ML16348A454
| ML16348A454 | |
| Person / Time | |
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| Site: | Waterford  |
| Issue date: | 08/25/2016 |
| From: | Entergy Operations |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16256A115 | List:
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| References | |
| W3F1-2016-0053 | |
| Download: ML16348A454 (34) | |
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WSES-FSAR-UNIT-3 11.1-1 Revision 14 (12/05)(DRN 03-2065, R14)11.0 Radioactive Waste ManagementThis chapter provides information on the design basis radioactive source terms and the systems utilized to manage radioactive wastes. The waste management systems include the liquid waste management system, the gaseous waste management system, the solid waste management system, and the process
and effluent radiological monitoring and sampling systems. 11.1 Source Terms(DRN 03-2065, R14)11.1.1 DESIGN BASIS SOURCE TERMS 11.1.1.1 Maximum Fission Product Activities in Reactor CoolantMaximum fission product activities have been used as design basis source terms for shielding and facilities design and for calculating the consequences of postulated accidents. The isotopes chosen for consideration in the maximum case are those which are significant for design purposes by reason of a
combination of energy, half-life, or abundance. The mathematical model used to determine the maximum concentration of nuclides in the Reactor Coolant System (RCS) 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 fuel pellet region, the mass balance includes fission product production by direct fission yield, by parent fission product decay, and by neutron activation, while the removal includes decay, neutron activation, and escape to the coolant.In the coolant region, the analysis includes the fission product production by escape from the fuel through defective fuel rod cladding, parent decay, by coolant purification, by feed and bleed operations (for fuel burnup), by leakage, and other feed and bleed operations such as startups and shutdowns, as well as load follow operation. The expression derived to determine the fission product inventory in the fuel pellet region is : (DRN 00-1045, R11-A; 03-2065, R14)i,piiii,p iii ii,pN)(N)f(PFY dt dN1 111 (1) (DRN 00-1045, R11-A) and in the primary coolant region is:
1111i,ciiii,pii,cN)CVRf(NDv dt dN (2) i,cto i i i N W L CCC)(W Q1(DRN 03-2065, R14) where:N = population (atoms) F = average fission rate (fission/MWt-sec)
WSES-FSAR-UNIT-3 11.1-2 Revision 14 (12/05)Y = fission yield of nuclide (atoms/fission) P = core power (MWt)
= decay constant (sec
- 1) = microscopic capture cross-section (cm
- 2) = thermal neutron flux (neut/cm 2 sec)(DRN 00-1045, R11-A) = escape rate coefficient (sec 1)(DRN 00-1045, R11-A)f = branching fraction t = time (sec)
D = defective fuel rod cladding (fraction)
CVR = core coolant volume to reactor coolant volume ratio Q = purification flowrate during power cycle (lb./sec)
W = reactor coolant mass during power cycle (lb.)
= resin efficiency of Chemical and Volume Control System (CVCS) ion exchanger for a given nuclide
C o = initial boron concentration (ppm) C = boron reduction rate by feed and bleed compensating for fuel burnup (ppm/sec) L = leakage or other feed and bleed from reactor coolant (lb./sec)
Subscripts P = pellet region c = coolant region i = i th nuclide i-l = precursor to i th nuclide(DRN 03-2065, R14)The above equations were used to determine the fission products in the reactor coolant; however, for the maximum reactor coolant specific activities, removal terms such as leakage and gas stripping were not
considered.Escape rate coefficients are used to represent the overall release from the fuel pellet to the coolant.Escape rate coefficients for noble gases, halogens and other elements are given in Table 11.1-1. Note: the following is historical information pursuant to NEI 98-03, which is identified by a designation of "Start" and "End". (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 11.1-3 Revision 14 (12/05)(DRN 03-2065, R14)
Start of Historical Information The escape rate coefficient is an empirical value that was derived from experiments initiated by Bettis Atomic Power Laboratory and run in the NRX and MTR reactors. The escape rate coefficients were obtained from test rods which were operated at high linear heat rates. The linear heat rates were uniform over the test sections of 10.25 inch in length. The exact linear heat rates were not precisely known but postirradiation inspection showed that some test specimens had experienced centerline melting. Later
test were done in Canada (4) to determine the effect of rod length on the release of fission gases and iodines from defective fuel rods. A byproduct of these experiments was the relationship between linear
heat rate and escape rate coefficient.
End of Historical Information (DRN 03-2065, R14)Section 4.2 contains a detailed discussion of fuel performance and the basis for the defective cladding percentage utilized. The values of parameters used to calculate the reactor coolant maximum fission product activities are shown in Table 11.1-1. The maximum fission product activities are presented in Table 11.1-2. 11.1.1.2 Normal Operating Source Terms Including Anticipated Operational Occurrences(DRN 03-2065, R14)The data in Table 11.1-3 represent the expected normal fission product activities for the plant with no gas stripping. The activities for this case are based on ANS-18.1 (31) and is intended for use in evaluating only normal operations including anticipated operational occurrences. The assumptions used to calculate
these activities are presented in Table 11.2-12. (DRN 03-2065, R14)11.1.2 DEPOSITED CORROSION PRODUCTS The activity of radioactive crud (Reactor Coolant System corrosion products) and its thickness on Reactor Coolant System surfaces are evaluated using measured data from various operating pressurized water
reactors. Even though these reactors have different water chemistries and different materials in contact with the reactor coolant, the crud activity is remarkably similar. The half-lives, reactions, and gamma decay energies for each of the long-lived isotopes are listed in Table 11.1-4. The long-lived isotopes are those
significant isotopes remaining after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The radioactive crud originates on in-core and out-of-core surfaces. The crud plates out on the in-core surfaces and re-erodes after a short irradiation period. This irradiation period or core residence time (in seconds) is determined by the following expressions (see Appendix 11.1A for the derivation of these equations): circulating crud (DRN 00-1045, R11-A)ciTi res AAA n t)67.16(11 1 (sec) (3)
(DRN 00-1045, R11-A)
WSES-FSAR-UNIT-3 11.1-4 Revision 14 (12/05) and deposited crud (in core) t-11n 1-A (16.67)
(sec) (4)res j iwhere: A i ,A j= the crud activities for each isotope (dpm/mg-crud) i = the activation rate (d/gm-sec)
A c= the core surface area (cm
- 2) = decay constant (sec
- 1) A T = total primary system area (cm
- 2) 16.67 = units conversion min-mgm/gm-sec (DRN 00-1045, R11-A) (DRN 00-1045, R11-A)
The activation cross section i is as follows: gm/cm, AN)o/w()o/a(iioi i i 2 (5) where: (a/o)i = the isotopic abundance (w/o)i = the elemental abundance in the crud or the elemental abundance in the base metal
N o = Avogadro number (0.6023 x 10 24 a/gm-mole) (DRN 03-2065, R14)
[A]i = the atomic weight of isotope (i), gpm (DRN 03-2065, R14)i= the microscopic cross section(cm 2)Circulating crud is taken to be all crud in the reactor coolant. Deposited crud is taken to be all crud which plates out in-core surfaces.
