ML16257A146
| ML16257A146 | |
| Person / Time | |
|---|---|
| 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: ML16257A146 (16) | |
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WSES-FSAR-UNIT-3 15B-i Revision 14 (12/05)
APPENDIX 15B CONTAINMENT LEAKAGE AND DOSE CALCULATION MODELS TABLE OF CONTENTS SECTION TITLE PAGE 15B.1 CONTAINMENT LEAKAGE AND DOSE CALCULATION MODELS 15B-1 (DRN 04-704, R14)15B.1.1 GOVERNING EQUATIONS 15B-1 15B.1.2 PRIMARY CONTAINMENT LEAKAGE PATHWAYS 15B-4 15B.1.3 SECONDARY STEAMING RELEASE MODEL 15B-5 15B.2 TEDE DOSE CALCULATIONS 15B-6 15B.3 MAIN CONTROL ROOM DOSE CALCULATION MODELS 15B-7 15B.4 REFERENCES 15B-9 (DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-ii Revision 14 (12/05)
APPENDIX 15B
LIST OF TABLES
TABLE TITLE 15B-1 ISOTOPE PROPERTIES
(DRN 04-704, R14) 15B-2 ICRP 30 ISOTOPE PROPERTIES
15B-3 POWERS 10% NATURAL DEPOSITION MODEL
(DRN 05-1551, R14) 15B-4 EFFECTIVE DOSE EQUIVALENT WEIGHTING FACTORS (DRN 04-704, R14; 05-1551, R14)
WSES-FSAR-UNIT-315B-iiiAPPENDIX 15BLIST OF FIGURESFIGURETITLE15B-1CONTAINMENT LEAKAGE DOSE MODEL WSES-FSAR-UNIT-3 15B-1 Revision 14 (12/05)
APPENDIX 15B 15B.1 CONTAINMENT LEAKAGE AND DOSE CALCULATION MODEL (DRN 04-704, R14)
The RADTRAD computer code, as documented in NUREG/CR-6604, was developed for the US NRC in support of the Alternative Source Term project. The code calculates the time dependent concentration of airborne radioactivity in the containment atmosphere, integrated release of activity to the environment, and the Total Effective Dose Equivalent (TEDE), to individuals at specified locations. Only TEDE doses
are required to be calculated to demonstrate conformance to 10CFR50.67, Accident Source Terms.
The dose conversion factors used to determine TEDE doses are based on Federal Guidance Reports
FGR-11 and FGR-12.
15B.1.1 Governing Equations
The radiological consequences of an accident in a nuclear reactor depend upon the quantity of the radioactive material that escapes to the environment or enters the control room. RADTRAD is designed to calculate doses at offsite locations, such as the low population zone, and in the control room. As material is transported through the containment and other buildings, credit is given for several natural and engineered removal mechanisms. Containment sprays remove iodine. The flow between buildings or rooms may be through HEPA filters or charcoal filters. Leakage to the environment may occur. Aerosols can deposit on surfaces within rooms and also in connection paths. (DRN 05-1551, R14)
The basic equation for radionuclide transport and removal is the same for all compartments. The governing equation for the number of nuclide n , in compartment i , during time step m is provided with all source and sink terms shown for clarity. (DRN 05-1551, R14) i m i , n v m i , v v , n l n v m i , n S N N dt d -m i , n m)forced (j , i m j , wu m n , dep m n , spr n m)conv (j , i L)u j (j N F)t ()t (F100 +m j , n m)conv (j , i m)forced (j , i m j , wi L)i j (j N F F100 1 2 1 2 n n T)(ln where, m j , n N = number of atoms of nuclide n in compartment i during time step m , m i , v N = number of atoms of nuclide v in compartment i during time step m , v , n n v1 = fraction of nuclide v that decays to nuclide n (dimensionless), (DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-2 Revision 14 (12/05)
(DRN 04-704, R14) n = radiological decay constant for nuclide n (s-1), v = radiological decay constant for nuclide v (s-1), 2 1 n T = half-life of nuclide n (s), m)conv (j , i F = volume normalized convection (leakage) air flow rate from compartment j to i , (s-1), m)forced (j , i F = volume normalized forced (filtered) air flow rate from compartment j to i , (s-1), L = number of compartments in the plant model, j Vol = volume of compartment j (m 3), )t (m n , spr = time dependent spray removal coefficient for nuclide n (s-1), )t (m n , dep = time dependent natural deposition removal coefficient for nuclide n (s-1), m i , n S = source injection rate of nuclide n to compartment i during time step m (atoms/s), and m j , wi = filter efficiency associated with nuclide n and pathway from j to i (%). A number of these terms are not applicable to all volumes. For example, fission product removal due to containment spray and/or natural deposition is only credited for reactor containment volumes.
