ML19248C238

Issuance of Amendment No. 201 Regarding Technical Specification Changes to Eliminate Requirements of Iodine Removal System

ML19248C238
Person / Time
Site:	Saint Lucie
Issue date:	11/20/2019
From:	Natreon Jordan Plant Licensing Branch II
To:	Moul D Florida Power & Light Co
Wentzel M
References
EPID L-2018-LLA-0301
Download: ML19248C238 (16)

Text

November 20, 2019 Mr. Don Moul Vice President, Nuclear Division and Chief Nuclear Officer Florida Power & Light Company Mail Stop: NT3/JW 15430 Endeavor Drive Jupiter, FL 33478

SUBJECT:

ST. LUCIE PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENT NO. 201 REGARDING TECHNICAL SPECIFICATION CHANGES TO ELIMINATE THE REQUIREMENTS OF THE IODINE REMOVAL SYSTEM (EPID L-2018-LLA-0301)

Dear Mr. Nazar:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 201 to Renewed Facility Operating License No. NPF-16 for St. Lucie Plant, Unit No. 2. The amendment changes the technical specifications in response to the application from Florida Power & Light Company dated November 9, 2018, as supplemented by letter dated May 17, 2019. The amendment revises the technical specifications to eliminate the requirements for the Iodine Removal System and revises the surveillance requirements for the trisodium phosphate dodecahydrate basket.

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

Sincerely, G~

Natreon J. Jordal'i,~~ct Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389

Enclosures:

1. Amendment No. 201 to NPF-16

2. Safety Evaluation cc: Listserv

FLORIDA POWER & LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO, FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST. LUCIE PLANT UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 201 Renewed License No. NPF-16

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

A. The application for amendment by Florida Power & Light Company dated November 9, 2018, as supplemented by letter dated May 17, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended 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 Atomic Energy Act of 1954, as amended, 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.

Enclosure 1

2. Renewed Facility Operating License No. NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 3.B to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 201, are hereby incorporated in the renewed license. FPL shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~hy Undine Shoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: November 20, 201 9

ATTACHMENT TO LICENSE AMENDMENT NO. 201 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 ST. LUCIE PLANT, UNIT NO. 2 DOCKET NO. 50-389 Replace page 3 of Renewed Facility Operating License No. NPF-16 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Remove Insert VII VII 3/4 5-5 3/4 5-5 3/4 6-17 3/4 6-17

neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

D. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission's regulations: 10 CFR Part 20, Section 30.34 of 10 FR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR 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 below:

A. Maximum Power Level FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 3020 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 201, are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

Renewed License No. NPF-16 Amendment No. 201

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY ..................................................................... 3/4 6-1 CONTAINMENT LEAKAGE ....................................................................... 3/4 6-2 CONTAINMENT AIR LOCKS ..................................................................... 3/4 6-9 INTERNAL PRESSURE ........................................................................... 3/4 6-11 AIR TEMPERATURE ............................................................................... 3/4 6-12 CONTAINMENT VESSEL STRUCTURAL INTEGRITY .......................... 3/4 6-13 CONTAINMENT VENTILATION SYSTEM .............................................. 3/4 6-14 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY AND COOLING SYSTEMS ............................. 3/4 6-15 DELETED ................................................................................................. 3/4 6-17 3/4.6.3 CONTAINMENT ISOLATION VALVES ............................................................ 3/4 6-19 3/4.6.4 DELETED DELETED ................................................................................................. 3/4 6-24 DELETED ................................................................................................. 3/4 6-25 3/4.6.5 VACUUM RELIEF VALVES .............................................................................. 3/4 6-26 3/4.6.6 SECONDARY CONTAINMENT SHIELD BUILDING VENTILATION SYSTEM .......................................... 3/4 6-27 SHIELD BUILDING INTEGRITY .............................................................. 3/4 6-30 SHIELD BUILDING STRUCTURAL INTEGRITY ..................................... 3/4 6-31 ST. LUCIE - UNIT 2 VII Amendment No. +G, -%4, 201

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS <continued}

2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show. no evidence of structural distress or corrosion.

