ML20247A537
| ML20247A537 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 05/01/1998 |
| From: | Dugger C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3F1-98-0076, W3F1-98-76, NUDOCS 9805060134 | |
| Download: ML20247A537 (33) | |
Text
- _ _ - _ _ - _ _ - - - - - _ - _ _ - _ _ _ - - _
y Enti gy Operations. Inc.
O/
Killona, LA 700( 6 0751 Tel 604 739 (460 Charles M. Dugger ce Pres der' opwates W3F198-0076 A4.05 PR May 1,1998 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Request for Additional Information (RAl) Regarding Technical SpeciGcation Change Request NPF-38-193 Gentlemen:
1 By letters dated March 27,1997 and supplemented by letters dated April 3,1997, July 21,1997, October 23,1997, December 12,1997, January 21,1998, January 29,1998 and March 23,1998, Waterford 3 proposed to amend Operating License NPF-38 to increase spent fuel storage capacity and increase the maximum fuel enrichment. The NRC staff requested additionalinformation, in a letter dated April 2, 1998, regarding the radiological assessment of the proposed changes. The requested information is included in the enclosures to this submittal.
The information in this submittal does not affect the previously provided determination of no significant hazards.
l
,.. <3 p 9805060134 980501 PDR ADOCK 05000382~
P PDR i
--.__ -.______ - _ _ _ a
9 Request for Additional Information (RAI)Regarding l
Technical Specification Change Request NPF-38-193 l
Page 2 May 1,1998 I
{
Should you have any questions or comments concerning this request, please contact Roy Prados at (504) 739-6632.
Very truly yours, I
i
/
l C.M. Dugger Vice President, Operations
. Waterford 3 CMD/RWP/rtk
Enclosures:
Affidavit Enclosures 1 & 2 Attachments (w/ Enclosures & Attachments) cc:
E.W. Merschoff, NRC Region IV C.P. Patel, NRC-NRR (w/o Enclosures & Attachments) cc:
J. Smith N.S. Reynolds NRC Resident inspectors Office Administrator Radiation Protection Division (State of Louisiana)
American Nuclear insurers l
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION in the matter of
)
)
Entergy Operations, incorporated
)
Docket No. 50-382 Waterford 3 Steam Electric Station
)
AFFIDAVIT Charles Marshall Dugger, being duly sworn, hereby deposes and says that he is Vice President Operations - Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Additional Information Regarding Technical Specification Change Request NPF-38-193; that he is familiar with the content ther90f; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
s
/
Charles Marshall Dugger Vice President Operations - Waterford 3 STATE OF LOUISIANA
)
) ss PARISH OF ST. CHARLES
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Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this / " day of
//z u
,1998.
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- 4) with the values in the table based on the assumptions given in Table 11.2-12, (Attachment 5). Neither these estimated releases, nor the actual releases, are affected by the number of stored fuel assemblies.
SUMMARY
OF FUEL HANDLING BUILDING SHIELDING CALCULATIONS increasing the number of fuel assemblies stored in the Spent Fuel Pool (SFP) does not have' a large affect on the dose rates in adjacent areas accessible to personnel. There is some increase in dose rate due to the more-tightly-packed arrangement of the fuel assemblies in the high-density racks, the closer Iccation of the racks to the pool walls, and higher burnup values assumed. However, the largest component of any dose rate increases results from the conservative peaking factor value (1.8) selected for use in the design analyses of the high-density racks. Many of the dose rate evaluations also conservatively consider some number of fuel assemblies to have a minimum cooling i
time of 3 days (the Waterford 3 Technical Specification, T.S. 3.9.3, prevents fuel discharge from the reactor for at least 3 days following reactor shutdown). This is very conservmive, since fuel transfer rates limit the number of fuel assemblies which could be discharged following the mandatory three day cooling time.
