[Federal Register Volume 89, Number 148 (Thursday, August 1, 2024)]
[Notices]
[Pages 62806-62811]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2024-16978]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-400; NRC-2024-0125]
Duke Energy Progress, LLC; Shearon Harris Nuclear Power Plant,
Unit 1; Exemption
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice; issuance.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) has issued an
exemption in response to a February 6, 2024, request, as supplemented
by letters dated April 3 and June 7, 2024, from Duke Energy Progress,
LLC (the licensee). The exemption relieves the licensee from NRC
regulations requiring specific reactor protection system cables at
Shearon Harris Nuclear Power Plant, Unit 1, to meet certain
requirements of the Electrical and Electronics Engineers (IEEE)
Standard 279-1971, ``Criteria for Protection Systems for Nuclear Power
Generating Stations.''
DATES: The exemption was issued on July 26, 2024.
ADDRESSES: Please refer to Docket ID NRC-2024-0125 when contacting the
NRC about the availability of information regarding this document. You
may obtain publicly available information related to this document
using any of the following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2024-0125. Address
questions about Docket IDs in Regulations.gov to Stacy Schumann;
telephone: 301-415-0624; 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 ``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, at 301-415-4737,
or by email to [email protected]. The ADAMS accession number for
each document referenced in this document (if that document is
available in ADAMS) is provided the first time that it is mentioned in
this document.
NRC's PDR: The PDR, where you may examine and order copies
of publicly available documents, is open by appointment. To make an
appointment to visit the PDR, please send an email to
[email protected] or call 1-800-397-4209 or 301-415-4737, between 8
a.m. and 4 p.m. eastern time (ET), Monday through Friday, except
Federal holidays.
FOR FURTHER INFORMATION CONTACT: Michael Mahoney, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-3867; email: [email protected].
SUPPLEMENTARY INFORMATION: The text of the exemption is attached.
Dated: July 29, 2024.
[[Page 62807]]
For the Nuclear Regulatory Commission.
Michael Mahoney,
Project Manager, Plant Licensing Branch II-2, Division of Operating
Reactor Licensing, Office of Nuclear Reactor Regulation.
Attachment: Exemption
Nuclear Regulatory Commission
Docket No. 50-400
Duke Energy Progress, LLC
Shearon Harris Nuclear Power Plant, Unit 1
Exemption
I. Background
Duke Energy Progress, LLC (Duke Energy, the licensee) is the holder
of Renewed Facility Operating License No. NPF-63, which authorizes
operation of Shearon Harris Nuclear Power Plant, Unit 1 (Harris). 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) now or hereafter in effect. The facility consists of a
pressurized-water reactor located in Wake and Chatham Counties, North
Carolina.
II. Request/Action
By application dated February 6, 2024 (Agencywide Documents Access
and Management System (ADAMS) Accession No. ML24037A284), as
supplemented by letters dated April 3, 2024 (ML24094A105), and June 7,
2024 (ML24159A746), Duke Energy, pursuant to Title 10 of the Code of
Federal Regulations (10 CFR) Section 50.12, ``Specific exemptions,''
requested an exemption from a provision in the Institute of Electrical
and Electronics Engineers (IEEE) Standard (std) 279-1971, ``Criteria
for Protection Systems for Nuclear Power Generating Stations,'' that is
required by CFR 50.55a(h)(2), ``Protection systems,'' for Harris.
Specifically, the exemption request would remove the requirement for
the Harris reactor protection system (RPS) cables that terminate within
turbine control system (TCS) Cabinet G (1TCS-CAB-G) to be independent
and physically separated in accordance with IEEE 279-1971, Section 4.6,
``Channel Independence.'' The licensee stated that application of the
regulation in this circumstance would not serve the underlying purpose
of the rule and is not necessary to achieve the underlying purpose of
the rule. The exemption request was submitted for review under the
NRC's Risk-Informed Process for Evaluations (RIPE).
