ML18249A019

Issuance of Amendment Nos. 341 and 323 Technical Support Center Relocation

ML18249A019
Person / Time
Site:	Cook
Issue date:	11/13/2018
From:	Dietrich A Plant Licensing Branch III
To:	Gebbie J Indiana Michigan Power Co
Dietrich A, 415-3826
References
EPID L-2017-LLA-0375
Download: ML18249A019 (26)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 13, 2018 Mr. Joel P. Gebbie Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, Ml 49106

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 341 and 323 RE: TECHNICAL SUPPORT CENTER RELOCATION (EPID L-2017-LLA-0375)

Dear Mr. Gebbie:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 341 to Renewed Facility Operating License No. DPR-58 and Amendment No. 323 to Renewed Facility Operating License No. DPR-7 4 for the Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, respectively. The amendments consist of changes to the license in response to your application dated November 7, 2017 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML17317A454 ), as supplemented by letters dated January 19, 2018 (ADAMS Accession No. ML18024A430), and August 14, 2018 (ADAMS Accession No. ML18232A063).

The amendments consist of changes to the Emergency Plan for the Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2. The amendments revise the Emergency Plan to move the Technical Support Center to a different location in a new facility located within the existing protected area.

A copy of the related safety evaluation is also enclosed. A notice of issuance will be included in the Commission's biweekly Federal Register notice.

Sin~~

Allison W. Dietrich, Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316

Enclosures:

1. Amendment No. 341 to DPR-58

2. Amendment No. 323 to DPR-74

3. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 341 License No. DPR-58

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Indiana Michigan Power Company (the licensee) dated November 7, 2017, as supplemented by letters dated January 19, 2018, and August 14, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, by Amendment No. 341, Renewed Facility Operating License No. DPR-58 is hereby amended to authorize revision to the Emergency Plan for the Donald C. Cook Nuclear Plant, Unit No. 1, as set forth in Indiana Michigan Power Company's application dated November 7, 2017, as supplemented by letters dated January 19, 2018, and August 14, 2018, and evaluated in the NRC staff's safety evaluation dated November 13, 2018.

Enclosure 1

3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION

~ J,, ~ , ,_,__

Ho K. Nieh, Director Office of Nuclear Reactor Regulation Date of Issuance: November 13, 2018

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 323 License No. DPR-74

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Indiana Michigan Power Company (the licensee) dated November 7, 2017, as supplemented by letters dated January 19, 2018, and August 14, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;

8. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, by Amendment No. 323, Renewed Facility Operating License No. DPR-74 is hereby amended to authorize revision to the Emergency Plan for the Donald C. Cook Nuclear Plant, Unit No. 2, as set forth in Indiana Michigan Power Company's application dated November 7, 2017, as supplemented by letters dated January 19, 2018, and August 14, 2018, and evaluated in the NRC staff's safety evaluation dated November 13, 2018.

Enclosure 2

3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION Ho K. Nieh, Director Office of Nuclear Reactor Regulation Date of Issuance: November 13, 2018

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 341 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 323 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316

1.0 INTRODUCTION

By application dated November 7, 2017 (Reference 1), as supplemented by letters dated January 19, 2018 (Reference 2), and August 14, 2018 (Reference 3), Indiana Michigan Power Company (the licensee) requested changes to the Emergency Plan for the Donald C. Cook Nuclear Plant (CNP), Unit Nos. 1 and 2.

The supplemental letters dated January 19, 2018, and August 14, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 2, 2018 (83 FR 169).

The proposed changes would revise the CNP Emergency Plan to move the Technical Support Center (TSC) to a different location in a new facility located within the existing protected area. A TSC is a facility from which the licensee staff provides plant management and technical support to plant operations personnel during emergency conditions, relieves the reactor operators of peripheral duties and communications not directly related to reactor system manipulations, prevents congestion in the control room (CR), and is able to perform the functions of the emergency operations facility (EOF) until the EOF is staffed and ready to respond, if needed.

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulatory requirements and guidance during its review of the proposed changes.

Enclosure 3

2.1 Regulations Section 50.47, "Emergency plans," of Title 10 of the Code of Federal Regulations (10 CFR),

sets forth emergency plan requirements for nuclear power plant facilities. The regulations in 10 CFR 50.47(b) establish the standards that the onsite and offsite emergency response plans must meet for the NRC staff to make a finding that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Planning standard 10 CFR 50.47(b)(8) specifically requires that adequate emergency facilities and equipment to support the emergency response are provided and maintained.

The regulations in 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 19, "Control Room," require that:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem [roentgen equivalent man] whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided ( 1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Additionally, 10 CFR Part 50, Appendix A, Criterion 19, states:

[H]olders of operating licenses using an alternative source term under§ 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv [Sievert] (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 for the duration of the accident.

The regulations in 10 CFR Part 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," paragraph IV, include requirements for the content of emergency plans. The regulation in 10 CFR Part 50, Appendix E, paragraph IV.E.8.a(i),

specifies the requirement for a licensee onsite TSC from which effective direction can be given and effective control can be exercised during an emergency.

The regulation in 10 CFR 50.67, "Accident source term," establishes acceptance criteria for design-basis accident radiological analyses. The regulations in 10 CFR 50.67 state that the applicant's analysis must demonstrate with reasonable assurance that: (1) an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE; (2) an individual located at any point on the outer boundary of the low population zone who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose in excess of 0.25 Sv (25 rem) TEDE; and (3) adequate radiation protection is provided to permit

access to and occupancy of the CR under accident conditions, without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.

2.2 Regulatory Guidance Regulatory Guide (RG) 1.101, Revision 2 (Reference 4), endorses Revision 1 to NUREG-0654/FEMA-REP-1 (Reference 5). NUREG-0654, as amended, provides specific acceptance criteria in Section 11.H, "Emergency Facilities and Equipment," for complying with the standards set forth in 10 CFR 50.47(b)(8). NUREG-0654,Section II, "Planning Standards and Evaluation Criteria," Evaluation Criterion 11.H.1, states:

Each licensee shall establish a technical support center and onsite operations support center (assembly area) in accordance with NUREG-0696.

