8.
SPECIAL RADIOCHEMICAL STUDIES

8.1 ESTIMATION OF NOBLE GAS TRANSIT TIME IN THE BWR TURBINE SYSTEM

The decay of the short-lived noble gases and buildup of their daughter product activities in the steam and condensate are unique features of the BWR system. The noble gas activities, which are exhausted from the Steam Jet Air Ejector (SJAE), decay during the transit in the steam path, and they and their daughter activities follow many separate paths through the low pressure turbine and the heater drain system, while ≈1% of the high-pressure steam is used to drive the SJAEs. The daughter activities will either be washed out by the condensed steam or deposited on the turbine and heater surfaces. In order to estimate the parent decay and daughter buildup in the steam paths in the turbine, the transit time as well as the activity release rate must be known. It is estimated that a major portion (≈60%) of the gaseous activities passes the turbine in a few seconds, but the “apparent” transit time, defined for the gaseous parents by the equation ti=(1/λi)ln(Aoi/Ai)*, may vary from several seconds to a few minutes, depending on the half-life of the individual gaseous isotopes.

The “apparent” transit time, T, can be calculated from the parent-daughter pair activities:

where

A2

=

daughter activity transport rate (μCi/s)

A1

=

parent activity release rate (μCi/s) measured at the same SJAE sample point

*  

e.g., in more familiar form, . In case of multiple pathes,

where Fj is the fraction of Aoi passing through the steam path j.



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Radiochemistry in Nuclear Power Reactors 8. SPECIAL RADIOCHEMICAL STUDIES 8.1 ESTIMATION OF NOBLE GAS TRANSIT TIME IN THE BWR TURBINE SYSTEM The decay of the short-lived noble gases and buildup of their daughter product activities in the steam and condensate are unique features of the BWR system. The noble gas activities, which are exhausted from the Steam Jet Air Ejector (SJAE), decay during the transit in the steam path, and they and their daughter activities follow many separate paths through the low pressure turbine and the heater drain system, while ≈1% of the high-pressure steam is used to drive the SJAEs. The daughter activities will either be washed out by the condensed steam or deposited on the turbine and heater surfaces. In order to estimate the parent decay and daughter buildup in the steam paths in the turbine, the transit time as well as the activity release rate must be known. It is estimated that a major portion (≈60%) of the gaseous activities passes the turbine in a few seconds, but the “apparent” transit time, defined for the gaseous parents by the equation ti=(1/λi)ln(Aoi/Ai)*, may vary from several seconds to a few minutes, depending on the half-life of the individual gaseous isotopes. The “apparent” transit time, T, can be calculated from the parent-daughter pair activities: where A2 = daughter activity transport rate (μCi/s) A1 = parent activity release rate (μCi/s) measured at the same SJAE sample point *   e.g., in more familiar form, . In case of multiple pathes, where Fj is the fraction of Aoi passing through the steam path j.

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Radiochemistry in Nuclear Power Reactors A smooth curve of T vs. the half-life of the parent nuclide from the data obtained in an operating BWR is shown in Figure 8–1. Qualitatively, the curve would be expected to flatten out for short half-lives as a limiting point is reached where only a single flow delay is significant, and all other delays are so long relative to the half-life that they do not contribute to the release. The apparent transit time for other gaseous activities can be obtained from this curve. If the transit time from the outlet of the reactor pressure vessel (RPV) to the SJAE is known, the gaseous activity release rate at the RPV can be estimated. Similarly, using the parent-daughter relationship, the delay time in the condensate (hot-well) can be estimated from the noble gas daughter and granddaughter activities. In the reactor coolant, the delay time in the sample line or any specific system can also be estimated from the parent-daughter pair activities. Some major parent-daughter pairs and their decay constants have been summarized in Table 2–1. Figure 8–1. Apparent Transit Time for Gaseous Activities from RPV to SJAE

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Radiochemistry in Nuclear Power Reactors 8.2 NO-CLEANUP TEST IN A BWR A no-cleanup test is generally referred to as a test of measuring the water chemistry and the distribution of fission products in the primary coolant system when the RWCU system is removed from service. During a no-cleanup test, a number of practical operation parameters can be measured, such as the mass of reactor coolant, water cleanup half-time, iodine carryover, impurity input rate, etc. A typical example of no-cleanup test data is shown in Figure 8–2. In this test, the water conductivity was allowed to increase to 2.4 μS/cm before the cleanup system was returned to service. By measuring the decrease in water conductivity after return to service, the cleanup half-time can be estimated (Figure 8–3). Also, by knowing the water cleanup flow rate and efficiency, the apparent reactor coolant mass may be calculated (it should not be a surprise that the measured reactor coolant mass during operation may not be consistent with that estimated from the design drawing). The radioactivity buildup in the coolant may be easily calculated. Some examples of calculations are compared with the measured data below. 8.2.1 Na-24 and Cl-38 Activity Buildup The total active impurity at equilibrium condition during normal operation has been shown in Section 2.7, Equation 2–17: (8–1) During the time of no-cleanup service, the production of active nuclides Nnc can be described by Equation 8–2: (8–2)