Dose rates in areas adjacent to the pools from freshly-discharged fuel are almost entirely the result of short-lived fission products that have reached saturation concentrations in the core. The gamma source term is proportional to the core specific power and the corresponding assumed peaking factor. Fuel burnup (70,000 Mwd /mtU assumed) has only a minor effect on dose rates to adjacent accessible areas. Long-lived fission products depend on fuel burnup rather than specific power and generally contribute littta to adjacent area dose rates. Short-lived fission products are the major contributors.. Mjacerbrea dose rates and decay very rapidly. For example, one-l year-cooled f0el will have a gamma source term less than 6% of the source term used in the " hot" assembly analyses; it is approximately 1% for five-year-cooled fuel and approximately 0.6% for ten-year-cooled fuel.
The dose rate from fuel in transit within the pool is a function of the gamma source term i
of the single assembly being moved. This contributor to dose rate is independent of the.
capacity of the pool. Increased dose rates are attributable to the necessity to move fuel closer to the pool walls to access the peripheral cells of the SFP racks and to place assemblies in the additional cells located in the Cask Storage Pit (the current SFP racks have set backs of from approximately twelve to thirty-eight inches from the SFP walls and there are currently no racks in the Cask Storage Pit).
Dose Rates From Stored Fuel All dose rates from stored fuel are based on assemblies with an initial enrichment of 5.5% and a fuel exposure of 70,000 Mwd /mtU. Cooling times and peaking factors vary as specified below.
Enclosuro 2 to W3F1-98-0076 Page 2 of 7 Calculation S1 The dose rate in accessible room areas below the SFP was determined considering a full SFP. The fuelin the SFP was considered to consist of 48 " hot" assemblies (3 day cooled,1.80 peaking factor) and the remainder of the SFP was considered to be filled with assemblies, which are 1 year cooled,1.00 peaking factor.
This calculation simulates the dose rates for the first assemblies unloaded from the reactor vessel during a refueling outage. This calculation is conservative because 48 assemblies could not be removed from the reactor vessel three days following reactor shutdown (Technical Specification, T.S. 3.9.3, prevents fuel discharge from the reactor until at least 3 days following reactor shutdown) and assuming 1801 assemblies (1849 - 48) with a cooling time of 1 year greatly increases the source term in the SFP.
R_esult S1 The dose rate at elevation -30.00' which is five feet above the floor in i
the room below the SFP is 4.17 mrem /hr. This is within the Zone 111 I
radiation level of less than 15 mrem /hr as shown in Attachments 1 and 2.
Calculation S2 The dose rate in accessible room areas below the SFP was determined, for a full core discharge (217 fuel assemblies), with a full SFP..The fuel in the pool is considered to consist of 116 assemblies (5 day cooled,1.80 peaking factors),101 assemblies (5 day cooled, average peaking factors of 1.40), and the remainder of the pool is considered to be filled with assemblies which are 1 year cooled,1.00 peaking factor. This calculation simulates the dose rates immediately following a full core discharge (unloading 4 assemblies per hour starting 3 days after reactor shutdown results in the full core being unloaded 126.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after reactor shutdown).
Result S2 The dose rate at elevation -30.00' which is five feet above the floor in the room below the SFP is 11.7 mrem /hr. This is within the Zone Ill radiation level of less than 15 mrem /hr as shown in Attachments 1 and 2.
Calculation S3 The dose rate outside the Fuel Handling Building east wall from fuel stored in the SFP racks was determined. The eastern-most row (outer row assemblies next to the wall) is considered to contain 1 year cooled assemblies. The remaining assemblies which form the source (interior rows) are considered to contain 3 day cooled,1.80 peaking factor assemblies. This calculation simulates the FHB external dose rate (administrative controls will ensure that fuel loaded into peripheral cell locations is cooled for a minimum of one year).
Enclosura 2 to W3F1-98-0076 Page 3 of 7
]
Result S3 The dose rate outside the Fuel Handling Building eart wall is 0.18 mrem /hr. This dose rate is less than the Radiation Area posting requirements of 2.5 mrem /hr.