15B.1.1.1 Containment Spray Equations
Sprays may remove aerosols and iodine. It is assumed that noble gases are not affected by containment sprays. The impact of containment sprays to organic iodine is also neglected. The containment spray removal coefficient is modeled in accordance with NUREG-0800, the U.S. N.R.C. Standard Review Plan, Section 6.5.2 guidance.
Removal of particulate (aerosol) iodine due to containment spray is modeled using the following formulas
for the spray removal coefficient:
Vol D E hF)t (m n), part (spr 2 3 where
F = spray flow rate (1750 gpm), Vol = containment volume (2.677E+06 ft 3), h = fall height of the drop (150 feet), and E/D = ratio of a dimensionless collection efficiency to the average drop diameter.
(DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-3 Revision 301 (09/07)
(DRN 04-704, R14)
Since the removal of particulate materials depends ma rkedly upon the relative sizes of the particles and the spray drops, it is convenient to combine paramet ers that cannot be known.
It is conservative to assume (E/D) to be 10 per meter initially (i.e., 1% efficiency for spray drop of 1 millimeter in diameter), changing abruptly to one per meter after the aerosol mass has been depleted by a factor of 50. This results is a particulate spray removal coefficient of 3.596 hr
-1 until the decontamination factor of 50 is reached, and 0.3596 hr
-1 thereafter. (EC-5000081470, R301)
The Waterford 3 Large Break LOCA offsite and control room intake dose calculations conservatively
neglect the removal of elemental iodine from the c ontainment atmosphere via the containment sprays.
However, for the isolated cases (e.g., CVAS filter shine) where it is credited Standard Review Plan guidance is used. The first-order removal c oefficient by spray for elemental iodine is: (EC-5000081470, R301)
D)V(TFK)t(g mn),elem(spr 6 where, K g = gas-phase mass-transfer coefficient, T = time of the fall of the drop, F = containment spray flow rate, V = containment volume, and D = mass-mean diameter of the spray drops.
The SRP states that maximum decontamination fa ctor which may be applied is 200. The removal coefficient for Waterford 3 for elemental iodine was determined to be 20/hr.
15B.1.1.2 Natural Deposition Equations
Reduction of airborne radioactivity by natural deposit ion within the containment may be credited. The methodology used differs between elemental and particu late iodine. As with containment sprays, deposition of organic iodine is neglected.
During injection, the removal of elemental iodine by wall deposition may be estimated by the first-order estimate from the Standard Review Plan VolAK welem,w where, w,elem = the first-order removal coefficient by wall deposition, A = the wetted surface area for the primary containment, Vol = the containment building net free volume, and
K w = the mass-transfer coefficient.
All available experimental data is enveloped if K w is taken to be 4.9 meters per second. If this is assumed then the resulting removal coefficient is 0.4 hr
-1.
For particulate/aerosol iodine, the Powers 10% aeros ol deposition model is specified. This model is described in section 2.2.2.1 of NUREG/CR-6604. The lower bound of this deposition model (10 th percentile) is specified. The Powers 10% model is normalized to reactor ther mal power as presented in Table 15B-4. (DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-4 Revision 301 (09/07)
(DRN 04-704, R14) 15B.1.2 PRIMARY CONT AINMENT LEAKAGE PATHWAYS Post accident containment leakage is limited by Technical Specifications to 0.5 percent of the containment volume per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50 percent of this value for the duration of the
accident. As documented in FSAR Section 6.2, contai nment pressure is reduced to less than half of its peak value by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the event. From a dose evaluation standpoint, this leakage can take any of
the three following pathways:
a) Leakage to the Shield Building annulus wh ich will be treated by the Shield Building Ventilation System.
b) Leakage to controlled ventilation area of the Reactor Auxiliary Building which will be treated by the Controlled Ventilation Area System (once through filtration).
c) Bypass Leakage.