III. Discussion
The regulations in 10 CFR 50.55a(h)(2) state:
For nuclear power plants with construction permits issued after
January 1, 1971, but before May 13, 1999, protection systems must
meet the requirements in IEEE Std 279-1968, ``Proposed IEEE Criteria
for Nuclear Power Plant Protection Systems,'' or the requirements in
IEEE Std 279-1971, ``Criteria for Protection Systems for Nuclear
Power Generating Stations,'' or the requirements in IEEE Std 603-
1991, ``Criteria for Safety Systems for Nuclear Power Generating
Stations,'' and the correction sheet dated January 30, 1995. For
nuclear power plants with construction permits issued before January
1, 1971, protection systems must be consistent with their licensing
basis or may meet the requirements of IEEE Std. 603-1991 and the
correction sheet dated January 30, 1995.
Duke Energy requested an exemption from IEEE 279-1971, Section 4.6,
as required by 10 CFR 50.55a(h)(2), for specific RPS cables at Harris.
Contrary to the requirements in IEEE 279-1971, Section 4.6, the safety-
related RPS cables that terminate within TCS Cabinet G are not
independent and physically separated from the non-safety-related TCS
cables. The licensee requested the exemption in order to maintain the
current configuration of the TCS circuitry at Harris.
Pursuant to 10 CFR 50.12, the NRC may, upon application by any
interested person or upon its own initiative, grant exemptions from
requirements of 10 CFR part 50 when: (1) the exemptions are authorized
by law, will not present an undue risk to the public health and safety,
and are consistent with the common defense and security, and (2)
special circumstances, as defined in 10 CFR 50.12(a)(2), are present.
The licensee states that the special circumstances associated with its
exemption request are that the ``application of the regulation in this
circumstance would not serve the underlying purpose of the rule and is
not necessary to achieve the underlying purpose of the rule.''
The exemption request was submitted for review under the RIPE,
which is described in the NRC's ``Guidelines for Characterizing the
Safety Impact of Issues,'' Revision 2 (referenced henceforth as SIC)
(ML22088A135). The Office of Nuclear Reactor Regulation (NRR) temporary
staff guidance (TSG) document TSG-DORL-2021-01, Revision 3
(ML23122A014), provides the framework and guidance for the staff to
implement the streamlined processing of exemption requests from NRC
requirements submitted under RIPE. Use of RIPE for exemption requests
is limited to issues for which the safety impact can be modeled using
probabilistic risk assessment (PRA) and shown to have a minimal safety
impact per SIC. RIPE is based on the application of pre-existing risk-
informed criteria that allows for the staff's review and disposition of
the submittal to be streamlined and efficient.
As described in the SIC, all the following must apply in order to
characterize an issue as having a minimal safety impact and qualify for
consideration under the RIPE:
The issue contributes less than 1 x 10-\7\/year
to core damage frequency (CDF);
The issue contributes less than 1 x 10-\8\/year
to large early release frequency (LERF);
The issue has no safety impact or minimal safety impact in
accordance with the SIC; and
Cumulative risk is assessed based on plant-specific CDF
and LERF. Cumulative risk is acceptable for the purposes of this
guidance if baseline risk remains less than 1 x 10-\4\/year
for CDF and less than 1 x 10-\5\/year for LERF once the
impact of the proposed change is incorporated into baseline risk.
RIPE exemption requests must also include defense-in-depth (DID)
and safety margin considerations assessed by the integrated decision-
making panel (IDP).
Requests for changes made under the RIPE are reviewed by the NRC
staff in a manner consistent with the principles of risk-informed
decision-making outlined in Regulatory Guide 1.174, Revision 3, ``An
Approach for Using Probabilistic Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to the Licensing Basis''
(ML17317A256), which includes ensuring that the proposed change is
consistent with DID philosophy, maintains sufficient safety margins, is
consistent with the Commission's Safety Goal Policy Statement, and
includes performance monitoring strategies.
Conformance With the RIPE Minimal Safety Impact Criteria
The licensee considered the RIPE screening questions contained in
the SIC and concluded that the requested exemption would not have a
more than minimal impact on safety. Considerations for each of the five
screening questions are discussed below.
1. Does the issue result in an adverse impact on the frequency of
occurrence of an accident initiator or result in a new accident
initiator?
[[Page 62808]]
In Section 4.4 of the exemption request, the licensee states that
the issue does not result in an adverse impact on the frequency of
occurrence of an accident initiator or result in a new accident
initiator because the cables impacted by the issue are associated with
the solid state protection system (SSPS), which provides the logic to
develop reactor trip and emergency safety feature actuation signals
(ESFAS). The licensee also states that the SSPS provides a mitigation
function and does not initiate an accident or create a new accident
initiator.
