ML21131A127

Issuance of Amendment No. 343 Modifications to Technical Specification 3.6.1.3, Primary Containment Isolation Valves (Pcivs)

ML21131A127
Person / Time
Site:	FitzPatrick
Issue date:	08/09/2021
From:	Justin Poole Plant Licensing Branch 1
To:	Rhoades D Exelon Generation Co
Poole J
References
EPID L-2020-LLA-0145
Download: ML21131A127 (14)

Text

August 9, 2021 Mr. David P. Rhoades Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ISSUANCE OF AMENDMENT NO. 343 RE: MODIFICATIONS TO TECHNICAL SPECIFICATION 3.6.1.3, PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS) (EPID L-2020-LLA-0145)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 343 to Renewed Facility Operating License No. DPR-59 for the James A.

FitzPatrick Nuclear Power Plant (FitzPatrick). The amendment consists of changes to the technical specifications (TSs) in response to your application dated June 30, 2020, as supplemented by letters dated November 10, 2020, and January 26, 2021.

The amendment revises the containment venting flow path as described in the FitzPatrick TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Surveillance Requirement 3.6.1.3.1.

A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Justin C. Poole, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 343 to DPR-59

2. Safety Evaluation cc: Listserv

EXELON FITZPATRICK, LLC AND EXELON GENERATION COMPANY, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 343 Renewed Facility Operating License No. DPR-59

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

A. The application for amendment by Exelon FitzPatrick, LLC and Exelon Generation Company, LLC (collectively, the licensees) dated June 30, 2020, as supplemented by letters dated November 10, 2020, and January 26, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-59 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 343, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James G. James G. Danna Date: 2021.08.09 Danna 13:36:13 -04'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 9, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 343 JAMES A. FITZPATRICK NUCLEAR POWER PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following page of the renewed facility operating license with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page Page 3 Page 3 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Page Insert Page 3.6.1.3-7 3.6.1.3-7

(4) Exelon Generation Company pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct, source and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus, components or tools.

(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Exelon Generation Company is authorized to operate the facility at steady state reactor core power levels not in excess of 2536 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 343, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection Exelon Generation Company shall implement and maintain in effect all provisions of the approved fire protections program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated November 20, 1972; the SER Supplement No. 1 dated February 1, 1973; the SER Supplement No. 2 dated October 4, 1974; the SER dated August 1, 1979; the SER Supplement dated October 3, 1980; the SER Supplement dated February 13, 1981; the NRC Letter dated February 24, 1981; Technical Specification Amendments 34 (dated January 31, 1978), 80 (dated May 22, 1984), 134 (dated July 19, 1989),

135 (dated September 5, 1989), 142 (dated October 23, 1989), 164 (dated August 10, 1990), 176 (dated January 16, 1992), 177 (dated February 10, 1992), 186 (dated February 19, 1993), 190 (dated June 29, 1993), 191 (dated July 7, 1993), 206 (dated February 28, 1994), and 214 (dated June 27, 1994); and NRC Exemptions and associated safety evaluations dated April 26, 1983, July 1, 1983, January 11, 1985, Amendment 343 Renewed License No. DPR-59

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 343 EXELON FITZPATRICK, LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59

1.0 INTRODUCTION

By letter dated June 30, 2020, as supplemented by letters dated November 10, 2020, and January 26, 2021 (Agencywide Documents Access and Management System (ADAMS)

Accession Nos. ML20182A198, ML20315A245, and ML21026A111, respectively), Exelon Generation Company, LLC (Exelon, the licensee) submitted a license amendment request (LAR) for changes to the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) Technical Specifications (TSs).

The proposed changes would revise the description of the containment venting flow path as in the FitzPatrick TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), Surveillance Requirement (SR) 3.6.1.3.1. Specifically, the licensee proposed to revise the TS 3.6.1.3 configuration requirements for the reactor building suction valves during inerting and de-inerting the containment atmosphere.

The supplements dated November 10, 2020, and January 26, 2021, 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) staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 25, 2020 (85 FR 52374).

2.0 REGULATORY EVALUATION

2.1 System Description The Standby Gas Treatment System (SGTS) consists of two identical, physically and electrically separated air filtration trains. Each full-capacity train consists of a demister, a prefilter, an electrical heating coil, a high-efficiency particulate air (HEPA) filter, an activated charcoal absorber, and a HEPA after filter. The demister in each train is designed to remove entrained water droplets and mist from the entering air stream.

Enclosure 2

The secondary containment utilizes features of the SGTS to mitigate the consequences of the postulated loss-of-coolant accident (LOCA) inside the drywell, as well as a refueling accident inside the secondary containment. One feature of the SGTS is to provide a negative pressure barrier which minimizes the ground level release of fission products by exfiltration. A second feature related to the SGTS is the exhausting of the secondary containment atmosphere through the SGTS filter trains to an elevated release point which aids in dispersion of the effluent by atmospheric diffusion.