WSES-FSAR-UNIT-3 11.1-5 Revision 14 (12/05)(DRN 03-2065, R14)Note: The following is historical information pursuant to NEI 98-03, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 99-2361, R11)The measured averages and maximum crud activities (dpm/mg-crud) as taken from Reference 6 through 19 for those reactors considered in the determination of the core residence times are as shown in Table 11.1-5. The average and maximum core residence times as determined by the above expressions, the activation rates in Table 11.1-6 and the system parameters in Table 11.1-7 are as shown in Table 11.1-8.
As all the Fe-59 residence times are long, its activity (A j) is assumed saturated. The averages (t res) of the maximum residence times are also given in Table 11.1-8. (DRN 99-2361, R11)
End of Historical Information (DRN 00-1045, R11-A)
The calculated crud activities (A i) for Waterford 3 are determined utilizing the averages of the maximum core residence times (t res) in Table 11.1-8, the activation rate in Table 11.1-9, a core-to-primary surface area ratio of 3.72, and the following equation: (DRN 00-1045, R11-A)crudmg/dpm),06.0(A A)e1(A T c]t[iiresi (6) (DRN 00-1045, R11-A)
As the averages of the maximum residence times (t res) are used in general these are a factor of two to four higher than a straight average residence time. The resulting calculated crud activities will be conservative. These calculated crud activities of the long-lived isotopes considered for Waterford 3 are
as shown in Table 11.1-10. These calculated crud activities are appplied to both the circulating crud and out-of-core deposited crud.Using the average crud level in the reactor coolant (75 ppb) of those operating reactors shown in Table
11.1-5, and the calculated crud activities (dpm/mg-crud) as shown in Table 11.1-10, the average isotopic activities in the primary coolant are determined by the following expression: (DRN 03-2065, R14)A A 6075 x 10 2.7 x 10 1 x 10,Ci/cm (7) i-9-5+33where:(DRN 99-2361, R11)
= is density of water (g/cm 3)(DRN 99-2361, R11) 2.7 x 10 5 converts disintegrations/sec to micro-curies. (DRN 03-2065, R14)The average calculated activities in the primary coolant using the above expression are shown in Table 11.1-11. The maximum coolant activities can be higher due to crud bursts during shutdowns or changes in power. However, these bursts occur over short periods of time, and therefore, the average values are
more reasonable to use for long-term operation. (DRN 03-2065, R14)
The equilibrium thickness of radioactive crud film (mg-crud/cm
- 2) has been determined by two methods: a) The direct measurement of the film during maintenance and/or tests in operating reactors.
WSES-FSAR-UNIT-3 11.1-6 Revision 14 (12/05)b) Calculating crud film thickness from measured dose rates and specific activities (dpm/mg-crud) of deposited crud. The equilibrium crud film thickness for various Reactor Coolant System areas are as shown in Table 11.1-12.The calculated crud activities in this section are reasonable values and together with measured plateout thicknesses match measured shutdown dose rates around various equipment associated with operating reactors. However, the crud levels do have rather wide variations as shown in Table 11.1-5 for operating reactors and many combinations of activites and plateout thicknesses could reproduce the measured shutdown dose rates. It is for this reason that the crud activities are periodically reviewed as more measured crud activities, plateout thicknesses and dose rates become available. (DRN 03-2065, R14)The conservative evaluation of the above operating data yields circulating crud concentrations (Table 11.1-11) which are lower for all isotopes than the concentrations given in ANS-18.1. ANS-18.1 is intended for use in evaluating only normal operations including anticipated operational occurrences.
However, the ANS-18.1 values for circulating crud are used as design source terms as well as for normal operations, since they yield a design that is more conservative than that resulting from the derived values
based on operating experience. The circulating corrosion product activity concentrations in the reactor
coolant are listed in Tables 11.1-2 and 11.1-3. (DRN 03-2065, R14)11.1.3 NEUTRON ACTIVATION PRODUCTS 11.1.3.1 Nitrogen-16 ActivityNitrogen-16 is produced by the 0-16 (n,p) N-16 reaction which beta decays, emitting high energy gamma 78 percent of the time. The gamma energies are 6.13 MeV, 73 percent of the time and 7.10 MeV, 5 percent of the time. The N-16 half-life is 7.13 sec. The threshold for the reaction is 10.2 MeV. (DRN 03-2065, R14)
The N-16 activity at the reactor vessel outlet nozzle is 7.66 x 10 6 disintegrations/cm 3-sec based on maximum flow. This activity is based on the following expressions and reactor parameters: (DRN 03-2065, R14)
Activity (dis/cmsec) 1-ee1-e (8) 3-t-t-t cR Twhere:A = activity (DRN 03-2065, R14) = is the reaction rate 7.20 x 10 7 (dis/cm 3-sec)
- t C = is the core transit time (0.672 sec) t T = is the total primary loop transit time (8.361 sec) t R = is the time from the core outlet to the point of interest (0.677 sec to outlet nozzle) (DRN 03-2065, R14) = is the decay constant (0.097/sec)
WSES-FSAR-UNIT-3 11.1-7 Revision 14 (12/05)(DRN 03-2065, R14)Similar calculations based on minimum flow conditions gave a Nitrogen-16 activity at the reactor vessel outlet nozzle of 8.19 x 10 6 disentegrations/cm 3-sec.(DRN 03-2065, R14)11.1.3.2 Carbon-14 Production(DRN 99-2361, R11)Carbon-14 is produced in the RCS by activiation of 0-17 and N-14 isotopes. The greatest amount of C-14 gas is produced by the 0-17 (n,) C-14 reaction, a lesser amount of C-14 is produced by the N-14 (n,p) C-14 reaction. The production of C-14 from both sources can be calculated by using the following equation:(DRN 99-2361, R11)(DRN 03-2065, R14)Q =N m t p s oo(DRN 03-2065, R14) where: N o = atom concentration in the RCS water, atoms/kg H 2 Oo = thermal cross-section, cm 2(DRN 03-2065, R14) = thermal neutron flux, 4.96 x 10 13 n/cm 2-sec(DRN 03-2065, R14)m = mass of core water, 2.22 x 10 4 kg t = 3.15 x 10 7 sec/yr p = plant capacity factor, 0.8 s = 1.03 x 10 22 Ci/atoms Q = production rate, Ci/yr (DRN 03-2065, R14)(DRN 03-2065, R14)(DRN 99-2361, R11; 03-2065, R14)
For C-14 production from O-17 activation, N o = 1.3 x 10 22 atoms 0-17/kg (H 2 O) and o = 2.4 x 10
-25 cm 2 is used in the above equation. The production rate is 8.92 Ci/yr. For C-14 production from N-14 activation N o = 2.75 x 10 20 atoms, N-14/kg (H 2 O) and o = 1.8 x 10
-24 cm 2 is used in the above equation. The production rate is 1.42 Ci/yr. (DRN 99-2361, R11)
The annual production of C-14 from these sources will be 10.34 curies. (DRN 03-2065, R14)11.1.4 TRITIUM PRODUCTION IN REACTOR COOLANT The principal sources of tritium production in a PWR are from ternary fission and neutron induced reactions in boron, lithium and deuterium that are present in the coolant, borated shim rods, and control element assemblies (CEA). The tritium produced in the coolant contributes immediately to the overall tritium activity while the tritium produced by fission and neutron capture in the CEAs and borated shim rods contributes to the overall tritium activity via release through the cladding.