3. Verifying that a minimum total of 173 cubic feet of solid granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

4. Verifying that when a representative sample of 70.5 +/- 0.5 grams of TSP from a TSP storage basket is submerged, without agitation, in 10.0 +/- 0.1 gallons of 120 +/- 10°F borated water representative of the RWT, the pH of the mixed solution is raised to greater than or equal to 7 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

f. In accordance with the Surveillance Frequency Control Program, during shutdown, by:

1. Verifying that each automatic valve in the flow paths actuates to its correct position on SIAS and/or RAS test signals.

2. Verifying that each of the following pumps start automatically upon receipt of a Safety Injection Actuation Test Signal:

a. High-Pressure Safety Injection pumps.

b. Low-Pressure Safety Injection pumps.

c. Charging Pumps

3. Verifying that upon receipt of an actual or simulated Recirculation Actuation Signal: each low-pressure safety injection pump stops, each containment sump isolation valve opens, each refueling water tank outlet valve closes, and each safety injection system recirculation valve to the refueling water tank closes.

g. By verifying that each of the following pumps develops the specified total developed head when tested pursuant to the INSERVICE TESTING PROGRAM:

1. High-Pressure Safety Injection pumps.

2. Low-Pressure Safety Injection pumps.

h. By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:

1. During valve stroking operation or following maintenance on the valve and prior to declaring the valve OPERABLE when the ECCS subsystems are required to be OPERABLE.

ST. LUCIE - UNIT 2 3/4 5-5 Amendment No. 94-, W, 400, 4-ae,~.~.489,201

DELETED ST. LUCIE - UNIT 2 3/4 6-17 Amendment No. 94, ~ . 201

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 201 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 FLORIDA POWER & LIGHT COMPANY, ET AL.

ST. LUCIE PLANT, UNIT NO. 2 DOCKET NO. 50-389

1.0 INTRODUCTION

By application dated November 9, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18316A028), as supplemented by letter May 17, 2019 (ADAMS Accession No. ML19137A052), Florida Power & Light Company (the licensee) requested changes to the technical specifications (TSs) for St. Lucie Plant, Unit No. 2 (St. Lucie 2), which are contained in Appendix A of Renewed Facility Operating License No. NPF-16. The licensee proposed to eliminate the requirements of the Iodine Removal System (IRS) and modify the surveillance requirement (SR) associated with the trisodium phosphate dodecahydrate (TSP) basket.

The supplement dated May 17, 2019, 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 the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 2, 2019 (84 FR 24).

2.0 REGULATORY EVALUATION

2.1. System Description 2.1.1 Iodine Removal System The IRS, operating in conjunction with the Containment Spray System, serves to remove fission products from a post-accident containment atmosphere. An independent IRS train consisting of a constant-volume metering pump, solenoid-operated isolation valve, IRS tank, and associated piping and valves, is connected to each containment spray loop. The IRS removes radioiodines from the containment atmosphere following a loss-of-coolant accident (LOCA) by adding controlled amounts of hydrazine to containment spray water. The solenoid-operated isolation valves open and the hydrazine pumps start upon receipt of containment spray actuation signal.

Hydrazine is injected into the suction side of each containment spray pump until a low level switch in the hydrazine storage tank simultaneously stops the pumps and closes the solenoid-Enclosure 2

operated valves. The system is designed to be fully automatic but is capable of local-manual control.

2.1.2 Trisodium Phosphate Dodecahydrate Post-LOCA pH control at St. Lucie 2 is accomplished through the use of TSP, which is stored in 16 stainless steel mesh baskets located in the vicinity of the containment sump. The TSP is dissolved by borated water from containment spray as the post-LOCA water level increases, thereby raising pH. Mixing is achieved as the solution is continuously recirculated from the sump to the spray nozzles. Based on analyses performed by the licensee, the use of TSP stabilizes the post-LOCA containment sump pH above 7, which assures that the iodine removed from the containment atmosphere remains in solution in the containment sump.