Calculation S4 The dose rate outside the Fuel Handling Building east wall from fuel stored in the SFP racks was determined. The eastern-most row (outer row of storage cells next to the wall) is considered to be empty (containing only water as a shielding material). The remaining rows of assemblies which form the source (interior rows) are considered to j
contain 3 day cooled,1.80 peaking factor assemblies. Since Uranium is a better shielding material than water, this calculation was performed as a worst case permutation of calculation S3 to ensure that the external dose rates are acceptable when fuel is not present in the outer rows. Additionally, a calculation was performed for a dose point opposite from a location where a single water hole (empty ell) is present in the outer row and other locations in the outer row contain 1 year cooled assemblies. This additional calculation was performed as a check on another possible permutation of calculation S3.
Result S4 The dose rate with the eastern-most row assumed to be water is 2.34 mrem /hr. The dose rate with a single water hole and 1 year cooled assemblies in the outer row is 0.61 mrem /hr. These doses are less than the Radiation Area posting requirements.
Calculation S5 The dose rate in the pipe chase at the north SFP wall was determined.
The northern-most row (outer row of assemblies next to the wall) is considered to contain 1 year cooled fuel with peaking factors of 1.0.
The remaining assemblies which form the source (interior rows) are considered to be 3 day cooled,1.80 peaking factor assemblies. This calculation simulates the pipe chase dose rate (administrative controls will ensure that fuel loaded into peripheral cell locations is cooled for a minimum of one year).
Result S5 The dose rate in the pipe chase at the wall north of stored fuel is 0.17 mrem /hr. This is within the Zone 11 radiation level of less than 2.5 mrem /hr as shown in Attachments 1 and 2.
Calculation S6 The dose rate in the pipe chase at the wall north of fuel stored in the SFP racks was determined. The northern-most row (outer row of assemblies next to the wall) is considered to be empty (containing only water as a shielding material). The remaining rows of assemblies
Enclosuro 2 to W3F1-98-0076 Page 4 of 7 which form the source (interior rows) are considered to contain 3 day cooled,1.80 peaking factor assemblies. Since Uranium is a better shielding material than water, this calculation was performed as a worst case permutation of calculation S5 to ensure that the pipe chase dose rates are acceptable when fuel is not present in the outer rows.
l This calculation verifies that removal of the 1 year cooled assemblies will not exceed the radiation zone restrictions. Additionally, a L
calculation was performed for a dose point opposite from a location where a single water hole (empty cell) is present in the outer row and other locations in the outer row contain 1 year cooled assemblies.
l This additional calculation was performed as a check on another possible permutation of calculation S5.
Result S6 The dose rate with the northern-most row assumed to be water is 2.15 mrem /hr. The dose rate with a single water hole and 1 year assemblies in the outer row is 0.56 mrem /hr. This is within the Zone ll radiation level of less than 2.5 mrem /hr as shown in Attachments 1 and 2.
Calculation S7 The dose rate in the pipe chase at the wall north of fuel stored in the Refueling Canal racks was determined. All assemblies are considered to be 1 year cooled with 1.00 peaking factors.~ This calculation simulates the dose rate in the pipe chase north of the Refueling Canal.
Result S7 The dose rate in the north pipe chase from fuel stored in the Refueling Canal is 0.05 mrem /hr. This is within the Zone ll radiation level of less than 2.5 mrem /hr as shown in Attachments 1 and 2.
i Calculation S8 The dose rate in the fuel pool pump room from the Refueling Canal racks fully filled with fuel was determined. All assemblies are considered to be 1 year cooled with 1.00 peaking factors. This calculation simulates the dose rate in the fuel pool pump room which is to the West of the Refueling Canal at the +1.00' elevation.
Result S8 The dose rate in the fuel pool pump room from fuel stored in the Refueling Canal is 0.05 mrem /hr. This is within the Zone ll radiation level of less than 2.5 mrem /hr as shown in Attachments 1 and 2.
