NucE 431W Core Design Presentation Bayshore Unit 1 Reload Core Design Group 13 Michael Bertino Michael Stachnik Submitted to: Dr. K Ivanov Dr. M. Avramova.

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NucE 431W Core Design Presentation

Bayshore Unit 1 Reload Core DesignGroup 13

Michael Bertino

Michael Stachnik

Submitted to:

Dr. K Ivanov

Dr. M. Avramova

Mentor: Chris Wagener

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Table of Contents

• Introduction• Loading Pattern Development• Safety Analysis• Operational Data • Thermal Hydraulics Analysis• Conclusion

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INTRODUCTION

Student Objectives:• To be able to use the codes and methods used by Westinghouse to

generate a core loading pattern for a cycle 13 Bayshore Unit 1 reactor and perform reload design analysis.• Perform an analysis for operational conditions and safety

requirements for our core loading pattern.• Perform a thermo-hydraulic analysis on our core loading pattern for

both steady state and transient conditions.

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What is ANC?

• ANC (Advanced Nodal Code)• Multidimensional nodal code. • Licensed by NRC for PWR analysis• Calculates:

1. Core reactivity2. Assembly power3. Rodwise Power4. Reactivity coefficients5. Core depletion6. Control rod and fission product worths

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Reactor Core Design

CE 2-PWR Loop Core• Thermal Power=2700 MWt• 4 Control Rods• 217 assemblies• 5 guide tubes

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Plant Description

• Inlet core temperature is programmed to vary from 532 to 549 from 0 % to 100 % power.• Control rods move from 0 to 137 steps withdrawn.• Rod insertion limits are a function of power.• Full power upper limit of the axial shape index (ASI) is -8 %.

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Loading Pattern Development

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Loading Pattern Development

There are four criteria that must be evaluated to development an initial loading pattern (LP)1. Energy(cycle length)

2. F peaking factor, ARO3. HZP, MTC (all power levels)4. Fuel Inventory

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Loading Pattern Parameters

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Gadolinia Burnable Absorber

• 68 feed assemblies • 36 assemblies at 4.013 w/o U235

• 20 assemblies at 4.420 w/o U235

• 12 assemblies at 4.365 w/o U235

• Mixed with UO2, displaces fuel from the core.• Complex depletion chain.• No placement restrictions.• Optimized to reduced peaking within the assembly

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Summary of ANC Runs to Final LP

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Final Core Loading Pattern

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Energy, Cycle Length Requirement

• EFPD is defined as the total amount of energy produced from BOC to EOC.• Boron Concentration must be reduced to 10 ppm at HZP conditions• The final burnup step is used to calculate the EFPD of the LP:

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• The final burnup step SL213_BE15 concentration target to around 10 ppm for an EFPD of 468.2 but the boron concentration of the input was 16 ppm with an EFPD of 469.51. This limit is confirmed at the final burnup step as it should be.

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FH Limit Confirmation

• FΔH is the normalized enthalpy rise in a given subchannel as the water flows from the bottom of the core to the top of the core. • Represents a localized power with in the core (local power > average power)• The peaking factor (FΔH) is defined in ANC by:

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• Below is the C-FDH for the 150 BU, the hottest step.

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• The figure below is the 12th step showing a peak in FΔH due to the boron burning up.

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MTC Limit Confirmation

• The moderator temperature coefficient is defined as the reactivity change per one degree change in the fuel temperature. In a PWR, the moderator is water in the liquid form and the basic units of MTC are pcm/degree temperature.• The MTC of water is negative at most conditions due to as

temperature increaseswater density decreasesmoderation decreasesless reactivity• The more boron that is dissolved in the moderator, the more positive

the MTC will be. Water has a naturally negative temperature coefficient while boron has a naturally positive one.

