ML17089A235

30441R00032, Revision B, Relap Accident Analysis and Fraptran Target Rod Transient Analysis Design Calculation Report

ML17089A235
Person / Time
Site:	University of Missouri-Columbia
Issue date:	01/27/2017
From:	Bolin J, Hon R General Atomics
To:	Office of Nuclear Reactor Regulation, Nordion (Canada), US Dept of Energy, National Nuclear Security Admin, Univ of Missouri - Columbia
References
DE-NA0002773 30441R00032, Rev B
Download: ML17089A235 (70)

Text

"

I

';;t: '

ATTACHMENT 7 30441 R00032 Revision B I

REACTOR-BASED MOLYBDENUM-99 SUPPLY SYSTEM PROJECT RELAP ACCIDENT ANALYSIS AND FRAPTRAN TARGET ROD TRANSIENT ANALYSIS DESIGN CALCULATION REPORT Prepared by General Atomics for the U.S. Department of Energy/National Nuclear Security Administration and Nordion Canada Inc.

Cooperative Agreement DE-NA0002773 GA Project 30441 WBS 1160

+GENERAL ATOMICS

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report REVISION HISTORY Revision Date Description of Changes A 01NOV16 Information Issue Initial Release B 27JAN17 RELAP5 model revision and new accident analyses POINT OF CONTACT INFORMATION PREPARED BY:

Name Position Email Phone J. Bolin Engineer John.Bolin@ga.com 858-455-2467 R. Hon Engineer Ryan.Hon@ga.com 858-455-4374 APPROVED BY:

Name Position Email Phone B. Schleicher Chief Engineer Bob.Schleicher@ga.com 858-455-4733 K. Murray Project Manager Katherine.Murray@ga.com 858-455-3272 K. Partain Quality Engineer Katherine.Partain@ga.com 858-455-3225 DESIGN CONTROL SYSTEM DESCRIPTION

~

R&D DISC QA LEVEL SYS DV&S D DESIGN D T&E N I NIA D NA ii

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B TABLE OF CONTENTS REVISION HISTORY .......................*............................................................................................ii POINT OF CONTACT INFORMATION .*.*.******..*......................*..........................................*........ii DESIGN CONTROL SYSTEM DESCRIPTION *...**.*........*.....*....*..................*.......****....**...****..*.* ii ACRONYMS *......*.......*..*.*....*.*..**.*.*.**..*.*.....*.*.*...*..**..**.....*.*...***...*..........................*................* vi 1 OBJECTIVE .............**............*...................*..*.......**....*.*..........................................*.....*.. 1 1.1 Analysis Objectives for Reactivity Transients ............................................................ 1 1.2 Analysis Objectives for Loss of Target Coolant and Loss of Target Flow ................. 1 2 APPLICABLE DOCUMENTS ........................................................................................... 3 3 METHODS ....................................*..................*....*...........................................................3 3.1 RELAP5 ..................................................................................................................... 3 3.2 FRAPCON ................................................................................................................. 4 3.3 FRAPTRAN ............................................................................................................... 5 4 ASSUMPTIONS ....*............................*....*..**................*.*.........................*..*....*.*......*.*...*. 6 4.1 RELAP5 Generic Code Assumptions ........................................................................ 6 4.2 FRAPCON Generic Code Assumptions .................................................................... 6 4.3 FRAPTRAN Generic Code Assumptions .............................. ~ ............... ;................... 8 4.4 Reactivity Transient Model Assumptions .................................................................. 9 4.5 Loss of Target Coolant Model Assumptions ............................................................ 10 4.6 Loss of Target Flow Model Assumptions ................................................................ 10 5 INPUTS .........................*...*.*...*.*.........*.......*.................................................................*. 10 5.1 Target Pellet, Cladding and Cartridge Data ............................................................ 10 5.2 Target Steady State Power Distribution .................................................................. 14 5.3 RELAP5 Hydrodynamic Components ..................................................................... 21 5.4 RELAP5 Pump Model ............................................................................................. 28 5.5 RELAP5 Heat Exchanger Model ............................................................................. 28 5.6 Steady State Inputs from RELAP ............................................................................ 29

5. 7 Reactivity Transient Power History ......................................................................... 32 6 RESULTS AND CONCLUSIONS ....**.*....*..*....*...*..**..*..*....................*.....*..........***.*..**.* 33 6.1 Rapid Insertion of Positive Reactivity ...................................................................... 33 6.2 Control Blade Withdrawal ........................................................................................ 38 6.3 Loss of Target Cciolant ............................................................................................ 42 6.3.1 Pipe Break out of the Reactor Pool .................................................................43 6.3.2 Pipe Break in the Reactor Pool ....................................................................... 50 6.4 Loss of Target Flow ................................................................................................. 56 7 CALCULATIONS FILES .................................................................................................. 58 7.1 RELAP5 Files .......................................................................................................... 58 7 .2 FRAPCON Files ...................................................................................................... 59 7 .3 FRAPTRAN Files .................................................................................................... 60 7.4 EXCEL Files ............................................................................................................ 61 7.5 MATHCAD Files ................................................................................................*..... 61 8 REFERENCES **.*.***....*****.......*****..**.*.**.*...**.*...*.*.*.*.**..*..*..*.....*.*.*.......**....*.*..*...........*.. 63 iii

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report LIST OF FIGURES Figure 1. Target cooling system ................................................................................................... 2 Figure 2. Target pellet dimensions ............................................................................................. 11 Figure 3. Cladding dimensions .................................................................................................. 11 Figure 4. Axial power factor for target rod I ............................................................................... 15 Figure 5. Axial power factors for target rod 1............................................................................. 16 I I I Figure 6. Axial power factors for target rods thru and thru ............................................ 17 Figure 7. Axial power factors for target rod * ........................................................................... 18 Figure 8. Axial power factors for target rod ..................................................................~ ......... 19 Figure 9. Axial power factors for target rods* thru

• and

• thru * * .................................... 20 Figure 10. Reactor power, fuel and cladding temperatures vs. time for a positive reactivity step insertion of 600 pcm ........................................................................................................ 34 Figure 11. Cladding strains at peak pellet location during 600 pcm reactivity insertion ............. 36 Figure 12. Peak target pellet temperature during 600 pcm reactivity insertion .......................... 37 Figure 13. Pellet OD and cladding temperatures at peak pellet location during 600 pcm reactivity insertion ........................................................................................................... 38 Figure 14. Power transient during 30 pcm per second reactivity insertion ................................ 39 Figure 15. Cladding strains at peak pellet location during 30 pcm per second reactivity insertion40 Figure 16. Peak target pellet temperature during 30 pcm per second reactivity insertion ......... 41 Figure 17. Pellet OD and cladding temperatures at peak pellet location during 30 pcm per second reactivity insertion ............................................................................................... 42 Figure 18. Pipe break locations out of the reactor pool ............................................................. 43 Figure 19. Mass flow transient during LOCA out of the reactor pool without target DHRS valves45 Figure 20. Target power during a LOCA out of the reactor pool without target DHRS valves .. .46 Figure 21. Coolant temperatures during a LOCA out of the reactor pool without target DHRS valves ..............................................................................................................................47 Figure 22. Maximum cladding ID temperatures during a LOCA out of the reactor pool without target DHRS valves .........................................................................................................48 Figure 23. Mass flow transient during a LOCA out of the reactor pool with target DHRS valves49 Figure 24. Maximum cladding ID temperatures during a LOCA out of the reactor pool with target DHRS valves ......................................................................................................... 50 Figure 25. Pipe break location in reactor pool ................. ~ ......................................................... 51 Figure 26. Mass flow transient during a LOCA in the reactor pool ............................................ 52 Figure 27. Target power during a LOCA in the reactor pool ...................................................... 53 Figure 28. Maximum cladding ID and coolant temperatures during a LOCA in the reactor pool54 Figure 29. Maximum cladding OD temperature profile in rod *during a LOCA in the reactor pool ................................................................................................................................. 55 Figure 30. *Cladding fractional strains in rod *during a LOCA in the reactor pool ................... 55 Figure 31. Mass flow transient during loss of pump flow ........................................................... 57 Figure 32. Maximum cladding ID and coolant temperatures during loss of pump flow .............. 58 iv

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B LIST OF TABLES Table 1 Target Pellet and Cladding Dimensions ........................................................................ 12 Table 2 Additional Target Pellet Dimensions ............................................................................. 12 Table 3 Target Spring and Plenum Data ................................................................................... 13 Table 4 Flow Areas and Diameters for Target Rods .................................................................. 13 Table 5 Target Rod Power ......................................................................................................... 20 Table 6 RELAP5 Volumes ......................................................................................................... 21 Table 7 RELAP5 Junctions with Non-Zero K-Loss Factors ....................................................... 25 Table 8 RELAP5 Pump Parameters .......................................................................................... 28 Table 9 RELAP5 Heat Exchanger Parameters .......................................................................... 29 Table 10 Cladding Surface Temperature at 100% Power and 100% Flow in Kelvin ................. 30 Table 11 Cladding Surface Temperature at 115% Power and 100% Flow in Kelvin ................. 30 Table 12 Cladding Surface Temperature at 115% Power and 85% Flow in Kelvin ................... 31 Table 13 Target Rod* Coolant Pressure in Pascals ............................................................... 32 Table 14 Target Rod* Mass Flux ............................................................................................ 32 Table 15 Excel Files ...................................................................................................................61 v

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report ACRONYMS Acronym Description ANL Argonne National Laboratory BOL Beginning of Life CHF Critical Heat Flux CHFR Critical Heat Flux Ratio DHRS Decay Heat Removal System EOL End of Life GA General Atomics HX Heat Exchanger ID Inner Diameter LEU Low Enriched Uranium LOCA Loss of Coolant Accident LOFA Loss of Flow Accident LOPF Loss of Pump Flow LSSS Limiting Safety System Setting LWR Light Water Reactor MTU Metric Ton Uranium MURR University of Missouri Research Reactor OD Outer Diameter pcm per cent mille =10-s PNNL Pacific Northwest National Laboratory SAR Safety Analysis Report SGE Selective Gas Extraction SI Systeme International (metric units)

TA Target Assembly TCS Target Cooling System USN RC United States Nuclear Regulatory Commission vi

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B 1 OBJECTIVE The objective of this report is to analyze and present the results of transients and potential accidents associated with the irradiation of target rods for the production of the medical isotope Mo-99. The target rods utilize low enriched uranium (LEU) to produce Mo-99 as a fission product, which is then extracted from the target rod post-irradiation for use in medical diagnostic procedures. The LEU targets will be irradiated in the reflector region of the University of Missouri Research Reactor (MURR). The target assembly to be installed in the reflector region will be equipped with its own independent cooling system, both for normal cooling of the rods during routine operation, as well as a natural circulation cooling system that will provide the required cooling to the LEU target rods in the case of a loss of flow or loss of coolant accidents.

The normal cooling system draws and discharges its cooling water from and to the MURR pool.

1.1 Analysis Objectives for Reactivity Transients Two reactivity transients were analyzed: rapid insertion of positive reactivity, and control blade withdrawal. The reactivity transient is driven by the MURR core and is defined in the MURR Safety Analysis Report (SAR) submitted for relicensing (References 1 and 2). The steady-state conditions of the target rod are calculated by FRAPCON which determines thermal expansion, pellet-clad interaction, pellet relocation, and fission gas release as described in Section 3.2.