The NRC staff reviewed the licensee's consideration of this
screening element and concluded that the issue does not adversely
impact the frequency of occurrence of an accident initiator or result
in a new accident initiator because the SSPS provides a mitigation
function and does not initiate an accident.
2. Does the issue result in an adverse impact on the availability,
reliability, or capability of structures, systems, or components (SSCs)
or personnel relied upon to mitigate a transient, accident, or natural
hazard?
In Section 4.4 of the exemption request, the licensee states that
the issue does not result in an adverse impact on the availability,
reliability, or capability of SSCs or personnel relied upon to mitigate
a transient, accident, or natural hazard because the safety-related
protection trains will remain fully capable of performing their
intended functions. The licensee's conclusion is based on an evaluation
that reviewed potential sources of electrical anomalies and the
mitigation techniques used to reduce the probability of an event
occurring that could impact plant equipment. The electrical anomaly
evaluation is described in Section 4.1 of the exemption request and
included evaluation of the cabinet design, cabinet location, electrical
grounding, power source design, signal attenuation due to cable length,
equipment qualification, cable routing, previous testing of low-level
instrument wiring, plant operating experience, and the requirements in
IEEE 384-1974, ``IEEE Trial-Use Standard Criteria for Separation of
Class 1E Equipment and Circuits.'' The evaluation concludes that there
are no credible electrical anomaly events which could impact either
train of safety-related equipment from performing its design basis
function.
The license stated that the turbine trip logic connects to the SSPS
and RPS through four control relays that use redundant contacts from
the reactor trip breaker. In addition, the licensee stated the reactor
trip breaker auxiliary contacts provide indication of a reactor trip to
the turbine trip system (TTS) and that an open or short of the contacts
used for the non-safety related portion of the circuit would not
prevent a reactor trip from occurring, if required, because the
auxiliary contacts are not in the direct electrical path of the reactor
trip breakers. The cables and conduits for each of these circuits
follow the separation criteria requirements except for Terminal Box B
and TCS Cabinet G. The isolation between the TCS and the RPS/SSPS
trains is provided in the RPS and SSPS cabinets.
In the exemption request, the licensee stated that if a short
circuit were to occur, the impact would be limited to a single train of
the TTS and that multiple shorts would be needed to impact both TTS
trains. In its supplement dated June 7, 2024, the licensee stated:
A fault of the TTS cables could impact the non-safety-related
automatic turbine trip on reactor trip function. For example, a
fault could cause a short circuit which could bypass the SSPS
turbine trip output relay contacts, thus preventing the turbine from
tripping. If this were to occur and a reactor trip occurred,
Operations would trip the turbine manually by the Main Control Board
turbine trip switch per step 2 of [Harris] Emergency Operating
Procedure EOP-E-0, ``REACTOR TRIP OR SAFETY INJECTION.''
Under 10 CFR 50.62(c)(1), each pressurized-water reactor must have
equipment, from sensor output to final actuation device, that is
diverse from the reactor trip system, to automatically initiate the
auxiliary feedwater system and initiate a turbine trip under conditions
indicative of an anticipated transient without scram (ATWS). Harris
complied with this requirement by installing ATWS mitigation system
actuation circuitry (AMSAC). The NRC staff notes that AMSAC would
remain available to trip the turbine if an ATWS were to occur.
The NRC staff reviewed the licensee's consideration of this
screening element and determined that an adverse impact to the
availability, reliability, or capability of SSCs relied upon to
mitigate a transient, accident, or natural hazard exists because the
separation and channel independence requirements of IEEE 279-1971 are
not met in TCS Cabinet G. However, the licensee's evaluation of the TCS
circuitry demonstrates that, while the exemption would rely on non-
safety-related equipment to prevent potential electrical anomalies from
propagating to safety-related components, the TCS design is robust and
configured such that any electrical perturbations are unlikely. Should
an electrical short condition result in failure of an automatic turbine
trip, pre-existing procedurally directed operator actions are available
to manually initiate the required turbine trip.