The SGTS configuration consists of the drywell and torus venting through parallel pipelines containing redundant PCIVs. Downstream of the outboard PCIVs, the drywell (24-inch) and Torus (20-inch) pipelines are connected to a common pipeline to direct the containment atmosphere to the SGTS for treatment and discharge via the main stack. Prior to reaching the SGTS filter train, the flow path branches to parallel 6-inch and 12-inch pipelines containing isolation valves. With the 12-inch line closed, the 6-inch pathway is used (during Modes 1, 2, and 3) to: (1) Provide the primary flow path for containment atmosphere maintenance during normal plant operation and (2) Prevent overpressurization of the SGTS when the 20-inch and/or 24-inch PCIVs are open for containment venting, should an accident occur during the venting process.

During normal operation, the SGTS provides primary flow for containment atmosphere maintenance during inerting, de-inerting, pressure control, as low as reasonably achievable (ALARA) or air quality considerations for personnel entry, with either 20-inch and/or 24-inch PCIVs open for venting. During accident conditions, upon receipt of an initiation signal the 20-inch and/or 24-inch PCIVs will close, and at least one SGTS filter train fan will start up and all valves in that train will open to draw air from the isolated reactor building.

2.2 Description of the Proposed Changes The licensee describes the current method for venting (inerting and de-inerting) the containment atmosphere as time consuming due to inefficient flow from the containment to SGTS. The current method for venting (inerting and de-inerting) is to align the containment (drywell) with the SGTS through a 6-inch pathway, while maintaining an open flow path between the reactor building and the SGTS HEPA filters. In this configuration, flow through the SGTS filter train is a combination of containment and reactor building atmospheres. This valve configuration restriction is included as a note to SR 3.6.1.3.1 of TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs).

The proposed change will allow venting the containment during inerting and de-inerting with the reactor building suction valves in the closed position, resulting in the entire flow through the SGTS filter train from containment through the 6-inch pathway. The licensee states that flow through the SGTS filter train remains approximately the same with the proposed lineup. Since the flow is entirely from containment in the proposed lineup, there is a reduction in containment venting duration.

2.3 Description of the TS Changes SR 3.6.1.3.1 requires verification that each 20-inch and 24-inch primary containment vent and purge valve is closed at a frequency in accordance with Surveillance Frequency Control Program. The associated note to SR 3.6.1.3.1 states:

---

Note-----------------------------------------------

Not required to be met when the 20 inch and 24 inch primary containment vent and purge valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open, provided the full-flow line to Standby Gas Treatment (SGT)

System is closed and one or more SGT System reactor building suction valves are open.

The licensee is proposing to delete the phrase and one or more SGT System reactor building suction valves are open. If approved, the authorized containment vent path through the 6-inch pathway would remain the same but with the SGT reactor building suction valves closed.

2.4 Regulatory Requirements and Guidance The NRC regulatory requirements in Title 10 of the Code of Federal Regulations (10 CFR) 50.36(a)(1) require an applicant for an operating license to include in the application proposed TS in accordance with the requirements of 10 CFR 50.36.

The regulation, 10 CFR 50.36(b), requires TSs to be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.

The regulation, 10 CFR 50.36(c)(3), Surveillance requirements, requires that TSs include requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The regulation, 10 CFR 50.55a(f)(4), Inservice testing standards requirement for operating plants, states, in part, that valves that are within the scope of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) must meet the inservice testing (IST) requirements set forth in the ASME OM Code and that valves that are within the scope of the ASME OM Code but are not classified as ASME Boiler and Pressure Vessel Code (BPV Code) Class 1, Class 2, or Class 3 may be satisfied as an augmented IST program.

3.0 TECHNICAL EVALUATION

The primary containment is inerted with nitrogen to maintain low oxygen concentrations during normal plant operations. Containment inerting is normally achieved by first supplying nitrogen to the torus until the required oxygen concentration is achieved. Once the torus is inerted, nitrogen is supplied to the drywell until the required oxygen concentration is achieved.

Additional nitrogen is then supplied to the drywell until the required drywell to torus differential pressure is achieved. During the containment inerting process, the gas purged from the torus and drywell is processed by the SGTS prior to being exhausted to the atmosphere.

During the containment de-inerting process, fresh air is supplied to the containment with a vent and purge supply fan and the containment gas is processed by the SGTS prior to being exhausted to the atmosphere.

A demister loop seal is located on each SGTS filter train for removal of any suspended water droplets in the air stream prior to entering the charcoal filters. The demister loop seal must remain intact following a LOCA and not result in a leak path of radioactive iodine out of the HEPA filter assembly. If the loop seal fails, resulting in a leak path through the seals, air would essentially bypass the charcoal filters and reduce the effectiveness of the SGTS filter assembly.