While FRAPCON can calculate the thermal boundary condition of the target rod , it was determined that heat transfer calculation assumes that the water properties are at the high pressures and temperatures typical of light water reactors and therefore not suitable for the low pressures and temperatures within the target assembly. For this reason, the thermal boundary conditions in FRAPCON are defined by steady state RELAPS runs at the appropriate power and flow conditions. FRAPCON results are then used as input to a transient FRAPTRAN run which simulates the reactivity transient. The objectives of the FRAPTRAN analysis are to assess cladding strain due to internal pressure and pellet-clad interaction, to assess peak pellet temperature, and to determine minimum critical heat flux ratio.

1.2 Analysis Objectives for Loss of Target Coolant and Loss of Target Flow The loss of target coolant and loss of target flow accidents are analyzed with a RELAPS model that includes both target assemblies (30441000211 , 30441000207, 30441000210, 30441000308, 30441000313, 30441 R00031 ), the reactor pool , the complete target cooling system, and the target decay heat removal system (30441 R00019). The model includes the target cooling system pump and heat exchanger, and target decay heat removal system (OHRS) valves needed for the transition from normal to natural circulation cooling flow. The target cooling system is depicted in Figure 1. The RELAPS analyses are used to assess the performance of the cooling systems both with and without the target OHRS valves operating, to 1

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report assess changes in cladding and pellet temperatures, and to determine minimum critical heat flux ratio (CHFR) using thermal hydraulic results from the accident simulation.

MURR Secondary Coolant System Connections Target Cooling Pumps Target Cooling Heat Exchangers Target Decay Heat Flexible Removal Valves .--- couplings Pump Discharge Pipes Typical 3" ID Size Pipes

~-- Flex Lines

~-- Target Assemblies Figure 1. Target cooling system 2

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report 2 APPLICABLE DOCUMENTS A list of applicable documents is provided below.

30441 R00031 3 METHODS 3.1 RELAP5 The transient analysis code RELAP5 (Reference 3) has been developed for the United States Nuclear Regulatory Commission (USNRC) to provide analysis to support rulemaking, licensing audit calculations, evaluation of accident mitigation strategies, evaluation of operator guidelines, and experiment planning analysis. The version RELAP5/MOD3.3 was developed jointly by the NRC and a consortium of several other organizations with the goal of developing a code suitable for the analysis of all transients and accidents in light water reactor (LWR) systems.

RELAP5 is a standard tool used by the nuclear industry for the analysis of LWR systems. The code models the behavior of the reactor core and cooling system for various transient situations including loss of coolant, transient without scram, loss of offsite power, loss of feed water, and loss of flow. There has been extensive verification and validation (Reference 3) of the code which gives confidence in the quality of its results. While developed for light water reactor 3

ATTACHMENT 7 ..,

RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B analysis, the code is also general enough to simulate a wide variety of thermal hydraulic systems containing mixtures of water, steam, noncondensables, and solute.

The hydrodynamic model in RELAPS is a one-dimensional, transient, two-fluid model for flow of a two-phase steam-water mixture that can contain noncondensable components in the steam phase and a soluble component in the water phase. The two-fluid equations of motion used in the model are formulated in terms of volume and time averaged parameters of the flow.

Transverse flow effects, such as friction and heat transfer, are formulated in terms of bulk properties using empirical transfer coefficient formulations. Additionally specific models have been developed for particular flow situations. The system model is solved numerically using a semi-implicit finite-difference technique.

For heat transfer across boundaries, RELAPS makes use of heat structures. Heat structures are represented by one-dimensional heat conduction in rectangular, cylindrical, or spherical geometry. The code is able to model a variety of heat transfer situations including fuel pins with nuclear heating, conduction between pipe walls, and heat transfer across steam generator tubes. The time dependence of heat sources can be inputted by the user or obtained from reactor kinetics and various boundary conditions can be used. The heat structures surfaces connected to hydrodynamic volumes contain correlations for convective, nucleate boiling, transition boiling, and film boiling heat transfer from the wall to the water as well as heat transfer from the water to the wall including condensation.

Control systems are provided within the code to model events in transient situations.

Parameters at certain points in the system can be used to cause events elsewhere to happen.

For example these features can be used to cause a pump to trip or a valve to open at a certain point in time.

The power behavior of a nuclear reactor can be modeled using the included point reactor kinetics model. This model computes both immediate power from fission and neutron moderation as well as decay power from the fission products undergoing radioactive decay. The model is adequate in situations where the spatial power distribution remains constant. The user has a choice of which decay power model to use. A detailed description of the models and methods used in RELAPS can be found in the user's manual (Reference 3). GA has verified the code for use on this project ).

3.2 FRAPCON FRAPCON-4.0 (Reference 4) is a steady state fuel performance code developed by Pacific Northwest National Laborato,.Y (PNNL) for the USNRC. The code has been developed for calculating LWR fuel behavior up to a rod average burnup of 62 gigawatt days per metric ton of uranium (GWd/MTU). The code calculates only "steady-state" situations, including slow power 4

ATTACHMENT7 RELAP Accident Analysis and .FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B ramp ups and long periods of constant power, typical to that of LWR operation. A number of parameters significant for fuel rod analysis are calculated by the code including fuel and cladding temperatures, cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, fuel densification, fission gas release, and rod internal gas pressure. The code can also be used to generate the initial conditions required for transient fuel behavior analysis with the FRAPTRAN code.

FRAPCON uses an iterative process to calculate all of the interrelated effects of fuel rod performance. A typical calculation will start by computing the initial conditions of the problem from the inputs given. The temperature distribution throughout the fuel rod will then be computed. Given the temperature distribution, any deformations to the fuel and cladding will be calculated: The deformation of the fuel and cladding will cause differences in the rate of heat transfer throughout the rod, requiring the fuel temperature distribution and deformation to be iterated upon until the gap temperature difference has converged. Next the fission gas release and gas pressure are computed for the rod. This change in gas pressure will have an effect on the temperature distribution of the rod. Another iterative loop is used until the gas pressure in the rod is converged after which time is advanced to the next step.

The material properties used in FRAPCON come from a database that relates important system and material properties to one anoth~r. For example, the fuel thermal conductivity is a function of temperature, density and bumup and the cladding stress-strain relationship is a function of temperature, strain rate, cold work, and neutron fluence. There are various models and methods that the code uses in order to calculate pellet redistribution, gas release, crud disposition, etc.,

some of which the user can specify for specific problems. A detailed description of the methods and models used in FRAPCON can be found in the user's manual (Reference 4). GA has verified the code for use on this project (30441 R00025).

3.3 FRAPTRAN FRAPTRAN-2.0 (Reference 5) is a code developed by PNNL for the USNRC that calculates the performance of LWR fuel rods during transient and accident situations. These accident conditions for pressurized water reactors include reactivity-initiated accidents and loss-of-coolant accidents. For these situations, FRAPTRAN models heat conduction, heat transfer *from cladding to coolant, elastic/plastic fuel and cladding deformation, cladding oxidation, fission gas release, and fuel rod gas pressure. FRAPTRAN does not model fuel densification, fuel swelling, cladding creep, and cladding irradiation growth, which vary slowly with time. These parameters are obtained from the FRAPCON code, which uses the same materials database, to set the conditions at the start of the transient.

In order to calculate a solution, FRAPTRAN begins with calculating the temperatures of the fuel and cladding followed by the fuel rod gases. Next the stresses and strains on the fuel and 5

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report cladding are calculated. Then the fission gas .release and gas pressure are calculated. Since the temperatures, pressures, and stresses are all related to one another they are iterated upon until the temperature profile converges. After the temperature converges cladding oxidation and ballooning are calculated.

FRAPTRAN uses finite difference techniques to calculate the variables which influence fuel rod performance. These variables are calculated at different user defined radial and axial positions with different axial positions taken to be independent of one another (stacked one-dimensional solution). A detailed description of the methods and models used in FRAPTRAN can be found in the user's manual (Reference 5). GA has verified the code for use on this project 4 ASSUMPTIONS 4.1 RELAP5 Generic Code Assumptions RELAP5 is designed for use in analyzing the interactions of system components and does not provide detailed simulations of fluid flow within components. It therefore has limited capabilities in modeling multi-dimensional effects in heat transfer, fluid flow, and reactor kinetics. Some of the major assumptions present in the code are as follows:

• One dimensional flow with cross-flow junctions to allow some multi-dimensional effects

• Limited geometry specification of components, with a coarse nodalization scheme

• One dimensional heat flux

• Point reactor kinetics For a more detailed analysis of all the assumptions involved with RELAP5 the reader is encouraged to consult the code manual (Reference 3).

4.2 FRAPCON Generic Code Assumptions FRAPCON-4.0 has been developed specifically for the analysis of oxide fuel rods in light water reactors. As such there are some limitations inherent within the code due to the scope of the problem analyzed. The following are a list of those limitations:

• The code is limited to uranium oxide fuel types namely: uranium dioxide (U02), mixed oxide ((U,Pu)02), urania-gadolinia (U02*Gd203), and uranium dioxide with zirconium diboride (ZrB2) coatings.

• The cladding is limited to Zircaloy-2, Zircaloy-4, MS, ZIRLO, and Optimized ZIRLO.

6

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

• The conditions of the system analyzed should be similar to those found in typical light and heavy water reactor systems.

• The code utilizes models based on steady state conditions.

• Gas release models are based upon slow power ramp data on the order of 0.1 days or greater and do not reflect the release rates for situations with rapid power changes.

Rapid power changes are analyzed using FRAPTRAN.

• Only small cladding deformations can be calculated (<5% strain).

Additionally the code has been found to slightly over predict the cladding strain when under 65 GWd/MTU burn up (Reference 4 ).

For the calculation of heat transfer within the code, the following assumptions are as used:

• Heat conduction is ignored in the axial direction. This is a common assumption for systems with a large length to diameter ratio.

• Heat conduction in the azimuthal direction is ignored. This is a common simplification in nuclear safety analyses.

• During each time step boundary conditions are held constant.

• Steady-state flow is assumed.

• Fuel rod is modeled as a right circular cylinder surrounded by water as a coolant.

• The radial power profile within a fuel pellet is based off of fuel type, reactor type, and burnup.

For the cladding deformation model the following assumptions are used:

• Incremental theory of plasticity

• Prandtl-Reuss flow rule

• Isotropic work-hardening

• Thick wall cladding (thick wall approximation formula is used to calculate stress at mid-wall)

• No axial slippage at the fuel/cladding interface 7

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

• Bending strains and stresses in cladding are negligible and beyond the scope of the computer code

• Axisymmetric loading and deformation of cladding The fuel deformation model is based upon the following assumptions:

• Thermal expansion, swelling, and densification are the only sources for fuel deformation

• No resistance to expansion of fuel

• No creep deformation of fuel

• Isotropic fuel properties Lastly, the assumptions specific to the internal gas pressure model are:

• The perfect gas law is applicable (PV =nRT).

• Gas pressure is the same throughout the fuel rod due to interconnectecj gaps and cracks.

• Gas in the fuel cracks is at the average fuel temperature.

