Safety Review Models on Radioactive Source Term Design For PWR Waste Treatment Systems

Transcription

1 Safety Review Models on Radioactive Source Term Design For PWR Waste Treatment Systems Xu Mingxia Waste Management Division, Nuclear Safety center, State Environmental Protection Administration, Beijing China The source of all liquid, gaseous and solid radioactive waste in the pressure water reactor (PWR) nuclear power plant (NPP) originate from leakage of fission products out of the fuel rods into the primary coolant and neutron activation of materials within and around the primary coolant system and reactor vessel. The source term design used to determine the concentrations of radionuclides in the reactor coolant, which could be: (1) A conservative source term which predicts the maximum concentration of nuclides in the Reactor Coolant System establishes the design basis of the various onsite processing systems for the purpose of defining system capacity and shielding requirements and (2) A realistic source term for the purpose of evaluating the reasonably expected inventories and releases of radionuclides under normal operating condition, including anticipated operational occurrences. This paper will discuss the safety review models on source term design for the PWR waste system mainly on the basis of the conception of the conservative source term. 1. Design Basis of Fission Product Activities The Mathematical method used to determine the maximum concentration of nuclides in the reactor coolant system involves a group of linear, first order differential equations. These equations are obtained by applying a mass balance for production and removal for the fuel pellet region as well as the coolant region. In the coolant region, the analysis includes the fission product production by escape from the fuel through defection fuel rod cladding, parent decay in the coolant, and neutron activation of coolant fission products. Removal is by decay, by coolant purification, by feed and bleed operations (for fuel burn-up), by leakage, and other feed and bleed operations such as startups and shutdowns, as well as load follow operation. In these calculations, the cladding defects in the equivalent of 1 percent of the fuel rods are assumed to be present at initial core loading and uniformly distributed throughout the core. Similar defects are assumed to be present in all reload regions. The fission product escape rate coefficients are, therefore, base on an average fuel temperature. The fission product activity in the reactor coolant during operation with defects in the cladding of the fuel rods is computed using the following differential equations: For parent nuclide in the coolant, dn ci /dt = R i N Fi /M c - [λ i +D i + Q L /M c (Ψ i + DF i - 1)/DF i ] N ci (1) For daughter nuclides in the coolant, dn cj /dt = R j N Fj /M c + f i λ i N ci - [λ j + D j + (Q l /M c ) ( Ψ j + DF j -1)/DF j ] N cj (2) Where: N c - Concentration of nuclide in the reactor coolant (atoms/gram) N F - Population of nuclide in the fuel (atoms) T - Operating time (seconds) R - Nuclide release coefficient (1/sec.) = F υ F - Fraction of fuel rods with defective cladding υ - Nuclide escape rate coefficient (1/sec.) M c - Mass of reactor coolant (grams) λ - Nuclide decay constant (1/sec.) D - Dilution coefficient by feed and bleeds = β/(b 0 -β t ) 1/DF (1/sec.) B 0 - Initial boron concentration (ppm) β - Boron concentration reduction rate (ppm/sec.) DF - Nuclide demineralizer decontamination factor Q L - Purification or letdown mass flow rate (grams/sec.) ψ - Nuclide volume control tank stripping fraction f - Fraction of parent nuclide decay events that result in the formation of the daughter nuclide Subscript i refers to the parent nuclide. 1