The licensees current method for venting (inerting and de-inerting) is to align the containment (drywell) with the SGTS through a 6-inch pathway, while maintaining an open flow path between the reactor building (reactor building valves open) and the SGTS filter trains. The licensees current calculations contain analyses for venting with reactor building valves open and with reactor building valves closed configurations (Case 1 and Case 2, respectively), as presented in the Evaluation of the Proposed Change in Section 3.0 Technical Evaluation of Attachment 1 to the LAR. Case 1 evaluates the current valve lineup with reactor building suction valves open at the time of a postulated LOCA. Case 2 evaluates valve lineup with reactor building suction valves closed and opened at the time a LOCA is postulated to occur. Current design calculations with these reactor building suction valves closed (Case 2) concluded an overpressure condition for demister loop seals exists. To address the potential overpressure condition of demister loop seals, the licensee venting process is currently performed with reactor building supply valves in an open position. As a result, SR 3.6.1.3.1 restrictions exist to prevent potential overpressurization of the SGTS demister loop seals, should a LOCA occur during containment venting.

The licensee has determined that the current process and valve alignment, required per the TS note, results in inefficient and lengthy containment venting. The licensee indicated that while the flow through the SGTS remains approximately the same, the new proposed lineup will route flow from the primary containment only and reduce the containment venting time by half. To address the venting configuration concern, the licensee performed a reanalysis of Case 2 with valve arrangement having the reactor building suction valves closed. The reanalysis modified assumptions and evaluated the SGTS ability to withstand the LOCA-induced loads during the brief periods of containment venting. Based on results of the reanalysis, the licensee is proposing to modify the valve arrangement in order to vent containment with the reactor building supply valves closed. The licensees Case 2 reanalysis concludes that the proposed SR operating restriction would no longer be required as the pressure does not exceed the demister loop seal design.

The original analysis was evaluated using RELAP5/MOD 3.2.1. The reanalysis was performed using the more recent RELAP5/MOD 3.3 version. In addition, the new analysis made certain assumptions that modeled flow more accurately through the SGTS fans and modeled fan flow during a LOCA, but no physical modification was made to the system. One assumption is based on a recalculation of the initial fan flow input to account for resistance of piping and better represent actual conditions. A second assumption relates to the method of defining LOCA flow through both trains. As defined in Updated Final Safety Analysis Report (UFSAR)

Section 5.3.3.4, Standby Gas Treatment System, upon receipt of an initiation signal at least one filter train fan will start up and all valves in that train will open to draw air from the isolated reactor building. As discussed in the application, the original (Case 2) analysis maintained a fixed flow (equal to the fan flow) through the operating train of SGTS. However, the reanalysis in the application models flow as free flow through both trains during a LOCA. The reanalysis

also includes enhancements in modeling the inlet and outlet connections to better represent the pressure difference across the HEPA filters and noted only a relatively minor impact on results.

As shown in the table (Section 3.0 of the LAR), the reanalysis results conclude that the reanalyzed Case 2 for demister loop seal meets the acceptance criteria and no longer results in an overpressure condition.

To ensure the licensee evaluated the ability of an SGTS worst-case scenario during a LOCA, the NRC staff requested the licensee to describe operation and response to a LOCA with one and two filter trains operating. One of the updated model assumptions is to allow free flow through both trains when evaluating overpressurization. By letter dated June 30, 2020, Case 2 assumption is that the overpressure condition is driven by the pressure surge generated in the system by the drywell pressure change being higher than previously modeled. It was not clear how worst-case overpressure conditions for filter train and demister loop seals were defined and considered in the analysis. By letter dated January 26, 2021, in response to an NRC request for additional information (RAI) dated December 17, 2020 (ADAMS Accession No. ML20352A275),

the licensee described modeling of the worst-case condition with a single train in operation prior to the LOCA and the opposite (isolated) train opening following a LOCA signal. Typically, during inerting/de-inerting, the SGTS operates on a single train with the inlet and outlet valves on the opposite train closed. Before a LOCA, the flow is driven through the system by one SGTS train with its fan in operation; the alternate train is isolated. Following a LOCA signal (low-low-low reactor water level or high drywell pressure), the inlet valve on the opposite train opens followed by the discharge valve. Valve opening time is based on the LOCA signal to the valve. The valves in the train which is in operation prior to the LOCA remain open. With a single train initially operating, pressure will be biased (higher) on the operating train until the opposite train isolation valves are opened. This maximizes the pressure at the SGTS filter train for the LOCA scenario. As stated in RAI response, the licensee provided details of the case analyzed and SGTS response to LOCA during the inert/de-inert configuration and concluded that the pressures remain below the design of the SGTS filter train and demister loop seal.