WSES-FSAR-UNIT-3 11.1-8 Revision 14 (12/05)11.1.4.1 Activation Sources of TritiumThe activation reactions producing tritium are as shown in Table 11.1-13. The tritium production from reactions five and six (B-11 and N-14 sources) is insignificant due to low cross-section and/or abundance and can be neglected. Reactions one through four from B-10, lithium, and deuterium are the major
sources of tritium in the coolant, CEAS, and borated shim rods.
The tritium production from the above sources is determined by the following expressions:
Tritium formation rate = Production rate - decay (10)N-dt dN a(DRN 99-2361, R11)(11)(t)at timeatoms/cm),e-(1 N 3 t aactivity (curies) V N x 2.7 x 10 (12) (1-e) V x 2.7 x 10-11 a-t-11where:a = the production rate (atoms/cm 3-sec)(DRN 99-2361, R11) t = the reactor operating period of interest (DRN 03-2065, R14) V = the effective core water volume, 2.0841 x 10 7 cm 3 2.7 x 10-11 converts disintegrations/sec to curies (DRN 03-2065, R14)The appropriate parameters used in the calculation are as shown in Table 11.1-14. Based on these parameters, the tritium produced from activation sources in the reactor coolant for one equilibrium fuel
cycle are included in Table 11.1-15. 11.1.4.2 Tritium From FissionThe ternary fission production of tritium in the core is expressed simply by:
dN dt YF - N (13)N YF (1-e), atoms at time (t) (14)-tactivity (curies) N x 2.7 x 10 YF (1-e) 2.7 x 10 (15)-11-t-11 WSES-FSAR-UNIT-3 11.1-9 Revision 14 (12/05) where: Y = the tritium fission yield (tritium atoms per fission) F = the fission rate (f/sec) t = the reactor operating period of interest, 2.7 x 10
-11 converts disintegrations/sec to curies (DRN 03-2065, R14)
Tritium as a product of fission (1) (20) has a yield of 8.0 x 10
-5 atoms/fission for U-235 and a yield of 2.6 x 10-4 atoms/fission for U-238, Pu-239 and Pu-241. The amount of tritium that is released through fuel cladding can be indirectly determined using measured tritium levels from operating PWR's subtracting the calculated tritium activity produced by neutron capture in the reactor coolant, and attributing the remaining tritium activity to release from the cladding of the fuel rods, borated shim rods, and CEAS. Due to the large number of fuel rods as compared to the number of borated shim rods and CEAs within the core during operation, any amount of tritium released to the system will be principally from the fuel rods. The total amount of tritium produced per fuel cycle can be determined by summing the total tritium discharged in the gaseous, liquid, and solid waste discharges of the plant and the tritium inventories in the RCS and other waste or refueling tanks that can contain tritium at the end of the fuel cycle of interest. This method has been used to analyze operating data from various PWRs (6) (21) thru (26).The results of the analysis are shown in Table 11.1-16. Buildup of plutonium in the fuel with burnup was accounted for in the analysis. An average expected two percent tritium release from the fuel and a maximum five percent design value are used to estimate the annual tritium production in Table 11.1-15. (DRN 03-2065, R14)11.1.5 TRITIUM CONCENTRATION AND RELEASES (DRN 03-2065, R14)Table 11.1-2 presents the tritium concentration in the reactor coolant under design basis conditions. The value is based on the production terms from the maximum case presented in Table 11.1-15 and removal due to decay alone. This approach is highly conservative since tritium will be lost due to evaporation from the fuel pool primary to secondary leakage and releases to the circulating water discharge. In addition, at equilibrium it is assumed that tritium is uniformly mixed within the reactor coolant, spent fuel
pool and refueling water storage pool. (DRN 03-2065, R14)The following model was used to derive the tritium concentration in the primary coolant. (DRN 99-2361, R11)
A = I (264)
Z V (DRN 99-2361, R11) where: A = tritium concentration in coolant (Ci/g) I = input of tritium into the coolant as presented in Table 11.1-15 (Ci/yr)
Z = tritium removal coefficient due to radioactive decay (.0693 yr
-1)(DRN 99-2361, R11) V = the combined volume of the primary coolant (5.5 x 10 4 gal) the spent fuel pool (312,000 gal) and the refueling water pool (601,000 gal). (DRN 99-2361, R11)
WSES-FSAR-UNIT-3 11.1-10 Revision 14 (12/05)(DRN 03-2065, R14)Table 11.1-3 presents the primary coolant tritium concentration during normal operation, including anticipated operational occurrences. This value is based on ANS-18.1 and NUREG 0017 and is consistent with the concentration which is obtained using the above model but applying a removal
coefficient (Z) which includes evaporation from the pool surface and leakage. (DRN 03-2065, R14)(DRN 99-2361, R11)11.1.6 SPENT FUEL POOL FISSION PRODUCT AND CORROSION PRODUCT ACTIVITIES (30) (DRN 99-2361, R11)Spent fuel pool maximum and expected fission and corrosion product specific activities are given in Table 11.1-17 for the start of the refueling period. It is assumed that upon shutdown for refueling, the RCS is cooled down for a period of approximately two days. During this period, the primary coolant is letdown through the purification filter, purification ion exchanger, and volume control tank. This serves two purposes; removing the noble gases in the volume control tank avoids large activity releases to the containment following reactor vessel head removal, and the ion exchanger and filtration reduces dissolved fission and corrosion products in the coolant which would otherwise enter the spent fuel pool and refueling water cavity. At the end of this period, the coolant above the reactor vessel flange is partially drained. The reactor vessel head is unbolted and the refueling water cavity is filled with approximately 365,900 gal of water from the refueling water storage pool. The remaining reactor coolant volume containing radioactivity is then mixed with water in the refueling cavity and spent fuel pool. After refueling, the spent fuel pool is isolated and the water in the refueling cavity is returned to the refueling water storage pool. This series of events determines the total activity to the spent fuel pool. The specific activities of the radionuclides given in Table 11.1-17 are based upon a volume of 292,000 gal. These values are initial values which will be reduced by decay during refueling as well as by operation of the
