
meets the RG 1.174, Revision 2, risk guidelines for a ``very small''
change. Based on this, the NRC staff concludes that the licensee's
integrated evaluation of the proposed change is acceptable.
3.3 Technical Evaluation Conclusion
Based on its review of the information provided by the licensee,
the NRC staff concludes that the licensee's request to credit a
shorting switch does not create any new accident precursors because
the plant's operation remains the same in that fire protection for
structures, systems, and components important to safe shutdown
continues to be provided, and fire damage continues to be limited so
that one of the redundant trains is free of fire.
The NRC staff also concludes that the licensee's evaluations are
acceptable and in accordance with RG 1.174, Revision 2, and that the
risk increase of the proposed change meets the RG 1.174, Revision 2,
risk guidelines for a ``very small'' change. Based on this, the NRC
staff concludes that the licensee's integrated evaluation of the
proposed change is acceptable.
3.4 Authorized by Law
The exemption would allow the licensee to rely on the installed
shorting switch and other fire protection DID features instead of
providing separation in accordance with 10 CFR Part 50, Appendix R,
Section III.G.2. As stated above, 10 CFR 50.12 allows the NRC to
grant exemptions from the requirements of 10 CFR Part 50. The NRC
staff has determined, as described in Section 3.7 below, that
special circumstances exist to grant the proposed exemption and that
granting of the licensee's proposed exemption will not result in a
violation of the Atomic Energy Act of 1954, as amended, or
[[Page 21315]]
the Commission's regulations. Therefore, the exemption is authorized
by law.
3.5 No Undue Risk to Public Health and Safety
The underlying purposes of 10 CFR Part 50, Appendix R, is to
provide reasonable assurance of fire protection safe shutdown
capability. As discussed in Sections 3.1 and 3.2 above, the NRC
staff found that the crediting of a shorting switch permitted by the
proposed exemption does not create any new accident precursors or
degrade detection systems because the plant's operation remains the
same and the installed shorting switch provides an acceptable level
of protection as compared to that provided by compliance with the
regulation.
Because no new accident precursors are created by the proposed
exemption, which would allow the licensee to use, or take credit
using a risk-informed approach, for an installed shorting switch to
ensure that one redundant train is free of fire damage, the
probability of postulated accidents is not significantly increased,
and reasonable assurance of fire protection of safe shutdown
capability is maintained. Therefore, the NRC staff concludes that
the consequences of postulated accidents are not significantly
increased, and there is no undue risk to public health and safety.
3.6 Consistent with Common Defense and Security
The proposed exemption would allow the licensee to rely on the
installed shorting switch instead of providing separation required
by 10 CFR Part 50, Appendix R, Section III.G.2. The NRC staff
concludes that this change to the plant design has no relation to
security issues, therefore, the common defense and security is not
impacted by this exemption.
3.7 Special Circumstances
Special circumstances, in accordance with 10 CFR 50.12, are
present whenever an application of the regulation in the particular
circumstances is not necessary to achieve the underlying purpose of
the rule. The underlying purpose of 10 CFR part 50, Appendix R,
Section III.G.2, is to provide reasonable assurance of fire
protection of safe shutdown capability by providing a means to
ensure that one of the redundant trains of systems necessary to
achieve and maintain hot shutdown conditions is free of fire damage.
The technical evaluation above demonstrates that the shorting switch
and DID features provide reasonable assurance that the underlying
purpose of the rule is met because the licensee demonstrated that
the installed shorting switch provides an acceptable level of
protection that is similar to that provided by compliance with the
regulation. The licensee performed a deterministic engineering
analysis and demonstrated that the proposed change is consistent
with the DID philosophy and maintains sufficient safety margins. The
licensee also assessed the technical adequacy of the PRA analysis
and evaluated the risk impact of the proposed change and compared
the results to the RG 1.174, Revision 2, risk guidelines, and also
defined the implementation of the proposed change and of a
monitoring program to ensure that no unexpected adverse safety
degradation occurs due to the proposed change. Therefore, the NRC
staff concludes that since the underlying purpose of 10 CFR 50,
Appendix R, Section III.G.2 (i.e., ensuring one of the redundant
trains of Drywell Spray is free of fire damage), is achieved, the
special circumstances required by 10 CFR 50.12 for the granting of
an exemption from 10 CFR part 50, Appendix R, Section III.G.2,
exist.
