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 Table of Contents 
ORIGINAL ARTICLE
Year : 2018  |  Volume : 41  |  Issue : 3  |  Page : 143-147  

Calculation of dose rates due to loss-of-coolant accident in open-pool spent-fuel storage


Department of Reactors, Nuclear Research Center, Atomic Energy Authority, Cairo, Egypt

Date of Submission04-May-2018
Date of Decision03-Aug-2018
Date of Acceptance23-Aug-2018
Date of Web Publication19-Nov-2018

Correspondence Address:
Dr. Amr Abdelhady
Department of Reactors, Nuclear Research Center, Atomic Energy Authority, Cairo
Egypt
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Source of Support: None, Conflict of Interest: None


DOI: 10.4103/rpe.RPE_31_18

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  Abstract 


The objective of the spent-fuel storage pool is shielding the worker and public from radiation emitted by radioactive decay in the spent fuel and providing a barrier for any radioactive release. In open-pool multipurpose reactor, the spent-fuel storage pool is connected to the main pool through the transfer channel. It was prepared to store 528 spent-fuel elements distributed in two racks that constructed one above the other. The loss-of-coolant accident (LOCA) in spent-fuel storage pool could result in rising of the radiation dose in the reactor building as the water level in the pool falls. The value of the radiation dose rate depends on the height of the water level above the spent fuels, and the number of spent-fuel elements stored in the storage pool during LOCA. The dose rate calculations were carried out starting from the minimum height which the water level could drop above the spent-fuel storage racks. The calculations were carried for two cases as follows: the full capacity of both racks and the full capacity of the lower rack only. Monte Carlo N-Particle Transport MCNP5 code was used to calculate the radiation dose rate above the storage pool and in the control room. The results show that the dose rate in the control room would be lower than the permissible limit when the water level height was 270 and 140 cm for the two cases, respectively. The dose rate above the storage pool would be lower than the permissible limit when the water height above the racks is higher than 385 cm in the first case, and 290 cm for the second case.

Keywords: Loss-of-coolant accident, Monte Carlo N-particle transport code, radiation dose, spent-fuel storage


How to cite this article:
Abdelhady A. Calculation of dose rates due to loss-of-coolant accident in open-pool spent-fuel storage. Radiat Prot Environ 2018;41:143-7

How to cite this URL:
Abdelhady A. Calculation of dose rates due to loss-of-coolant accident in open-pool spent-fuel storage. Radiat Prot Environ [serial online] 2018 [cited 2018 Dec 10];41:143-7. Available from: http://www.rpe.org.in/text.asp?2018/41/3/143/245794




  Introduction Top


Avoiding exposure to high radiation dose is necessary for nuclear reactors. High radiation dose rate causes that the worker would receive an accumulated dose which may exceed the permissible annual limit. It also results in making some important places in the reactor are inaccessible. The radiation dose level in the open-pool reactor may rise due to the following accidents: Fuel element withdrawal from the reactor core,[1] ejection of irradiated materials from the reactor core,[2] and decreasing in water level in the reactor pools due to the loss-of-coolant accident (LOCA). Evaluating the radiation dose level resulting from fuel element withdrawal and radioactive materials ejection from the core has been studied. In this study, the dose rate resulting from the LOCA in spent-fuel storage pool of an open-pool reactor would be studied.

A multipurpose reactor (MPR) is an open pool-type reactor, of the power of 22 MW, has two pools; the main pool and the spent-fuel pool (SFP). The main pool contains the reactor core, the irradiation grid, a part of the core cooling circuit, a part of the pool cooling system, and the radial and tangential irradiation tubes. A transfer channel connects the two pools to transport the spent-fuel and radioactive materials from the main pool to the SFP underwater surface to avoid rising in radiation dose during the transport process. Spent nuclear fuel is highly radioactive, and water is efficient for both radiations shielding and cooling; hence, spent fuel is stored at the bottom of the SFP until it's inert enough to be moved to permanent spent-fuel storage. A shielding water of 4.6-m thickness is sufficient to conserve the radiation dose rate in the reactor building at level less than the permissible limit at the full capacity of SFP.[3] Decreasing in water level due to malfunction or accidents results in rising of the radiation dose rate in the reactor building. LOCA is one of the accidents may occur in nuclear reactors and results from breaking or rupture in cooling system pipes or neutrons irradiation tubes which leads to decreasing in water level in the reactor pools.[4] This would result in decreasing the water level in the SFP and consequently raising the radiation dose levels that would prevent access to the reactor hall and the control room. The dose rate is a function of water level above the storage racks, and the number of spent-fuel elements stored in the SFP during LOCA.

Monte Carlo N-Particle Transport (MCNP) 5 code[5] was used to calculate the radiation dose rates in the reactor hall and the control room resulting from LOCA in SFP. Then, it could be used to determine the water level heights which corresponding to the permissible dose rate limit above the SFP and in the control room.


