Medical Isotope P d ti F ilit Production Facility (MIPF) Construction Permit ApplicationPermit Application

NRC Public MeetingMarch 26, 2015

Our MissionProvide a reliable supply of medical di ti d th ti

Ou ss o

diagnostic and therapeutic radioisotopes in the United States

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AgendagTopic Presenter

Introduction Carmen Bigles

Project & Site Overview Caroline Schlaseman

Design Progress Veronica Garea

Environmental Report Progress Vanessa Newton

QA and SGI Programs Caroline Schlaseman

Questions NRC Staff PublicQuestions NRC Staff, Public

Conclusion Carmen Bigles

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Our TeamCoquí

U.S. company that will own and operate the MIPF

Ou ea

INVAPDesign Authority for the MIPF

Hogan LovellsNuclear regulatory and environmental attorneys

Gresham, Smith and Partners (GS&P)Consultant for siting, environmental data collection,

d i t l t (ER) tiand environmental report (ER) preparation

MPR AssociatesCoquí Owner’s Engineer and consultant

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Project &Project &Site OverviewSite Overview

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MIPF DescriptionFacility includes:

Two pool-type low-enriched

esc p oo poo ype o e c ed

uranium (LEU) reactorsRadioisotope processing plant

i i iWaste conditioning plantSupport services and administrative offices

Designed and constructed by INVAP with Designed and constructed by INVAP with proven technologies tailored to Coquí’sspecific needsspecific needsLEU targets produce Mo-99 and I-131Planned Mo-99 capacity of MIPF is 7,000 Planned Mo 99 capacity of MIPF is 7,000 six-day curies per week

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MIPF Site S eNorth Central FloridaAlachua CountyyTown of AlachuaProgress Corporate ParkZoned: “Corporate Park”Available skilled work f (U i it f Fl id force (University of Florida and high tech industry area)area)Nearest population center is Gainesville, FL 10 miles SEis Gainesville, FL 10 miles SE

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Attractive Site FeaturesAttractive Site FeaturesJanuary 2015 completed transaction with University of Florida Foundation for 25-acre parcel of land on which Coquí MIPF will be built

“The University of Florida Foundation is proud to beThe University of Florida Foundation is proud to be a partner in this exciting venture with Coquí RadioPharmaceuticals, a partnership that will provide opportunities for collaboration with UFprovide opportunities for collaboration with UF engineering and medical researchers, and bolster the development of high-paying jobs and infrastructure for a biotech research park in Alachua County.”

--Tom Mitchell, Vice President for Development and , pAlumni Affairs at the University of Florida

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Attractive Site FeaturesAttractive Site FeaturesKnowledgeable local emergency responders due to University of Florida Training Reactor (UFTR)Existing state emergency plan due to power reactors in Florida

Coquí Siteq~25 Acres

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Attractive Site FeaturesAttractive Site Features106 feet above mean sea levelTopography slightly slopedL t d i Located in Compact Region with reliable with reliable disposal of Class A, B, and C radioactive waste

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Administration BuildinggArtist’s Rendering

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Administration BuildinggArtist’s Rendering

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Design Progressg g

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INVAP Contract SignedINVAP Contract SignedNovember 2014 signed contract with INVAP to design MIPF in Alachua FL INVAP to design MIPF in Alachua, FL

Argentine nuclear engineering firm Over 30 years of nuclear development Over 30 years of nuclear development Over 15 reactors and related facilities across the world including production across the world, including production of medical isotopes:

OPAL reactor in AustraliaETRR-2 reactor in EgyptNUR reactor in Algeria

Coquí MIPF will use technology similar to

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Coquí MIPF will use technology similar to that employed in OPAL facility in Australia

INVAP Design MilestonesINVAP Design MilestonesDetailed Division of Responsibilities (DOR) Detailed Division of Responsibilities (DOR) for Preliminary Safety Analysis Report (PSAR) sections established(PSAR) sections established

Schedule of key milestones:1. Preliminary Design Review Meeting—April 20152. Critical Design Review Meeting—August 20153 Submittal of PSAR portion of Construction Permit 3. Submittal of PSAR portion of Construction Permit

Application 4th Quarter 2015

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Environmental Report pProgressg

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Environmental ReportEnvironmental ReportSite environmental monitoring performed Site environmental monitoring performed throughout last year and is continuingEnvironmental Report (ER) approximately Environmental Report (ER) approximately 85% completeER submittal planned for early 4th p yQuarter 2015 (~2 months prior to PSAR)ER to be subject of next NRC public j pmeeting in May-June timeframe

