ML24114A126
| ML24114A126 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/15/2024 |
| From: | Wolf Creek |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 000347 | |
| Download: ML24114A126 (1) | |
Text
WOLF CREEK
TABLE OF CONTENTS
CHAPTER 11.0
RADIOACTIVE WASTE MANAGEMENT
Section Page
11.1 SOURCE TERMS 11.1-1
11.1.1 RADIOACTIVE CONCENTRATIONS AND RELEASES 11.1-1 11.1.2 SHIELDING 11.1-1 11.1.3 ACCIDENT ANALYSIS SOURCE TERMS 11.1-1
App. 11.lA PARAMETERS FOR CALCULATION OF SOURCE 11.lA-1 TERMS FOR EXPECTED RADIOACTIVE CONCEN-TRATIONS AND RELEASES
11.2 LIQUID WASTE MANAGEMENT SYSTEMS 11.2-1
11.2.1 DESIGN BASES 11.2-1
11.2.1.1 Safety Design Basis 11.2-1 11.2.1.2 Power Generation Design Bases 11.2-1
11.2.2 SYSTEM DESCRIPTION 11.2-1
11.2.2.1 General Description 11.2-1 11.2.2.2 Component Description 11.2-6 11.2.2.3 System Operation 11.2-9
11.2.3 RADIOACTIVE RELEASES 11.2-13
11.2.3.1 Sources 11.2-13 11.2.3.2 Release Points 11.2-13 11.2.3.3 Dilution Factors 11.2-14 11.2.3.4 Estimated Doses 11.2-14
11.2.4 CALCULATED BASIS FOR LIQUID SOURCE TERMS 11.2-14
11.2.5 SAFETY EVALUATION 11.2-15 11.2.6 TESTS AND INSPECTION 11.2-15 11.2.7 INSTRUMENTATION DESIGN 11.2-15 11.
2.8 REFERENCES
11.2-15
11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 11.3-1
11.3.1 DESIGN BASES 11.3-1
11.0-i Rev. 29 WOLF CREEK
TABLE OF CONTENTS (Continued)
Section Page
11.3.1.1 Safety Design Basis 11.3-1 11.3.1.2 Power Generation Design Bases 11.3-1
11.3.2 SYSTEM DESCRIPTIONS 11.3-2
11.3.2.1 General Description 11.3-2 11.3.2.2 Component Description 11.3-4 11.3.2.3 System Operation 11.3-6
11.3.3 RADIOACTIVE RELEASES 11.3-7
11.3.3.1 Sources 11.3-7 11.3.3.2 Release Points 11.3-8 11.3.3.3 Dilution Factors 11.3-8 11.3.3.4 Estimated Doses 11.3-8
11.3.4 SAFETY EVALUATION 11.3-9 11.3.5 TESTS AND INSPECTIONS 11.3-9 11.3.6 INSTRUMENTATION APPLICATION 11.3-9 11.
3.7 REFERENCES
11.3-12
11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4-1
11.4.1 DESIGN BASES 11.4-1
11.4.1.1 Safety Design Bases 11.4-1 11.4.1.2 Power Design Bases 11.4-1
11.4.2 SYSTEM DESCRIPTION 11.4-3
11.4.2.1 General Description 11.4-3 11.4.2.2 Component Description 11.4-4 11.4.2.3 System Operation 11.4-5 11.4.2.4 Packaging, Storage, and Shipment 11.4-9
11.4.3 SAFETY EVALUATION 11.4-10 11.4.4 TESTS AND INSPECTIONS 11.4-11 11.4.5 INSTRUMENTATION APPLICATION 11.4-11
Appendix 11.4A Interim Onsite Storage 11.4A-1
11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5-1
11.5.1 DESIGN BASES 11.5-1
11.0-ii Rev. 29 WOLF CREEK
TABLE OF CONTENTS (Continued)
Section Page
11.5.1.1 Safety Design Bases 11.5-1 11.5.1.2 Power Generation Design Bases 11.5-2 11.5.1.3 Codes and Standards 11.5-3
11.5.2 SYSTEM DESCRIPTION 11.5-3
11.5.2.1 General Description 11.5-3 11.5.2.2 Liquid Monitoring Systems 11.5-6 11.5.2.3 Airborne Monitoring Systems 11.5-12 11.5.2.4 Safety Evaluation 11.5-18
11.5.3 EFFLUENT MONITORING AND SAMPLING 11.5-19 11.5.4 PROCESS MONITORING AND SAMPLING 11.5-19
11.0-iii Rev. 29 WOLF CREEK
TABLE OF CONTENTS (Continued)
LIST OF TABLES
Number Title
11.1-1 Reactor Coolant and Secondary Coolant Specific Activities 0.12-Percent Fuel Defects
11.1-2 Annual Effluent Releases - Liquid
11.1-3 Comparison of the Design to Regulatory Positions Of Regulatory Guide 1.112, Revision 0, Dated April, 1976, Titled "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors"
11.1-4 Reactor Coolant and Secondary Coolant Shielding Source Terms - 0.25 Percent Fuel Defects
11.1-5 Primary Coolant Activity Concentrations
11.1-6 Contained Sources of the Radioactive Waste Management Systems and Large Potentially Radioactive Outside Storage Tanks
11.lA-1 Plant Data for Source Term Calculations
11.lA-2 Parameters Used in the Calculation of Estimated Activity in Liquid Wastes
11.lA-3 Description of Major Sources of Gaseous Releases
11.lA-4 Characteristics of Release Points and Releases
11.2-1 Liquid Waste Processing System Equipment Principal Design Parameters
11.2-2 Tank Uncontrolled Release Protection Provisions
11.0-iv Rev. 34 WOLF CREEK
TABLE OF CONTENTS (Continued)
Number Title
11.2-3 Deleted
11.2-4 Deleted
11.2-5 Deleted
11.2-6 Deleted
11.2-7 Deleted
11.2-8 Deleted
11.2-9 Deleted
11.2-10 Deleted
11.2-11 Deleted
11.2-12 Liquid Waste Management System Instrumentation Principal Design Parameters
11.3-1 Gaseous Waste Processing System Major Component Description
11.3-2 Deleted
11.3-3 Deleted
11.3-4 Deleted
11.0-v Rev. 14 WOLF CREEK
TABLE OF CONTENTS (Continued)
Number Title
11.3-5 Gaseous Waste Processing System Instrumentation Design Parameters
11.4-1 Design Comparison to Branch Technical Position ETSB 11-3 Revision 2, "Design Guidance for Solid Radioactive Waste Management System Installed in Light-Water-Cooled Nuclear Power Reactor Plants
11.4-2 Estimated Expected and Maximum Annual Activities of the Influents to the Solid Radwaste Solidification System, Curies (Historical)
11.4-3 Estimated Maximum Annual Quantities of Solid Radwaste (Historical)
11.4-4 Estimated Expected and Maximum Annual Activities of Solid Radwaste Shipped, Curies (Historical)
11.4-5 Solid Radwaste System - Component Description
11.4A Interim On-Site Storage Facility
11.5-1 Liquid Process Radioactivity Monitors
11.5-2 Liquid Effluent Radioactivity Monitors
11.5-3 Airborne Process Radioactivity Monitors
11.5-4 Airborne Effluent Radioactivity Monitors
11.5-5 Power Supplies for Process and Effluent Monitors
11.0-vi Rev. 32 WOLF CREEK
CHAPTER 11 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet Title Drawing #*
11.1A-1 0 Liquid Waste Treatment Systems Block Diagram 11.1A-2 1 System Decontamination Factors 11.1A-2 2 System Decontamination Factors 11.1A-2 3 System Decontamination Factors 11.1A-2 3A System Decontamination Factors 11.1A-2 4 System Decontamination Factors 11.1A-2 4A Deleted 11.1A-2 5 System Decontamination Factors 11.1A-2 6 System Decontamination Factors 11.1A-2 7 System Decontamination Factors 11.1A-3 0 Potential Gaseous Release 11.2-1 1 Liquid Radwaste System M-12HB01 11.2-1 2 Liquid Radwaste System M-12HB02 11.2-1 3 Liquid Radwaste System M-12HB03 11.2-1 4 Liquid Radwaste System M-12HB04 11.2-1 5 Radioactive Liquid Release Flow Diagram 11.3-1 1 Gaseous Radwaste System M-12HA01 11.3-1 2 Gaseous Radwaste System M-12HA02 11.3-1 3 Gaseous Radwaste System M-12HA03 11.3-2 0 Deleted 11.3-3 0 Compressor Package Instruments 11.3-4 0 Hydrogen Recombiner Instruments 11.4-1 1 Solid Radwaste System M-12HC01 11.4-1 2 Solid Radwaste System M-12HC02 11.4-1 3 Solid Radwaste System M-12HC03 11.4-1 4 Solid Radwaste System M-12HC04 11.4-2 0 Deleted
11.0-vii Rev.31 WOLF CREEK
CHAPTER 11.0
RADIOACTIVE WASTE MANAGEMENT
11.1 SOURCE TERMS
This section presents the design bases for determining the source terms for radioactive releases from the plant, for shielding within the plant, and for accident analysis performed in Chapter 15.0. The source terms used for releases, shielding, and accident analyses are based on 0.12, 0.25, and 1.0 percent fuel defects, respectively.
Actual release data is contained in Annual Radioactive Effluent Release Reports filed with the NRC in accordance with Offsite Dose Calculation Manual (ODCM) requirements.
11.1.1 RADIOACTIVE CONCENTRATIONS AND RELEASES
Reactor coolant and secondary coolant specific activities for an assumed 0.12-percent fuel defects and an assumed 100 pounds per day primary-to-secondary leakage are listed in Table 11.1-1. The basis for calculating these sources is Regulatory Guide 1.112. Compliance with Regulatory Guide 1.112 is discussed in Table 11.1-3. Appendix 11.1A provides a description of the input used.
The decontamination factors applied are based on Regulatory Guide 1.112. A description of liquid leakage rates, process paths, and associated component activity levels is contained in Section 11.2 and Appendix 11.1A. A description of gaseous leakage rates, process paths, and associated activity levels is contained in Appendix 11.1A and Sections 11.3 and 9.4. In-plant airborne activity concentrations and other data regarding the ventilation systems are provided in Sections 12.3 and 12.4.
11.1.2 SHIELDING
Reactor coolant and secondary coolant source terms used for shielding are based on 0.25-percent fuel defects. The source terms and the parameters used to calculate the source terms are given in Table 11.1-4 and Appendix 11.1A, respectively. Table 11.1-6 provides the isotopic composition of the contained sources for radioactive waste management systems and for large, potentially radioactive outside storage tanks.
11.1.3 ACCIDENT ANALYSIS SOURCE TERMS
Chapter 15.0 provides a complete discussion and a listing of the source terms for each accident analyzed.
11.1-1 Rev. 34 WOLF CREEK
TABLE 11.1-1 Specific Activities - 0.125% Fuel Defects(1)Reactor Coolant and Secondary Coolant
Class 1 Reactor Coolant Secondary Coolant
/g109Ci/gm /g109Ci/gm Kr-83m 6.93E-02 2.40E-06 Kr-85m 2.83E-01 8.88E-06 Kr-85 1.18E+00 3.70E-05 Kr-87 1.84E-01 5.77E-06 Kr-88 5.33E-01 1.67E-05 Kr-89 1.51E-02 4.64E-07 Xe-131m 4.26E-01 1.34E-05 Xe-133 3.63E+01 1.14E-03 Xe-133m 6.71E-01 2.17E-05 Xe-135m 7.55E-02 1.08E-05 Xe-135 1.23E+00 4.00E-05 Xe-137 2.80E-02 8.62E-07 Xe-138 1.02E-01 3.19E-06 Total noble gas 4.11E+01 1.30E-03 Class 2 Br-83 1.36E-02 2.48E-05 Br-84 7.28E-03 5.48E-06 Br-85 8.57E-04 7.50E-08 I-130 4.47E-03 1.20E-05 I-131 3.50E-01 1.06E-03 I-132 3.93E-01 7.47E-04 I-133 6.16E-01 1.74E-03 I-134 9.40E-02 1.03E-04 I-135 3.60E-01 8.81E-04 Total halogens 1.84E+00 4.75E-03
Rev. 16 WOLF CREEK
TABLE 11.1-1 (Sheet 2)
Specific Activities - 0.125% Fuel Defects(1)Reactor Coolant and Secondary Coolant
Class 3 Reactor Coolant Secondary Coolant
µCi/gm µCi/gm Rb-86 3.56E-03 1.96E-05 Rb-88 6.70E-01 3.40E-04 Rb-89 3.07E-02 1.35E-05 Cs-134 2.93E-01 1.62E-03 Cs-136 3.52E-01 1.93E-03 Cs-137 2.42E-01 1.34E-03 Cs-138 1.57E-01 1.34E-04 Total Cs, Rb 1.75E+00 5.40E-03 Class 4 N-16 1.31E+02 3.12E-10 Water activation product Class 5 H-3 3.50E+00 2.19E+00 Tritium Class 6 Cr-51 1.90E-03 5.83E-06 Mn-54 3.10E-04 9.53E-07 Fe-55 1.60E-03 4.92E-06 Fe-59 1.00E-03 3.07E-06 Co-58 1.60E-02 4.92E-05 Co-60 2.00E-03 6.15E-06 Sr-89 6.39E-04 3.55E-06 Sr-90 2.38E-05 1.31E-07 Sr-91 8.42E-04 3.55E-06 Y-90 1.85E-04 4.92E-07 Sr-92 6.48E-06 2.27E-08 Y-91m 4.94E-04 1.87E-06 Y-91 7.14E-05 2.23E-07
Rev. 13 WOLF CREEK
TABLE 11.1-1 (Sheet 3)
Specific Activities - 0.125% Fuel Defects(1)Reactor Coolant and Secondary Coolant
Class 6 Reactor Coolant Secondary Coolant
µCi/gm µCi/gm Y-93 5.46E-05 1.45E-07 Zr-95 8.15E-05 2.50E-07 Nb-95 8.17E-05 2.51E-07 Mo-99 1.02E-01 3.07E-04 Tc-99m 9.43E-02 2.85E-04 Ru-103 6.69E-05 2.05E-07 Ru-106 2.06E-05 6.34E-08 Rh-103m 6.64E-05 2.05E-07 Rh-106 2.06E-05 3.21E-10 Ag-110m 1.64E-04 5.03E-07 Te-125m 7.40E-05 2.27E-07 Te-127m 3.69E-04 1.13E-06 Te-127 1.63E-03 4.42E-06 Te-129m 1.29E-03 3.95E-06 Te-129 1.70E-03 3.64E-06 Te-131m 3.19E-03 9.30E-06 Te-131 1.79E-03 2.45E-06 Te-132 3.74E-02 1.13E-04 Te-134 4.62E-03 4.27E-06 Ba-137m 2.29E-01 1.25E-03 Ba-140 5.17E-04 1.58E-06 La-140 1.69E-04 5.60E-07 Ce-141 7.92E-05 2.43E-07 Ce-143 6.91E-05 2.03E-07 Ce-144 5.87E-05 1.80E-07 Pr-143 7.67E-05 2.36E-07 Pr-144 5.87E-05 1.80E-07 Total other isotopes 5.05E-01 2.07E-03 Note (3)
(1) Refer to Table 11.1A-1 for assumptions.
(2) For the secondary side, the noble gas activities are for the steamphase; all other activities are for the steam generator water (3) Lower blowdown rates result in higher secondary system activities.activities.
A 60-gpm blowdown will result in a total of 5.85E-2 µCi/gm (excluding noble gases, N-16, and tritium) in the steam generator.
A maximum blowdown rate was used in this table.
Rev. 13
WOLF CREEK
TABLE 11.1-3
COMPARISON OF THE DESIGN TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.112, REVISION 0, DATED APRIL, 1976, TITLED "CALCULATION OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS AND LIQUID EFFLUENTS FROM LIGHT-WATER-COOLED POWER REACTORS"
Regulatory Guide 1.112 Position WCGS
- 1. Each application for a per- 1. Inplant control meas-mit to construct a nuclear power ures to maintain radioactive reactor should include in-plant releases as low as is rea-control measures to maintain sonably achievable have been releases of radioactive materials incorporated in the design.
in liquid and gaseous effluents to the environment as low as is reasonably achievable in accor-dance with the requirements of Paragraph 20.1(c) of 10 CFR Part 20 and of Paragraph 50.34a, Para-graph 50.36a, and Appendix I of 10 CFR Part 50. For gaseous effluents, such measures could include storage for decay of noble gases removed from the pri-mary coolant and charcoal adsor-bers or HEPA filters to remove radioiodine and radioactive par-ticulates released from building ventilation exhaust systems. For liquid effluents, such measures could include storage for decay, demineralization, reverse osmosis, and evaporation.
- 2. The method of calculation 2. Parameters of NUREG-described in NUREG-0016 and NUREG- 0017 are used as discussed 0017 and the parameters presented in Appendix 11.lA. The in Chapter 2 of each report should method of calculation des-be used to calculate the quanti- cribed in NUREG-0017 has ties of radioactive materials in been generally followed.
gaseous and liquid effluents from light-water-cooled nuclear power reactors.
- 3. If methods and parameters 3. Justification for used in calculating source terms use of assumptions other are different from those given than those used in NUREG-in NUREG-0016 and NUREG-0017, 0017 are provided in they should be described in detail Appendix 11.lA.
and in the Environmental Report the basis for the methods and para-meters used should be provided.