The NRC staff also requested that the licensee describe what impact, if any, an accident environment (e.g., temperature, wet steam) will have on the SGTS filter trains. In its response dated January 26, 2021, the licensee stated that PCIVs will close within 6 seconds after initiation of a LOCA and during that period the drywell mixture of steam, nitrogen, and air increases linearly from 230 degrees Fahrenheit (°F) to 275 °F. The SGTS filter train is designed to be heat resistant to at least 250 °F. The licensee noted that the fluid mixture from the drywell has to navigate 500 feet of relatively cold pipe before it gets to the SGTS filter trains. Given the short duration of 6 seconds and the intervening components driving cooling/humidity reduction, the licensee concluded that the accident environment should not have an impact on the SGTS filter train post-LOCA capability. The NRC staff reviewed the RAI response and concludes that the qualitative arguments provided by the licensee are reasonable and acceptable, as piping and condensation during the short duration of elevated temperature provide additional cooling and margin to minimize risk of overtemperature condition. In addition, it is unlikely that a LOCA would occur during venting operations and the safety risk presented is extremely small.

The NRC staff reviewed the licensees evaluation of the SGTS system response to a LOCA demonstrating the demister loop seal will not be overpressurized and SGTS will perform its safety function with the new venting method and valve configuration. The analysis in the application, as supplemented, shows that the proposed valve configuration does not impact the current ability of the reactor building suction valves to perform their safety function upon receipt of a safety signal. Based on the licensees reanalysis related to SGTS and the ability of the demister loop seal to withstand a LOCA, the NRC staff finds that venting with a new valve

configuration will continue to be performed in a safe manner and avoid loop seal overpressure condition. Therefore, removing TS restrictions to have reactor building suction valves in the open position during inerting/de-inerting processes is acceptable.

In the June 30, 2020, letter, under the discussion of the affected valves, the licensee specifically named a single reactor building suction valve, 01-125-MOV-12, despite the request impacting the configuration of two valves (01-125-MOV-11 in addition to 01-125-MOV-12). In an RAI dated October 19, 2020 (ADAMS Accession No. ML20294A037), the NRC staff asked the licensee whether the second valve was 01-125-MOV-11 and if the statements made about 01-125-MOV-12 were applicable to the other valve and also whether this other valve had other unique attributes that required consideration in the review of the LAR. In its response dated November 10, 2020, the licensee confirmed that the second valve was 01-125-MOV-11 and there were no unique attributes that required consideration.

In the October 19, 2020, RAIs, the NRC staff also requested information on whether the performance requirements for these two valves would remain bounded after implementation of the requested change, if their status as augmented valves in the FitzPatrick IST Program would change as a result of this amendment request, and how the closed valves would be tested for operational readiness if the proposed amendment were issued. In its response dated November 10, 2020, the licensee confirmed that there is no adverse impact to the performance of the valves, their status in the IST Program was not expected to change, and the valves would continue to be stroke time tested as required by the IST Program.

Based on the information provided by the licensee, in the June 30, 2020, submittal and in response to the NRC staffs RAIs, the NRC staff concludes that there is assurance that the reactor building suction valves will be operationally ready to perform their required functions.

The NRC staff finds that the performance of the reactor building suction valves 01-125-MOV-11 and 01-125-MOV-12 will not be adversely impacted by the activities requested to be authorized by the proposed amendment in that these valves will continue to be subject to the requirements of the ASME OM Code, as required by 10 CFR 50.55a(f)(4).

The LAR includes the licensees reanalysis that demonstrated that the revised configuration in the proposed SR note will have no adverse impact on the operation of the SGTS filter train during accident conditions. The new analyses show that the overpressure concern addressed in the current SR 3.6.1.3.1 note is no longer applicable and, therefore, there is no need to retain the requirement that one or more reactor building valves be open to the SGTS filter train.

Therefore, the NRC staff concludes that the TS, as revised, will continue to meet the 10 CFR 50.36(b) requirement that TSs be derived from the analyses and evaluation in the safety analysis report, as amended.

The NRC staff also determined that the proposed change to SR 3.6.1.3.1 does not remove or alter any SR requirements, but removes a restriction on configuration of components during containment atmospheric control operations denoted by the associated note in SR 3.6.1.3.1.

The licensee demonstrated that the revised configuration with the SGTS suction valves closed, which would be allowed by deletion of a phrase in the note, will have no adverse impact on the operation of the SGTS filter train during accident conditions. Therefore, the NRC staff concludes that the change to SR 3.6.1.3.1 continues to demonstrate that LCO 3.6.1.3 is being met and, therefore, meets the requirements of 10 CFR 50.36(c)(3) for inclusion of requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on May 18, 2021. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a surveillance requirement. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (August 25, 2020; 85 FR 52374).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: G. Curran N. Hansing Date: August 9, 2021

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