Dose Rates From Fuel in Transit The following dose rates are based on single fuel assemblies. Fuelin transit is suspended from the Spent Fuel Handling Machine with the top of the fuel assembly
_ _ -_____ ______ _ __- - __ - - - - - ---___- -___-- -___u
. to W3F1-98-0076 Page 5 of 7 approximately seven feet below the surface of the water. In these calculations, three types of fuel assemblies have been considered:
l 1.
" Hot" fuel (5.5% initial enrichment, 70,000 Mwd /mtU burnup, 3-day cooling,1.8 peaking factor).
2.
Bounding aged fuel from Batches A0,80, CO, C1, C2, D0, D1, D2 and D3 (Batch D0,3.79% initial enrichment,48,000 Mwd /mtU burnup,7-year cooling).
3.
Bounding Batch A0 fuel (1.87% initial enrichment,16,000 Mwd /mtU,11-year cooling). Batch A0 fuelis the oldest (longest cooling time in the SFP) fuel in the SFP.
Three types of fuel assemblies were used in the various analyses because during the reracking the Cask Storage Pit must be almost completely filled with spent fuel to allow for the removal of the existing racks in the SFP and the installation of the new racks.
The oldest (Batch A0) fuel in the SFP will be placed, in the Cask Storage Pit, closest to Gate #3A to minimize the dose rate in areas adjacent to the North side of the Cask Storage Pit. Seven year cooled and eleven year cooled fuel assemblies are used for the reracking " design basis" calculations. These are the oldest (longest cooling time since discharge from the reactor) assemblies in the SFP. The " hot" fuel assembly represents the normal Spent Fuel Pool operation " design basis" calculation. During normal SFP operation the seven rows r,f storage spaces, in the Cask Storage Pit, adjacent to Gate #3A will either be left empty or will be used to store fresh, unirradiated fuel. If, for any reason, the need arises to store irradiated fuel closer (than seven spaces) to Gate #3A during norma! SFP operation the appropriate shielding calculations will be performed, prior to pl acing fuel into these cells, to ensure that dose rates will remain acceptable. The dose rates determined below are only for fuel in transit (top of fuel assembly approximately seven feet below the surface of the water).
C_a. lculation T1 The dose rate in the Cask Decontamination Area from a single assembly moved (suspended approximately 7' underwater from the Spent Fuel Handling Machine) in the SFP and located against the west pool wall opposite the Cask Decontamination Area was determined.
This calculation was performed with bounding aged fuel (3.79% initial enrichment,48,000 Mwd /mtU burnup,7-year cooling). This calculation simulates the dose rate in the Cask Decontamination Area from an assembly during loading into the SFP storage spaces next to the wall that separates the SFP from the Cask Decontamination Area.
Result T1 The dose rate in the Cask Decontamination Area is 18.0 mrem /hr.
This is within the Zone IV radiation level of less than 100 mrem /hr as shown in Attachments 1 and 2.
. to W3F1-98-0076 Page 6 of 7 Calculation T2 The dose rate in the Cask Decontamination Area from a single assembly moved (suspended approximately 7' underwater from the Spent Fuel Handling Machine) in the Cask Storage Pit was determined. This calculation (three cases, each representing a type of -
fuel as described above) determines the minimum separation distance between the three " design basis" fuel assemblies and Gate #3A.
Result T2 The dose rate limit based on the current zone designation is less than I
100.0 mrem /hr. The separation distance of the fuel assembly from the gate leading to the Cask Decontamination Area (Gate #3A) has been calculated such that the dose rate in the area (Cask i
Decontamination Area) does not exceed the 100.0 mrem /hr limit. The dose rate from " hot" fuel is 52.2 mrem /hr. The " hot" fuel is no closer i
than above the eighth row of cells from the gate (Gate #3A). The dose i
rate from bounding aged fuelis 3.97 mrem /hr. The bounding aged fuelis no closer than above the fourth row of cells from the gate. The dose rate from the bounding Batch A0 fuelis 22.6 mrem /hr. The bounding Batch A0 fuel is no closer than above the second row of cells from the gate. These results are within the Zone IV radiation level of 1
less than 100 mrem /hr as shown in Attachments 1 and 2.