NUREG-0696 (Reference 6) describes the facilities and systems to be used by nuclear power plant licensees to improve responses to emergencies. NUREG-0696 also establishes criteria for the NRC staff to use in evaluating whether an applicant/licensee has met the requirements of paragraph IV.E.8 of Appendix E to 10 CFR Part 50 and General Design Criteria 19, "Control Room," in Appendix A to 10 CFR Part 50. NUREG-0696, Section 2.6, "Habitability," states:

Since the TSC is to provide direct management and technical support to the control room during an accident, it shall have the same radiological habitability as the control room under accident conditions. TSC personnel shall be protected from radiological hazards, including direct radiation and airborne radioactivity from inplant sources under accident conditions, to the same degree as control room personnel. Applicable criteria are specified in General Design Criterion 19; Standard Review Plan 6.4; and NUREG-0737, "Clarification of TMI Action Plan Requirements," Item 11.B.2.

NUREG-0737, Supplement 1 (Reference 7), Section 8.2, "Technical Support Center," provides additional guidance on the requirements for emergency response facilities.

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (Reference 8), provides guidance to the NRC staff in performing safety reviews of construction permit or operating license applications.

NUREG-0800 SRP Section 2.3.4 (Reference 9) addresses atmospheric dispersion models used to calculate atmospheric dispersion factors (x/Q) for postulated accidental radioactive and hazardous airborne releases.

NUREG-0800 SRP Section 6.4 (Reference 10) covers the review of CR ventilation and the CR layout to ensure that plant operators are adequately protected against the effects of accidental releases of toxic and radioactive gases.

RG 1.23 (Reference 11) provides guidance concerning criteria for the onsite meteorological measurements program for the collection of basic meteorological data needed to support plant licensing and operation.

RG 1.194 (Reference 12) provides guidance on determining x/Q values in support of design-basis CR radiological habitability assessments at nuclear power plants.

RG 1.140 (Reference 13) specifies the acceptance criteria for testing normal filtration systems and is the applicable guidance for the TSC filtration system.

RG 1.183 (Reference 14) provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms (ASTs); the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. RG 1.183 establishes an acceptable AST and identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. RG 1.183 also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

Amendment Nos. 332 and 314 to Renewed Facility Operating License Nos. DPR-58 and DPR-74, respectively, for CNP, Unit Nos. 1 and 2 (Reference 15), authorized the adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-490, Revision O (Reference 16),

and the implementation of a full-scope AST.

3.0 TECHNICAL EVALUATION

In its review, the NRC staff considered all of the regulatory requirements and guidance documents cited above. In particular, the NRC staff reviewed the proposed TSC against the following evaluation criteria provided in Section 2, "Technical Support Center," of NUREG-0696:

• Function (Section 2.1)

• Location (Section 2.2)

• Staffing and Training (Section 2.3),

• Size (Section 2.4)

• Structure (Section 2.5)

• Habitability (Section 2.6)

• Communications (Section 2. 7)

• Instrumentation, Data System Equipment and Power Supplies (Section 2.8)

• Technical Data and Data System (Section 2.9)

• Records Availability and Management (Section 2.10)

The NRC staff also reviewed the proposed CNP TSC against the criteria provided in Section 8.2 of Supplement 1 to NUREG-0737.

3.1 Function The guidance in Section 2.1 of NUREG-0696 specifies that the TSC should be able to fulfill the following functions:

• Provide plant management and technical support to plant operations personnel during emergency conditions;

• Relieve the reactor operators of peripheral duties and communications not directly related to reactor system manipulations;

• Prevent congestion in the CR; and

• Perform EOF functions following the declaration of an Alert, Site Area Emergency, or General Emergency classification until the EOF is functional.

The application states that when activated, the TSC functions include:

• Support for the affected CR's emergency response efforts;

• Continued evaluation of event classification;

• Assessment of the plant status and potential offsite impact;

• Coordination of emergency response actions within the Protected Area;

• Communication with the NRC via the Emergency Notification System; and

• Activation of the Emergency Response Data System (ERDS) 1 or ensuring it is activated.

The licensee's submittal states that the proposed TSC will continue to provide plant management and technical support to plant operations personnel during emergency conditions and relieve operations staff of peripheral duties and communications not directly related to system manipulations. Additionally, the TSC, in support of the CR, can perform the functions of the EOF until the EOF is activated, if needed. Performance of EOF functions by the TSC is not expected and would occur only if the EOF was somehow delayed in its activation or became unavailable.

As part of its evaluation, the NRC staff physically walked down the facility on July 16 and July 17, 2018, to verify the physical size, layout, and capabilities of the proposed CNP TSC to effectively provide plant management and technical support to plant operations during an emergency. Additionally, the staff observed a drill performed by the licensee on July 17, 2018, to validate the capabilities of the proposed TSC to perform its functions as described above.

The NRC staff has determined, based on its review of the licensee's submittal, and the NRC staff's walkthrough and observation of a drill at the proposed TSC, that the size, layout, and capabilities of the proposed TSC are adequate and will serve to enhance the functions of the TSC as currently described in the CNP Emergency Plan. Therefore, the NRC staff has concluded that the proposed TSC is consistent with the guidance in Section 2.1 of NUREG-0696, and meets the standards of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to TSC functional capabilities.

3.2 Location The guidance in Section 8.2.1, "Requirements," of NUREG-0737, states, in part, that the TSC will be "located within the site protected area so as to facilitate necessary interaction with control room, OSC [operations support center], EOF, and other personnel involved with the emergency."

The guidance in Section 2.2 of NUREG-0696 states, in part:

1 EROS is a direct near real-time electronic data link between the licensee's onsite computer system and the NRC Operations Center that provides for the automated transmission of a limited data set of selected parameters. The EROS supplements the existing voice transmission over the Emergency Notification System by providing the NRC Operations Center with timely and accurate updates of a limited set of parameters from the licensee's installed onsite computer system in the event of an emergency.