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QA and SGI QProgramsg

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Quality Assurance (QA)Quality Assurance (QA)Coquí Quality Assurance Program Coquí Quality Assurance Program Description (QAPD) developedMPR’s initial audit of INVAP’s QA program MPR s initial audit of INVAP s QA program complete

Program complies with ANSI/ANS 15.8 per gNUREG-1537No audit findings; two recommendations

I R 0 C í QAPD id A ilIssue Rev. 0 Coquí QAPD mid-AprilFuture audits to be conducted for procurement and construction phases

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procurement and construction phases

Safeguards Information (SGI) g ( )Program

Coquí Safeguards Information (SGI) P b i d l dProgram being developed

10 CFR Part 3710 CFR Part 7310 CFR Part 73

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QuestionsQ

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Conclusion

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ConclusionConclusion1. Land transfer from UF Foundation completed 1. Land transfer from UF Foundation completed

January 20152. Significant progress on MIPF design since last

NRC meeting in October 2014NRC meeting in October 20143. Environmental Report (ER) development on-

going and will be topic of next NRC meeting in May-June

4. Progress on “enterprise” programs, e.g., QA and SGI on-goingSGI on going

Planning for 4th Quarter 2015 submittal of C í MIPF C t ti P it A li ti

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Coquí MIPF Construction Permit Application

Closed Meeting Presentation #1(redacted for public record)

Nuclear Design for COQUI Reactor

1

26 March 2015 1

Preliminary design

6 FA + 42 MoBe+ ~100 Be FA 100 Be FA

Mo + BeBe

2

Core designOPAL COQUI

5 CR - Inside Control Guide BoxOPAL COQUI

4 CR - Inside Control Guide Box

Zircalo chimne16 FA/21 FPZircaloy chimney Zircaloy chimney 6 FA/24 FP

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Fuel Assembly data

FA t MTR U Si t Fl t ll l l t

y

FA type MTR - U3Si2 meat – Flat parallel platesUranium density [g/cm3] 4.8235U enrichment 19.75 wt%Clad material Al 6061Clad material Al 6061Coolant channel gap [mm] 2.45

COQUI OPALNumber of plates 24 21

22 i l t 19 i l tFuel plate dimensions [mm](thickness x width x height)

22 inner plates 1.40 x 75 x 5502 outer plates

1.55 x 75 x 550

19 inner plates 1.35 x 75 x 6502 outer plates

1.50 x 75 x 6501.55 x 75 x 550 1.50 x 75 x 650Fuel meat dimensions [mm](thickness x width x height) 0.67 × 65 × 500 0.61 × 65 × 615

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Molybdenum Targetsy g

COQUI OPALMo production (6-days Ci) 7000 ~ 2000Total power in Mo plates (MW) 2 8 1 0Total power in Mo plates (MW) 2.8 1.0# of irradiation positions 42 12Total # of plates 420 96 (192 equiv.)p ( q )q”max (W/cm2) 200 120Minimum velocity in Mo (m/s) 6.0 3.5Flow direction DownwardMaximum allowable DP (kPa) 70

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Primary Coolant System (Forced Convection)

PCS flow Closure flow

Flap valves ClosedOutlet pipe

f

Flap valves Closed

Inlet pipe ChimneyInlet pipe Chimney

FA + CR + CRGBInlet plenum

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Shutdown Mode (Natural Circulation)

Pool

Flap valves OpenOutlet pipe

Flap valves Open

Inlet pipe ChimneyInlet pipe Chimney

FA + CR + CRGBInlet plenum

No flow reversal

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Reflector and Irradiation positionsOutlet pipe COQUIp p

OPALCOQUI

Riser

Irradiation positionsIrradiation positions

D2O tank

Reflector

Be blocks

Neutron beamsReflector

NO Neutron beams

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Thermal-hydraulic data

COQUI OPAL

Primary Coolant SystemCOQUI OPAL

Power /FA (MW) 1.13 1.25

Volumetric Power q”’(W/cm3) 295 306Volumetric Power q (W/cm3) 295 306

Surface Heat Flux q”max (W/cm2) 240 220

Minimum velocity (m/s) 8 2 8 6Minimum velocity (m/s) 8.2 8.6

Flow direction Upward Upward

Total flowrate (m3/h) 1000 2000Total flowrate (m /h) 1000 2000

Total core DP (kPa) 200 230

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Primary Coolant System

100 m3/h

1000 m3/h

900 m3/h

100 m3/h to RSPCS

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Thermal-hydraulic data

Reactor Pool & Service Pool Coolant SystemTotal power in Mo plates (MW) 2 8Total power in Mo plates (MW) 2.8Surface Heat Flux q”max (W/cm2) 200Minimum velocity in Mo (m/s) 6.5y ( )Flow direction DownwardTotal Be + Mo flowrate (m3/h) ~ 1500Maximum allowable DP (kPa) 70