l Calculation T3 The dose rate in the Train Bay Area from a single assembly moved (suspended approximately 7' underwater from the Spent Fuel Handling Machine) in the Cask Storage Pit was determined. The dose rate point in this calculation is farther away from the source than the point in Calculation T2 and receives additional shielding from Gate #4 (the gate north of the Cask Decontamination Area; see Attachments 1 and 2 for the physical arrangement). The separation distance of the fuel assembly from Gate #3A (the gate leading to the Cask Decontamination Area) will be the same as determined in Calculation T2. This calculation simulates the dose rate in the Train Bay from an assembly moved in the Cask Storage Pit.
Result T3 The dose rate from " hot" fuel is 0.75 mrem /hr. The dose rate from bounding aged fuelis 0.02 mrem /hr. The dose rate from bounding Batch A0 fuelis 0.05 mrem /hr. The bounding aged fuel and Batch A0 fuel results are within the Zone I radiation level of less than.25 mrem /hr as shown in Attachments 1 and 2. The " hot" fuel result is slightly above the Zone I level.
. to W3F1-98-0076 Page 7 of 7 Calculation T4 The dose rate at a point 3'-0" above the operating floor level of elevation +46.00', just north of the Cask Decontamination Area, from a single assembly moved (suspended approximately 7' underwater from the Spent Fuel Handling Machine) in the Cask Storage Pit was determined. The dose rate point in this calculation is much farther from the source than the dose rate point in Calculation T2 (see Attachments 1 and 2 for the physical arrangement). The separation distance of the fuel assembly from the gate (Gate #3A) will be the same as determined in Calculation T2. This calculation simulates the dose to a person standing at the North side of the Cask Decontamination Area at the +46.00' elevation.
Result T4 The dose rate from " hot" fuel is 0.30 mrem /hr. The dose rate from bounding aged fuelis 0.01 mrem /hr. The dose rate from bounding Batch A0 fuelis 0.04 mrem /hr. This is within the Zone ll radiation level of less than 2.5 mrem /hr as shown in Attachments 1 and 2.
Calculation T5 The dose rate at the top of the air gap between Gates #3A and #3B (Gate #3A forms the north boundary of the Cask Storage Pit, and Gate
- 3B is located to the north of Gate #3A) from a single assembly moved (suspended approximately 7' underwater from the Spent Fuel Handling Machine)in the Cask Storage Pit was determined. The separation distance of the fuel assembly from Gate #3A will be such that the dose rate at the top of thc air gap is less than 100.0 mrem /hr. The purpose of this calculation is to simulate the dose rate to the operator of the spent fuel handling machine.
Result T5 The dose rate from " hot" fuel is 99.4 mrem /hr. For this calculation, the " hot" fuel was allowed to approach as closely as a location above the sixth row of cells from Gate #3A; in reality the " hot" assembly will not be allowed closer than over the eighth row of cells from Gate #3A.
The dose rate from bounding aged fuelis 1.58 mrem /hr. The bounding aged fuel is no closer than above the fourth row of cells from the gate. The dose rate from bounding Batch A0 fuelis 21.0 mrem /hr. The bounding Batch A0 fuelis no closer than above the second row of cells from the gate. The bounding aged fuel result is within the current Zone 11 radiation level of less than 2.5 mrem /hr as shown in Attachments 1 and 2. The " hot" fuel and Batch A0 fuel results are above the current Zone 11 level (see item 2 response in ).
The calculation techniques used to develop the results discussed above were very conservative. Actual dose rates will be significantly lower than those shown.