The total amount of all penetration leakage is limited by a further Technical Specification applied to those lines which are potential sources of bypass leakage. (EC-5000081470, R301)
Several relaxations of analysis assumptions in the ca lculations for offsite dose and for main control room personnel dose (due to both inhalation and shine) have not been relaxed for EQ dose calculations.
These include:
- Not crediting SBVS charcoal filtration
- Assuming a 6 minute positive pressure period for the Shield Building
- Allowing Shield Building Maintenance Hatch leakage at 2 days.
These relaxations have been included in the calculations for offsite dose and for main control room personnel dose (due to both inhalation and shine), conservatively increasing these calculated doses.
These assumptions are not included in EQ dose calculations. (EC-5000081470, R301) 15B.1.3 SECONDARY STEAMING RELEASE MODEL (EC-5000081470, R301)
For events that involve fuel failure and steaming rel eases, the source term of FSAR Table 12.2-12A is utilized. This is the source term developed to det ermine the gap fission product activities in peak power fuel rods.
(EC-5000081470, R301)
A maximum allowable peak linear heat rate of 6.
3 kw/ft rod average power is assumed for burnups exceeding 54 GWD/MTU, per RG 1.183 footnote 11.
Most sequences for which the source term dose (o ffsite and Main Control Room) is dominated by the secondary side steaming release pathway (e.g., C EA Ejection, SBLOCA, Inside Containment MSLB) assume a constant primary-to-secondary side leak ra te. The mass release rate is conservatively evaluated at 350°F conditions, corresponding to Shutdown Cooling entry conditions.
For events where the steam generator (SG) is not in a faulted condition, the pressure differential between
the RCS and the secondary side across the SG tubes is about the same or less than that under normal operating conditions. Thus, for non-faulted SG's w hen there is no large primary-to-secondary pressure differential, the operational leakage value specified in Technical Specifications is the appropriate primary-to-secondary leakage value to assume in radiologica l analyses. The actual leak rate during these transients would be the same or, because of r educed differential pressure between primary and secondary sides, less than that for normal operating conditions. Under cert ain conditions, secondary pressure could exceed primary pressure, resulting in backflow from the SG to the RCS. Each event scenario with non-faulted SG's has slightly different assumptions per taining to the primary-to-secondary (DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-5 Revision 301 (09/07)
(DRN 04-704, R14) leak rate, however at nominal RCS conditions a value of 150 gal/day is assumed, except for Small Break LOCA for which the Technical Specification value of 75 gal/day is assumed. For the MSLB or FWLB cases where a significant pressure drop exists across the SG tubes, a larger (540 gal/day) leakage rate is assumed.
Except for SGTR, the dose analyses assuming prim ary-to-secondary leakage assume the following RCS masses:
34,260 Pressurizer Liquid Mass (lbm) 4,016 Pressurizer Steam Mass (lbm) 395,502 Non-Pressurizer Liquid Mass (lbm)
For purposes of determining activity concentration, only the non-pressurizer liquid mass is considered.
For purposes of determining the steaming rate due to cooldown, the liquid and steam masses of the pressurizer are also considered.
For SGTR, a CENTS analysis is performed for the cool down. A total initial RCS mass of 467,000 lb. is assumed for that analysis.