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• The MTC is checked at HZP conditions for the core, RELPOW=0 /• The MTC should also have

no xenon, DEPLETE= HDALL, NAXE /

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• The calculation to solve for the MTC of the first case to see if the limit was below 0.50 pcmF

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Loading Pattern Limit Confirmation

Margins Target Values Actual Values

Energy 468.2 EFPD 469.51 EFPD

ARO peaking factor(Fdh)

1.635 1.627

HZP, MTC (all power levels)

.50 pcm/F .37 pcm/F

Fuel inventory 68 Feeds 68 Feeds

Boron Concentration(ppm)

10 ppm 16 ppm

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Safety Analysis

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Westinghouse RSAC Process

• Reload Safety Analysis Checklist.• Transient analyst does calculations to determine damage to the core

and environment in case of accident.• Core Designer must confirm that reload design does not violate

assumed values. • Always go to the extreme, worst case scenario• Safety calculations done in conservative manner.(most limiting

condition)

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Safety Analysis

Westinghouse RSAC Process:1. Rodded FH2. Rod Ejection Accident3. Shutdown Margin

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Rodded FH

• Rodded FH must be met when the leading control rod bank is in to is insertion limit (RIL). • The RIL is the deepest possible insertion for any rod bank. This is done

to make sure there is enough rod worth left to shut down the core in case of accident or emergency.

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Rodded FH

Input file for rodded FH:• Xenon must be reconstructed• Xenon must be skewed to save time

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Rodded FH

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Rodded FH

• Below is the Rodded FH output from E-SUM:

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Rod Ejection Accident

• A mechanical failure where a control rod is ejected from the core• Causes large power increase, fuel and clad temperature increase and

increase in DNB• The limits found are the ejected rod worth Δρ(E), and the ejected rod

hot channel factor FQ(E)

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• There are two limits evaluated at two conditions. 1. BOC, HFP, ARO, equilibrium xenon2. EOC, HFP, ARO, no xenon3. BOC, HZP, ARO, equilibrium xenon4. EOC, HZP, ARO, non xenon

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Rod Ejection Accident, HFP

• The most limiting rod ejection is at the EOC• The most limiting

FQ is at BOC• Sample input deck

from BOC RELPOW=1.00

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The output E-SRW for HFP BOC

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The output E-SRW for HFP EOC

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Rod Ejection Accident, HZPThe rod ejection for HZP is the same thing except RELPOW=0 /

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Rod Ejection Accident, HZP

The output E-SRW for HZP BOC

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The output E-SRW for HZP EOC

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Shutdown Margin• The amount of reactivity in the core at subcritical following a trip. • Shows that the operators will be able to safely shut down the core. • There are components that affect the SM in ANC:1. Doppler Defect-fuel pellet temperature increases with power and so does

resonance absorption due to Doppler. Doppler decrease reactivity2. Voids-Local boiling in the moderator can also cause small voids to form. Voids

decrease reactivity but collapsing gives a small increase. 3. Axial Flux redistribution-enthalpy in the core rises causing a flux tilt towards

the bottom of the core. When its goes from HFP to HZP(power defect) there is no rise in enthalpy causing the flux to shift to the top of the core which increases reactivity.

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Shutdown Margin

4. Power Defect- amount of total reactivity associated with a change in power. It is larger at EOC because the MTC is more negative due to less boron. So reactivity increases. 5. Rod insertion allowance- cannot assume a full worth of control rod banks. The core may have only partially inserted rods at trip. The reactivity depends on how much rod worth there is. 6. Variable Moderator Temperature- The moderator temperature is greater at HFP so when it trips to HZP causes the temperature to decrease causing a spike in reactivity.