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Radiochemistry in Nuclear Power Reactors Figure 8–2. Variation of Conductivity and pH Value in Reactor Water During a No-Cleanup Test

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Radiochemistry in Nuclear Power Reactors Figure 8–3. Variation of Reactor Water Conductivity with RWCU Returning to Service

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Radiochemistry in Nuclear Power Reactors By replacing Equation (8–2) can be integrated to become (8–3) Replacing from Equation (8–1), (8–4) The total activity in the reactor water at time t after the cleanup service is turned off, can be calculated by the summation of Equation 8–4 and the following two items: — active nuclide produced during normal operation — activation of parent nuclide input to reactor water during normal operation. (8–5) Equation 8–5 predicts that the concentrations of Na-24 and Cl-38 activities should increase with time after the RWCU system is out of service for a period of a few half-lives. The observed increases of Na-24 and Cl-38 activity concentrations are shown in Figure 8–4. The data are also compared with the calculated concentrations from Equation 8–5.

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Radiochemistry in Nuclear Power Reactors Figure 8–4. Variation of Na-24 and Cl-38 Concentrations in Reactor Water during a No-Cleanup Test

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Radiochemistry in Nuclear Power Reactors 8.2.2 Measurement of Iodine Steam Transport The magnitude of iodine carryover can be measured directly from the ratio of iodine concentration in the condensate to that in the reactor water. However, in a reactor with forward-pumped heater drains, the iodine carryover cannot be measured directly from the condensate because ~75% of the iodine originally carried over by steam is condensed with steam in the high pressure turbine and forward-pumped back to the reactor (see Section 3.4, Figure 3–5). Another technique for measuring the iodine carryover is to determine the steady-state concentration of iodine in reactor water during normal steady power operation with and without the RWCU system in service. When the RWCU system is removed from the service βc=0 in Equation 3–23, the iodine concentration will increase and reach a higher steady state concentration, Cnc (Figure 8–5), (8–6) From the ratio of (8–7) βs and the iodine carryover ε can be calculated, where all symbols have been defined in Section 3.2.3.

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Radiochemistry in Nuclear Power Reactors Figure 8–5. Variation of Iodine Activity Concentrations in Reactor Water during a No-Cleanup Test

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Radiochemistry in Nuclear Power Reactors 8.3 RADIOCHEMISTRY OF IODINE 8.3.1 Chemical Forms of Radioiodine in PWR Coolant The PWR coolant chemistry is more complex than BWR chemistry, particularly during reactor shutdown. During normal operation, the coolant is maintained at pH~9 (25°C) and under reducing conditions with hydrogen. Boration of the coolant during shutdown brings the water pH to ~6, and in some reactors, H2O2 is added to oxidize the corrosion film in order to remove the cobalt activities from the primary system surfaces (Section 4.3.2.) The behavior of iodine activities in the PWR coolant has been studied recently by Voilleque(1). Most of the iodine activities are found in the reduced form (iodide), as expected, during normal operation, but immediately after shutdown, due to boration and H2O2 addition, volatile forms of iodine (mainly I2) have been found to vary from a few percent to approximately 40%. A few percent of organic iodide are generally found after shutdown. A few percent of iodate are also observed after shutdown, but in some cases, up to 70% of iodine activities have been reported when H2O2 is added to the coolant. 8.3.2 Chemical Forms of Radioiodine in BWR Coolant* The chemical forms of radioactive iodine in reactor water have been measured in a number of reactors during normal operation and during shutdown.(2) As shown in Table 8–1, during reactor operation, 60–90% of the iodine in reactor water was found as iodide (I−) and perhaps HIO, and the remainder was essentially iodate . Only traces of I2 and organic iodine were detected and these traces were most likely cross contamination from the other fractions in analytical procedures (see Figure 8–6). *   More recently, the chemical behavior of radioiodine in BWR coolant under hydrogen water chemistry conditions has been studied by Lin and the results have been published elsewhere.(16)