The onsite TSC is to provide facilities near the control room for detailed analyses of plant conditions during abnormal conditions or emergencies by trained and competent technical staff....

. . . the walking time from the TSC to the control room shall not exceed 2 minutes ....

. . .there should be no major security barriers between the two facilities other than access control stations for the TSC and control room ....

The CNP TSC is currently located in the Unit Nos. 1 and 2 CR complex. The licensee's submittal states that the proposed TSC will be located within the protected area in the northeast corner of the boundary on the second floor of the new TSC/North Access Building (NAB). The TSC/NAB houses the security control point for access to the protected area on the first floor.

The new location has an estimated walking time of approximately 6 minutes from the Unit Nos. 1 and 2 CR complex, which is a departure from the 2-minute criteria provided in NUREG-0696. This timing includes leaving the Unit Nos. 1 and 2 CR complex, walking through the plant and between buildings, and going to the second floor of the new TSC/NAB. Figure 3, "Overview of Protected Area Arrangement Showing TSC/NAB," in Enclosure 2 to the licensee's November 7, 2017, letter, illustrates the location of the proposed TSC in relation to the Unit Nos. 1 and 2 CR complex.

The licensee's submittal states that there are no major security barriers between the two facilities, but there is a security delay fence that is located between the proposed TSC and the CR. The fence performs a defense-in-depth security function and is not considered to be a major security barrier, because the only time the barrier's personnel gates are closed is when a security-related event is occurring. In this case, TSC Emergency Response Organization (ERO) responders would be directed to go to an alternative location. If there is no security-related event occurring, ERO responders are able to walk through the open gates.

Other than this fence, there are no security barriers between the two facilities other than security card readers for the TSC and CR.

NUREG-0800 includes a statement that advanced communication capabilities may be used to satisfy the 2-minute travel time. Significant enhancements in data acquisition and communications technologies have reduced the need for direct face-to-face communications between the TSC and CR. The licensee's submittal states that management interaction and technical information exchange will be accomplished using plant computer and communication (telephone, radio, etc.) systems that provide means to directly contact the CR. The capabilities to review and evaluate technical data such as plant parameter display information are provided in the TSC from real-time systems that receive their inputs from the same sources as the CR.

Based on its review of the licensee's submittal and the NRC staff's walkthrough and observation of a drill at the proposed TSC, the NRC staff has determined that given the technological and communications capabilities, layout, and increased size of the new facility, the TSC's new location will effectively* support CNP's emergency response, consistent with the intent of the guidance in Section 2.2 of NUREG-0696, Section 8.2, of Supplement 1 to NUREG-0737, and NUREG-0800. Therefore, the location of the proposed TSC meets the standards of 10 CFR 50.47(b )(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50.

3.3 Staffing and Training The guidance in Section 2.3 of NUREG-0696 states, in part, "The licensee-designated TSC staff shall consist of sufficient technical, engineering, and senior designated licensee officials to provide the needed support to the CR during emergency conditions." Section 2.3 also discusses the need for training programs for TSC personnel, as well as participation in activation drills.

The licensee is not proposing any changes to ERO staffing as currently described in the CNP Emergency Plan. The proposed TSC staffing will continue to have the capability to support a simultaneous emergency at both Unit Nos. 1 and 2.

The licensee's submittal further states that the training of the ERO will continue to be maintained as currently described in the CNP Emergency Plan and Emergency Plan Implementing Procedures (EPIPs). Changes specific to the relocation of the TSC and the setup of new work stations will be incorporated into personnel training specific to ERO assignments.

Familiarization training for the proposed TSC has been conducted, and the NRC staff observed a practice drill in the proposed TSC on July 17, 2018, which demonstrated the ERO staff's ability to perform their required functions in the proposed TSC.

Based on its review of the licensee's submittal and the NRC staff's walkthrough and observation of a drill at the proposed TSC, the NRC staff has determined that the staffing and training of TSC staff, including ERO response times and periodic drills and exercises, remains unchanged from that currently described in the CNP Emergency Plan and Emergency Plan Implementing Procedures. As such, the NRC staff has concluded that the staffing and training of the proposed TSC is consistent with the guidance in Section 2.3 of NUREG-0696 and, therefore, meets the standards of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to the staffing and training of TSC personnel.

3.4 Size The guidance in Section 2.4 of NUREG-0696 recommends that the TSC be sized to accommodate a minimum of 25 individuals, including 20 designated licensee personnel and 5 NRC personnel, with enough space to allow a minimum of 75 square feet (ft)/person (1,875 square feet). The licensee's submittal states that the proposed TSC plant evaluation team room and associated work spaces will consist of approximately 5,015 square ft, in contrast to the existing TSC, which is 1,211 square ft in size. A detailed comparison between the current and proposed TSC is provided in Table 1, "TSC Resource Comparison Overview," in Enclosure 2 to the licensee's November 7, 2017, letter, and was updated in the licensee's August 14, 2018, letter (Reference 1 and Reference 3, respectively).

The TSC will consist of a central command center with workspaces for designated personnel surrounding its perimeter, as well as a dedicated NRC conference room, a work room, and a dose assessment room. The TSC also includes restrooms with a shower facility; a break room; two storage rooms; a data room; a records room for copiers and printers; and a heating, ventilation, and air conditioning (HVAC)/electrical room that supports only the TSC. Diagrams of the proposed TSC were included in Figure 2, "TSC Resource Layout Diagram," in to the licensee's November 7, 2017, letter, and were updated in Figure 1, "Updated TSC Floor Plan," of the licensee's August 14, 2018, letter (Reference 1 and Reference 3, respectively).

The submittal describes, and the NRC staff observed during its walkthrough and observation of a drill at the proposed TSC, that the proposed TSC includes a designated NRC conference room, which has dedicated work spaces and a conference table, as well as double-seating workstations in the plant evaluation team room to facilitate NRC counterpart seating following deployment of an NRC regional site team.

