Home About us Editorial board Search Ahead of print Current issue Archives Submit article Instructions Subscribe Contacts Login 
Home Print this page Email this page Small font size Default font size Increase font size Users Online: 867


 
 Table of Contents 
TECHNICAL NOTE
Year : 2016  |  Volume : 39  |  Issue : 3  |  Page : 165-169  

Health physics operating experience in transporting cleanup spent resin and verification of methodology for estimating the shielding thickness of cask


1 Department of Atomic Energy, Health Physics Section, Tarapur Atomic Power Station 1 and 2, Tarapur Maharashtra Site, Nuclear Power Corporation of Ltd, Mumbai, Maharashtra, India
2 Department of Atomic Energy, Health Physics Section, Health Safety and Environment Group, Nuclear Power Corporation of Ltd, Mumbai, Maharashtra, India

Date of Web Publication30-Nov-2016

Correspondence Address:
Bahadur Singh K Rautela
Health Physics Section, Tarapur Atomic Power Station 1 and 2, Tarapur Maharashtra Site, NPCIL, Palghar - 401 504, Maharashtra
India
Login to access the Email id

Source of Support: None, Conflict of Interest: None


DOI: 10.4103/0972-0464.194963

Rights and Permissions
  Abstract 

Cleanup (CU) demineralizer of TAPS 1 and 2, containing mixed bed resins, is used to maintain reactor water chemistry. Once the bed is saturated, spent resin needs to be transferred to resin fixation facility for fixation and subsequent disposal in near-surface disposal facility. This is done by transporting the spent resin in a shielded resin transportation cask (SRTC) from radwaste building. The SRTC was manufactured with shielding thickness, based on methodology proposed by Rautela et al., “Shielding adequacy of proposed cask for transporting TAPS 1 and 2 CU spent resins.” Extensive radiation monitoring was carried out during the entire transfer and transportation process. Observed radiation levels on cask surface were within the accepted and calculated radiation levels. This paper describes verification methodology proposed by Rautela et al. and health physics operating experience during first CU spent resin transfer process at TAPS 1 and 2.

Keywords: AIRDUCT code, health physics experience, shielding adequacy, spent resin


How to cite this article:
Rautela BK, Sahu SR, Raveendran P S, Kothare PK, Mitra SR, Murthy BR. Health physics operating experience in transporting cleanup spent resin and verification of methodology for estimating the shielding thickness of cask. Radiat Prot Environ 2016;39:165-9

How to cite this URL:
Rautela BK, Sahu SR, Raveendran P S, Kothare PK, Mitra SR, Murthy BR. Health physics operating experience in transporting cleanup spent resin and verification of methodology for estimating the shielding thickness of cask. Radiat Prot Environ [serial online] 2016 [cited 2020 Oct 29];39:165-9. Available from: https://www.rpe.org.in/text.asp?2016/39/3/165/194963


  Introduction Top


Cleanup (CU) demineralizer system of TAPS 1 and 2 (boiling water reactor) plays a vital role in maintaining the reactor water chemistry. It reduces fuel failure, by minimizing the deposition of crud/water impurities on fuel surfaces. The system also helps in reducing the secondary sources of beta and gamma radiations by removing corrosion products, fission/activation products, and impurities from primary system.

This system is always kept in service, irrespective of whether reactor is in operation or shutdown condition. The demineralizer which is a mixed bed resin, containing both cation and anion resins, is provided in the system to remove soluble and insoluble impurities from the reactor water. Once depleted, the highly radioactive resin needs to be replaced. Earlier, the resins were transferred to a storage tank at radwaste (RW) building and thereafter transferred to spent resin storage tank for interim storage. As per the current industrial standards, practice of interim storage of spent resin was discontinued, and it was decided to transport CU spent resins directly from TAPS to resin fixation facility (RFF) at away-from-reactor (AFR) facility building of TAPS 1 and 2 for fixation by polymerization. The polymerized resins will then be disposed in reinforced cement concrete trenches of near-surface disposal facility. To achieve this, a resin transportation vessel (RTV) and a shielded resin transportation cask (SRTC) were fabricated. As this intrasite transfer of resins was carried out for the 1st time at the site, it was necessary to check the shielding adequacy of the cask before actual transfer.

This paper describes SRTC shielding verification methodology and health physics operating experience during first CU spent resin transfer process at TAPS 1 and 2.