The NRC staff concluded that the adverse impact of not meeting the
separation and channel independence requirements of IEEE 279-1971,
Section 4.6, for the RPS cables that terminate within TCS Cabinet G on
the availability, reliability, or capability of SSCs or personnel
relied upon to mitigate a transient, accident, or natural hazard is not
more than minimal because (1) the design of the TCS ensures it is
unlikely that an electrical anomaly event could occur that would
prevent either train of safety-related equipment from performing its
design basis function, (2) not meeting separation and channel
independence requirements would not impact the reactor trip breakers
because the turbine trip logic is not directly electrically connected
the reactor trip breakers, and (3) operator actions and AMSAC would
remain available to trip the turbine in the unlikely event that a fault
prevented the turbine trip from occurring automatically.
3. Does the issue result in an adverse impact on the consequences
of an accident sequence?
In Section 4.4 of the exemption request, the licensee stated that
the issue does not affect the safety-related design functions of the
SSPS or RPS. The licensee also states the design function of the SSPS
to mitigate an accident is not impacted and therefore the consequences
of any accident previously evaluated are not impacted. In its
supplement dated June 7, 2024, the licensee stated that a fault of the
TTS cables could impact the non-safety-related automatic turbine trip
on reactor trip function, but procedurally directed operator actions
would remain available to manually trip the turbine if needed.
The NRC staff reviewed the licensee's consideration of this
screening element and concluded that the proposed exemption does not
result in an adverse impact on the consequences of an accident because
the proposed exemption does not prevent the ability of the safety-
related systems to perform their design functions.
4. Does the issue result in an adverse impact on the capability of
a fission product barrier?
In Section 4.4 of the exemption request, the licensee stated that
the issue does not affect operating limits, the fuel, reactor coolant
system (RCS), or modify the containment boundary in any way. The cables
are located outside the containment building and do not result in
revising or challenging a design
[[Page 62809]]
basis limit for a fission product barrier (i.e., numerical limiting
value for controlling the integrity of the fuel cladding, reactor
coolant pressure boundary, and/or containment) as described in the
Updated Final Safety Analysis Report. Furthermore, the licensee stated
the proposed exemption does not prevent the ability of the safety-
related systems to perform their design functions.
The NRC staff reviewed the licensee's consideration of this
screening element and concluded that the proposed exemption does not
result in an adverse impact on the capability of a fission product
barrier because the proposed exemption does not prevent the ability of
safety-related systems, including RCS and containment, to perform their
design functions or alter any design-basis limits.
5. Does the issue result in an adverse impact on DID capability or
impact in safety margin?
In Section 4.4. of the exemption request, the licensee stated that
there is no adverse impact on DID and safety margins because there are
no credible events that would prevent both trains of safety-related
equipment from fulfilling their design-basis functions. The licensee's
conclusion is based on an evaluation of the potential for electrical
anomalies described in Section 4.1 of the exemption request, which
included evaluation of the cabinet design, cabinet location, electrical
grounding, power source design, signal attenuation due to cable length,
equipment qualification, cable routing, previous testing of low-level
instrument wiring, plant operating experience, and the requirements in
IEEE 384-1974. The evaluation concluded that there are no credible
electrical anomaly events which could impact either train of safety-
related equipment from performing its design-basis function.
The licensee stated that, based on the evaluation that established
there are no credible events that would impact both trains of safety-
related equipment from performing its design-basis function, the key
aspects of IEEE 279-1971 for single failure criterion and channel
integrity are maintained. The licensee also stated that while the
common connection for the ``A'' and ``B'' trains in the TCS does
challenge the channel independence requirement of IEEE 279-1971,
Section 4.6, there is not a credible reduction in the ability of the
safety-related systems to perform their intended design functions. The
licensee further stated that exemption to the IEEE 279-1971, Section
4.6, requirement will not impact the ability of the safety-related
protection trains to remain fully capable of performing their intended
design functions in generating the signals associated with actuating
reactor trip and engineered safeguards, as required by IEEE 279-1971.
In its response to screening question 2, the licensee stated that
the turbine trip logic connects to the SSPS and RPS through four
control relays that use redundant contacts from the reactor trip
breaker and that an open or short of the contacts used for the non-
safety related portion of the circuit would not prevent a reactor trip
from occurring, if required, because the auxiliary contacts are not in
the direct electrical path of the reactor trip breakers. The licensee
also stated that the isolation between the TCS and the RPS/SSPS trains
is provided in the RPS and SSPS cabinets. Further, the licensee stated
that if a short circuit were to occur, the impact would be limited to a
single train and the ability to trip the turbine would not be lost. In
its supplement dated June 7, 2024, the licensee stated that a fault of
the TTS cables could impact the non-safety-related automatic turbine
trip on reactor trip function, but procedurally directed operator
actions would remain available to manually trip the turbine if needed.
