April 17, 2007
Rev.0
Japan Nuclear Technology InstituteAnalysis on Criticality ‘Accident’ Occurred at Shika 1 of Hokuriku ElectricPower Company

1. Overview
In March 2007, it was revealed that Shika 1 core had become
criticality during outage due to unexpected withdrawal of 3 control
rods in June 1999.

Therefore, Japan Nuclear Technology Institute(JANTI) made core performance analyses using information provided
from Hokuriku Electric Power Company (Hokuriku EPCO).
Analysis result showed that in the conservative condition of control
rods withdrawal speed with the associated reactivity inserted(standard case), it was possible that the core had been in prompt
criticality.

The power decreased instantly (in 0.3 second) following

rapid increase to 14% (230MW) of rated thermal power after 6 secondsof (delayed) criticality.

The maximum enthalpy increase during peak

power period was calculated to be 13cal/gUO2, which is well below fuel
PCMI failure threshold of 85cal/gUO2(*1).

Also, the maximum fuelenthalpy was calculated to be 49cal/gUO2, which is below limit value of
230cal/gUO2(*2) in accident or 92cal/gUO2(*3) in abnormal transient
during operation. In some cases such as with low control rod
withdrawal speed, the core status did not result in prompt criticality,
and stayed in delayed criticality.
*1)

Threshold

value

for

PCMI

(Pellet-Cladding

MechanicalInteraction) failure*2) Threshold to prevent occurrence of mechanical energy bypressure impact resulted from fuel failure due to pellet melting
and vaporization.
*3) Threshold to prevent fuel failure due to high temperature rapture,1

melting and nil-ductility of cladding
2. Analysis Condition(1)

Determination of Analysis Condition
Timeline of input parameter determination is shown in Table 1.
In the determination process, sensitivity analysis was made on theassociated parameter to focus on parameters with high priority, sinceexact value was not available at first.

Then, after checking analyses

condition of Hokuriku EPCO, analysis condition of standard case wasdetermined.

Also, variable range was determined for inserted

reactivity and control rods withdrawal speed in analyses.
(2)Power Distribution in the CoreAt Shika 1, criticality occurred as 3 out of 89 control rods werewithdrawn.

The situation can be understood that ‘small partial

core’ was constituted inside the full core (Figure 1).

The power

distribution had shape of top peak as shown in Figure 2.

In thepartial core where control rods were withdrawn, 70% of power was
generated in 4% of the full core volume.

Kinetics importance of thepartial core was estimated to be equivalent to the full core.

Thus,

JANTI analysis was made on the partial core shown in Figure 1.
(3)

Inserted Reactivity and Control Rods Speed in Standard CaseIn the analyses, inserted reactivity and control rods withdrawal
speed were considered as parameter.

For both parameters, basic

values (standard case) were determined as follows.
-

Keff in Standard Case: 1.0079-

Control Rods Speed in Standard Case: 47mm/s

Inserted reactivity (Keff) of the core in the standard case wasdetermined based on the analysis by Hokuriku EPCO.
Control rods withdrawal speed in the standard case wasdetermined as practically fastest speed based on mockup test by2

Hokuriku EPFO.
(4)

Affect due to Inserted Reactivity and Control Rods Speed
In standard case, inserted reactivity (0.0079ΔK =1.3$) is above β(0.0060=1$).

However, as extent of power increase depends on thereactivity insertion rate, it was not to be concluded that there
was/was not prompt criticality occurred just based on value of
inserted reactivity in excess of β.

Therefore, sensitivity analyses

were made to identify cases which result in prompt criticality.
For inserted reactivity, analyses were made on the core withhigher/lower reactivity by 0.5$ each in addition to the standard caseconsidering accuracy of analysis code.
- Standard Case: 1.3$
- High Reactivity: 1.8$ (0.5$ higher)- Low Reactivity: 0.81$ (0.5$ lower)
For control rods movement, the following three withdrawal speeds
were selected for analysis.
- Standard Case: 47mm/s- High Speed: 76mm/s (normal operation speed of control rods)- Low Speed: 16mm/s (assumed average speed of control rod(26-39) from 0pos. to 16pos. during 77
seconds between start of control rod(s)withdrawal signal and initiation of scram
signal.