Background

The CU spent resins were planned to be transported inside RTV, made of carbon steel (CS) of thickness 6 mm, housed inside a SRTC, made of lead, and sandwiched between CS plates of 8 mm on either sides.[1] The other physical parameters of RTV and SRTC are listed in [Table 1]. Thickness of lead used for fabrication of SRTC (65 mm) was calculated using a gamma shielding computer code AIRDUCT, developed by Health Physics Division, Bhabha Atomic Research Centre (BARC) for cylindrical geometries.[2] [Figure 1] shows a sketch of RTV housed inside SRTC.
Table 1: Design data of resin transportation vessel and shielded resin transportation cask

Click here to view
Figure 1: Sketch diagram of resin transportation vessel housed inside shielded resin transportation cask

Click here to view


The shielding calculation was carried out in the year 2009 based on the specific activity of the CU spent resin, i.e., 0.925 MBq/g, during that period. The composition of spent resin had 60 Co as most predominant radionuclide with a contribution of 90% followed by 137 Cs: 5%–6%,54 Mn and 134 Cs: 1%–2%. The RTV and SRTC together provided an effective shielding of 2TVL for an average photon energy of 1 MeV. Fabrication of cask was completed in the year 2012 and qualified by radiometry tests. The calculated maximum radiation level on external surfaces of SRTC was 0.48 mGy/h (with resins of specific activity of 0.925 MBq/g), well below the acceptable level.[3]


  Methodology Top


The actual spent resins transfer was carried out in the year 2015. Spectroscopic analysis of this resins was carried out in order to estimate the radionuclide composition and specific activity of the CU spent resin bed which was to be transported inside RTV. The specific activity of the resin batch which was to be transported in SRTC was found to be 1.36 MBq/g. Gamma shielding code, AIRDUCT, for cylindrical geometries, was used for reassessment of shielding adequacy checks in view of higher specific activity and different radionuclide compositions. Estimation of radiation levels with varying resin filled height and at specific distances from SRTC external surfaces was also done.

Empty SRTC was loaded on a trailer and was shifted to RW cask yard area. Mock-up trials of the entire resin transportation task were carried out using inactive resins. [Figure 2] shows a sketch diagram of RTV housed inside SRTC during resin transfer process. In view of the 1st time exercise at the station, it was decided to transfer CU spent resins to RFF at AFR in two campaigns. Due to high radioactivity content of the resins, additional protective measures were taken for radiation levels monitoring, contamination control, and optimization of personnel exposures. A special monitoring system having multidetectors with centralized data acquisition capability was installed to monitor radiation levels during the entire campaign. Radiation detectors of this system were installed at different locations around SRTC, resin transfer hose, and RW yard gate to monitor radiation level trending during the transfer. In addition, radiation survey of the selected locations in supervised area around RW building was carried out before, during, and after filling resin inside RTV to monitor the change in radiological background. [Figure 3] shows a layout of RW building outside area during resin transfer process. Radiation levels at supervised areas were within the stipulated limits. As an as low as reasonably achievable (ALARA) effort, RW yard floor was covered with one layer of polyvinyl chloride sheet and one layer of tarpaulin to contain the spread of contamination in case of any spillage due to any inadvertent occurrence of system failure. The inlet flexible hose and all flange joints on the top of SRTC were covered with lead sheets having 6 mm thickness to reduce the radiation level in a probable situation of resin being stuck in these joints after the resin transfer and flushing. Occupancy at resin transfer yard and vehicle route were restricted to prevent any unplanned exposure. The area around the resin transfer yard was cordoned during the resin filling, and special care was taken to keep the route unoccupied during transfer.
Figure 2: Sketch diagram of resin transportation vessel housed inside shielded resin transportation cask during resin transfer process

Click here to view
Figure 3: Layout of radwaste building outside area during resin transfer process

Click here to view



  Results and Discussion Top


The percentage composition of various radionuclides in the current resin, to be transferred, was observed to contain 60 Co with a contribution of 84%.54 Mn,58 Co,65 Zn, and 106 Ru contribution was 1%–6%.137 Cs was found to be contributing 0.39% [Table 2].[4]
Table 2: Composition of spent resin and activity

Click here to view


Calculated radiation level on contact and at one foot distance of SRTC (with half filled) were 0.96 mGy/h and 0.87 mGy/h respectively. These values were in very good agreement with the observed values of 1.0 mGy/h and 0.80 mGy/h, respectively, on contact and at one foot distance of SRTC. Similarly, calculated radiation level at outside supervised area was found to vary between 0.1 and 0.5 µGy/h against observed radiation level of 0.1–0.6 µGy/h during resin transfer work [Table 3]. Thus, it is observed that the calculated values are in very good agreement with the observed values. This validates the shielding adequacy of the SRTC, and the thickness of which was calculated using AIRDUCT code.
Table 3: Calculated and observed radiation levels on shielded resin transportation cask external surfaces

Click here to view


Although the cask was fabricated using data of a lower specific activity of CU spent resin bed, it is observed that the calculated maximum radiation level on external surface of the cask is 0.96 mGy/h, which is within the acceptable limit of 2 mGy/h on contact of shipment package.[5] Furthermore, the calculations show that the radiation levels on cask externals will be within the acceptable limits even when the entire CU spent resin bed is transferred in a single campaign.

