Azizul Khakim - 2009.11.12

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Majelis Nuklir TITech (MaNuk TIT) By: Azizul Khakim Tokyo, November 12, 2009

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History 1954 -1958: Panitia Negara untuk Penyelidikan Radioaktivitas dilatarbelakangi oleh adanya percobaan ledakan nuklir pada tahun 1950-an oleh beberapa negara terutama Amerika Serikat di beberapa kawasan Pasifik, sehingga menimbulkan kekhawatiran tentang jatuhnya zat radioaktif di wilayah Indonesia. Tugas dari panitia ini adlah untuk menyelidiki akibat percobaan ledakan nuklir, mengawasi penggunaan tenaga nuklir dan memberikan laporan tahunan kepada pemerintah. 1958 – 1964: Lembaga Tenaga Atom Tugasnya untuk melaksanakan riset di bidang tenaga nuklir dan mengawasi penggunaan tenaga nuklir di Indonesia. 1964 – 1997: Badan Tenaga Atom Nasional (BATAN) Tugas BATAN adalah untuk melaksanakan riset tenaga nuklir dan mengawasi penggunaan tenaga nuklir di Indonesia. Pengawasan penggunaan energi nuklir tersebut dilaksanakan oleh unit yang berada di bawah BATAN, yang terakhir pada Biro Pengawasan Tenaga Atom (BPTA). 1997 – Sekarang: Badan Tenaga Nuklir Nasional (BATAN) dan Badan Pengawas Tenaga Nuklir (BAPETEN) melalui UU No 10/1997 tentang Ketenaganukliran telah memberikan kewenangan bagi BAPETEN untuk melaksanakan fungsi pengawasan terhadap penggunaan tenaga nuklir, yang meliputi perizinan, inspeksi dan penegakan peraturan. UU Ketenaganukliran juga mensyaratkan pemisahan antara badan pengawas, BAPETEN, dan badan peneliti, BATAN.

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Tupoksi Tupoksi ‡ Pembuatan Peraturan ‡ Perizinan ‡ Inspeksi Kegiatan Penunjang Pengawasan ‡ Penegakan Peraturan ‡ Pengkajian Sistem Pengawasan ‡ Kesiapsiagaan Nuklir

Struktur organisasi BAPETEN www.bapeten.go.id

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Inspection

Assessment / Analyses Licensing

Regulation

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Thermal hydraulic Analyses of MTR Type Research Reactor By: Azizul Khakim

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Description of the MTR type RR: Fuel plate of U3Si2-Al ‡ 40 Fuel Element (21 plates) ‡ 8 Control element (15 plates + absorber) ‡ Reflector: Beryllium. ‡ 30 MW of Nominal power ‡ Downward forced convection of 800 kg/s. ‡

Safety criteria: Maximum fuel design temp.: 200°C ‡ Maximum clad design temp.: 145°C ‡ Min. Safety Margin against Flow Instability (S): 1.48 ‡

ηC S= ηE

ηE: experimental Bubble Detachment Parameter of 22.1 cm3K/Ws.

[ Ts ( z ) − Tc ( z )]V ( z ) η ( z) = q" ( z )

where: q”: Heat flux, w/cm2 V : Coolant velocity, cm/s z : distance from coolant inlet channel, cm Ts, Tc: Saturated temp. and coolant bulk temp., K

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Conditions to be analyzed: RIA at Power Range of 1 MW ‡ RIA at Natural Circulation of 0.3 MW ‡ LOFA ‡

The code: PARET/ANL code

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RIA at Power Range Initial power: 1 MW ‡ Initiation: inadvertent CRs withdrawalÆ fast (+) reactivity into the core ‡ Single failureÆ 1st trip signal (Floating Limit Value) fails to scram ‡ 2nd trip signal (Over Power) scrams the Rx. ‡ Delay time from trip signal to CR Drop: 0.5s ‡ Downward forced normal cooling ‡ Transient starts at t=5 s ‡

50

40

45

35

40 35

30

30

25

25

20

S

POWER, MW

10 45

20

15

15

10

10

5

5

0

0 0

5

10

15

20

25

TIME, S Pow er

S

200

TEMPERATURE,

150 TFuel 100

TClad TCoolant

50

0 0

5

10

15 TIME, S

20

25

11

Result of RIA at Power Range Steady state condition (the first 5 s):

Power, MW Max. fuel temp., °C Max. coolant temp. in hot channel, °C T (s)

Transient condition:

19.3 23.4 23.9 23.9 23.9 23.9 23.94

The 1st trip signal (FLV), % pw chg The 2nd trip signal (over power), % Peak power, MW Max. fuel temp., °C Max. clad temp., °C Min. S Max. coolant temp. in hot channel, °C