Rev. 0 WOLF CREEK
TABLE 11.1-4
REACTOR COOLANT AND SECONDARY COOLANT SHIELDING SOURCE TERMS - 0.25 PERCENT FUEL DEFECTS(1)
Class 1 Reactor Coolant Secondary Coolant
µCi/gm µCi/gm Kr-83m 1.39E-01 4.80E-06 Kr-85m 5.66E-01 1.78E-05 Kr-85 2.35E+00 7.40E-05 Kr-87 3.68E-01 1.15E-05 Kr-88 1.07E+00 3.35E-05 Kr-89 3.03E-02 9.28E-07 Xe-131m 8.53E-01 2.68E-05 Xe-133 7.26E+01 2.28E-03 Xe-133m 1.34E+00 4.33E-05 Xe-135m 1.51E-01 2.16E-05 Xe-135 2.45E+00 7.99E-05 Xe-137 5.59E-02 1.72E-06 Xe-138 2.04E-01 6.37E-06 Total noble gas 8.21E+01 2.60E-03 Class 2 Br-83 2.73E-02 4.96E-05 Br-84 1.46E-02 1.10E-05 Br-85 1.71E-03 1.50E-07 I-130 8.93E-03 2.41E-05 I-131 6.99E-01 2.11E-03 I-132 7.85E-01 1.49E-03 I-133 1.23E+00 3.48E-03 I-134 1.88E-01 2.07E-04 I-135 7.19E-01 1.76E-03 Total halogens 3.68E+00 9.14E-03
Rev. 13 WOLF CREEK
TABLE 11.1-4 (Sheet 2)
Reactor Coolant and Secondary Coolant Specific Activities - 0.25% Fuel Defects(1)
Class 3 Reactor Coolant Secondary Coolant
µCi/gm µCi/gm Rb-86 7.13E-03 3.91E-05 Rb-88 1.34E+00 6.80E-04 Rb-89 6.15E-02 2.71E-05 Cs-134 5.87E-01 3.25E-03 Cs-136 7.05E-01 3.86E-03 Cs-137 4.85E-01 2.68E-03 Cs-138 3.14E-01 2.68E-04 Total Cs, Rb 3.50E+00 1.08E-02 Class 4 N-16 1.31E+02 3.12E-10 Water activation product Class 5 H-3 3.50E+00 2.19E+00 Tritium Class 6 Cr-51 1.90E-03 5.83E-06 Mn-54 3.10E-04 9.53E-07 Fe-55 1.60E-03 4.92E-06 Fe-59 1.00E-03 3.07E-06 Co-58 1.60E-02 4.92E-05 Co-60 2.00E-03 6.15E-06 Sr-89 1.28E-03 7.10E-06 Sr-90 4.76E-05 2.63E-07 Sr-91 1.68E-03 7.10E-06 Y-90 3.70E-04 9.83E-07 Sr-92 1.30E-05 4.54E-08 Y-91m 9.88E-04 3.74E-06 Y-91 1.43E-04 4.47E-07
Rev. 13 WOLF CREEK
TABLE 11.1-4 (Sheet 3)
Reactor Coolant and Secondary Coolant Specific Activities - 0.25% Fuel Defects(1)
Class 6 Reactor Coolant Secondary Coolant
µCi/gm µCi/gm Y-93 1.09E-04 2.89E-07 Zr-95 1.63E-04 5.01E-07 Nb-95 1.63E-04 5.02E-07 Mo-99 2.05E-01 6.15E-04 Tc-99m 1.89E-01 5.69E-04 Ru-103 1.34E-04 4.11E-07 Ru-106 4.13E-05 1.27E-07 Rh-103m 1.33E-04 4.10E-07 Rh-106 4.13E-05 6.43E-10 Ag-110m 3.28E-04 1.01E-06 Te-125m 1.48E-04 4.55E-07 Te-127m 7.38E-04 2.27E-06 Te-127 3.25E-03 8.85E-06 Te-129m 2.58E-03 7.90E-06 Te-129 3.40E-03 7.28E-06 Te-131m 6.37E-03 1.86E-05 Te-131 3.58E-03 4.91E-06 Te-132 7.48E-02 2.25E-04 Te-134 9.24E-03 8.55E-06 Ba-137m 4.58E-01 2.51E-03 Ba-140 1.03E-03 3.17E-06 La-140 3.38E-04 1.12E-06 Ce-141 1.58E-04 4.86E-07 Ce-143 1.38E-04 4.05E-07 Ce-144 1.17E-04 3.61E-07 Pr-143 1.53E-04 4.71E-07 Pr-144 1.17E-04 3.61E-07 Total other isotopes 9.86E-01 4.07E-03 Note (3)
(1) Refer to Table 11.1A-1 for assumptions.
(2) For the secondary side, the noble gas activities are for the steam phase; all other activities are for the steam generator water activities.
(3) Lower blowdown rates result in higher secondary system activities. A 60-gpm blowdown will result in a total of 1.17E-1 µCi/gm (excluding noble gases, N-16, and tritium) in the steam generator. A maximum blowdown rate was used in this table.
Rev. 13 WOLF CREEK
TABLE 11.1-5
Primary Coolant Activity Concentrations(1)
Nuclide RCS Activity* (Ci/gram) Nuclide RCS Activity* (Ci/gram)
Br-83 9.86E-02 Sr-89 4.04E-03 Br-84 4.88E-02 Sr-90 2.59E-04 Br-85 5.75E-03 Y-90 7.34E-05 I-127 (grams) 1.24E-10 Y-91m 3.01E-03 I-129 7.17E-08 Sr-91 5.60E-03 I-130 4.65E-02 Y-91 5.64E-04 I-132 3.39E+00 Sr-92 1.31E-03 1-134 7.30E-01 Y-92 1.13E-03 Kr-83m 4.62E-01 Y-93 3.82E-04 Kr-85m 1.83E+00 Zr-95 7.02E-04 Kr-85 1.00E+01 Nb-95 7.03E-04 Kr-87 1.19E+00 Mo-99 8.94E-01 Kr-88 3.29E+00 Tc-99m 8.22E-01 Kr-89 9.30E-02 Ru-103 7.43E-04 I-131 3.28E+00 Rh-103m 7.44E-04 Xe-131m 3.74E+00 Ru-106 3.25E-04 Xe-133m 5.54E+00 Ag-110m 3.34E-03 I-133 5.04E+00 Te-125m 6.03E-04 Xe-133 3.08E+02 Te-127m 4.49E-03 Xe-135m 6.15E-01 Te-127 1.52E-02 I-135 2.85E+00 Te-129m 1.46E-02 Xe-135 8.12E+00 Te-129 1.63E-02 Xe-137 2.18E-01 Te-131m 4.18E-02 Xe-138 7.59E-01 Te-131 1.63E-02 Rb-86 4.33E-02 Te-132 3.43E-01 Rb-88 4.08E+00 Te-134 3.51E-02 Rb-89 1.89E-01 Ba-140 4.55E-03 Cs-134 4.82E+00 La-140 1.53E-03 Cs-136 4.35E+00 Ce-141 6.94E-04 Cs-137 2.68E+00 Ce-143 5.46E-04 Cs-138 1.16E+00 Pr-143 6.46E-04 Ce-144 5.37E-04
- Results include a fuel management multiplier of 1.04 (1) Refer to Table 11.1A-1 for assumptions.
Rev. 34 WOLF CREEK
TABLE 11.1-6
CONTAINED SOURCES OF THE RADIOACTIVE WASTE MANAGEMENT SYSTEMSAND LARGE POTENTIALLY RADIOACTIVE OUTSIDE STORAGE TANKS
Component: Refueling Water Diameter, ft: 40.0 Location: Outside Height, ft: 34.5 Source volume, gal (1): 133,600Storage Tank
Class 1 Inventory (2)Ci Concentration (3)Ci/gm Class 5 Inventory (2)Ci Concentration(3)
µCi /gm
Kr-83m NEG NEG H-3 3.79E+03 2.5E+0 Kr-85m NEG NEG Kr-85 NEG NEG Kr-87 NEG NEG Class 6 Kr-88 NEG NEG Kr-89 NEG NEG Cr-51 3.47E-05 2.29E-08 Xe-131m NEG NEG Mn-54 6.99E-06 4.62E-09 Xe-133m NEG NEG Fe-55 3.66E-05 2.42E-08 Xe-133 NEG NEG Fe-59 1.99E-05 1.32E-08 Xe-135m NEG NEG Co-58 3.36E-04 2.22E-07 Xe-135 NEG NEG Co-60 4.58E-05 3.03E-08 Xe-137 NEG NEG Sr-89 5.92E-05 9.78E-09 Xe-138 NEG NEG Sr-90 1.92E-06 3.17E-10 Sr-91 NEG NEG Total noble gas NEG NEG Y-89m 5.33E-09 NEG Y-90 1.76E-05 2.90E-10 Class 2 Y-91m NEG NEG Y-91 1.17E-05 1.93E-09 Br-83 NGE NGE Y-93 NEG NEG Br-84 NGE NGE Zr-95 1.25E-06 8.27E-10 Br-85 NGE NGE Nb-95m 1.06E-06 7.01E-10 I-130 NGE NGE Nb-95 1.31E-06 8.65E-10 I-131 2.34E-02 3.87E-06 Mo-99 1.59E-03 2.62E-07 I-132 3.57E-04 5.89E-08 Tc-99m NEG NEG I-133 4.55E-05 7.52E-09 Ru-103 8.81E-07 5.82E-10 I-134 NEG NEG Ru-106 2.26E-07 1.49E-10 I-135 NEG NEG Rh-103m NEG NEG Rh-106 NEG NEG Te-125m 5.97E-07 3.95E-10 Total halogens 2.38E-02 3.94E-06 Te-127m 6.07E-06 4.01E-09 Te-127 6.09E-06 4.03E-09 Class 3 Te-129m 2.67E-05 1.76E-08 Te-129 1.71E-05 1.13E-08 Rb-86 3.38E-05 5.59E-09 Te-131m 3.41E-08 2.25E-10 Rb-88 NEG NEG Te-131 6.22E-08 4.11E-11 Cs-134 1.39E-02 2.30E-06 Te-132 8.65E-05 5.72E-08 Cs-136 4.45E-03 7.35E-07 Ba-137m 9.55E-03 1.58E-06 Cs-137 1.01E-02 1.67E-06 Ba-140 2.56E-05 4.22E-09 La-140 2.90E-05 4.78E-09 Total Cs, Rb 2.85E-02 4.71E-07 Ce-141 1.10E-05 1.82E-09 Ce-143 7.26E-08 1.20E-11 Class 4 Ce-144 6.19E-06 1.02E-09 Pr-143 6.52E-07 1.08E-09 N-16 NEG NEG Pr-144 6.20E-06 1.02E-09
Total other 1.19E-02 2.29E-06 isotopes
Notes:
(1) For liquid vessels, this is based(3) Source is based on 0.25 percent on at least 80 percent of vesselfuel defects usable volume
(2) Source is based on 1.0 percentfuel defectsNEG - negligible
Rev. 14 WOLF CREEK
TABLE 11.1-6 (Sheet 2)
Component: Boron Recycle Holdup Diameter, ft: 21 Location: Radwaste BuildingTankAorB Height, ft: 31 Source Volume, gal (1): 44,800
Inventory (2) Concentration (3) Inventory (2) Concentration(3)
Class 1 Ci µCi /gm Class 5 Ci µCi /gm
Kr-83m 5.02E-01 7.40E-04 H-3 5.92E+02 3.50E+00 Kr-85m 4.93E+00 7.27E-03 Kr-85 1.59E+03 2.35E+00 Tritium Kr-87 9.06E-01 1.34E-03 Kr-88 5.80E+00 8.56E-03 Class 6 Kr-89 3.10E-03 4.57E-06 Xe-131m 3.35E+02 4.94E-01 Cr-51 5.48E-03 3.23E-05 Xe-133m 1.40E+02 2.06E-01 Mn-54 1.12E-03 6.62E-06 Xe-133 1.68E+04 2.47E+01 Fe-55 5.88E-03 3.47E-05 Xe-135m 1.06E-01 1.56E-04 Fe-59 3.16E-03 1.86E-05 Xe-135 4.40E+01 6.49E-02 Co-58 5.36E-02 3.16E-04 Xe-137 6.97E-03 1.03E-05 Co-60 7.38E-03 4.35E-05 Xe-138 9.38E-02 1.38E-04 Sr-89 1.65E-02 2.44E-05 Sr-90 7.04E-04 1.04E-06 Total noble gas1.89E+04 2.78E+01 Y-90 5.15E-03 7.59E-06 Sr-91 5.23E-06 7.71E-09 Y-91M 3.79E-05 5.59E-08 Class 2 Y-91 1.87E-03 2.76E-06 Sr-92 1.26E-05 1.86E-08 Br-83 2.88E-03 4.25E-06 Y-92 4.76E-05 7.02E-08 Br-84 3.40E-04 5.01E-07 Y-93 4.73E-05 6.97E-08 Br-85 3.61E-06 5.32E-09 Zr-95 2.17E-03 3.20E-06 I-129 2.18E-07 3.22E-10 Nb-95 2.39E-03 3.52E-06 I-130 4.89E-03 7.20E-06 Mo-99 5.72E-01 8.43E-04 I-131 4.96E+00 7.32E-03 Tc-99M 5.26E-01 7.75E-04 I-132 7.90E-02 1.16E-04 Ru-103 1.66E-03 2.45E-06 I-133 1.13E+00 1.66E-03 Rh-103M 5.28E-06 7.78E-09 I-134 7.25E-03 1.07E-05 Ru-106 6.00E-04 8.84E-07 I-135 2.09E-01 3.08E-04 Rh-106 1.46E-08 2.15E-11 Ag-110M 4.72E-03 6.96E-06 Total halogens 6.39E+00 9.43E-03 Te-125M 1.94E-03 2.86E-06 Te-127M 1.02E-02 1.51E-05 Te-127 1.10E-02 1.63E-05 Class 3 Te-129M 3.10E-02 4.57E-05 Te-129 1.98E-02 2.91E-05 Rb-86 8.92E-01 1.32E-03 Te-131M 8.13E-03 1.20E-05 Rb-88 6.01E+00 8.86E-03 Te-131 6.35E-05 9.36E-08 Rb-89 1.12E-02 1.64E-05 Te-132 2.45E-01 3.61E-04 Cs-134 1.05E+02 1.54E-01 Te-134 2.75E-04 4.06E-07 Cs-136 7.65E+01 1.13E-01 Ba-137M 8.25E+01 1.22E-01 Cs-137 8.72E+01 1.29E-01 Ba-140 9.16E-03 1.35E-05 Cs-138 8.64E-02 1.27E-04 La-140 9.09E-03 1.34E-05 Ce-141 1.89E-03 2.79E-06 Total Cs, Rb2.75E+02 4.06E-01 Ce-143 1.94E-04 2.86E-07 Pr-143 1.52E-03 2.23E-06 Class 4 Ce-144 1.69E-03 2.50E-06 Pr-144 1.69E-03 2.50E-06 N-16 NEG NEG Total other isotopes8.41E+01 1.24E-01
Notes:
(1) Tank liquid usable volume is 44800 gal.
(2) Based on 1.00% fuel defects.
(3) Based on 0.25% fuel defects.
Rev. 14 WOLF CREEK
TABLE 11.1-6 (Sheet 3)
Component: Spent Resin StorageDiameter, ft: 7 Location: Radwaste BuildingTank (Primary)Height, ft: 10.7Source volume, ft3 (1): 280
Inventory (2) Concentration (3) Inventory (2) Concentration(3)
Class 1 Ci Ci /gm Class 5 Ci µCi /gm
Kr-83m NEG NEG H-3 NEG NEG Kr-85m NEG NEG Kr-85 NEG NEG Kr-87 NEG NEG Class 6 Kr-88 NEG NEG Kr-89 NEG NEG Cr-51 2.99E+01 3.90E+00 Xe-131m NEG NEG Mn-54 2.91E+01 3.80E+00 Xe-133m NEG NEG Fe-55 1.93E+02 2.52E+01 Xe-133 NEG NEG Fe-59 2.49E+01 3.26E+00 Xe-135m NEG NEG Co-58 6.10E+02 7.98E+01 Xe-135 NEG NEG Co-60 2.56E+02 3.34E+01 Xe-137 NEG NEG Sr-89 9.80E+00 2.67E+00 Xe-138 NEG NEG Sr-90 1.35E+00 3.67E-01 Sr-91 NEG NEG Total noble gas NEG NEG Y-90 1.33E+00 3.62E-01 Y-91m NEG NEG Class 2 Y-91 2.18E+00 5.93E-01 Y-93 NEG NEG Br-83 NEG NEG Zr-95 2.12E+00 2.77E-01 Br-84 NEG NEG Nb-95m 2.11E+00 2.76E-01 Br-85 NEG NEG Nb-95 3.00E+00 3.92E-01 I-130 5.80E-01 1.57E-01 Mo-99 1.36E+02 3.71E+01 I-131 1.17E+03 3.16E+02 Tc-99m NEG NEG I-132 5.20E+01 7.51E+00 Ru-103 9.98E-01 1.31E-01 I-133 1.76E+02 4.80E+01 Ru-106 9.89E-01 1.29E-01 I-134 9.08E-01 2.47E-01 Rh-103m NEG NEG I-135 2.83E+01 7.73E+00 Rh-106 NEG NEG Te-125m 9.18E-01 1.20E-01 Total halogens 1.43E+03 3.80E+02 Te-127m 1.50E+01 1.96E+00 Te-127 1.52E+01 1.99E+00 Class 3 Te-129m 2.69E+01 3.51E+00 Te-129 1.72E+01 2.25E+00 Rb-86 7.91E-01 2.15E-01 Te-131m 1.83E+00 2.39E-01 Rb-88 1.39E+00 3.80E-01 Te-131 NEG NEG Cs-134 1.78E+03 4.85E+02 Te-132 5.15E+01 6.74E+00 Cs-136 8.91E+01 2.43E+01 Ba-137m 1.40E+03 3.81E+02 Cs-137 1.48E+03 4.03E+02 Ba-140 1.63E+00 4.44E-01 La-140 1.77E+00 4.82E-01 Total Cs, Rb 3.35E+03 9.13E+02 Ce-141 1.28E+00 3.48E-01 Ce-143 NEG NEG Class 4 Ce-144 3.00E+00 8.15E-01 Pr-143 4.25E-01 1.16E-01 N-16 NEG NEG Pr-144 3.00E+00 8.15E-01
Total other isotopes2.89E+03 5.93E+02
Notes:
(1) For liquid vessels, this is based(3) Source is based on 0.25 percent on 80 percent of vessel usable fuel defects.
volume.
(4) Liquid activities are obtained by multi-(2) Source is based on 0.12 percent fuelplying inventory and concentration by .001.
defects and 1 year accumulated activity.