Fuel Pool System. 11.1.7 LEAKAGE SOURCES Systems containing radioactive liquids are potential sources of leakage to the environment. Table 11.1-18 provides a listing of leakage values from valves and pumps. Leakage of primary coolant into the containment building atmosphere, which is ultimately exhausted to the environment at times of containment purge, is assumed to be one percent per day of the primary coolant noble gas activity and
.001 percent per day of the iodine activity in the primary coolant. An additional potential source of gaseous discharge results from coolant leakage into the Reactor Auxiliary Building. A leakage rate of 180 lbs/day of a mixture of hot and cold primary coolant leakage is assumed, with an iodine and noble gas partition factor of .0075 and 1.0 respectively.The liquid from these leakage sources is collected and
processed in the Liquid Waste Management System which is described in Section 11.2. (DRN 03-2065, R14)Primary to secondary leakage can result in the buildup of radionuclides in the secondary coolant and Steam Generator Blowdown System (SGBS). Under normal operation a leakage rate of 75 gal/day is assumed. This activity can ultimately result in discharge of small amounts of liquid and gaseous wastes to the environment. The discharge of liquid waste can result from liquid leakage to the Turbine Building sump and the release of portions of blowdown. It is assumed that leakage to the Turbine Building sump is five gpm. and that all of steam generator blowdown is processed and returned to the secondary coolant
system. The SGBS is discussed in Section 10.4.8. (DRN 03-2065, R14)Gaseous releases from the secondary side can result from main steam leakage, the gland seal system exhaust and the discharge of noncondensible gases from the SGBS flash tank. Overall main steam leakage is assumed to be approximately 1700 lbs/hr. and originates from many sources, each too small WSES-FSAR-UNIT-3 11.1-11 Revision 14 (12/05)to identify. Turbine gland seal steam flow is sent to a gland steam condenser resulting in negligible discharges. Since all noncondensible gases from the SGBS flash tank are vented to the condenser, these releases are also negligible. (DRN 99-2361, R11)
The above leakage rates and partition coefficients are based on the recommendations and experience
presented in NUREG 0017 (28).(DRN 99-2361, R11) Releases inside the plant are handled by the appropriate ventilation system. Containment air purification and cleanup systems are described in Section 9.4.5. Auxiliary and turbine building ventilation systems are discussed in Section 9.4 and continuous radiation monitors are discussed in Section 12.3.4. The source
terms used as design bases for evaluating these systems are provided in Subsection 12.2.2.
Means of controlling leakage are discussed in Section 5.2.
11.1.8 STEAM GENERATOR ACTIVITY MODEL (DRN 99-2361, R11; 03-2065, R14)The radionuclide concentrations in the steam generators under normal operating conditions, including anticipated operational occurrences, are presented in Table 11.1-3 and are based on models
recommended in ANS-18.1 (31) and NUREG 0017 (28).(DRN 99-2361, R11; 03-2065, R14)The radionuclide concentrations in the steam generators under design basis conditions are based on the assumptions presented in Table 11.1-19. The activities are presented in Table 11.1-20. (DRN 03-2065, R14)Assuming constant reactor coolant activities and a negligible contribution from the condensate, the equilibrium radioisotope concentrations in the steam generator are given by the following equation: BVPFF QA A sgiilsg i p i sg(DRN 03-2065, R14) where: sg i A = the total activity of isotope i in the steam generator liquid at time t, Ci(DRN 03-2065, R14)Q = primary to secondary leakage, vol/sec A i p = primary coolant activity of isotope l, Ci F sg = steam generator flow rate, vol/sec l i PF = steam generator partition factor for the i th isotope i = decay constant for the i th isotope, sec
-1 V sg = steam generator liquid volume B = steam generator blowdown rate, vol/ sec (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 11.1-12 Revision 14 (12/05)
SECTION 11.1: REFERENCES 1. Meck, M.E. and Rider B.F., "Summary of Fission Product Yields for U 235 , U 238 , Pu 241 atThermal, Fission Spectrum and 14 MeV Neutron Energies," APED-5398 , Class 1, March 1, 1968. 2. Lippincott, E.P., Pitner, A.L., and Kellogg, L.S., "Measurement of 10 B (n,t) Cross Section in a Fast Neutron Spectrum," HEDL-TME-73-49,. May 1973. 3. "Neutron Cross Sections," BNL 325 Supplement No. 2 , May 1964. 4. G.M. Allison and H.K. Rae, "The Release of Fission Gases and Iodines from Defected UO 2 Fuel Elements of Different Lengths" AECL-2206, June 1965. (DRN 03-2065, R14)
- 5. Deleted (DRN 03-2065, R14)6. Grant, P.J., et al., "Oconee Radiochemistry Survey Program," RDTPL-75-4 , May 1975. 7. Weisman, J., and Bartnoff, S., "The Saxton Chemical Shim Experiment," WCAP-3269-24 , July 1965.8. "Large Closed-Cycle Water Reactor Research and Development Program," Progress Report, WCAP-3269-13 , April 1, 1965 - June 30, 1965. 9. "Corrosion Product Behavior in Stainless-Steel-Clad Water Reactor Systems," Nuclear Applications, Vol. 1 , October 1965. 10. Abrams, C.S., and Salterelli, E.A., "Decontamination of the Shippingport Atomic Power Station," WAPD-299, January 1966. 11. Weingart E., "Radiation Buildup on Mechanisms and Thermal Barriers," WAPD-PWR-TE-145, June 1963. 12. Indian Point 1 Semi-Annual Operations Reports, September 1966, September 1967, March 1968, September 1968. 13. "Test Data Sheets, Maine-Yankee Core Crud Removal," CENPD-113 , August 13, 1973. 14. Uhl, D.L., "Oconee Radiochemistry Survey Program," Semi-annual Report July-December 1973, May 1974. 15. Uhl, D.L., "Oconee Radiochemistry Survey Program," Semi-annual Report January-June 1974, May 1975. 16. Connecticut Yankee Monthly Operating Reports, February 1968, March 1968, June 1968, July 1968, December 1968, January 1969, March-May 1969, August 1969, October 1969, December 1969, March 1970, October 1970, November 1970. 17. San Onofre Monthly Operating Reports, December 1969, January 1970, January-March 1971, June-September 1971, November 1971.