Value

1 50.2 45.9 7.0 114 40.26 185.3 138.0 2.25 98.95

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RIA at Natural Convection Initial power: 0.3 MW ‡ Initiation: inadvertent CRs withdrawalÆ fast (+) reactivity into the core ‡ Trip signal: period of 5 s. ‡ Natural circulation cooling ‡ Transient starts at t=5 s ‡

30

1.1 1

25 20

0.9

15

0.8 0.7

10 5

0.6

0

0.5 0.4

-5 -10

0.3

-15

0.2 0.1

-20 -25

0

Period, S

Power, MW

13 1.2

-30 0

2

4

6

8

10

12

14

16

18

20

Time, S Pow er

Period

240

1.2 1.1

0.9 160

Power (MW)

0.8 0.7

120

0.6 0.5

80

0.4

MFR (Kg/s.m2), Temp. (°C

200

1

0.3 40

0.2 0.1 0

0 0

5

10

15

20

Time, s Pow er

Flow Rate

TCoolant

TClad

TFuel

Result of RIA at Natural Convection Steady state condition (the first 5 s): Power, MW Max. fuel temp., °C Max. coolant temp. in hot channel, °C Coolant flow rate, kg/s.m2 T (s) Transient condition: 17.0 Trip signal (period), s

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value 0.3 72.2 59.0 126.7 5

17.5 Peak power, MW

1.136

17.5 Max. fuel temp., °C

113.1

17.5 Max. clad temp., °C

112.8

18.0 Max. coolant temp. in hot channel, °C 18.2 Max. coolant flow rate, kg/s.m2

69.8 212.0

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LOFA Initial power: 30 MW ‡ Initiation: all primary pumps simultaneously off ‡ Trip signal: low flow trip signal ‡ Transient starts at t=5 s ‡

16

100

140 1

130

2

80

120

70

110

60

100

50

90

40

80

30

70

20

60

10

50

0

40

-10

0

10

20

30

40

50

60

70

Time (s) Pow er

MFR

T Fuel

TCoolant

80

90

100

30

Temperature (C)

Power (%), MFR (%)

90

Result of LOFA

17

Steady state condition (the first 5 s):

Value:

Max. fuel temp., °C

125.8

Max. coolant temp. in hot channel, °C Min. S

64.4 7.8

T (s)

1st critical point:

7.9

Trip signal (Low Flow), %

8.0

Max. fuel Temp., °C

8.2

Min. S

8.4

Reactor trip

8.43

Max. coolant temp. in hot channel, °C

85 136.4 3.6 87.0

2nd critical point:

82.1

Flow reversal (stagnant flow)

0

86.1

Max. fuel Temp., °C

115.4

86.3

Max. coolant temp. in hot channel, °C

107.0

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Conclusion Thermal hydraulic safety analyses for MTR type RR have been conducted for major DBA. ‡ No safety criterion is exceeded for major DBA. ‡

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Neutronic Calculation of MTR Type Research Reactor with MCNP By: Azizul Khakim

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Codes and Background Codes: ‡ MCNP-4b (Monte Carlo N-Particles): 3-D core calculation, with ENDF/B-VI & B-V ‡ ORIGEN2: FP inventory & burn up calculation Background: ‡ To support the decision making during the licensing process of fuel replacement from U3O8-Al to U3Si2-Al

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Features of MCNP ‡ ‡ ‡ ‡ ‡

Generalized-geometry Time-dependent Couple n/p/e Monte Carlo Transport code Continuous energy; n:10-11 – 20 MeV; p/e: 10-3 – 1000 MeV By simulating individual particle instead of solving transport equation, as deterministic method does

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Description of MTR Pool type Reactor ‡ Fuel plate of U3Si2-Al ‡ 40 Fuel Element (21 plates) ‡ 8 Control element (15 plates + absorber) ‡ Reflector: Beryllium ‡ Moderator: H2O ‡ Enrichment: 19.75% ‡ Cladding material: AlMg2 ‡ Absorber: AgInCd ‡ Nominal power: 30 MW ‡

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Description of MTR (cont’d) FE

CE

core

Reactor

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Verification of MCNP input with exp’tal data The experimental data of the 1st core and 1st criticality are used to verify the MCNP input. ‡ The 1st criticality is achieved when the core is composed of 9 U3O8-Al FEs, 6 CEs when RR at 475 mm ‡

‡

The 1st core is composed of 12 U3O8-Al FEs

and 6 CEs

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Verification of MCNP input with exp’tal data of 1st core and 1st criticality 3-D Diffusion Calculations