Based on its review of the licensee's submittal, and the NRC staff's walkthrough and observation of a drill at the proposed TSC, the NRC staff has determined that the proposed TSC will be of sufficient size to accommodate and support licensee ERO personnel, equipment, and documentation in the TSC, as well as NRC resident inspectors and regional site team personnel. As such, the NRC staff has concluded that the size of the proposed TSC is consistent with the guidance in Section 2.4 of NUREG-0696 and, therefore, meets the standards of 10 CFR 50.47(b )(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to adequate sizing of the proposed TSC to perform its designated functions.

3.5 Structure The guidance in Section 2.5 of NUREG-0696 provides the following criteria for a TSC structure:

The TSC complex must be able to withstand the most adverse conditions reasonably expected during the design life of the plant, including adequate capabilities for ( 1) earthquakes, (2) high winds (other than tornadoes), and (3) floods.

The TSC need not meet seismic Category I criteria or be qualified as an engineered safety feature (ESF). Normally, a well-engineered structure will provide an adequate capability to withstand earthquakes. Winds and floods with a 100-year recurrence frequency are acceptable as a design basis. Existing buildings may be used to house the TSC complex if they satisfy the above minimum criteria.

The licensee's submittal states that the TSC/NAB is built to the requirements of the 2009 Michigan Building Code and American Society of Civil Engineers 7, "Minimum Design Loads and Associated Criteria for Buildings and Other Structures," as called for in the Michigan Building Code. The TSC/NAB is also designed to meet the requirements of the 2012 International Building Code Site Class D, Design Category C, and Occupancy Category IV. Therefore, the licensee stated in its submittal that the TSC/NAB is "able to withstand the most adverse conditions, reasonably expected during the design life of the plant including adequate capabilities for high winds (other than tornadoes) and floods with a 100-year recurrence frequency and earthquakes as required by NUREG-0696."

Based on its review of the licensee's submittal, and the NRC staff's walkthrough of the proposed TSC, the NRC staff has determined that the physical structure of the proposed TSC is consistent with the guidance in Section 2.5 of NUREG-0696 and, therefore, meets the standards of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to the structural integrity of the proposed TSC.

3.6 Habitability The guidance in Section 2.6 of NUREG-0696 states, in part:

[The TSC] shall have the same radiological habitability as the control room under accident conditions. TSC personnel shall be protected from radiological hazards, including direct radiation and airborne radioactivity from inplant sources under accident conditions, to the same degree as control room personnel.

The TSC ventilation system shall function in a manner comparable to the control room ventilation system.

The guidance in Section 2.6 of NUREG-0696 also states, in part:

Sufficient potassium iodide [Kl] shall be provided for use by TSC and control room personnel. The capacity of the installed TSC ventilation shall be independent of these thyroid-blocking provisions.

Equipment that protects personnel shall be provided in the TSC for the staff who must travel between the TSC and the control room, or the EOF, under adverse radiological conditions. Protective equipment also shall be provided to allow TSC personnel to continue to function during the presence of low-level airborne radioactivity or radioactive surface contamination. Anticontamination clothing and respiratory protective gear are examples of equipment that shall be provided.

This equipment shall be properly maintained to assure availability during an emergency.

During the NRC staff's walkthrough of the proposed TSC, the NRC staff verified that sufficient quantities of Kl, portable radiation monitoring equipment, and personal protective equipment is being maintained in the TSC. The NRC staff also reviewed the equipment inventory list included in CNP procedure EPP-2080-ER0-001, "Emergency Response Resource Readiness."

The licensee's submittal states:

Similar to the current TSC, the new TSC is designed to meet the protected envelope functional requirements for habitability and ventilation similar to the CR as identified in NUREG-0696 and Section 11.B.2 of NUREG-0737. A stand-alone HVAC system will be provided. Shielding will be provided to significantly reduce the effects of external shine as well as TSC HVAC [heating, ventilation, and air conditioning] Unit filter unit shine following an accident. Radiation monitoring is provided to indicate radiation dose rates as well as airborne radioactivity levels.

The NRC staff's evaluation regarding the licensee's compliance with this guidance is set forth below.

3.6.1 TSC Dose Consequence Analysis By letter dated January 19, 2018, the licensee submitted Document No. RD-13-02, Revision 2, "Technical Support Center Loss of Coolant Accident Radiological Analysis." Revision 2 was written to incorporate the updated AST, as well as changes in the atmospheric dispersion factors, to model the 30-day TEDE to emergency response occupants in the relocated TSC

from a design-basis dose consequence loss-of-coolant accident (LOCA). The use of the design-basis dose consequence LOCA dose consequence analysis is appropriate since the assumed magnitude of the release for this accident bounds all other design-basis dose consequence analyses.

The NRC staff notes that the design-basis dose consequence LOCA analysis is intended to be based upon a major accident or possible event, resulting in dose consequences not exceeded by those from any accident considered credible. Historically, this accident analysis, which is performed to show compliance with the dose criteria specified in 10 CFR 50.67, is referred to as the maximum hypothetical accident. It should be noted that the requirements of 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,"

ensure that the emergency core cooling system will prevent significant core damage during a design-basis LOCA. Notwithstanding the requirements of 10 CFR 50.46, the maximum hypothetical accident for dose consequence determinations deterministically assumes a substantial core melt with an appreciable release of fission products into the containment.

Therefore, the maximum hypothetical accident is a conservative surrogate to enable a deterministic evaluation of the response of a facility's engineered safety features (ESFs). All design-basis dose consequence accident analyses are performed in an intentionally conservative manner in order to compensate for known uncertainties in accident progression, activity product transport, and atmospheric dispersion.

3.6.1.1 Source Term and Transport The licensee incorporated the source term used in the current licensing basis LOCA accident dose consequence analysis as documented in calculation RWA-1313-002, "Cook Nuclear Plant AST Radiological Analysis Core and [Reactor Coolant System (RCS)] Source Terms,"

Revision 0, dated June 12, 2015. The licensee followed all aspects of the guidance outlined in RG 1.183, Regulatory Position 3, regarding the reactor core inventory, release fractions, and timing for the evaluation of its dose consequence LOCA. The radioactivity released into the containment is assumed to terminate at the end of the early in-vessel phase that occurs 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the onset of a LOCA.