Rev. 14 WOLF CREEK
TABLE 11.1-6 (Sheet 4)
Component: Secondary Liquid Waste SystemDiameter, ft: 12 Drain Collector Tank A or B Location: Turbine Building Height, ft:22.75 Source volume, gal (1): 12,600
Class 1 Inventory (2)CiConcentration (3)Ci /gmClass 5Inventory (2)CiConcentration(3)
µCi /gm
Kr-83m NEG NEG H-3 1.66E-01 3.49E-03 Kr-85m NEG NEG Kr-85 NEG NEG Kr-87 NEG NEG Class 6 Kr-88 NEG NEG Kr-89 NEG NEG Cr-51 1.89E-09 3.98E-11 Xe-131m NEG NEG Mn-54 4.28E-10 8.99E-12 Xe-133m NEG NEG Fe-55 1.72E-09 3.60E-11 Xe-133 NEG NEG Fe-59 1.27E-09 2.67E-11 Xe-135m NEG NEG Co-58 1.70E-08 3.57E-10 Xe-135 NEG NEG Co-60 1.93E-09 4.05E-11 Xe-137 NEG NEG Sr-89 3.54E-09 1.86E-11 Xe-138 NEG NEG Sr-90 7.13E-11 3.75E-13 Sr-91 1.36E-09 7.13E-12 Total noble gasNEG NEG Y-90 2.58E-11 1.36E-13 Y-91m 9.21E-10 4.84E-12 Class 2 Y-91 5.51E-10 2.90E-12 Y-93 7.00E-11 3.68E-13 Br-83 1.68E-08 8.83E-11 Zr-95 8.53E-11 1.79E-12 Br-84 9.95E-10 5.23E-12 Nb-95m 1.24E-11 2.61E-13 Br-85 NEG NEG Nb-95 8.47E-11 1.78E-12 I-130 4.44E-08 2.34E-10 Mo-99 5.97E-07 3.14E-09 I-131 1.49E-05 7.85E-08 Tc-99m NEG NEG I-132 4.58E-07 2.68E-09 Ru-103 4.23E-11 8.89E-13 I-133 1.17E-05 6.17E-08 Ru-106 8.54E-12 1.79E-13 I-134 3.87E-08 2.03E-10 Rh-103m NEG NEG I-135 2.26E-06 1.19E-08 Rh-106 NEG NEG Te-125m 2.13E-11 4.47E-13 Total halogens 2.94E-05 1.55E-07 Te-127m 2.13E-10 4.48E-12 Te-127 3.84E-10 8.06E-12 Class 3 Te-129m 1.27E-09 2.66E-11 Te-129 8.71E-10 1.83E-11 Rb-86 7.52E-10 3.95E-12 Te-131m 1.47E-09 3.08E-11 Rb-88 2.91E-09 1.53E-11 Te-131 2.75E-10 5.77E-12 Cs-134 2.28E-07 1.20E-09 Te-132 1.84E-08 3.86E-10 Cs-136 1.13E-07 5.94E-10 Ba-137m 1.56E-07 8.21E-10 Cs-137 1.65E-07 8.66E-10 Ba-140 1.72E-09 9.02E-12 La-140 1.44E-09 7.56E-12 Total Cs, Rb5.10E-07 2.68E-09 Ce-141 7.04E-10 3.70E-12 Ce-143 1.26E-10 6.63E-13 Class 4 Ce-144 3.57E-10 1.88E-12 Pr-143 3.50E-10 1.84E-12 N-16 NEG NEG Pr-144 3.61E-10 1.90E-12
Total other isotopes8.12E-07 5.01E-09
Notes:
(1) For liquid vessels, this is based(3)Source is based on 0.25 percent on 84 percent of vessel usable fuel defects.
Volume. NEG - negligible
(2) Source is based on 1.0 percent fuel defects.
Rev. 14 WOLF CREEK
APPENDIX 11.1A
PARAMETERS FOR CALCULATION OF SOURCE TERMS FOR EXPECTED RADIOACTIVE CONCENTRATIONS AND RELEASES
11.1A.1 Regulatory Guide 1.112 provides guidelines for developing radioactive source terms. The following parameters and models are used to calculate radioactive source terms for the evaluation of radioactive waste treatment systems in determining the impact of radioactive effluents on the environment.
Figure 11.1A-1 shows a block diagram of liquid releases, and Table 11.1A-2 and Figure 11.1A-2 provide the volume, radioactivity level, and decontamination factors (DF) for each liquid path.
Figure 11.1A-3 shows a block diagram of gaseous releases, and Tables 11.1A-3 and 11.1A-4 provide the volume, radioactivity level, and DF for each gaseous path.
11.1A.2 The basic plant data for the source term calculations are provided in Table 11.1A-1.
Table 11.1A-5 provides summary GALE Code input data.
The following sections discuss the detailed design of waste systems:
- a. Chemical and volume control 9.3.4
- b. Gaseous radwaste 11.3
- c. Liquid radwaste 11.2
- d. Boron recycle 9.3.6
- e. Secondary liquid waste 10.4.10
- f. Steam generator blowdown 10.4.8
The plant ventilation systems are discussed in Section 9.4.
11.1A-1 Rev. 14 WOLF CREEK
TABLE 11.1A-1 PLANT DATA FOR SOURCE TERM CALCULATIONS
I. Reactor Power, MWt 3565 x 1.02 = 3636 II. Fuel Data
- a. Number of fuel assemblies 193
- c. Enrichment, w/o 5.0
- d. Operation time, days 510
- e. Fuel with defects, % 1.0, 0.25, 0.125 III. Plant Parameters
- a. Reactor coolant average temperature, °F 593.2
- b. System pressure, psia 2250
- c. Letdown rate, gpm 75
- d. Mixed bed demineralizer volume, ft33 30
- e. Cation demineralizer volume, ft 30
- f. Cation demineralizer effective flow, gpm 7.5
- g. Volume control tankLiquid volume, ft3 200
3 Vapor volume, ft 200 Pressure, nominal, psig 115-1250-30
- h. Chemical and volume control systemTemperature, °FSee Figure 11.1A-2 parameter (Sheet 1) and Table11.1A-2
- i. Boron recycle system parameters (Sheet 2) and TableSee Figure 11.1A-2 11.1A-2 IV. Secondary System Parameters
- a. Steam flow rate, 107 lbs/hr 1.592
- b. Secondary side water, 105 lbs 3.82
- c. Steam fraction in the secondary 0.08
- d. Moisture carryover fraction from the steam generator0.25
- e. Primary to secondary leak rte, gpm 1
- f. Steam generator blowdown rate, gpm 360
Rev. 13 WOLF CREEK
TABLE 11.1A-1 (Sheet 2)
V. Liquid Waste Processing Systems
- 1. Liquid radwaste system design parameters See Figure 11.1A-2 (Sheets 3,4,5) and Table 11.1A-2
- 2. Secondary liquid waste system design See Figure 11.1A-2 parameters (Sheet 7) and Table 11.1A-2
VI. Gaseous Waste Processing System
Gaseous radwaste system design parameters See Figure 11.1A-3 and Tables 11.1A-3
& 4
VII. Ventilation and Exhaust Systems
HVAC system design parameters See Figure 11.1A-3 and Tables 11.1A-3
& 4
Rev. 31
~LF CREEK
(7.5gpmJ Divert to
~cycle System (1.840 gpd 0 1.0 PCA) 2 Ve ntto Ga ~seous Radwaste System letdown Return to (76 11Pn1 @ 1.0 PCA) Reactor Coolant System
12ECONT/!MINATIO~ FACTQB&
Cesium & Other Iodine Rubidium Nuclides
- 1. Mixed Bad Deminaralizars 10 2 10
- 2. Cation Bed Damineralizer* 1 10 10
- 3. Reactor Coolant Filter 1 1 1
4, Volume Control Tank (a)
System OF 10 '20 102
(a) For noble gases, a value of 0.25 Is built into the GALE code for the y parameter for the case of continuous VCT pi.R'ging.
OPDATED SAFETY WOLP CREEK ANALYSIS REPORT
FISURE 11.1A-2 Rev. 0 SYSTEM DECONTAMINATION FACTORS (SHEET 1)
.) ~ _)
WOLF CREEK
Laundry & Hot Showers ____ .. 1 2 I ' (
- Plant Discharge (450gpd)
(Built GALE code) into the
- .r
DECONTAMINATION FACTORS Cesium& Other
--Iodine Rubidium Nuclides
- 1. Laundry and Hot Shower Tank
- 2. Laundry and Hot Shower Filter 1 1 1 1 1 1
System DF <NOTE 1) 1 1 1
Decay Times*
L + H.S. Tank Collection Time T c "'~ 0.4 x.!~*ooo*= 8.9 days '0.4 x 10,000 ... 0.7 day Liquid Radwaste
- Tp* 5,:760 Laundry Train REV.8
. . UPDATED SAFETY WOLF CREEK ANALYSIS REPORT
- The GALE coda does not usa thasa decay credit factors.
FIGURE 11.1A-2 1.> VOLUMES ARE EXTREMELY CONSERVATIVE. SYSTEM DECONTAMINATION FACTORS LAUNDRY IS PROCESSED OFFSITE.
NO CONTAMINATED INFLUENTS ARE NORMALLY CSHEET 5)
RECEIVED BY THE L 8c HST.
~ ~
LowTDS 1 2 - -WOLF CREEK (12,857 gpd} .. 1
~
- 11 - 12 -..
~ 13 - -- Secondary Cycle 9 -
~
HighTDS 3 ~ - 10 (4.286 gpd} 6 7 ...... Pia ntDischarge
I i*~.-** - ~ -
- I !.@*
Secondary I Side Floor 4 5 I -
- Plant Discharge Drains -*
( 7200 . gpd) DECONTAMINATION FACTORS
Cesium & Other Iodine Rubidium Nuclides
- 1. Low TDS Collector Tank
- 2. Low TDS Filter 1 1 1
- 3. High TDS Collector Tank
- 4. Oil Interceptor
- 5. SLW Drain Collector Tank
- 6. SLW Filter 1 1 1
- 7. SLW Evaporator (available only for high TDS) 1o3 1o4 1o4
- 8. SLW Charcoal Adsorber
- 9. SLW Demineralizer (C)
- 10(1o2) 10(2) 10(1o2)
- 10. SLW Monitor Tank (Low TDS)
System OF- High TDS 1o4 1o5 UP LowTDS 1o2 2 1o2 Secondary Liquid 11.. SL W Rodlotton t.tonltor RE *95 Waste System
) Rev. 5 t2. Wostewoter Treatment F'ocllity WOLF CREEK
- 13. Lime Sludge Pond UPDATED SAFETY ANALYSIS REPORT
(a) Processing will be subject to chemistry requirements. FIGURE 11.1A-2 (b) No credit is taken for collection and processing times.
(c) Second number indicates Low TDS DF.
SYSTEM DECONTAMINATION FACTORS
SHEET 7 (
Wolf Creek
11.2 LIQUID WASTE MANAGEMENT SYSTEMS
Several systems within the plant serve to control, collect, process, handle, store, recycle, and dispose of liquid radioactive waste generated as a result of normal plant operation, including anticipated operational occurrences. This section discusses the design and operating features and performance of the liquid radwaste system and the performance of other liquid waste management systems which are discussed in other sections.
11.2.1 DESIGN BASES
11.2.1.1 Safety Design Basis
Except for two containment penetrations and the component cooling water side of the reactor coolant drain tank heat exchanger, the liquid radwaste system (LRWS) is not a safety-related system.
SAFETY DESIGN BASIS ONE - The containment isolation valves in the LRWS are selected, tested, and located in accordance with the requirements of 10 CFR 50, Appendix A, GDC-56, and 10 CFR 50, Appendix J, Type C testing.
11.2.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The LRWS, in conjunction with other liquid waste management systems, is designed to meet the requirements of the discharge concentration limits of 10 CFR 20 and the ALARA dose objective of 10 CFR 50, Appendix I.
POWER GENERATION DESIGN BASIS TWO - The LRWS uses design and fabrication codes consistent with quality group D (augmented), as assigned by Regulatory Guide 1.143, for radioactive waste management systems.
POWER GENERATION DESIGN BASIS THREE - Liquid effluent discharge paths are monitored for radioactivity and isolated upon detection of unacceptable radioactivity.
11.2.2 SYSTEM DESCRIPTION
11.2.2.1 General Description
This section describes the design and operating features of the LRWS. The performance of the LRWS, in conjunction with other liquid waste management systems, is discussed in Section 11.2.3. Detailed descriptions of other liquid waste management systems are provided in the following sections:
11.2-1 Rev. 13 Wolf Creek
- a. Boron recycle 9.3.6
- b. Steam generator blowdown 10.4.8
- d. Secondary liquid waste 10.4.10
The piping and instrumentation diagram for the LRWS is shown in Figure 11.2-1.
The LRWS collects and processes radioactive or potentially radioactive waste water. The LRWS consists of two subsystems designated as drain channel A and drain channel B. Drain channel A is for processing water which could be recycled and drain channel B is for processing water which would normally be discharged. Equipment drains and waste streams are segregated to prevent the intermixing of the liquid wastes. Tritiated waters (CRW), potentially radioactive nontritiated waste (DRW), and detergent waste (SRW) are discussed in Section 9.3.3. A drain system is also provided inside the containment to collect drainage and leakage and transfer it to an appropriate tank.
Operating experience has shown that operating dose rates and overall release of radioactivity to the environment are minimized by not recycling triatiated water to the Reactor Makeup Water Storage Tank (RMWST). This method of operation eliminates the potential for contamination of secondary systems while degassing the Reactor Makeup Water System (BL) water in the Demineralized Water Makeup Storage and Transfer System (AN).
The various waste streams are processed as follows:
BORON RECYCLE SYSTEM - The bulk of the radioactive liquid discharged from the reactor coolant system is processed by the Boron Recycle System as described in Section 9.3.6. This water is transferred from a Recycle Holdup Tank to the LRWS for processing by the Liquid Radwaste Processing Skid as indicated in Figure 11.1A-2.
TRITIATED WASTES - These consist of reactor coolant which has been exposed to the atmosphere and has become aerated. This waste consists of equipment drains, leakoffs, and overflows from tritiated systems (e.g., CVCS and reactor coolant samples which have not been chemically contaminated). This waste is typically collected in the floor and equipment drain system, transferred to the waste holdup tank and processed in the Liquid Radwaste Processing skid prior to entering the waste evaporator condensate tank, waste monitor tanks or secondary liquid waste monitor tanks. The processed wastes are analyzed for chemical and radioactive content in the waste evaporator condensate tank, waste monitor tanks (WMTs) or secondary liquid waste monitor tanks prior to being discharged.
11.2-2 Rev. 22 Wolf Creek
HIGH LEVEL CHEMICAL WASTE - High level chemical waste consists of plant samples which have been chemically contaminated and decontamination solutions used in the decontamination tanks located in the hot machine shop. These wastes are collected in the chemical drain tank. The contents are received and sampled by chemistry to ensure that no highly contaminated chemical solutions are allowed to enter the floor drain system. This is done by analyzing for conductivity and PH. (If an abnormal parameter exists the contents are drained in small quantities to the floor drain system to allow for dilution).
The chemical drain tank contents are processed by draining its contents to the Floor Drain Tanks for dilution then processed by the LRPS.
CONTROLLED ACCESS AREA FLOOR DRAINAGE - Controlled access area floor drain wastes are miscellaneous liquid wastes collected by the floor drain system within the radiologically controlled areas of the plant. The controlled access areas are radiation zones B through E and include the containment, auxiliary building, fuel building, radwaste building, hot machine shop, and the access control areas of the control building.
Floor drainage consists of miscellaneous leakage from systems within the above areas. Generally, the amount of highly radioactive reactor coolant leakage into the drain system is very small. The bulk of the water originates as leakage from nonradioactive or slightly radioactive systems, such as the service water and component cooling water systems. In addition to system leakage, the floor drain systems collect decontamination water used for area washdowns, spent fuel cask decontamination, and laboratory equipment decontamination and rinses. Highly contaminated chemical solutions are not allowed to enter the floor drain system in large volumes, and, therefore, are directed to the chemical drain tank for processing. During maintenance, equipment drains from nontritiated systems are directed to the floor drain system. Large volumes of component cooling water are not drained to the floor drain system to prevent contamination of the LRWS by corrosion inhibitors.
The floor drain tanks are processed through the liquid radwaste demineralizer skid. The FDT may contain chemical contaminants, mild decontamination solutions, organics, etc. Filtration and ion exchange are capable of providing the required purity for environmental discharge. Relatively small volumes of exchange media are consumed in comparison to the volumes of solidified concentrates generated by evaporator bottoms processing. Since the processed water is not recycled, it is not necessary to deaerate for discharge to the environment.
11.2-3 Rev. 22 Wolf Creek
The liquid waste charcoal adsorber (LWCA) should be used only if the presence of organics is detected. If the waste in the FDT has a low level of dissolved solids, an activity of less than 10-5 mCi/cc, and the operator intends to discharge, the floor drain tank filter, liquid waste charcoal adsorber, waste evaporator condensate filter, and waste monitor tank demineralizer in series may be used to process the waste effectively. This method of processing can also be employed when abnormally large volumes of floor drain wastes are to be processed. When the effluent has not been processed, it should be directed to an aerated waste monitor tank.
A second floor drain tank is available to allow one tank to be isolated and sampled prior to feeding the processing system while the other tank is available to receive wastes. The second floor drain tank also provides greater system storage volumes which will minimize inventory problems by providing greater surge capacity during periods of abnormal waste generation or equipment outages.
When processing floor drain waste it is highly desirable to operate with a known influent quality to ensure optimum system performance. This is normally accomplished by isolating the floor drain tank to be processed and withdrawing a sample to determine its chemical properties. The operator selects the appropriate process equipment.
If the sample indicates relatively clean waste (less than 25 ppm TDS without organic or boric acid contamination), it can be effectively processed through the demineralizer train. Waste is processed with the Liquid Radwaste Processing Skid. With known influent chemistry, the optimum process can be selected.
LAUNDRY AND PERSONNEL DECONTAMINATION WASTE - Laundry waste is generated by the radioactive contamination of protective clothing and gear. The use of vendor provided laundry services is employed to process laundry waste. The hot shower in the access control area is used only for personnel decontamination; consequently, its use should be infrequent.
The washing machine water supplies have been disconnected. Dryers, washing machines and the washing machine hot water heater tank have been removed.
Therefore, no laundry can be performed on site and no laundry water will be generated for processing through Radwaste Systems.
11.2-4 Rev. 31 Wolf Creek
The waste from personnel decontamination is collected in the chemical and detergent waste systems detergent drain tank and then transferred to the laundry and hot shower tank. Also, they may be transferred to the monitor tanks for discharge. Suspended solids are removed by strainers and filters located at the beginning of the processing train. The Laundry and Hot Shower Tank (LHST) contents are normally not reprocessed due to the small amount of water that would be recycled. The system generates low volumes due to contaminated laundry being processed offsite through vender services.
All tanks which contain or may contain concentrations of radioactivity have provisions to prevent the uncontrolled release of the fluid. Table 11.2-2 indicates the provisions made for each tank.
The system is designed to handle the occurrence of equipment faults of moderate frequency such as:
- a. Malfunction in the LWPS
Malfunction in this system could include such things as pump or valve failures or evaporator failure. Because of pump standardization throughout the system, a spare pump can be used to replace most pumps in the system.
There is sufficient surge capacity in the system to accommodate waste until the failures can be fixed and normal plant operation resumed.