Core Configuration

First Criticality

Full Core (CRs all up)

Exp’ment Batan-3 Diff Data &WIMSD4

MCNP & Citation-3D ENDF/B-VI &WIMSD4

0.99816

0.99172

1.00238 ± 0.002

C/E

0.998

0.992

1.00238

Keff

1.08466

1.08179

1.09714 ± 0.0002

0.993

0.99

1.001

0.92508

0.96987

0.91875 ± 0.0013

Keff

1.0

1.09242

C/E

Full Core Keff (CRs all down)

-

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Verification of MCNP input with exp’tal data of CRs Calibration Cal’d CR’s post (cal’d CR’s level / other CRs’) Keff C-8 C/E (0 mm / 290 mm) Keff E-9 C/E (0 mm / 284 mm) Keff F-8 C/E (0 mm / 293 mm) Keff C-5 C/E (0 mm / 288 mm) Keff F-5 C/E (0 mm / 290 mm) Keff D-4 C/E (0 mm / 282 mm)

Exp’ment Data 1.00008 1.00008 1.00008 1.00008 1.00008 1.00008

Calculation with MCNP 1.00291 ± 0.00199 1.0028 1.00065 ± 0.00127 1.00057 0.9998 ± 0.00148 0.9997 1.00329 ± 0.00125 1.0032 1.00102 ± 0.00158 1.0009 1.00169 ± 0.00121 1.0016

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Typical working core (TWC) analyses Burn up distributions at BOC & EOC as proposed by the Operating Organization are used in TWC analyses. ‡ Refueling every 615 MWD with 5 FEs/1 CE. ‡ Burn up limit 56%. ‡ Max. radial power peaking factor: 2.6 (OLC) ‡ Max. axial power peaking factor: 1.6 (OLC) ‡

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Result of TWC calculation No

TWC Condition

MCNP, ρ(%) Diff. Code, ρ(%)

1

BOC, equilibrium Xe

6.25

2

BOC, Cold, w/o Xe (Max excess ρ)

9.43

9.7

3

Xe Reactivity

-3.18

-3.7

4

EOC, w/o Xe (Fully Up)

6.89

5

Reactivity change in one cycle

-2.54

6

EOC, cold, w/o Xe (fully down)

-5.26

7

Control rods reactivity

-12.15

-13.8

8

Shutdown margin

-2.72

-4.1

9

Void reactivity coeff., (∆k/k)/%void

–1.29×10-3

-1.34×10-3

10

Max. radial power peaking factor

1.26

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Axial power peaking factor

*all CEs fully up. **all CEs 50% withdrawn

1.35*;1.61**

-2.5

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having yet to be calculated: Fuel temp. coefficientÆ need other code to generate nuclear data at higher than 300K (e.g.: NJOY, PREPRO) ‡ OSR (One Stuck Rod) criteria (<-0.5 ÅOLC) ‡ Various axial CEs combinations have yet to be analyzed to determine max. axial PPF. ‡

Uncertainty Accuracy of ORIGEN2 code. ‡ Familiarity with data & system (data available in SAR is somewhat inadequate) ‡

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Conclusion Decision making during licensing process of Rx modification should be supported with independent analyses by RB. ‡ For comprehensive neutronic calculation, MCNP should be supported by other codes. ‡

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9. 液体金属冷却高速増殖炉におけるボイド反応度に関する研究

9. Study on Void Reactivity of Liquid Metal Fast Breeder Reactor

By: Azizul Khakim

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Background & Purpose ‡

Background: „

The future FBR fuel should include Minor Actinides (MAs) because they are contained in LWR discharged fuel and burden the environments.

„

Problem: MA increases the sodium void reactivity (safety concern)

‡

Purpose: to observe the parameters and design characteristic that induce both (-) & (+) void reactivity effects to be taken into account during the reactor core design to achieve reasonably low positive or even negative void reactivity.

‡

Calculation: 3-D continuous energy Monte Carlo method of MVP Code with JENDL-3.3. The Reactor core is modeled in heterogeneous 3-D geometry.