4.3 FRAPTRAN Generic Code Assumptions FRAPTRAN-2.0 uses the results from FRAPCON to obtain starting conditions for transient calculations and has many of the same assumptions. However, new assumptions are needed as the system analyzed is no longer at steady-state. The major assumptions of the FRAPTRAN code not included in, or different from the FRAPCON assumptions, are as follows.

For the heat transfer model the following assumptions are used for the transient calculation in addition to the steady state assumptions from FRAPCON:

• Steady-state critical heat flux (CHF) correlations are assumed to be valid during transient conditions.

• Steady-state cladding surface heat transfer correlations are assumed to be valid during transient conditions.

• The coolant in the system is. water or another coolant that can be modeled with altered heat transfer coefficients.

8

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B For the calculation of the plenum temperature for transient conditions the model is based on the following assumptions:

• The temperature of the top surface of the fuel stack is independent of the plenum gas temperature.

• The plenum gas is well mixed by natural convection.

• Axial temperature gradients in the spring and cladding are small.

The clad deformation model has the following assumptions different from FRAPCON:

• No low-temperature creep deformation of cladding

• Thin wall cladding (uniform temperature, stress and strain)

Lastly, the transient fuel rod gas pressure model is based off of the following assumptions:

• The gas behaves as a perfect gas.

• The gas flow past the fuel column is a quasi-steady process.

• The gas flow is compressible and laminar.

• The gas flow past the fuel column ean be analyzed as Poiseuille flow (force balance only).

• Gas expansion in the plenum and ballooning zone is an isothermal process.

• The entire fuel-cladding gap can be represented as one volume containing gas at a uniform pressure.

• The flow distance is equal to the distance from the plenum to the centroid of the fuel-cladding gap.

• The minimum cross-sectional area of flow is equivalent to an annulus with inner radius equal to that of the fuel pellet radius and a radial thickness of 25 µm (0.98 mil).

4.4 ReactiVity Transient Model Assumptions For both reactivity transients, the power transient in the target rods is assumed to be proportional to the power transient in the MURR core. The power distribution in the target rods during the reactivity transients is assumed to be the same as the steady-state power 9

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report distribution. The reactivity transients Were assumed to occur either at the beginning or end of a

- irradiation in which mid-week shutdowns also occur.

4.5 Loss of Target Coolant Model Assumptions_

Because the target cooling system operates at a relatively low pressure of around -

-

• a high pressure pipe break is not credible. A mechanical deformation due to some heavy object striking the cooling water pipe is assumed to be credible and that it would take at least 0.5 seconds for the pipe to be dislocated enough to form an offset break. Two worst-case hypothetical break locations out of the reactor pool are analyzed: before the wye which affects both target assemblies, and at the joint after the last flex pipe before the cooling line enters the reactor pool. One break location within the reactor pool is analyzed at the joint connecting the flex pipe to the target assembly inlet pipe. These break locations are explained in more detail in Section 6.

Only one target DHRS valve is assumed to be open in each target cooling line. Analyses are also performed where none of the target DHRS valves operate to prove that they are not needed. RELAP5 only models gap closure due to thermal expansion so it overpredicts peak pellet temperatures. Therefore, RELAP5 is only used to predict changes in peak pellet temperature and FRAPCON is used to predict steady state peak pellet temperatures.

4.6 Loss of Target Flow Model Assumptions The assumed pump characteristics, particularly its inertia and torque requirements, determine the coastdown of the pump. No flywheel or other special means are assumed which would extend the coastdown of the pump.

Only one valve is assumed to be open in each target cooling line. Analysis was also performed where none of the target DHRS valves operate to prove that they are not needed. RELAP5 only models gap closure due to thermal expansion so it overpredicts peak pellet temperatures.

Therefore, RELAP5 is only used to predict changes in peak pellet temperature and FRAPCON is used to predict steady state peak pellet temperatures.

5 INPUTS 5.1 Target Pellet, Cladding and Cartridge Data Pellet dimensions are explicitly provided as input for both FRAPCON and FRAPTRAN and are used to ~determine pellet volume and initial gap between pellet and cladding for RELAPS. Pellet and cladding dimensions and tolerances obtained from two drawings are used to provide necessary inputs and establish minimurn and maximum gaps between the pellet and cladding

). Pellet dimensions are depicted in Figure 2 and cladding 10

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B dimensions are depicted in Figure 3. The minimum gap of occurs when the maximum pellet outer diameter (OD) is matched with the minimum cladding inner diameter (ID).

Similarly, the maximum gap of occurs when the minimum pellet OD is matched with the maximum cladding ID. The nominal gap of occurs with nominal pellet and cladding dimensions. Table 1 gives the pellet and cladding dimensions that are used in the minimum, nominal and maximum gap cases. The minimum cladding thickness was used in the minimum gap case to maximize cladding strain. The maximum cladding thickness was used in the maximum gap case to maximize pellet temperature. The nominal gap dimensions are used to model the average of the remaining rods in both the minimum and maximum gap cases that are not explicitly modeled.

Figure 2. Target pellet dimensions Figure 3. Cladding dimensions 11

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 1 Target Pellet and Cladding Dimensions Parameter Pellet outer diameter Cladding inner diameter Cladding thickness Cladding outer diameter Additional pellet dimensions listed in Table 2, and shown in Figure 2, include those needed to define the pellet dishing and chamfer.

Table 2 Additional Target Pellet Dimensions Parameter 1--~~~~~~~~~~-+-~~

Dimension (SI)

~~--~~

Dimension (US)

~~~

Pellet dish depth Pellet end-dish shoulder width Chamfer height Chamfer width Pellet height and the cladding has a surface roughness of The pellet density is - of theoretical with an enrichment of - weight percent (30441000211 ).

The cladding will have a cold work o f - according to the Zircaloy-4 tubing supplier.

Each target rod nominally contains

• pellets with a total length of . The upper plenum above the pellet stack contains a spring as specified in the assembly and spring drawings . Table 3 presents the relevant spring and plenum data used for input.

Each target rod cartridge contains

• target rods with a pitch of

. The nominal diameter of the coolant holes in the cartridge is

• but an upper limit of was used which conservatively increases the flow area and reduces the flow velocity within the target cartridge.

12

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B Table 3 Target Spring and Plenum Data Parameter Value Plenum length cold Plenum spring outer diameter Plenum spring wire diameter Plenum spring number of turns Rod fill gas pressure Rod fill gas helium For all three gap sizes and associated cladding outer diameter (OD), the flow area, equivalent hydraulic diameter, and equivalent heated diameter were calculated using MATHCAD. The calculations depend on the location of the target rods because the corner rods differ from the interior rods within the target cartridge with respect to power profile. The results of these calculations are presented in Table 4 and are used primarily in the RELAPS input but also in the FRAPCON and FRAPTRAN input for target rod

• In target assembly 1, the maximum power rod and maximum power density rod are both interior rods so that the remaining rods are

• corner rods and - interior rods. In target assembly 2, the maximum power rod is a corner rod and the maximum power density rod is an interior rod so that the remaining rods are

• corner rod and . . interior rods.

Table 4 Flow Areas and Diameters for Target Rods Parameter

• Flow Area Hydraulic Diameter Heated Diameter Interior Rod Flow Area Hydraulic Diameter Heated Diameter 13

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 4 Flow Areas and Diameters for Target Rods Parameter Interior Rods and Flow Area Only Only Nominal Nominal Hydraulic Diameter Used Used Heated Diameter Interior Rods and Flow Area Only Only Nominal Nominal Hydraulic Diameter Used Used Heated Diameter 5.2 Target Steady State Power Distribution Steady state power distributions are based on detailed MCNPS analyses (30441 R00031 ).

MCNPS produced results for each target rod in both target assemblies using 25 axial nodes.

The input for RELAPS uses 20 axial nodes to represent the stack of

• target pellets. The MCNPS results were averaged in a piece-wise fashion to generate 20 data points. The RELAPS model explicitly analyzes rods 5 and 6 in target assembly 1 and rods 17 and 22 in target assembly 2. The remaining rods in each target assembly are averaged together in the RELAP5 model. The rods with the peak power density - are obtained from the Extreme MCNP6 case 30 (30441 R00031 ). The peak power density case has more control blade insertion which makes the power density more peaked below the core mid-plane. The other rods modeled in RELAP5 are based on the Maximum MCNP6 case 44 with less control blade insertion which maximizes both rod and assembly power. Rod

• has the peak power density and is analyzed in RELAPS, FRAPCON, and FRAPTRAN. The FRAPCON and FRAPTRAN input needs 21 data points at the mesh lines of the 20 axial nodes. Figure 4 through Figure 9 presents the axial power factors generated by MCNP6 along with the factors used in RELAPS. Figure 7 also includes the axial power factors used in FRAPCON and FRAPTRAN for target rod *

• 14

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

+ MCNP

• RELAP 1.4 1.3

• *~

1.2

....! 1.1 l;J

~ ti .

CO.I

~ 0.9 Q

Q..

'i 0.8 **

• ~

<: 0.7

• 0.6 0.5 ***

0.4

• Height (cm)

Figure 4. Axial power factor for target rod I 15

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

• MCNP

• RELAP 1.4 1.3 1.2 I.I **

• • '~ ~

'tj

~ 1

~ 0.9 Q

~ ***

• Q..

'; 0.8

'>(

< 0.7 0.6 **

0.5 **

0.4 Height (cm)

Figure 5. Axial power factors for target rod I

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report 1.4 1.3

't 1.2 1.1

.... , ti . . .. . ti

~

... l

• ~- *

~ 0.9 Q

Q..

-; 0.8

• ~

< 0.7 0.6 0.5 0.4 Height (cm)

Figure 6. Axial power factors for target rods I thru I and I thru

• 17

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

+ MCNP RELAP + FRAPCON 1.4 1.3 1.2

....!!s c:.I 1.1

++ +

e: 1 t-.

... +

~ 0.9 0

Q.

i 0.8

< 0.7 +

0.6 0.5 0.4

• +

Height (cm)

Figure 7. Axial power factors for target rod

• 18

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

+MCNP

• RELAP 1.4 1.3 1.2

, ..... ' ti ..... ' .. ..

't

=

1.1

""... 0.9

~

~ ... ... * ...

0 c...

-; 0.8

'>(

< 0.7 0.6 0.5 0.4

...................... Height (cm)

Figure 8. Axial power factors for target rod*

19

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

+ MCNP

• RELAP 1.4 1.3 1.2

~ 1.1 , ..... ... . . . . *

~

~

a.

Q

~ 0.9 Q,.

~*

~ 0.8

I(

< 0.7 0.6 0.5 0.4 Height (cm)

Figure 9. Axial power factors for target rods* thru

• and* thru

• The total target rod powers are presented in Table 5 and conservatively include 2-sigma uncertainties from statistical error, density, enrichment, and target rod position (30441 R00031 ).

I I Target rods and* have uncertainties of 4.26%. Target rods and* have uncertainties of 3.52%. All of the remaining rods represent have an uncertainty of 2.27% based on the total target power. The target rod power for the average rods * ** *-

• and - are based on the nominal gap pellet volume.

Table 5 Target Rod Power Target Rods t----..-~~~~~~~-t-~~~

Power per rod

~~~--1 (Average) 20

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B FRAPCON and FRAPTRAN *power input for target rod

• uses the RELAPS value divided by the pellet stack height of - and converted to units of kW/m. This value is then modified by an axial power factor to get the linear power at a given height.