In addition, the NRC staff notes that AMSAC would remain available to
trip the turbine if an ATWS were to occur, such as due to multiple
shorts occurring (which is outside of the single failure proof design
criteria).
The NRC staff reviewed the licensee's consideration of this
screening element and determined that the licensee describes a
potential adverse impact to DID and safety margins because the channel
independence requirements of IEEE 279-1971 are not met in TCS Cabinet
G. However, the licensee's evaluation of the TCS circuitry demonstrates
that, while the exemption would rely on non-safety-related equipment to
prevent potential electrical anomalies from propagating to safety-
related components, the TCS design is robust and configured such that
any electrical perturbations are unlikely. In the unlikely event that
an electrical condition results in failure of an automatic turbine
trip, procedurally directed operator actions are available to manually
trip the turbine. The use of pre-existing procedurally controlled
operator actions to provide diversity and DID for this unlikely
scenario does not result in the over-reliance on programmatic measures.
The NRC staff concluded that the adverse impact of not meeting the
separation and channel independence requirements of IEEE 279-1971,
Section 4.6, for the RPS cables that terminate within TCS Cabinet G on
DID capability and safety margins is not more than minimal because (1)
the design of the TCS ensures it is unlikely that an electrical anomaly
event could occur that would prevent either train of safety-related
equipment from performing its design-basis functions, (2) not meeting
channel independence requirements would not impact the reactor trip
breakers because the turbine trip logic is not directly electrically
connected the reactor trip breakers, and (3) operator actions and AMSAC
would remain available to trip the turbine in the unlikely event that a
fault prevented the turbine trip from occurring automatically.
Implementation of an IDP
The licensee has been approved to adopt 10 CFR 50.69, ``Risk-
informed categorization and treatment of structures, systems and
components for nuclear power reactors,'' by license amendment No. 174,
issued September 17, 2019 (ML19192A012), as revised by license
amendment No. 188, issued January 19, 2022 (ML21316A248). The licensee
established a multi-disciplinary IDP to evaluate the proposed exemption
request. The IDP membership included personnel from site engineering,
operations, PRA, safety analysis, and licensing. Therefore, the NRC
staff concludes that Harris used an acceptable IDP in support of the
proposed exemption request per the RIPE guidance in TSG-DORL-2021-01.
Use of an Acceptable/Approved PRA Model
Harris has adopted risk-informed initiative Technical
Specifications Task Force (TSTF) traveler TSTF-505, Revision 2,
``Provide Risk-Informed Extended Completion Times--RITSTF Initiative
4b,'' for the risk-informed completion time program, as approved by
license amendment No. 184, issued April 2, 2021 (ML21047A314). The
Harris PRA model used to support the risk-informed completion time
license amendment includes internal events, internal flooding, and fire
hazards. The Harris PRA model does not include high winds, external
flooding, or seismic hazards due to meeting screening criteria as part
of the approval of its risk-informed completion time license amendment.
There are no concerns in this exemption request specifically related to
high winds, external flooding, or seismic hazards. Implementation of
the TSTF-505 license amendment and associated license conditions have
been completed. Therefore, the NRC staff concludes that Harris used a
technically acceptable PRA model in support of the
[[Page 62810]]
proposed exemption request per the RIPE guidance in the SIC.
Evaluation of PRA Results
The licensee quantitatively assessed the risk significance of
maintaining the current TCS circuitry at Harris with the proposed
exemption using a surrogate to represent the potential for a hot short
to fail the ability of (1) the turbine to trip upon a reactor trip, (2)
the reactor to trip upon a valid RPS signal, and (3) the ESFAS to
actuate upon a valid actuation. The surrogate basic event was applied
in the logic model where turbine trips, RPS signal failures, and ESFAS
actuations were modeled. The surrogate basic event probability was
based on the conditional probability of a hot short to occur during a
conservative selection of fires that impact either train of SSPS. The
risk results were 1.6 x 10-\8\/year for CDF and less than 1
x 10-\10\/year for LERF. These results satisfy the RIPE
criteria of contributing less than 1 x 10-\7\/year to CDF
and 1 x 10-\8\/year to LERF. Cumulative risk results were
4.1 x 10-\5\/year for CDF and 3.5 x 10-\6\/year
for LERF. Therefore, cumulative risk for Harris remains less than the
RIPE criteria of 1 x 10-\4\/year for CDF and 1 x
10-\5\/year for LERF. The NRC staff concludes that these
results satisfy the RIPE criteria for a minimal increase in risk for
the proposed exemption.