No leakage or spillages occurred during the transfer of spent resin to cask. Floor contamination of RW yard was checked and found to be <3 kBq/m 2 (i.e., <1/6th of the allowable limit [6]) after the each campaign.

Radiation levels observed at outside area which falls under supervised category are also in very good agreement with the calculated values. It is observed that radiation levels even during resin transfer task, at outside areas, are well within the stipulated acceptable limits of 1 µGy/h for supervised areas.[6]

Estimation of radiation levels on external surfaces of SRTC using AIRDUCT code was considered for planning, preparedness for the execution of resin transfer job, and effective implementation of ALARA measures.


  Conclusions Top


As the observed radiation levels on external surfaces of SRTC (when RTV is filled up to 50% capacity) are in very good agreement with the values calculated using AIRDUCT code, it is concluded that the SRTC fabricated for transporting spent resins is adequately shielded during actual resin transfer works.

As the calculated radiation levels on SRTC surface (when RTV is filled up to its full capacity) are well below the acceptable radiation level (i.e., 2 mGy/h on contact of shipment package), it is concluded that the fabricated SRTC is adequately shielded to handle CU spent resins even when the RTV is filled up to its full capacity.

The AIRDUCT code has proved to be a vital tool for estimation of anticipated radiation levels at the work location for the 1st time work, which helps in better planning and better preparedness for the execution of exposure intensive jobs (such as resin transfer job) and effective implementation of ALARA measures Thus, it is concluded that the first-ever task carried out of resin transportation at TAPS 1 and 2 was successful, and the radiation protection procedures followed for it are adequate for such future campaigns.

Acknowledgements

The authors would like to express their gratitude to Shri VS Daniel, station director TAPS 1 and 2, Shri M Joshi, chief superintendent TAPS 1 and 2, and Shri AK Kundu, senior operation engineer, for providing the facilities to conduct this study. The authors are also thankful to Shri MR Somayajulu, chief engineer and Shri K Venkataramana, Adl. Chief engineer of health physics, HS and E group, NPCIL-HQ, for providing the encouragement and technical overview. The authors would like to express their gratitude to Shri VK Sharma Ex-Head SSS, HPD-BARC and Ms Chitra S, SSS, HPD, BARC, for the useful discussions and guidance.

Financial support and sponsorship

Nil.

Conflicts of interest

There are no conflicts of interest.

 
  References Top

1.
Resin Cask Design Data: Ref No. TAPS-1& 2/01150//2007/M/; 22 November, 2007.  Back to cited text no. 1
    
2.
Sharma VK. AIRDUCT: A Shielding Code for the Calculation of Radiation Sources in Cylindrical Geometry, HPD, 90.  Back to cited text no. 2
    
3.
Rautela BK, Singh AP, Sahu SR, Phadnis PS. Shielding Adequacy of Proposed Cask for Transporting TAPS 1&2 Clean – Up Spent Resin, IARP-NC on Recent Advancement in Radiation Dosimetry, Souvenir/Abstract Book. Indian Association For Radiation Protection: IARP; 2010. p. 38, 79.  Back to cited text no. 3
    
4.
Clean-up Resin Analysis Report from TAPS Station Chemist; 18 May, 2015.  Back to cited text no. 4
    
5.
AERB Safety Code On Safe Transport of Radioactive Material, Code No. AERB/NRF-TS/SC-1(Rev-1).  Back to cited text no. 5
    
6.
AREB Safety Manual on Radiation Protection for Nuclear Facilities, Manual No. AERB/NF/SM/O-2.  Back to cited text no. 6
    


    Figures

  [Figure 1], [Figure 2], [Figure 3]
 
 
    Tables

  [Table 1], [Table 2], [Table 3]



 

Top
   
 
  Search
 
Similar in PUBMED
   Search Pubmed for
   Search in Google Scholar for
 Related articles
Access Statistics
Email Alert *
Add to My List *
* Registration required (free)

 
  In this article
Abstract
Introduction
Methodology
Results and Disc...
Conclusions
References
Article Figures
Article Tables

 Article Access Statistics
    Viewed528    
    Printed5    
    Emailed0    
    PDF Downloaded102    
    Comments [Add]    

Recommend this journal