0
Special Section Papers

Development of Technologies and Safety Systems for Pressurized Heavy Water Reactors in India

[+] Author and Article Information
S. Banerjee

Mem. ASME
Bhabha Atomic Research Center,
Mumbai 400085, India
e-mail: sbanerjee@barc.gov.in

H. P. Gupta

Bhabha Atomic Research Center,
Mumbai 400085, India
e-mail: hpgupta@barc.gov.in

1Corresponding author.

Manuscript received July 18, 2016; final manuscript received December 7, 2016; published online March 1, 2017. Assoc. Editor: Thambiayah Nitheanandan.

ASME J of Nuclear Rad Sci 3(2), 020902 (Mar 01, 2017) (13 pages) Paper No: NERS-16-1072; doi: 10.1115/1.4035435 History: Received July 18, 2016; Revised December 07, 2016

The technology of pressurized heavy water reactors (PHWRs) which was developed with prime objectives of using natural uranium fuel, implementing on power fuelling, utilizing mined uranium most effectively, and achieving excellent neutron economy has demonstrated impressive performance in terms of high capacity factors and an impeccable safety record. The safety features and several technology advancements evolved over the years in which Indian contributions that are considerable are briefly discussed in the first part of the paper. Unique features of PHWR such as flexibility of fuel management, distribution of pressure boundaries in multiple pressure tubes (PTs), and a large inventory of coolant-moderator heat sink in close proximity of the core provide inherent safety and fuelling options to these reactors. PHWRs, in India have demonstrated to have the advantage of lower capital cost per megawatt even in small size reactors. Low burn up associated with natural uranium fuel, higher level of tritium in the heavy water coolant, and a slightly positive coolant void coefficient in present generation PHWRs have all been addressed in the design of advanced heavy water reactor (AHWR). The merit of adopting closed fuel cycle with partitioning of minor actinides in reducing the burden of radio-toxicity of nuclear waste and of deploying light water reactors (LWRs) in tandem with PHWRs in the evolving nuclear fuel cycle in India are also discussed.

Copyright © 2017 by ASME
Your Session has timed out. Please sign back in to continue.