(26-39) was at peaking power

location as shown in Figure 2)
Reactivity insertion rate in each case is shown in Figure 3.
(5)

Analysis Code
JANTI analyses were made using multi regional nuclear-thermal
hydraulics combined kinetics code EUREKA-2.

In EUREKA-2, both

nuclear and thermal hydraulics feedback (Doppler feedback andcoolant temperature feedback) can be treated simultaneously.

3

Inthe analyses by Hokuriku EPCO, kinetics calculation was made byreactivity insertion events analysis code APEX considering Doppler
feedback, followed by analysis using thermal hydraulics code SCAT
with APEX result as input.

Comparison of calculation method isshown in Table 3-1.
In the analyses by JANTI, EUREKA-2 code is used primarily
because of its high performance during peak power period that ischaracteristicofreactivity

insertion

accidents.

Meanwhile,

EUREKA-2 tends to calculate conservative results for fuel enthalpy
after power peak due to its assumption of constant power
distribution, and due to its incapability of calculation in boilingcondition etc.

Therefore, in JANTI analyses, result of fuel enthalpy

after power peak was considered as reference.

Also, as EUREKA-2

is not capable of calculation in boiling condition, higher core pressure
value was used for calculation to avoid boiling.
3. Analysis Result(1)Analysis Result of Standard Case
Trend of power are shown in Figure 4-1 and Figure 4-2.

Instandard case, 1.1$ of net reactivity was inserted, and rapid powerincrease due to prompt criticality occurred 6 seconds after (delayed)
criticality.

But because of inherent reactivity feedback mechanism

as shown in Figure 4-3, the power decreased instantly (in 0.3 second)
following rapid increase to 14% of rated power (230 MW).

Then, thepower became stable about 0.3% of rated power (4MW).
The maximum enthalpy increase during peak power was calculated
to be 13cal/gUO2, which is well below fuel PCMI failure threshold of
85cal/gUO2.

The analysis results were similar to the ones by

Hokuriku EPCO as shown in Table 2.

The fuel enthalpy increased

gradually to become maximum value of 49cal/gUO2 after 10 secondsof rapid power increase, which is well below limit value of
230cal/gUO2 during accident or 92cal/gUO2 during abnormal

4

transients during operation.
Maximum pellet temperature was about 700℃ as shown in Figure
4-4, and the maximum coolant temperature in the partial core was
the boiling temperature at core outlet as shown in Figure 4-5.
(2)Affect of Inserted Reactivity
Calculation results are shown in Table 4, Figures 5-1 and 5-2.

Ineach case, control rods withdrawal speed was considered to be
47mm/s.

In the cases where reactivity of more than 1$ was inserted,

the results were well below fuel PCMI failure threshold while prompt
criticality was observed.

In the case of large reactivity insertion,less than 0.1$ of difference was observed in the net inserted
reactivity as compared with the standard case.
(3)

Affect of Control Rods Movement Speed
Calculation results are shown in Table 5, Figures 6-1 and 6-2.
While reactivity of 1.3$ were inserted, the net inserted reactivity was
below 1$ in the case of low control rods withdrawal speed, and rapid
power increase was not observed.

In all cases, the results were well

below fuel PCMI failure threshold.
(4)

Affect of Analysis Model
Sensitivity analysis was made using “zero” coolant temperature
coefficient. The result is shown in Table 3-2.

No large affect due todifference in treatment of coolant temperature reactivity coefficient
was observed.
4.

Conclusion
Analyses were made on various conditions for criticality ‘accident’occurred at Shika 1 using inserted reactivity and control rods
withdrawal speed as variable parameters.