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Radiochemistry in Nuclear Power Reactors Table 8–1 CHEMICAL FORMS OF RADIOIODINE IN BWR PRIMARY COOLANT (%) Reactor (Date)   I− I2 Organic During Normal Operation   Millstone-1 (7/1972) 80 20 <1 <1 Humbolt (12/1972) 78 10 8 4 Nine Mile Point-1 (12/1972) 65 34 1 <1 Nine Mile Point-1 (4/1973) 52 48 <1 <1 Nine Mile Point-1 (7/1974) 79 18 3 <1 Monticello (7/1974) 77 23 <1 <1 Monticello (9/1975) 84 12 4 <1 Brunswick-2 (5/1977) 67 33 <1 <1 FitzPatrick (8/1985) 90 10 <1 <1 FitzPatrick (HWC) (8/1985) 99 1 <1 <1 Hatch-1 (7/1986) 83 17 <1 <1 Hatch-1 (HWC) (7/1986) 90 10 <1 <1 Nine Mile Point-1 (5/1987) 40 60 <1 <1 Nine Mile Point-1 (HWC) (5/1987) 95 5 <1 <1 During Reactor Shutdown   Nine Mile Point-1 (4/1973) 2 98 <1 <1 Nine Mile Point-1 (3/1974) 4 96 <1 <1 Monticello (9/1975) 63 30 3 <1 Monticello (2/1976) 18 82 <1 <1 Note: The distribution of I− and was found to vary from plant to plant, most likely depending on the metallic ion concentration, particularly Cu ions, in reactor water. Similar results of measurements have been reported in the literature(5,6,7)

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Radiochemistry in Nuclear Power Reactors Figure 8–7. Variation of Cm-242 to Zr-95 Activity Ratio in Coolant as a Function of Average Core Burnup

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Radiochemistry in Nuclear Power Reactors 8.5 APPLICATION OF Na-24 TRACER IN FLOW MEASUREMENTS The measurement of water flow by a radioisotope dilution technique is made by injecting a radioactive tracer solution into the flowing process stream and, after subsequent mixing, sampling the process stream to determine the concentration of tracer in the stream. By knowing the tracer injection rate and the concentrations in the reference solution and the process stream, the flow rate of the process stream can be easily determined by the following equation: (8–8) where F = water flow rate being determined f = injection rate of reference solution c = tracer concentration in reference solution C = tracer concentration in process stream Application of this technique has been widely accepted in the nuclear industry, particularly using the Na-24 tracer in feedwater flow measurements. It has been reported by Holloway and Gilbert(14) that the accuracy of ±0.2% relative to a calibrated reference flow nozzle (by ASME) can be obtained. The most obvious advantage of using the Na-24 tracer is the accuracy and high sensitivity of the gamma-ray spectrometric measurement of the Na-24 activity. In order to achieve a high precision and accuracy of flow measurement, there are a number of parameters that should be carefully determined during a measurement. These parameters are: Activity concentrations in the reference solution and water samples. Activity counting time, decay correction and counting geometry correction. Sample time and sample volume (mass). Injection time and injection rate. The combined uncertainty of these parameters should be less than 0.2% in a single measurement.

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Radiochemistry in Nuclear Power Reactors The only drawback of using the Na-24 tracer is that a heavy shielding is often required to protect the technicians from radiation exposure, particularly during the preparation of highly concentrated tracer solution. A typical procedure of the flow measurement involves the following steps: Preparation of the Na-24 Reference Solution The reference solution is prepared by dissolving an irradiated Na2CO3 or NaNO3 containing ~200 mCi Na-24 in an appropriate container with dilute acid and demineralized water (total volume is about 2 liters). After thoroughly mixing, the solution is transferred to two clean 1-L polyethylene bottles. Determination of the Na-24 Concentration in Reference Solution The Na-24 concentration is accurately measured. In order to minimize the counting geometric correction error, the activity sample for counting is prepared identically to the sample collected from the feedwater. Multiple samples should be prepared to obtain the best result. The typical Na-24 concentration in the reference solution is about 2×106 cpm/g. Injection of the Na-24 Reference Solution The Na-24 reference solution, contained in an 1-L polyethylene bottle, is placed on the platform of a top-loading analytical balance. The solution is injected at a constant rate (approximately 10 g/min) with a minipump into the feedwater upstream of the feedwater pump. The injection rate is determined by determining the weight loss of the solution bottle at predetermined time intervals, generally every 5 minutes. A typical sampling period is ~30 minutes, and total duration of the injection is normally 2 hours for one test.