References

Bhabha, H. J. , and Prasad, N. B. , 1958, “ A Study of the Contribution of Atomic Energy to a Power Programme in India,” 2nd UN International Conference on ‘Peaceful Uses of Atomic Energy,’ Geneva, Switzerland, Sept. 1–13, pp. 89–101.
Ramanna, R. , 1987, “ Indian Nuclear Programme: Achievements and Prospects,” J. Korean Nucl. Soc., 19(8), pp. 213–219.
Rastogi, B. P. , 1989, “ Reactor Physics Computer Code Development for Neutronic Design, Fuel Management, Reactor Operation, and Safety Analysis of PHWRs,” BARC Report No. BARC-1442.
Krishnani, P. D. , and Srinivasan, K. R. , 1981, “ A Method for Solving Integral Transport Equation for PHWR Cluster Geometry,” Nucl. Sci. Eng., 78(1), pp. 97–103.
Jain, V. K. , and Gupta, H. P. , 1986, “ Analysis of Super Delayed Critical Transients in Thermal Reactors Using 2 and 3-D Adiabatic and IQS Methods,” Ann. Nucl. Energy, 13(3), pp. 115–125. [CrossRef]
Fernando, M. P. S. , 2012, “ Development of 3-D Space Time Kinetics Model for Coupled Neutron Kinetics and Thermal Hydraulics,” IAEA Workshop on Advanced Code Suite for Design, Safety Analysis, and Operation of Heavy Water Reactors, Ottawa, Canada, Oct. 2–5.
Askew, J. R. , Fayers, F. J. , and Kemshell, P. B. , 1966, “ A General Description of the Lattice Code WIMS,” J. Br. Nucl. Energy Soc., 5(4), pp. 564–585.
Nuclear Data Services, 2014, “ WIMS Nuclear Data Libraries,” IAEA, Vienna, Austria, accessed Dec. 3, 2016, http://www-nds.iaea.org/wimsd
Srinivasan, K. R. , 1996, “ Reactor Physics Methods for Design and Analysis of Heavy Water Moderated Reactors,” National Conference on Radiation Shielding and Protection, IGCAR, Kalpakkam, India, June 26–28, pp. 93–98.
Banerjee, S. , and Mukhopadhaya, P. , 2007, “ Phase Transformations: Examples From Titanium and Zirconium Alloys,” Pergamon Materials Series, Elsevier, Amsterdam, The Netherlands.
Singh, R. N. , Kishore, R. , Sinha, T. K. , and Banerjee, S. , 2000, “ Tensile Properties of Zr-2.5 Nb Pressure Tube Alloy Between 25 and 8000 C,” BARC Report No. BARC/2000/E/029.
De, P. K. , John, J. T. , Raman, V. V. , and Banerjee, S. , 1993, “ Stress Distribution and Hydride Orientation in Zr 2.5 Nb 0.5 Cu Garter Spring Under Complex Loading,” J. Nucl. Mater., 203(2), pp. 94–111. [CrossRef]
Bajaj, S. S. , and Gore, A. R. , 2006, “ The Indian PHWR,” Nucl. Eng. Des., 236(7–8), pp. 701–722. [CrossRef]
Soni, K. L. , Arpana , Mohan, L. R. , Nema, M. K. , and Mahajan, S. C. , 1997, “ Liquid Poison Injection System (LPIS) for Kaiga 1&2, & RAPP 3&4, 220 MWel PHWRs,” Workshop on Reactor Shutdown System, IGCAR, Kalpakkam, India, Mar. 4–6, pp. IV.3.1–IV.3.10.
Bhardwaj, S. A. , 2006, “ The Future 700 MWe Pressurized Heavy Water Reactors,” Nucl. Eng. Des., 236(7–8), pp. 861–871. [CrossRef]
Chatterjee, S. K. , Srinivasan, G. R. , Das, M. , Prakash, P. , and Mulgund, S. , 1994, “ Containment Design of Indian PHWRs-Evolution and Future Trends,” 3rd International Conference on Containment Design and Operation, Toronto, ON, Canada, Vol. 1, Oct. 19–21.
Muktibodh, U. C. , 2011, “ Advanced Nuclear Reactor Technology for Near-Term Deployment,” International Workshop, Vienna, Austria, July 4–8.
Kakodkar, A. , Kushwaha, H. S. , and Dutta, B. K. , 1989, “ Structural Evolution of Containment,” Nucl. Eng. Des., 117(1), pp. 33–44. [CrossRef]
Singh, T. , Singh, R. K. , and Ghosh, A. K. , 2008, “ Axisymmetric Global Structural Analysis of Barc Prestressed Concrete Containment Model for Beyond Design Pressure,” BARC Report No. BARC-2008/E/020.
Prasad, Y. S. R. , 1996, “ Indian Nuclear Power Programme: Challenges in PHWR Technology,” IAEA Technical Committee Meeting on Advances in Heavy Water Reactor Technology, Mumbai, India, Jan. 29–Feb. 1, pp. 29–46.
Puri, R. K. , and Singh, M. , 2009, “ BARCIS-BARC Channel Inspection System,” International Conference on Peaceful Uses of Atomic Energy, New Delhi, India, pp. 647–648.
Van Hong, L. , and Chatterjee, B. , 2002, “ Large LOCA Analysis of Indian Pressurised Heavy Water Reactor-220 MWel,” Nucl. Sci. Tech., 1(1), pp. 12–17.
Gupta, H. P. , and Jain, V. K. , 1993, “ Development of Computer Codes for the Analysis of Reactivity Induced Transients in PHWRs,” I Technical Coordination Meeting (TCM) on Advances in Heavy Water Reactors, Toronto, ON, Canada, June 7–10, pp. 187–189.
Lawande, Q. V. , Yadav, R. D. S. , and Gupta, H. P. , 2002, “ Transient Analysis in a 500 MWel PHWR,” National Conference on Nuclear Reactor Safety, Mumbai, India, Nov. 25–27, p. 305.
A-6 Interim report, “ A-6 Report on Interim Report on Safety Evaluation of 700 MWel Indian PHWRs KAPP3&4 and RAPP7&8 Post Fukushima Event,”, Nuclear Power Corporation of India Limited (NPCIL), India, accessed Dec. 3, 2016, https:\\www.npcil.nic.in /pdf/A6.pdf
Nuclear Power Corporation, 2015, “ Plants Under Operation, All Plant,” Nuclear Power Corporation of India Limited, India, accessed Dec. 3, 2016, https:\\www.npcil.nic.in
Soni, R. , Prasad, P. , Vijay Kumar, S. , Chhatre, A. , and Dwivedi, K. P. , 2005, “ Fuel Technology Evolution for Indian PHWRs,” International Conference on WWER Fuel Performance Modelling, and Experimental Support, Albena, Bulgaria, Sept. 19–23.
Bhardwaj, S. A. , Kumar, A. N. , Prasad, P. N. , and Ravi, M. , 2003, “ Fuel Performance Design and Development,” 8th International Conference on CANDU Fuel, Canadian Nuclear Society, Toronto, ON, Canada, Sept. 21–24, pp. 98–108.
Bhardwaj, S. A. , and Das, M. , 1986, “ Fuel Design Evolution in Indian PHWRs,” International Symposium on Improvements in Water Reactor Fuel Technology and Utilisation, Stockholm, Sweden, Sept. 15–19, IAEA, Austria, pp. 129–136.
Prasad, P. N. , Tripathi, R. M. , Kumar, A. N. , Ray, S. , and Dwivedi, K. P. , 2010, “ Fuel Element Designs for Achieving High Burn-Ups in 220 MW(e) Indian PHWRs,” Report No. IAEA-TECDOC-1654, pp. 75–81.
Balakrishnan, K. , Majumdar, S. , Ramanujam, A. , and Kakodkar, A. , 2002, “ The Indian Perspective on Thorium Fuel Cycles,” Report No. IAEA-TECDOC-1319, pp. 257–265.
Gupta, H. P. , Menon, S. V. G. , and Banerjee, S. , 2008, “ Advanced Fuel Cycles for Use in PHWRs,” J. Nucl. Mater., 383(1–2), pp. 54–62. [CrossRef]
Boczar, P. G. , Chan, P. S. W. , Dyck, G. R. , Ellis, R. J. , Jones, R. T. , and Sullivan, J. D. , 2002, “ Thorium Fuel-Cycle Studies for CANDU Reactors,” Report No. IAEA-TECDOC-1319, pp. 25–41.
Banerjee, S. , Gupta, H. P. , and Bhardwaj, S. A. , 2016, “ Nuclear Power From Thorium: Different Options,” Curr. Sci., 111(10), pp. 1607–1623. [CrossRef]
Balakrishnan, K. , 1994, “ Optimisation of Initial Fuel Loading of the Indian PHWR With Thorium Bundles for Achieving Full Power,” Ann. Nucl. Energy, 21(1), pp. 1–9. [CrossRef]
Balakrishnan, K. , and Kakodkar, A. , 1992, “ Preliminary Physics Design of Advanced Heavy Water Reactor (AHWR),” Report No. IAEA-TECDOC-638, pp. 70–77.
Sinha, R. K. , and Kakodkar, A. , 2006, “ Design and Development of AHWR-The Indian Thorium Fuelled Innovative Nuclear Reactor,” Nucl. Eng. Des., 236(1), pp. 683–700. [CrossRef]
Kannan, U. , and Krishnani, P. D. , 2013, “ Energy From Thorium—An Indian Perspective,” Sadhana, 38(5), pp. 817–837.
Mukesh, Kumar , Nayak, A. K. , Jain, V. , Vijayan, P. K. , and Vaze, K. K. , 2013, “ Managing a Prolonged Station Blackout Condition in AHWR by Passive Means,” Nucl. Eng. Tech., 45(5), pp. 605–612. [CrossRef]
Balu, K. , Purushotham, D. S. C. , and Kakodkar, A. , 1998, “ Closing Fuel Cycle–A Superior Option for India,” Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors, Victoria, Canada, Apr. 28–May 1, pp. 25–34.
Wattal, P. K. , 2013, “ Recycling Challenges of Thorium-Based Fuel,” International Thorium Energy Conference, (IThEC-13), Cern, Geneva, Switzerland, Oct. 27–31, pp. 171–175.
Balu, K. , and Ramanujam, A. , 1999, “ Reprocessing and Recycling of U/Pu–A Safer Option for Optimum Utilisation of Resources in the Nuclear Fuel Cycle,” Radiation Protection in Nuclear Fuel Cycle: Control of Occupational and Public Exposures, R. K. Pushparaja , P. R. Sangurdekar , and T. Kurien , eds., Indian Association of Radiation Protection, Kakrapar, India, pp. 57–62.

