LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 11, Radioactive Waste Management
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SECTION 11 RADIOACTIVE WASTE MANAGEMENT The purposes of this section performance evaluation of the are 1) to provide a complete description Radioactive Waste Treatment Systems and 2) and to demonstrate that, under normal plant the anticipated operational will be in conformance regulations. occurrences, with applicable operating conditions and during radioactive releases from the plant Nuclear Regulatory Commission (NRC) 11.1 SOURCE TERMS Source terms describe the quantity, chemical species, and timing of radioactive materials released from the core and/or primary coolant system for a specific accident. The amounts of radioactive materials which are produced and stored in the reactor system are discussed in this section. These sources have been calculated for the design basis accidents and for normal operation using the ORIGEN computer code for nuclide concentrations and activities from fuel depletion. 11.1.1 Determination of Activity in Reactor Core The total core activity was originally calculated for Salem with the ORIGEN computer code. The alternative source term (AST, Ref. 1) analysis required additional radionuclide groups than originally calculated by ORIGEN. The numerical values for core activity (60 isotopes) used in the AST LOCA analysis in UFSAR Section 15.4.1 are shown in Table 11.1-1 and were obtained from the default nuclide inventory file from the RADTRAD (3.02 and 3.03) computer program (Refs. 5, 6 & 7). The LOCA dose analysis is based on these 60 isotope activities and a core thermal power of 3632 MWt (=3459 x 1.05). The aerosol inventory is multiplied by 1.10 in some analysis applications to make the RADTRAD default file conservative with respect to the SNGS plant-specific ORIGEN inventory file. 11.1-1 SGS-UFSAR Revision 28 May 22, 2015 I 11.1.2 Activities in The Fuel Rod Gap 'l'he activity contained in the space (gap) between the fuel pellets and the cladding is released when the cladding is breached. Rod failure is typically caused by high fuel temperature and primary system depressurization. 'l'he fraction of core activity assumed to be in the gap can vary depending on the specific application. Gap activity is the primary source term for the locked rotor, rod ejection and fuel handling accidents. The gap activity basis is discussed as part of the assumptions described in the specific accident section of Chapter 15. 11.1-2 SGS-UF'SAR Revision 23 October 17, 2007 * *
- 11.1.3 Fuel Handling Sources The inventory of fission products in a fuel assembly is dependent on the power rating of the assembly. The parameters used for calculation of the highest rated assembly in the core to be discharged are provided in Section 15.4.6. 11.1-3 SGS-UFSAR Revision 20 May 6, 2003 I 11.1.4 Reactor Coolant Fission Product Activities The parameters used in the calculation of the reactor coolant fission product concentrations, including pertinent information concerning the expected coolant cleanup flow rate, demineralizer effectiveness, and volume control tank noble gas stripping behavior, are presented in Table 11.1-7. The results of calculations are presented in Table 11.1-8. The table lists nuclides of fission and corrosioin products which are significant from a shielding standpoint as well as those nuclides which are listed in ANS standard ANSI/ANS-18.1-1984. The values tabulated are the maxLmums that occur during the fuel cycle from startup through the equilibrium cycle. In these calculations, small cladding defects in the equivalent of one percent of the fuel rods are assumed to be present at the 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, based upon 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 nuclides in the coolant: SGS-UFSAR 11.1-4 Revision 16 January 31, 1998 For daughter nuclides in the coolant: Where: He NF t R F v Me l OF QL
- f = = = = = = = = :s = = = Concentration of nuclide in the reactor coolant (atoms/gram) Inventory of nuclide in the fuel (atoms) Operating (seconds) Nuclide release coefficient (1/sec) -F
- v Fraction of fuel rods with defective cladding Fission product escape rate coefficient (1/sec) Mass of reactor coolant (grams) Nuclide decay constant (1/see) Nuclide demineralizer decontamination factor Purification or letdown mass flow rate (grams/sec) Nuclide volume control tank stripping fraction Fraction of parent nuclide decay events that result in the formation of the daughter nuclide D = Dilution coefficient for feed and bleed (1/sec) = : p 1 BD-p. I DF B0 = Initial boron concentration (ppm) p = Boron concentration reduction rate (ppm/sec) and where: SGS-UFSAR subscript i refers to the parent nuclide subscript j refers to the daughter nuclide 11 .. 