2 Subscript j refers to the daughter nuclide. The parameters used in the calculation of the reactor coolant fission product concentrations, including pertinent information concerning the expected coolant cleanup flow rate, demineralized effectiveness, and volume control tank noble gas stripping behavior etc., are presented in the design documents or reference books. All the parameters could be found easily, but the F (the fraction of fuel rod with defective cladding) in the design of PWR NPP did not meet the requirement 1% cladding defect in standard review plant (SRP) nowadays. Therefore, the F will be discussed as below. 2. Selection of the parameter F The system capability to process wastes should be at design basis fission product leakage level, i.e., from 1% of the fuel producing power in a PWR, so as to meet the requirement of Standard Review Plan (SRP). Nowadays thanks to the improvement of the quality of the fuel assembly, the damage percentage of the fuel was greatly reduced. The actual damage percentage of the fuel assembly was more less than 0.1%from the feed back of the experience of operational PWRs in the world. Facing the reality, most of the PWR designers, taking the advantage of the fuel quality improvement, use the damage percentage of less than 1%instead of 1% in the source term design. Table 1. Selection of failed fuel fraction for design basis in PWRs Country United States China(QinShan1) France* Russia Failed fuel Fraction % ~ * 37GBq I-131 equivalent, which equals about 0.25% fuel damage was adopted as design basis in France. Fuel failed fraction 1% is taken as PWR design basis by United States designer in Now PWR does not be constructed any more in the United States. No doubt, above design considerations (except in U.S.) in deferent PWRs could meet the demand of the waste treatment capacity during reactor normal operating condition and anticipated operational occurrence. But when the accident happened on the condition of the 1 % fuel damage, those systems may not meet the requirement of SRP. Therefore, two safety review models were established. 3. Safety Review Models 3.1 Model I Considering the short period feature of the accident, even on the condition of 1% fuel damaged, the waste systems which designed based on less than 1% fuel defect could also function properly an radioactive inventory in discharge effluent may keep within the regulation limits. If may not, it is necessary to modify the systems or to take proper measurements. 2

3 An example analysis: The fuel defect 0.2% of gas untightness and fuel 0.02% contact with coolant were used in calculation of design basis of fission products in 1000 MWe PWR. In this NPP, fission product migration in primary circuit is given in Fig 1. Axial hole volume 0.2% of gas untightness Diametrical clearance, joints of tablet Fuel 0.02% contact with coolant Internal surface of fuel rod Decontamination in volume and boron control system Primar y coolant Purification system Surface contamination of fuel rods Primary Circuit debalance water Fig 1. Fission products migration in primary circuit The sources of primary coolant fission products during power operation of the unit are considered as followings: - Surface contamination of external shells of fuel rods; In the initial period of reactor operation, the level of coolant contamination by fission products is determined by discharge to the circuit at the consume of kinetic energy of fission fragments of uranium-235 which is present on external surfaces of fuel rods as contamination, originating from their manufacturing, and in an impurity of natural uranium in zirconium shells. -Defective fuel rods with gas untightness and significant damages; At normal operation conditions of the reactor tightness of fuel rod shells can be broken owing to different processes of corrosion-fatigue type. This results in micro-cracks first, and then large defects in shells, which are followed by increase in fission products discharge from fuel rods to the primary coolant. The following mechanisms of fission product discharge from uranium dioxide to fuel rod gas clearance are used: - Discharge at the consume of kinetic energy of fragment during uranium-235 fission; - Discharge at the consume of knocking-out of fission products from fuel surface by taking off fission fragments; - Discharge of fission products from structural changes area During operation of the reactor, which has untight fuel rods in its core, increase of radioactivity of some fission products in the primary coolant after decrease or increase in reactor power is frequently observed. This increase of radioactivity (spikeeffect) caused by an additional discharge of gaseous and volatile fission products, accumulated in gas clearance of untight fuel rods, is taken into account for transient condition of the power unit. Solution balance in the primary circuit realized by their compute code. 3