Figures

Grahic Jump Location
Fig. 4

Schematic of zone control units, shut down system (SDS)-1 and shut down system (SDS)-2 for 540 and 700 MWel PHWRs

Grahic Jump Location
Fig. 3

Schematic of liquid poison injection system (LPIS) for 220 MWel PHWRs

Grahic Jump Location
Fig. 2

Schematic of automatic liquid poison addition system (ALPAS) and gravity addition of boron (GRAB) for 220 MWel PHWRs

Grahic Jump Location
Fig. 1

Schematic of secondary shutdown system (SSS) for 220 MWel PHWRs

Grahic Jump Location
Fig. 5

Schematic of containments of (a) RAPS, (b) KAPS, (c) Kaiga standardized 220 MWel PHWR, and (d) Crane putting steam generator through containment opening

Grahic Jump Location
Fig. 6

Schematic of containment spray system of 700 MWel PHWR

Grahic Jump Location
Fig. 9

Schematic of coolant channel and associated moderator of 540 MWel PHWR (axial view)

Grahic Jump Location
Fig. 7

(a) Calandria with support rods and (b) calandria-end shield integral assembly

Grahic Jump Location
Fig. 8

(a) Fuel handling system and (b) schematic of passive decay heat removal (PDHR) system of 700 MWel PHWR

Grahic Jump Location
Fig. 10

Flow diagram of accident analysis

Grahic Jump Location
Fig. 11

Normalized power as function of time for loss of coolant accident (LOCA) benchmark problem

Grahic Jump Location
Fig. 12

Relative power versus time in loss of regulation accident (LORA) of 540 MWel PHWR

Grahic Jump Location
Fig. 13

(a) Emergency power supply scheme in a nuclear power plant and (b) fault tree for emergency power supply system in a nuclear power plant (courtesy P. V. Varde, BARC)

Grahic Jump Location
Fig. 14

(a) Simplified illustration showing event tree approach for accident sequence and core damage frequency (CDF) evaluation for off-site power supply failure scenario and (b) core damage frequency (courtesy P. V. Varde, BARC)

Grahic Jump Location
Fig. 15

Availability and capacity factors [26] (Courtesy NPCIL, India.)

Grahic Jump Location
Fig. 16

Variation of fissile inventory of U235 and U233 in gm/kg as function of burn-up when 3% enriched UO2 and ThO2 is used in PHWRs [34]

Tables

Errata

Discussions

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In