Evaluation of the Need for Risk Management Actions
Evaluation of the RIPE screening questions and the PRA results
confirm that the proposed exemption results in a minimal safety impact.
For these results, the SIC guidance states that risk management actions
must be considered to offset the risk increase for the NRC staff to
review under RIPE. Section 4.3 of the exemption request states that a
review of industry operating experience related to the issue did not
identify any specific modifications necessary to assess and/or bound
the impact of the issue on quantitative risk. Therefore, the NRC staff
concluded that no risk management actions were identified or required.
Assessment of Performance Monitoring Strategies
Section 4.1 of the exemption request states that the TSC was
upgraded in 2018 but the cables in question have not been moved since
original plant construction. Both the previous and current designs
energize the control circuits continuously so that a loss of power
would result in a turbine trip. The previous design tested the circuit
quarterly. The current design cycles the control relays weekly, and
this test has been performed once a week for over 5 years. There have
been no instances of spurious control circuit anomalies attributed to
the TCS trip relays cycling on and off.
The NRC staff concluded that the existing performance monitoring
strategies will ensure no deficiencies exist that would challenge the
conclusions in the licensee's evaluation of the proposed exemption.
A. The Exemption is Authorized by Law
The NRC has the authority under 10 CFR 50.12 to grant exemptions
from the requirements of Part 50 upon demonstration of proper
justification. The licensee has requested an exemption to the
requirement in 10 CFR 50.55a(h)(2) requiring protection systems meet
the requirements of IEEE 279-1971, Section 4.6, for safety-related RPS
cables that terminate within TCS Cabinet G. As discussed below, the NRC
staff determined that special circumstances exist, which support
granting the proposed exemption. Furthermore, granting the exemption
would not result in a violation of the Atomic Energy Act of 1954, as
amended, or the NRC's regulations. Therefore, the exemption is
authorized by law.
B. The Exemption Presents No Undue Risk to Public Health and Safety
The NRC staff has concluded that the exemption represents low risk,
is of minimal safety impact, and that adequate DID and safety margins
are preserved. The NRC staff concluded that the licensee's submittal
demonstrates that the design of the TCS is robust against electrical
failures that would prevent the RPS from performing their intended
functions with the proposed exemption. Thus, granting this exemption
request will not pose undue risk to public health and safety.
C. The Exemption Is Consistent With the Common Defense and Security
The NRC staff has evaluated the licensee's exemption request and
concluded that the licensee's submittal demonstrates that the design of
the TCS is robust against electrical failures that would prevent the
RPS from performing their intended functions with the proposed
exemption. The NRC staff also concluded that adequate DID and safety
margins will be preserved with the requested exemption. Further, the
exemption does not involve security requirements and does not create a
security risk. Therefore, the exemption is consistent with the common
defense and security.
D. Special Circumstances
Special circumstances, in accordance with 10 CFR 50.12(a)(2)(ii),
are present whenever application of the regulation in the circumstances
would not serve the purpose of the rule or is not necessary to achieve
the purpose of the rule. The licensee has requested a limited scope
exemption from 10 CFR 50.55a(h)(2) that would only apply to the RPS
cables that terminate within TCS Cabinet G. Specifically, the exemption
request would remove the requirement for the RPS cables that terminate
within TCS Cabinet G to be independent and physically separated in
accordance with IEEE 279-1971, Section 4.6. The underlying purpose of
IEEE 279-1971, Section 4.6, is to ensure the capability of the safety-
related system to accomplish its safety function during normal and
accident conditions and reduce the likelihood of interactions between
channels during maintenance operations or in the event of a channel
malfunction.