11.2-5 Rev. 27 Wolf Creek
- b. Excessive leakage in reactor coolant system equipment
The system is designed to handle a 1-gpm reactor coolant leak in addition to the expected leakage of 50 lb/day (Ref. 1) during normal operation, which is discussed in Section 5.2.5. Operation of the system is almost the same for normal operation, except that the load on the system is increased. A 1-gpm leak into the reactor coolant drain tank is handled automatically. If the 1-gpm leak enters the waste holdup tank, operation is the same as normal, except for the increased load on the system. Abnormal liquid volumes of reactor coolant resulting from excessive reactor coolant or auxiliary building equipment leakage (in excess of 1 gpm) can also be accommodated by the floor drain tank and processed by the LWPS.
- c. Excessive leakage in the auxiliary system equipment
Leakage of this type could include water from steam side leaks and fan cooler leaks inside the containment which are collected in the containment sump and sent to the floor drain tank. Other sources could be component cooling water leaks, service water leaks, and secondary side leaks. This water enters the floor drain tank and is processed and discharged as during normal operation.
11.2.2.2 Component Description
Codes and standards applicable to the LRWS are listed in Tables 3.2-1 and 11.2-
- 1. The LRWS is designed and constructed in accordance with quality group D (augmented). The LRWS is housed within a seismically designed building.
Regulatory Guide 1.143 is complied with to the extent specified in Table 3.2-5.
REACTOR COOLANT DRAIN TANK PUMPS - Due to the relative inaccessability of the containment and the loop drain requirements, two pumps are provided. One pump provides sufficient flow for normal tank operation with one pump for standby.
WASTE EVAPORATOR FEED PUMP - One standard pump is used. The waste evaporator feed pump supplies feed to the evaporator and the liquid radwaste demineralizer skid (LRDS). The pump is shut off when low level is reached in the waste holdup tank.
11.2-6 Rev. 14 Wolf Creek
WASTE EVAPORATOR CONDENSATE TANK PUMP - The waste evaporator condensate tank pump is a transfer pump. One standard pump is used to transfer the contents of the waste condensate tank to the waste monitor tanks.
CHEMICAL DRAIN TANK PUMP - One standard pump is used to recirculate the liquid back to the chemical drain tank for mixing prior to sampling.
LAUNDRY AND HOT SHOWER TANK PUMP - One standard pump is used to transfer the water to the waste monitor tank.
FLOOR DRAIN TANK PUMPS - Two standard pumps are available to transfer the contents of the floor drain tanks to the waste monitor tank. The pumps are cross-connected to the pump from either floor drain tank. The pumps can also be used to supply the LRDS.
WASTE MONITOR TANK PUMPS - One standard pump is to be used for each tank to discharge water from the plant site or for recycle if further processing is required. The pump may also be used for circulating the water in the waste monitor tank in order to obtain uniform tank contents and hence a representative sample before discharge. The pump can be throttled to achieve the desired discharge rate.
REACTOR COOLANT DRAIN TANK HEAT EXCHANGER - The reactor coolant drain tank heat exchanger is a U-tube type with one shell pass and four tube passes. Although the heat exchanger is normally used in conjunction with the reactor coolant drain tank, it can also cool the pressurizer relief tank from 200 to 120°F in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
REACTOR COOLANT DRAIN TANK - One tank is provided to collect leakoff type drains inside the containment at a central collection point for further disposition through a single penetration via the reactor coolant drain tank pumps.
Only water which can be directed to the recycle holdup tanks enters the reactor coolant drain tank. The tank is provided with a hydrogen or nitrogen cover gas. The water must be compatible with reactor coolant.
Sources of water entering the reactor coolant drain tank include the reactor vessel flange leakoff, reactor coolant pump number two seal leakoffs, and the excess letdown heat exchanger flow. No continuous leakage is expected from the reactor vessel flange during operation.
11.2-7 Rev. 29 Wolf Creek
The tank maintains a constant level to minimize the amount of gas sent to the gaseous waste processing system and also to minimize the amount of hydrogen or nitrogen required. The level is maintained by using a proportional control valve in the discharge line. This valve operates, on a signal from a level controller, to maintain a constant level by discharging normally to the recycle system. The remainder of the flow is recirculated to the tank.
WASTE HOLDUP TANK - One atmospheric pressure tank is provided outside the containment to collect equipment drainage, pump seal leakoffs, recycle holdup tank overflows, and other water from tritiated, aerated sources.
WASTE EVAPORATOR CONDENSATE TANK - One tank originally used to collect condensate from the waste evaporator which has been abandoned in place. This tank is now used for temporary water storage during outages or whenever a large surge of non-recyclable water occurs. The tanks damaged diaphragm has been removed.
CHEMICAL DRAIN TANK - One tank is provided to collect chemically contaminated tritiated water from the laboratories.
LAUNDRY AND HOT SHOWER TANK - One atmospheric pressure tank is used to collect laundry and hot shower drainage.
FLOOR DRAIN TANKS - Two atmospheric pressure tanks are used to collect floor drainage from the reactor plant operations.
WASTE MONITOR TANKS - The two atmospheric waste monitor tanks are provided for monitoring liquid discharges from the plant site. Each tank is sized to hold a volume large enough such that sampling requirements are minimized, thus minimizing laboratory effluent.
WASTE EVAPORATOR REAGENT TANK - One tank is used for adding chemicals to the plant for such things as cleaning of the waste evaporator tubes.
WASTE EVAPORATOR CONDENSATE DEMINERALIZER - One mixed bed demineralizer with nonregenerative hydrogen-hydroxide resin is provided to remove ionic contaminants from the waste condensate.
WASTE MONITOR TANK DEMINERALIZER - One mixed bed demineralizer with nonregenerative hydrogen-hydroxide resin is provided to remove trace contaminants from the water in the floor drain tank.
FILTERS - The filters provided are of a disposable-type cartridge.
11.2-8 Rev. 29 Wolf Creek
The methods employed to change filters and screens are dependent on activity levels. Filters are valved out of service, drained to the appropriate tank, and vented locally. If the radiation level of the filter is low enough, it is changed manually. Filter handling is discussed in Section 11.4.
STRAINERS - Strainers are provided in the discharge of the laundry and hot shower pump and the floor drain tank pumps to remove large particulate matter and thus prevent clogging of the downstream lines and filters.
WASTE EVAPORATOR - The waste evaporator is abandoned in place.
LIQUID RADWASTE PROCESSING SKID (LRPS)- The LRPS consists of a vendor supplied skid containing a chemical injection system, filtration unit and a series of demineralizer vessels. Based on the chemical and/or isotopic analysis of the waste stream, the processing skid may use every component available, or bypass those components not needed. The processes include filtration, reverse osmosis, and/or demineralization. Filtration removes large complex radioactive isotopes not easily removed by ion exchange from plant radioactive wastewater.
Reverse osmosis only allows water and selected ions to pass through a membrane.
Demineralization provides filtration and selective ion exchange. Following filtration, the radioactive contaminants or other solids left in solution are removed by reverse osmosis or demineralization.
11.2.2.3 System Operation
The LRWS operation is manually initiated, except for some functions of the reactor coolant drain subsystem. The system includes adequate control equipment to protect the system components and instrumentation and alarm functions to provide operator information to ensure proper system operation.
All pumps in the system have low level shutoffs, and all filters, strainers, and demineralizers have differential pressure indication to indicate fouling.
Operation of the LRWS is essentially the same during all phases of normal reactor plant operation; the only differences are in the load on the system.
The following sections discuss the operation of the system in performing its various functions. In this discussion, the term "normal operation" should be taken to mean all phases of operation, except operation under emergency or accident conditions. The LRWS is not regarded as a safety-related system.
REACTOR COOLANT DRAIN TANK SUBSYSTEM OPERATION - Normal operation of the reactor coolant drain subsystem is automatic and requires no operator action.
The system can be put in the manual mode, if desired. The leakage rate of reactor coolant pump No. 2 seal leakoffs, reactor vessel flange leakoffs, and discharges from the excess letdown heat exchanger into the reactor coolant drain tank (RCDT) can be estimated by putting the system
11.2-9 Rev. 22 Wolf Creek
in the manual mode, stopping operation of the reactor coolant drain tank pump, and watching the rate of level change. The reactor coolant drain tank pump normally discharges to the boron recycle system. These drains can also be processed in the waste holdup tank. The level in the RCDT is maintained by running one RCDT pump continuously and using a proportional control valve (LCV-1003) in the discharge line. This valve operates on a signal from the RCDT level controller to limit the flow out of the subsystem. The remainder of the flow is recirculated to the RCDT. The RCDT heat exchanger is sized to maintain the RCDT contents at or below 170°F, assuming an in-leakage of 10 gpm at 600°F.
A venting system is provided to prevent wide pressure variations in the RCDT.
Hydrogen or nitrogen cover gas is supplied from the service gas system and is automatically maintained between 2 and 6 psig by pressure-regulating valves.
PCV-7155 maintains a minimum tank pressure by admitting hydrogen or nitrogen, while PCV-7152 maintains maximum tank pressure by venting the RCDT to the gaseous radwaste system. The hydrogen is supplied from no more than two 194 SCF bottles, to limit the amount of hydrogen gas which might be accidentally released to the containment atmosphere. The RCDT vents to the gaseous radwaste system to limit any releases of radioactive gases.
The reactor coolant drain subsystem may also be used in the pressurizer relief tank (PRT) cooling mode of operation. In this mode, the level control valve in the discharge line to the recycle evaporator feed demineralizers (LCV-1003),
the isolation valve at the discharge of the reactor coolant drain tank (HV-7127) and the isolation valve in the reactor coolant drain tank recirculation line (HV-7144) are all closed. The PRT contents are circulated through the reactor coolant drain tank heat exchanger, via valve BB-HV-8031 and the reactor coolant drain tank pumps, prior to returning to the PRT via valve BB-HV-7141.
In this mode of operation, the RCDT heat exchanger is capable of cooling the PRT contents from 200°F to 120°F in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. As an alternative to returning the cooled fluid to the PRT, the fluid may be directly transferred to the recycle holdup tanks in the boron recycle system. In any and all cases of PRT cooling, the PRT is vented to less than 50 psig to prevent overpressurization of the RCDT subsystem.
The reactor coolant drain subsystem may be used to drain the reactor coolant loops by first venting the reactor coolant system, then connecting the spool piece in the RCDT pump suction piping. The design objective of this mode of operation is to drain the RCS to the midpoint of the reactor vessel nozzles in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with both RCDT pumps running. In this mode, valve HV-7144 is
11.2-10 Rev. 15 Wolf Creek
closed and, in order to maximize flow capability, the RCDT discharge level control valve (LCV-1003) may be bypassed during RCS draining operations. If automatic RCDT level control is desired, then the flow path through LCV-1003 may be used.
The reactor coolant drain subsystem may be used to drain down portions of the refueling pool which cannot be drained by the residual heat removal pumps. In this mode of operation, the RCDT heat exchanger may be bypassed and the RCDT level control valve (LCV-1003) may be bypassed to maximize flow through the fuel pool cooling and cleanup system to the refueling water storage tank. An alternate drain line is provided from the refueling pool to the containment sump to route decontamination chemicals away from the RCDT subsystem and minimize the possibility of contaminating any systems downstream of the RCDT pumps.
DRAIN CHANNEL "A" SUBSYSTEM OPERATION - Waste is accumulated in the waste holdup tank until a sufficient quantity exists to warrant processing. The Waste Holdup Tank contents are normally processed for discharge by the Liquid Radwaste Processing Skid. Processed effluent is not returned to the RMWS.
Demineralized LRWS effluent is discharged.
WASTE EVAPORATOR OPERATION - The waste evaporator is abandoned in place.
11.2-11 Rev. 19
Wolf Creek
DRAIN CHANNEL "B" SUBSYSTEM OPERATION - Normally, one floor drain tank is aligned to receive the discharge from the floor and equipment drain system, while the other tank is being used to supply waste to the processing system.
This procedure allows the waste to be sampled and pH adjusted prior to processing to ensure optimum system performance.
If the waste in the floor drain tank has a low total dissolved solids content
(<25 ppm), an activity of less than 10-5 mCi/cc, and does not contain significant organics, it may be processed using the liquid waste charcoal adsorber and waste monitor tank demineralizer in series, and directed to waste monitor tanks.
Any planned releases from the system must be weighted with all other unit radioactive liquid releases to ensure that the local releases do not exceed the ODCM limits at the boundary of the restricted area.
LAUNDRY SUBSYSTEM OPERATION - Waste from the personnel decontamination shower is directed by gravity drain to the detergent drain tank located in the basement of the control building. This waste is pumped to the LHST where it is sampled, prior to being processed. If discharge of the LHST contents is desired and the tank contents are found to be of acceptable quality for discharge, the fluid may be transferred to the Secondary Liquid Waste Monitor Tanks or Waste Monitor Tank "B" by way of the Laundry and Hot Shower Tank Basket Strainer and Filter.
The vendor provided laundry services for contaminated laundry is employed.
This helps prevent the spread of highly contaminated particles throughout the laundry water system.
The laundry water stored in the laundry water storage tank may also be directed to the LHST for reprocessing or to the waste monitor tank "B" or one of the secondary liquid waste monitor tanks. Any planned releases from this system must be weighed with all other radioactive liquid releases to ensure total releases do not exceed the ODCM limits at the boundary of the restricted area.
The LRWS is operated so that the waste discharges are segregated. Waste monitor tank "B" is normally aligned for laundry water while waste monitor tank "A" is normally aligned for demineralized floor drains. Laundry water is normally low radioactivity waste, and does not require treatment other than the removal of organics. Provision is made to demineralize the laundry water, via the waste monitor tank demineralizer, prior to discharge, if necessary.
11.2-12 Rev. 31 Wolf Creek
Floor drain wastes are relatively dirty and may contain moderately high radioactivity. Treatment of floor drain wastes prior to discharge consists of options for Ozone Injection, Ultra Filtration, Reverse Osmosis and demineralization. These options are provided using the (ZERO) liquid waste processing components.
The chemical drain tank (CDT) receives chemically contaminated tritiated water from the plant sample stations, and chemically contaminated decontamination wastes. Contents of the tank are sampled as process initiation levels are reached then drained to the FDT subsystem to dilute any high conductivity prior to being processed by the liquid waste process system. A high level alarm is provided from the CDT for operator information.
11.2.3 RADIOACTIVE RELEASES
This section describes the estimated liquid release from the plant for normal operation and anticipated operational occurrences.
11.2.3.1 Sources
Section 11.1 and Appendix 11.1A provide the bases for determining the contained sources inventory and the normal releases.
A survey has been performed of liquid discharges from different Westinghouse pressurized water reactor plants. The results are presented in Table 11.2-17 of Reference 2. The data includes radionuclides released on an unidentified basis, and are all within the permissible concentration for the release of liquid containing all unidentified radionuclide mixtures.
11.2.3.2 Release points
Radioactive plant wastes are treated inside the power block, where the majority of radioactive material is concentrated for offsite disposal. Water containing small concentrations of radioactivity is discharged from the power block to the environment as plant effluent. The effluent normally discharges from the plant into the circulating water discharge piping, which dilutes the power block effluent and conveys it to the cooling lake. The point of discharge into the cooling lake for these effluents is at the circulating water discharge structure (See Figure 11.2-1). Three other potential discharge points to the cooling lake are directly from the lime sludge pond, the oily waste separator, and the Technical Support Center. The Technical Support Center decontamination shower would only be used by E-Plan personnel if access control and rad waste showers were unavailable. These three pathways have no dilution. Further discussion of concentrations of radioactivity in the cooling lake from normal operational releases is provided in Section 11.2.3.3. A discussion of concentrations of radioactivity in the cooling lake from accidental release of liquid effluents is discussed in Section 2.4.12.
11.2-13 Rev. 23 Wolf Creek
This low level radioactive liquid effluent is stored in the power block in the primary and secondary waste monitor tanks (two each, four total) and the steam generator blowdown surge tank. Each of these tanks feeds into the liquid radwaste discharge line, which is connected to the circulating water discharge piping (See Figure 11.2-2). Tank discharge is initiated manually in all cases.
The minimum flow of dilution water which conveys the power block radioactive effluent to the cooling lake is 5,000 gpm. In the event that the dilution flow is less than 5,000 gpm, release of radioactive power block effluent is prohibited and is terminated through automatic controls at a point inside the power block.
Circulating water pumps and service water pumps provide dilution to discharge from the power block. The release of radioactive effluent from the power block is automatically terminated when no Circulating Water Pumps are in service.
Minimum dilution flow necessary for the discharge of radioactive effluents is established through administrative controls to ensure compliance with Federal discharge limits.
11.2.3.3 Dilution Factors
Liquid radioactive releases are normally diluted by cooling water with a flow rate of 1114 cfs and service water with a flow rate of 90 cfs for a total discharge of 1204 cfs. This is the normal dilution assumed for dose calculations to the maximum individual interacting with the cooling lake environment.
11.2.3.4 Estimated Doses
Preoperational estimates of doses from liquid effluents were shown to be in conformance with 10CFR50, Appendix I requirements. Actual dose from liquid effluents during plant operation are calculated using the approved methodology presented in the Offsite Dose Calculation Manual (ODCM). The ODCM describes the methods used for calculating concentration of radioactive material in the environment and the estimated potential offsite doses associated with liquid and gaseous effluents. The ODCM also specifies controls for release of liquid and gaseous effluents to ensure compliance with NRC regulations.
11.2.4 CALCULATIONAL BASIS FOR LIQUID SOURCE TERMS
The Wolf Creek Generating Station, Unit No. 1 uses the mixed bed demineralizer option shown in Item 5 of Figure 11.1A-2 (Sheet 2). The original GALE code input and annual liquid effluent releases are shown in Tables 11.2-10 and 11.2-11 respectively.
11.2-14 Rev. 23 Wolf Creek
11.2.5 SAFETY EVALUATION
Except for two associated containment penetrations and the CCW pressure boundary integrity at the reactor coolant drain tank, the LRWS is not a safety-related system.
SAFETY EVALUATION ONE - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability.
11.2.6 TESTS AND INSPECTION
Preoperational testing is discussed in Chapter 14.0.
The operability, performance, and structural and leaktight integrity of all system components are demonstrated by continuous operation.
11.2.7 INSTRUMENTATION DESIGN
The system instrumentation is described in Table 11.2-12 and shown on Figure 11.2-1.
The instrumentation readout is located mainly on the waste processing system panel in the radwaste building. Some instruments are read locally.
All alarms are shown separately on the waste processing system panel and further relayed to one common waste processing system annunciator on the main control board.
The waste processing system pumps are protected against loss of suction pressure by a control setpoint on the level instrumentation for the respective vessels feeding the pumps. The reactor coolant drain tank pumps and the spent resin sluice pump are, in addition, interlocked with flow rate instrumentation and stop operating when the delivery flows reach minimum setpoints.
Differential pressure indicators with local readout are provided for filters, strainers, and demineralizers.