1-5 Revision 16 January 31, 1998 11.1.5 Tritium Production 11.1.5.1 General -overall sources Tritium is formed from several sources, the most abundant of which is the fissioning of uranium, which yields tritium as a ternary fission product. Tritium atoms are generated in the fuel at a rate of approximately 8 x 10-5 atoms per fission, or 1.05 x 10-2 curies/mwt/day. Boron-bearing control rods can also be a potential source of tritium. These potential sources of tritium are only present in the reactor coolant to the extent that they diffuse through the fuel or control rod cladding. A direct source of tritium is the reaction of neutrons with dissolved boron in the reactor coolant. Neutron reactions with lithium are also a direct source of tritium. Lithium is present in a pressurized water reactor (PWR) for pH control and as a product of boron reactions with neutrons. An extremely small amount of tritium is also produced by neutron reactions with naturally occurring deuterium in light water. 11.1.5.2 Specific Individual Sources of Tritium 11.1.5.2.1 Ternary Fissions -Clad Diffusion Because of the mode of operation of the PWR to minLmize any liquid or gaseous discharges from the plant, it has been possible to very accurately determine the buildup of tritium from various sources in the plant and to identify their origin. A program was undertaken by Westinghouse to determine the source of tritium in the reactor coolant in operating plants with both stainless steel and zircaloy cladding. This program clearly indicated that with the current generation of Westinghouse reactors with zircaloy-clad fuel, 1 percent or less of the tritium produced in the fuel will diffuse through the cladding into the coolant. For those plants containing stainless steel cladding, operational data have shown that as high as 80 percent of the ternary tritium produced will diffuse through the cladding. The tritium concentration in the reactor coolant in those plants having stainless steel and zircaloy fuel cladding has been substantially different. Tritium concentrations at Yankee-Rowe (600 MWt), which has stainless steel cladding, has ranged from 11.1-6 SGS-UFSAR Revision 16 January 31, 1998
- -----about 4.5 to 5 essentially throughout the core cycle. A total discharge from the plant during the core cycle of -1500 curies of tritium was reported in the monthly operating reports. In addition, with the stainless cores, there has been a continuing source of tritium to the reactor coolant during the power coastdown period when all the boric acid has been removed from the system. This information, in particular, substantiates the high tritium diffusion through the stainless steel clad. The experience at the R. E. Ginna plant has been substantially different. This plant operates at 1455 MWt and has zircaloy cladding. After approximately 8 months of operation at Ginna, the tritium concentrations were less than 0.3 in the reactor coolant and the monthly discharges averaged -5 curies/month. The experiences at Benzau and Zori ta were comparable. An extensive program to follow the buildup of tri tiurn in the Ginna plant was initiated, and the results indicated a potential source from the core which is 1 percent or less of the ternary fissions generated in the fuel. Based on this experience, the tritium sources during the operation of a PWR can be very accurately predicted. In the past, Westinghouse has assumed that 30 percent of the tritium from ternary fissions would diffuse through the zircaloy fuel. This value was used as a basis for systems and operational design and is clearly conservative. 11.1.5.2.2 Tritium Produced from Boron Reactions The neutron reactions with boron resulting in the production of tritium are: 10 B (n, 2a) T 10 ( ) . 7 B n, a Ll {n, na) T 11 9 B (n, T) Be SGS-UFSAR 11.1-7 Revision 6 February 15, 1987 (n, a) T Of the above reactions, only the first two contribute signi to the tritium production in a PWR. The B11 (n, T) Be9 reactions have a threshold of 14 Mev and a cross section of "'5 mb. Since the number of neutrons produced at this energy is less than 109 n/cm2-sec the tritium produced from this reaction is negligible. The B10(n, d) reaction may be neglected since has been found to be unstable. 11.1.5.2.3 Tritium Produced from Lithium Reactions The neutron reactions with lithium in the production of tritium are: Li7 (n, na) T Li6(n, a) T In Westinghouse designed reactors, li thi urn is used for pH adjustment of the reactor coolant. The reactor coolant is maintained at a maximum state level of 3. 