Enhanced Safety Analysis Code 

Suites for the New Reactor 

Design at NCNR

Richard Leos

NIST Center for Neutron Research

Reactor Operations and Engineering Group
Status of the NBSR

• The lifetime of the National Bureau 

of Standards Reactor (NBSR) will be 

coming to an end sometime in the 

next few decades.

• NBSR Main Characteristics:

• High-Enrichment Uranium (HEU) fuel: 93 wt%

• U3O8 + Al

• Vertically Split Fuel Element

• Full Power: 20 MW

• D2O Coolant , Moderator, Reflector
Overview of the New Reactor

• Reactor Core Characteristics:

• Low-Enrichment Uranium (LEU) fuel: 19.75 wt%

• U3Si2 + Al

• Horizontally Split Core 

• Two Cold Neutron Sources (CNS)

• Full Power: 20 MW

• H2O Coolant, Moderator

• D2O Reflector
Parameter Changes for New Reactor

Parameters NBSR Split Core

Half thickness of fuel meat (c) 0.0254 cm 0.033 cm

Length of fuel meat (H) 55.88 cm 60 cm

Number of fuel elements 30 18

Views of Single Channel

Top View

Side View
Introduction to PARET/ANL Code

• Program for the Analysis of REactor Transients developed by Argonne National 

Laboratory (ANL)

• Intended primarily for the safety analysis of test and research reactors that use 

plate-type (flat) fuel elements

• Based on an evaluation of the coupled thermal, hydrodynamic, and nuclear 

effects of the core

• Program calculates Critical Heat Flux Ratio (CHFR) using the Mirshak Correlation.
Critical Heat Flux & Onset of Flow Instability

Critical Heat Flux (CHF)

• The thermal limiting condition 

where a phase change occurs during 

heating which decreases efficiency 

of heat transfer causing localized 

overheating of the heating surface.

Onset of Flow Instability (OFI)

• Excursive flow instability due to the 

onset of net vapor generation in the 

coolant channel.

Probability 

Level CHFR OFIR

90% 1.301 1.310

95% 1.391 1.403

99.9% 1.778 1.828

Thermal Limits

Criteria

(A) (B) (C)
Objectives

• Upgrade the critical heat flux ratio (CHFR) calculations for the PARET/ANL output 

by using the Sudo-Kaminaga correlation

• Determine the safety margins for various transient cases based on the critical 

heat flux ratio (CHFR) and onset of flow instability ratio (OFIR)
Sudo-Kaminaga Correlation

• Used to calculate CHFR for vertical 

rectangular channels of a research 

reactor

• Possesses greater geometric similarities 

to the new reactor

• Considers effects:

o Pressure

o Inlet sub-cooling

o Outlet sub-cooling

o Channel configuration

o Mass flux

o Flow direction

• qqCCCCCC,1

∗

= 0.005 GG∗ 0.611

• qqCCCCCC,2

∗

=

AA

AAHH

GG∗ ΔTTssssss,iiii

∗

• qqCCCCCC,3

∗

= 0.7 AA

AAHH

WW

λλ

1+ ρρgg

ρρll

1

4

2 (1.0 + 3.0ΔTTssssss,iiii

∗ )

• qqCCCCCC,4

∗

= 0.005 GG∗ 0.611 5000

GG∗ ΔTTssssss,0

∗

• qqCCCCCC

′′

= qqCCCCCC

∗ hffff λλ(ρρll − ρρ )ρρ gg

• CCCCCCCC =

qqCC CC

′′

qqmmmmmmmm

′′
Sudo-Kaminaga Correlation (Cont.)