I i
i ATTACHMENT 1 TO W3F1-98-0076
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5 ATTACHMENT 3 TO i W3F1-98-0076 J l
ATTACHMENT 3 WSES-FSAR-UNIT-3 I TABI.E 11.1-17 FISSION AND CORROSION PRODUCT ACTIVITIES IN THE SPENT FUEL POOL (pCi/ gram) Nuclide Expected Maximum H-3 1.0 (0) 9.5 (0) Br-84 0. O. Kr-85m 3.7 (-7)a 4.0 (-6) i Kr-85 3.5 (-4) 1.2 (-2) { Kr-87 1.5 (-15) 3.1 (-14) { Kr-88 5.1 (-9) 1.0 (-7) Rb-88 0. O. Sr-89 6.7 (-6) 1.2 (-4) i Sr-90 2.0 (-7) 6.7 (-6) j Y-90 1.8 (-7) 9.9 (-6) J Y-91 3.9 (-4) 5.5 (-4) Y-91m O. O. Sr-91 3.3 (-7) 2.5 (-6) Mo-99 5.0 (-3) 1.6 (-2) Ru-103 8.6 (-7) 9.6 (-5) Ru-106 2.0 (-7) 5.9 (-6) Te-129 0. O. I-131 4.5 (-3) 4.6 (-2) I-132 9.4 (-10) 1.2 (-8) l Te-132 3.2 (-4) 5.0 (-3) I-133 1.3 (-3) 1.4 (-2) I-134 0. O. Cs-134' 7.4 (-4) 2.8 (-3) l I-135 2.1 (-5) 2.4 (-4) Cs-136 3.4 (-4) 4.4 (-4) Cs-137 5.3 (-4) 1.1 (-2) Xe-131m 2.3 (-4) 4.9 (-3) 1 Xe-133 3.2 (-2) 6.0 (-1) Xe-135m O. O. Xe-135 2.5 (-5) 6.7 (-4) Xe-138 0. O. Ba-140 3.8 (-6) 1.3 (-4) ) La-140 1.2 (-6) 6.4 (-5) l Pr-143 8.8 (-7) 1.2 (-4) l Ce-144 6.5 (-7) 9.0 (-5) l Z -95 1.2 (-6) 1.3 (-4) l Cr-51 3.7 (-5) 3.7 (-5) Mn-54 6.3 (-6) 6.3 (-6) } Co-58 3.2 (-4) 3.2 (-4) Fe-59 2.0 (-5) 2.0 (-5) { Co-60 4.1 (-5) 4.1 (-5) i a ( ) denotes power of 10 l
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ATTACHMENT 5 TO W3F1-98-0076 1
r.... - ATTACH' MENT 5 l WSES-FSAR-UNIT-3 TABLE 11.2-12 (Sheet.1 of 3) 1 PRINICIPAL PARAMETERS AND CONDITIONS USED IN RADIOLOGICAL EFFLUENT EVAllMTIONS l 1. Design Thermal Power Level 3560leit 2. Plant Load Factor 0.8 3. Percent Failed Fuel 0.12 3 4. Primary System j Mass of Coolant 455.%0 lbs. Average Letdown Rate 40 gpm Average Letdown Rate through 203 Demineralizers 8 gpm l Shim Bleed Rate 0.54 gpm Steam Generator Leak Rate to Secondary Side 100 lbs./ day Leakage Rate to Auxiliary Bldg. 160 lbs./ day Source Term to Containment Bldg. Noble Gases it of noble gas coolant inventory l lodine 0.001% of iodine coolant inventory Letdown Stripping (no stripping) 5. Secondary System Steam Flow Rate 1.51 x 107 lbs./hr Hass of Steam in each Steam Generator 1.31 x 104 lbs. Mass of Water in each Steam Generator 1.64 x 105 lbs. Number of Steam Generators 2 Hass of Secondary Coolant (includes Condenser hotwell) 2.8 x 106 lbs. Rate of Steam Leakage to Turbine Bldg. 1700 lbs./hr Steam Generator Blowdown Rate 60 gpm 6. Containment Volume 2.7 x 106 ft.3 7. Frequency of Containment Purge 24 purges /yr. f with 16 hrs. of l recirculation through filters prior to each l purge 1 8. Gaseous Waste Management System j Number of Tanks 3 Decay Tank Volume 600 ft.3 (each) Fill Time 60 days Holdup Time 60 days