The licensee's licensing basis dose consequence LOCA includes dose contributions from the following potential radioactive material release pathways:

• Containment purge

• Containment leakage

• ESF leakage to the auxiliary building

• ESF leakage to the Reactor Water Storage Tank A more detailed discussion of the licensee's current licensing basis LOCA source term and transport, including the basis for the NRC staff's acceptance, can be found in the safety evaluation associated with Amendment Nos. 332 and 314 incorporating the full-scope AST (Reference 15).

3.6.1.2 TSC Habitability Using the current licensing basis source term and transport assumptions for the dose consequence LOCA, the licensee evaluated the contributions to the TEDE to occupants of the TSC from the following constituents: the infiltration of airborne radioactivity into the TSC, the

direct dose from the accumulation of radioactive material on the TSC filter system, and the direct dose from the external cloud of airborne radioactive materials surrounding the TSC.

3.6.1.2.1 Infiltration of Airborne Radioactivity To limit the infiltration of airborne radioactivity, the TSC emergency ventilation system is designed to provide a filtered makeup air flow rate to maintain the TSC habitability envelope at a positive pressure. The makeup flow rate is specified as a minimum of 650 cubic feet per minute (cfm) to a maximum of 1,000 cfm. For the TSC volume of 71,448 ft 3 , the minimum pressurization makeup flow rate of 650 cfm will result in more than 0.5 volume changes per hour which, as enumerated in SRP 6.4, allows for the least restrictive frequency for periodic verification. In addition, the TSC ventilation system includes airlocks and outdoor air bubble-tight dampers to further reduce the intake of unfiltered in-leakage. The licensee assumed an unfiltered in-leakage of 25 cfm from the time of system actuation until the end of the 30-day analysis period. This assumed unfiltered infiltration is based on 10 cfm for each air-locked entryway and 5 cfm for the intake damper. Assuming a 10 cfm infiltration for an airlock is conservative, based on SRP 6.4 guidance. Footnote 4 in SRP 6.4 states the following regarding infiltration for a pressurized CR:

Normally 5 Us (10 cfm) infiltration is assumed for conservatism. This flow could be reduced or eliminated if the applicant provides assurance that backflow (primarily as a result of ingress and egress) will not occur. This may mean installing two-door vestibules or equivalent.

The assumption of a constant 5 cfm of infiltration though a bubble-tight damper is also a conservative assumption. Typically, bubble-tight testing standards allow for leakage up to a 1/16 inch bubble in 1 second. A bubble growth rate of 1/16 inch per second equates to a flow rate of less than 0.001 cfm.

The TSC filtered makeup pressurization system includes a pre-high efficiency particulate air (HEPA) filter, a 2-inch charcoal absorber bed, and a post-HEPA filter. In accordance with RG 1.140 guidance, the licensee assigned filter efficiencies of 99, 95, and 95 percent for particulates, elemental iodine, and organic iodide, respectively.

The licensee's dose consequence analysis assumes that the TSC filtered makeup pressurization system will be activated 53 minutes after accident initiation. The NRC staff notes that this conservative assumption results in a significant increase in the calculated dose since the assumed AST will have developed to a substantial degree during the 53 minutes following accident initiation. In reality, it is highly likely that due to degrading plant conditions, the TSC will be activated prior to the onset of any significant release of radioactive material to the environment.

3.6.1.2.2 Direct Shine Dose from the TSC Filtration System The licensee used conservative assumptions to evaluate the dose to TSC occupants from radioactive materials retained on the TSC filtration system. For conservatism, the licensee maximized the calculated dose from the TSC filters by assuming that the HEPA and charcoal filters were 100 percent efficient and that that the TSC emergency makeup flow rate is initiated at the start of the postulated accident.

The licensee used the MicroShield computer code to evaluate the direct dose due to TSC filter loading. To ensure that the calculated doses through wall ventilation penetrations as documented in the calculation of record were conservative, the licensee performed a more detailed evaluation of the filter dose contribution using the Monte Carlo N-Particle Transport Code. The results of the evaluation using the more precise Monte Carlo N-Particle Transport Code indicated that the dose results based on the MicroShield program in the calculation of record are conservative.

3.6.1.2.3 Direct Shine Dose from the External Cloud of Airborne Radioactive Materials The licensee modeled the cloud shine dose in a conservative manner. The licensee modeled the cloud concentration in discrete time steps using bounding assumptions for the individual time steps. For example, the licensee used the 8-hour cloud concentration for the entire time period from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This approach adds a significant degree of conservatism by ignoring the effects of decay during each individual time step.

3.6.2 Atmospheric Dispersion Estimates The relocation of the TSC requires a recalculation of the x/Q values used in the dose calculation that demonstrates the habitability of the new TSC.

The licensee's license amendment request (LAR) involves evaluation of the radiological consequences for the TSC airborne dose for the following contributors: containment leakage, ESF leakage, containment purge, and refueling water storage tank (RWST) back leakage. The LAR identifies the following release pathways for the contributors: containment leakage pathway, RWST pathway (RWST back leakage), and unit vent pathway (ESF leakage and containment purge). The licensee developed new x/Q values for each release pathway for the TSC receptor in the calculations made for the radiological consequence assessments. The NRC staff reviewed the licensee's new atmospheric dispersion analyses as described below.

3.6.2.1 Meteorological Data In support of the November 7, 2017, LAR, the licensee provided supplemental information on January 19, 2018, regarding the atmospheric dispersion analysis. Hourly onsite meteorological data from calendar years 2002, 2004, 2005, 2007, and 2010 were used in the analysis. The meteorological data were formatted for the ARCON96 atmospheric dispersion code (Reference 17) in order to calculate updated x/Q values for the TSC. This format contained hourly data on wind speed, wind direction, and atmospheric stability class taken from the 10 m and 60 m levels of the onsite meteorological tower.