5 +/- 0.15 ppm lithium by the addition of Li70H and by a cation demineralizer included in the Chemical and Volume Control demineralizer will remove any excess of lithium such as could be the B10(n, a) Li7 reaction. The Li6 (n, a) T reaction is controlled by limiting the impurity in the This in OH used in the reactor coolant and by lithiating the demineralizers with 99.9 atom percent 11.1.5.2.4 Control Rod Sources In a fixed burnable poison rod, there are two primary sources of tritium generation: the B10 {n, 2a) T and the P10 (n, a) Li 7 (n, na) T reactions. Unlike the coolant where the level is controlled at a maximum steady state level of 3.5 +/- 0.15 ppm, there is a buildup of 11.1-8 SGS-UFSAR in the burnable Revision 24 May 11, 2009 poison rod. The burnable poison rods are required during the first year of operation only. During this time the tritium production is 72 curies/pound a10* There are no tritium sources in Ag-In-Cd control rods. 11.1.5.2.5 Tritium Production from Deuterium Reactions Since the fraction of naturally occurring deuterium in water is less than 0.0015, the tritium produced from this reaction is negligible (less than 1 curie per year) 11.1.5.2.6 Total Tritium Sources Tritium sources released to the reactor coolant are listed in Table 11.1-9, based on 12 months of operation at full power (3558 MWt} and a 0.8 load factor. Included in Table 11.1-9 is the amount of tritium produced in the reactor for all of the nuclear reactions described above. Two columns of values of tritium released to the reactor coolant are given in Table 11.1-9, namely, a design value and an expected value. The design values are based on a release of 30 percent of the tritium produced being diffused through the fuel cladding. The present values are based on operating experience at existing PWR facilities where the data have indicated that the previous design values were unduly conservative. Based on this experience, the tritium released to the reactor coolant for a typical 3558 MWt reactor is reduced from -3815 to -690 curiesjyear. Basic parameters employed to calculate the tritium inventory are given in Table 11.1-10. 11.1-9 SGS-UFSAR Revision 6 February 15, 1987 11.1.6 Volume Control Tank Activity The radiation sources in the volume control tank (VCT) are basad on a nominal operating level in the tank of 200 cubic feet in the liquid phase and 200 cubic feet in the vapor phase, and on the stripping fractions given in Table 11.1-7, assuming no VCT purge. Table 11.1-11 lists the activities for the vapor phase of the VCT with clad defects in 1 percent of the fuel rods. 11.1.7 Gas Decay Tank Activity The isotopic maximum inventories are determined in the RCS and VCT. Since there is no continuous purge from the control tank, the activity values are obtained based on the following considerations:
- At shutdown the radiogas inventory of the VCT is instantaneously transferred to the GOT.
- The RCS (operating with one percent fuel defects) is stripped to the VCT at the maximum letdown with a stripping fraction of 1.0 over a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> period at the end of which the VCT is again instantaneously transferred to the GOT.
- The radioactive sources are calculated at the point where Xr-88 is a maximum in the GDT. This provides a lLmiting gamma source. The GDT activities for noble gas nuclides are presented in Table 11.1-12. 11.1-10 SGS-UFSAR Revision 16 January 31, 1998 11.1.8 Activity in Recirculated Sump Water The concentration of iodine isotopes in the recirculation loop at initiation of recirculation phase after the design basis loss-of-coolant accident (LOCA) was replaced with an al terna ti ve source term (AST) pursuant to Section 50. 67 of Title 10 of the Code Of Federal Regulations ( 10 CFR 50. 67), "Accident Source Term", and the potential radiological consequences were re-evaluated. The guidance provided in Regulatory Guide 1.183, Alternative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2002, was used in the re-evaluation. The sump water volume is 328,600 gallons and 40% of the total core value of Iodine (in accordance with Reg. Guide 1.183, Table 2) is released to the containment sump. Of this total, 4. 85%, which is elemental, is subject to becoming airborne in proportion to a flashing rate. The remainder is assumed to remain waterborne since the sump water pH is maintained> 7. The radioactivity in the containment would be an additional source of radiation to the auxiliary building following a LOCA. The residual heat removal loop source and the containment source are used to calculate post-accident radiation doses in the Auxiliary Building. The radioactivity leaking out of the recirculation flow path in the Auxiliary Building is Engineered Safety Features (ESF) Leakage and is assumed to be a total of 0.45 gpm in the Section 15.4.1 accident analysis. 11.1.9 References for Section 11.1 1. Regulatory Guide 1.183, Alternative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2002 2. CCC-217, ORIGEN Isotope Generator and Depletion Code, Matrix Exponential Method, April, 1975. 3. Radioactive Source Term for Nominal Operation of Light Water Reactors, ANSI/ANS-18.1-1984, American Nuclear Society, December, 1984. 4. Section 50. 67 of Title 10 of the Code Of Federal Regulations ( 10 CFR 50.67), "Accident Source Term" 5. "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", NUREG/CR-6604, USNRC, April 1998 6. "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", NUREG/CR-6604 Supplement 1, USNRC, June 8, 1999 7. "RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation," W.C. Arcieri (ITSC), NUREG/CR-6604 Supplement 2 (October 2002) 11.1-11 SGS-UFSAR Revision 28 May 22, 2015