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Shutdown Margin

Use six cases for ANC input:1. K1- Base Case at Burnup of Interest (BOC or EOC)

• EOC, boron should be set to 0 ppm• BOC, boron should be constant

2. K2-Rods at Insertion Limits • Lead bank is inserted which means less rod worth out of the core • Less negative reactivity upon trip

3. K3-Over-power/Over Temperature, Skew Power to Top of Core• Increase core power from 100%to 105% • Higher power means that the initial temperature will be higher and it will increase power defect• Xenon is skewed so that the AO shifts to most positive side(xenon to the bottom of the core,

shifts power to the top• Increases the worth of partially inserted rods and the power defect

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Shutdown Margin

4. K4-Trip to Zero Power• Holds the Xenon, boron and D bank and it goes from HFP to HZP

5. K5-Full Core-All Rods In• All rods inserted in full core

6. K6-Worst Stuck Rod Out• Removes the worst stuck rod

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Shutdown Margin

To calculate the shutdown margin we took the worst stuck rod case at BOC and EOC

This meets the limit for the shutdown margin by 1.866%.

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Shutdown Margin

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Shutdown Margin BOC WORTHS (pcm) EOC WORTHS (pcm)

POWER DEFECTS 1483.7 2548.22

Void Effects 50 50

Total Control Bank Requirement(1)

1533.7 2598.22

SDMCALC 7777.97 8489.18

Less 10% (2) 7000.18 7640.26

Available SDM (2)-(1) 5466.49 5042.03

Required SDM 3600 3600

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Operational Data

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Operational Data

Calculations that must be performed to make sure the core is running at normal conditions1. Rod Worth2. Xenon Worth3. Differential Boron Worth4. Isothermal Temperature Coefficient5. Critical Boron Concentration

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Rod Worth

Measured at BOC, HZPThe rodworth is found using the boron dilution method:1. Core at BOC, HZP, ARI, subcritical (high CB)2. Withdraw rods (ARO)

3. Dilute CB to ARO critical boron, CBARO

4. Insert lead control bank (Bank D)5. Insert Remaining Control Banks, One at a Time, in Normal Sequence

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Rod Worth

• As more rod worth is inserted, the fraction of rated thermal power decreases.• The lead bank in our case is bank5

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Rod Worth

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Rod Worth

Configuration CB (ppm) Inserted Worth (ppm)

ARO 1605 -----

D 1479 126

D+C 1436 43

D+C+B 1336 100

D+C+B+A 1247 89

D+C+B+A+BANK1 1197 50

D+C+B+A+BANK1+BANK6 884 313

D+C+B+A+BANK1+BANK6+BANK7

503 381

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Xenon Worth

• Reactivity due to the absorption on neutrons Xe-135 in the core. • A fission product with a high absorber worth produced from the

decay of I-135

• Xe-135 is removed by two effects:1. Absorption by thermal neutron flux2. Radioactive decay

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Xenon Worth

Xenon reactivity after startup and trip is what is calculated using ANC1. Startup• BOC,MOC,EOC at 50 % and 100 % Power

2. Trip • BOC,MOC,EOC at 50 % and 100 % Power

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Xenon Worth

Speed up calculations core is collapsed from 3-D to 2-DDeplete the Xenon for over 100 hours

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Xenon Worth

Use the Eigenvalues to calculate the reactivity

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Xenon Worth, Startup

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Xenon Worth, Startup

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Xenon Worth, After Trip

~9Hours

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Xenon Worth, After Trip

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Differential Boron Worth

• Change in reactivity due to a unit change in boron concentration. • Calculated at both HZP and HFP

varying by ±25 ppm.• Input sample from HZP job file • At HZP, differential boron

worths were calculated at a wide range of boron concentrations

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DBW -

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• Shows the amount of soluble boron throughout the cycle at HFP

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• HZP DBW for 5 different boron concentrations over the length of the cycle

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Isothermal Temperature Coefficient

• ITC is defined as the change in reactivity of a core with a change in core temperature.

• DTC is defined as the reactivity change per one degree change in the fuel temperature. In a PWR, the fuel is UO2 in ceramic form and the basic units of DTC are in pcm/F or pcm/C.

• It will follow the behaviors of both the MTC and DTC.• DTC remains constant, ITC will follow the MTC, boron decreases so does ITC.