IV. Environmental Considerations.
The NRC staff determined that the issuance of the requested
exemption meets the provisions of categorical exclusion 10 CFR
51.22(c)(9) because the exemption is from a requirement, with
respect to the installation or use of a facility component located
within the restricted area, as defined in 10 CFR part 20 and the
issuance of the exemption involves: (i) No significant hazards
consideration; (ii) no significant change in the types or
significant increase in the amounts of any effluents that may be
released offsite; and (iii) no significant increase in individual or
cumulative occupational radiation exposure. Therefore, in accordance
with 10 CFR 51.22(b), no environmental impact statement or
environmental assessment need be prepared in connection with the
NRC's issuance of this exemption. The basis for the NRC staff's
determination is provided in the following evaluation of the
requirements in 10 CFR 51.22(c)(9)(i)-(iii).
Requirements in 10 CFR 51.22(c)(9)(i)
The NRC staff evaluated whether the exemption involves no
significant hazards consideration by using the standards in 10 CFR
50.92(c), as presented below:
1. Does the requested exemption involve a significant increase
in the probability or consequences of an accident previously
evaluated?
No. The proposed exemption would allow the licensee to rely on
the installed shorting switch instead of providing physical
separation in accordance with 10 CFR Part 50, Appendix R, III.G.2 to
protect structures, systems or components important to safe shutdown
of the plant in the event of a fire. The licensee performed a risk
impact assessment for installation of the shorting switches rather
than physically separating the control circuitry in accordance with
the 10 CFR Part 50, Appendix R, III.G.2 separation requirements. For
the assessment, the risk was evaluated by estimating the change in
fire risk between an Appendix R-compliant configuration and the as-
installed and as-operated configuration of the shorting switches.
Based on its review of the licensee's exemption request, the NRC
staff concludes that the licensee's evaluations are acceptable and
in accordance with Regulatory Guide (RG) 1.174, ``An Approach for
Using Probabilistic Risk Assessment in Risk-Informed Decisions on
Plant-Specific Changes to the Licensing Basis,'' Revision 2, and
that the risk increase of the proposed change meets the RG 1.174,
Revision 2, risk guidelines for a ``very small'' change.
The installation of the shorting switch does not alter plant
operation or affect fire detection capability because fire
protection for structures, systems, and components important to safe
shutdown continues to be provided, and fire damage continues to be
limited so that one of the redundant trains is free of fire damage
and, therefore, would not alter the consequences of any accident
previously evaluated.
Therefore, the exemption does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the requested exemption create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The underlying purposes of 10 CFR Part 50, Appendix R,
III.G.2 is to provide reasonable assurance of fire protection safe
shutdown capability. The exemptions' crediting of a shorting switch
and defense in depth measures does not create any new accident
precursors because the plant's operation and fire detection
capability remains the same.
Therefore, the exemption does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the requested exemption involve a significant reduction
in a margin of safety?
No. The installation of the shorting switch and reliance on
defense in depth measures does not alter plant operation and does
not impact any safety margins because codes and standards or their
alternatives approved by the NRC are met, and the safety analysis
acceptance criteria described in the licensing basis are met.
Therefore, the exemption does not involve a significant
reduction in a margin of safety.
Based on the evaluation above, the NRC staff has determined that
the proposed exemption involves no significant hazards
consideration. Therefore, the requirements of 10 CFR 51.22(c)(9)(i)
are met.
Requirements in 10 CFR 51.22(c)(9)(ii) and (iii)
The proposed exemption would allow the Monticello Nuclear
Generating Plant to maintain a hot shutdown train of Drywell Spray
free of fire damage by using a method that is different from one of
the acceptable methods listed in 10 CFR Part 50, Appendix R, Section
III.G.2. Specifically, In lieu of meeting these protection
requirements, the licensee has installed a shorting switch
modification on the control circuitry for one motor-operated valve
(MOV) in each division of the Drywell Spray system to reduce the
risk impact of a fire-induced multiple spurious operation that fails
both MOVs. In addition, the licensee will rely on fire protection
DID features such as administrative controls, plant design features,
fire protection inspections, installed fire detection and
suppression systems, and passive fire protection features. The
exemption does not modify plant operation because fire protection
for structures, systems, and components important to safe shutdown
continues to be provided, and fire damage continues to be limited so
that one of the redundant trains of Drywell Spray is free of fire
damage. Thus the exemption does
[[Page 21316]]
not result in a significant change in the types or amount of
effluents that may be released and does not result in any additional
occupational exposure. Therefore, the requirements of 10
CFR51.22(c)(9)(ii) and (iii) are met.