11.
2.8 REFERENCES
- 1. NUREG-0017, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors" (PWR-GALE Code), NRC, April 1976, pg. 6-1.
- 2. "Appendix D to RESAR-3S, Liquid Waste Management System,"
WCAP 8665, March 1976.
11.2-15 Rev. 23 Wolf Creek
- 3. Attachment to Concluding Statement of Position of the Regulatory Staff. Public Rule-making Hearing on: Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Power Stations, USAEC, Docket No. RM-50-2, February 20, 1974.
- 4. Fletcher, J. F., and W. L. Dotson (compilers). HERMES-A Digital Computer Code for Estimating Regional Radiological Effects from the Nuclear Power Industry, USAEC. Report HEDL-TME-71-168, Hanford Engineering Development Laboratory, 1971.
- 5. Final Environmental Statement Concerning Proposed Rule Making Action: Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as Practicable:" for Radioactive Material in Light-Water-Cooled, Nuclear Power Reactor Effluents, USAEC Report WASH-1258, Washington, D.C., July 1973.
- 6. Lyon, R. J., Shearin, R. L., 1976, EPA-520 Radionuclide Accumulation in a Reactor Cooling Lake: USEPA, Office of Radiation Programs.
- 7. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, Office of Standards Development.
- 8. Regulatory Guide 1.113, Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I, Office of Standards Development.
- 9. Simpson, D. B., McGill, B. L., 1980, NUREG/CR-1276 User's Manual for LADTAP II Computer Program: U.S.N.R.C. and Oak Ridge National Laboratory.
11.2-16 Rev. 23 WOLF CREEK
TABLE 11.2-1
LIQUID WASTE PROCESSING SYSTEM EQUIPMENT PRINCIPAL DESIGN PARAMETERS
Reactor Coolant Drain Tank Pumps
Number 2
Type Horizontal centrifugal
Design pressure, psig 150
Design temperature, F 200
Design flow, gpm
Point l 100
Point 2 150
Design head, ft
Point 1 260
Point 2 250
Material Stainless steel
Design code MS
Waste Evaporator Feed Pump
Number l
Type Canned centrifugal
Design pressure, psig 150
Design temperature, F 200
Design flow, gpm
Point 1 35
Point 2 100
Design head, ft
Point l 250
Point 2 200
Rev. 16
WOLF CREEK
TABLE 11.2-1 (Sheet 2)
Material Stainless steel
Design code (1) MS
Waste Evaporator Condensate Pump Number 1
Type Canned centrifugal
Design pressure, psig 150 Design temperature, F 200
Design flow, gpm
Point 1 35 Point 2 100
Design head, ft
Point 1 250 Point 2 230
Material Stainless steel
Design code MS Chemical Drain Tank Pump
Number 1
Type Canned centrifugal Design pressure, psig 150
Design temperature, F 200
Design flow, gpm Point 1 35
Point 2 100
Design head, ft Point l 250
Point 2 230
Rev. 0
WOLF CREEK
TABLE 11.2-1 (Sheet 3)
Material Stainless steel
Design code MS
Laundry and Hot Shower Tank Pump Number l
Type Horizontal centrifugal
Design pressure, psig 150 Design temperature, F 200
Design flow, gpm
Point 1 35 Point 2 100
Design head, ft
Point 1 250 Point 2 230
Material Stainless steel
Design code MS Floor Drain Tank Pumps
Number 2
Type Horizontal centrifugal Design pressure, psig 150
Design temperature, F 200
Design flow, gpm Point 1 35
Point 2 100
Design head, ft Point l 250
Point 2 230
Rev. 0
WOLF CREEK
TABLE 11.2-1 (Sheet 4)
Material Stainless steel
Design code MS
Waste Monitor Tank Pumps Number 2
Type Canned centrifugal
Design pressure, psig 150 Design temperature, F 200
Design flow, gpm
Point 1 35 Point 2 100
Design head, ft
Point 1 250 Point 2 230
Material Stainless steel
Design code MS Laundry Water Storage Tank Pump
Number 1
Type Inline centrifugal Design pressure, psig 150
Design temperature, F 200
Design flow, gpm 35 Design head, ft 81
Material Stainless steel
Design code MS
Rev. 0
WOLF CREEK
TABLE 11.2-1 (Sheet 5)
Reactor Coolant Drain Tank Heat Exchanger
Number 1
Type U-tube Estimated UA, Btu/hr-F 70,000
Design flow, lb/hr
Shell 112,000 Tube 44,600 (See *)
Temperature in, F
Shell 105 Tube 180 (See *)
Temperature out, F
Shell 125 Tube 130
Material
Shell Carbon steel Tube Stainless steel
Design code
Shell side ASME Section III Tube side ASME Section VIII
- At Operating temp. 170 ° F, Flow is 55,581 #/hr
Reactor Coolant Drain Tank
Number 1
Type Horizontal
Usable volume, gal 350
Design pressure, psig* 100
Design temperature, F 250
- External design pressure is 60 psig.
Rev. 16
WOLF CREEK
TABLE 11.2-1 (Sheet 6)
Material Stainless steel Design code (1) ASME Section VIII
Waste Holdup Tank
Number l Type Vertical
Usable volume, gal 10,000
Design pressure Atmospheric Design temperature, F 200
Material Stainless steel
Design code (1) ASME Section VIII (no code stamp)
Waste Evaporator Condensate Tank Number 1
Type Vertical
Usable volume, gal 5,000 Design pressure, psig +0.433
Design temperature, F 200
Material Stainless steel Design code ASME Section VIII (no code stamp)
Chemical Drain Tank
Number 1
Type Vertical Usable volume, gal 600
Design pressure, psig +0.5
Design temperature, F 200 Material Stainless steel
Rev. 0
WOLF CREEK
TABLE 11.2-1 (Sheet 7)
Design code ASME Section VIII (no code stamp)
Laundry and Hot Shower Tank
Number 1
Type Vertical Usable volume, gal 10,000
Design pressure, psig +0.5
Design temperature, F 200 Material Stainless steel
Design code ASME Section VIII (no code stamp)
Floor Drain Tanks
Number 2 Type Vertical
Usable volume, gal 10,000
Design pressure, psig +0.5 Design temperature, F 200
Material Stainless steel
Design code ASME Section VIII (no code stamp)
Laundry Water Storage Tank Number l
Type Vertical
Usable volume, gal 10,000 Design pressure Atmospheric
Design temperature, F 200
Material Stainless steel Design code ASME Section VIII (no code stamp)
Rev. 0
WOLF CREEK
TABLE 11.2-1 (Sheet 8)
Waste Monitor Tanks
Number 2
Type Vertical Usable volume, gal 5,000
Design pressure, psig +0.5
Design temperature, F 200 terial Stainless steel
Design code ASME Section VIII (no code stamp)
Waste Evaporator Reagent Tank
Number 1 Type Vertical
Usable volume, gal 5
Design pressure, psig 150 Design temperature, F 200
Material Stainless steel
Design code ASME Section VIII Waste Evaporator Condensate Demineralizer
Number 1
Type Flushable Design pressure, psig 300
Design temperature, F 250
Design flow, gpm 120 Resin volume, ft3 max. 39
Material Stainless steel
Design code (1) ASME Section VIII
Rev. 0
WOLF CREEK
TABLE 11.2-1 (Sheet 9)
Waste Monitor Tank Demineralizer
Number l
Type Flushable Design pressure, psig 300
Design temperature, F 250
Design flow, gpm 120 Resin volume, ft3 max. 39
Material Stainless steel
Design code (1) ASME Section VIII Liquid Waste Charcoal Adsorber
Number 1
Type Flushable Design pressure, psig 150
Design temperature, F 200
Design flow rate, gpm 35 Charcoal volume, ft3 42
Material Stainless steel
Design code ASME Section VIII Laundry and Hot Shower Charcoal Adsorber
Number 1
Type Flushable Design pressure, psig 150
Design temperature, F 200
Design flow rate, (gpm) avg./max. 4/10 Charcoal volume, ft3 10
Material Stainless steel
Design code ASME Section VIII
Rev. 0
WOLF CREEK
TABLE 11.2-1 (Sheet 10)
Waste Evaporator Feed Filter
Number l
Design pressure, psig 300 Design temperature, F 250
Design flow, gpm 250
P at design flow, unfouled, psi 5 Particle Retention (see note 2 of Table 9.3-13)
Material Stainless steel
Design code (1) ASME Section VIII Waste Evaporator Condensate Filter (FHB10)*
Number l
Design pressure, psig 300 Design temperature, F 250
Design flow, gpm 250
P at design flow, unfouled, psi 5 Particle retention (See Note 2 of Table 9.3-13)
Material Stainless steel
Design code (1) ASME Section VIII Laundry and Hot Shower Tank Filter (FHB07)*
Number l
Design pressure, psig 300 Design temperature, F 250
Design flow, gpm 250
P at design flow, unfouled, psi 5 Particle retention (See Note 2 of Table 9.3-13)
- See comments on Sheet 2 of Table 9.3-13.
Rev. 11
WOLF CREEK
TABLE 11.2-1 (Sheet 11)
Material Stainless steel Design code (1) ASME Section VIII
Waste Monitor Tank Filter (FHB08)*
Number l Design pressure, psig 300
Design temperature, F 250
Design flow, gpm 250
P at design flow, unfouled, psi 5
Particle retention (See Note 2 of Table 9.3-13)
Material Stainless steel
Design code (1) ASME Section VIII
Floor Drain Tank Filter (FHB06)*
Number 1
Design pressure, psig 300
Design temperature, F 250
Design flow, gpm 250
P at design flow, unfouled, psi 5
Particle retention (See Note 2 of Table 9.3-13)
Material Stainless steel
Design code (1) ASME Section VIII
Rev. 10
WOLF CREEK
TABLE 11.2-1 (Sheet 12)
Liquid Radwaste Demineralizer Skid
Number 1 Design flow rate, gpm 50
Nominal Design pressure, PSIG Maximum 150
Design temperature, F Maximum 150
Material MS Design code ASME Section VIII
Laundry and Hot Shower Tank Strainer
Number l Design pressure, psig 150
Design temperature, F 200
Design flow, gpm 35
P at design flow, unfouled, psi 0.2
Basket perforation size, inch 1/16
Material Stainless steel
Design code ASME Section VIII
Floor Drain Tank Strainer
Number 1
Design pressure, psig 150
Design temperature, F 200
Design flow, gpm 35
P at design flow, unfouled, psi 0.2
Basket perforation size, inch 1/16
Material Stainless steel
Design code ASME Section VIII
Rev. 8
WOLF CREEK
TABLE 11.2-1 (Sheet 13)
Waste Evaporator (2)
Number 1
Steam design pressure, psig 50 Design feed flow, gpm 15
Feed concentration, boron, ppm 10-2,500
Bottoms concentration, boron, ppm 7,200-21,000 Material (for concentrates) Incoloy 825 (or equivalent)
Design code ASME Section VIII/TEMA C
(1) Table indicates that the required code is based on its safety-related importance as dictated by service and functional requirements and by the consequences of their failure. Note that the equipment may be supplied to a higher principal construction code than required.
(2) Equipment is abandoned in place.
Rev. 19
WOLF CREEK
TABLE 11.2-3 (Historical Information)
CALCULATED LIQUID EFFLUENT DISCHARGE CONCENTRATIONS FROM ROUTINE OPERATION
pCi/1 Release Circulating Cooling Isotopea Ci/yr Waterb Lakec LeRoyd
1H 3 4.10+002 2.38+004 2.34+004 7.38E+002 24CR 51 9.00-005 1.62-004 7.83-005 2.47E-006 25MN 54 1.20-004 1.22-003 1.11-003 3.50E-005 26FE 55 9.00-005 2.39-003 2.30-003 7.26E-005 26FE 59 5.00-005 1.17-004 7.02-005 2.22E-006 27CO 58 1.30-003 4.09-003 2.88-003 9.09E-005 27CO 60 9.80-004 4.07-002 3.98-002 1.26E-003 35BR 83 3.00-005 2.79-005 9.40-008 2.97E-009 42MO 99 1.80-003 1.84-003 1.67-004 5.27E-006 43TC 99M 1.70-003 1.60-003 1.33-005 4.20E-007 52TE 129M 7.00-005 1.39-004 7.42-005 2.34E-006 53I 131 9.50-002 1.12-001 2.40-002 7.57E-004 52TE 132 6.10-004 6.30-004 6.19-005 1.95E-006 53I 132 1.70-003 1.58-003 5.11-006 1.61E-007 53I 133 3.00-002 2.87-002 8.23-004 2.60E-005 55CS 134 8.10-003 1.79-001 1.71-001 5.40E-003 53I 135 5.20-003 4.89-003 4.55-005 1.44E-006 55CS 136 2.10-003 2.81-003 8.54-004 2.70E-005 55CS 137 7.30-003 5.03-001 4.96-001 1.57E-002 40ZR 95 1.40-004 4.14-004 2.84-004 8.96E-006 41NB 95 2.00-004 4.06-004 2.20-004 6.94E-006 37RB 86 2.00-005 3.03-005 1.17-005 3.69E-007 44RU 103 2.00-005 4.35-005 2.49-005 7.86E-007 44RU 106 2.40-004 2.90-003 2.68-003 8.46E-005 47AG 110M 4.00-005 3.46-004 3.09-004 9.75E-006 58CE 144 5.20-004 4.98-003 4.50-003 1.42E-004 38SR 89 2.00-005 5.12-005 3.26-005 1.03E-006 52TE 127M 1.00-005 4.32-005 3.39-005 1.07E-006 52TE 127 2.00-005 1.88-005 2.46-007 7.76E-009 52TE 129 5.00-005 4.65-005 7.49-008 2.36E-009 53I 130 1.00-004 9.49-005 1.63-006 5.14E-008 52TE 131M 3.00-005 2.90-005 1.18-006 3.72E-008 93NP 239 2.00-005 2.01-005 1.47-006 4.64E-008
aM = Metastable
bBased solely on dilution by the circulating water discharge and buildup of radionuclides over 40 year plant life.
cBased on dilution by the circulating water discharge and buildup of radionuclides in the cooling lake over 40 year plant life.
dConcentration of radionuclides at the LeRoy water works intake.
Based on dilution by circulating water discharge and build-up of radionuclides in the cooling lake over 40 year plant life and additional dilution in the Neosho River.
Rev. 19
WOLF CREEK
TABLE 11.2-4 (Historical Information)
BIOACCUMULATION FACTORS (pCi/kg per pCi/liter)
FRESHWATER SALTWATER ELEMENT FISH INVERTEBRATE FISH INVERTEBRTATE
H 9.0E-01 9.0E-01 9.0E-01 9.3E-01 C 4.6E 03 9.1E 03 1.8E 03 1.4E 03 NA 1.0E 02 2.0E 02 6.7E-02 1.9E-01 P 1.0E 05 2.0E 04 2.9E 04 3.0E 04 CR 2.0E 02 2.0E 03 4.0E 02 2.0E 03 MN 4.0E 02 9.0E 04 5.5E 02 4.0E 02 FE 1.0E 02 3.2E 03 3.0E 03 2.0E 04 CO 5.0E 01 2.0E 02 1.0E 02 1.0E 03 NI 1.0E 02 1.0E 02 1.0E 02 2.5E 02 CU 5.0E 01 4.0E 02 6.7E 02 1.7E 03 ZN 2.0E 03 1.0E 04 2.0E 03 5.0E 04 BR 4.2E 02 3.3E 02 1.5E-02 3.1E 00 RB 2.0E 03 1.0E 03 8.3E 00 1.7E 01 SR 3.0E 01 1.0E 02 2.0E 00 2.0E 01 Y 2.5E 01 1.0E 03 2.5E 01 1.0E 03 ZR 3.3E 00 6.7E 00 2.0E 02 8.0E 01 NB 3.0E 04 1.0E 02 3.0E 04 1.0E 02 MO 1.0E 01 1.0E 01 1.0E 01 1.0E 01 TC 1.5E 01 5.0E 00 1.0E 01 5.0E 01 RU 1.0E 01 3.0E 02 3.0E 00 1.0E 03 RH 1.0E 01 3.0E 02 1.0E 01 2.0E 03 TE 4.0E 02 6.1E 03 1.0E 01 1.0E 02 I 1.5E 01 5.0E 00 1.0E 01 5.0E 01 CS 2.0E 03 1.0E 03 4.0E 01 2.5E 01 BA 4.0E 00 2.0E 02 1.0E 01 1.0E 02 LA 2.5E 01 1.0E 03 2.5E 01 1.0E 03 CE 1.0E 00 1.0E 03 1.0E 01 6.0E 02 PR 2.5E 01 1.0E 03 2.5E 01 1.0E 03 ND 2.5E 01 1.0E 03 2.5E 01 1.0E 03 W 1.2E 03 1.0E 01 3.0E 01 3.0E 01 NP 1.0E 01 4.0E 02 1.0E 01 1.0E 01
- Regulatory Guide 1.109
Rev. 19
WOLF CREEK
TABLE 11.2-9 (Historical Information)
APPENDIX I CONFORMANCE
SUMMARY
TABLE FOR LIQUID EFFLUENTS
Type of Dose Design Calculated Point of Dose Liquid Effluents Objectivea Doseb Evaluation Dose to total body 3 mrem/yr 2.51 mrem/yr b Point of from all pathways per site Discharge, Cooling Lake Dose to any organ 10 mrem/yr 3.63 mrem/yr c Same as above from all pathways per site
aAppendix I design objectives from Sections II.A, II.B, II.C, and II. D (by Annex, RM50-2) of Appendix I, 10CFR Part 50; considers doses to maximum individual.
bMaximum dose to an individual from all liquid pathways.
cMaximum dose to a teen liver from all liquid pathways.
Rev. 19
r .. --.. *-- c . ~." -*- --------'---.. ~--- .. ----... -.r*~- *:-.--.-,...,.. --_,..-~--*. :------*---.---.---r*--. -:
I WOLF ~cEK I o I I
I
I I cnrr* &: I . SITE UNDERGROUND PIPING
z* '
en I
RADIOACTIVE LIQUID RELEASE
~6
144" CIRCULATING WATER <AND SERVICE
- WATER DISCHARGE>
CIRCULATING WATER DISCHARGE STRUCTURE
42" WARMING LINE~ WARMING LINE TO
. .. CIRCULATING WATER SCREENHOUSE r REV. 12
I. WOLF Cl\\BBK UPDATED SAFETY ANALYSIS REPORT ..
FIGURE 11.2-1 I RADIOACTIVE FLOW LIQUID DIAGRAM RELEASE
- -. - .. - , -- -. -'---~ .. - ---- " - ------- . ---- ----. - . -- . -- .. -. ---. -" ---. ~ ---.. --.. -....... -~s~~~ ~~ . J WOLF CREEK
11.3 GASEOUS WASTE MANAGEMENT SYSTEMS
The gaseous radwaste system (GRWS) and the plant ventilation exhaust systems control, collect, process, store, and dispose of gaseous radioactive wastes generated as a result of normal operation, including anticipated operational occurrences. This section discusses the design, operating features, and performance of the GRWS and the performance of the ventilation systems. The plant ventilation exhaust systems accommodate other potential release paths for gaseous radioactivity due to miscellaneous leakages, aerated vents from systems containing radioactive fluids, and the removal of noncondensables from the secondary system. Systems which handle these gases are not normally considered gaseous waste systems and are discussed in detail in other sections. These systems are included here to the extent that they represent potential release paths for gaseous radioactivity.