WSES-FSAR-UNIT-3 11.1-13 Revision 14 (12/05)
SECTION 11.1: REFERENCES (Cont'd) 18. Yankee Rowe Monthly Operating Reports, February-June 1969, August-December 1969, January-December 1970, January 1972, April-July 1972. 19. "Large Closed-Cycle Water Reactor Research and Development Program," Progress Report WCAP-3620-12 , January 1, 1965 March 31, 1965. 20. ANL-7450 Chemical Engineering Division Research Highlights, May 1967 - April 1968.
- 21. Omaha Semi-Annual Reports, 1973-1975.
- 22. Maine Yankee Daily Log Sheets, June 1972 - April 1975.
- 23. Point Beach Semi-Annual Reports, June 1971 - January 1974.
- 24. H.B. Robinson Semi-Annual Reports, June 1971 - January 1975.
- 25. Ginna Semi-Annual Reports, June 1971 - January 1975.
- 26. "Source Term Data for Westinghouse Pressurized Water Reactors," WCAP-8253 , May 1974. 27. Data Transmittal at C-E, KWU July 1975, Experts Meeting No. 192 at Erlangen. (DRN 99-2361, R11)28. NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR - GALE Code). 29. WSES-3 cycle 9 Nitrogen-16 and Tritium Production, QR-319-04 transmitted by memo dated May 14, 1997, R.B. Lang to Gary Dolese. 30. C-CE-4034, W.D. Mahinney (CE) to R.K. Stampley (Ebasco), dated February 16, 1977. (DRN 99-2361, R11) (DRN 03-2065, R14)31. ANSI/ANS-18.1 - 1999, Radioactive Source Term for Normal Operation for Light Water Reactors
- 32. "ORIGEN2 - Isotope Generation and Depletion Code - Matrix Exponential Method", CCC-371, Oak Ridge National Laboratory, Oak Ridge, Tennessee, September 1983 (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-1 Revision 14 (12/05)
BASIS FOR REACTOR COOLANT RADIOISOTOPE CONCENTRATIONS Parameter Value (DRN 03-2065, R14)
Reactor power level, (Kwt) (a)3735 Equilibrium fuel cycle, equivalent, full power days (a) 590 Average thermal fission rate, fission/Mw-sec (a) 3.12E+16 Average thermal neutron flux, n/cm 2 - sec (a) 4.96E+13 Reactor coolant mass (include pressurizer and CVCS), lbm 5.56E+05 Percent failed fuel (a), % 1.0 Purification flowrate, gal/min 40 BOL Boron Concentration, ppm 2000 Boron reduction rate due to fuel burnup, ppm/sec 3.865E-05 Ion exchanger removal DF (b) Eff (%)(b) Br, I and Te 122 99.18 Xe, Kr and H-3 1 0 Rb and Cs 2 50 All Others 50 98 Fission product escape rate coefficients (sec
-1) Noble gases 6.5E-08 Cs, Rb 1.3E-08 Halogens 1.3E-08 Ba, Sr 1.0E-11 Te 1.0E-09 Mo 2.0E-09 All others 1.6E-12 Cold shutdowns None a. Parameter used in analysis of maximum activities only. Value for normal operations including anticipated operational occurrences is assumed to be consistent with ANS-18.1. b. Nuclides are also removed from the letdown flow via the CVCS lithium removal ion exchanger. This ion exchanger is used in series with the CVCS purification ion exchanger during approximately 20% of the core cycle. Credit for this reduction is not taken in determining the maximum RCS specific activity. (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-2 Revision 14 (12/05)
MAXIMUM REACTOR COOLANT RADIOISOTOPE CONCENTRATION ONE PERCENT FAILED FUEL, NO CONTINUOUS GAS STRIPPING OF VOLUME CONTROL TANK Nuclide Activity (Ci/cm 3)(a)Nuclide Activity (Ci/cm 3)(a)(DRN 99-2361, R11; 03-2065, R14)H-3 9.5E+00 (DRN 99-2361, R11)Br-84 2.10E-02 I-132 9.50E-01 Kr-85m 1.10E+00 Te-132 3.60E-01 Kr-85 5.50E+00 I-133 5.60E+00 Kr-87 8.70E-01 I-134 4.80E-01 Kr-88 2.40E+00 Cs-134 5.20E-01 Rb-88 2.40E+00 I-135 2.70E+00 Sr-89 5.90E-03 Cs-136 9.80E-02 Sr-90 3.70E-04 Cs-137 7.90E-01 Y-90 1.30E-04 Xe-131m 5.40E+00 Y-91 8.80E-04 Xe-133 3.30E+02 Y-91M 3.80E-03 Xe-135m 6.80E-01 Sr-91 6.40E-03 Xe-135 7.40E+00 Zr-95 9.80E-04 Xe-138 5.70E-01 Mo-99 5.10E-01 Ba-140 7.70E-03 Ru-103 3.70E-04 La-140 3.30E-03 Ru-106 1.70E-04 Pr-143 1.00E-03 Te-129 1.10E-02 Ce-144 8.50E-04 I-131 4.80E+00 Mn-54 1.89E-04 Co-60 4.06E-04 Cr-51 5.37E-03 Fe-59 7.26E-05 Co-58 1.76E-02 a. At 70~ F b. The RCS activity limits provided in Technical Specification Section 3/4.4 are more restrictive than the activities presented in this table that are based on one percent fuel failure. (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-3 Revision 14 (12/05)