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Radiochemistry in Nuclear Power Reactors Sample Collection The sample is collected at the downstream of the feedwater pump, normally at the regular feedwater sampling station. The water containing the diluted Na-24 activity is passed through two sets of three cation exchange membranes to quantitatively collect the Na-24 in water. Approximately 10 kg of water is collected in a 30 minute period. The exact mass of sample processed through the filter poles is determined by collecting the filter effluent in tared containers and reweighing. The filter holder is a Millipore high pressure filter holder similar to that used in the regular corrosion product sample solution. Radioactivity Measurement The Na-24 activity collected on cation membranes is measured with a Ge(Li) detector interfaced to a multichannel gamma-ray spectrometer and a single-channel analyzer. The single-channel analyzer is used to simultaneously measure the gross gamma counting rate detected above the energy of 1.7 MeV. The multichannel analyzer is used to measure the activity of the Na-24 at different photopeaks and other potential background activities in the feedwater. All samples collected from the feedwater are not counted until about 5 hours after sampling to allow the Cs-138 activity which may be present in the sample to decay away. Generally, the background is found to be less than 0.1% of the sample activity. The gross counting rate is generally used in the flow rate calculation because of its smaller uncertainty in counting statistic. Calculation The feedwater flow rate is calculated by Equation 8–8. It is important to include all the possible uncertainties in each parameter, as described earlier, in the overall estimate of uncertainty.

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Radiochemistry in Nuclear Power Reactors 8.6 IDENTIFICATION OF DEFECTIVE FUEL* In light-water reactors, there are tens of thousands of fuel rods in a reactor. The individual fuel rods and the fuel assemblies are fabricated with exceptional care, to preclude the escape of fission products through even minor cladding defects. However, with such a large number of fuel rods it is possible for cladding failure to develop during an operating cycle. Failure of fuel cladding is indicated when fission-product activity in reactor cooling water rises above normal background levels. (Background radiation is due to uranium impurities previously deposited on reactor internals, or to uranium present in core components). In operating BWRs, reactor-coolant iodine and fission-gas concentrations in the off-gas system are measured during operation. An on-line, continuously reading off-gas monitor is also used to estimate fission-gas release from the fuel. For PWRs, iodine in the coolant is the basis for evaluating fuel integrity (Section 3.3). The fission product concentrations are determined by analyzing grab samples of coolant or gaseous effluent with gamma-spectrometric techniques (see Appendix C). Identification of defective fuel is usually accomplished during scheduled refueling outages by a procedure known as “fuel sipping”. Three methods are used: wet sipping, dry sipping, and vacuum sipping. All three feature the use of radiation detectors and associated electronics to measure radionuclide releases from defective fuel rods in an isolated sipping container. A comparison of various parameters for three sipping methods is shown in Table 8–7. It is desirable to identify and remove all defective fuel from the reactor core to prevent the release of fission products into the coolant, and consequently release to the environment. At the same time, however, it is important not to discard expensive sound fuel prematurely. Obviously, precise detection of fuel integrity is particularly desirable. *   Information obtained from Reference 17.

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Radiochemistry in Nuclear Power Reactors Table 8–7 COMPARISON OF FUEL-SIPPING METHODS (Reproduced with Permission, Power Magazine, Ref. 17) Method Fission-Product Driving Force Typical Nuclides Analysis Method Holding Point Overall Sipping Rate* Water Requirements Wet Temperature I-131 I-132 Cs-134 Gamma spectrometry of sample 30–60 min 3–5 bundles/hr out-of-core, 4–6 in-core Demin, 100–200 gal/bundle out-of-core, none in-core Dry Temperature Xe-133 Kr-85 Gamma spectrometry of sample 10–60 min 3–5 bundles/hr None Vacuum Pressure Xe-133 Kr-85 On-line beta scintillation (no samples) 10 min 3–5 bundles/hr out-of-core, 6–8 near-core Demineralized or condensate, 65 gal/bundle *Limited by grapple operator, assuming fuel must be moved from core location.

