A Case Study on Application of HAZOP for Intermediate Level Radioactive Liquid Waste

Treatment Facility

Nilay J. Jasu, Kiran Kumar, Anand Gangadharan, Neelima S. Tomar, B. B.Verma and Kanwar Raj Waste Management Division, Nuclear Recycle Group

Bhabha Atomic Research Centre, Mumbai, India- 400085 jasunj@barc.gov.in

Abstract

An Ion Exchange (lEX) process is being adopted for treatment of

intermediate level radioactive liquid waste (ILW) streams generated

during reprocessing of spent nuclear fuel. In this process separation

of Cesium and Strontium from waste is achieved using highly

selective organic resins developed indigenously in BARC

laboratories. The process splits the waste into two major

components, firstly a decontaminated low active emuent stream for

dilution & discharge and secondly the highly radioactive

concentrated liquid stream for further immobilization. It is a step

towards implementation of the wealth from waste concept, where

the radioactive cesium bearing streams are stripped of the cesium

component to generate purified cesium product for irradiation

applications.

This system is planned to be retrofitted in existing hot cells of Waste

Immobilization plant, Trombay by utilizing the available

infrastructure. It is designed for remote operations of all the

different process sequences which involves a detailed process

control logistics and has various interfaces with the existing plant

features. HAZOPwas considered to be the appropriate tool to assess

the safety of the system prior to proceeding with the detailed

engineering and operation. Considering the design intent, the

entire process flow diagram was split into important segments like;

the transfer operations oflLW streams; the cesium sorption cycle,

resin bed regeneration cycle comprising of elution & conditioning

of the bed ; transfer of the highly radioactive cesium concentrate

streams; handling & disposal of the decontaminated emuents. The

radiochemical nature of solutions necessitated the equipment-wise

HAZOP methodology. The gadgets, valves, storage tanks, transfer

modes etc, for each process segment were evaluated individually. To

consider the variation of process parameters relevant to each study

node, detailed list of the same was prepared from operation point of

view. Apart from standard process parameters, certain specific

parameters related to radioactive contamination, radiation field

etc. were also considered. Application of standard guide words was

done to identify any reasonable or practical process variations.

Detailed study was done to find all possible causes and

consequences. These provided an in-depth knowledge of all possible

combinations of undesired events or maioperations. The necessary

978-1-4244-8343-3/10/$26.00 ©2010 IEEE 619

safeguards to be incorporated in design and operation were derived

and fmalized based on the Hazop study. As an outcome, we could

deduce the requirements of additional interlocks, alarms and

arrive at a detailed passing valve sequence analysis. The authors

feel that the study can be done on discrete and continuous process

path for each system and it is to be done at the design stage. It is

further deduced that Fault tree analysis for hazards and

equipments is to be subsequently followed based on the criticality

and probability of process variation.

Keywords- intermediate level radioactive liquid waste; Cesium and

Strontium from waste; highly selective organic resins; wealth from

waste; retrofitted in existing hot cells; interfaces with the existing

plantfeoJures; safety of the system; equipment-wiseJlAZOP methodology;

causes and consequences; undesired events or maloperations.

I. INTRODUCTION

In the process of contributing towards the growth and progress of human civilization, all industrial activities inevitably give rise to some unwanted waste product. This is true for nuclear industry as well. However, the radioactive waste produced during nuclear power generation differs from other waste as it contains small quantities of radioactive materials. Further, the safe and effective management of radioactive waste has been given utmost importance from the very inception of nuclear industry in India. Three basic philosophies followed in the management of the liquid radioactive waste are 1) Concentrate and Contain 2) Delay and Decay and 3) Dilute and Disperse. Present technologies for radioactive waste management lay great emphasis on the confinement of radioactivity involving recovery and recycle and immobilization of residue in inert solid matrices. Therefore, greater emphasis is being laid on the concept of wealth from waste, i.e. recovering the radioactive components from the waste streams and thereby utilizing them for other useful purposes e.g. irradiation. The scope ofthis paper is limited to the discussion related to management of intermediate level liquid waste by recovery of useful isotopes like 137Cs and 90Sr implementing

2010:r International Conference on Reliability, Safety & Hazard (ICRESH-201O)

wealth from waste concept. The safety analysis study using Hazop is dealt in this paper, which was done to ascertain the safety features in design and to obtain the necessary measures to combat operational hazards in the management ofILW streams.