5.3 RELAP5 Hydrodynamic _Components The steady state RELAPS model of the target cooling system (Figure 1) does not include the target DHRS valves. The steady state model also does not include the target cooling system pump or heat exchanger. The steady state RELAPS model uses time dependent volumes and time dependent junctions to replace the pump and heat exchanger.

The RELAPS models for the loss of coolant accident (LOCA) and loss of pump flow (LOPF) accident are essentially the same except for the trip logic and the addition of break valves to the LOCA model. The volume data for the hydrodynamic components in the RELAPS model are presented in Table 6. Volume 80 has a flow area expander so 3 different areas are used.

Volume 102 has a flow area reducer in that pipe so 3 different areas are used. Other volumes in the target assembly also have area changes associated with their flow paths.

Table 6 RELAP5 Volumes Volume Area Length Angle Roughness Description m2) (0)

ID (m) (m)

Pump 075-01 4.769-3 0.1382 90.0 Pump outlet to HX inlet 080-01 4.769-3 0.986 90.0 1.9-6 080-02 4.769-3 0.410 0.0 1.9-6 080-03 4.769-3 0.225 0.0 1.9-6 080-04 6.376-3 0.990 0.0 1.6-6 080-05 8.213-3 1.340 0.0 1.3-6 Heat exchanger (HX) 090-01 -31.79 090-02 -31.79 090-03 -31.79 090-04 -31.79 090-05 -31.79 090-06 -31.79 090-07 -31.79 090-08 -31.79 090-09 -31.79 090-10 -31.79 HX outlet to wye 100-01 8.213-3 0.733 0.0 1.3-6 100-02 8.213-3 0.612 0.0 1.3-6 100-03 8.213-3 1.485 0.0 1.3-6 100-04 8.213-3 1.181 0.0 1.3-6 100-05 8.213-3 0.225 0.0 1.3-6 100-06 8.213-3 0.842 10.95 1.3-6 21

ATTACHMENT 7 REI.AP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 6 RELAP5 Volumes Volume Area Length Angle Roughness Description (0)

ID (m2) (m) (m) 100-07 8.213-3 0.150 0.0 1.3-6 Wye to isolation valve 102-01 8.213-3 0.308 0.0 1.3-6 .

102-02 8.213-3 0.138 0.0 1.3-6 102-03 6.376-3 0.244 0.0 1.6-6 102-04 4.769-3 0.549 0.0 1.9-6 102-05 4.769-3 0.479 0.0 1.9-6 102-06 4.769-3 0.673 0.0 1.9-6 Isolation valve to natural 104-01 4.769-3 0.418 0.0 1.9-6 circulation valve tee 104-02 4.769-3 0.305 0.0 1.9-6 104-03 4.769-3 0.230 0.0 1.9-6 104-04 4.769-3 0.230 0.0 1.9-6 104-05* 4.769-3 0.305 0.0 1.9-6 104-06 4.769-3 1.380 0.0 1.9-6 104-07 4.769-3 1.108 0.0 1.9-6 104-08 4.769-3 1.032 0.0 1.9-6 104-09 4.769-3 0.230 -90.0 1.9-6 104-10 4.769-3 0.573 -90.0 1.9-6 104-11 4.769-3 0.573 -90.0 1.9-6 104-12 4.769-3 0.077 -90.0 1.9-6 From tee to nat. circ. valve 106-01 2.027-3 0.170 0.0 1.9-6 From tee to tee 108-01 4.769-3 0.077 -90.0 1.9-6 108-02 4.769-3 0.077 -90.0 1.9-6 From tee to nat. circ. valve 110-01 2.027-3 0.170 0.0 1.9-6 Natural circulation valve tee to 112-01 4.769-3 1.000 -90.0 1.9-6 TA 112-02 4.769-3 1.000 -90.0 1.9-6 112-03 4.769-3 0.872 -90.0 1.9-6 112-04 4.769-3 0.406 0.0 1.9-6 112-05 4.769-3 0.508 -90.0 1.9-6 112-06 4.769-3 0.406 0.0 1.9-6 112-07 4.769-3 0.770 -90.0 1.9-6 112-08 4.769-3 0.865 -90.0 1.9-6 112-09 4.769-3 0.305 -90.0 1.9-6 112-10 4.769-3 0.443 -90.0 1.9-6 TA top to lower plenum 114-01 -90.0 114-02 -90.0 114-03 -90.0 Lower plenum 115-01 0.0 From lower plenum to target 116-01 90.0 rod holder 116-02 90.0 Target rod holder 117-01 0.0 Target r o d s - 121-01 90.0 121-02 90.0 90.0 90.0

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B Table 6 RELAP5 Volumes Volume Area Length Angle Description ID (m2) (m) (0) 121-05 90.0 121-06 90.0 121-07 90.0 121-08 90.0 121-09 90.0 121-10 90.0 121-11 90.0 121-12 90.0 121-13 90.0 121-14 90.0 121-15 90.0 121-16 90.0 121-17 90.0 121-18 90.0 121-19 90.0 121-20 90.0 121-21 90.0 Target rod I 122-01 90.0 122-02 90.0 122-03 90.0 122-04 90.0 122-05 90.0 122-06 90.0 122-07 90.0 122-08 90.0 122-09 90.0 122-10 90.0 122-11 90.0 122-12 90.0 122-13 90.0 122-14 90.0 122-15 90.0 122-16 90.0 122-17 90.0 122-18 90.0 122-19 90.0 122-20 90.0 122-21 90.0 Target rod I 123-01 90.0 123-02 90.0 123-03 90.0 123-04 90.0 123-05 90.0 90.0 23

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 6 RELAP5 Volumes Volume Angle Roughness Description (0)

ID (m) 123-07 90.0 123-08 90.0 123-09 90.0 123-10 90.0 123-11 90.0 123-12 90.0 123-13 90.0 123-14 90.0 123-15 90.0 123-16 90.0 123-17 90.0 123-18 90.0 123-19 0.0 123-20 90.0 123-21 90.0 Upper target rod positioner 125-01 0.0 Target hold-down and diffuser 135-01 90.0 135-02 90.0 135-03 90.0 135-04 90.0 135-05 0.0 Pool inlet to pump inlet 302-01 4.769-3 0.689 90.0 1.9-6 302-02 4.769-3 0.689 90.0 1.9-6 302-03 4.769-3 0.689 90.0 1.9-6 302-04 4.769-3 0.689 90.0 1.9-6 302-05 4.769-3 0.230 90.0 1.9-6 302-06 4.769-3 1.019 0.0 1.9-6 302-07 4.769-3 1.066 0.0 1.9-6 302-08 4.769-3 0.920 0.0 1.9-6 302-09 4.769-3 0.936 0.0 1.9-6 302-10 4.769-3 1.437 0.0 1.9-6 302-11 4.769-3 0.305 0.0 1.9-6 302-12 4.769-3 0.230 0.0 1.9-6 302-13 4.769-3 0.230 0.0 1.9-6 302-14 4.769-3 0.305 0.0 1.9-6 302-15 4.769-3 0.196 0.0 1.9-6 302-16 4.769-3 1.167 0.0 1.9-6 302-17 4.769-3 1.167 0.0 1.9-6 302-18 4.769-3 0.413 0.0 1.9-6 302-19 4.769-3 0.354 0.0 1.9-6 302-20 4.769-3 0.309 10.06 1.9-6 302-21 4.769-3 1.079 0.0 1.9-6 302-22 4.769-3 0.664 0.0 1.9-6 302-23 4.769-3 0.407 0.0 1.9-6 24

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B Table 6 RELAP5 Volumes Volume Area Length Angle Roughness Description ID (m2) (m) (0) (m) 302-24 4.769-3 0.483 0.0 1.9-6 Reactor pool 500-01 7.0686 10.00 -90.0 1.0-3 500-02 7.0686 0.893 -90.0 1.0-3 500-03 7.0686 0.154 -90.0 1.0-3 500-04 7.0686 0.154 -90.0 1.0-3 500-05 7.0686 0.154 -90.0 1.0-3 500-06 7.0686 1.225 -90.0 1.0-3 500-07 7.0686 3.941 -90.0 1.0-3 500-08 7.0686 0.340 -90.0 1.0-3 500-09 7.0686 0.103 -90.0 1.0-3 Secondary side HX inlet 400-01 1.0 1.0 0.0 0.0 Secondary side HX 406-01 31.79 406-02 31.79 406-03 31.79 406-04 31.79 406-05 31.79 406-06 31.79 406-07 31.79 406-08 31.79 406-09 31.79 406-10 31.79 Secondary side HX outlet 408-01 1.0 1.0 0.0 0.0 Hydrodynamic components 102 through 136 are duplicated for components 202 through 236 to model flow through the second target assembly. The only difference is in the flow areas in the target assembly volumes 221, 222, and 223 (per Table 4). The junction data for hydrodynamic components with non-zero K-loss factors are presented in Table 7.

Table 7 RELAP5 Junctions with Non-Zero K-Loss Factors Junction Area Kloss Kloss Description m2 ID forward reverse Pump inlet 75-01 4.769-3 0.04 0.04 Pump outlet 75-02 4.769-3 0.04 0.04 Pump outlet to HX inlet 80-01 4.769-3 0.111 0.111 80-02 4.769-3 0.200 0.200 80-03 4.769-3 0.187 0.187 HX inlet 85-00 8.213-3 0.500 0.500 Heat exchanger (HX) 90-01 90-02 25

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 7 RELAP5 Junctions with Non-Zero K-Loss Factors Junction Area K Loss K Loss Description ID (m2) forward reverse 90-04 90-05 90-06 90-07 90-08 90-09 HXoutlet 95-00 8.213-3 0.500 0.500 HX outlet to wye 100-01 8.213-3 0.097 0.097 100-02 8.213-3 1.000 1.000 100-03 8.213-3 0.050 0.050 100-04 8.213-3 0.097 0.097 100-05 8.213-3 0.650 0.650 100-06 8.213-3 0.650 0.650 WyetoTAl 101-01 8.213-3 0.100 0.100 WyetoTA2 101-02 8.213-3 0.100 0.100 Wye to Isolation valve 102-01 8.213-3 0.050 0.050 102-02 8.213-3 0.187 0.187 Isolation valve 103-00 4.769-3 0.400 0.400 Isolation valve to natural 104-01 4.769-3 0.750 0.750 circulation valve tee 104-02 4.769-3 0.750 0.750 104-03 4.769-3 0.116 0.116 104-04 4.769-3 0.750 0.750 104-05 4.769-3 0.750 0.750 104-06 4.769-3 0.116 0.116 104-07 4.769-3 0.020 0.020 104-08 4.769-3 0.116 0.116 From tee to nat. circ. valve 105-0101 4.769-3 0.200 0.200 105-0102 2.027-3 0.800 0.800 Natural circulation valve 107-00 2.027-3 1.600 1.100 From tee to nat. circ. valve 109-0101 4.769-3 0.200 0.200 109-0102 2.027-3 0.800 0.800 From natural circulation valve tee 112-03 4.769-3 0.116 0.116 to TA 112-04 4.769-3 0.116 0.116 112-05 4.769-3 0.116 0.116 112-06 4.769-3 0.116 0.116 112-08 4.769-3 0.750 0.750 112-09 4.769-3 0.750 0.750 Target assembly (TA) inlet 113-00 TA top to lower plenum 114-01 114-02 Lower plenum 115-01 115-02 After lower plenum 116-01 26

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B Table 7 RELAP5 Junctions with Non-Zero K-Loss Factors Junction Area K Loss K Loss Description ID (m2) forward reverse Target rod holder inlet 117-01 Inlet to target rods - 117-02 Inlet to target rod 6 117-03 Inlet to target rod 5 117-04 Exit from target rods - 125-01 Exit from target rod 6 125-02 Exit from target rod 5 125-03 Upper target rod positioner 125-04 Target hold-down and diffuser 135-01 135-02 135-03 135-04 Diffuser exit to pool 136-00 Pool inlet to cooling water 301-00 4.769-3 0.500 0.500 Pool inlet to pump inlet 302-05 4.769-3 0.116 0.116 302-06 4.769-3 0.116 0.116 302-08 4.769-3 0.040 0.040 302-09 8.213-3 0.116 0.116 302-10 4.769-3 0.560 0.560 302-11 4.769-3 0.560 0.560 302-12 4.769-3 0.116 0.116 302-13 4.769-3 0.560 0.560 302-14 4.769-3 0.560 0.560 302-17 4.769-3 0.560 0.560 302-18 4.769-3 0.560 0.560 302-20 4.769-3 0.116 0.116 302-21 4.769-3 1.000 1.000 302-22 4.769-3 0.116 0.116 302-23 4.769-3 0.050 0.050 Pool water I air surface 500-01 7.0686 5.00 5.00 Secondary side HX 406-01.