11.3.1 DESIGN BASES
11.3.1.1 Safety Design Basis
The GRWS and other gaseous waste management systems serve no safety-related function.
11.3.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The GRWS and the ventilation exhaust systems are designed to meet the requirements of the discharge concentration limits of 10 CFR 20 and the as low as reasonably achievable dose objective of 10 CFR 50, Appendix I.
POWER GENERATION DESIGN BASIS TWO - The GRWS includes design features to preclude the possibility of an explosion where a potential for an explosive mixture exists.
POWER GENERATION DESIGN BASIS THREE - The GRWS uses design and fabrication codes consistent with quality group D (augmented), as assigned by Regulatory Guide 1.143 for radioactive waste management systems.
POWER GENERATION DESIGN BASIS FOUR - The ventilation exhaust system complies with Regulatory Guide 1.140 to the extent specified in Table 9.4-3.
POWER GENERATION DESIGN BASIS FIVE - Gaseous effluent discharge paths are monitored for radioactivity.
11.3-1 Rev. 13 WOLF CREEK
POWER GENERATION DESIGN BASIS SIX - The Radwaste Building (including the Waste Bale Drumming Area) is equipped with a monitored ventilation system which ensures that the potential release pathways are controlled and monitored as per 10 CFR 50, Appendix A, in case of a breach of container.
11.3.2 SYSTEM DESCRIPTIONS
11.3.2.1 General Description
This section describes the design and operating features of the GRWS. The performance of the GRWS and other plant gaseous waste management systems with respect to the release of radioactive gases is discussed in Section 11.3.3.
Detailed descriptions of the plant ventilation systems and main condenser evacuation system are presented in Sections 9.4 and 10.4.2, respectively.
The piping and instrumentation diagram for the GRWS is shown in Figure 11.3-1.
The main flow path in the GRWS is a closed loop comprised of two waste gas compressors, two catalytic hydrogen recombiners, six gas decay tanks for normal power service, and two gas decay tanks for service at shutdown and startup.
The system also includes a gas decay tank drain collection tank, drain pump, four gas traps to handle normal operating drains from the system, and a waste gas drain filter to permit maintenance and handle normal operating drains from the system. All of the equipment is located in the radwaste building.
The closed loop has nitrogen for a carrier gas. The primary influents to the GRWS are combined with hydrogen as the stripping or carrier gas. The hydrogen that is introduced to the system is recombined with oxygen, and the resulting water is removed from the system. As a result, the bulk of all influent gases is removed, leaving trace amounts of inert gases, such as helium and radioactive noble gases to build up.
The primary source of the radioactive gas is via the purge of the volume control tank with hydrogen, as described in Section 9.3.4. The operation of the GRWS serves to reduce the fission gas concentration in the reactor coolant system which, in turn, reduces the escape of fission gases from the reactor coolant system during maintenance operations or through equipment leakage.
Smaller quantities are received, via the vent
11.3-2 Rev. 11 WOLF CREEK
connections, from the reactor coolant drain tank, the pressurizer relief tank, and the recycle holdup tanks.
Since hydrogen is continuously removed in the recombiner, this gas does not build up within the system. The largest contributor to the nonradioactive gas accumulation is helium generated by a B10 (n,a)Li7 reaction in the reactor core. The second largest contributors are impurities in the bulk hydrogen and oxygen supplies. Stable and long-lived isotopes of fission gases also contribute small quantities to the system gas volume accumulation.
Operation of the system is such that fission gases are distributed throughout the six normal operation gas decay tanks. Separation of the GRWS gaseous inventory in several tanks assures that the allowable site boundary dose will not be exceeded in the event of a gas decay tank rupture. Radiological consequences of such a postulated rupture are discussed in Section 15.7.1.
The GRWS also provides the capacity for indefinite holdup of gases generated during reactor shutdown. Nitrogen gas from previous shutdowns is contained in the shutdown gas decay tank for use in stripping hydrogen from the reactor coolant system. The shutdown tank is normally at low pressure and is used to accept relief valve discharges from the normal operation gas decay tanks.
For all buildings where there is potential airborne radioactivity, the ventilation systems are designed to control the release. Where applicable, each building has a vent collection system for tanks and other equipment which contain air or aerated liquids. The condenser evacuation system discharge is filtered and discharged to the unit vent in addition to the discharges from the reactor building, auxiliary building, and fuel building. The radwaste building, which houses the GRWS, has its own release vent. The turbine building has an open ventilation system, and the steam packing exhaust discharges outside the turbine building.
The vent collection systems receive the discharge of vents from tanks and other equipment in the radwaste and auxiliary buildings which contain air or aerated liquids. These components contain only a very small amount of fission product gases. Prior to release via the radwaste or auxiliary building ventilation system, the gases are monitored, as described in Section 11.5, and passed through a prefilter, HEPA filter, charcoal filter, and another
11.3-3 Rev. 14 WOLF CREEK
HEPA filter in series which reduce any airborne particulate radioactivity to negligible levels and provide a decontamination factor of at least 10 for radioactive iodines and 100 for particulates. Expected efficiencies for iodine removal are better than 99 percent for elemental iodine and 95 percent for organic iodine at 70-percent relative humidity. However, for gaseous effluent release calculations, 70-percent efficiency is conservatively used for radioiodine isotopes.
Although plant operating procedures, equipment inspection, and preventive maintenance are performed during plant operations to minimize equipment malfunction, overall radioactive release limits have been established as a basis for controlling plant discharges during operation with the occurrence of a combination of equipment faults of moderate frequency. These faults include operation with fuel defects in combination with steam generator tube leaks and malfunction of liquid or gaseous waste processing systems or excessive leakage in reactor coolant system equipment or auxiliary system equipment. Operational occurrences such as these can result in the discharge of radioactive gases from various plant systems. These unscheduled discharges may be from plant systems which are not normally considered gas processing systems or from a gas decay tank after a 90-day holdup period. These potential sources are tabulated in Table 11.1-2. The bases for assumed releases, the factors which tend to mitigate the release of radioactivity, and the release paths are given in Appendix 11.1A.
A further discussion of the gaseous releases from the plant is provided in Section 11.3.3.
11.3.2.2 Component Description
Codes and standards applicable to the GRWS are listed in Tables 3.2-1 and 11.3-
- 1. The GRWS is designed and constructed in accordance with quality group D (augmented). The GRWS is seismically designed to the requirements of Reg. Guide 1.143, as discussed in Table 3.2-5. The GRWS is housed within a building also seismically designed to the requirements of Reg. Guide 1.143. The GRWS design complies with Regulatory Guide 1.143, as specified in Table 3.2-5.
WASTE GAS COMPRESSOR - The waste gas compressor is a water-sealed centrifugal displacement unit which maintains continuous circulation of nitrogen around the waste gas loop. The compressor is provided with a mechanical shaft seal to minimize water leakage. The compressor moisture separator normal water level is maintained to keep the shaft immersed at all times.
11.3-4 Rev. 0 WOLF CREEK
Two waste gas compressor packages are provided. One compressor is normally used, and the other compressor is on standby. The packages are self-contained and skid-mounted. Construction is primarily of carbon steel.
CATALYTIC HYDROGEN RECOMBINER - The catalytic recombiner disposes of hydrogen brought into the GRWS. This is accomplished by adding a controlled amount of oxygen to the recombiner which reacts with the hydrogen as the gas flows through a catalyst bed. The control system for the recombiner is designed to preclude the possibility of a hydrogen explosion. This is further discussed in Section 11.3.6.
Two hydrogen recombiner packages are provided. One recombiner is normally used, and the other is on standby. The packages are self-contained and skid-mounted. The recombiner is located in the system where the hydrogen concentration and pressure are optimum with respect to hydrogen removal.
DECAY TANK - Eight gas decay tanks are provided, six for normal power operation and two for service at shutdown and startup. The tanks are of the vertical-cylindrical type and are constructed of carbon steel.
MISCELLANEOUS COMPONENTS - The gas decay drain collection tank provides a collection point for condensation drained from the gas decay tanks, recombiners, and gas compressors.
All control valves, with the exception of those on the recombiner, are provided with bellow seals to minimize the leakage of radioactive gases through the valve bonnet and stem. Valves on the recombiner package are provided with leakoffs. The leakoff port was removed and capped on the Feed Gas Pressure Control Valve for SHA01A A Hydrogen recombiner skid. This leak off line remains intact for the B Hydrogen Gas Recombiner skid.
Relief valves have soft seats and are exposed to pressures which are normally less than two-thirds of the relief valve set pressure. The relief valves of the major components discharge to the shutdown tanks. This permits decay and controlled disposal of all discharges less than about 3,000 scf. The relief valves are designed to relieve full flow from both waste gas compressors.
To maintain leakage from the system at the lowest practicable level, diaphragm-type manual valves are used throughout the waste gas system. For low temperature, low pressure service valves with a synthetic rubber-type diaphragm are used. This application includes all parts of the system, except the recombiners. Because of the high temperature that may exist in the recombiner, globe type valves with a metal diaphragm seal in the stem are used. There should be no measurable stem leakage from either type of valve.
11.3-5 Rev. 26 WOLF CREEK
The gas decay tank drain pump directs water from the gas decay drain collection tank (due to condensation or maintenance) to the waste holdup tank or recycle holdup tanks. It is used when there is insufficient pressure in the gas system to drive the fluid. All parts of the pump in contact with the drain water are of austenitic stainless steel. The pump is a canned-motor type.
The waste gas drain filter is a disposable cartridge filter provided to prevent particulate matter, including rust, from entering the LRWS and BRS. Parts of the filter in contact with the drain water are of austenitic stainless steel.
The waste gas traps are designed to prevent gases from leaving the GRWS. There are four gas traps - two in the gas decay tank drain line and one each in the recombiner drain lines and compressor drain lines.
The component description for the ventilation systems is provided in Section 9.4.
11.3.2.3 System Operation
Operation of the ventilation systems is described in Section 9.4. The following is a description of the GRWS.
NORMAL OPERATION - During normal power operation, nitrogen gas, with contained fission gases, is circulated around the GRWS loop by one of the two compressors. Fresh hydrogen gas is introduced into the volume control tank where it is mixed with fission gases stripped from the reactor coolant by the action of the volume control tank letdown line spray nozzle. The gas is vented from the volume control tank into the circulating nitrogen in the waste gas system, at the compressor suction. Normal operational mode of the system is dependent on the reactor coolant system (RCS) gas concentration and the RCS status. A purge of the Volume Control Tank is performed as directed by Chemistry. During a VCT purge using the same Gas Decay Tank is advantageous.
However, switching GDTs may be required, depending on the high operating pressure parameters of the system.
The resulting mixture of nitrogen, hydrogen, and fission gases is pumped by one of the compressors to one of the two catalytic hydrogen recombiners where enough oxygen is added to react with and reduce the hydrogen to a low residual level. Water vapor formed in the recombiner by the hydrogen and oxygen reaction is condensed and removed, and the cooled gas stream (now composed primarily of nitrogen, helium, and fission gases) is discharged from the recombiner, routed through a gas decay tank, and sent back to the compressor suction to complete the loop circuit.
Only one gas decay tank is valved into the waste gas loop at any time. By switching tanks when tank pressure nears the upper operating parameters, this will allow for more decay time for the gases stored in the tanks. This practice will result in fewer radioactive curies released.
11.3-6 Rev. 11 WOLF CREEK
If it has been determined that excessive nitrogen buildup is occurring within the system or when other occurrences require it, one tank can be valved out of service and allowed to decay for a period of 90 days, and then discharged.
STARTUP - At plant startup, the system is first flushed free of air and filled with nitrogen at atmospheric pressure. One compressor, one recombiner, and one shutdown decay tank are in service. The reactor is at the cold shutdown condition. Fresh hydrogen is charged into the volume control tank, and the volume control tank vent gas mixes with the circulating nitrogen in the GRWS.
This circulating mixture enters the compressor suction, passes through the recombiner and shutdown gas decay tank, and returns to the compressor suction.
When the reactor coolant system hydrogen concentration is within operating specifications, the shutdown gas decay tank is isolated and the gas flow directed to one of the gas decay tanks provided for normal power operation.
Gases accumulated in the shutdown tank will be retained for reuse during hydrogen stripping from the reactor coolant system during subsequent shutdown operations.
SHUTDOWN AND DEGASSING OF THE REACTOR COOLANT SYSTEM - Plant shutdown operations are essentially startup operations in reverse sequence. The volume control tank hydrogen purge is maintained until after the reactor is shut down and coolant fission gas concentrations have been reduced to specified level.
During this operation, hydrogen purge flow may be increased to speed up coolant degassing. The gas decay tank in service for normal power operation is valved out, and a nitrogen purge from the shutdown tank to the volume control tank is begun. The shutdown tank is placed in the process loop at the compressor discharge so that the gas mixture from the volume control tank vents to the compressor suction and passes through the shutdown tank and to the recombiner where hydrogen is removed and returned to the compressor suction. The nitrogen purge continues until the reactor coolant hydrogen concentration reaches the required level. Degassing is then complete, and the reactor coolant system may be opened for maintenance or refueling.
11.3.3 RADIOACTIVE RELEASES
This section describes the estimated gaseous release from the plant for normal operation and anticipated operational occurrences.
11.3.3.1 Sources
Section 11.1 and Appendix 11.1A provide the bases for determining the contained source inventory and the normal releases.
11.3-7 Rev. 0 WOLF CREEK
11.3.3.2 Release Points
Potential release paths for gaseous radioactivity are illustrated schematically in Appendix 11.1A. The general location of potential gaseous radioactivity release points is depicted in Figure 1.2-1. A description of potential release points for radioactive gaseous effluents is given in Appendix 11.1A, along with the physical characteristics of the gaseous effluent streams. Release points from the gaseous waste processing systems are shown on Figure 11.1A-3.
11.3.3.3 Dilution Factors
The annual average dilution factors used in evaluating the release of gaseous radioactive effluents are derived and justified in Section 2.3.
11.3.3.4 Estimated Doses
The GASPAR computer code, which calculates doses due to normal gaseous effluents in accordance with Regulatory Guide 1.109, was used to determine the doses listed herein. This code was validated and verification is maintained on file.
The doses due to normal gaseous effluents from WCGS are listed in Tables 11.3-2, 3 and 4. Doses attributable to radioactive iodines and particulates at the controlling sector Exclusion-Restricted Area boundary are contained within Table 11.3-3 (Hypothetical Worst Case). Doses from iodines and particulates at the controlling residence are contained within Table 11.3-4 (Controlling Existing Resident). Table 11.3-2 contains doses from noble gases at the Exclusion-Restricted Area boundary.
The doses in these tables were calculated assuming intermittent purge operation. Intermittent purge mode release rates were taken from Section 11.1.
The values of the dispersion and deposition coefficients, X/Q (non-decayed),
X/Q (depleted and non-decayed) and D/Q used in the calculations were taken from Section 2.3 and Table 2.3-75. A comparison of the half lives of the radionuclides released to the time needed for released nuclides to disperse to any point within the 5-mile radius of interest shows that the effect of decay during this dispersion period is negligible. Thus, the values for X/Q (decayed) and X/Q (decayed and depleted) were taken to be equivalent to the corresponding X/Q (non-decayed) and X/Q (depleted and non-decayed) values.
11.3-8 Rev. 31 WOLF CREEK
A survey of the area within a five-mile radius of the site was conducted during June 1980 and was used to determine the pathways present at the controlling locations. A 1986 survey of the same area indicates the pathways present at the controlling locations are still the same. X/Qs for the controlling locations were used in calculating doses from iodines and particulates as well as noble gases.
The total doses for Table 11.3-3 and 11.3-4 were calculated by summing the doses from each pathway present. It was conservatively assumed that all age groups were present at each controlling location.
Doses due to noble gases and radioactive iodines and particulates in no case exceed 10 CFR 50 Appendix I limits.
Actual doses from gaseous effluent during plant operation will be calculated using the approved methodology presented in the Offsite Dose Calculation Manual.
11.3.4 SAFETY EVALUATION
The GRWS serves no safety-related function.
11.3.5 TESTS AND INSPECTIONS
Preoperational testing is described in Chapter 14.0.
The operability, performance, and structural and leaktight integrity of all system components are demonstrated by continuous operation.
11.3.6 INSTRUMENTATION APPLICATION
The GRWS instrumentation, as described in Table 11.3-5, is designed to facilitate automatic operation and remote control of the system and to provide continuous indication of system parameters.
The instrumentation readout is located mainly on the waste processing system panel in the radwaste building. Some instruments are read where the equipment is located. Alarms are shown separately on the waste processing system panel and further relayed to one common waste processing system annunciator on the main control board of the plant. Where suitable, instrument lines are provided with diaphragm seals to prevent fission gas outleakage through the instrument.
Figure 11.3-3 shows the location of the instruments on the compressor package.
11.3-9 Rev. 5 WOLF CREEK
The compressors are interlocked with the seal water inventory in the moisture separators and trip off on either high or low moisture separator level. During normal operation, the proper seal water inventory is maintained automatically.
Figure 11.3-4 indicates the location of the instruments on the recombiner installation.
The catalytic recombiner system is designed for automatic operation with a minimum of operation attention. Each package includes two online gas analyzers, one to measure hydrogen and oxygen in and one to measure hydrogen and oxygen out. The analyzers are the primary means of recombiner control.
Each of these online gas analyzers is independently controlled. In the event that these analyzers are declared inoperable, operation of the system may continue provided grab samples are taken and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both oxygen channels or both the inlet oxygen and inlet hydrogen channels inoperable, oxygen supply is suspended to the recombiner. Addition of waste gas to the system may continue provided grab samples are taken and analyzed at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations.
The GRWS is designed to operate with hydrogen concentrations above 4 percent by volume. Flammable mixtures of gases in the system are prevented by monitoring and controlling the oxygen concentration to appropriate levels. The setpoints for oxygen concentration in the catalyst bed inlet stream are 3 percent for the hi-alarm and 3.5 percent for the hi-hi alarm and isolation of the oxygen supply. The setpoint for oxygen concentration downstream of the catalyst bed is 60 ppm oxygen for the hi-hi alarm and isolation of inlet oxygen supply.