2.2 Licensee's Proposed Changes The licensee proposed to delete TS 3/4.6.2.2, "Iodine Removal System (IRS)," in its entirety and make the conforming change to the TS index. In addition, the licensee proposed to revise SR 4.5.2.e.4 as indicated below (deletions shown in stricken text and additions underlined):

4. Verifying that when a representative sample of 70.5 +/- 0.5 grams of TSP from a TSP storage basket is submerged, without agitation, in 10.0 +/- 0.1 gallons of 120 +/- 10°F borated waterffem representative of the RWT [refueling water tank], the pH of the mixed solution is raised to greater than or equal to 7 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2.3 Regulatory Review The NRC staff reviewed the licensee's application to determine whether (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 the activities proposed 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 the health and safety of the public. The NRC staff considered the following regulatory requirements, guidance, licensing, and design-basis information during its review of the proposed changes.

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) states, in part, that each applicant for an operating license shall include in the application proposed TSs in accordance with the requirements of 10 CFR 50.36, "Technical Specifications."

Section 50.36(c) of 10 CFR requires that the TSs include items in the following categories related to station operation: ( 1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; and (5) administrative controls.

Paragraph 50.36{c){3) states, in part, that SRs are 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.

Section 50.67 of 10 CFR, "Accident source term," (b )(2), states:

The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) 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) 1 total effective dose equivalent (TEDE).

(ii) 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 radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent {TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

Appendix A, "General Design Criteria (GDC) for Nuclear Power Plants," to 10 CFR Part 50 establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The following GDC are applicable for this review:

• GDC 14, "Reactor coolant pressure boundary," states that the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

• GDC 19, "Control room," states, in part, 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 LOCAs.

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 whole body, or its equivalent to any part of the body, for the duration of the accident.

• GDC 41, "Containment atmosphere cleanup," states that systems to control fission products, hydrogen, oxygen, and other substances that may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

1 The use of 0.25 Sv (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a reference value, which can be used in the evaluation of proposed design-basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," Revision 4 (ADAMS Accession No. ML070190178),

provides guidance to the NRC staff for the review of containment atmosphere fission product cleanup functions.

SRP Branch Technical Position (BTP} 6-1, "pH for Emergency Coolant Water for Pressurized Water Reactors" (ADAMS Accession No. ML063190011) provides guidance for establishing the minimum value of pH in post-accident containment sprays. It says in part:

The criteria for pH level of postaccident emergency coolant water to reduce the probability of stress-corrosion cracking of austenitic stainless steel components, nonsensitized or sensitized, nonstressed or stressed, are as follows:

[...]

2. For the spray water recirculated from the containment sump, the higher the pH in the 7.0 to 9.5 range, the greater the assurance that no stress corrosion cracking will occur.

SRP Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms,"

Revision O (ADAMS Accession No. ML003734190), provides guidance to the NRC staff for the review of alternative source term amendment requests. SRP Section 15.0.1 states that the NRC reviewer should evaluate the proposed change against the guidance in Regulatory Guide (RG) 1.183.

RG 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (ADAMS Accession No. ML003716792), provides the methodology for analyzing the radiological consequences of several design-basis accidents to show compliance with 10 CFR 50.67. RG 1.183 provides guidance to licensees on acceptable application of alternate source term (also known as the accident source term) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted alternate source term.

The regulatory requirements applicable to the evaluation of radiological consequences from which the NRC staff based its acceptance are the reference values in 10 CFR 50.67, the accident-specific guideline values in Regulatory Position 4.4 of RG 1.183, and Table 1 of SRP Section 15.0.1.