V. Conclusions.
Accordingly, the Commission has determined that, pursuant to 10
CFR 50.12, the exemption is authorized by law, will not present an
undue risk to the public health and safety, and is consistent with
the common defense and security. Also, special circumstances are
present in that application of the regulation is not necessary to
achieve the underlying purpose of the rule. Therefore, the
Commission hereby grants Northern States Power Company, doing
business as Xcel Energy, an exemption from the requirements of 10
CFR 50, Appendix R, Section III.G.2, for Monticello Nuclear
Generating Plant, to allow the use of a shorting switch to ensure
that one redundant train of Drywell Spray is free of fire damage to
achieve and maintain hot shutdown conditions in the event of a fire.
VI. Availability of Documents.
The documents identified in the following table are available in
ADAMS.
------------------------------------------------------------------------
Document ADAMS Accession No.
------------------------------------------------------------------------
Risk-Informed Request for Exemption from ML17096A599
10 CFR 50, Appendix R, III.G.2
Requirements for Multiple Spurious
Operations of Drywell Spray Motor-
Operated Valves.
Request for additional information RE: ML17293A091
Monticello Request for Exemption from
Appendix R Requirements (CAC NO.
MF9586; EPID L-2017-LLE-00012).
Response to Request for Additional ML17324B361
Information regarding Risk-Informed
Request for Exemption from 10 CFR 50,
Appendix R, III.G.2 Requirements for
Multiple Spurious Operations of Drywell
Spray Motor-Operated Valves (CAC No.
MF9586).
Monticello Nuclear Generating Plant ML14119A216
Triennial Fire Protection Inspection
Report 05000263/2014008.
Letter of Intent to Transition to 10 CFR ML053460342
50.48(c)--National Fire Protection
Association Standard NFPA 805.
``Performance-based Standards for Fire
Protection for Light Water Reactor
Electric Generating Plants.'' 2001
Edition.
Notice of Withdrawal of Letter of Intent ML102000433
to Transition to 10 CFR 50.48(c)''.
NRC Regulatory Issue Summary 2007-06 ML070650428
Regulatory Guide 1.200 Implementation.
NEI 05-04, Rev. 2 Process for Performing ML083430462
Internal Events PRA Peer Reviews Using
the ASME/ANS PRA Standard.
NEI 07-12 [REV 1] Fire Probabilistic ML102230070
Risk Assessment (FPRA) Peer Review
Process Guidelines.
NUREG/CR-7150, Vol. 2 Joint Assessment ML14141A129
of Cable Damage and Quantification of
Effects from Fire (JACQUE-FIRE).
Browns Ferry Nuclear Plant, Units 1, 2, ML15212A796
And 3--Issuance of Amendments Regarding
Transition to a Risk-Informed,
Performance-Based Fire Protection
Program in Accordance with 10 CFR
50.48(c) (CAC NOS. MF1185, MF1186, AND
MF1187).
Arkansas Nuclear One, Unit 1--Issuance ML16223A481
of Amendment Regarding Transition to a
Risk-Informed, Performance-Based Fire
Protection Program in Accordance with
10 CFR 50.48(c) (CAC NO. MF3419).
Regulatory Guide 1.189 ``Fire Protection ML092580550
for Nuclear Power Plants,'' Revision 2.
Regulatory Guide 1.174 ``An Approach for ML100910006
Using Probabilistic Risk Assessment in
Risk-Informed Decisions on Plant-
Specific Changes to the Licensing
Basis,'' Revision 2.
Regulatory Guide 1.200 ``An Approach for ML090410014
Determining the Technical Adequacy of
Probabilistic Risk Assessment Results
for Risk-Informed Activities,''
Revision 2.
Monticello Nuclear Generating Station: ML16047A273
Evaluation of Risk Significance of
Permanent Integrated Leak Rate Test
Extension.
------------------------------------------------------------------------
Dated at Rockville, Maryland, this 1st day of May 2018.
For the Nuclear Regulatory Commission.
Gregory F. Suber,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2018-09801 Filed 5-8-18; 8:45 am]
BILLING CODE 7590-01-P