II. INTERMEDIATE LEVEL LIQUID WASTE

A. Source o/Waste The categorization of different waste forms, as per Indian Atomic Energy Regulatory Board (AERB), with respect to their radioactive characteristics is as shown in Table I. In the chop leach process in reprocessing of the spent fuel, the dejacketting of the fuel elements produces alkaline Al bearing declad wastes and various neutralized concentrates obtained during waste evaporation cycle. Both of these wastes comprise to form the main source of intermediate level waste streams. The closed fuel cycle option adopted by India, emphasizes on spent fuel reprocessing to recover plutonium and unused uranium and recycling to the power reactors. During the reprocessing of the spent fuel, liquid radioactive wastes of various compositions and different level of radioactivity is generated. Broadly they are classified as high level (-Ci/L), intermediate level (-mCi/L) and low level waste hlCi/L). ILW streams contribute to about 7% of the total waste volumes generated and 4% of the total radioactivity, when qualitatively and quantitatively compared with the total high level and low level wastes produced in fuel reprocessing.

Charaderioti •• of alkaline reproo:e .. ing ..... te

High coneenlmlio,. 0/ Na .alb

137 90 RJUlioactivity, _iIIly due to C. and Sr

106 125 Tracer 0/ RK, Sb, l'K

TABLE I. CATEGORIZATION OF RADIOACTIVE WASTE As PER AERB GUIDELINES

Classification Solid (contact Liquid (p-y Gaseous (p-y

dose rate) Sp. Activity) Sp. Activity) Cat-I < 200 mRIhr < 1O.6Ci/ m3 < 10.10 Ci/m3

>10.6 Citm3 > 10.10 Cit m3 >200 mRIhr

&::; 10.3 Cit & < 10.6 Cit Cat-II & ::;2R1hr

m3 m3

>10.6 Citm3

Cat-III >::;2R1hr &::; 10.3 Cit > 1O.6Citm3

m3

a -Bearing >10.3 Citm3 Cat-IV

waste (>100 &::; 101 Cit ---nCi/g) m3

Cat-V ----- > 101 Citm3 ----

Due to the compositional variation, these wastes are segregated for effective treatment. ILW is characterized by the presence of 137 Cs as the maj or radionuclide, which amounts to 98- 99% of the total radionuclides. Other elements present are 9OSr,106Ru, 12SSb and inactive AI. Typically the stream has a 137 Cs activity ranging

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from 5- 50 mCiIl and is characterized by high sodium Icesium ratio (104 -lOS) along with high total solids content , majority in the form ofNaN0

3 amounting to 100- 300gm/l.

B. Process PrespectiYe Pretreatment, treatment and immobilization techniques are applied for management of radioactive waste wherein volume reduction factor (VRF) and decontamination factor (DF) are the key parameters in choice of technologies. Conventional pretreatment technologies like evaporation, membrane separation, ion exchange (IX), compaction, incineration etc. are considered for this purpose. The concentrates obtained from the primary treatment processes are conditioned to an acceptable waste form, i.e. fixing the radio nuclides in appropriate immobilization matrix. Currently employed management practices emphasizes on volume reduction and conversion into a stable waste form for disposal and storage, with physical confinement of radionuclides to restrict their mobility. The salient features which are considered for selection of such conditioning techniques depend on chemical and radiochemical toxicity of the waste, viability of the process and throughput, effective volume reduction and conversion into a product of good radiological, thermal and chemical stability. These conditioned products were then disposed either in near surface disposal facilities (NSDF) with engineered and natural barriers following the regulatory guidelines or stored for ultimate future disposal in geological formation.