Average energy loss : Σ ΔE =

Phenomena during voiding

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Σ el ΔE el + Σ inel ΔE inel E

where ⎛ A−1⎞ α=⎜ ⎟ ⎝ A+1⎠

ΔE el = 12 (1 − α ) E n ;

2

and 2

‡ ‡ ‡

Leak = ∫ J .n dS

0.5 Normalized Flux

Increase in neutron leakage (-) Spectrum hardeningÆη⇑ Æ(+) Reduction of Na capture (+) Change in self shielding

‡

A + 1⎤ ⎛ A ⎞ ⎡ ΔE inel = E n − ⎜ ⎟ ⎢ En − Q A ⎥⎦ ⎝ A+1⎠ ⎣ Q : excited energy level

S

J = − D∇φ D=

1 3Σ s (1 − 2

3A Σ s = Σ Cs + Σ fs + Σ ss

)

0.4 0.3 0.2 0.1 0.0 1.E-04

1.E+02

Flooded

1.E+04

1.E+06

1.E+08

Voided

5 4.5

1.E+01

4

1.E+00

3.5

η= no. neutron released per neutron absorbed

η(E) = ν

3

1.E-01

η(E)

σc (barn)

1.E+00

Energy (eV)

Na Capture XS

1.E-02

σ f (E) σ f (E) + σ c (E)

2.5 2

1.E-03

1.5

1.E-04

1

1.E-05 1.E-06 1.E-04

1.E-02

0.5

1.E-02

1.E+00

1.E+02

1.E+04

1.E+06

1.E+08

0 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 1.E+08 Energy (eV)

E (eV)

U-235

U-238

Pu-239

Pu-240

Am-241

Pu-241

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VOID MODELING ‡

‡

Assumption: generated by excessive fuel heating under accident conditions. Void Location: „

axially active fuels and above

„

inner and outer core

„

Inside channel box

„

does not occur in control assembly positions

‡

Void fraction: homogeneously 100%

‡

Liquid fraction above the core =Total Liquid flow area/total area

b

a

c

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Base case core configuration ‡

Homogeneous core configuration

‡

Radial peaking factor: 1.26

‡

Sodium void reactivity: 1.600 %∆k/k'

Electrical power, MW Thermal power, MW Ave./Max linear power, kW/m CORE PARAMETERS In/outer Core height, m Pu Fissile In/out enrichment Fuel/Sodium/Structure, % FUEL ASSEMBLY In/Out Driver Assembly Fuel Type Bond material Pin diameter, mm Clad & Duct material Pin Pitch, mm T/B F.G Plenum, m No. Pins per Assembly Duct Flat-to-Flat, mm Duct Thickness, mm Duct Pitch, mm

1200 3000 28 / 48 1.0/1.0 9.5%/11.5% 37/34/29 150 / 216 (U0.8Pu0.2)O2 He 8.5 SS 9.8 0.15 / 0.85 271 173 3 179

BLANKET Material

UO2

Pin Dia., mm

8.5

Top/Bot. length, m

Normalized neutron flux distribution

0.3/0.3

No. Rad. blanket Ass.

150

Control Material

B4C

No. Control Assembly Shield

31 B4C

36

Effect of MAs inclusion ‡

MAs build up with burn-up

‡

MAs inclusion: 4.71% of Pu

‡

Pu/Np237/Am241/Am243/Cm244: 95.0/0.5/2.0/1.0/1.0

‡

η of MAs up as spectrum hardens

‡

Sodium void reactivity: 1.689 %∆k/k’

37

Parametric study (1/3) Case 1: Heterogeneous core config. ‡

by interchanging the same no. of FAs in the inner region with the blanket assemblies

‡

The number FAs & blanket assemblies are the same as those in the homogeneous config.

‡

Radial peaking factor: 1.26 Æ 1.71

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Parametric study (2/3) Case 2: Step core ‡

2 cases of 70 cm & 60 cm inner core height are calculated

‡

Radial peaking factor: 1.26 Æ 1.34 (70 cm inner height)

Case 3: Elimination of up. gas plenum ‡

70 cm-inner step core is used

‡

The upper plenum is eliminated

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Parametric study (3/3) Case 4: shorter upper blanket (15 cm) ‡

70 cm-inner step core is used

‡

Upper blanket: 30 cm Æ 15 cm

Case 5: Reduction of radial blanket ‡

70 cm-inner step core is used

‡

Radial blanket: 2 Æ 1

Result summary for parametric study Case Ref MA 1 2.a 2.b 3 4 5

Flooded Condition Parameter Keff Std. Dev Base case core 1.07003 0.0117% MAs inclusion of 4.71% of Pu 1.06314 0.0124% Heterogeneous configuration 1.02036 0.0127% Step core: a. 70 cm in. core H 1.03549 0.0121% b. 60 cm in. core H 1.02077 0.0135% Elimination of up. G. plenum 1.03549 0.0124% Shorter up. blanket (15 cm) 1.03745 0.0129% Radial blanket reduction 1.03349 0.0133%

Voided Condition Keff Std. Dev 1.08867 0.0122% 1.08258 0.0122% 1.02907 0.0127% 1.04862 0.0126% 1.03221 0.0140% 1.04736 0.0134% 1.04865 0.0124% 1.04619 0.0129%