Hot Zero Power steam generator inventories (241,450 lbm per SG) are assumed for most events. The
HZP inventory is larger than the HFP inventory of 153,700 lbm and is considered more representative of the SG mass present in an intact SG that is being cooled to shutdown cooling entry conditions, and maximizes the initial activity present for the case of releases from a faulted SG. For the SGTR, a low
initial SG level (corresponding to 106,300 lbm per SG) wa s conservatively assumed since this results in earlier uncovery of the top of the SG tubes, which results in increased flashing and a lower decontamination factor. This is a very conservative value that corre sponds approximately to the reactor trip setpoint on Steam Generator Level Low. (EC-3277, R301)
For events where ADV's from both SG's are credited fo r steaming (e.g., CEA Ejection), it is assumed that there is an equal release from each ADV. (EC-3277, R301)
(EC-5000081470, R301)
The events explicitly analyzed that involve assumed releases from the ADV's include Small Break LOCA, SGTR, CEA Ejection, Inside Containment Main Steam Line Break, and the combined scenario of Feedwater Line Break / Outside Containment Main Steam Line Break. For secondary steaming release scenarios other than SGTR and FWLB/OC-MSLB, it is assumed that operators act at 30 minutes into the event to begin plant cooldown to shutdown cooling ent ry conditions of 350°F. A maximum cooldown rate of 100°F is assumed. This is the maximum allow ed cooldown rate for forced circulation conditions and conservatively bounds the plant procedural limits of 50° F/hour for natural circulation conditions. A rapid cooldown increases the releases early in the event, thus increasing t he releases subject to the higher X/Q's assumed for the initial two hours of the event. (EC-5000081470, R301)
The event duration for these secondary steaming cases is assumed to be 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This would bound the time required for shutdown cooling initiation (and t hus for terminating steaming releases from the SG for decay heat removal) based on a nominal cooldown rate of 40°F/hour for natural circulation conditions. Thus, for most events, it is assumed that the plant is rapidly cooled to shutdown cooling entry conditions, but that plant conditions are maintained by steaming at those conditions for up to 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event.
For the FWLB/OC-MSLB, described in FSAR section 15.2.
3.1, it is assumed that operator action to cool the plant starts at 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and shutdown cooling in itiation is assumed at 8.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> into the event.
Steam releases for individual secondary steaming events are presented in the input table for the specific events. Note that because similar cooldown sc enarios are used, results depend on whether or not a Loss of Offsite Power (LOOP) is assumed or on how many reactor coolant pumps are assumed operating (and providing energy input to the RCS) during the cooldown. (DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-6 Revision 301 (09/07)
(DRN 04-704, R14)
The cooldown scenario for SGTR is described specifically in FSAR Section 15.6.3.2. (DRN 04-704, R14)
(DRN 05-1551, R14) 15B.1.4 ENGINEERING SAFETY FEATURES (ESF) LIQUID LEAKAGE (DRN 05-645, R14; EC-5000081470, R301; EC-3277, R301)
The large break LOCA dose model also accounts fo r liquid leakage from ESF injection systems as required by Regulatory Guide 1.183. Specifically, a 0.5 gpm leak of safety injection sump water is assumed to occur within the CVAS boundary, therefore this leakage represents a filtered release. This leakage is assumed to begin roughly 23.4 minutes in to the event. For offsite and control room submersion/inhalation dose calculations, a ten percent (10%) flashing fraction is assumed, i.e., ten percent of the iodine in the leakage is assumed to go ai rborne. The ten percent fl ashing fraction is also assumed for determination of EQ doses. The ten percent flashing fraction is also assumed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the calculation of filter shine doses for Main Control Room personnel. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a flashing
fraction of two percent (2%) is assumed for the filt er shine dose calculations for Main Control Room personnel.* Note that the 2 percent value is roughly a factor of ten greater than the maximum that would be expected from the post-LOCA sump temperature profile. For off-site and control room
submersion/inhalation dose calculations the ESF l eakage is directed to the environment with no credit being taken for mixing or dilution in the reactor auxiliary building. For filter shine doses, mixing in the reactor auxiliary building is credited. Dose rates are determined based on a deterministic model of
radionuclide buildup on the CVAS filter train, and t he total dose to control room operators is then determined using simple numerical integration techniques. (EC-5000081470, R301; EC-3277, R301)
- In the Safety Evaluation Report for Amendment 198, the NRC did not find the assumption of two percent airborne iodine acceptable but approv ed the filter shine analysis results due to compensating conservatisms. The NRC reco mmended that the flashing fraction be increased to ten percent or the lower value of two percent be justified when the large break LOCA filter shine dose analyses are revised in the future. (DRN 05-645, R14; 05-1551, R14)
(DRN 04-704, R14) 15B.2 TEDE DOSE CALCULATIONS
The Alternative Source Term dose methodology r egulated by 10CFR50.67 has dose acceptance criteria in Total Effective Dose Equivalent (TEDE).