• Bituminization: Conventionally, for conditioning of ILW, predominant technologies such as bituminization, cementation, polymerization techniques were applied. In India the first two methods were used in the earlier decades of waste management. Bituminization of ILW has been practiced in Waste Immobilization Plant (WIP) Tarapur earlier. The waste stream with additives was blended with bitumen emulsion and then the excess water content was reduced by evaporation in wiped film evaporator and subsequently poured in bituminized waste product drums. Crucial for smooth processing were parameters like choice of bitumen, processing features of temperature, waste to bitumen ratio, off gas clean up etc. Compatibility of high salt bearing wastes with bitumen was of concern, since ILW being laden with nitrates of various cations has an inbuilt possibility of leading to formation of an explosive mixture. Exothermic reaction between nitrates present in the waste streams and bitumen had a risk of fire hazard. Process safety was ensured with proper checklisting and safety interlocks. Operational and maintenance problems, high temperature requirement to maintain the desired flow characteristics, lower radiation stability of the product and a negative volume reduction led to search and adoption of alternative processes.

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• Cementation: Cement and cement composites have been used as an inorganic matrix usually for conditioning of ILW which has been a widely accepted technique for immobilization of aqueous waste because of ease of operation, low temperature requirements, and good radiation stability. Various types of mixers, like in-situ mixer, in-drum mixer with reusable or once through agitator, in line mixer or cone mixers are employed for intimately mixing waste and cement slurry or dry powder. The waste product is then cast into a solid monolith in drums. The type of mixing, time of mixing, ratio of waste to cement, quantity of bleed water, curing temperature and time are important parameters to assure the product quality. Cementation was initially planned for use in Waste Immobilization Plant (WlP) Trombay. However other options were developed due to problems of negative volume reduction factor of the immobilized product, limitation of the total waste loading, and incompatibility with certain chemicals present in IL W.

• Ion Exchange: Selective ion exchange has a direct benefit of lower final waste volumes generated for disposal. Commercially available organic/ inorganic ion exchange resins already find their application in various stages of the fuel cycle. The scheme adopted at WIP, Trombay utilizes indigenously developed resorcinol formaldehyde polycondensate resin for cesium sorption with a very high efficiency. RFPC resin has very high selectivity; good exchange capacity; processability; regenarability and radiation stability under the process conditions. The phenolic groups of the resin ionize in alkaline condition and provide cation exchange sites for uptake of Cs + ions with very high selectivity. In series, removal of 90Sr is carried out using commercially available macroporous poly styrene divinyl benzene based chelating imino diacetic acid group.

Earlier, IL W by lEX process was treated in a 1 6 I column at WlP Tarapur employing this method. A scaled up version was implemented at Trombay in a mobile treatment unit, where Tank Remediation by Ion Exchange (TRlX) campaign was performed employing similar lEX process

using 100 I columns, which successfully treated 2500m3 of ILW. Few kilocuries of 1 37 Cs activity were recovered by this process. The design and operational experience gained led to the conceivement of a processing facility with remote operations dedicated for ILW treatment at BARC, Trombay. This has a distinct benefit to improve upon the reduction of radiation exposures during O&M, ease of operation and to implement the recovery of multi kilo curie of this fission product for use in brachytheraphy, blood irradiation, hygenisation of sewage sludge, food irradiation etc.

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III. SYSTEM FOR HAZOP: lEX TREATMENT OF ILW

As indicated above the Ion Exchange (lEX) Process is planned to be used for treatment of large volumes of legacy waste and presently generated fresh intermediate level waste produced from the research reactor fuel reprocessing. The IX columns consisting of 100 1 of organic resins RFPR and IDAR will be used. The Ion Exchange treatment facility is set up in the boosting station (Pump House) of Waste Immobilization Plant, Trombay with an objective to serve as a dedicated and independent facility for the treatment of ILW. This will enable Pump House to serve as an IL waste treatment facility in addition to original intent of serving as an intermediate boosting station for transferring the radioactive liquid wastes of different categories (HLW & ILW) to WIP, Trombay. The main benefits includes recovery of Cesium from ILW and its transfer to WIP for further processing; retention of the existing purpose of Pump House and retrofitting this new process and optimum utilization of the existing underground radioactive waste transfer trench lines.