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Radiochemistry in Nuclear Power Reactors The method most commonly used for locating malfunctioning fuel elements is wet sipping. Key to monitoring here is the leaching of fission products—primarily iodine-131 and cesium isotopes—from fuel rods into an isolated volume coolant. In a BWR, the primary advantage of wet sipping is that it can sometimes be done without removing fuel from the core. Most light-water reactors use out-of-core wet sipping. (For PWRs, in fact, in-core wet sipping is not possible.) This requires lifting the fuel assemblies up and out of the core and placing them in sealed containers filled with water. To reduce radionuclide background levels, the pool water is replaced with demineralized water. After a 30–60 minute holding period, the water is sampled and analyzed for the presence of specific radioactive fission products. Injection of hot water into the sipping container to accelerate the activity release has been reported(18). Dry sipping, as described, depends on forcing the expulsion of fission gas through fuel-cladding defects. This is accomplished by allowing the decay heat from fission to increase fuel bundle temperature. The bundle is removed from the core, placed in a sealed sipping container, and the coolant is displaced with air to expose the fuel rods. When the required temperature is reached, the stagnant air is sampled and analyzed for gaseous fission product activity. Since fission-gas atoms migrate much more readily than iodine and cesium nuclides, better sensitivity is achieved than in the wet method. The main drawback to dry sipping is potential overheating of the cladding surrounding the fuel rods. This results from the absence of water that would otherwise remove heat generated in the decay of fission products. Due to this safety concern, little development effort has been expended on optimization of this technique. Vacuum sipping was developed for use with BWRs. It takes the advantage of high sensitivity of measuring the gaseous fission product activity released from cladding defects under partial vacuum in the sipping container. The vacuum-sipper containers are designed so that they can be installed in the control-rod-blade storage rack in the fuel pool, or on the guide rod inside the BWR reactor cavity. This minimizes the time required to transport fuel to and from the sipping containers.

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Radiochemistry in Nuclear Power Reactors A schematic diagram of a vacuum-sipper system is shown in Figure 8–8. A typical vacuum-sipping cycle starts when a fuel assembly is inserted into the container through the top, which is then closed and sealed by a pneumatically actuated system. Pool water is flushed out of the container with an air-water purge. A gas space is established in the lid above the bundle by aspirating the water out of the lid bottom; at the same time, air is supplied to the top port at a pressure slightly less than the pool pressure outside the lid. The trapped air is then isolated and the space evacuated through a scintillation detector. The remaining low-pressure gas volume is then recirculated through the monitor and back to the can for a predetermined time (3–5 min). After sampling, the container is purged, flooded, and opened for bundle removal. More detailed description of this system has been reported by Green(17).

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Radiochemistry in Nuclear Power Reactors Figure 8–8. Schematic Diagram of a Vacuum-Sipper System (Reproduced with Permission, Power Magazine, Ref. 17)

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Radiochemistry in Nuclear Power Reactors 8.7 REFERENCES (1) P.G.Voilleque, Nucl. Tech., 90, 23 (1990). (2) C.C.Lin, J. Inorg. Nucl. Chem., 42, 1043 (1980). (3) R.L.Hepplette and R.E.Robertson, Proc. Ray. Soc., A252, 273 (1959). (4) E.A.Moelwyn-Hughes, Proc. Ray. Soc., A220, 386 (1953). (5) Yu.V.Kuznetsov, et al., Russian Radiochemistry, 23, 744 (1981). (6) S.W.Duce and J.W.Mondler, Trans. Am. Nucl. Soc., 40, 617 (1985) (7) E.Linden and D.J.Turner, “Carryover of Volatile Iodine Species in Some Swedish and American BWRs”, Water Chemistry of Nuclear Reactor System 3, Vol. I, III (1983) BNES, London. (8) J.H.Keller, “An Evaluation of Materials and Techniques Used for Monitoring Airborne Radioiodine.” 12th AEC Air Cleaning Conference, Vol. I, p. 322 (1972). (9) E.C.Potter and G.M.W.Mann, J. Br. Corr., 1, 2E (1965). (10) E.A.Pelletier and R.T.Hempbill, “Nuclear Power Plant Related Iodine Partition Coefficients”, EPRI NP-1271 (December 1979). (11) C.P.Pelletier, Compiler, “Results of Independent Measurements of Radioactivity in Process Systems and Effluent at Boiling Water Reactors”, U.S. AEC, Docket RM-50–2, ALAP Exhibit No. 22, May 1973. (12) J.M.Skarpelos and R.S.Gilbert, “Technical Derivation of BWR 1971 Design Basis Radioactive Material Source Terms”, NEDO-10871 (July 1973) (Addendum A). (13) C.C.Lin, “Transuranic Isotopes in BWR Coolants”, May, 1975 (unpublished). (14) J.H.Holloway and R.S.Gilbert, “Qualification of the Na-24 Tracer Method for Feedwater Flow Nozzle Calibration”, NEDM-12454 (September 1973). (15) W.B.Wilson, et al. Nucl. Safety, 29, 177 (1988). (16) C.C.Lin “Chemical Behavior of Radioiodine in BWR Systems. (II) Effects of Hydrogen Water Chemistry”, Nucl. Tech. 97, 71 (1992). (17) T.A.Green, “Monitoring Fuel for Defects”, Power, P. 56, (August 1978). (18) S.Uchida, et al. Nucl. Tech., 40, 79 (1978).

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