1.
BRIEF DESCRIPTION OF NUCLEAR POWER REACTOR SYSTEMS AND PRIMARY COOLANT CHEMISTRY

1.1 BOILING WATER REACTOR (BWR)

The direct cycle boiling water reactor nuclear system (Figure 1–1) is a steam generating system consisting of a nuclear core and an internal structure assembled within a pressure vessel, auxiliary systems to accommodate the operational and safeguard requirements of the nuclear reactor, and necessary controls and instrumentation. High-purity water is circulated through the reactor core, serving as moderator and coolant. Saturated steam is produced in the reactor core, separated from recirculation water, dried in the top of the vessel, and directed to the steam turbine generator. The turbine employs a conventional regenerative cycle with condenser deaeration and condensate demineralization.

The reactor core, the source of nuclear heat, consists of fuel assemblies and control rods contained within the reactor vessel and cooled by the recirculating water system. A typical 1220 MWe BWR/6 core consists of 748 fuel assemblies and 177 control rods, forming a core array about 5 meters in diameter and 4.3 meters high. The power level is maintained or adjusted by positioning control rods up and down within the core. The BWR core power level is further adjusted by changing the recirculation flow rate without changing the control rods positions.

The boiling water reactor requires substantially lower primary coolant flow through the core than pressurized water reactors. The core flow of a BWR is the sum of the feedwater flow and the recirculation flow.

The function of the reactor water recirculation system is to circulate the required coolant through the reactor core. The system consists of two or more loops external to the reactor vessel, each loop containing a pump with a directly coupled water-cooled (air-water) motor, a flow control valve, and two shutoff valves.

High-performance jet pumps located within the reactor vessel are used in the BWR recirculation system. The jet pumps, which have no moving parts, provide a continuous internal circulation path for the total core coolant flow.



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Radiochemistry in Nuclear Power Reactors 1. BRIEF DESCRIPTION OF NUCLEAR POWER REACTOR SYSTEMS AND PRIMARY COOLANT CHEMISTRY 1.1 BOILING WATER REACTOR (BWR) The direct cycle boiling water reactor nuclear system (Figure 1–1) is a steam generating system consisting of a nuclear core and an internal structure assembled within a pressure vessel, auxiliary systems to accommodate the operational and safeguard requirements of the nuclear reactor, and necessary controls and instrumentation. High-purity water is circulated through the reactor core, serving as moderator and coolant. Saturated steam is produced in the reactor core, separated from recirculation water, dried in the top of the vessel, and directed to the steam turbine generator. The turbine employs a conventional regenerative cycle with condenser deaeration and condensate demineralization. The reactor core, the source of nuclear heat, consists of fuel assemblies and control rods contained within the reactor vessel and cooled by the recirculating water system. A typical 1220 MWe BWR/6 core consists of 748 fuel assemblies and 177 control rods, forming a core array about 5 meters in diameter and 4.3 meters high. The power level is maintained or adjusted by positioning control rods up and down within the core. The BWR core power level is further adjusted by changing the recirculation flow rate without changing the control rods positions. The boiling water reactor requires substantially lower primary coolant flow through the core than pressurized water reactors. The core flow of a BWR is the sum of the feedwater flow and the recirculation flow. The function of the reactor water recirculation system is to circulate the required coolant through the reactor core. The system consists of two or more loops external to the reactor vessel, each loop containing a pump with a directly coupled water-cooled (air-water) motor, a flow control valve, and two shutoff valves. High-performance jet pumps located within the reactor vessel are used in the BWR recirculation system. The jet pumps, which have no moving parts, provide a continuous internal circulation path for the total core coolant flow.

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Radiochemistry in Nuclear Power Reactors Figure 1–1. Direct Cycle Boiling Water Reactor System With Forward-Pumped Heater Drains

