[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