The dose acceptance crit eria is as follows:
(i) An individual located at any point on t he boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 re m) total effective dose equivalent (TEDE).
(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulti ng from the postulated fission product release (during the entire period of its passage), woul d not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(iii) Adequate radiation protection is provi ded to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
TEDE doses are the sum of the "Deep Dose Equivalent" (DDE) for external exposures and the
"Committed Effective Dose Equivalent
" (CEDE) for internal exposures.
TEDE = DDE + CEDE (DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-7 Revision 301 (09/07)
(DRN 04-704, R14)
The DDE is calculated using a "semi-infi nite" model and is determined as follows:
D c,n = C n (t)dt(DCF), where
C n (t) = the instantaneous concentration of nuclide n , DCF = is the dose conversion factor for nuclide n , and dt = the time interval of interest.
The concentration is assumed to be constant in a ll directions (above the ground). The DCFs are based on Federal Guidance Report (FGR12).
The CEDE doses are the sum of the products of the weighting factors (w T) applicable to each of the body organs or tissues that are irradiated, and the co mmitted dose equivalent to these organs or tissues.
TTT EHwH where H T is the committed dose equivalent, or the dos e equivalent to tissues of reference (T) that will be received from an intake or radioactive material by an individual during the 50-year period following the intake. The DCFs to determine these doses are based on FGR 11, which is in turn based on ICRP Publication 30, "Limits for Intake s of Radionuclides by Workers."
15B.3 MAIN CONTROL ROOM CALCULATIONS
15B.3.1 CONTROL ROOM VENTILATION MODEL
The Control Room Air Conditioning System has a single normal air intake and two widely separated air
inlets for emergency operation. The system is a z one isolation system with filtered recirculated air and provisions for maintaining positive pressure in the control room envelope under accident conditions.
Intake air for pressurization is filtered before entering the control room envelope.
During normal operation, air is drawn into the syst em through the single normal air intake. Air handling units condition the air for distribution into the control room envelope, and exhaust fans discharge air from
the control room envelope to the environment.
The Control Room Air Conditioning System is automat ically isolated on a Safety Injection Actuation Signal (SIAS) or by high radiation detected outside t he normal air intake. Automatic isolation involves shutting the normal outside air intake and exhaust isolation valves, stopping the normally operating exhaust fans, opening recirculation dampers, and st arting both emergency filtration units in the recirculation mode. Operation of only one emergency filtration unit is required. The control room operators may then initiate the pressurization m ode of operation by opening a valve in one of the emergency air intakes. The intake air is ducted to the operating filtration unit before being discharged to
the control room envelope. The emergency filtration uni t continues to recirculate air within the control room envelope while providing filtered pressurization air flow. (EC-5000081470, R301)
Various control room and Control Room Air Conditioning System parameters are input to the control room dose analyses. Dependent on the particular scenario, a minimum (168,500 ft
- 3) or a maximum (220,000 ft 3) control room envelope volume is assumed in the c ontrol room dose evaluation. Recirculation flow through the filter of the operating emergency filtrati on unit is assumed to be 3,800 Cubic Feet per Minute (CFM). Consistent with Technical Specification limits, filter effici ency of 99% is assumed for aerosol, elemental, and organic iodine. No filtration of noble gases is credited. The pressurization flow rate is assumed to be 225 CFM.
(DRN 04-704, R14; EC-5000081470, R301)
WSES-FSAR-UNIT-3 15B-8 Revision 301 (09/07)
(DRN 04-704, R14; EC-5000081470, R301)
Unfiltered in-leakage to the control room envelope distributes through the volume and is processed through the emergency filtration unit with filter effi ciencies assumed as above. Control room out-leakage is assumed to be equal to the unfiltered in-leakage.
Unfiltered inleakage applies the most conservative at mospheric dispersion factor for determining the dose consequences from radionuclide leakage.