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Radiochemistry in Nuclear Power Reactors The BWR operates at a constant pressure of 1050 psi (or 68 kg/cm2) and maintains a constant steam pressure, similar to most fossil boilers. The coolant temperature is maintained at approximately 280°C. A small portion of the recirculation water is removed from the reactor through the reactor recirculation pump suction line, passed through the heat exchangers and a reactor water cleanup (RWCU) system, and returned through the feedwater line. The purpose of the RWCU system is to maintain high reactor water quality by removing soluble and insoluble impurities, including corrosion products and radioactive species. In addition, the system provides a means for water removal from the primary system during periods of increasing water volume. A summary of BWR water quality specifications is presented in Table 1–1. The main objectives for water quality control in a BWR system are to: Minimize the potential for stress corrosion cracking of structural materials. Minimize corrosion and corrosion product release from the primary system surfaces. Minimize fuel cladding failures due to zircaloy corrosion. Minimize corrosion product deposition and activation on the fuel cladding surfaces. Minimize the radiation field buildup on the system component surfaces due to deposition of activated corrosion products. More detailed discussion on the BWR water chemistry quality control and guidelines can be found elsewhere(1). As a result of water radiolysis-gas stripping in the core and recirculation, the reactor recirculation water contains dissolved oxygen and hydrogen peroxide in the concentration range from ~100 to ~300 ppb and somewhat less than stoichiometric concentration of dissolved hydrogen. Under this highly oxidizing environment, the radioactive impurities are normally found in the higher oxidation states in the coolant. Recently, hydrogen water chemistry (HWC) was developed to mitigate the intergranular stress corrosion on structure materials. In this technology, hydrogen gas is injected into the coolant through the feedwater system to suppress the water radiolysis in the core region and to control the dissolved oxygen concentration in the coolant at a very low level. More discussion on HWC is presented in Chapter 6.

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Radiochemistry in Nuclear Power Reactors Table 1–1 BWR WATER QUALITY SPECIFICATIONS   Typical Warranty Limits Operational Practices   Normal Operating Limits Maximum Operating limits Suggested* Administrative Limits • Reactor Water     • Power Operation       – Conductivity (μS/cm) 1.0 10.0 0.2     – Chloride (ppb) 200 500 20     – pH 5.6–8.6 4.6–9.6 6.1–8.1     – Silica (ppb) 200 — 100     – Total Copper (ppb) 20 — 10   • Startup & Hot Standby       – Conductivity (μS/cm) 1.0 2.0 0.2     – Chloride (ppb) 100 100 20     – Total Copper (ppb) 20 — 10   • Shutdown       – Conductivity (μs/cm) 2.0 5.0 2.0     – Chloride (ppb) 100 500 50     – pH 5.3–8.6 4.6–9.6 5.3–8.6     – Silica (ppb) 200 — 100     – Total Copper 20 — 10 • Feedwater     • Power Operation       – Metallic Impurities (ppb) 15 60 —     —Iron (ppb)         • Insoluble 10 40 2.0       • Soluble 1.0 2.0 0.5     – Total Copper (ppb) 0.5 2.0 0.1     – Oxygen (ppb) 35±15 110±90 25±5     – Conductivity (μs/cm) 0.005 0.1 0.06   • Startup and Hot Standby       – Conductivity (μs/cm) 0.1 0.15 0.08     – Total Copper (ppb) 1.0 — 0.2 • Condensate Treatment System Effluent     • Oxygen (ppb) 35±15 110±90 25±5   • Conductivity (μS/cm) 0.065 0.1 0.06   • Iron (ppb)       – Insoluble 10. — 2.0     – Soluble 1.0 — 0.5   • Total Copper (ppb) 0.5 — 0.1 *J.M.Skarpelas, GE Nuclear Energy, Private Communication (1990)

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Radiochemistry in Nuclear Power Reactors 1.2 PRESSURIZED WATER REACTOR (PWR) The primary side of a PWR system consists of a series of interrelated systems which directly or indirectly interface with the reactor itself. Each of the associated systems has its own specific functions which contribute to the fundamental safe operation and control of the reactor. The major component of the primary side is the reactor vessel, which houses the reactor core. The reactor is operated at 2200 psi (or 142.5 kg/cm2) with the coolant temperature at 350°C. Unlike the BWR system, there is no boiling in the core region. Associated with the reactor vessel is a piping system through which the reactor coolant is pumped. The energy is transferred by the coolant from the reactor core to the steam generators where secondary steam is subsequently produced and routed to the unit turbine generator and main steam system. The primary coolant system also includes the reactor, an electrically heated pressurizer, a pressurizer quench tank and inter-connecting piping, and the chemical and volume control system (Figure 1–2). The pressurizer maintains primary coolant system operating pressure and compensates for changes in primary coolant volume during load changes. The quench tank is designed to receive and condense the normal discharges from the pressurizer relief valves and prevent the discharge from being released to containment. The chemical and volume control system is designed to allow the operators to control the volume of primary coolant as well as its chemical composition through a dual interface with the primary coolant system. The major use of this system is to control the primary coolant boron concentration as a function of power level and core life. It is also used to control the reactor coolant pH through the addition or removal of lithium hydroxide. Some fission and activation products can be removed by the system’s purification demineralizers. The system is designed to allow the addition or removal of boron in the form of boric acid, lithium hydroxide and hydrogen during normal operation. Hydrogen gas is dissolved in the reactor coolant to scavenge any trace amounts of dissolved oxygen which may be present in the coolant.