406-02 406-03 406-04 406-05 406-06 406-07 406-08 406-09 27

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B 5.4 RELAP5 Pump Model RELAPS includes pump curves for two different models of pump. Because the exact model and associated performance data for the target cooling system pump remains to be specified by the supplier of the cooling system skid, the Westinghouse pump was selected with input data appropriate for this application. The pump head and velocity are adjusted in the RELAPS model to match the flow rate through the target assembly. K-loss factors within the target assembly are also adjusted to match the pressures predicted by FLUENT (30441 R00038). The pump parameters in Table 8 produce the desired characteristics but would be updated as the pump design is finalized and cooling system pressure drops are verified.

Table 8 RELAP5 Pump Parameters Parameter Value Rated pump speed Pump speed ratio Rated pump flow Rated pump head Rated pump torque Pump moment of inertia 5.5 RELAP5 Heat Exchanger Model A plate heat exchanger will likely be selected by the subcontractor supplying the target cooling system skid and its performance will meet the specified requirements. The RELAPS model can approximate a plate heat exchanger and the parameters in Table 9 closely match the expected inlet and outlet conditions at the design flow rate.

28

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B Table 9 RELAP5 Heat Exchanger Parameters Parameter Value Primary inlet temperature Primary outlet temperature Secondary inlet temperature Secondary outlet temperature Primary inlet pressure Primary outlet pressure Secondary inlet pressure Secondary outlet pressure Primary flow rate Secondary flow rate Flow area Heat transfer area Height 1.70m Flow gap 2mm Flow length 2.54m Plate thickness lmm Surface roughness K-loss per turn 5.6 Steady State Inputs from RELAP Steady state RELAP5 runs were performed for the - gap and - gap cases at nominal (100%) flow of - through the target assembly at 100% and 115% reactor and target power. Steady state RELAP5 runs were also performed at 85% flow for the - gap and

- gap cases at 100% power and 115% power. RELAP5 cladding surface temperatures at the node centers for target rod

• were averaged or extrapolated to produce the mesh line temperatures required by FRAPCON. Table 10 through Table 12 present the RELAP5 results and FRAPCON input values in Kelvin. RELAP5 coolant pressures along target rod *are presented in Table 13. The mass flux in the flow channel around target rod* is presented in Table 14.

29

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 10 Cladding Surface Temperature at 100% Power and 100% Flow in Kelvin Table 11 Cladding Surface Temperature at 115% Power and 100% Flow in Kelvin RELAP Result 30

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 12 Cladding Surface Temperature at 115% Power and 85% Flow in Kelvin gap RELAP Result FRA I pu 31

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Table 13 Target Rod

• Coolant Pressure in Pascals Table 14 Target Rod. Mass Flux 100% Power 100% Flow 5.7 Reactivity Transient Power History The MURR reactivity transient for a 600 per cent mille (pcm) rapid reactivity insertion is described in the MURR SAR for relicensing and summarized below in Section 6.1. MURR provided General Atomics (GA) with an EXCEL file containing the power prediction for this transient. It was manipulated within the EXCEL file to provide the data in the format needed by FRAPTRAN.

The MURR reactivity transient for a 30 pcm per second control blade withdrawal is also described in the MURR SAR for relicensing and summarized below in Section 6.2. The slow reactivity addition coupled with negative temperature feedback produces a gradual power ramp 32

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441R00032/B which was added to the MURR EXCEL spreadsheet and manipulated to provide the data in the format needed by FRAPTRAN.

6 RESULTS AND CONCLUSIONS 6.1 Rapid Insertion of Positive Reactivity For the Insertion of Excess Reactivity accident analysis, the licensed maximum power level of 10 MW was used in the reactor SAR as the starting assumption since MURR does not, nor can it legally, operate above this power level. NUREG-1537, Part 2, page 13-9, Standard Review Plan and Acceptance Criteria, states for Insertion of Excess Reactivity accident, that "The accident scenario assumes that the reactor has a maximum load of fuel (consistent with the technical specifications), the reactor is operating at full licensed power, and the control system ..." The accident was re-analyzed at a much more conservative starting power level (11 .5 MW) than is required by NUREG-1537 and the results are provided below. 11 .5 MW was chosen, instead of the Limiting Safety System Setting (LSSS) set point of 12.5 MW, since the rod run-in system will initiate a rod run-in at 11.5 MW (Technical Specification 3.2.f.1) and shutdown the reactor prior to reaching the LSSS SCRAM set point of 125%.

For the reactor SAR analysis of the Insertion of Excess Reactivity accident, the temperature coefficient used was -6.0 x 10*5 .6k/k and not -7.0 x 10*5 .6k/k as stated in the previous version of the reactor SAR. The third paragraph on Page 13-17 of the SAR lists the various reactivity coefficients assumed for the Insertion of Excess Reactivity accident analysis.

For both the reactor SAR analyses, as well as for the updated analysis presented here, the control blade insertion times are based on the current and relicensing Technical Specification 3.2.c requirement of insertion to the 20% withdrawn position in less than 0.7 seconds. So the insertion rate was calculated based on shim control blades travelling from 26 inches (fully withdrawn) to 5.2 inches (20% withdrawn or 80% inserted) in 0.7 seconds. This is a conservative assumption since monthly control blade drop time verifications performed at MURR have always yielded insertion times of 0.6 seconds or less.

Similar to the reactor SAR analysis, the Reactivity Transient Analysis program PARET 0/7.5),

maintained and distributed by the Nuclear Engineering Division of Argonne National Laboratory (ANL) was used. For the Insertion of Excess Reactivity accident analysis, two channels were modeled in PARET; a hot channel representing worst-case conditions inside the core and an average channel representing the rest of the core experiencing "average" conditions. As indicated earlier, the transient was started from an initial power level of 11 .5 MW with core coolant flow rate as well as core coolant inlet temperatures set at their LSSS values of 3,200 gpm and 155°F, respectively. Also, pressurizer pressure was at 75 psia (LSSS value).

Since the Insertion of Excess Reactivity transient was analyzed from a starting power level of 33

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report 11 .5 MW, the rod run-in that would be initiated by the rod run-in system at 11 .5 MW was bypassed and only the high power SCRAM set point of 12.5 MW was modeled. Also, a delay of 150 milliseconds was incorporated into the control blade SCRAM model so that the control blades would only start to insert 0.15 seconds after the power level had exceeded the SCRAM set point of 12.5 MW.

The results of a step reactivity insertion of 600 pcm (+0 .006 6k/k) are shown below in Figure 10.

As expected , due to the higher starting core power level, much lower core coolant flow rate and much higher than normal core coolant inlet temperature conditions assumed for this updated analysis, the peak power during the transient momentarily reaches approximately 37.4 MW compared to a value of approximately 33.0 MW reported in prior SAR analysis for the same 600 pcm step reactivity insertion.

t

- POWER MW

] '0 350

~ 30.00 l

- Tclad'C Tf max"C l 300 u

250 ~

i 25.00 Q,I Q

Q.

20.00

...=

200 ~

Q,I su c.

15.00 150 E t'I Q,I ~

~ 10.00 100 5.00 so 0.00 0 0 .00 0.50 1.00 1.50 2.00 2.50 3.00 Time (seconds)

Figure 10. Reactor power, fuel and cladding temperatures vs. time for a positive reactivity step insertion of 600 pcm The power generation in the target assemblies would follow the same proportional power transient that the reactor core experiences because the target assemblies are driven by the neutron flux generated by the reactor core.

34

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report The target rod analysis for the positive reactivity step insertion examines the maximum powered target rod at the beginning and end of a - irradiation. A steady state analysis of the

- irradiation was performed using FRAPCON to establish the initial conditions for the transient analysis. The FRAPCON analysis assumed that the first and last day of the . .

. . irradiation are at 115% power and the rest of the operating period is at 100% power. The FRAPCON analysis also assumes that the target assembly flow rate is 85% of nominal during the first and last day of operation. Mid-week and weekend shutdowns are also included in the power history for the target assembly. The transient analysis was performed using the FRAPTRAN code. FRAPCON and FRAPTRAN are two NRC-sponsored computer codes that can model the steady state and transient thermal-mechanical behavior of LWR oxide fuel.

Phenomena modeled by the codes include heat transfer through fuel and cladding to coolant, cladding elastic/plastic deformation, fuel-cladding mechanical interaction, fission gas release, rod internal pressure, and cladding oxidation.

The FRAPCON steady state analysis defines the bumup, U02 and cladding deformation, and fission gas release that form the starting point for the FRAPTRAN analysis. The FRAPTRAN analysis increases the power to 115% and decreases the flow rate to 85% within 40 seconds.

Twenty-one seconds later, the 600 pcm (+0.006 6k/k) step reactivity insertion is simulated. The target cooling water inlet temperature to the target assembly was assumed to be at 102°F (38.9°C) which is higher than the maximum allowable temperature of 90°F (32.2°C) exiting the heat exchanger. The maximum cladding strain occurs with the minimum gap tolerance of

. The maximum cladding strain also occurs at the end of the -

irradiation due to the effects of bumup on gap closure and fission gas release. The FRAPCON and FRAPTRAN analysis also use the maximum tolerance on the scallop diameter that forms the flow area of the target cartridge. This tolerance results in the minimum velocity consistent with the minimum gap. The analysis also uses the maximum pellet outer diameter, minimum cladding inner diameter, minimum cladding thickness, and minimum cladding outer diameter.

The strain transient that the cladding is predicted to undergo during the positive step reactivity insertion is shown in Figure 11. The fractional hoop strain increases from about 0.50% to just under 0.76% but remains under the 1% strain limit.

35

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report 0.008 0.006 0.004 c

.c

~

0.002

';; - Hoop Strain g

i 0 - Axial Strain E

t;i;.