The NRC staff also considered relevant information in Chapter 15 of the St. Lucie 2 UFSAR, "Accident Analyses" (ADAMS Accession Nos. ML19101A081 and ML19101A083), which describes the design-basis accidents and the evaluation of their radiological consequences.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's application to determine if the proposed changes are consistent with the guidance, regulations, and plant-specific design and licensing basis information discussed in Section 2.3 of this safety evaluation.

3.1 Deletion of TS 3/4.6.2.2 The licensee proposed to delete TS 3/4.6.2.2 and make conforming changes to the TS index.

The removal of the IRS has the potential to impact the following three aspects of a post-LOCA environment: containment sump pH, corrosion of materials, and radiological consequences.

The NRC staff's evaluation of the proposed change on these three aspects follows.

3.1.1 Containment Sump pH The original St. Lucie 2 licensing basis credited the use of spray additives (i.e., hydrazine) for removal of containment atmosphere radioiodine during a postulated post-LOCA environment.

In support of the September 24, 2012, amendment approving the extended power uprate of St. Lucie 2 (ADAMS Accession No. ML12235A463), the licensee conducted an evaluation that demonstrated that the post-LOCA sump pH will be maintained at or above 7. The licensee stated in the current license amendment request that the effects of hydrazine injection were not included in this analysis and that the pH will be sufficiently basic without considering the effects of hydrazine injection.

The NRC staff evaluated the license amendment request to determine whether the removal of the IRS, and therefore, the means to verify hydrazine injection during a postulated post-LOCA environment, will affect the ability of the post-LOCA sump water to retain radioiodine and prevent its re-evolution. The NRC staff notes that the licensee performed an analysis that demonstrated hydrazine injection has a negligible impact on post-LOCA sump water pH, radioiodine retention, and radioiodine re-evolution. Per SRP Section 6.5.2, chemical additivities (e.g., hydrazine) do not have a significant effect on aerosol particle removal, as this process is largely mechanical. As such, the licensee's proposed change will not impact its ability to meet the requirements of GDC 41. Therefore, the NRC staff finds that it is acceptable to remove the IRS requirements from the TSs, as hydrazine injection does not have a significant effect on radioiodine removal. Additionally, because hydrazine is a weak base and will not have a significant effect on the post-LOCA sump water pH, the NRC staff finds it acceptable to remove the IRS requirements from the TSs, as they pertain to maintaining post-LOCA sump water pH.

3.1.2 Corrosion of Materials In its license amendment request, the licensee stated that materials for American Society of Mechanical Engineers Code Class 2 and 3 components are selected for compatibility with the reactor coolant system and containment spray. The licensee also stated that following a LOCA, the sources of water will be pH neutral or slightly acidic. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the postulated accident, the water will be at a neutral pH in accordance with BTP 6-1. As a result, the probability of stress corrosion cracking will be reduced. Additionally, the licensee stated that hydrazine is not credited as a corrosion inhibitor and Jhat the maximum sump pH will be approximately 8.1.

The NRC staff evaluated the license amendment request to determine whether the removal of the IRS from the TSs, and therefore, the means to verify hydrazine injection during a postulated post-LOCA environment, will impact the potential for corrosion of materials in containment.

Because the water in the sump will be maintained at an approximately neutral pH without hydrazine injection, metallic materials in containment will not be significantly affected from a corrosion standpoint by the elimination of hydrazine injection. Additionally, as noted in BTP 6-1, "For the spray water recirculated from the containment sump, the higher the pH in the 7.0 to 9.5 range, the greater the assurance that no stress corrosion cracking will occur." This provides

reasonable assurance that the spray water will not induce stress corrosion cracking. Further, the NRC staff finds that the licensee's proposed change does not impact its ability to meet the requirements of GDC 14 for assuring the low probability of abnormal leakage or failure of the reactor coolant pressure boundary and safety-related structures. Therefore, the NRC staff finds that the licensee's request to remove the IRS requirements from the TSs is acceptable as it pertains to inhibiting corrosion of certain components within containment in a post-LOCA environment.