The NRC staff previously completed a detailed review related to the acceptability and representativeness of the 2002, 2004, 2005, 2007, and 2010 onsite hourly meteorological data for the CNP, Unit Nos. 1 and 2, adoption of TSTF-490, Revision 0, "Deletion of E-bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification," and the implementation of full-scope AST (Reference 15). Based on this review, the staff considers the onsite meteorological dataset from calendar years 2002, 2004, 2005, 2007, and 2010 suitable for use in making calculations for the atmospheric dispersion analyses used to support this LAR.

3.6.2.2 TSC Atmospheric Dispersion Estimates In support of the LAR, the licensee used the computer code ARCON96 to estimate xtQ values for the TSC for potential accidental releases of radioactive material. RG 1.194 endorses the ARCON96 model for determining xtQ values to be used in the design-basis evaluations of CR radiological habitability.

The ARCON96 code estimates xtQ values for various time-averaged periods ranging from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 30 days. The meteorological input to ARCON96 consists of hourly values of wind speed, wind direction, and atmospheric stability class. The xtQ values calculated through ARCON96 are based on the theoretical assumption that material released to the atmosphere will be normally distributed (Gaussian) about the plume centerline. A straight-line trajectory is assumed between the release points and receptors. The diffusion coefficients account for enhanced dispersion under low wind speed conditions and in building wakes.

The hourly meteorological data are used to calculate hourly relative concentrations. The hourly relative concentrations are then combined to estimate concentrations ranging in duration from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 30 days. Cumulative frequency distributions prepared from the average relative concentrations and the relative concentrations that are exceeded no more than 5 percent of the time for each averaging period are determined.

The dispersion coefficients used in ARCON96 have three components. The first component is the diffusion coefficient used in other NRC models such as PAVAN (Reference 18). The other two components are corrections to account for enhanced dispersion under low wind speed conditions and in building wakes. These components are based on analysis of diffusion data collected in various building wake diffusion experiments under a wide range of meteorological conditions. Because the dispersion occurs at short distances within the plant's building complex, the ARCON96 dispersion parameters are not affected by nearby topographic features such as bodies of water. Therefore, the staff finds the licensee's use of the ARCON96 dispersion parameter assumptions to be acceptable for use in estimating xtQ values for the TSC for potential accidental releases. to the January 19, 2018, supplement to this LAR includes Table 5.1: ARCON96 Meteorological Input, the meteorological file names, the lower and upper height measurements, and units of wind speed. Section 5.2.2 of this supplement describes the source input values of vertical velocity, stack flow, stack radius, and diffusion coefficients. Table 5.2: ARCON96 Source Input, includes the release type, release height, and building area for each release pathway. Table 5.4: ARCON96 Receptor Input, includes the distances to receptor, intake height, elevation difference, and direction to source for each release pathway. Table 5.6 lists the xtQ values from the ARCON96 output for the 0-2, 2-8, 8-24, 24-96, and 96-720 hour time intervals for each of the release pathways.

The NRC staff confirmed the licensee's atmospheric dispersion estimates by running the ARCON96 computer model and obtaining similar results. Both the NRC staff and the licensee used a ground-level release assumption for each of the release pathway-receptor combinations, as well as the previously discussed source-receptor distances, directions, heights, and area values. Based on the results of its confirmatory analysis, the staff finds the licensee's TSC xtQ values to be acceptable for use in the radiological consequence assessments.

3.6.2.3 Atmospheric Dispersion Estimate Conclusion The NRC staff reviewed the guidance, assumptions, and methodology used by the licensee to assess the x/Q values associated with postulated releases from the potential release pathways.

The NRC staff found that the licensee used methods, consistent with regulatory guidance identified in Section 2.0 of this safety evaluation. The licensee used onsite meteorological data that complied with the guidance of RG 1.23. The inputs and assumptions used to calculate the TSC x/Q values were also consistent with the guidance of RG 1.194. Therefore, on the basis of this review of the atmospheric dispersion analysis, NRC staff finds the licensee's proposed x/Q values acceptable for use in calculating the radiological consequences assessments associated with this LAR.

3.6.3 Habitability Conclusion The results of the licensee's calculation indicate that the TSC personnel will be protected from radiological hazards, including direct radiation and airborne radioactivity from inplant sources under accident conditions, to the same degree as CR personnel (radiation exposures shall not exceed 5 rem TEDE for the duration of the postulated accident). Based on its review of the licensee's dose calculation and the associated atmospheric dispersion estimates, the NRC staff finds reasonable assurance that the proposed TSC will provide occupants an adequate level of radiological protection under design-basis accident conditions.

Based on its review of the licensee's dose calculation and the associated atmospheric dispersion estimates, the NRC staff has determined with reasonable assurance that the proposed TSC will provide occupants an adequate level of radiological protection under design-basis accident conditions. In addition, during the NRC staff's walkthrough of the proposed TSC and drill observation, the NRC staff verified that sufficient quantities of Kl, portable radiation monitoring equipment, and personal protective equipment are being maintained in the TSC. As such, the NRC staff has concluded that the habitability of the proposed TSC is consistent with the guidance in Section 2.6 of NUREG-0696 and, therefore, meets the standards of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to habitability for TSC ERO personnel..

3. 7 Communications The guidance in Section 2. 7 of NUREG-0696 specifies that the TSC will be the primary onsite communications center during an emergency at a nuclear power plant and that it needs to include reliable primary and backup communications to the CR, OSC, EOF, and NRC, as well as State and local operations centers.

The licensee's submittal states that it intends to maintain the same or enhanced communications features at the proposed TSC, which include: station voice over internet protocol (VoIP) telephones, business local area network (LAN) terminals and personal computers with LAN and standalone printers, Gaitronics Plant Public Address System, portable radio communication system, Health Physics Network and Emergency Notification System lines, plant private automatic branch exchange connections for dedicated lines to the CR, fax (telecopier), and commercial telephone service lines.