Thus the oxygen supply to the recombiner would be terminated before the concentration in the GRWS would reach levels favorable for hydrogen flammability.
Since the GRWS is designed to operate with hydrogen concentrations up to 6 percent by volume, up to 3 percent oxygen is necessary for operation of the catalytic recombiner. Termination of oxygen feed at 2 percent as suggested by regulatory guidance is inappropriate. Further, since the minimum oxygen concentration necessary to support combustion at 4 percent by volume hydrogen concentrations is 5 percent, the hi-alarm setpoint of 3 percent provides sufficient margin (i.e., 60 percent of the limit) to flammability.
A multipoint temperature recorder monitors temperatures at several locations in the recombiner packages.
The process gas flow rate is measured by an orifice located upstream of the recombiner preheater. Local pressure gauges indicate pressure at the recombiner inlet and oxygen supply pressure.
The following controls and alarms are incorporated to maintain the gas composition outside the range of flammable and explosive mixtures:
11.3-10 Rev. 21 WOLF CREEK
- a. If the recombiner feed concentration exceeds 6 percent by volume, a high-hydrogen alarm sounds to warn that all hydrogen entering the recombiner is not reacted. This alarm is followed by a second alarm indicating high hydrogen in the recombiner discharge. These alarms warn of a possible hydrogen accumulation in the system.
- b. If the hydrogen concentration in the recombiner feed reaches 9 percent by volume, a high-high hydrogen alarm sounds, the oxygen feed is terminated, and the volume control tank hydrogen purge flow is terminated. These controls limit the possible accumulation of hydrogen in the GRWS to 3 percent by volume.
- c. If the oxygen concentration in the recombiner feed reaches 3 percent by volume, an alarm sounds and oxygen feed flow is limited so that no further increase in flow is possible. This control maintains the system oxygen concentration at 3 percent or less, which is below the flammable limit for hydrogen-oxygen mixtures.
- d. If the oxygen concentration in the recombiner feed reaches 3.5 percent by volume, an alarm sounds and the oxygen feed flow is terminated.
- e. If hydrogen in the recombiner discharge exceeds 0.25 percent by volume, an alarm sounds. This alarm warns of high hydrogen feed, possible catalyst failure, or loss of oxygen feed.
- f. If oxygen in the recombiner discharge exceeds 60 ppm, an alarm sounds and oxygen feed is terminated. This control prevents any accumulation of oxygen in the system in case of hydrogen recombiner malfunction.
- g. On low flow through the recombiner, oxygen feed is terminated. This control prevents an accumulation of oxygen following system malfunction.
- h. High discharge temperature from the cooler-condenser (downstream from the reactor) terminates oxygen feed.
This protects against loss of cooling water flow in the cooler-condenser.
11.3-11 Rev. 10 WOLF CREEK
- i. High temperature indication by any one of six thermocouples in the catalyst bed limits oxygen feed so that no further increase is possible.
- j. High temperature indication at the recombiner reactor discharge terminates oxygen feed to the recombiner.
11.
3.7 REFERENCES
Published References
- 1. Eckerman, K.F. and Lash, D G, 1978, GASPAR version marked "revised 8/19/77": U S Nuclear Regulatory Commission, Radiological Assessment Branch.
- 2. Eckerman, K.F., Congel, F.J., Roecklein, A.K. and Pasciak, W.J., 1980, NUREG-0597 Users Guide to GASPAR Code: U.S.
Nuclear Regulatory Commission, Radiological Assessment Branch.
Personal References
1 Warminski, N C, 1979, Horticulture agent for the Sedgwick County Extension Office of the Kansas State University Cooperative Extension Service, Wichita, Kansas, telephone conversation (25, 26 January), written communication (29 January).
11.3-12 Rev. 10 WOLF CREEK
TABLE 11.3-1
GASEOUS WASTE PROCESSING SYSTEM MAJOR COMPONENT DESCRIPTION
Water Gas Compressors
Type Centrifugal Quantity 2 Design pressure, psig 150 Design temperature, F 180 Operating temperature, F 70 to 130 Design suction pressure, N2 at 130 F, psig 0.5 Design discharge pressure, psig 110 Design flow, N2 at 130 F, scfm 40 Material Carbon steel Design code (1) ASME VIII/D (augmented)
Seismic design In accordance with Table 3.2-1
Gas Decay Tanks
Type Vertical Quantity 8 Design pressure, psig 150 Design temperature, F 180 Volume, each, ft3 600 Material of construction Carbon steel Design code (1) ASME VIII/D (augmented)
Seismic design In accordance with Table 3.2-1
Recombiners
Type Catalytic Quantity 2 Design pressure, psig 150 Design temperature, F (2)
Design flow rate, scfm 50 Operating discharge pressure, psig 20 Operating discharge temperature, F 70 to 140 Material of construction Stainless steel Design code (1) ASME VIII/D (augmented)
Seismic design In accordance with Table 3.2-1
(1) Table indicates the required code based on its safety-related importance as dictated by service and functional requirements and by the consequences of their failure.
Note that the equipment may be supplied to a higher principal construction code than required.
(2) Varies by component in the recombiner package, but exceeds operating temperatures by 100 F.
Rev. 0 WOLF CREEK
TABLE 11.3-2
Deleted Table
Rev. 14 WOLF CREEK
TABLE 11.3-3
Deleted Table
Rev. 14
c* c* (
~F CREEK
TO RECOMBIHER
COMPRESSOR tROM VOLUME CONTROL TANK
MOISTURE SEPARATOR
T - TEMPERATURE MEASUREMENT SEAL WATER P
- PRESSURE MEASUREMEHT COOlER RETURN L - LEVEl MEASUREMENT
OPDA'l"BD SAF~TY WOLF CREEK ANAL'fSlS REPORT
FIGURE U. 3-3 Rev. 0 COMPRESSOR PACKAG~ INSTRUMENTS WOlF CREEK OXYGEN
HEATER CATALYTIC FROM GAS REACTOR
COMPRESSOR
TO GAS ANALYZER
PHASE SEPARATOR COOLER/CONDENSOR
TO GAS DECAY TAHK TO GAS T - TEMPERATURE MEASUREMENT ANALYZER P - PRESSURE MEASUREMENT F - FLOW MEASUREMENT
WOLF CRBBB:
UPDA~ED SAFE~Y ANALYSIS REPORT FIGURE 11. 3-L.f HYDROGEN RECOHBINER INSTRUMENTS
Rev.21 WOLF CREEK
11.4 SOLID WASTE MANAGEMENT SYSTEM
The solid radwaste system (SRS) is designed to meet the functional requirements of the solid waste management system. The SRS is designed to collect, process, and package low-level radioactive wastes (LLW) generated as a result of normal plant operation, including anticipated operational occurrences, and to store this packaged waste until it is shipped offsite to a waste processor for treatment and/or disposal or to a licensed burial site. The process and effluent radiological and sampling systems are described in Section 11.5.
11.4.1 DESIGN BASES
11.4.1.1 Safety Design Bases
The SRS performs no function related to the safe shutdown of the plant, and its failure does not adversely affect any safety-related system or component; therefore, the SRS has no safety design bases.
11.4.1.2 Power Design Bases
POWER GENERATION DESIGN BASIS ONE - The SRS is designed to meet the following objectives:
- a. Provide remote transfer and hold-up capability for spent radioactive resins from the chemical and volume control system, fuel pool cooling and cleanup system, boron recycle system, liquid radwaste system, steam generator blowdown system, and secondary liquid waste system and for spent radioactive activated charcoal from the liquid radwaste system and the secondary liquid waste system.
- b. Provide a means to semiremotely remove and transfer the spent filter cartridges from the filter vessels to the solid radwaste processing system in a manner which minimizes radiation exposure to operating personnel and the spread of contamination.
- c. Provide a means for compacting and packaging miscellaneous dry radioactive materials, such as paper, rags, and contaminated clothing.
- d. Provide a means for dewatering primary and secondary resin storage and shipment offsite.
11.4-1 Rev. 18 WOLF CREEK
POWER GENERATION DESIGN BASIS TWO - The SRS is designed and constructed in accordance with Regulatory Guide 1.143, as described in Table 3.2-5, and Branch Technical Position ETSB 11-3, as described in Table 11.4-1. The seismic design classification of the radwaste building, which houses the solid waste management system, and the seismic design and quality group classification for the system components and piping are provided in Section 3.2.
POWER GENERATION DESIGN BASIS THREE - The SRS design parameters are based on the radionuclide concentrations and volumes consistent with reactor operating experience for similar designs and with the source terms of Section 11.1.
POWER GENERATION DESIGN BASIS FOUR - Collection, packaging, and storage of radioactive wastes are to be performed so as to maintain any potential radiation exposure to plant personnel during system operation or during maintenance to "as low as is reasonably achievable" (ALARA) levels, in accordance with the intent of Regulatory Guide 8.8 in order to maintain personnel exposures well below 10 CFR 20 requirements. Design features incorporated to maintain ALARA criteria include remote system operation, remotely actuated flushing, and equipment layout permitting the shielding of components containing radioactive materials. Additionally, access to the solid waste processing and storage areas is controlled to minimize personnel exposure.
POWER GENERATION DESIGN BASIS FIVE - The onsite storage facilities for solid wastes have a capacity for temporary storage of solid wastes resulting from approximately 5 years of plant operation. Temporary onsite storage and shipping offsite of solid radwaste do not present a radiation hazard to persons onsite or offsite, for either normal conditions or extreme environmental conditions, such as tornados, floods, or seismic events. Greater detail on interim on-site storage is provided in section 11.4.A.
POWER GENERATION DESIGN BASIS SIX - The SRS is designed to meet the requirements of General Design Criterion 60 of 10 CFR 50, Appendix A.
Packaging and shipment of radioactive wastes is performed in accordance with the requirements of 10 CFR 61, 10 CFR 71, 49 CFR 173, and applicable state regulations.
POWER GENERATION DESIGN BASIS SEVEN - Temporary storage, on a concrete slab or within a building addition located West of the IOS facility and South of the Radwaste Building provides temporary indoor/outdoor storage of large waste material which becomes activated during reactor operation. Each stored item will be unique, therefore procedures for storing items outdoors will be determined on a case by case basis.
11.4-2 Rev. 13 WOLF CREEK
11.4.2 SYSTEM DESCRIPTION
11.4.2.1 General Description
The SRS consists of the following subsystems which are illustrated in the piping and instrumentation diagrams provided in Figure 11.4-1:
- a. Dry waste system
- b. Resin handling system
- c. Filter handling system
- d. Waste disposal system
The activity of the influents to the SRS is dependent on the activities of the various fluid systems, such as the boron recycle system, secondary liquid waste system, liquid waste management system, chemical and volume control system, fuel pool cooling and cleanup system, floor and equipment drain system, and the steam generator blowdown system. Reactor coolant system activities and the decontamination factors for the systems given above also determine theinfluent activities to the solid radwaste system.
Table 11.4-2 lists the estimated expected and maximum activities of waste to be processed on an annual basis and their physical form and source. The isotopic makeup and curie content of the expected influents to the SRS are given in Table 11.4-2. The estimated maximum annual quantities of solid radwaste generation are presented in Table 11.4-3. The estimated annual expected and maximum curie and isotopic content is presented in Table 11.4-4, for each waste category. Packaged waste volumes are based on the following:
- a. Waste content volume in Table 11.4-3, when based on packaging in 55-gallon and solidified with concrete, are:
(1) 3.5 ft3 primary spent resin, primary charcoal, and primary evaporator bottoms per drum
(2) 4.8 ft3 secondary spent resin and charcoal per drum
(3) 5.3 ft3 secondary evaporator bottoms
(4) 4.0 ft3 chemical waste per drum
(5) 1 filter cartridge per drum
(6) 7.5 ft3 shipped volume per drum (including cement
- b. Disposal volumes are based on packaging in the following typical containers:
11.4-3 Rev. 8 WOLF CREEK
Waste Stream Container Type Container Volume Primary Resin PL8-120 120.3 cuft Secondary Resin PL14-215 205.8 cuft Filters PL6-80 83.4 cuft DAW 85 Gallon Drum 11.6 cuft 79 Gallon Drum 10.8 cuft 55 Gallon Drum 7.5 cuft B-25 Box 96 cuft
Section 11.1 and Appendix 11.1A provided the bases for determination of liquid source terms which are used to calculate the solid waste source terms The sources presented in Tables 11.4-2 and 11.4-4 are conservatively based on Section 11.1, Appendix 11.1A and the following additional information:
- a. As a basis for the activities given in Table 11.4-4, 30 days decay is assumed.
- b. The miscellaneous dry and compacted waste volume will reflect the historical increases since the issuance of Case 6 in Table 2-49 of WASH-1258, July 1973.
11.4.2.2 Component Description
Codes and standards applicable to the SRS are listed in Tables 3.2-1 and 11.4-
- 5. The SRS is housed within a seismically designed building. Regulatory Guide 1.143 is complied with to the extent specified in Table 3.2-5.
SRS component parameters are presented in Table 11.4-5. The following is a functional description of the major system components:
SPENT RESIN STORAGE TANK (PRIMARY) - Provides for storage and decay of the spent resins from the demineralizers in the chemical and volume control system, fuel pool cooling and cleanup system, boron recycle system, and liquid radwaste system.
SPENT RESIN STORAGE TANK (SECONDARY) - Provides for storage and decay of the spent resins and spent activated charcoal from the demineralizers and charcoal adsorbers in the steam generator blowdown system, secondary liquid waste system, and charcoal adsorbers in the liquid radwaste system.
EVAPORATOR BOTTOMS TANK (PRIMARY) - Provides for storage, decay, sampling, and chemistry control of the concentrated wastes from the liquid radwaste system.
EVAPORATOR BOTTOMS TANK (SECONDARY) - Provides for storage, decay, sampling, and chemistry control of the concentrated wastes from the secondary liquid waste system.
11.4-4 Rev. 14 WOLF CREEK
SPENT RESIN SLUICE PUMPS (PRIMARY AND SECONDARY) - Provides the motive flow to transfer spent resin or spent activated charcoal from the various demineralizers or adsorbers to the appropriate spent resin storage tank.
EVAPORATOR BOTTOMS TANK PUMPS (PRIMARY AND SECONDARY) - Are available to transfer the concentrated liquid wastes from the evaporator bottoms tanks to the solid radwaste disposal station.
ACID ADDITION TANK AND METERING PUMP - Provides chemistry control to the chemical drain tank, and floor drain tank.
CAUSTIC ADDITION TANK AND METERING PUMP - Provides chemistry control to the chemical drain tank, floor drain tank, waste holdup tank, evaporator bottoms tank (primary), and evaporator bottoms tank (secondary).
RESIN CHARGING TANKS - Provides remote means of gravity sluicing clean resin and activated charcoal into the demineralizer and adsorber units.
WASTE DISPOSAL STATION - The waste disposal station provides the capability to transfer primary/secondary spent resins and evaporator bottoms, and liquid radwaste demineralizer skid spent resins, to a HIC for storage/shipping. A return header provides a path for decanted water to be returned to the liquid radwaste system or the Secondary Spent Resin Storage Tank or the Primary Spent Resin Storage Tank. The waste disposal station also provides necessary interface support requirements for mobile vendor processing systems.
RADWASTE BRIDGE CRANE - A crane, remotely operated from the solid radwaste control console, which provides the means of moving containers to the processing area, from the processing area to the solid waste storage area, and from the solid waste storage area to the shipping area. The crane is equipped with a television camera system to facilitate the remote handling operation.
DRY WASTE COMPACTORS - Hydraulic power mechanical ram devices that are used to reduce the volume of compressible dry wastes by a factor of approximately five.
They are designed with exhaust fan and filter to control the airborne dust during dry waste compaction operations.
11.4.2.3 System Operation
11.4.2.3.1 Waste Disposal System
The waste disposal station provides the capability to transfer primary/secondary spent resins and evaporator bottoms, and liquid radwaste demineralizer skid spent resins, to a HIC for storage/shipping. A return header provides a path for decanted water to be returned to the liquid radwaste system or the Secondary Spent Resin Storage Tank or the Primary Spent Resin Storage Tank. The waste disposal station also provides necessary interface support requirements for mobile vendor processing systems.
11.4-5 Rev. 14 WOLF CREEK
Evaporator concentrates are stored in either the evaporator bottoms tank (primary) or the evaporator bottoms tank (secondary). Each tank is provided with a mixer, and the piping system contains a relatively high flow pump for recirculation of the tank's contents to maintain the concentrates in the homogeneous state. Each tank is supplied with external strip heaters, and all piping that can contain the concentrated waste is heat traced to preclude crystallization and eventual plugging within the piping system.
Spent resins are stored in either the primary or secondary resin storage tank.
Each tank is supplied with nitrogen gas for sluicing the spent resin to the waste disposal station. Spent resin from the liquid radwaste demineralizer skid is also sluiced to the waste disposal station using Reactor make-up water or the associated system pump. Spent resins are normally sluiced into a High Integrity Container (HIC) for disposal. Resins are dewatered in accordance with the Process Control Program using approved procedures.
The waste disposal station area consists of a segmented concrete shield with nine inch walls, capable of containing the largest anticipated HIC, 60 inch diameter and 73 inch height, with 630 curies of activitie without disturbing normal operations.
The waste disposal station utilizes the necessary system controls to prevent improper system operation to preclude the spillage of waste. Because of these system design features, waste spillage is not anticipated although provisions are made for processing waste spillage. A drain system is provided in the waste disposal station for handling waste spillage. Provisions are also contained in the drain system to feed waste to a mobile vendor solidification system/mobile vendor resin dewatering system.
11.4.2.3.2 Dry Waste System
Low-level dry wastes are collected in drums at appropriate locations throughout the plant, as dictated by the volume of these wastes generated during operation or maintenance. Dry wastes, which can be compressed by a factor of five to minimize the volume, may be compacted in 55-gallon drums with a dry waste compactor. Compactors are located in the radwaste building and the auxiliary building. The dry waste compactors have an integral shroud which directs any airborne dusts created by the compaction operation through an exhaust fan and filter, and then to the respective building's ventilation system.
The filled drums are sealed and moved to the storage area in the radwaste building, or other designated areas, where they are stored until shipment offsite.
11.4-6 Rev. 11 WOLF CREEK
Dry wastes can also be processed/compacted offsite by contractor as part of the shipment and waste disposal contract. The low level dry waste collected can be placed in a NRC/DOT approved waste container (e.g., sea van) which is shipped offsite when filled. The container is placed outside the radwaste building within the radiological controlled area.