WSES FSAR-UNIT-3 Table 11.2 12 (Sheet 2 of 3) Revision 7 (10/94) PRINICIPAL PARAMETERS AND CONDITIONS USED IN RADIOLOGICAL EFFLUENT EVALUATIONS l 9. Iodine Partition Factors { Auxiliary Building Leakage 0.0075 Steam Leakage to Turbine Bldg. 1.0 Steam Generator (carryover) 0.01 Main Condenser 0.15 10. Steam Generator Blowdown Flash Tank Routed to Condenser i 11. Decontamination Factors (2) Demineralized Anion CsRb Other Nuclides i Mixed Bed: Primary Coolant Letdown 10 2 10 (L13 80 ) 3 Ra& aste (H+ OH-) 102 (10)(1) 2(10) 102 (10) Evaporator Condensate 10 10 10 Polishing Cation Bed (any system) 1(1) 10(10) 10(10) Anion Bed (any system) 102 (10) 1(1) 1(1) Powdex (any system) 10(10) 2(10) 10(10) All Nuclides Evaporators (3) Except Iodine Iodine l Miscellaneous Radwaste 104 103 Boric Acid Recovery 103 102 Detergent Wastes 102 102 (1) For demineralizers in series, the DF for the second demineralized is given in parenthesis. l (2) l These DFs differ from DFs presented in Tables 11.2 6, 11.2 7 and 11.2 8 in that they take less credit for the effectiveness for filters and evaporators in orde' '.o place a conservative upper limit on liquid effluent from the plant. (3) The Waste Concentrator was abandoned per DC 3188. l t _._____ _______ _ - - _ - _ _ - - - - - - - -
lc. WSES-FSAR-UNIT-3 Table 11.2-12 (Sheet 3 of 3) PRINICIPAL PARAMETERS AND CONDITIONS USED IN RADIOLOGICAL EFFLUENT EVALLMTIONS J 12. Liquid Waste Streams Collection Decay ' Stream Flow Rate Fraction Fraction Time-Time Gal / Days of PCA Discharged (Days) (Days) Shim Bleed 7.78(+02) 1.000 0.120 26.000 1.300 Equipment Drains 5.00(+01) 0.100 0.120 90.000 1.300 Clean Wastes 6.22(+02) 1.000 0.120 32.400 1.300 i Dirty Wastes 1.38(+03) 0.075 1.000 2.300 0.100 Blowdown 8.69(+04) 0.000 2.300 0.100 i Untreated Blowdown 0. 1.000 0.0 0.0 Stream Decontamination Factors 1 Iodine Cesitan Others Shin Bleed 1.00(+05) 2.00(+04) 1.00(+05) Equipment Drains 1.00(+04) 2.00(+04) 1.00(+05) Clean Wastes' 1.00(+04) 2.00(+04) 1.00(+05) Dirty Wastes 1.00(+04) 2.00(+04) 1 00(+05) B1owdown 1.00(+01) 2.00(+00) 1.00(+01) Untreated Blowdown 1.00( 00) 1.00( 00) 1.00( 00) i 13. Filtration of Airborne Effluents Decontamination Factors lodines Particulate Auxiliary Bldg. Vent 10 100 Reactor Bldg. Vent 10 100 Turbine Bldg. Unfiltered (open Turbine Building) Air Ejector 10 100 Decay Tanks Unfiltered 14. X/Q's and D/Q's based on a ground release model only (i.e., no mixed mode release). 15. A filtration efficiency of 90% for the RAB filter during all normal operation, including Containment Building purge. 16. Iodine charcoal filter efficiency of 50% for the ARRS inside the Containment. } () Denotes power of 10 l l ~}}