Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor

Rahman, Mohammad Mizanur; Akond, Mohammad Abdur R.; Basher, Mohammad Khairul; Huda, M.Q.

doi:10.4236/wjnst.2014.42013

Cited by 16 publications

(3 citation statements)

References 3 publications

Supporting

Mentioning

Contrasting

Unclassified

Order By: Relevance

“…The COOLOD-N2 computer code [2] is used for analysing thermal hydraulics of the steady-state research reactors operation. Both rod-type and plate-type fuel elements can be used in the calculation model [4]. The maximum fuel temperature is calculated based on one-dimensional heat conduction with heat production inside the fuel element.…”

Section: Thermal Hydraulic Core Model For Coolod-n2mentioning

confidence: 99%

Optimization for fuel loading and reactivity initiated accident analysis for TRR-1/M1

Tiyapun¹,

Wetchagarun²

2023

J. Phys.: Conf. Ser.

View full text Add to dashboard Cite

TRR-1/M1 reactor core consists of two types of standard TRIGA fuel, 8.5% and 20% weight uranium. Periodically, the reactor core configuration is rearranged in order to optimize its performance. The objective of this work is to show that it is safe to operate TRR-1/M1 at 1 MW steady state condition for the new core configurations of which the 8.5% weight fuels in the B ring are replaced by high burn up 20% weight fuel elements. To ensure sufficient safety margins of the reactor operation, thermal–hydraulic computer codes, i.e., COOLOD-N2 and EUREKA-2/RR, were coupled with neutronic calculation based on Monte Carlo N-Particle Transport concept in order to analyse some crucial operating parameters, e.g., coolant and fuel temperature in the reactor core as well as Departure from Nucleate Boiling Ratio (DNBR) along the fuel axial direction of the hottest channel. From this study, it can be concluded that no significant effect on the reactor operation was found when replacing the 8.5% weight fuel elements in the B ring with the high burnup 20% weight fuels. Transient phenomena of Reactivity Initiated Accident (RIA) were also performed for the rapid removal of a sample from the reactor core with reactivity insertion rate of $1.00/s. It is found that the negative temperature coefficient of TRIGA fuel provided safety for the reactor even during the prescribed reactor transients. The calculation showed slightly increase in the fuel temperature during the short period of reactivity insertion which caused insignificant energy release to heat up the fuel during the accident. The maximum fuel temperature found during the accident in this study was well below TRR-1/M1 safety.

show abstract

Section: Thermal Hydraulic Core Model For Coolod-n2mentioning

confidence: 99%

Optimization for fuel loading and reactivity initiated accident analysis for TRR-1/M1

Tiyapun¹,

Wetchagarun²

2023

J. Phys.: Conf. Ser.

View full text Add to dashboard Cite

show abstract

“…Code ini telah dibandingkan dengan eksperimen SPERT-I dan SPERT-II untuk sistem air ringan dan air berat [3]. Di samping itu PARET/ANL juga dibandingkan dengan RELAP5/MOD3 terhadap 10 MW IAEA Research Reactor [4] untuk empat kasus transien, yaitu transien LOF (Loss of Flow) cepat, transien LOF lambat, transien penyisipan reaktivitas lambat dan transien penyisipan reaktivitas cepat [4,14]. Secara keseluruhan kesesuaian antara PARET/ANL dan RELAP Code untuk serangkaian benchmark tersebut sangat baik, dan hasilnya sesuai dengan hasil guidebook.…”

Section: Pendahuluanunclassified

Analisis Keselamatan Termohidrolik Bulk Shielding Reaktor Kartini

Khakim¹

2015

Tri Dasa Mega

View full text Add to dashboard Cite

ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK) yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET. ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs) important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the spent fuels cooling with natural convection mode. The calculations results concluded that natural convection cooling mode can adequately cools down the decay heat without significant increase in fuel temperatur. Keywords: bulk shielding, spent fuels, natural convection, PARET.

show abstract

“…Several studies have been performed on nuclear reactor to ensure adequate safety, temperature control with an integrated safety system (Gharib et al, 2011;Hossain et al, 2019;IAEA, 2002;Sunday et al, 2013;Vojackova et al, 2017;Nain et al, 2019). The literature shows that the thermal-hydraulic models through hot channel fuel centreline temperature play a significant role to safety-related parameters within the design limit (Rahman et al, 2014). However, the investigation was performed using computer code and the data were far to compromise the safety of the reactor.…”

Section: Introductionmentioning

confidence: 99%