The dose evaluations are performed such that the control room envelope can be either in the recirculation mode or pressurization mode from the beginning of the
event. For the pressurization flow, it is assumed that the control room operator selects the more favorable air intake, therefore the lowest atmospheric dispersion factor is reduced by a factor of four as allowed by S.R.P. Section 6.4 and Regulatory Guide 1.183. (EC-5000081470, R301)
15B.3.2 CONTROL ROOM DOSE CALCULATION
The dose to a hypothetical individual in the cont rol room is calculated based on the time-integrated concentration in the control room compartment. The ai r immersion dose in the control room is adjusted to a "finite" model using a "geometry factor":
, GF DCFdt)t(CD n CRn,c where GF = the geometry factor =
3380 1173.CR V Inhalation doses in the control room are calculated as follows:
n,i CRn,t xBRxOFxDCF)t(CD where BR = breathing rate for control room operators, V = control room volume, and OF = control room occupancy factor.
The breathing rate assumed is 3.47x10
-4 m 3 per second for the main control room. The control room occupancy factor is assumed to be 1.0 for the firs t 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 0.6 from 1 to 4 days, and 0.4 for the remaining duration of the event.
(DRN 04-704, R14)
WSES-FSAR-UNIT-3 15B-9 Revision 301 (09/07)
(DRN 04-704, R14) 15B.4
REFERENCES:
15B-1 ICRP, "Limits for Intakes of Radionucli des by Workers," ICRP Publication 30, 1979.
15B-2 K.F. Eckerman et. al., "Limiting Values of Radionuclide Intake and Air Concentration and dose Conversion Factors for Inhalat ion, Submersion and Ingestion," Federal Guidance Report 11, Environmental Protection Agency, 1988.
15B-3 K.F. Eckerman and J.C. Rym an, "External Exposure to Radionuclides in Air Water and Soil," Federal Guidance Report 12, Envi ronmental Protection Agency, 1993.
(EC-5000081470, R301)
(DRN 04-704, R14; EC-5000081470, R301)
WSES-FSAR-UNIT-3 TABLE 15B-1 Revision 14 (12/05)
ISOTOPE PROPERTIES(DRN 04-704, R14)
Isotope Half-life (s) (MeV/dis) (MeV/dis)I-131 6.9466E+05 0.381 0.194 I-132 8.2800E+04 2.283 0.496 I-133 7.4880E+04 0.608 0.410 I-134 3.1560E+03 2.624 0.623 I-135 2.3796E+04 1.580 0.367 Kr-83m 6.5880E+03 0.002 0.037 Kr-85m 1.6127E+04 0.159 0.258 Kr-85 3.3830E+08 0.002 0.251 Kr-87 4.5780E+03 0.793 1.324 Kr-88 1.0224E+04 1.957 0.364 Xe-131m 1.0282E+06 0.020 0.143 Xe-133m 1.8904E+05 0.042 0.190 Xe-133 4.5317E+05 0.045 0.135 Xe-135m 9.1740E+02 0.432 0.098 Xe-135 3.2724E+04 0.247 0.316 Xe-138 1.0199E+03 1.128 0.672 (DRN 04-704, R14)
WSES-FSAR-UNIT-3 TABLE 15B-2 Revision 14 (12/05)
ICRP30 ISOTOPE PROPERTIES(DRN 04-704, R14)