Sudo-Kaminaga Correlation Scheme

Dimensionless Mass Flux

Dimensionless Critical Heat Flux
OFI Criteria: Saha-Zuber

• Mass flow rate criteria is based on Peclet number: PPPP = GGDDhCCpppp

kk ≤ 70,000

• For low mass flux (Pe ≤ 70,000): NNNN = qq′′DDh

kk (TTssssss−TTλλ) = 455

• For high mass flux (Pe ˃ 70,000): SSSS = qq

GGCCpppp(TTssssss−TTλλ) = 0.0065

• Heat flux for OFI and OFIR: qqqqOO OO = �

455 ∗ hffff ∗ kkff ∗ (TTssssss − TTλλ), PPPP ≤ 70,000

0.0065 ∗ GG ∗ CCpp ∗ TTssssss − TTλλ , oo oooooooo

• OO OO =

qq OO OO

qq mmmmmmmm
Coding for Calculations

Define 

starting node 

& time step

Apply CHFR 

& OFIR 

calculations

Arrange 

data in an 

array

More nodes 

to analyze?

Yes

End 

Loop

Define next 

node & 

time step

No
10 11 12 13 14 15

0

5

10

15

20

25

30

35

Power During RIA

time (s)

Power (MW)

Power Applied

Power when Scram occurs

Simulated Transient Cases

Reactivity Insertion Accident (RIA)

• Positive reactivity insertion in the core 

that may be caused by experiments 

removed from the core

Loss of Flow Accident (LOFA)

• A core heat up due to malfunction of 

the cooling system even if the reactor 

power is operating at nominal value

• Parameters

• Reactor operates at full power (20 MW)

• Flow decay modeled as exponential decay 

function

• Scram occurs when flow decay is reduced by 

15%

24
Case 1: Small Reactivity Insertion Accident

• Initiating Power: 2 W

• Reactivity insertion at slow ramp rate: $0.1/s

Variation of the Minimum CHFR Variation of the Minimum OFIR 

6 8 10 12 14 16 18 20 22

100

101

102

103

104

105

106

Time (s)

MCHFR

Sudo-Kaminaga Correlation

Mirshak Correlation

Limiting MCHFR for LEU

0 5 10 15 20 25

100

101

102

103

104

105

106

107

108

Time (s)

Minimum OFIR

MOFIR

Limiting MOFIR for LEU
Case 2: Large Reactivity Insertion Accident

• Initiating Power: 20 MW (full power)

• Reactivity insertion: $1.5 in 0.5s 

Variation of the Minimum CHFR Variation of the Minimum OFIR 

0 5 10 15 20 25

100

101

102

Time (s)

MCHFR

Sudo-Kaminaga Correlation

Mirshak Correlation

Limiting MCHFR for LEU

0 5 10 15 20 25

100

101

102

103

Time (s)

Minimum OFIR

MOFIR

Limiting MOFIR for LEU
Case 3: Slow Loss of Flow Accident

• Decay constant (T): 25 s

Variation of the Minimum CHFR Variation of the Minimum OFIR 

0 10 20 30 40 50 60 70 80 90

100

101

102

103

Time (s)

MCHFR

Sudo-Kaminaga Correlation

Mirshak Correlation

Limiting MCHFR for LEU

0 10 20 30 40 50 60 70 80 90

100

101

102

103

Time (s)

Minimum OFIR

MOFIR

Limiting MOFIR for LEU
Case 4: Fast Loss of Flow Accident

• Decay constant (T): 1 s

Variation of the Minimum CHFR Variation of the Minimum OFIR 

0 5 10 15 20 25

100

101

102

Time (s)

MCHFR

Sudo-Kaminaga Correlation

Mirshak Correlation

Limiting MCHFR for LEU

0 5 10 15 20 25

100

101

102

Time (s)

Minimum OFIR

MOFIR

Limiting MOFIR for LEU
Summary on All Cases

Sudo-Kaminaga Mirshak

Case # MCHFR MCHFR MOFIR

1 2.81 2.17 6.06

2 3.29 2.50 7.18

3 3.63 2.99 6.99

4 3.85 2.80 3.69

All MCHFR values are above 1.778 (Thermal Limit) 

All MOFIR values are above 1.828 (Thermal Limit)
Conclusion

• The results for these cases show that there is at least a 99.9% 

probability of no fuel damage.

• The new reactor under these parameters can be deemed safe.
Acknowledgements

• Advisors: Dr. Zeyun Wu & Dr. Robert Williams

• SURF Directors: Dr. Julie Borchers and Dr. Joseph Dura