AVERAGE REACTOR COOLANT RADIOISOTOPE CONCENTRATION*(NO GAS STRIPPING)
PRIMARY SECONDARY (a)NUCLIDE (Ci/ml)(Ci/ml)(DRN 03-2065, R14)
Cr-51 6.57E-03 3.32E-07 Mn-54 3.43E-03 1.67E-07 Fe-55 2.57E-03 1.26E-07 Fe-59 6.38E-04 3.07E-08 Co-58 9.82E-03 4.88E-07 Co-60 1.14E-03 5.67E-08 Np-239 4.23E-03 1.96E-07 (DRN 99-2361, R11)
H-3 1.00E+00 1.00E-03 (DRN 99-2361, R11)
Sr-89 2.98E-04 1.46E-08 Y-91 1.11E-05 5.39E-10 Mo-99 1.25E-02 5.91E-07 Tc-99m 6.86E-03 2.08E-07 Te-129m 4.03E-04 1.99E-08 Te-129 3.09E-02 3.95E-07 Te-131m 2.70E-03 1.19E-07 I-131 4.16E-03 2.03E-07 Te-132 3.35E-03 1.58E-07 I-132 8.02E-02 1.61E-06 I-133 4.49E-02 1.93E-06 I-134 1.28E-01 1.29E-06 Cs-134 7.32E-05 3.57E-09 I-135 8.14E-02 2.67E-06 Cs-136 1.67E-03 8.16E-08 Cs-137 1.05E-04 5.25E-09 Kr-85M 1.98E-02 nil (b)Kr-85 9.71E-01 nil Kr-87 2.10E-02 nil Kr-88 2.22E-02 nil Xe-131m 1.03E+00 nil Xe-133m 8.89E-02 nil Xe-133 3.82E-02 nil Xe-135m 1.60E-01 nil Xe-135 8.31E-02 nil Xe-137 4.19E-02 nil Xe-138 7.52E-02 nil (DRN 99-2361, R11)(a) These values are for the steam generator. Main steam activity is 0.001 of the steam generator liquid activity for all particulate isotopes, 0.01 for all halogens and 1.0 for noble gas and
tritium. (DRN 99-2361, R11)(b) As identified in ANS-18.1 (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-4 Revision 14 (12/05) (DRN 03-2065, R14)Note: The following is historical information pursuant to NEI-98-03, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 03-2065, R14)
PHYSICAL PARAMETERS OF LONG-LIVED ISOTOPES IN CRUD Isotope T 1/2(days)-1 Parent Reaction
/dis E(mev) Co-60 5.26yr 3.6.
(-4) Co-59 n 2.00 1.25 Co-58 71.4d 9.73
(-3) Ni-58 n,p 1.00 0.81 Mn-54 313d 2.21
(-3) Fe-54 n,p 1.00 0.84 Cr-51 27.8d 2.49
(-2) Cr-50 n 0.10 0.32 Fe-59 45d 1.54
(-2) Fe-58 n 1.00 1.18 Zr-95 65.5d 1.06
(-2) Zr-94 n 2.00 0.75 (DRN 03-2065, R14)
End of Historical Information (DRN 03-2065, R14)
WSESFSARUNIT3 Table 11.15 Revision 14 (12/05) MEASURED RADIOACTIVE CRUD ACTIVITY (dpm/mgcrud)(DRN 03-2065, R14)Note: The following is historical information pursuant to NEI9803, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 03-2065, R14) Crud Reactor Co60 Co58 Mn54 Cr51 Fe59 Hf181 Zr95 Cu64 ppb Ref _________________________________________________________________________________________________________________________________________________________Conn. Yankee (a) Ave 9.1(+6)(c) 9.9(+7) 2.3(+6) 1.3(+7) 2.8(+6) 85 16 Max 2.5(+7) 4.0(+8) 1.2(+7) 3.6(+7) 1.5(+7) 4000 San Onofre, (a) Ave 2.0(+6) 2.2(+7) 1.4(+6) 3.1(+6) 6.7(+5) 90 7 Max 2.0(+7) 1.2(+8) 4.2(+6) 2.0(+7) 3.8(+6) 4000 Yankee Rowe, (a) Ave 6.7(+6) 3.3(+7) 4.5(+6) 1.7(+7) 5.5(+6) 6.6(+5) 70 18,9 Max 2.1(+7) 1.2(+8) 1.9(+7) 1.4(+8) 1.8(+7) 1.8(+6)
Saxton, (b) Ave 4.3(+6) 2.7(+7) 3.9(+6) 9.0(+7) 1.2(+6) 55 7,8 Max 2.2(+7) 1.5(+8) 1.4(+7) 1.1(+8) 6.0(+6). 250 9,19 Shippingport, (a) Ave 2.3(+7) 2.8(+6) 1.3(+6) 2.2(+6) 1.8(+6) 5.2(+5) 7.0(+5) 75 10,11 Max 4.8(+7) 3.2(+6) 1.7(+6) 2.2(+6) 1.8(+6) 7.6(+5) 9.7(+5)
Indian Point I, (a)Ave 1.8(+6) 4.6(+6) 7.7(+5) 5.7(+6) 2.2(+6) 1.5(+5) 2.3(+5) 3.1(+9) 77 12 Max 2.9(+6) 9.1(+6) 2.0(+6) 8.2(+6) 3.3(+6) 4.2(+5) 1.2(+10)
MaineYankee, (a) Ave 1.29(+6) Max 2.2(+6) 4.53(+7) 9.70(+5) 4.24(+7) 2.03(+6) 7.26(+6) 41 13 Oconee, (a) Ave 2.8(+6) 5.1(+7) 5.5(+5) 2.9(+7) 2.4(+5) 5.6(+6) 25 Max 2.3(+7) 1.9(+8) 1.1(+7) 1.5(+8) 1.7(+6) 8.7(+6) 100 14,15 Average Crud (ppb) 68 (d) 75
- a. Circulating crud. b. Deposited crud on fuel rods with exception of Cr51 (ave, max) and Fe59 (ave) which are circulating. C. () denotes power of 10.