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Reactor Pool & Service Pool Coolant System

Outlet pipe

h ldBe + Mo holder

Outlet plenum B flOutlet plenum Be reflector

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Reactor Pool & Service Pool Coolant System

1500 m3/h

3/1600 m3/h

From PCS 100 m3/h

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Core preliminary design

Be + Mo holder

BeBe

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AcronymsFA = Fuel Assembly

Mo = Target (Molybdenum)

Be = Beryllium

CR = Control Rod

FP Fuel PlateFP = Fuel Plate

PCS = Primary Coolant System

RSPCS = Reactor & Service Pools Coolant System

CRGB = Control Rod Guide Box

q”’ = Volumetric Powerq = Volumetric Power

q” max = Surface Heat Flux

DP = Differential Pressure

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F ll O Di i Follow-On Discussion Topics from Previous Topics from Previous

Meeting with NRCgClosed Meeting Presentation

#2#2(redacted for public record)

March 26, 2015

Discussion Topics

Follow-On Discussion TopicsFollow On Discussion Topics1. Downward flow through fuel assemblies;

now upward flownow upward flow2. Critical Heat Flux (CHF)3. Power Densityy4. LOCA5. Maximum Hypothetical Accident (MHA),

including Ventilation

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Safety Design Criteria

Critical Heat Flux RatioComponents of Safety Marginp y g

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Safety Design Criteria

Critical Heat Flux Ratio“Safety limits should preclude flow Safety limits should preclude flow instabilities in the hottest channel and ensure that the minimum departure from

l t b ili ti (DNBR) i t l t nucleate boiling ratio (DNBR) is at least 2.0 (which has been an acceptable margin to the onset of nucleate boiling).”g g)

NUREG 1537 Part 1, page 4CHF ratio will normally be ~ 2.2No nucleate boiling at any time duringsteady-state operationFlow oscillation avoided

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Flow oscillation avoided

Power Density – Fuel

Similar to OPALU d fl i F l Upward flow in Fuel:

No flow reversal/instabilities

COQUI OPALPower /FA (MW) 1.13 1.25

Volumetric power q´´´ (W/cm3) 295 306

Surface heat flux q”max (W/cm2) 240 220

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Power Density – Targets

Similar to OPALD d fl i T tDownward flow in Targets:

COQUI OPALTotal power (MW) 2.8 1.0

Surface heat flux q”max (W/cm2) 200 120

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LOCALOCA

LOCA“In many non-power reactor designs, the loss-of-

l t id t ( OC ) i f coolant accident (LOCA) is of no consequence because decay heat in the fuel is so small as to be incapable of causing fuel failure. In some higher power reactors (normally greater than 2 MW), an engineered reactors (normally greater than 2 MW), an engineered safety feature, such as an emergency core cooling system, may need to be operable for some time after reactor shutdown to remove decay heat in the event of a LOCA Some initiators of LOCAs are the following:of a LOCA. Some initiators of LOCAs are the following:1. failure or malfunction of some component in the primary

coolant loop2. failure or malfunction of an experimental facility, such as 2. failure or malfunction of an experimental facility, such as

a beam tube3. failure or leak of the reactor coolant boundary”

NUREG 1537 Part 1, Chapter 13.1.3

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LOCALOCA

LOCA

1. Failure or malfunction of some 1. Failure or malfunction of some components in the PCS

All process penetrations to the pool are p p pat Level 7m (23 ft) or aboveLow pressure, low temperature, low

di ti fi ldradiation fieldFailure of pump casing, seals, unions is considered within the design basisconsidered within the design basisSiphon Effect Breakers (SEB) Flap Valves

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Flap Valves

LOCALOCA

LOCA

2 Failure or malfunction of an 2. Failure or malfunction of an experimental facility, such as a beam tubebeam tube

No beam tubesN i t l f ilitiNo experimental facilitiesTargets and reflector cooled by RSPCS

M h i l d i i t id ti l Mechanical design requirements identical to PCS, pool penetration at 7m (23ft), etc.

