Influence of PT-CT contact on PHWR fuel channel thermal behaviour under accident condition - An experimental study

被引：13

作者：

Ajay, Ketan ^{[1
]}

Kumar, Ravi ^{[1
]}

Mukhopadhyay, Deb ^{[2
]}

Gokhle, Onkar ^{[2
]}

Gupta, Akhilesh ^{[1
]}

Das, Arup K. ^{[1
]}

机构：

[1] Indian Inst Technol, Dept Mech & Ind Engn, Roorkee 247667, Uttar Pradesh, India

[2] Bhabha Atom Res Ctr, Reactor Safety Div, Mumbai 400085, Maharashtra, India

来源：

NUCLEAR ENGINEERING AND DESIGN | 2020年 / 361卷

关键词：

PHWR; LOCA; ECCS; PT; CT; Moderator; Sagging deformation; COOLANT CHANNEL; BREAK; HEAT;

D O I：

10.1016/j.nucengdes.2020.110543

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

An experimental investigation has been performed to simulate a scenario of a beyond design basis accident like LOCA along with the un-availability of ECCS. The investigation is aimed to study the thermal behaviour of a single channel of PHWR during sagging deformation. The sagging deformation of PT was simulated by making a direct physical contact between PT and CT. The fuel channel assembly of a length of 1.4 m was immersed in water which simulates Moderator. The decay heat liberated after reactor shutdown was simulated by Joule heating of the 37-fuel pin elements. The steady state circumferential temperature distribution in the 37-fuel pin bundle simulator, PT and CT at different axial position has been obtained. It was found that there is a significant circumferential temperature gradient in the fuel channel. The maximum and minimum temperature was obtained respectively at top and bottom nodes of PT. However, an opposite trend was observed in CT. The maximum temperature difference of 185.7 degrees C and 10.2 degrees C has been found between top and bottom surfaces of PT and CT respectively. A similar behaviour to that of PT was discerned in the 37-pin fuel element. However, fuel pin of ring-1 has an insignificant effect of PT-CT contact. It was observed that around 84 percent of simulated decay heat is transferred to the simulated moderator.

引用

页数：10

共 12 条

[1] Experimental investigation of radiation heat transfer in coolant channel under impaired cooling scenario for Indian PHWR [J].

Ajay, Ketan ;

Kumar, Ravi ;

Mukhopadhyay, Deb ;

Gupta, Akhilesh ;

Das, Arup K. ;

Gokhale, Onkar .

NUCLEAR ENGINEERING AND DESIGN, 2019, 347 :45-52

[2]

[Anonymous], 1953, MECH ENG

[3] The Indian PHWR [J].

Bajaj, SS ;

Gore, AR .

NUCLEAR ENGINEERING AND DESIGN, 2006, 236 (7-8) :701-722

[4] The future 700 MWe pressurized heavy water reactor [J].

Bhardwaj, SA .

NUCLEAR ENGINEERING AND DESIGN, 2006, 236 (7-8) :861-871

[5] CORRELATING EQUATIONS FOR LAMINAR AND TURBULENT FREE CONVECTION FROM A HORIZONTAL CYLINDER [J].

CHURCHILL, SW ;

CHU, HHS .

INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER, 1975, 18 (09) :1049-1053

[6]

Gupta S.K., 1997, A Study of the Indian PHWR Reactor Channel Under Prolonged Deteriorated Flow Conditions, P331

[7] Thermal analysis of severe channel damage caused by a stagnation channel break in a PHWR [J].

Mukhopadhyay, D ;

Majumdar, P ;

Behera, G ;

Gupta, SK ;

Raj, VV .

JOURNAL OF PRESSURE VESSEL TECHNOLOGY-TRANSACTIONS OF THE ASME, 2002, 124 (02) :161-167

[8] Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition [J].

Nandan, Gopal ;

Sahoo, P. K. ;

Kumar, Ravi ;

Chatterjee, B. ;

Mukhopadhyay, D. ;

Lele, H. G. .

NUCLEAR ENGINEERING AND DESIGN, 2010, 240 (10) :3504-3512

[9] A HIGH-TEMPERATURE CREEP MODEL FOR ZR-2.5 WT-PERCENT NB PRESSURE TUBES [J].

SHEWFELT, RSW ;

LYALL, LW ;

GODIN, DP .

JOURNAL OF NUCLEAR MATERIALS, 1984, 125 (02) :228-235

[10] ON THE THERMAL-ANALYSIS OF PRESSURE TUBE CALANDRIA TUBE CONTACT IN CANDU REACTORS [J].

SHOUKRI, M ;

CHAN, AMC .

NUCLEAR ENGINEERING AND DESIGN, 1987, 104 (02) :197-206

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