  Description of Loss-of-Coolant Accidentaccident in Spent-Fuel Pool Top


SFP is a cylinder of 7.5-m height and 3.5-m diameter constructed of a stainless steel surrounded with heavy concrete to verify safe radiological conditions in radial direction. The water in the SFP is dematerialized water and provides radiological shielding in vertical direction for workers beside the storage pool and adjacent areas.[3]

The spent fuel is stored in stainless steel racks that are submerged with approximately height of 4.6 m of water above the top of the stored fuel. The SFP contains two storage racks, and each rack has rectangular arrangement of 12 × 22 positions. The first rack (lower rack) locates in the bottom of the pool, and when it reaches to the full capacity, the following spent-fuel elements would be stored in the second rack (upper rack) that located above the first one.

In MPR, loss of water inventory through connected systems can be caused by a leakage through pumps seals, heat exchanger cracks, failed welds, and cracks or breaks in the piping. In case of rupture in the pipes of SFP cooling system, the water surface in SFP will reach a level still verifying the radiological safety point of view because the inlet and outlet of the cooling system are located at levels near the top of the storage pool. The second pathway for loss of SFP coolant inventory is the leakage through temporary gates in the transfer channel between the main and storage pools as shown in [Figure 1]. Thus, during a loss of water in the main pool would drop the water level in the SFP in case of malfunction in temporary gate of the transfer channel.
Figure 1: Vertical section in the spent-fuel storage pool

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In case of loss of water reposition system, the water level in SFP would continue to decrease to level equivalent to the bottom level of the transfer channel (=80 cm above the upper rack).


  Spent-Fuel Element Inventory Top


The spent fuel removed from the core would be stored for 2 days in a basket beside the core and then would be transported to the SFP through transfer channel. ORIGEN2.1 code[6] was used to calculate the inventory for a 235U mass of 404.7 g per fuel element. A continuous irradiation during 275 days to a power of 0.759 MW was considered and burn up of 103350 MWD/TU was achieved which corresponds to the maximum extraction value. The inventory of the spent fuel was determined after 2 days of cooling after fuel element extraction from the core. All fuel elements in SFP were considered at the maximum inventory at the time of the accident for conservative calculations.

The photon source intensity was calculated after 2 days of cooling for conservative calculations. The photon spectrum was determined using the ORIGEN2.1 18-group photon energy structure as shown in [Table 1]. These photon source intensity data were then introduced into an MCNP5 model to calculate the dose rate in reactor hall and control room.
Table 1: Fuel element photon intensity with energy after 2 days of cooling

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The Monte Carlo N-particle transport 5 model

The MCNP model for SFP includes the racks filled with spent-fuel elements distributed in two levels immersed in water and surrounded with a cylindrical wall of stainless steel and heavy concrete in radial direction. In the storage pool, two racks of a rectangular arrangement of 12 × 22 positions were spent-fuel element can be loaded as shown in [Figure 1].

The dose rate was studied in the study for the water level ranged between 400 cm and 80 cm above the top of the upper rack. The minimum value of the water level is corresponding to the minimum height that water could be dropped due to LOCA. The maximum value is chosen depending on the dose rate level in the control room which verifying the permissible limit.

F5 tally was located at the center of the top of the storage pool and in the control room, as shown in [Figure 2], to determine the dose rate with conjunction with the dose function card.[7]
Figure 2: Top view of the location of the spent-fuel storage pool and the control room in the reactor hall

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The weight window technique was used as a variance reduction technique to decrease the time consumption in calculations. The relative errors for the dose rate point detectors often were <5%. Some relative errors of the dose rate calculations reach 10%, especially in case of dose rate calculations, in the control room with the water level height of more than 300 cm.


  Results and Discussion Top


The expected dose rate, at full-capacity storage during normal operation, must be determined to ensure the radiological safety of the spent-fuel storage design. During normal operation, the water surface is located at the level of 460 cm above the top of the second rack. It was found that the dose rate at the top of the SFP is 1.18E-8 Sv/h which considered lower than the permissible limit (10 μSv/h).[8]

[Figure 3] and [Figure 4] represent the dose rates above the SFP and in the control room during the water level drop from the normal level (460 cm above the rack) to level of 80 cm for two cases, namely both bottom and top rack filled and the second case only bottom rack filled. The dose rate above the SFP is mainly coming from the radiation attenuated by the water layer above the racks and also from the scattered radiations from the SFP walls. The dose rate in the control room is considered a scattered dose rate since there is no direct exposure between the operator located in the control room and the spent-fuel storage.
Figure 3: Dose rates at the top of the spent-fuel storage pool during the loss-of-coolant accident

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Figure 4: Dose rates in the control room during the loss-of-coolant accident

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In [Figure 3] and [Figure 4], the graph of the first case is divided into two regions with different decreasing rate; before and after the water level height of 80 cm. The difference in decreasing rate because the photons with energy <0.1 Mev [which representing 40% of the total photons intensity as shown in [Table 1] have high attenuation coefficients relative to those of energy more than 0.1 Mev, and hence, with increasing the water level height, the low-energy photons would attenuate rapidly and would diminish when the water level height would reach 80 cm. After 80 cm, the decreasing rate depends only on the photons of energy more than 0.1 Mev which have very low attenuation coefficients relative to those of energy <0.1 Mev.