Analyses result showed

possibility of being prompt criticality in cases with conservative

5

condition.

On the other hand, in cases such as with low control rod

withdrawal speed, the core did not result in prompt criticality, andstayed in delayed criticality.

For all analyzed cases, the maximum

enthalpy increase during peak power period was well below fuel
PCMI failure threshold.

Results of JANTI analyses were equivalent

to the ones by Hokuriku EPCO.

Also, the reference analyses resultof maximum fuel enthalpy was below threshold value in accident orin abnormal transient during operation for all analyzed cases.
5.

Attachment(1)

Summary of “Treatment of High Burn-up Fuels in Reactivity
Insertion Accident of Light Water Reactor Generation Facilities”(2)

Summary of “Safety Analyses Review Guide for Light Water Reactor
Generation Facilities”(3)

Summary of “Reactivity Insertion Accident Review Guide for Light
Water Reactor Generation Facilities”(4)

Definition of Terms

6. Reference“Report Regarding Criticality Accident at Shika 1 NPS”
(April 6, 2007, Hokuriku Electric Power Company)“LWR Reactivity Insertion Accident Code EUREKA-2”(JAERI-M 84-074, May 1984, Japan Atomic Energy Research Institute)“BWR Analyses Method of Reactivity Insertion Accident”(HLR-012R3, February 1999, Hitachi Co.)

6

Table 1

Timeline of Input Parameter Determination by JANTIFrom March 20, 2007: Start consideration of core analysis【Phase I (Preliminary analysis using 3 region core model)】- Sensitivity Analysis using assumed Inserted Reactivity, Control
Rod Withdrawal Speed, Reactivity Coefficient etc.
Parameter

Considered Value

Inserted Reactivity ($)

1.2, 1.4, 1.6

Control Rod Speed (mm/s)

10, 30, 60, 100

Doppler Coefficient (Δk/k/℃)-2×10-5

Coolant Temperature
Coefficient (Δk/k/℃)-1.0×10-4

Coolant Speed (cm/s)

10- Consideration of analysis condition of Hokuriku EPCO, analysis
condition of standard case was determined as follows except forcontrol rods withdrawal speed.
Parameter

Considered Value

Inserted Reactivity ($)

1.3

Control Rod Speed (mm/s)

Per mockup test

Doppler Coefficient (Δk/k/℃)-2×10-5

Coolant Temperature
Coefficient (Δk/k/℃)-4.0×10-5

Coolant Speed (cm/s)

10**Determined by sensitivity analysis (4cm/s and 10cm/s).
From March 24, 2007
【Phase II (Analysis of Partial Core based on 3-D Core Analysis)】- Setup of partial core with 5 horizontal & 10 axial regions.
- Sensitivity analysis of inserted reactivity and control rod speed.
① Inserted Reactivity ($): 0.81, 1.3, 1.8 etc.
② Control Rods Withdrawal Speed (mm/s):16, 47, 76 etc.
-

Additional analysis with zero coolant temperature coefficient.

7

Table 2

Comparison of Analyses Input and Result①Analyses Condition (Initial Condition)
JANTI

Hokuriku EPCO

Analyzed Core

Partial Core
(34 fuels)

Full Core(368 fuels)

Initial Power

0.7E-6%

1E-6%

Reactivity Excess
Reactivity
Feedback

0.0079Δk(Standard Condition)
Doppler Coefficient
Coolant Temperature
Coefficient

0.0079Δk (1)Doppler Coefficient②Analyses Result
JANTI(Standard
Condition)

Hokuriku
EPCO (1)Peak Power（Fraction to Rated Power）

14％

15％

－

Max. Enthalpy Increase during
Peak Power [cal/gUO2]

13

13

85 (2)

Max. Fuel Enthalpy [cal/gUO2]

49

41

230 (3)92 (3)

Threshold(1)

Among conditions considered by Hokuriku EPCO, reactivity excess
estimated by cold criticality test results was selected.