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LOCALOCA

LOCA

3 Failure or leak of the reactor 3. Failure or leak of the reactor coolant boundary:

L k th h l3.1 Leak through pool

3.2 Leak through balance of PCS

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LOCALOCA

LOCAReactor Pool (tank)( )

Service Pool (tank)(tank)

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LOCALOCA

LOCA

[insert figure of Pool and Concrete Block][insert figure of Pool and Concrete Block]

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LOCALOCA

LOCA

3.1Leak through pool (continued): 3.1Leak through pool (continued): Pool is a free-standing stainless steel, cylindrical tank, not a linery ,Designed to ASME B&PV Code, Section IIISeismic lateral loads carried bySeismic lateral loads carried bysurrounding reinf. concrete structure100% weld inspectionpQualified weldersLeak detection system

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Leak detection system

LOCALOCA

LOCA

3 1Leak through pool (continued): 3.1Leak through pool (continued): CRD penetrations in bottom of pool are accessible via small room below poolaccessible via small room below pool

Small penetrations, proven sealsLow leak rate if seal(s) fail( )Water tight CRD room remains closed unless CRD maintenance is requiredIf complete failure of CRD penetrations is assumed, flooding of CRD room does not uncover the core

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uncover the core

LOCALOCA

LOCA

3 2 Leak through balance of PCS:3.2 Leak through balance of PCS:Consequence of failures is not significant, due to low pressure and temperature due to low pressure and temperature and minimum radioactivityFailure of seals, unions, pump casing: a u e o sea s, u o s, pu p cas g: considered within the design basis

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MHA

Maximum Hypothetical Accident“This limiting accident is named the This limiting accident is named the maximum hypothetical accident (MHA) for nonpower reactors; the details are reactor specific. Because the MHA is not expected to occur, the scenario need not be entirely credible The initiating event and the credible. The initiating event and the scenario details need not be analyzed, but the potential consequences should be analyzed and evaluated.”

NUREG 1537 Part 1, Chapter 13, p 13-2

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MHA – Reactor

Maximum Hypothetical AccidentReactor scenarioseac o sce a os1. A specified fraction of fuel in the core melts. 2. Cladding is stripped from a specified fraction of

the core fuel plates or elements.3. The fuel encapsulation bursts, releasing gaseous

fission products to the pool or the air fission products to the pool or the air. 4. A fueled experiment melts or fails

catastrophically in the pool or in the air.NUREG 1537 Part 1, Chapter 13, p 13-3

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MHA – Reactor

MHA – Scenario Selection1. A specified fraction of fuel in the core melts. spec ed ac o o ue e co e e s

Can be used to bound other scenarios2. Cladding is stripped from a specified fraction of

fthe core fuel plates or elementsBounded by #1

3 The fuel encapsulation bursts releasing gaseous 3. The fuel encapsulation bursts, releasing gaseous fission products to the pool or the air

Not applicable4. A fueled experiment melts or fails

catastrophically in the pool or in the airApplicable to Mo targets

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Applicable to Mo targetsIn pool bounded by #1; in air not credible

MHA - Reactor

Fuel and Target Failure in Air Prevented

[insert figure of arrangement of pools][insert figure of arrangement of pools]

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Specified fraction of one MHA – Reactor

Specified fraction of one core’s fuel and targets meltsg

Accident AssumptionsOperating cycle lengthFull initial powerFull initial powerMaximum target inventory availableTiming of fuel melt and credit for decay after scram

Release Assumptions100% of noble gases of affected fuel & targetsFraction of Cs & I retain by melt or scrubbed by poolFraction of Cs & I retain by melt or scrubbed by poolMinimum distance of about X meters to publicConservative, site appropriate meteorology

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R t MHA S MHA – Reactor

Reactor MHA Summary Acceptance Criterion: staff and public dose does

t d ti 10 CFR P t 20 li itnot exceed respective 10 CFR Part 20 limitsAccident Scenario: Specified fraction of one core’s fuel and targets melts underwaterg

Bounds other scenariosRadioisotope propagation from fuel conservativeRadioisotope release to environment conservativeRadioisotope release to environment conservative

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MHA – RPP

RPP MHA SummaryAcceptance Criterion: staff and public Acceptance Criterion: staff and public dose does not exceed respective 10 CFR Part 70 limitsAccident Scenario: Inventory driven

Still being defined based on evolution of design of the production linesRelease of radioactive materials throughventilation system

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MHA – WCP

WCP MHA SummaryAcceptance Criterion: staff and public Acceptance Criterion: staff and public dose does not exceed respective 10 CFR Part 70 limitsAccident Scenario: Inventory driven

Still being defined based on the evolutiondesign of the WCP

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AcronymsFA = Fuel Assemblyy

Mo = Target (Molybdenum)

Be = Beryllium

CRD = Control Rod Drive

FP = Fuel Plate

PCS = Primary Coolant SystemPCS = Primary Coolant System

RSPCS = Reactor & Service Pools Cooling System

q”’ = Power Density

q” max = Surface Heat Flux

DP = Differential Pressure

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