• In ANC the most limiting case is at BOC HZP where the boron is at the highest.

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• Input deck and E-Sum for ITC

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ITC must be calculated by doing it the old fashion way, by hand. MTC and DTC are found by ANC and ITC is found using the equation below:

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Critical Boron Concentration

• Critical boron concentration is found at BOC, HZP, ARO, no xenon• It is done to predict cycle length and reactor control. • It has a heavy effect on MTC and Xenon Worth.• Good agreement between the measured value and the value

predicted by the design code. Gives accuracy of the design model of the reactor. • NRC says that the difference should not exceed 1 %.

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Critical Boron Concentration

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Thermal-Hydraulic Analysis

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Why Thermal-Hydraulic Analysis is Needed• A thermal-hydraulic analysis is necessary for any core design before

the design is implemented. • The goal is to determine the range of operation and the conditions

that the reactor can safely operate without resulting in fuel failure over the reactor life considering both steady-state and anticipated transient operation.• With thermal hydraulic analysis, the temperature distribution

throughout the core can be determined with a given fission power distribution and coolant inlet condition.

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Core Design Subchannel Code (CDSC)• CDSC is a three-dimensional thermal hydraulic code that solves for

mass flow, quality, void fraction, fluid temperature, and pressure for each subchannel.• CDSC also models assembly-to-assembly mixing as well as

subchannel-to-subchannel mixing.• CDSC assumes homogeneous equilibrium two-phase flow (no slip and

same temperatures for each phase).• The output provides a 3D enthalpy and flow distribution and data for

the evaluation of thermal safety limits.

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Boundary Conditions Nominal Power

• Inlet Mass Flux: 3033.889kg//s• System Exit Pressure: 15MPa• Inlet Temperature: • Power Input: 485676.91 w/• Spacer Grid Loss Coefficient: 0.8• 3D power Distribution

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Geometry of the Subchannel Analysis• Flow Channel Dimensions• Rod Dimensions• Flow Channel Gap and Distance• Heated and Wetted Parameters

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Overpower Uncertainties

• 5% decrease in Inlet Mass Flux: 2882.21kg//s• 50psia Increase in Exit Pressure: 15.34MPa• in Inlet Temperature: • Power Input Increased until DNBR of 1.3: 734000.91w/• 10% Increase in Spacer Grid Loss Coefficient: 0.88• Decrease of Pitch by 0.006 inches in Hottest Subchannel: 0.5in

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Nominal and Overpower Cases

Type Power Hottest Rod Hottest Channel Axial Location (m) Min DNB Ratio

Nominal 100% 16 22 2.8935-3.0382 3.3969

Overpower+unc 151% 28 22 3.1828-3.3275 1.3007

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Bayshore Reactor Vessel at Nominal Power and Overpower

Water Out at 327.13 CWater In at 278 CWater In at 285 C Water Out at 344.79 C

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Coolant Temperature

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Fuel Temperature in Nominal and Overpower

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Cladding Temperature

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Departure From Nucleate Boiling Ratio

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Void Fraction

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Mass Flux

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Conclusion

• Learned how to develop a loading pattern under restrictions.• Made sure our core was safe using the RSAC process.• Calculated operational data to confirm that our core performed

properly at all phases of the cycle.• Performed thermal-hydraulics calculations to evaluate the

subchannels and rods to nominal and overpower conditions.

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References

1.Dr. K. Ivanov, Dr. M. Avramova. Nuclear Engineering 431W: Nuclear Reactor Core Design. Department of Mechanical and Nuclear Engineering, The Pennsylvania State University: Spring 2014.2. C. Wagener. Core Design Training Course. Westinghouse Electric Company, presented to Penn State University: Spring 2014

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QUESTIONS?

NucE 431W Core Design Presentation Bayshore Unit 1 Reload Core Design Group 13 Michael Bertino Michael Stachnik Submitted to: Dr. K Ivanov Dr. M. Avramova.

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