(2)

Threshold value for fuel PCMI failure per “Treatment of HighBurn-up Fuels in Reactivity Insertion Accident of Light Water
Reactor Generation Facilities”(3)

Limit value in accidents/abnormal transients during operation per“Reactivity Insertion Accident Review Guide for Light Water Reactor
Generation Facilities”

8Table 3-1

Comparison of Analysis Method

Analysis by JANTI

Analysis by Hokuriku EPCO

・ Calculation of both nuclear and ・ Calculation of Nuclear feedback
thermalhydraulics
feedback(Kinetics
Calculation(Doppler coefficient and coolant
considering Doppler Coefficient)
was calculated by APEX.
temperature
coefficient)by
EUREKA-2.
・ Input APEX result into thermal
hydraulics code SCAT to
・ Input value of pressure wascalculate fuel thermal power.
increased as calculation needed to
be done with no void condition.
・ Power distribution change isconsidered.
・ Accuracyofheat
removalcalculation is not comparable to
SCAT.
・ Power distribution is constant.

Table 3-2 Sensitivity Analysis of Affect of Coolant Temperature CoefficientStandard Condition

With “Zero” Coolant
Temperature
Coefficient

Net Inserted Reactivity
[$]

1.11

1.11

Peak Power（Fraction to Rated Power）

14％

15％

Max. Enthalpy Increaseduring Peak Power [cal/gUO2]

13

13(Reference) Max. Fuel
Enthalpy [cal/gUO2]

49

53

9

Table 4

Analyses Result (Affect of Reactivity Inserted to the Core)

Analyzed Case

Large

Standard

Small

Inserted Reactivity

1.8$

1.3$

0.81$

Net Inserted Reactivity
[$]

1.15

1.11

0.81

Peak Power（Fraction to Rated Power）

23％

14％

1％

Max. Enthalpy Increaseduring Peak Power[cal/gUO2]

15

13

－

(Reference) Max. Fuel
Enthalpy [cal/gUO2]

66

49

－

- In all cases, control rods withdrawal speed is 47mm/s。
- In all cases, maximum enthalpy increase is well below 85cal/gUO2,
threshold value of fuel PCMI failure during reactivity insertion accident.
- Larger reactivity being inserted, increase of net inserted reactivity issmall due to reactor core inherent feedback effect.
(Analysis Result by Hokuriku EPCO)- Control Rod Withdrawal Speed: 47mm/s- Peak Power: 15%- Max. Enthalpy Increase during Peak Power: 13cal/gUO2- Max. Fuel Enthalpy : 41cal/gUO2

10

Table 5

Analyses Result (Affect of Control Rod Withdrawal Speed)

Analyzed CaseHigh Speed

Standard

Low Speed

Control Rod Withdrawal
Speed

76mm/s

47mm/s

16mm/sNet Inserted Reactivity
[$]

1.17

1.11

0.98

Peak Power（Fraction to Rated Power）

28％

14％

3％

Max. Enthalpy Increaseduring Peak Power[cal/gUO2]

17

13

－

(Reference) Max. Fuel
Enthalpy [cal/gUO2]

50

49

－

- In all cases, inserted reactivity is 1.3$.
- In all cases, maximum enthalpy increase is well below 85cal/gUO2,
threshold value of fuel PCMI failure during reactivity insertion accident.
- Even if reactivity of more than 1 $ is inserted by control rods withdrawal,
net inserted reactivity can be below 1 $ based on withdrawal speed ofcontrol rods. (reactivity insertion speed)(Analysis Result by Hokuriku EPCO)- Control Rod Withdrawal Speed: 47mm/s- Peak Power: 15%- Max. Enthalpy Increase during Peak Power: 13cal/gUO2- Max. Fuel Enthalpy : 41cal/gUO2