A. Process Schematic In adopted process schematic, as shown in Fig 1 , the cesium entrapped in the course of sorption cycle is later removed from the resin by nitric acid streams. This acidic stream, the eluate, is purified from other radiochemical contaminants and primarily cesium rich, with gross specific activity in the range of 2.5Ci/1 and N a/ Cs ratio of 1 02. This product stream will be transferred to WIP for immobilization in glass matrix to generate cesium rich vitrified product, which will be subsequently converted to glass pencils required for irradiation purposes. The conditioned resin which is reconverted to exchangeable form during the regeneration cycle by sodium hydroxide solutions will be redeployed for sorption of cesium in further cycles. Thus, in consecutive loading and regeneration cycles large volumes of waste will be stripped of Cesium and treated by a single batch of

resin. In each loading cycle, about 2000 Ci of 137CS is loaded in the resin. A VRF of 50 to 100 in each cycle, with overall DF more than 10,000, is obtained depending on the input stream concentration. The decontaminated streams, namely the effiuents are expected to be of average specific activity of 5 Xl O· 2mCili. The effiuents containing trace amount of short lived radionuc1ides are subjected to chemical treatment in c1ariflocculator for decontamination by co-precipitation methods. Fig. 2 gives the further downstream of the effiuents prior to discharge.

2010 r'International Conference on Reliability, Safety & Hazard (ICRESH-201O)

TABLE II. COMPOSmON OF VARIOUS STREAMS IN TIlE lEx FACILITY

Characteristics

Parameter Feed Effluents from Ion Eluate Exchange

10- 13 10- 13 0.4 Molar pH (If)

Specific Activity Gross �;y 4 - 50 mCi/lt 7xlO-3 -2.0xlO-2 Average (mCi/lt) mCillt (J37Cs) : 0.24

Cillt

137CS 3.9 - 49 :5 5xlO-3 mCillt Peak. mCillt Sp.Activity

(J37Cs): 2.3Ci/lt

90Sr 0.04 - 0.5 0.35- 0.45 mCillt Nil mCillt

The sludge generated is concentrated in decanters, where VRF of lOis achieved_ Decant and supernatant from clariflocculator is mixed and discharged after dilution maintaining the specific activity of 1 0-4 mCi/1 i.e. within permissible discharge limits. The concentrated sludges are subsequently immobilized in cement matrices and CWP will be transferred for disposal in NSDF. Table II gives the composition of the various process streams.

Main Storage

Tank

Figure 1. Treatment Schematic ofIEX Facility

Ef1luents for Down

stream Procculng

Figure 2. Schematic of Downstream Treatment of Effiuents

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B. lEX System description The Ion exchange system is retrofitted utilizing the infrastructure available at Pump House. The shielded cells were suitably augmented to house the IX columns, valves and pumps maintaining the radiation levels within allowable ranges and various interfaces with the existing facility were taken due care so as to avoid any undesired events of spread of contaminants or rise in radiation field. Existing service distribution system, off gas and ventilation systems are planned to be used. Certain changes are incorporated with an objective of retention of the present purpose of Pump House. The IX system comprises of two feed pumps, one polypropylene prefilter, three numbers of SS 304L IX columns, one polypropylene resin trap, seven storage tanks, associated remotely operable valves, various remote handling equipments as spring loaded and pneumatically actuated mechanisms for flange connections, fixed position and movable trolleys equipped with drives and roller conveyers for shifting of the equipments, associated shielding cubicles with remotely operable sliding doors etc. The controlling of the operation and monitoring of the parameters will be carried out from the centralized control room involving the PLC based control panel. There will be closed! open loop control with interlocks for process parameter control. This will also provide graphics with static and dynamic display of process parameters in different display modes and operation status of control elements. There are two RFPR columns and one IDAR column. The first column acts for the major sorption and second is the polishing one. After each loading cycle, interchange of the leading and polishing column is planned. All the process operations necessary to implement the different operation sequence are planned to be carried out remotely. Interfaces with other high level radioactive svstems were to be looked into precisely and take necessary precautionary measures so as to obtain a detailed feedback prior to the finalization of engineering safety aspects.