IsotopeEffective CloudshineSv-s/Bq-m 3Effective InhaledSv-s/Bq-m 3 IsotopeEffective CloudshineSv-s/Bq-m 3Effective InhaledSv-s/Bq-m 3Co-58 4.760E-14 2.940E-09 Te-131m 7.463E-14 1.758E-09 Co-60 1.260E-13 5.910E-08 Te-132 1.030E-14 2.550E-09 Kr-85 1.190E-16 0.000E-00 I-131 1.820E-14 8.890E-09 Kr-85m 7.480E-15 0.000E-00 I-132 1.120E-13 1.030E-10 Kr-87 4.120E-14 0.000E-00 I-133 2.940E-14 1.580E-09 Kr-88 1.020E-13 0.000E-00 I-134 1.300E-13 3.550E-11 Rb-86 4.810E-15 1.790E-09 I-135 8.294E-14 3.320E-10 Sr-89 7.730E-17 1.120E-08 Xe-133 1.560E-15 0.000E-00 Sr-90 7.530E-18 3.510E-07 Xe-135 1.190E-14 0.000E-00 Sr-91 4.924E-14 4.547E-10 Cs-134 7.570E-14 1.250E-08 Sr-92 6.790E-14 2.180E-10 Cs-136 1.060E-13 1.980E-09 Y-90 1.900E-16 2.280E-09 Cs-137 2.725E-14 8.630E-09 Y-91 2.600E-16 1.320E-08 Ba-139 2.170E-15 4.640E-11 Y-92 1.300E-14 2.110E-10 Ba-140 8.580E-15 1.010E-09 Y-93 4.800E-15 5.820E-10 La-140 1.170E-13 1.310E-09 Zr-95 3.600E-14 6.390E-09 La-141 2.390E-15 1.570E-10 Zr-97 4.432E-14 1.171E-09 La-142 1.440E-13 6.840E-11 Nb-95 3.740E-14 1.570E-09 Ce-141 3.430E-15 2.420E-09 Mo-99 7.280E-15 1.070E-09 Ce-143 1.290E-14 9.160E-10 Tc-99m 5.890E-15 8.800E-12 Ce-144 2.773E-15 1.010E-07 Ru-103 2.251E-14 2.421E-09 Pr-143 2.100E-17 2.190E-09 Ru-105 3.810E-14 1.230E-10 Nd-147 6.190E-15 1.850E-09 Ru-106 1.040E-14 1.290E-07 Np-239 7.690E-15 6.780E-10 Rh-105 3.720E-15 2.580E-10 Pu-238 4.880E-18 7.790E-05 Sb-127 3.330E-14 1.630E-14 Pu-239 4.240E-18 8.330E-05 Sb-129 7.140E-14 1.740E-10 Pu-240 4.750E-18 8.330E-05 Te-127 2.420E-16 8.600E-11 Pu-241 7.250E-20 1.340E-06 Te-127m 1.470E-16 5.810E-09 Am-241 8.180E-16 1.200E-04 Te-129 2.750E-15 2.090E-11 Cm-242 5.690E-18 4.670E-06 Te-129m 3.337E-15 6.484E-09 Cm-244 4.910E-18 6.700E-05 (DRN 04-704, R14)
WSES-FSAR-UNIT-3
TABLE 15B-3 Revision 14 (12/05) (DRN 04-704, R14)
EFFECTIVE DOSE EQUIVALENT WEIGHTING FACTORS (ICRP 30)
Organ/tissue W T Gonads 0.25 Breast 0.15 Red Marrow 0.12 Lungs 0.12 Thyroid 0.03 Bone Surface 0.03 Remainder
- 0.30
(DRN 05-1551, R14)
- A value of WT = 0.06 is applicable to each of the five remaining organs or tissues (such as liver, kidneys, spleen, brain, small intestine, upper large intestine, lower large intestine, etc., but excluding skin, lens of the eye, and the extremities) receiving the highest doses (DRN 04-704, R14; 05-1551, R14)
WSES-FSAR-UNIT-3 TABLE 15B-4 Revision 14 (12/05) (DRN 04-704, R14)
POWERS 10% NATURAL DEPOSITION MODELPhase Time Interval (s) Correlation (hr-1)gap0 - 1800 (0 - 0.5 hrs)
)(10 260.3 0182.0)10 (6 MWt P xgap 1800 - 6480 (0.5 - 1.8 hrs) 1000)(938.0 exp 1 0645.0)10 (MWt Pearly in-vessel 1800 - 6480 (0.5 - 1.8) 1000)(910.0 exp 1 0326.0)10 (MWt P gap + early in-vessel 6480 - 13680 (1.8 - 3.8 hr) 1000)(869.0 exp 1 094.0)10 (MWt P gap + early in-vessel 12680 - 49680 (3.8 - 13.8 hr)
)(10 15.10 0811.0)10 (6 MWt P xgap + early in-vessel 49680 - 80000 (13.8 - 22.22 hr) 1000)(384.2 exp 1 094.0)10 (MWt P (DRN 04-704, R14)