- d. Does not include Oconee data. (DRN 03-2065, R14)
End of Historical Information (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-6 Revision 14 (12/05)
ACTIVATION RATES(DRN 03-2065, R14)Note: The following is historical information pursuant to NEI-98-03, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 03-2065, R14)
Activation Rates,i (d/gm-sec)
________________________________________________________________ Reactor Co-60 Co-58 Mn-54 Cr-51 Fe-59 Zr-95 A T /A C ________________________________________________________________________________________________________________
Conn. Yankee (a) 1.90(+10) (b) 7.00(+10) 1.40(+9) 2.90(+10) 1.90(+8) -
4.10 San Onofre (a) 1.90(+10) 7.00(+10) 1.40(+9) 2.90(+10) 1.90(+8) -
4.10 Yankee Rowe 1.70(+10) 1.50(+10) 4.34(+9) 1.90(+10) 3.50(+8) 7.50(+8) 3.13
Saxton 8.00(+9) 1.00(+10) 2.95(+9) 1.30(+10) 2.40(+8) -
5.26 Indian Point 1 6.6(+9) 1.3(+10) 3.7(+9) 1.1(+10) 2.0(+9) 1.4(+8) 4.53 Maine-Yankee 6.5(+9) 6.1(+10) 5.2(+8) 1.9(+10) 6.3(+7) 3.8(+8) 5.44 Oconee 1.3(+10) 1.00(11) 3.1(+9) 9.8(+10) 9.5(+8) 3.1(+9) 4.00
________________________________________________________________________________________________________________
- a. Conn. Yankee and San Onofre fluxes and area ratios assumed the same
- b. () denotes power of 10. (DRN 03-2065, R14)
End of Historical Information (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-7 Revision 14 (12/05)
SYSTEM PARAMETERS(DRN 03-2065, R14)Note: The following is historical information pursuant to NEI-98-03, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 03-2065, R14)
__________________________________________________________________________________________________________________________ Steam Generator Therma; Flux Fast Flux Reactor Tubing Core Cladding (n/cm 2-sec) (n/cm 2-sec)A T/A C__________________________________________________________________________________________________________________________
Conn. Yankee Inconel Stainless Steel 4.0(+13) (a) 1.8(+14) 4.10 San Onofre (b) Inconel Stainless Steel 4.4(+13) 1.8(+14) 4.10 Yankee Rowe Stainless Steel Zircaloy 3.9(+13) 2.6(+14) 3.13 Saxton Stainless Steel Stainless Steel 1.8(+13) 1.2(+14) 5.26 Shippingport Stainless Steel Zircaloy 5.1(+13) 1.5(+14) 2.44 Indian Point I Stainless Steel Stainless Steel (c) 1.5(+13) 1.5(+14) 4.53 Maine-Yankee Inconel Zircaloy 3.6(+13) 1.6(+14) 5.44 Oconee Inconel Zircaloy 3.6(+13) 1.5(+14) 4.00
__________________________________________________________________________________________________________________________ a. () denotes power of 10. b. San Onofre fluxes and area ratio assumed same as Conn. Yankee.
C. Zircaloy box around each fuel assembly. (DRN 03-2065, R14)
End of Historical Information (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-8 Revision 14 (12/05)
AVERAGE AND MAXIMUM RESIDENCE TIMES, DAYS(DRN 03-2065, R14)Note: The following is historical information pursuant to NEI-98-03, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 03-2065, R14)
________________________________________________________________________________________ Reactor Co-60 Co-58 Mn-54 Cr-51 Fe-59 (a) Zr-95 ________________________________________________________________________________________
Conn. Yankee Ave 92 10 54 1 Sat - -
Max 262 51 390 4 Sat - -
San Onofre Ave 20 2 32 1 18 - -
Max 207 13 104 2 Sat - -
Yankee Rowe Ave 58 13 25 2 111 6 Max 185 56 116 19 Sat 17 Saxton Ave 25 5 10 38 38 - -
Max 136 30 38 54 Sat - -
Shippingport Ave 115 1 8 1 10 2 Max 246 1 11 1 10 3 Indian Point I Ave 58 3 7 2 115 13 Max 94 6 19 2 Sat 24 Maine-Yankee Ave - - - - - 34 Max 87 7 84 9 Sat -
Oconee Ave 41 3 5 1 1 12 Max 356 13 118 4 8 66 Ave of Max T res 166 (c) 22 110 12 Sat 29 140 23 110 13 Sat 20
__________________________________________________________________________________________ a. Fe-59 isotope reaches saturation before erosion from core surfaces. b. Included in Zr-95 average of Maximum (T res). c. Lower values do not include Oconee data. (DRN 03-2065, R14)
End of Historical Information (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-9 Revision 14 (12/05) (DRN 03-2065, R14)
ACTIVATION RATES Reaction Activation Rate i activations/gm-sec 59 Co (n, )60 Co 9.34E+09 58 Ni (n, p) 58Co 1.10E+11 54 Fe (n, p) 54Mn 1.05E+09 50 Cr (n, )51 Cr 2.78E+10 58 Fe (n, )59 Fe 9.86E+07 94 Zr (n, )95 Zr 7.45E+08(DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-10 Revision 14 (12/05) (DRN 03-2065, R14)
LONG-LIVED CRUD ACTIVITY
__________________________________________________________________________________________
Isotope T res (days) Half Life Act.(dpm/mg)
__________________________________________________________________________________________Co-60 166 5.26 y 7.89E+06Co-58 22 71.4 d 3.42E+08 Mn-54 110 313 d 3.67E+06 Cr-51 12 27.8 d 1.04E+08 Fe-59 267 45 d 1.41E+06 Zr-95 29 65.5 d 2.86E+06
__________________________________________________________________________________________(DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-11 Revision 14 (12/05) (DRN 03-2065, R14)
REACTOR COOLANT CIRCULATING CRUD ACTIVITY (a)________________________________________________________________________________________ Isotope Act, (Ci/cm 3)________________________________________________________________________________________ Co-60 2.66E-04 Co-58 1.15E-02 Mn-54 1.24E-04 Cr-51 3.52E-03 Fe-59 4.76E-05 Zr-95 9.66E-05 a. Reactor coolant temperature is 70°F. Crud level 75 ppb (DRN 03-2065, R14)