IV. SAFETY ANALYSIS METHODOLOGY USING HAZOP

HAZOP was considered to be the appropriate tool for performing the analysis, to assess the safety of the system prior to proceeding with the detailed engineering and operation. The study was conducted during the conceptual design phase where recommendations affecting the general design could be implemented. The process was slightly customized. HAZOP was performed on each process equipments considering all the online gadgets, the related valves and machineries, storage equipments, ion exchange columns and piping. On developing the engineering flow diagrams, P& ID and on deciding the detailed retrofit features in the existing layout of equipments and distribution systems the lEX system was splitted into following six maj or components depending on the importance of operation namely:

2010 2"" International Conference on Reliability, Safety & Hazard (ICRESH-201O)

a) Receipt of ILW from underground storage tanks of 1500 m3 capacity to batching tank of 10001 capacity.

b) Feed conditioning and sorption cycle of cesium and strontium in the forward and reverse modes which are executed by remote operation of 34 numbers of pneumatically operated valves in predetermined sequence and pneumatically operated IX columnflanges.

c) Resin bed regeneration including two broad steps of operation, firstly elution of the resin to remove the absorbed Cs ions and secondly conditioning of the bed to bring into reusable form for the next cycle. This is to be applied for all the three IX columns.

d) Effluent transfer system to store, handle and pump the effluents of lEX to further downstream processing of co­precipitation and concentration.

e) Vessel Off Gas system for providing necessary negative pressure in the process vessels and decontamination of off gases prior to discharge.

j) Transfer of high radioactive Cs rich eluate streamfrom lEX facility to immobilization facility for vitrification through underground trench lines and remotely operable transfer modes.

All the operation sequencing were chalked out initially and necessary operation of valves and other gadgets so as to implement desired routing of the waste streams were finalized, along with operation of services and auxiliary systems. Fig. 3 shows a typical process routing for the forward sorption cycle. Similar process routing drawings, 1 3 numbers, were prepared to study each of the process cycles of forward loading( IXI- IX2-IX3), reverse loading (IX2- IXI- IX3), elution and conditioning of three columns, prefilter bypass operation, backwashing of prefilter, air flushing of three columns, considering the different status of valves and flow routings. For each of the six systems as stated above, equipment- wise listing was done to consider all the safety related and process equipments like storage tanks, IX columns, prefilter, resin trap, control valves, pneumatic flanges, pneumatically operated normally open! closed valves, transfer pumps, steam jets, strainers etc. These equipments formed the "study nodes" in each system. During this analysis, certain assumptions were made regarding the correctness of instrumentation indications and availability of services of compressed air, high pressure steam, demineralised water etc. Before considering for the variation of process parameters, all the desired range of process parameters, required for the healthiness of the process, adequate decontamination and volume reduction were finalized. It was important to consider the reasonable variation of relevant process parameters as flow rate, level, pressure drop, density, temperature, pH, current etc using the standard guide words of 'no', 'more', 'less', 'as well as', 'reverse', 'other than' and 'part of. In addition to the above process parameters, specific activity of effluents, radiation field

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in operating areas, air borne activity and surface contamination were also considered since from radiological aspects the variation of these parameters were more important to be ascertained during regular operation of the facility. To explain one specific system, the resin bed regeneration has been considered as one of the most significant process sequence since it generates high radioactive liquid solution. The system study nodes as assumed were acid! alkali storage tank, feeding pump, IX column, air operated valves (typically open), air operated valves (typically closed), eluate I regenerant collection tank. Certain important process deviations considered during the study were no/less molarity of acid for elution, more flow than desired flow rate through column, alkali! water as well as acid flow, more specific activity of eluate (> 2.5 Ci!l), more pressure drop in the resin bed, more flow through typically closed AOV etc.