The VoIP system is battery powered with the batteries constantly serviced by a trickle charger powered from an alternating current station auxiliary power supply. In the event of a power

outage, the security diesel generator will provide power to the battery charger. The battery is capable of operating the VoIP system for approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when not being charged.

Based on its review of the licensee's submittal, and the NRC staff's walkthrough of the proposed TSC and observation of a drill, the NRC staff has determined that the proposed TSC has sufficient communications capabilities to support TSC functions and NRC response activities.

As such, the NRC staff has concluded that the proposed TSC is consistent with the guidance in Section 2.7 of NUREG-0696 and, therefore, meets the standards of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to reliable TSC voice and data communications capabilities.

3.8 Instrumentation, Data System Equipment. and Power Supplies The guidance in Section 2.8 of NUREG-0696 states, in part:

Equipment shall be provided to gather, store, and display data needed in the TSC to analyze plant conditions. The data system equipment shall perform these functions independent of actions in the CR and without degrading or interfering with CR and plant functions.

The TSC electrical equipment load shall not degrade the capability or reliability of any safety-related power source. Circuit transients or power-supply failures and fluctuations shall not cause a loss of any stored data vital to the TSC functions.

Sufficient alternate or backup power sources shall be provided to maintain continuity of TSC functions and to immediately resume data acquisition, storage, and display of TSC data if loss of the primary TSC power sources occurs.

The licensee's submittal states that the proposed TSC provides for the same level of instrumentation and data systems equipment as the current TSC and meets the intent for power supply requirements. It states:

The new TSC data system and instrumentation, providing inputs to the data used in the new TSC to analyze plant conditions, is routed through the same data server, the PSS [Plant System Server], that the current TSC uses for its data system and instrumentation inputs of plant data. For construction of the new TSC, redundant hardware data connections were created on the data server such that the data server currently has the capability to simultaneously provide inputs to both the current TSC and the new TSC data systems. The data system is designed such that it does not degrade or interfere with CR and plant functions.

The new TSC did not result in the need for any hardware changes to instrumentation and data system equipment and therefore does not affect the reliability of such.

Both the current TSC and new TSC normal power supplies meet the unavailability criteria established in the NUREG-0696 (Reference 2). The TSC/NAB building power is fed from a 12kV [kilo volts] off site power source and is supplied from the same power source that supplies the CNP 69kV Emergency Power. The current TSC has dual normal power supplies from each Train of Unit 1 balance of plant buses. The backup power source for the current TSC and

the new TSC is the Security Diesel generator. An evaluation was performed on the available capacity of the Security Diesel generator and it was determined that the generator has sufficient capacity to power the designated new TSC power loads in addition to the existing loads currently required of the Security Diesel generator. The current TSC and the new TSC both have a UPS [uninterruptible power supply] for TSC-only HVAC, utility receptacles/lights, and both have emergyoency lights with batteries. Similar to the current TSC, the new TSC UPS is a 30kVA [kilo-volt-ampere] UPS with a battery backup that will provide backup power for 30 minutes in the case of a loss of all AC power, which allows sufficient time for the Security Diesel to be started and loaded.

The NRC staff verified during its drill observation that TSC could satisfactorily obtain plant system and radiological data independent of actions in the CR and without degrading or interfering with CR and plant functions.

Based on its review of the licensee's submittal, and the NRC staff's walkthrough of the proposed TSC and its drill observation, the NRC staff has determined that the proposed TSC provides for reliable TSC instrumentation, data system equipment, and power supplies. As such, the NRC staff has concluded that the TSC instrumentation, data systems, and power supplies are consistent with the guidance in Section 2.8 of NUREG-0696 and, therefore, meet the standards of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to providing reliable equipment to gather, store, and display data needed in the TSC.

3.9 Technical Data and Data Systems The guidance in Section 2.9 of NUREG-0696 states, in part:

The TSC technical data system shall receive, store, process, and display information acquired from different areas of the plant as needed to perform the TSC function. The data available for display in the TSC must enable the plant management, engineering, and technical personnel assigned there to aid the control room operators in handling emergency conditions.

The data set available to the TSC data system must be complete enough to permit accurate assessment of the accident without interference from the control room emergency operation.

There is to be data storage and recall provided for the TSC data set.

The licensee's submittal states:

The current data systems for technical data to the TSC use replicated data through a data diode to ensure that the actual PPC [Plant Process Computer]

and the instrumentation supplying the data cannot be affected. The PPC data for each Unit is continuously replicated to the PSS, which functions as a read-only PPC. Meteorological data is available on the read-only PPC for use in dose assessment. The read-only PPC is accessed through the station business LAN.

The RMS [Radiation Monitoring System] data for each Unit is continuously replicated to the Plant Data Server (PDS) which functions as a read-only RMS.

Radiation monitoring software that is used by the ERO Radiological Assessment

Coordinator for on-site dose assessment, is accessed through the business LAN and uses technical data input from both the PSS and PDS.

The licensee's submittal indicated that the proposed TSC provides the same technical data and data systems as the current TSC, which would provide for continued evaluation and assessment of the emergency. As with the current TSC, the proposed TSC provides data through the business LAN. In addition, the proposed TSC provides additional workstation computers with LAN access beyond those provided in the current TSC.

Based on its review of the licensee's submittal, the NRC staff's walkthrough of the proposed TSC, and its drill observation, the NRC staff has determined that the TSC technical data and data systems are consistent with the guidance in Section 2.9 of NUREG-0696 and, therefore, meet the standards of 10 CFR 50.47(b)(8) and the requirements in Paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to data receipt, storage, processing, and display.

3.10 Records Availability and Management The guidance in Section 2.10 of NUREG-0696 states, in part:

The TSC shall have a complete and up-to-date repository of plant records and procedures at the disposal of TSC personnel to aid in their technical analysis and evaluation of emergency conditions. In particular, up-to-date as-built drawings of the plant systems are needed to diagnose sensor data, evaluate data inconsistencies, and identify and counteract fault plant system elements.