4 Fuel rod axial hole volume 0.2% of gas untightness, which was considered to cause nuclides Kr, Xe, I, and Cs release to primary coolant, and fuel pellet 0.02% contact with coolant, which mainly brings Sr, Rb, Te, and Tc release, were used in the calculation. It is obvious that the calculation results of fission product activity in primary coolant depending on above design parameters are much less than that depending on 1% fuel damaged. These source terms for design basis of waste treatment system can not be met the requirement of SRP. Therefore, the demonstration should be done by recalculation. Designer provides the recalculation results of the nuclide inventory of the annually discharged effluent in Table 2. In the recalculation, the following assumption were made: 1) The inventory of the primary coolant source terms comes from 1% fuel defect; 2) Before discharge, the wastes have been treated. Table 2. Radioactive Effluent Discharged per year Bq/GW (e) Nuclide Depending on 0.2% fuel Depending on 1%fuel State regulation limit defect defect Noble gases E14 1.0E15 Iodine 2.8E Aerosol E9 The conclusion is successful from this table due to the value in column 3 within the state regulation limit. Inspire of the values in column 3 are more higher than that in column 2, the design of waste treatment system are acceptable because the reactor should be shut down at that accident on 1% fuel damage condition and the accident release can be processed and recovered in a short time. 3.2 Mode II Since the waste system possess the large redundancy feature when the accident happened and 1% fuel was damaged all the radioactive wastes could be stored in system tanks temporally, let them decay and then treat them gradually. The treated effluent could not be discharged to environment till the design objectives for equipment necessary to control release of radioactive effluents have been met. If the existing system capacity could not meet the requirement mentioned above, the system should be modified or takes proper provision. An Example Analysis: Failed fuel 0.5% used on calculation of design basis of fission product activities in 300 MWe PWR. Reactor coolant volume is 147 m 3. The calculation method and mechanism of fission product release are the same as in the section 1. Main equipment of liquid waste processing system and gaseous waste processing system are presented in Table 3 and Table 4. Main Equipment Evaporator Table 3. Main equipment in liquid waste processing system T 1 T 1 T 1 T 2 collection Storage Condensate Collection tank tank tank tank 4 T 2 Condensate tank Discharg e tank Volume 2 T/h 37 m m 3 20 m m 3 24 m 3 24 m 3 Number *T1: The drain and leakage from CVCS, BRS etc. *T2: The drain from resin regeneration, component washing etc. Normally, it takes 2-3 months for processing liquid waste by an evaporator during normal operation condition per year. When the accident happened at 1% fuel failed the fission product activities are increased rapidly in primary coolant, and reactor should be shut down at that time. It is consume that all 147 m 3 of primary coolant is as liquid waste and could be transferred to T1 collection tanks, and then be treated gradually. It is obviously from Table 3 that these tanks have enough volume for storage of all the liquid waste. Table 4. Main components in gases waste treatment system Decay tank Compressor Number 6 Number 2 Volume m 3 21 Capacity m 3 /h 40 Pressure Mpa 0.7 Operation pressure Suction Mpa Discharge Mpa 2.94E-2 to 2.14E-3 0 to Two waste gases compressors are provided to circulate gases around the system loop. One is normally used

5 with the other on a standby basis. Gaseous wastes discharged to the vent heater are pumped into decay tank by compressor. The tank in service is pressurized to 0.7Mpa and storage in 60 days, and then gas effluents are released through a high stack during normal operation. When the accident happened on 1% fuel rod damaged, the decay time for nuclides (mainly Xe-133) may need increasing to 2 times of decay time to assure that the radio-activities in gas effluent discharged to atmosphere keep the same as that discharged in normal operational condition. At that time the capacity of this gasses waste treatment system is 2402m 3. The volume of gas waste produced in plant annually during normal operation condition is less than 2000m 3. Therefore, the gases waste treatment system is acceptable. These two safety review models have been adopted successfully in the review of waste treatment system in China s PWRs. The radioactive inventory in effluent from the operation of PWR is shown in Table 5. Complying with the evaluation findings in the review, the radioactive level of the gaseous effluent discharged annually was lower than 10% of the discharge limit defined in State Standard; the liquid effluent lower than 10% (except tritium) from the operational PWRs in China. Gases effluents Bq/a Table 5. Radioactive inventory in effluent discharged annually in China PWR NPP Qinshan Nuclear power plant Da YA Bay State regulation limit Year Bq/ GW(e).a Noble gases E15 Iodin * * * and particle Liquid effluents Tritium 4.82 Bq/a others 4.08 * It can not be detected E11 Reference (1) U.S. Nuclear Regulatory commission Standard Review Plan NUREG-0800 Rev.3 April (2) China State Standard Authorized Limits for Normalized Releases of Radioactive effluents from Nuclear Fuel Cycle GB (3) China State Standard Radioactive Source Term of PWR Nuclear Power Plant for Operational Conditions GB/T (4) Source Term Data for Westinghouse Pressurized Water Reactors WCAP-8253 July (5) AFCEN System Design and Construction Rules for 900 MWe PWR NPP in France RCC-P (6) Nuclear Safety Rules for Reactor Units of Nuclear Power Plants 1989 Russia. (7) Safety Analysis Report for QinShan Nuclear Power Plant Rev.2 June