Void reactivity (%∆k/k') 1.600 1.689 0.830 1.209 1.086 1.094 1.029 1.175

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Effect on Void Reactivity (%∆k/k') +0.089 -0.771 -0.391 -0.514 -0.115* -0.180* -0.035*

* Relative to case 2.a 5 4

5

most effective way to reduce

4

void reactivity without

2.b

significantly degrading neutron

1

3

3 Case

Æ Step core is concluded to be the

2.b

2.a

2.a

economy and other safety characteristic

1 -0.8

-0.7

-0.6

-0.5

-0.4

-0.3

-0.2

Effect on void reactivity (%∆k/k')

-0.1

0.0

0.1

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Core characteristics (1/3) 70 cm-inner step core is used 1. Voiding in the inner & outer core ‡

Voiding takes place in the inner or outer region only; the rest remains flooded

‡

Void reactivity: 0.62%∆k/k'(0.0041 %∆k/k'/FA) and 0.63 %∆k/k' (0.0029 %∆k/k'/FA) in the inner & outer core, respectively. 1.4

2. Void reactivity profile ‡

Æ linear Void reactivity coefficient =2.18x10-2 %∆k/k/%void.

1 ρ (%∆k/k)

‡

1.2

0.8 0.6 0.4 0.2 0 0

10

20

30

40

50

60

Void Fraction (%)

70

80

90

100

42

Core characteristics(2/3)

0.6 0.4 0.2

1.E-02

1.E+00

1.E+02

1.E+04

1.E+06

1.E+08

E (eV)

P/D up Æ ρex down and ρvoid up

1.5 1.4

P/D=1.2

550 540 530 520 510 500 490 480 470 460 450 440

Base case

W

1.3 1.2 1.1 1 0.9

4. Void reactivity at EOC Refueling batch In/out enrichment, wt% Pu Fissile/HM

1

4

Keff

366 (1 yr)

1.1

1.15

1.2

1.25

1.3

D

Breeding ratio

1.33

Burn up reactivity swing, %∆k/k

1.24 EOC

Ave. burn up, GWD/T.HM

19.7

32.8

Reactivity at EOC, % ∆k/k

2.82

1.58

Void reactivity, % ∆k/k’

1.63

1.82

Void reactivity

Core Witdh (cm)

P

4.5E+04

10. / 12.

BOC

1.05

P/D

4.0E+04 3.5E+04 3.0E+04 MA Mass (g)

Refueling interval, days

0.8

Core width (cm)

P/D up Æ FA size up Æ core size (W) up Æ coolant vol. fraction up Æ softer neutron spectrum

P/D=1.05

Keff, ∆ρ Void (%∆k/k')

‡

0.8

0 1.E-04

3. The effect of pin pitch ‡

Normalized flux

1

AM241

2.5E+04

NP237 AM243

2.0E+04

CM242

1.5E+04 1.0E+04 5.0E+03 0.0E+00 0.00

0.25

0.50 Year

0.75

1.00

43

‡

Voiding Æ energy shift Æ change in fuel XSÆ η(E)

‡

Keff decreases faster in flooded core than voided one Æ(∆keff/∆T)flood<(∆keff/∆T)void

‡

Approximated by K=Ta + b; where: „

a=-0.012395; b=0.1151568 (flooded)

„

a=-0.010577; b=0.117029 (voided)

Keff

5. Effect of voiding on Doppler reactivity

1.06

0.0E+00

1.055

-2.0E-06

1.05

-4.0E-06

1.045

-6.0E-06

1.04

-8.0E-06

1.035

-1.0E-05

1.03

-1.2E-05

1.025

-1.4E-05

1.02

-1.6E-05

1.015 0

500

1000

1500

2000

2500

3000

-1.8E-05 3500

Fuel Temperature (K) Keff Flooded

Keff Voided

(∆Keff/∆T)flood

(∆Keff/∆T)void

∆Keff/∆T

Core characteristics(3/3)

44

Conclusion ‡

The void reactivity increases by 0.19 %∆k/k‘ from BOC to EOC due to MAs build-up. Therefore, 70 cm-inner step core, providing -0.39 %∆k/k‘, can be employed to compensate MAs build-up.

‡

Upper blanket slash could be another way to compensate high void reactivity due to MAs build-up.

‡

Heterogeneous core configuration significantly brings negative effect on void reactivity, but it also degrades excess reactivity and exceeds allowable radial peaking factor.

‡

Voiding phenomenon changes the Doppler reactivity pattern. The Doppler reactivity under voided condition is less than under the normal condition.