3.1.3 Radiological Consequences Regarding the evaluation of the impacts to radiological consequences of postulated accidents, the NRC staff reviewed the current licensing basis radiological consequences of design-basis accidents, as reflected in the St. Lucie 2 UFSAR, to determine if any inputs or assumptions depend on hydrazine and the IRS. The NRC staff's review determined that hydrazine is not credited for maintaining post-LOCA sump pH, nor is it credited in managing post-accident iodine in the radiological consequences of design-basis accidents. The radiological consequences of design-basis accidents rely upon the TSP baskets to control pH and the containment spray system without hydrazine for iodine removal in the containment; therefore, removal of the IRS from the TSs will have no effect on the. design-basis accidents radiological consequence analyses. Because the resultant radiological doses for the design-basis accidents remain unchanged, the regulatory limits stated in 10 CFR 50.67 and the dose limits specified in GDC 19 continue to be met.

3.1.4 Confirming Change to the TS Index The NRC staff finds that the licensee's request to make a conforming change to the TS index is editorial in nature and acceptable.

3.2 Revision to SR 4.5.2.e.4 The licensee proposed to modify SR 4.5.2.e.4 to allow the water used for testing to be representative of the refueling water tank instead of being from the refueling water tank. The proposed change would allow the licensee to use a prepared solution of borated water that is representative of the refueling water tank water to verify the ability of the TSP to buffer the solution.

In its letter dated May 17, 2019, the licensee provided the typical refueling water tank chemistry parameters. Based on these parameters, the licensee identified boron, in the form of boric acid, as the only significant contributor to pH and buffering. As such, the licensee stated that boron would be added to the solution to a concentration of 2100-2150 parts-per-million when preparing a sample representative of the refueling water tank water. This value is consistent with the TS-controlled boron concentration in the refueling water tank of 1900-2200 parts-per-million.

The NRC staff finds that the use of a water sample representative of the refueling water tank water is acceptable because the licensee will account for species that contribute to the solution pH and buffering when preparing the water sample. Therefore, the NRC staff finds that the licensee's request to modify SR 4.5.2.e.4 is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the NRC staff notified the State of Florida official (Ms. Cynthia Becker, M.P.H., Chief of the Bureau of Radiation Control, Florida Department of Health) on September 11, 2019 (ADAMS Accession No. ML19262F640), of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the use of facility components located within the restricted area as defined in 10 CFR Part 20 or a SR. 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, which was published in the Federal Register on January 2, 2019 (84 FR 24), that the amendment involves no significant hazards consideration, and there has been no public comment on such finding. 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 aforementioned considerations, 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: K. Bucholtz A. Chereskin Date: November 20, 2019

SUBJECT:

ST. LUCIE PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENT NO. 201 REGARDING TECHNICAL SPECIFICATION CHANGES TO ELIMINATE THE REQUIREMENTS OF THE IODINE REMOVAL SYSTEM (EPID L-2018-LLA-0301) DATED NOVEMBER 20, 2019 DISTRIBUTION:

PUBLIC PM File Copy RidsACRS_MailCTR Resource RidsNrrPMStLucie Resource RidsNrrLALRonewicz Resource RidsRgn2MailCenter Resource RidsNrrDorlLpl2-2 Resource RidsNrrDmlrMccb Resource RidsNrrDraArcb Resource RidsNrrDssStsb Resource KBucholtz, NRR AChereskin, NRR ADAMS A ccess1on N o.: ML19248C238 *b*Y memoran d um **b*Y e-ma1*1 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DMLR/MCCB/BC*

NAME MWentzel LRonewicz SBloom DATE 9/11/2019 9/10/2019 9/03/2019 II OFFICE NRR/DRA/AR CB/BC* NRR/DSS/STSB/BC(A)** OGC (NLO)**

NAME KHsueh MHamm DRoth DATE 5/02/2019 9/13/2019 10/02/2019 OFFICE NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME UShoop NJordan (MWentzel for)

DATE 11/19/2019 11/20/2019 OFFICIAL RECORD COPY