- Radial Strain

-0 .002

-0 .004

-0.006 0 0.5 1.5 2 2.5 Time (seconds)

Figure 11. Cladding strains at peak pellet location during 600 pcm reactivity insertion Maximum target pellet temperatures occur with the maximum gap tolerance of

- rather than the minimum gap tolerance. Burnup effects reduce the gap so that the maximum pellet temperature occurs at the beginning of irradiation. The analysis also uses the minimum pellet outer diameter, maximum cladding inner diameter, maximum cladding thickness, and maximum cladding outer diameter in order to obtain the maximum gap tolerance or to maximize peak pellet temperature. The peak target temperature transient is shown in Figure 12. The peak target temperature increases from to slightly less than 2729°C (4944°F) in less than 0.4 seconds. This peak U02 temperature is 111°C (200°F) below the melting temperature of 2840°C (5144°F).

36

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B 0 0.5 1.5 2 2.5 Time (seconds)

Figure 12. Peak target pellet temperature during 600 pcm reactivity insertion The pellet outer diameter and cladding inner and outer diameter temperature at the axial location where the peak pellet centerline temperature occurs is shown in Figure 13. The cladding outer diameter temperature shows a modest increase in temperature of around 20°C (36°F). The pellet outer diameter and cladding inner diameter temperatures become more closely coupled due to closure of the gap between the pellet and cladding and an increase in interface pressure. Cladding hoop strain for the gap at the end of irradiation goes through a similar transient as for the just under 0.1 8%.

Peak heat generation within the target pellets is a factor of 3.74 greater than nominal. Heat capacity and thermal resistances result in the surface heat flux at the peak location being only a factor of 1.34 to 1.50 greater than nominal, depending on the initial gap size and burnup conditions. FRAPTRAN assesses critical heat flux (CHF) using one of five different CHF correlations. The MacBeth CH F correlation was chosen because of its validity at low pressures.

The MacBeth CHF ratio reaches a minimum value of between 2.1 6 to 2.24 depending on initial gap size and burnup conditions. The Bernath CHF ratio reaches a minimum of 1.67 to 1.87 depending on the case. No target damage or radiological release would occur from this accident.

37

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

~--

- Pellet OD

- Clad ID

- Clad OD 0 0.5 1 1.5 2 2.5 Time (seconds)

Figure 13. Pellet OD and cladding temperatures at peak pellet location during 600 pcm reactivity insertion 6.2 Control Blade Withdrawal As presented in section 13.2.2.1.2 of the MURR SAR, a positive reactivity ramp insertion rate of 30 pcm/s, which is the Technical Specification limit on the maximum rate of reactivity insertion for all four shim control blades operating simultaneously, was introduced to the reactor starting at subcritical cold conditions and at an initial power level of 10 MW. For subcritical cold conditions, the short period reactor SCRAM terminates the transient within 150 sec before the power has reached 64 watts. No target pellet or cladding damage would occur at such a low power.

For full power conditions, the high power SCRAM terminates the transient after 4.53 seconds when reactor power rises from 10.0 MW to the 12.5 MW high power SCRAM. The thermal-mechanical performance of the target rod during this transient was analyzed using FRAPCON and FRAPTRAN. The power transient that drives the heat generation in the target rods is 38

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B shown in Figure 14. The figure includes 1 sec of steady-state operation at 100% power and the power transient after reactor SCRAM .

I I 1 *- ...___

~

'5 >-

c

.2

~

l!

\

u..

0.1 I

0 1 2 3 4 s 6 7 Time (seconds)

Figure 14. Power transient during 30 pcm per second reactivity insertion The target rod analysis for the control blade withdrawal examines the maximum powered target rod at the beginning and end of a - irradiation. A steady state analysis of the -

. . irradiation was performed at 100% power using FRAPCON to establish the initial conditions for the transient analysis. The FRAPCON analysis also assumes that the target assembly flow rate is 100% of nominal during the - irradiation. Mid-week and weekend shutdowns are also included in the power history for the target assembly.

The FRAPCON steady state analysis defines the burnup, U0 2 and cladding deformation, and fission gas release that form the starting point for the FRAPTRAN analysis. The maximum cladding strain occurs with the minimum gap tolerance of . The maximum cladding strain also occurs at the end of the - irradiation due to the effects of burnup on gap closure and fission gas release. The strain transient that the cladding is predicted to 39

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report undergo during the control blade withdrawal is shown in Figure 15. The fractional hoop strain increases from about 0.24% to just over 0.56% but remains well under the 1% strain limit.

0.007 0.006 0.005 0.004

= 0.003 l:

v:i

"'; 0.002 - Hoop Strain i: - Axial Strain u

~

0.001 a.. - Radial Strain tor..

0

-0.001

-0.002

-0.003 0 2 3 4 5 6 7 Time (seconds)

Figure 15. Cladding strains at peak pellet location during 30 pcm per second reactivity insertion For the case using the gap between the pellet and cladding , the pellet centerline temperature increases less than at the beginning of irradiation. Burnup effects reduce the gap so that the maximum pellet temperature decreases during irradiation . The temperature of the cladding inner diameter increases to less than -

-

• and the cladding outer diameter increases by less than -

The FRAPTRAN analysis of the gap case predicts the maximum pellet temperatures due to the higher thermal resistance between the pellet and cladding. The maximum pellet temperature increases from to just over at the beginning of irradiation as shown in Figure 16. The pellet outer diameter and cladding inner and outer diameter temperature at the axial location where the peak pellet centerline temperature occurs is shown in Figure 17. The pellet outer diameter and cladding inner diameter temperatures become more closely coupled due to closure of the gap between the pellet and cladding and an increase in interface pressure. The cladding outer diameter 40

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B temperature shows a modest increase in temperature of around - to just under

. Cladding fractional hoop strain goes through a similar transient as shown in Figure 15 but increases from 0.13% to 0.14%. No target damage or radiological release would occur from this accident.

0 1 2 3 4 5 6 7 Time (seconds)

Figure 16. Peak target pellet temperature during 30 pcm per second reactivity insertion 41

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

- Pellet OD

- Clad ID

- Clad OD 0 I 2 3 4 5 6 7 Time (seconds)

Figure 17. Pellet OD and cladding temperatures at peak pellet location during 30 pcm per second reactivity insertion 6.3 Loss of Target Coolant This accident assumes a double-ended break of one of the pipes in the target cooling system.

The design and operation of the target cooling system are described in the Target System Cooling Calculation Report and depicted in Figure 1. Depending on the location of the break, the loss of coolant will cause either an increase or a decrease in flow at the flow meters on either pipe leg supplying cooling water to the two target assemblies. The hi/low flow setpoints are at 115% and 85% and result in a reactor SCRAM and open the target DHRS valves located just above the refueling bridge. The hi/low flow signals that cause a reactor SCRAM will also automatically shutdown the target cooling system pumps. No operator action is required. Pump coastdown after pump trip can mitigate the early phase of the transient unless the offset pipe break prevents pump flow from reaching the target assembly.

The pipe breaks in the reactor pool differ from pipe breaks out of the reactor pool particularly if the target DHRS valves are assumed to not operate. A pipe break in the reactor pool establishes a natural circulation loop between the target assembly and reactor pool regardless of whether the target DHRS valves open. If the break is out of the reactor pool, a natural 42

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report circulation loop is only established if the target DHRS valves open. Otherwise, the break prevents pool water from reaching the target assembly inlet and establishing a natural circulation loop.

The pipe break before the wye affects both target assemblies but a pipe break downstream of the wye at the joint with the flexible pipe has a lower cold leg volume and greater impact on the target assembly. The relative locations of these pipe breaks are shown in Figure 18. The water in the cold leg of the supply line is almost - cooler than the pool water. The draining of this water through the target assemblies mitigates the transient during the early phase of the loss of coolant accident (LOCA).

Figure 18. Pipe break locations out of the reactor pool Pipe breaks out of the reactor pool are covered in Section 6.3.1 and pipe breaks in the reactor pool are covered in Section 6.3.2.

6.3.1 Pipe Break out of the Reactor Pool The pipe break out of the reactor pool just after the flexible pipe is a more credible pipe break than other locations closer to the target assembly because the joint is more vulnerable and the 43

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B remaining pipe is of welded construction. The upper bridge structure protects the pipes where they bend downward into the reactor pool. The break is analyzed to occur in the supply line to target assembly (TA) 2 because it has the target rod with the highest power density, and the target rod with the highest power in addition to being the target assembly with the most power.

The LOCA for the Selective Gas Extraction (SGE) target was modeled using RELAP5 mod3.3 patch 03. RELAP5 was developed by the NRC to analyze thermal-hydraulic transients in pressurized water reactors. It can be used to analyze a variety of geometries and was used to analyze the LOCA and Loss of Flow Accident (LOFA) in Sections 13.2.3 and 13.2.4 of the MURR SAR for relicensing. The RELAP5 model for this accident includes both target assemblies, target cooling system pump and heat exchanger, and target DHRS valves. The double-ended guillotine rupture is modeled with three valves - two of the valves connect to the reactor pool on either side of the break, and the third valve connects the two ends of the pipe across the break. Prior to accident initiation, the valves to the reactor pool are closed and the valve across the break is open. When the break is initiated, the valve connecting the two pipe ends is closed and the two valves from each end connected to the reactor pool is opened. All three valves are assumed to completely change position in 0.5 seconds which is a reasonable assumption for a mechanically induced failure that displaces the two ends of the pipe since the piping is not at high pressure.

The flow transient caused by a LOCA resulting from a double ended pipe break is shown in Figure 19. The transient analysis starts with the reactor and target at 100% power and the target cooling system at 100% nominal flow to accurately simulate the expected response of the protection system. The target DHRS valves are assumed to not operate. The break causes an increase in the flow measurement in the pipe supplying TA 2 and a decrease in the flow rate supplying TA 1.

The reactor is scrammed at 0.05 seconds when the mass flow rate to TA 2 is greater than 115 percent. A low flow signal occurs shortly thereafter at 0.08 seconds for TA 1 when its flow rate is less than 85 percent. The target cooling system pumps are automatically shutdown on the high flow alarm. Control blade movement is delayed by 0.15 seconds after the reactor SCRAM signal is generated. The mass flow entering TA 2 falls quickly because of the pipe break. The flow through TA 1 remains above 1.0 kg/s (2.2 Ibis) for almost 20 seconds. The flow through TA 2 drops below 1.0 kg/s (2.2 lb/s) at around 9 seconds and experiences several slow flow reversals between 10 and 40 seconds after which boiling and chugging within both target assemblies is predicted.

44

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

- Target 1 Flow Meter

- Target 2 Flow Meter Target 1 Inlet Target 2 Inlet

• * * .. *

• Target 1 Outlet

• ** * * *

• Target 2 Outlet 0.001 0.01 0.1 1 10 100 1000 Time (sec)

Figure 19. Mass flow transient during LOCA out of the reactor pool without target DHRS valves In TA 2, rod* has the maximum peak power density and rod* has the maximum rod power.

In TA 1, rod I has the maximum peak power density and rod I has the maximum rod power though both rods are lower in power than the rods in TA 2. The heat generation in the target rods is shown in Figure 20. The heat generation drops at around 0.20 seconds when the control blades start inserting . The heat removal from the target rods presented in Figure 20 shows that stored energy remova l is significant over the first 10 seconds. Mass flow oscillations due to chugging and boiling are also reflected in the oscillations in heat removal.