F or each set of variation of process parameters, a detailed analysis was done to identify all the probable causes arising from the new system or the interfaces with the existing facility, which can lead to occurrence of these untoward events or potential hazards. In the case of occurrence of such events, the consequences were traced giving emphasis on the aspects which can lead to undesired eventsl maloperations such as, release of radioactive solutions to undesired areas of higher personnel occupancy, cross contamination of streams leading to unmanageable products, wrong sequence of operation leading to decrease of DF or VRF, radiological and industrial hazards etc. F or such cases, all the safeguards which can be provided in design so as to prevent these events from occurring were listed. In addition corrective measures and lor actions which can be taken from design stage to be incorporated as hard wired passive safety features were formalized. Problems which can happen during operation leading to undesired events were also enlisted.

Figure 3. Schematic of Typical Loading Cycle

20l0:r International Conference on Reliability, Safety & Hazard (ICRESH-201O)

V. RESULTS

Hazop study revealed the importance of careful waste characterization to ensure the selection of the treatment options. Inventory of the nature and type of radioactive elements, the individual concentrations and radiotoxicity, concentration of non radioactive elements and their interference in the process were also found to affect various process variations where it can lead to decrease of DF and lowering of VRF. Such a case was where, the process deviation was of "more specific activity of effluents" and the cause leading to this event results from abnormality in the waste generation cycle during reprocessing. So efficiency of the lEX process will also depend on the processing parameters of the ILW generator or extent of cooling of the irradiated fuel bundles prior to reprocessing. It may also be due to presence higher quantity of radionuclides, short or long lived, other than Cs and Sr like Ruthenium and Antimony, presence of which in ILW will lead to high specific activity of the effiuents after lEX. This helped in finalizing the sampling and analysis frequency essential during each cycle of operation. In the study, from causes of relevant process deviations like "less specific activity of eluate", "more pressure drop of columns" certain important parameters of selectivity, exchange capacity, resin particle sizes which were necessary for preliminary selection of resin could be inferred.

Incorporation of passive design safety features like gadgets, back flow preventing valves and equipments was found essential to avoid the events leading to unwanted process variations. Non return valves were inferred to be installed in six places e.g. for prevention of reverse flow of ILW into the service distribution headers, back flow of effiuents during the transfer to downstream processing facility, eluate transfer to eluate receiving tank, additive inlet lines, discharge of off gas from the air operated pump etc. Schemes necessary for handling of unmanageable streams during operation steps were accordingly implemented and additional design features were incorporated. As described earlier, the process steps involve sequential operation of various AOV's, and any maloperation of these can lead to a breach in multiple barriers of safety. An in-depth knowledge was obtained on undergoing an "analysis of valve passing scenario ", which proved very effective in determining the critical sets of valves which can lead to untoward events including radiological and chemical hazards. In this detailed analysis, the effect of passing of valves leading to higher specific activity of effiuents, presence of high level radioactive solution in unwanted destinations, spread of surface contamination in service areas, under utilization of acid! alkali during regeneration etc were identified. This calls for extra vigilance and effective quality assurance in inspection, installation and during operation. This enabled us to determine all possible routings of inactive and active liquids. In built safety features as felt necessary from the study were designed to be implemented,

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like the provision for forward and backward routings of all the streams, necessary bypassing provisions in the treatment options, provision of additional instrumentation to provide indication of presence of liquid in unwanted areas etc. An instance of process variation of "high level in effiuents monitoring tank" can have a consequence of liquid reporting in the sump area, for a safeguard it was decided to provide an alarm with conductivity type level probe for alerting the operator.