11

A

Partial Core

Withdrawn
ControlB
Rods

C

Peaking

Withdrawn Control Rods (3)Inserted Control Rods (86)A

C
B

Shika 1 Full Core
Control
Control
Control
Rods
Rods
Rods
(Inserted) (Withdrawn) (Inserted)

Analyzed Core (Partial Core)・ Number of Fuels：34 of 368 (9%)
・ Height：10/24 from Top（40%）
・ Volume：4% of Full Core・ Heat Generatiion：70% of Full Core・ Peaking of Partial Core：3.5
Figure 1

Analyzed Partial Core12

Withdrawn Control Rods
A（26-39、16pos）
B（30-39、20pos）
C（34-35、08pos）

Withdrawn Control Rods (3)Inserted Control Rods (86)

00pos

A

C
B

48pos
.
Control
Control
Control
Rods
Rods
Rods
(Inserted) (Withdrawn) (Inserted)

Figure 2

1

2

Axial Power Distribution(Relative Value)

Power Distribution of the Core

13

3

4

Figure 3

Reactivity Insertion Speed

14

300

60

50

40

Power（MW）

200

Standard Case：47mm/s

30Standard Case：Fuel Enthalpy
100

20

10

0

0
0

5

10

15

20

25

30

Time (Second)

Figure 4-1

Trend of Power (Standard Case)

1.E+3

Standard Case 47mm/s

1.E+2
1.E+1

Power （MW）

/ /

1.E+0
1.E-1
1.E-2
1.E-3
1.E-4
1.E-5

/ /

0

20

40

60

80

100

120
860

Time (Second)

Figure 4-2

Trend of Power (Standard Case)

15

140
880

160
900

2.0

Reactivity Inserted by Control Rods Withdrawal

1.5

Reactivity ($)

1.0

Net Reactivity = Total Effect of Reactivity

0.5

Reactivity due to Moderator Temperature Increase

0.0
-0.5
-1.0
1.3$
-1.5

Reactivity due to Doppler Effect (due to Fuel Temperature Increase)

-2.0
0

5

10

15

20

25

30

Time (Second)

Figure 4-3

Effect of Reactivity Feedback (Standard Case)
1.3$

800
反応度大
基本 1.3$
反応度小Fuel Temperature(℃)

600

400

200

0
0

5

10

15

20

25

Time (Second)

Figure 4-4

Trend of Pellet Temperature (Standard Case)

16

30

140

at 10s
at 27sat 50s
at 200s

Coolant Temperature （℃ ）

120

100

Boiling Temperature Corresponding to Core Pressure

80

60

40

Initial Temperature

20(Core Outlet）

0
1

2

3

4

5

6

7

8

9

10

Axial Node of Partial Core

(Note) As higher core pressure value was used to avoid boiling, portion of
coolant temperature appears to be above boiling temperature.

Figure 4-5

Axial Distribution of Coolant Temperature in Partial Core
(Standard Case)

17

60

500

50

400

Power （MW）

(Control Rods Speed：47mm/s)

300

40

Large Reactivity：1.8$
Standard Case：1.3$

30

Small Reactivity：0.81$

200

20

Standard Case：Fuel Enthalpy
100

10

0

0
0

5

10

15

20

25

30

Time (Second)

Figure 5-1

Trend of Power (Affect of Reactivity Inserted to the Core)

1.E+3
1.E+2

Power （MW）

1.E+1

/ /
/ /
/ /

1.E+0

(Control Rods Speed：47mm/s)
1.E-1

Large Reactivity：1.8$
Standard Case：1.3$

1.E-2

Small Reactivity：0.81$

1.E-3
1.E-4
1.E-5

/ /

0

20

40

60

80

100

860
120

880
140

900
160

Time (Second)

Figure 5-2

Trend of Power (Affect of Reactivity Inserted to the Core)

18

500

60

50

400

Power （MW）

(Inserted Reactivity：1.3$)300

40High Speed：76mm/s
Standard Case：47mm/s

30

Low Speed：16mm/s

200Standard Case：Fuel Enthalpy

20

100

10

0

0
0

5

10

15

20

25

30

Time (Second)