Coolant temperatures at the inlet and outlet of the two target assemblies are presented in Figure 21. Mass flow out of the target assemblies due to boiling are represented by outlet temperatures between 70 and 100°C (160 and 212°F). Mass flow back into the target assemblies are represented by the lower bounds of the temperature oscillations around S0°C (120°F). These flow oscillations also move water and thermal energy to the target assembly entrance which slowly raises that temperature.

45

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

~

.g Fission Power Fission Product Power I!

GI 0.1

• * * * * *

• Heat Remova l Rod 6 A.

~ - - - - Heat Removal Rod 5

• * ** * ** Heat Removal Rod 22

- - - - Heat Removal Rod 17 0.01 0.01 0.1 1 10 100 1000 Time (sec)

Figure 20. Target power during a LOCA out of the reactor pool without target DHRS valves 46

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

• * * * * *

• Target 1 Inlet

- - Target 1 Outlet

.. * .. *

• Targe t 2 Inlet

- - Target 2 Outlet 0 200 400 600 800 1000 1200 Time (sec)

Figure 21 . Coolant temperatures during a LOCA out of the reactor pool without target DHRS valves Pellet centerline temperatures do not increase during this LOCA because of the reactor SCRAM signal at 0.05 seconds, the U02 heat capacity, and the water flow through the target caused by draining of the cooling line. The maximum temperatures of the cladding inner diameter (ID) are shown in Figure 22. The cladding inner diameter temperature increases by 3.0°C (5.4 °F) prior to control blade insertion and afterwards, is lower than normal temperature. The increases in cladding inner diameter temperature at around 10 and 30 seconds coincide with the decreases in flow rate at those times. Boiling within the target assemblies keeps the maximum cladding temperature just above saturation temperature.

Because the pellet and cladding temperatures are near normal values, all cladding stresses and strains are also normal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOCA.

47

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

• * * * * *

• Target Rod 6

- - Target Rod 5

• * * * * *

• Target Rod 22

- - Target Rod 17 0.01 0.1 1 10 100 1000 Time (sec)

Figure 22. Maximum cladding ID temperatures during a LOCA out of the reactor pool without target DHRS valves If the one of the two target DHRS valves for each target assembly were to operate, the consequences of the LOCA out of the reactor pool are greatly improved. The flow transient due to a LOCA out of the reactor pool with the target DHRS valves working is shown in Figure 23.

The target inlet flow is the same as the target outlet flow and a small natural circulation flow is established after 20 seconds. TA 1 flow rate is higher around 10 seconds due to a small contribution of pump coastdown through the intact supply line to TA 1.

48

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

• * * * * *

• Target 1 Flow M eter

- - Target 2 Flow Meter

• * * * * *

• Target 1 Outlet

- - Target 2 Outlet 0 .001 0.01 0 .1 1 10 100 Time (sec)

Figure 23. Mass flow transient during a LOCA out of the reactor pool with target DHRS valves Pellet centerline temperatures also do not increase during this LOCA. The maximum temperatures of the cladding inner diameter (ID) are shown in Figure 24. The cladding ID temperature increases by 3.0°C (5.4°F) prior to control blade insertion and afterwards, is lower than normal operating temperature. Natural circulation is sufficient to prevent chugging and cladding temperatures steadily decrease after 20 seconds. Peak vapor fraction is less than 1%.

Because the pellet and cladding temperatures are near normal values, all cladding stresses and strains are also normal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOCA. The target DHRS valves can prevent boiling and chugging in the target assembly but are not necessary to assure cladding integrity.

49

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

• * * * * *

• Target Rod 6

- - Target Rod 5

• • * * * *

• Target Rod 22

- - Target Rod 17 0.01 0.1 1 10 100 Time (sec)

Figure 24. Maximum cladding ID temperatures during a LOCA out of the reactor pool with target DHRS valves 6.3.2 Pipe Break in the Reactor Pool The target cooling lines supplying water to the target assemblies are protected by the upper bridge structure above the pool and the refueling bridge located below the normal pool water level. The cooling lines also have lateral supports and are joined to the target assembly by a flexible joint. It is nearly incredible to postulate a mechanistic failure of these cooling pipes which results in a complete offset rupture within the assumed rupture time of 0.5 sec. If the break occurs up near the target DHRS valves then it is essentially the same as opening the target DHRS valves. The worst possible break location is the connection between the inlet pipe welded to the target housing and the flexible pipe as shown in Figure 25. Direct mechanical interaction in this area is very unlikely due to the congestion around this elevation.

50

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report Figure 25. Pipe break location in reactor pool The LOCA in the reactor pool used the same RELAPS model as the LOCAs out of the reactor pool except that the break location was relocated as shown in Figure 25. The break time to completely offset the rupture was also kept at 0.5 seconds although it is more likely that the break time would be either longer underwater or less than a complete offset rupture.

The flow transient caused by a LOCA in the reactor pool resulting from a double ended pipe break is shown in Figure 26. The transient analysis starts with the reactor and target at 100%

power and the target cooling system at 100% nominal flow to accurately simulate the expected response of the protection system. The target DHRS valves are assumed to not operate though their operation would be ineffective in this accident. The break causes an increase in the flow measurement in the pipe supplying TA 2 and a decrease in the flow rate supplying TA 1.

51

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

• * * * * *

• Target 1 Flow Meter

• * * * * *

• Target 2 Flow Meter

- - Target 1 Inlet

- - Target 2 Inlet

- - -

• Target 1 Outlet

- - -

• Target 2 Outlet 0.001 0.01 0 .1 1 10 100 Time (sec)

Figure 26. Mass flow transient during a LOCA in the reactor pool The reactor is scrammed at 0.09 seconds when the mass flow rate to TA 2 is greater than 115 percent. A low flow signal occurs shortly thereafter at 0.13 seconds for TA 1 when its flow rate is less than 85 percent. The target cooling system pumps are automatically shutdown on the high flow alarm. Control blade movement is delayed by 0.15 seconds after the reactor SCRAM signal is generated. The mass flow entering TA 2 falls quickly because of the pipe break. The flow through TA 2 drops to zero at around 0.6 seconds and experiences several boiling and chugging oscillation for the next few seconds. The flow through TA 1 remains above zero for about 3 seconds before it also experiences some boiling and chugging oscillations.

Natural circulation through TA 2 after 3 seconds is sufficient to prevent further boiling and chugging.

The heat generation in the target rods is shown in Figure 27. The heat generation drops at around 0.24 seconds when the control blades start inserting. Brief periods of transition and film boiling occur at the high heat flux locations of the target rods. The heat removal from the target rods presented in Figure 27 shows that stored energy removal is significant over the first 10 seconds. Mass flow oscillations due to chugging and boiling are also reflected in the oscillations in heat removal and occur while significant stored energy remains to be removed.

52

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

- Fission Power

- Fission Produ ct Power

• * * * ** Heat Removal Rod 17

- - Heat Re moval Rod 22 0 .01 0.1 1 10 100 Time (sec)

Figure 27. Target power during a LOCA in the reactor pool The maximum temperatures of the cladding inner diameter (ID) and coolant entering and exiting TA 2 are shown in Figure 28. Cladding temperatures start experiencing rapid increases at around 0.6 seconds which corresponds to the rapid drop in flow rate. Peak cladding temperatures reach . Peak cladding temperatures in rods - steadily decline after reactor scram except for a brief temperature excursion of around 50°C (90°F) between 6 and 7 seconds.

A FRAPTRAN analysis of cladding integrity during this LOCA in the reactor pool transient was performed using minimum and maximum pellet-clad gap tolerances of and . Cladding outer diameter (OD) temperatures as a function of time and axial location were obtained from RELAP5 and used as the boundary condition for the FRAPTRAN analysis. FRAPCON results used in the reactivity transients in Section 6.2 were used to define the bumup and fission gas release inputs for FRAPTRAN at beginning of life (BOL) and end of life (EOL) conditions. The maximum hoop strain occurs at EOL conditions in target rod

• with the maximum gap tolerance of The cladding temperatures used in the FRAPTRAN analysis for rod

• are presented in Figure 29. The maximum hoop strain of 0.42% at the peak strain location along with the radial and axial strains at that location are presented in Figure 30. These strains are much less than the 1% strain criteria so that cladding integrity is maintained. The external pressure at BOL conditions is less than 15% greater than the internal pressure so that buckling of the cladding is not possible.

53

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

• • • • • • • Rod 6

- - Rod 5

• • • • • • • Rod 22

- - Rod 17

• • • • • • • TA 2 Outlet

- - TA 2 Inlet 0.01 0.1 1 10 100 Time (sec)

Figure 28. Maximum cladding ID and coolant temperatures during a LOCA in the reactor pool 54

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report u

-- - O.Os bl)

......... O.Ss QJ

~ - - 0 .6s QJ a.. ......... 0 .7s I'll QJ - - 1.3s Cl.

E - 2.os

~

- - 2.Ss

---

3.0s

-- -4.0s 0 10 20 30 40 50 60 Axial Height (cm )

Figure 29. Maximum cladding OD temperature profile in rod

• during a LOCA in the reactor pool 0 .005 0.004 r*....

,.,. '*'\

c

~ 0 .003 I

I

' ~ ....*-.

.... , / '\

VI  ;

ii - - - Hoop Stra in 6

~ - Axial Stra in

"' 0 .002

u. .,. ... ** Rad ial Stra i n r.

.........~.,..,,,!

0 .001 ......,, ... _~ _ .....""' *-

• 0 0 1 2 3 4 Time (seconds) 55

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report 6.4 Loss of Target Flow The loss of target flow accident can be initiated by inadvertent valve closures, pump failures, or loss of electrical power. Pipe breaks discussed in the previous section also result in a loss of target flow. Inadvertent valve closure is prevented by securing the valves in the open position using a locking pin since there is no safety requirement associated with the operation of the manual isolation valves and are there only for maintenance purposes. Therefore, the most limiting initiating event is the loss of pump flow (LOPF) which can be caused by either loss of electrical power or pump failure. The LOPF event impacts both target assemblies and includes pump coastdown and fluid momentum to ease the transition from forced flow to natural circulation flow.

As shown and discussed in the Target System Cooling Calculation Report and depicted in Figure 1, the system has redundant 100% capacity pumps. Only one of the pumps is required to operate and the other pump is a backup. If the operating pump fails , there is no automatic switchover to the backup pump. Redundant flow signals in the target cooling system will initiate protective actions including reactor SCRAM when the flow either reduces to 85% of nominal or increase to 115% of nominal. Additional actions taken due to the flow reduction are opening of the target DHRS valves.

The LOPF was modeled using RELAP5 mod3.3 patch 03. The RELAP5 model is the same model used for the LOCA analysis described in Section 6.3.

The flow transient caused by this LOPF is shown in Figure 31. The transient analysis starts with the target cooling system at 100% power and 100% nominal flow to accurately simulate the expected response of the protection system. The LOPF accident is assumed to include a loss of secondary flow as one would expect during a loss of site power. The analysis assumes that the target DHRS valves do not open. The LOPF causes a decrease in the flow measurement in the pipes supplying TA 1 and TA 2. Pump coastdown takes about 10 seconds to complete.