WSES-FSAR-UNIT-3Table 11.1-12EQUILIBRIUM CRUD FILM THICKNESSThicknessLocation(mg/cm 2)Vessel Internals, Piping, SG Inlet Plenum1.00(+0)PressurizerLower Head6.5(-1)Surge Line1.20(+0)CEDM, Vessel Head ICI Tops3.00(-1)SG Tubing1.00(-l)
Regenerative HX3.50(-1)
Letdown HX3.00(-2)
Shutdown Cooling HX3.00(-2)
Spent Fuel Pool HX3.00(-3)
WSES-FSAR-UNIT-3 TABLE 11.1-13 Revision 14 (12/05) TRITIUM ACTIVATION REACTIONS Threshold Energy (DRN 03-2065, R14)Reaction (MeV) Cross-section (barns)(a)1.10 B (n, 2)3H 1.9 4.2E-02 2.7 Li (n, n)3H 3.9 3.85E-01 3.6 Li (n, )3H Thermal 9.50E+02 4.2 H (n, )3H Thermal 5.50E-04 5.11 B (n, 3 H)9Be 10.4 8.00E-06 6.14 N (n, 3 H)12c 4.3 3.00E-04 a. Threshold cross sections (3) (19)(DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-14 Revision 14 (12/05) PARAMETERS USED IN TRITIUM PRODUCTION DETERMINATION Parameters Value(DRN 03-2065, R14)
Core Channel Coolant Volume, cm 3 2.08E+07 Fission Rate, fission/MW-sec 3.12E+16 Fast Neutron Flux >0.625 ev, n/cm 2-sec 3.77E+14 Thermal Neutron Flux n/cm 2-sec 6.23E+13 Lithium concentration, ppm 2.2
! 0.15 Lithium-6 abundance, percent 0.1 Boron concentration, ppm 900 (a)Power level (Mwt) 3735 a. Cycle average for UO 2 core. (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-15 Revision 14 (12/05)
TRITIUM PRODUCTION IN REACTOR COOLANT(DRN 03-2065, R14)
Reaction Expected (Ci/yr)(b)Maximum (Ci/yr)(c)2 H(n, )3 H 7.67 7.67 6 Li (n, )3 H 271.26 271.26 (DRN 01-458, R11-A) 7 Li (n, )3 H(DRN 01-458, R11-A) 22.19 22.19 10 B (n, 2 )3 H 501.40 501.40 Fission (a) 502.32 1255.80Total 1304.84 2058.32 (Maximum RCS Flow)
Reaction Expected (Ci/yr)(b)Maximum (Ci/yr)(c)2 H(n, )3 H 7.56 7.56 6 Li (n, )3 H 267.62 267.62 7 Li (n, )3 H 21.90 21.90 10 B (n, 2 )3 H 494.68 494.68 Fission (a) 502.32 1255.80Total 1294.08 2047.56 a. For UO 2 core b. Two percent tritium release from fuel c. Five percent tritium release from fuel (DRN 03-2065, R14)
WSESFSARUNIT3 Table 11.116 (Sheet 1 of 2)
Revision 14 (12/05)
TRITIUM PRODUCTION AND RELEASE AT OPERATING PWRs(DRN 03-2065, R14)Note: The following is historical information pursuant to NEI9803, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 03-2065, R14)
Total Calculated Total (b) Calculated Produced due Release Measured Production due to Capture From Cycle Production (a) to Fissions in RCS Fuel Operating PWR No. (Ci/cycle) (Ci/cycle) (Ci/cycle) (Percent)Maine Yankee 1 305.3 11,720 370.0 Li conc approx 0 2 59.8 6,510 155.8 OmahaLi conc approx 0 1 192.6 6,100 153.9 0.6 PalisadesLi conc approx 0 1 440 10,890 343.8 0.9 Obrigheim, KWO
Li conc assumed 2 ppm 1 662 6,540 257.1 6.2 2 239 5,120 86.0 3.0 3 391 5,680 103.6 5.1 4 314 6,070 110.8 3.3 5 199 5,700 82.4 2.1 Stade, KKS Li conc assumed 2 ppm 1 408 10,490 300.3 1.0 2 131 8,050 157.4 Oconee Li conc 0.5 ppm 1 325 8,050 335.7 Ginna Li conc assumed 2 ppm 1(c) 1410 (e) 15,570 384.1 6.6 3 449 9,830 216.2 2.4 4 105 5,260 115.7 (DRN 03-2065, R14)
End of Historical Information (DRN 03-2065, R14)
WSESFSARUNIT3 Table 11.116 (Sheet 2 of 2)
Revision 14 (12/05)
TRITIUM PRODUCTION AND RELEASE AT OPERATING PWRs(DRN 03-2065, R14)Note: The following is historical information pursuant to NEI9803, which is identified by a designation of "Start" and "End".
Start of Historical Information (DRN 03-2065, R14)
Total Calculated Total (b) Calculated Produced due Release Measured Production due to Capture From Cycle Production (a) to Fissions in RCS Fuel Operating PWR No. (Ci/cycle) (Ci/cycle) (Ci/cycle) (Percent)
Point Beach I and 2Li conc assumed 2 ppm 11 (d) 943 11,470 381.9 4.9 12,21 1269 19,060 558.2 3.7 H.B. Robinson
Liconc assumed 2 ppm 1 777 12,090 373.1 3.3 2 604 11,980 288.2 2.6 3 247 7,050 204.4 0.6 a. Production is total measured tritium discharges plus measured system inventories b. Fission curies are based on appropriate cycle average fraction fission of U235, U238 and Pu C. Includes cycles 1A, IB, and 2
- d. (11) Unit No. I Cycle No. 1
- e. 1410 Ci accounted for under tritium measurement program.
(24) Only 800 Ci can be accounted for using plant discharges and inventories (DRN 03-2065, R14)
End of Historical Information (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-17 Revision 14 (12/05)
FISSION AND CORROSION PRODUCT ACTIVITIES IN THE SPENT FUEL POOL (Ci/gram)(DRN 01-458, R11-A; 03-2065, R14)
Nuclide WSES-FSAR-UNIT-3Table 11.1-18Revision 11-A (02/02)
EQUIPMENT LEAKAGE ASSUMPTIONS Valves(DRN 00-1045)
Seat Leakage 10 cc/hr/in. Seat Circumference Stem Leakage 10 cc/hr/in. Stem Circumference(DRN 00-1045)
Pumps Centrifugal50 cc/hr Positive Displacement 1 gallon/hr Pump Flanges30 cc/hr WSES-FSAR-UNIT-3 Table 11.1-19 Revision 307 (07/13)
BASIS FOR DESIGN BASIS ACTIVITIES FOR THE STEAM GENERATOR (DRN 03-2065, R14)Parameter Value Reactor Coolant Activities 1% failed fuel See Table 11.1-2 Primary to Secondary Leakage 0.75 gal/min Steam Generator Blowdown 165 gal/min (EC-8458, R307)Steam Generator Liquid Mass 3.61+05 lbs (EC-8458, R307)
Steam Generator Partition Factors Iodines 0.01 Particulates 0.001 Main Steam Rate 1.66E+07 lbs/hr (DRN 03-2065, R14)
WSES-FSAR-UNIT-3 Table 11.1-20 Revision 14 (12/05)
STEAM GENERATOR LIQUID ACTIVITIES (DESIGN BASIS)