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Radiochemistry in Nuclear Power Reactors Figure 1–2. Schematic of a Pressurized Reactor System

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Radiochemistry in Nuclear Power Reactors The function of the primary coolant is to remove heat from the core, moderate the core, and transfer the heat to the steam generators. In order to perform these functions efficiently, chemistry controls are required. The purpose of these controls is to: Minimize the number of control rod movements by use of chemical shim (i.e., boron) for controlling core reactivity throughout the life of the core. Minimize corrosion of primary system surfaces including the steam generator tubing. Minimize Zircaloy corrosion and fuel cladding failures. Minimize corrosion product deposition and activation on the fuel cladding surfaces. Minimize corrosion product buildup on the steam generator which could reduce heat transfer. Minimize total activated corrosion product inventory and transport and plant radiation field buildup. The vendor reactor coolant chemistry specifications may vary slightly from each other as shown in Table 1–2. Current Westinghouse specifications for the reactor coolant system are given in Table 1–3. Under the highly reducing chemistry environment, most of the radioactive impurities are expected to be present in the reduced or insoluble forms in the PWR coolant. More detailed discussion on the principals of PWR coolant chemistry control and guidelines can be found in References 3 and 5. 1.3 REFERENCES (1) BWR Owners Group, “BWR Water Chemistry Guidelines”, EPRI NP-3589-SR-LD, Special Report (April 1985). (2) A.Strasser et al, “Corrosion-Product Buildup on LWR Fuel Rods”, EPRI NP-3789 (April 1985). (3) PWR Primary Water Chemistry Guidelines Committee, “PWR Primary Water Chemistry Guidelines: Revision 2”, EPRI NP-7070 Final Report (November 1990). (4) S.Glasstone and A.Sesonske, “Nuclear Reactor Engineering,” D.Van Nostrand Company, Inc., Princeton, New Jersey (1963). (5) P.Cohen, “Water Coolant Technology of Power Reactors”, Gordon and Breach Science Publishers, New York (1969).

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Radiochemistry in Nuclear Power Reactors Table 1–2 VENDOR REACTOR COOLANT CHEMISTRY SPECIFICATIONS FOR POWER OPERATION (Reproduced with Permission, EPRI NP-3789, Ref. 2) Vendor 7LiOH (as ppm Li) Boron* (ppm) Hydrogen (cc/kg, STP) Westinghouse (W) 0.7–2.2 (formerly 0.2–2.2) 0–800 (0–1100 first cycle) 25–50 (normal operating range 30–40) Combustion Engineering (CE) 1.0–2.0** (formerly 0.2–1.0) <4400 25–50 (formerly 10–50) Babcock and Wilcox (B&W) 0.2–2.0 0–2270 15–40 *Concentration varies from high levels early in cycle to low levels at end of cycle. **Near end of life, when the deborating ion exchanger is placed in service (≈30 ppm B), the lithium range is 0.2–0.5 ppm.

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Radiochemistry in Nuclear Power Reactors Table 1–3 CURRENT WESTINGHOUSE SPECIFICATIONS FOR THE REACTOR COOLANT SYSTEM (Reproduced with Permission, EPRI NP-3789, Ref. 2) Chemistry Parameter Permissible Range Electrical Conductivity 1–40 μmho/cm pH 4.2–10.5 Dissolved Oxygen ≤5 ppb Chloride ≤0.15 ppm Fluoride ≤0.15 ppm Suspended Solids ≤1.0 ppm Hydrogen 25–50 cc/kg Boron 0–4000 ppm Lithium 0.7 to 2.2 ppm (recommendation depending on boron concentration) Silica ≤0.2 ppm Calcium ≤0.05 ppm Magnesium ≤0.05 ppm Aluminum ≤0.05 ppm

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