The licensee has supported that the design of the TCS is adequate
to ensure that the lack of independence and physical separation between
TCS and RPS cables in TCS Cabinet G is unlikely to prevent either
system from being able to perform their intended functions. In
addition, the licensee has also demonstrated that adequate DID and
safety margins will be preserved with the requested exemption. For
these reasons, the NRC staff finds that for this limited scope
exemption to the requirements of 10 CFR 50.55a(h)(2) for the safety-
related RPS cables that terminate within TCS Cabinet G, application of
the regulation in the particular circumstances is not necessary to
achieve the underlying purpose of the rule.
E. Environmental Considerations
The exemption requested by the licensee includes changes to
requirements with respect to installation or use of a facility
component located within the restricted area. The NRC staff determined
that the exemption meets the eligibility criteria for the categorical
exclusion set forth in 10 CFR 51.22(c)(9) because the granting of this
exemption involves: (i) no significant hazards consideration, (ii) no
significant change in the types or a 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
[[Page 62811]]
NRC's consideration of this exemption request. The basis for the NRC
staff's determination of each of the requirements in 10 CFR 51.22(c)(9)
is discussed below.
Requirements in 10 CFR 51.22(c)(9)(i)
The NRC staff evaluated the issue of no significant hazards
consideration using the standards described in 10 CFR 50.92(c), as
presented below:
1. Does the proposed exemption involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design of the TCS is robust against electrical failures that
would prevent the RPS from performing their intended functions with the
proposed exemption and does not modify how the plant is operated. The
proposed exemption does not affect any plant protective boundaries,
cause a release of fission products to the public, or alter the
performance of any SSCs important to safety.
Therefore, the proposed exemption does not result in a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed exemption create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The design of the TCS is robust against electrical failures that
would prevent the RPS from performing their intended functions with the
proposed exemption and does not modify how the plant is operated. In
addition, the TTS and RPS provide mitigation functions and do not
initiate accidents or create a new accident initiators.
Therefore, the proposed exemption does not create the possibility
of a new or different kind of accident from any previously evaluated.
3. Does the proposed exemption involve a significant reduction in a
margin of safety?
Response: No.
The design of the TCS is robust against electrical failures that
would prevent the RPS from performing their intended functions with the
proposed exemption and does not modify how the plant is operated. The
proposed exemption does not alter any setpoints for protective actions,
change the initial conditions for any accidents, or alter the
requirements of any SSCs important to safety.
Therefore, the proposed exemption does not involve a significant
reduction in a margin of safety.
The NRC staff concludes that the proposed exemption presents no
significant hazards consideration under the standards set forth in 10
CFR 50.92(c), and, accordingly, a finding of no significant hazards
consideration is justified (i.e., satisfies the provision of 10 CFR
51.22(c)(9)(i)).
Requirements in 10 CFR 51.22(c)(9)(ii)
The design of the TCS is robust against electrical failures that
would prevent the RPS from performing their intended functions with the
proposed exemption and does not modify how the plant is operated. The
proposed exemption does not alter any setpoints for protective actions,
change the initial conditions for any accidents, or alter the
requirements of any SSCs important to safety. The proposed exemption
will not significantly change the types or amounts of effluents that
may be released offsite. Therefore, the staff finds that the provision
of 10 CFR 51.22(c)(9)(ii) is satisfied.
Requirements in 10 CFR 51.22(c)(9)(iii)
The licensee's request supported that the exemption had either no
or a minimal safety impact for all accident initiator categories and
the NRC staff has concluded that the proposed exemption will not result
in an adverse impact on the frequency of existing accident initiators
or result in new accident initiators. The proposed exemption will not
significantly increase individual occupational radiation exposure, or
significantly increase cumulative public or occupational radiation
exposure. Therefore, the staff finds that the provision of 10 CFR
51.22(c)(9)(iii) is satisfied.
The NRC staff concludes that the proposed exemption meets the
eligibility criteria for the categorical exclusion set forth in 10 CFR
51.22(c)(9). 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 proposed granting of this
exemption.
IV. 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.
Therefore, the Commission hereby grants Duke Energy an exemption from
IEEE 279-1971, Section 4.6, as required by 10 CFR 50.55a(h)(2), for the
safety-related RPS cables at Harris that terminate within TCS Cabinet
G.
Dated: July 29, 2024.
For the Nuclear Regulatory Commission.
Michael Mahoney,
Project Manager, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2024-16978 Filed 7-31-24; 8:45 am]
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