Figure 6-1

Trend of Power (Affect of Control Rod Withdrawal Speed)

1.E+3
1.E+2
1.E+1

Power （MW）

/ /

1.E+0
1.E-1

(Inserted Reactivity：1.3$)

1.E-2

High Speed：76mm/s
Standard Case：47mm/s

1.E-3

Low Speed：16mm/s

1.E-4
1.E-5

/ /

0

20

40

60

80

100

120
860

880
140

900
160

Time (Second)

Figure 6-2

Trend of Power (Affect of Control Rod Withdrawal Speed)

19

Attachment 1
Summary of “Treatment of High Burn-up Fuels in Reactivity Insertion
Accident of Light Water Reactor Generation Facilities”
This document was approved by Nuclear Safety Commission of Japanon April 13, 1998. For reference, summary of this document wastranslated as follows;
Background
For the future safety review, consideration was made on treatment ofhigh burn-up fuels in reactivity insertion accident of light water reactors
based on detailed investigation result of domestic research outcome andoverseas examination results by safety standard sub-committee of nuclear
safety commission to finalize the conclusion in this report.
Threshold Value of Fuel Failure
The threshold value of fuel failure due to PCMI (Pellet-Cladding
Mechanical Interaction) is estimated as shown in the following table. The
threshold values are presented as maximum enthalpy increase during
peak power in conjunction with the associated pellet burn-up.
Maximum Enthalpy Increase
During Peak Power

Pellet Burn-up
Below

110cal/g･UO２

25,000MWd/t

Between 25,000MWd/t & 40,000MWd/t

85cal/g･UO２

Between 40,000MWd/t & 65,000MWd/t

50cal/g･UO２

Between 65,000MWd/t & 75,000MWd/t

40cal/g･UO２“Threshold Value Used in JANTI Analyses”
85cal/g･UO２ was used based on the report by Hokuriku EPCO.

20

Attachment 2
Summary of “Safety Analyses Review Guide for Light Water Reactor
Generation Facilities”
This review guide was approved by Nuclear Safety Commission ofJapan on August 30, 1990, and revised on March 29, 2001. For reference,summary of this review guide was translated as follows;
II Safety Analysis Review
1. Purpose of Safety Analysis Review
Appropriateness of fundamental principle for safety design of nuclear
facilities is reviewed per “Safety Design Review Guide.” “Safety Design
Review Guide” requires that structures, systems and components ofnuclear facilities should function as expected to maintain safety bothduring normal operation and during abnormal condition. Therefore,
review and analyses of “abnormal transients during operation” and“accidents” are needed to confirm appropriateness of fundamental
principle for safety design of nuclear facilities. This guide presents
events to be considered for safety design review, threshold of analyses
results, and conditions to be considered in analyses.
2. Scope of Analyses
2.1 Abnormal Transients during Operation
During reactor operation, events resulted from single failure/malfunction of equipment or single operator error that is expected during
operational life of nuclear facilities, or resulted from another contributors
to be expected in equivalent frequency are considered.
2.2 Accidents“Accidents” are the abnormal conditions that exceed “abnormaltransients during operation.” In spite of low frequency, events should be
regarded as “accidents” if there is potential of nuclear material releasefrom the facility, and consideration is needed in the standpoint of safety
review.

21

3. Selection of Events to be Reviewed
For both “abnormal transients during operation” and “accidents”,
events for safety review should be selected appropriately in accordancewith above mentioned purpose and scope of safety design review.