Radiation protection equipments of adequate detection ranges, were felt essential in critical areas and on service distribution systems at the interface of different radiation zones. Low range area gamma monitors on effiuent distribution header or regenerant addition system, strainer flushing line and continuous air monitors for detection of breach of pipings etc. has been planned to be incorporated. This was to ascertain any presence of radioactive solutions or air borne activities in amber areas or personnel occupied areas, and to detect failure of process systems. We could arrive at the requirement of providing low range area gamma monitor in four additional places in the facility to have a measure of identification and control of presence of radioactive fluids in unwanted places. Before proceeding with detailed engineering, it was quite an effective tool to determine the necessary process safety interlocks & alarms, operating ranges & critical parameters, requirement of additional limit switches & CCT V viewing which will help in ease of various operation sequences. This also helped in formulating various standard operation procedures, emergency operation procedures, requisition slips, checklists and administrative checks necessary to help in formation of documents and methods to enable safe operation manuals. A necessity was felt for isolation of process control panels from remote handling operation console. Certain administrative measures like check on batch size, hydrotesting on column replacement, suitable isolation of non operating valves and determining the sequence of such valve lists required in each parallel operation step, which were felt necessary for providing safeguards of process variation were deduced and decided to be implemented during operation. Checklisting was identified to be very important for selection of desired sequencing of valves, pump availability of acid or other chemicals. According to the necessity of various critical system failure, emergency operation procedures for handling of spillage of radioactive liquid, chemicals, combating situations of electrical supply and compressor failure were prepared.

The hazop study provided a gross picture of all the process variations of which critical events identification could be done and it was felt that further study will provide insight of the root causes of such events and in continuation the consequences which can affect the ILW management or safety of personnel and equipments involved can be deduced.

20l0:r International Conference on Reliability, Safety & Hazard (ICRESH-201O)

VI. SUMMARY

It is concluded that performing this study at the inception of detailed engineering activities provided us with various insight of the necessary design features that should be incorporated as passive safety measures, installation of gadgets and additional equipments to avoid any untoward events and a trouble shooting document to be implemented for process operation. It is also inferred that brainstonning sessions using this analysis tool provides with a qualitative study. As a necessity to quantify the probability of events and inferring the probabilistic nature of the process variation this is to be augmented with further improvised techniques.

This study provides with a first line approach in detennining various causes and consequences responsible for particular process variation. On obtaining the initial guideline, it will be further fruitful in performing a detailed root cause analysis of important events by the fault tree approach. These events can be finalized on the basis of hazop study and their criticality & probabilistic criterion. This study gives an in-depth knowledge of all possible combinations of undesired events or maloperations. This can help to serve as a guide for identifying a correct approach for handling the operational problems and combating them. It can, thereby, reduce the downtime of the treatment process, maintain the targets of production capacity, reduce personnel exposures as per ALARA, and prevent events at the budding stage, provide guidelines to implement important design features and safe operation practices.

This is afirst attempt made at WIP Trombay to apply hazop and perfonn the safety analysis for an intermediate level radioactive waste management facility. Although as a standard practice, it is

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used for conventional industries, here an approach was made to perfonn the study on equipment wise basis, since a single gadget can lead to radiological hazards. It is felt that the batch process can be split in continuous segments, and similar study can be done on assuming discrete continuous path in the different operation sequence of the treatment, and then apply the process parameter variation to assess the consequences. This can avoid unnecessary duplications or cumbersome nature of the problems in certain areas and a time saving approach towards the assessment study.

ACKNOWLEDGMENTS

Authors acknowledge the valuable help provided by Shri S.D.Bharambhe, Shri Keyur Pancholi, Smt. Helen Jeyaraj, Shri Guruprasad, Shri D. N. Patil and other colleagues of WIP Trombay, BARe in perfonning this study and for their important suggestions.

REFERENCES

[I] Kanwar Raj, K. K. Prasad and Bansal N.K., "Radioactive Waste Management Practices in India", Nuclear Engineering and Design, Elservier, 236 (2006), 914-930.

[2] S K Samanta et al., ''Nuclear Waste Management- Practices and Trends", IANCAS builetin, vol. 11, No. I,April 1997

[3] Rausand Marvin, "Hazard and Operability Study", System Reliability Theory, 2'" Edition, Wiley 2004.

[4] International Atomic Energy Agency, "Requirements and methods for low and intermediate Level Waste Package Acceptability", IAEA Tecbinal Document No. 864, Vienna 1996

[5] International Atomic Energy Agency, "Handling and Processing of Radioactive Waste from Nuclear Application", IAEA Tecbinal Report Series No. 402, Vienna 2001.