The licensee's submittal states that the proposed TSC will include an up-to-date repository of selected plant records, drawings, and procedures, consistent with the current TSC and NUREG-0696, to aid in technical evaluation of emergency conditions by the personnel located in the TSC. The plant records will be controlled and will be available via the business LAN from the Nuclear Documents Management System.

Based on its review of the licensee's submittal, and the NRC staff's walkthrough of the proposed TSC and its drill observation, the NRC staff has determined that the proposed TSC provides for records availability and management, consistent with the guidance in Section 2.1 O of NUREG-0696 and, therefore, meets the standards of 10 CFR 50.47(b)(8), and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, in support of the technical evaluation of emergency conditions.

3.11 Technical Evaluation Summary As discussed above, the NRC staff finds that the proposed relocation of the CNP TSC continues to meet the applicable planning standards of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50. Given the enhanced technological capabilities, layout, and increased size of the new facility, the proposed TSC will continue to effectively support CNP's emergency response in support of the functions set forth in NUREG-0696 and Supplement 1 to NUREG-0737. As such, the NRC staff finds reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency upon the licensee's relocation of the CNP TSC. Therefore, the NRC staff has concluded that the licensee's proposed relocation of the CNP TSC, as detailed in the licensee's letter dated November 7, 2017, as supplemented by letters dated January 19, 2018, and August 14, 2018, is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendments on September 4, 2018. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the CNP Emergency Plan, relate to changes in recordkeeping, reporting, or administrative procedures, or requirements, and result in changes to requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Letter from Q. S. Lies, Indiana Michigan Power Company, to U.S. Nuclear Regulatory Commission, "Donald C. Cook Nuclear Plant, Unit 1 and Unit 2, License Amendment Request Regarding Revision to the Emergency Plan for Technical Support Center Relocation,"*dated November 7, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17317A454).

2. Letter from Q. S. Lies, Indiana Michigan Power Company, to U.S. Nuclear Regulatory Commission, "Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 - Supplement to License Amendment Request Regarding Revision to the Emergency Plan for Technical Support Center Relocation," dated January 19, 2018 (ADAMS Accession No. ML18024A430).

3. Letter from Q. S. Lies, Indiana Michigan Power Company, to U.S. Nuclear Regulatory Commission, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Supplement to License Amendment Request Regarding Revision to the Emergency Plan for Technical Support Center Relocation," dated August 14, 2018 (ADAMS Accession No. ML18232A063).

4. RG 1.101, Revision 2, "Emergency Planning and Preparedness for Nuclear Power Reactors," dated October 1981 (ADAMS Accession No. ML090440294).

5. NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear

5. NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," dated November 1980 (ADAMS Accession No. ML040420012), as amended in March 2002 (ADAMS Accession No. ML021050240).

6. NUREG-0696, "Functional Criteria for Emergency Response Facilities," dated February 1981 (ADAMS Accession No. ML051390358).

7. NUREG-0737, "Clarification of TMI Action Plan Requirements," Supplement 1, "Requirements for Emergency Response Capability," dated January 1983 (ADAMS Accession No. ML102560009).

8. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," dated March 2007 (ADAMS Accession No. ML052340534).

9. NUREG-0800, SRP Section 2.3.4, Revision 3, "Short-Term Diffusion Estimates for Accidental Atmospheric Releases," dated March 2007 (ADAMS Accession No. ML070730398).

10. NUREG-0800, Standard Review Plan (SRP), Section 6.4, Revision 3, "Control Room Habitability System," dated March 2007 (ADAMS Accession No. ML070550069).

11. RG 1.23, Revision 1, "Onsite Meteorological Programs," dated March 2007 (ADAMS Accession No. ML070350028).

12. RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," dated June 2003 (ADAMS Accession No. ML031530505).

13. RG 1.140, Revision 3, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," dated August 2016 (ADAMS Accession No. ML16070A277).

14. RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000 (ADAMS Accession No. ML003716792).

15. Letter from A. Dietrich, U.S. Nuclear Regulatory Commission, to J. Gebbie, Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2-lssuance of Amendments Re: Adoption of TSTF-490, Rev. 0, 'Deletion of E-Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification,' and Implementation of Full Scope Alternative Source Term (CAC Nos. MF5184 and MF5185)," dated October 20, 2016 (ADAMS Accession No. ML16242A111).

16. TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Technical Specification," dated September 13, 2005 (ADAMS Accession No. ML052630462).

17. NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes," dated May 31, 1997 (ADAMS Accession No. ML17213A187).

18. NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design-Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations," dated November 30, 1982 (ADAMS Accession No. ML12045A149).

Principal Contributors: A. Marshall, NSIR J. Parillo, NRR J. White, NRO Date of Issuance: November 13, 2018

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 341 and 323 RE: TECHNICAL SUPPORT CENTER RELOCATION (EPID L-2017-LLA-0375) DATED NOVEMBER 13, 2018 DISTRIBUTION:

PUBLIC PM File Copy RidsNrrDorllpl3 Resource RidsNrrPMDCCook Resource RidsNrrLASRohrer Resource RidsACRS_MailCTR Resource RidsNrrDraArcb Resource RidsNsirDprRlb Resource RidsNroDlseRhm Resource RidsRgn3MailCenter Resource AMarshall, NSIR JWhite, NRO JParillo, NRR ADAMS Access1on No.: ML18249A019 *b,y memoran dum OFFICE N RR/DORL/LPL3/PM NRR/DORL/LPL3/LA NSI R/DPR/RLB/BC*

NAME ADietrich SRohrer JAnderson DATE 09/05/2018 09/17/2018 07/23/2018 OFFICE NRR/DRA/ARCB/BC* NRO/DLSE/RHM/BC* OGC-NLO NAME JDozier CCook STurk DATE 05/15/2018 07/02/2018 10/10/2018 OFFICE NRR/DORL/LPL3/BC NRR/DORL/D NRR/D NAME DWrona CErlanger HNieh (MEvans for)

DATE 10/18/2018 10/31/2018 11/13/2018 OFFICIAL RECORD COPY