22Attachment 3
Summary of “Reactivity Insertion Accident Review Guide for Light Water
Reactor Generation Facilities”
This review guide was approved by Nuclear Safety Commission ofJapan on January 19, 1984, and revised on August 30, 1990. For
reference, summary of this review guide was translated as follows;
Definition of Terms
・ Reactivity Insertion Accidents: Events accompanied by increase ofreactor power and the associated increase of fuel enthalpy due torapid insertion of basically more than 1$ of reactivity into reactor
in or near criticality.
・ Fuel Enthalpy: Radial average enthalpy of pellet. Summation of
initial enthalpy and increased enthalpy obtained by analysis of theevent. Fuel enthalpy is base value at 0℃.
・ Definition of “peak power” period is shown in (Figure 1). “Po” isinitial power, and “Pp” is peak power. “th” is the period while
power is above (Po+Pp)/2. “tp” is the time of peak power. Peakpower period “te” is defined as tp+th (time period of slashed zone.)

Peak Power Zone
Power

Time Period after initiation of Event(Figure 1)

Definition of “peak power” period in reactivity insertion
accidents
23

Purpose
Analyze increase of reactor power and the associated increase of fuelenthalpy due to rapid insertion of basically more than 1$ of reactivity into
reactor in or near criticality in order to confirm integrity of core andreactor coolant pressure boundary during “abnormal transient during
operation” and “accident”.
Threshold(1) “Abnormal Transient during Operation”
1)
Maximum fuel enthalpy should be within “Fuel Design Limit”
shown in (Figure 2).
2)
Pressure at reactor coolant pressure boundary should bewithin 110% of Maximum Operational Pressure.
(2) “Accident”
1)
Maximum fuel enthalpy should be within 230cal/g・UO2.
2)

Pressure at reactor coolant pressure boundary should bewithin 120% of Maximum Operational Pressure.
(3) During “abnormal transient during operation” and “accident”, reactor
shutdown capability and integrity of reactor pressure vessel should
not be affected by disturbance such as pressure impact resulted fromrupture of fuel with water intrusion.

Fuel
Enthalpy

Pressure Difference at fuel rod surface (kg/cm2)(Figure 2)

Fuel design limit at reactivity insertion accident

24

Attachment 4
Definition of Terms

Terms

Criticality

Explanation

Status of generated neutron from fission anddisappeared neutron from the core is in balance,and chain reaction is maintained. Status when
Effective Criticality Factor (Keff) is 1.

Effective Criticality Number of generated neutron from fission divided
Factor (Keff)by number of neutron disappeared from the core.
Prompt Criticality

Status when criticality is maintained with nocontribution of delayed neutron.

Delayed Criticality

Status when criticality is maintained with
contribution of both prompt and delayed neutrons.

Prompt Neutron

Neutron emitted almost simultaneously (within
10-4 second) during fission.

Delayed Neutron

Neutron emitted from collapse of fission products
after 0.4 second to 50-60 seconds of the original
fission.

Delayed Neutron
Fraction

Fraction of Delayed Neutron from total number ofneutrons emitted from fission.

Reactivity

Value of (Keff-1)/Keff, indicator of deviation fromcriticality.

Reactivity Excess

Value of (Keff-1), reactivity of over criticality.

25

Inserted Reactivity

Reactivity inserted in the core.

Net Inserted
Reactivity

Inserted reactivity
reactivity.

Feedback
Reactivity

Reactivity such as Doppler Reactivity andModerator Temperature Reactivity which has
suppression effect on the inserted reactivity.

Reactivity
Coefficient

Coefficient of reactivity change due to change offuel temperature or moderator temperature.

Doppler(Reactivity)
Coefficient

Reactivity change per fuel temperature change.
When fuel temperature increases, reactivity tends
to decrease because of increased neutron absorption
rate isotopes such as U-238.

Moderator
Temperature(Reactivity)
Coefficient

Reactivity change per moderator temperaturechange.

Core Inherent

When reactor power increases, reactivity will

subtractedbyfeedback

Safety Feature (Self decrease due to Doppler effect and others, which
Control Feature)
will lead to decrease of reactor power.
Fuel Enthalpy

Amount of heat accumulated per weight of fuel.

Power Peaking

Maximum power divided by average power.

26