It is noted that the graph of the second case (only bottom rack filled) has the same decreasing rate of the second region of the first case since it depends only on the photons of energy more than 0.1 Mev due to the water level height of more than 120 cm above the first rack.

[Figure 3] shows the dose rate would be less than the permissible limit at water level more than 385 cm above the top of the second rack for the first case and at water level more than 290 cm for the second case. It also shows that the maximum dose rate at the top of the SFP was 3330 and 0.678 Sv/h for the two cases, respectively, when the water level reaches the minimum water height above the upper rack (80 cm).

The most important point is determining the radiation dose rate in the control room where the operator on duty always might locate. [Figure 4] shows the dose rate in the control room as a function water level during LOCA. The dose rate would be less than the permissible limit for the case of full capacity of the lower rack (the second case) when the water level is more than 140 cm. In case of full capacity of the storage pool (the first case), the dose rate would be less than the permissible limit when the water level is more than 270 cm above the top of the racks.

It also shows that the maximum dose rate in the control room was 3.45 and 4.7E-4 Sv/h for the two cases, respectively, when the water level reaches the minimum water height above the upper rack (80 cm).


  Conclusions Top


Radiological consequence resulting from water level drop in the auxiliary pool due to LOCA was studied using MCNP code. The dose rates were conservatively calculated by assuming that all the spent-fuel elements stored recently in the storage pool. The calculations show that the dose rate above the pool would be 1.1E-8 Sv/h for normal operation. It was found that the minimum height that water level could drop after LOCA was 80 cm above the second rack of the spent-fuel elements. The maximum dose rate above the SFP and in the control room which corresponding to the minimum height of the water level after LOCA depending on the cases of SF storage as shown in the following as shown in [Table 2].
Table 2: Maximum dose rate above the SFP and in the control room for the two cases

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The dose rate in the control room would be lower than the permissible limit when the water level height above the racks is more than 270 cm and 140 cm for the first and the second cases, respectively. The dose rate above the SFP would be lower than the permissible limit when the water height above the racks is more than 385 cm and 290 cm for the first and the second cases, respectively.

Finally, all water level heights have been determined based on using recently stored fuel elements for conservative calculations. Hence, in the real case, the spent-fuel elements have been stored for a long time, and then, the water level heights would be lower than the conservative results.

Financial support and sponsorship

Nil.

Conflicts of interest

There are no conflicts of interest.



 
  References Top

1.
Abdelhady A. Dose rates from the accidental withdrawal of a fuel element from an open pool-type reactor. J Nucl Eng Radiat Sci 2018;4:021010-1.  Back to cited text no. 1
    
2.
Abdelhady A. Radiation dose distributions due to sudden ejection of cobalt device. Appl Radiat Isot 2016;115:208-11.  Back to cited text no. 2
    
3.
INVAP. FSAR, Final Safety Analysis Report of Open Pool Reactor. Argentina: INVAP S.E; 2003.  Back to cited text no. 3
    
4.
ECD Nuclear Energy Agency. Status Report on Spent Fuel Pools Under Loss-of-Cooling and Loss-of-Coolant Accident Conditions. Report NEA/CSNI/R 02 May, 2015. Paris, France: OECD Nuclear Energy Agency; 2015.p. 2.  Back to cited text no. 4
    
5.
X-5 MONTE CARLO TEAM, MCNP. A General Monte Carlo N-Particle Transport Code. Ver. 5, Vol. 1 & 2. University of California, USA: LANL; 2003.  Back to cited text no. 5
    
6.
Croff AG. User's Manual for the ORIGEN2 Computer Code. Technical Report No. ORNL/Tm-7175. Oak Ridge, TN: Oak Ridge National Laboratory; 1980.  Back to cited text no. 6
    
7.
American Nuclear Society. Neutron and Gamma-Ray Fluence-to-Dose Factors. ANSI/ANS-6.1.1-1991. La Grange Park, Illinois: American Nuclear Society, ANSI Standard; 1992.  Back to cited text no. 7
    
8.
International Commission on Radiation Protection ICRP60. Recommendations of the International Commission on Radiation Protection. Publication 60. Oxford, England: Pergamon Press, ICRP, 1991.  Back to cited text no. 8
    


    Figures

  [Figure 1], [Figure 2], [Figure 3], [Figure 4]
 
 
    Tables

  [Table 1], [Table 2]



 

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