The reactor is scrammed at 0.14 seconds when the mass flow rate drops below 85%. Control blade movement is delayed by 0.1 5 seconds after the reactor SCRAM signal is generated. As natural circulation progresses, the colder water in the target cooling system is slowly replaced with water at pool temperature which eventually results in a decline in natural circulation flow at around 600 seconds.

56

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

• * * * * *

• Target 2 Flow Meter

- - Target 2 Outlet 0.001 0.01 0.1 1 10 100 1000 Time (sec)

Figure 31 . Mass flow transient during loss of pump flow Maximum cladding ID temperatures and coolant temperatures are shown in Figure 32. Peak cladding temperatures rise slightly by but then steadily decline after reactor SCRAM . Coolant exit temperature from TA 2 rises slightly at around 5 seconds and again after 600 seconds. At 600 seconds the inlet temperature starts to rise as reactor pool water has mixed with the cooler target cooling water. Eventually the target inlet temperature will reach the maximum pool temperature of 50°C (120°F) at which point further increase in temperatures will cease.

Because the pellet and cladding temperatures are near normal values, all cladding stresses and strains are also normal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOPF. The target DHRS valves can improve the natural circulation cooling in the target assembly but are not necessary to assure cladding integrity.

57

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

• • • • • • • Rod 22 Clad ID

- - Rod 17 Clad ID

• * * * * *

• Target 2 Inlet

- - Target 2 Outlet Time (sec)

Figure 32. Maximum cladding ID and coolant temperatures during loss of pump flow 7 CALCULATIONS FILES All computer files used in support of this design calculation report are included in Windchill and listed in the following sections.

7.1 RELAPS Files Steady state RELAP5 runs were used to set the thermal boundary conditions for the reactivity transients in FRAPCON. The following filenames were used with file extensions of

<filename> .inp for input and <filename> .out for output.

• 1thruB_1O.OMw_mssnomF - constant 10 MW (100%) power with - gap and nominal flow

• 1thruB_1 O.OMW~SnomF - constant 10 MW ( 100%) power with - gap and nominal flow

• 1thruB_ 10 . 0MW~SnomF - constant 10 MW (100%) power with - gap and nominal flow

• 1thruB_ 11 . 5MW~SnomF - constant 11 .5 MW (115%) power with - gap and nominal flow 58

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B

• 1thruB_11.5MW--SSnomF - constant 11.5 MW (115%) power with - gap and nominal flow

• 1thruB_11.5Mw_mssnomF - constant 11.5 MW (115%) power with - gap and nominal flow

• 1thruB_11.5Mw_mss1owF - constant 11.5 MW (115%) power with-gap and low (85%) flow

• 1thruB_11.5MW~SlowF - constant 11.5 MW (115%) power with - gap and low (85%) flow The RELAP5 runs for the LOCA and LOPF use the following filenames and also include input and output files.

• 1thruB_10.0MW--LOPFwV - loss of pump flow at 10 MW (100%) power with target DHRSvalves

• 1thruB_10.0MW--LOPFnoV - loss of pump flow at 10 MW (100%) power without target DHRS valves

• 1thruB_10.0MW...LOCA1wV - loss of coolant out of reactor pool before wye at 10 MW (100%) power with target DHRS valves

• 1thruB_10.0MW--LOCA1noV - loss of coolant out of reactor pool before wye at 10 MW (100%) power without target DHRS valves

• 1thruB_10.0MW--LOCA2wV - loss of coolant out of reactor pool after flex line at 10 MW (100%) power with target DHRS valves

• 1thruB_10.0MW--LOCA2noV - loss of coolant out of reactor pool after flex line at 10 MW (100%) power without target DHRS valves

• 1thruB_10.0MW...LOCA3wV - loss of coolant in reactor pool at 10 MW (100%) power with target DHRS valves

• 1thruB_10.0MW...LOCA3noV - loss of coolant in reactor pool at 10 MW (100%)

power without target DHRS valves 7.2 FRAPCON Files Four FRAPCON cases were run to provide input for the FRAPTRAN reactivity transients and the LOCA in the reactor ool. The followin filenames were used with file extensions of 59

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report

<filename> .inp for input, <filename> .out for output, <filename> .plot for plotting, and

<filename>.rst for restart which is used by FRAPTRAN.

• frapcon._10.0MWSS3B - constant 10 MW (100%) target irradiation with - gap

• frapcon._10.0MWSS3B - constant 10 MW (100%) target irradiation with - gap

• frapcon.600pcm3B - 115-100-115 % target power history with - gap

• frapcon.600pcm3B - 115-100-115 % target power history with - gap 7.3 FRAPTRAN Files For each reactivity transient, the minimum and maximum gap between the pellet and cladding were run after 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of operation and after - of operation for a total of eight cases. The following filenames were used with file extensions of <filename>.inp for input,

<filename> .out for output, and <filename> .plot for plotting.

• fraptran.600pcmB_bolB_NewR - 600 pcm reactivity transient with - gap at BOL conditions

• fraptran.600pcmB_eol_NewR - 600 pcm reactivity transient with - gap at EOL conditions

• fraptran.600pcmB_bol_NewR - 600 pcm reactivity transient with - gap at BOL conditions

• fraptran.600pcmB_eolB_NewR - 600 pcm reactivity transient with - gap at EOL conditions

• fraptran.30pcmpers_bolB_NewR - 30 pcm per sec reactivity transient with -

.gap at BOL conditions

• fraptran.30pcmpers_eolB_NewR - 30 pcm per sec reactivity transient with -

gap at EOL conditions

• fraptran.30pcmpers_bolB_NewR - 30 pcm per sec reactivity transient with -

gap at BOL conditions

• fraptran.30pcmpers_eolB_NewR - 30 pcm per sec reactivity transient with -

gap at EOL conditions 60

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441 R00032/B FRAPTRAN was also used to assess cladding integrity during the LOCA in the reactor pool.

The following filenames were used and also include input, output, and plotting files.

• fraptran.CladTemp_Rod17_bolB - LOCA in pool, rod 17 w i t h - gap at BOL

• fraptran.CladTemp_Rod17_eolB - LOCA in po~I. rod 17 with - gap at EOL

• fraptran.CladTemp_Rod22_bolB - LOCA in pool, rod 22 with - gap at BOL

• fraptran.CladTemp_Rod22_eolB - LOCA in pool, rod 22 with - gap at EOL

• fraptran.CladTemp_Rod17_bolB - LOCA in pool, rod 17 with - gap at BOL

• fraptran.CladTemp_Rod17_eolB - LOCA in pool, rod 17 w i t h - gap at EOL

• fraptran.CladTemp_Rod22_bolB - LOCA in pool, rod 22 with - gap at BOL

• fraptran.CladTemp_Rod22_eolB - LOCA in pool, rod 22 with - gap at EOL 7.4 EXCEL Files i

EXCEL was used to do simple manipulations of data to prepare it for input to RELAPS, FRAPCON, and FRAPTRAN. Table 15 lists the EXCEL files generated and their purpose in the calculational scheme.

Table 15 Excel Files EXCEL filename Purpose Power-profile.xlsx Steady state power distribution 600pcm (Final) with Target.xlsx Reactivity transient power 7.5 MATHCAD Files MATHCAD was used to calculate the flow areas, equivalent hydraulic diameter and equivalent heated diameter for the minimum, nominal, and maximum gaps between the pellet and cladding. The following three MATHCAD files were generated:

• Areas - RevB.xmcd - minimum gap hydraulic calculations

• Areas - RevB.xmcd - nominal gap hydraulic calculations

• Areas - RevB.xmcd - maximum gap hydraulic calculations 61

ATTACHMENT7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design 30441 R00032/B Calculation Report MATH CAD was also used to calculate the minimum critical heat flux ratio using results from the LOCA and LOPF RELAP5 cases. The following two MATHCAD files were used:

• 1thruB_10.0MW--LOCA1noV.xmcd - loss of coolant out of reactor pool before wye at 10 MW (100%) power with target DHRS valves

• 1thruB_10,0MW--LOCA2noV.xmcd - loss of coolant out of reactor pool after flex line at 10 MW (100%) power without target DHRS valves

• 1thruB_10.0MW--LOCA3noV.xmcd - loss of coolant in reactor pool at 10 MW (100%)

power without target DHRS valves

• 1thruB_10.0MW--LOPFnoV.xmcd - loss of pump flow at 10 MW (100%) power without target DHRS valves Text files containing the RELAP5 data for select time points are as follows:

• 1thru_10.0MW--LOCA1noV_022.txt - at time 0.22 sec

• 1thru_10.0MW--LOCA1noV_030.txt - attime 0.30 sec

• 1thru_10.0MW--LOCA1noV_038.txt - at time 0.38 sec

• 1thru_10.0MW.JILOCA2noV_022.txt - at time 0.22 sec

• 1thru_10.0MW.JILOCA2noV_030.txt - at time 0.30 sec

• 1thru_10.0MW_JILOCA2noV_038.txt - at time 0.38 sec

• 1thru_10.0MW.JILOCA3noV_022.txt - attime 0.22 sec

• 1thru_10.0MW--LOCA3noV_030.txt - attime 0.30 sec

• 1thru_10.0MW--LOCA3noV_038.txt - at time 0.38 sec

• 1thru_10.0MW--LOCA3noV_040.txt - at time 0.40 sec

• 1thru_10.0MW.JILOCA3noV_050.txt - at time 0.50 sec

• 1thru_10.0MW.JILOCA3noV_060.txt - at time 0.60 sec

• 1thru_10~0MW.JILOCA3noV_070.txt - at time 0.70 sec 62

ATTACHMENT 7 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report

• 30441 R00032/B

• 1thru_10.0MW--LOCA3noV_080.txt - at time 0.80 sec

• 1thru_10.0MW--LOCA3noV_090.txt - at time 0.90 sec

• 1thru_10.0MW--LOCA3noV_100.txt - at time 1.00 sec

• 1thru_10.0MW--LOPFnoV_OOO.txt - at time 0.00 sec

• 1thru_10.0MW--LOPFnoV_030.txt - at time 0.30 sec 8 REFERENCES 1 Missouri University Research Reactor (M.URR) Safety Analysis Report, Vols. 1 and 2, August 18, 2006.

2 "Written communication as specified by 10 CFR 50.4(b)(1) regarding responses to the "University of Missouri at Columbia - Clarifications Need to Nuclear Regulatory Commission Staff Request for Additional Information Regarding the Renewal of Facility Operating License No. R-103 for the University of Missouri at Columbia Research Reactor (TAC No. ME1580)," dated December 17, 2015," letter from John L. Fruits to U.S. Nuclear Regulatory Commission, February 8, 2016.

3 RELAP5/MOD3.3 Code Manual (NUREG/CR-5535/Rev P3-Vol I - VIII), March 2006.

4 Geelhood, K. J., et al., "FRAPCON-4-0: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup," PNNL-19418, Vol.1, Rev. 2, September2015.

5 Geelhood, K. J., et al., "FRAPTRAN-2.0: A Computer Code for the Transient Analysis of I

Oxide Fuel Rods," (PNNL-19400, Vol.1 Rev.2), May 2016._

63

,*I.

' ~"' ' I ,!

ATTACHMENT 7

+ GSNSIUJL ATOMICS P.O. BOX 85608 SAN DIEGO, CA 92186-5608 (858) 455-3000