Large components and equipment which have been activated during reactor operation and which are not amenable to solidification or compaction are handled either by qualified plant personnel or by outside contractors specializing in radioactive materials handling, and are packaged in shipping casks or appropriate shipping packages of an appropriate size.
Dry noncompressible radwaste (such as hoses, buckets, etc.) will be packaged in approved containers and shipped as Low Specific Activity (LSA) or Type A waste.
11.4.2.3.3 Resin Handling System
The resin handling system provides the capability for remote removal of spent radioactive resin and activated charcoal from the demineralizer and charcoal adsorber vessels in the chemical and volume control system, fuel pool cooling and cleanup system, boron recycle system, liquid radwaste system, steam generator blowdown system, and secondary liquid waste system and to transfer them to the associated spent resin storage tank.
In the resin transfer mode, the spent resin sluice pumps take suction from the storage tank via a screened connection on the tank and pump water through the respective vessel to first backflush the resin and then sluice the resin to the spent resin storage tank. Positive indication that the resin has been sluiced to the spent resin storage tank is provided by an ultrasonic density element located in the spent resin sluice header. Alternate Sluice water may be provided by the Reactor Makeup Water system, if the sluice pumps are inoperable.
The spent resin storage tank (primary), which accepts resins from the reactor purification systems, is capable of accommodating at least 60 days' waste generation at normal generation rates. The spent resin storage tank (secondary), which accepts spent resin and spent activated charcoal from the remaining vessels, is capable of accommodating at least 30-days' waste generation at normal generation rates.
Spent resin and spent activated charcoal are transferred from the spent resin storage tanks to the waste disposal station by pressurizing the storage tank with nitrogen and supplying sluice water at the outlet nozzle on the tank.
Positive indication that resin has been transferred is provided by a local camera, monitoring at the container entry at the solid radwaste disposal station. Upon completion of the resin transfer, the tank is vented to the radwaste building ventilation system.
The empty demineralizer or charcoal adsorber vessels are filled with clean resin or activated charcoal by gravity sluicing from the resin charging tank into the associated vessels. The filling operations are performed remotely from the vessels being filled.
11.4-7 Rev. 10 WOLF CREEK
11.4.2.3.4 Filter Handling System
The filter handling system is a semiremote system which provides the capability to remove spent radioactive cartridge filters from their filter housings and to transport them to the solid radwaste processing area in the radwaste building.
The system, requires the operator to be in the proximity of the filters; however, they are protected by distance which minimizes operator exposure.
The filter handling system consists of long handled tools for removal of the filter housing top and assemblies. As necessary, shielded transport casks are used for transport and storage of the filter assembly.
The steps required by the operator for the removal of the filters are as follows:
- a. Using a monorail hoist, the shield plug above the filter housing is removed and set aside. Any time the plug hole is uncovered, the operators must take care to stay well away from the proximity of the hole, to avoid exposure. This necessitates that the monorail hoist be operated with a remote pendant controller.
- b. Using long-handled tools the operator loosens the housing head bolts and flips them back out of the way.
- c. With another tool, he engages the housing head and flips it back out of the way.
- d. The filter is lifted part way out of the housing and allowed to drip until it has decayed to an acceptable level. It is placed into a shielded cask or shielded storage location.
- e. A new cartridge is installed in the filter housing, either by reversing the previous sequence or, if filter housing radiation levels permit, by manually loading and securing the head.
11.4.2.3.5 Mixed Waste Handling System
Mixed waste (MW) is defined as radioactive waste that has hazardous characteristics or components as defined by 40 CFR 260/261. MW (liquid and solid) is collected in the plant and placed in the appropriate containers.
The MW will be processed (if required) and shipped for disposal. Radioactive content of the MWSF will be limited to prevent exceeding the limits in 10 CFR 20 and 10 CFR 50 Appendix I during normal operation, including anticipated operational occurrences.
11.4-8 Rev. 32 WOLF CREEK
11.4.2.4 Packaging, Storage, and Shipment
Solidified radwaste, or waste meeting the no free standing water criteria of Branch Technical Position ETSB 11-3 (i.e., dewatered), shall be stored in the Waste Bale Drumming Area. These wastes satisfy all applicable transportation and disposal requirements.
Wet radioactive waste, defined as any waste which does not meet receiving burial site free liquid requirements may be temporarily stored in the Waste Bale Drumming Area. Wet waste storage containers are designed to withstand the corrosive nature of the wet waste for the expected duration of the storage.
Temporarily stored wet waste will be processed (i.e., dewatered) or shipped to a waste processor for treatment prior to disposal.
DRY ACTIVE WASTE (DAW) - includes contaminated trash (paper, cloth, plastic, etc.)
SOLIDIFIED/DEWATERED WASTES - includes resin, filter cartridges and filter sludges transferred into HICs, and dewatered to less than 1% free standing water.
UNCOMPACTIBLE CONTAMINATED WASTE - other wastes not suitable for packaging in drums or HICs may be packaged in LSA boxes (B-25 or equivalent) or packaged into modular storage containers and stored on the temporary outdoor storage slab.
Spent resins, evaporator bottoms, spent charcoal, spent filter cartridges, and solid compactable waste such as contaminated paper, rags, and clothing are packaged in approved containers in accordance with 10CFR61 and shipped in accordance with applicable NRC (10CFR71) and DOT (49CFR173) regulations.
The 55-gallon drums used in the solid radwaste system meet the requirements of DOT approved containers.
Packaged solid radwaste is stored in the Waste Bale Drumming Area of the existing radwaste building prior to shipment offsite. The NRC/DOT approved waste container (e.g., sea van) is placed outside the radwaste building within the radiological controlled area prior to shipment offsite for processing.
The radwaste building storage areas have the ability to store 1,450 fifty-five gallon drums. However, other container sizes and storage configuration may be used.
Containers with radwaste are inventoried and their location recorded prior to being placed in storage.
Primary radwaste normally consists of:
- Spent resins, primary
- Filter cartridges, primary
Secondary waste normally consists of:
- Spent resins, secondary
- Filter cartridges, secondary
- Dry and compacted wastes
- Chemical wastes
11.4-9 Rev. 18 WOLF CREEK
Of the secondary waste, it is possible that most or all of it will be surveyed and released, rather than stored as radioactive waste.
Refer to Table 11.4-3 for Estimated Maximum Annual Quantities of Solid Radwaste.
11.4.3 SAFETY EVALUATION
Packaged solid radwastes containing, or potentially containing, significant quantities of radioactivity (i.e., spent resins, evaporator bottoms, are in a form that is highly resistant to release and spread of radioactivity during an extreme environmental event, such as a tornado or earthquake. This configuration provides, in effect, a double barrier against the release of radioactivity.
The containers that require radiation shielding are stored in the waste bale drum area which is resistant to tornados as described in Section 11.4-A. The containers with significant quantities of radioactivity remain in place during any extreme environmental event. The drums or other approved containers for the storage of dry active waste (DAW) have a low specific activity. See Section 11.4A for further details.
The packaged radwaste storage areas protect the containers from rainfall and corrosion. As described in Chapter 2.0, flooding is not a potential concern in grade-level buildings at the Wolf Creek site.
Although compacted and solidified wastes are expected to be stored onsite for some period of time prior to shipment, normally no credit other than 30-day decay is taken for radioactive decay realized by such storage when filling containers for shipping in accordance with 49 CFR 173 dose limitations. That is, once filled, containers can normally be shipped immediately, with the proper shielding, without exceeding Department of Transportation radiation limits. If 49 CFR 173 dose limitations cannot be met with the available shielding, however, the applicable containers are stored in the shielded storage area until the doses are acceptable for shipping in accordance with Department of Transportation requirements.
The normal onsite residence time for low level solid radwaste prior to shipping, such as dry compacted waste, steam generator blowdown spent resins, evaporator bottoms, spent charcoal, and ranges from several days to a few months. The normal onsite residence time for primary solid radwaste prior to shipping, such as primary spent resins and spent filter cartridges from the primary system, ranges from a few months to a few years. Onsite residence time is based on the initial activity of the container, the time required to have sufficient containers to completely load a transporting vehicle, the thickness of the shields available, the number of containers which can be stored in the available shipping casks, the availability of a transporting vehicle, and the availability of ultimate disposal facilities.
11.4-10 Rev. 30 WOLF CREEK
Solid radwaste is shipped from the site in Department of Transportation-approved containers by Department of Transportation-approved carriers.
Containers with any significant surface dose rate are moved remotely from the shielded storage areas to the transporting vehicle.
Radiation measurements made at the time of shipment of any radioactive waste material ensure that all shipments leave the site well within prescribed limits. Similarly, external contamination measurements are made to detect any potential release of radioactive material from the container prior to shipment.
Mixed waste will be stored in liquid and solid form in the MWSF. The total Curie content of the MWSF will be restricted accordingly to maintain doses to the maximally exposed individual during an extreme environmental event (e.g.
fire, tornado, etc.) below the applicable limits in 10 CFR 20 and 10 CFR 50.67.
11.4.4 TESTS AND INSPECTIONS
The SRS is in intermittent use throughout normal reactor operation. Periodic visual inspection and preventive maintenance are conducted using normal industry practice. Refer to Chapter 14.0 for information on preoperational and startup testing.
11.4.5 INSTRUMENTATION APPLICATION
Two control panels are provided for the equipment in the SRS which contains or processes potentially radioactive fluids or slurries. One control panel is located in the radwaste building control room and contains the instrumentation for the equipment which interfaces the influent systems (i.e., evaporator bottoms tank - primary, evaporator bottoms tank - secondary, spent resin storage tank - primary, and spent resin storage tank - secondary) and for the equipment used for process control (i.e., acid addition tank, acid addition metering pump, caustic addition tank, and caustic addition metering pump).
The second control panel (radwaste crane control panel) is located in a separate room in close proximity to the solid radwaste processing area. The control panel contains all instrumentation, including television monitors, required for remote operations. Pertinent instruments and controls for the transferring of the wastes from the tanks containing the wastes are duplicated on this panel so that the solid radwaste system operator can transfer the waste from these tanks to the waste disposal station.
11.4-11 Rev. 34
WOLF CREEK
TABLE 11.4-2
ESTIMATED EXPECTED AND MAXIMUM ANNUAL ACTIVITIES OF THE INFLUENTS TO THE SOLID RADWASTE SOLIDIFICATION SYSTEM, CURIES (Note 1}
(This Table is considered historical)
Spent Resins and And Evaporator Dry and Filter Cartridges Bottoms Chemical Charcoal Compacted Isotope (Primary} Tsecondary} -- Filters
Cr-51 3.0E+1 2.0E-2 9.8E-1 3.3E-4 2.3E-4 NEG Mn-54 2.9E+1 6. OE-3 4.5E-1 3.2E-4 1. 4E-4 NEG Fe-55 1.9E+2 2.5E-2 2.6E+O 1.4E-3 B.SE-4 NEG Fe-59 2.5E+1 1.5E-2 7.4E-1 3.7E-4 1.9E-4 NEG co-58 6.1E+2 2.2E-1 1.5E+1 7.1E-3 4.3E-3 NEG Co-60 2.6E+2 2.8E-2 3.2E+O 1.7E-3 1.1E-3 NEG Br-83 (1} NEG 1.7E-4 NEG 1.1E-5 NEG NEG Br-84 (1} NEG l.OE-5 NEG NEG NEG NEG Rb-86 (1} 7.9E-1 8.2E-4 3.2E-2 NEG NEG NEG RB-88 (1} 1.4E+O 3.0E-4 NEG NEG 1.1E-5 NEG Sr-89 (1} 9.8E+O 5.1E-3 2.8E-1 1.4E-4 7.3E-5 NEG Sr-90 (1} 1.4E+O 1.2E-4 1.7E-2 NEG NEG NEG Sr-91 (1} NEG 1.5E-4 3.6E-3 NEG NEG NEG Y-90 (1} 1.3E+O 1.1E-4 1.6E-2 NEG NEG NEG Y-91m (1} NEG 9.9E-5 2.4E-3 NEG NEG NEG Y-91 (1} 2.2E+O 8.3E-4 5.9E-2 2.4E-5 1.6E-5 NEG Zr-95 (1} 2.1E+O 1.1E-3 2.6E-2 3.3E-5 1.SE-5 NEG Nb-95 (1} 3.0E+O 1.2E-3 S.OE-2 5.7E-5 2.0E-5 NEG Nb-95m (1} 2.1E+O 9.0E-4 2.6E-2 3.4E-5 1.SE-5 NEG Mo-99 (1} 1.4E+2 1. 7E-1 4.4E+O 1.1E-3 1. OE-3 NEG Ru-103 (1} 1.0E+O 4.9E-4 1.3E-2 1.1E-5 NEG NEG Ru-106 (1} l.OE+O 1. 2E-4 7.1E-3 NEG NEG NEG Te-125m (1} 9.2E-1 2.6E-4 1.2E-2 NEG NEG NEG Te-127m (1} 1.5E+1 2.8E-3 l.SE-1 1.1E-4 9.4E-5 NEG Te-127 (1} 1.5E+1 3.0E-3 1. SE-1 1.1E-4 9.5E-5 NEG Te-129m (1} 2.7E+1 1.4E-2 4.1E-1 2.8E-4 2.1E-4 NEG Te-129 (1} 1.7E+1 9.0E-3 2.6E-1 1. BE-4 1.3E-4 NEG Te-131m (1} 1.8E+O 2.0E-3 2.7E-2 1.2E-5 1.4E- 5 NEG Te-131 (1} NEG 3.7E-4 4.8E-3 NEG NEG NEG Te-132 (1} 5.2E+1 S.OE-2 8.1E-1 3.1E-4 3.9E-4 NEG I-130 (1} S.BE-1 S.OE-4 1.6E-2 3.1E-5 NEG NEG I-131 (1} 1. 2El+3 l.OE+O 4.3E+1 7.2E-2 9.8E-3 NEG I-132 (1} 5.2E+l S.SE-2 8.7E-1 S.BE-4 4.4E-4 NEG I-133 (1} 1.BE+2 1.6E-1 5.5E+0 9.8E-3 1.5E-3 NEG
Rev. 32 I WOLF CREEK
TABLE 11.4-2 (Sheet 2)
(This Table is considered historical)
Spent Resins and Spent Resins And Evaporator Evaporator Dry and Filter Cartridges Filter Cartridges Bottoms *Bottoms Chemical Charcoal Compacted Isotope !Primary) ~-----!secondary)---- (Primary) (Secondary! Filters Nrui.t.e.
I-134 (1) 9.1E-1 3.9E-4 NEG 2.4E-5 NEG NEG I-135 (1) 2.8E+1 2.3E-2 6.1E-1 1.4E-3 2.4E-4 NEG Cs-134 (1) 1.8E+3 3.9E-1 3.9E+1 2.0E-2 1.3E-2 NEG cs-136 (1) 8.9E+1 1. OE-1 3.3E+O 8.6E-4 7.6E-4 NEG cs-137 (1) 1.5E+3 2.9E-1 3.0E+1 1.6E-2 l. OE-2 NEG Ba-137m (1) 1.4E+3 2.7E-1 2.8E+1 4.0E-2 9.6E-3 NEG Ba-140 (1) 1.6E+0 1.6E-3 5.6E-2 1.4E-5 1. 3E-5 NEG La-140 (1) 1.8E+0 1.7E-3 6.1E-2 1.6E- 5 1.4E-5 NEG Ce-141 (1) 1.3E+0 9.2E-4 4.1E-2 1.8E-5 l.OE-5 NEG Ce-144 (1) 3.0E+0 6.0E-4 4.7E-2 3.1E-5 1.5E-5 NEG Pr-143 (1) 4.3E-1 3.4E-4 1.5E-2 NEG NEG NEG Pr-144 (1) 3.0E+0 6.0E-4 4.7E-2 5.2E-3 1. 5E-5 NEG
Total 7.7E+3 2.9E+O 1.8E+2 1.8E-1 5.5E-2 NEG <5.0E+0
(1) Consistent with Section 11.1, the maximum activities would be obtained by multiplying the Curie Value given for the indicated isotopes by a factor of 2.
(2) The demineralizer skid resins, which discharge to the solid radwaste system, consists of activities from evaporator bottoms (primary) , evaporator bottoms (secondary) and non hazardous chemical waste.
Rev. 32 I WOLF CREEK
TABLE 11.4-3
ESTIMATED MAXIMUM ANNUAL QUANTITIES OF SOLID RADWASTE (This Table is considered historical)
Influent Volume to Solid Source Radwaste System Comments
Spent Resins
Primary 920 ft3 2 CVCS mixed, 1 CVCS cation, 1 BTRS, 1 fuel pool cleanup, 1 waste monitor, 1 waste evaporator condensate, 2 recycle evaporator feed, and 1 recycle evaporator condensate demineralizer beds. A conservative factor of 2 is applied.
Secondary* 2,000 ft3 24 steam generator blow-down demineralizer beds, 1 secondary liquid waste demi neralizer bed, 1 LRW charcoal adsorber bed, 1 SLW charcoal adsorber bed, and 1 laundry and hot shower charcoal adsorber bed.
Liquid Radwaste 154 ft3 Demineralizer Skid
Evaporator Bottoms
Primary 1,474 ft3 This includes 400 gpd from the waste holdup tank, 1140 gpd from the floor drain tank, 184 gpd
shim bleed, and 30 gpd reactor coolant drain tank (see Appen dix 11.1A}. Average boric acid concentration of reactor coolant assumed to be 1100 ppm.
Evaporator concentrates to 10 weight percent boric acid.
Rev. 32 WOLF CREEK
TABLE 11.4-3 (Sheet 2)
(This Table is considered historical)
Influent Volume to Solid Source Radwaste System Comments
Secondary* 22,026 ft3 Includes 7,200 gpd from turbine building floor drains and l condensate demineralizer vessel regeneration every 2 days, 17,940 gallon HTDS waste per regeneration, and 50 weight percent evaporator bottoms.
Filter Cartridges
Primary 239 cartridges/ Annual filter change-year (167 ft3) out numbers based on operational average of like systems:
FBG04A/B-20, FBGOS-1 FBG06-5, FBG07-1, FBM03A/B-26 I FEC01A/B-2 FEC02-l, FHA01-1, FHB06-73*, -FHB10-76 I FHBll-012, FHC01-3 I FHD01-1, FHD02-1, FHD03-l, FHD04-1, FHDOS-1, FHD06-1, FHD07-1, FHD08-l, FHE04-2, FHEOS-5, FHE06-3.
Secondary* 72 cartridges Annual filter change-out numbers based on operational averages of like systems:
FHB07 -7 I FHB08 -14 I FHC02-3, FHF04A/B-24 FHFOS-24.
Chemical Wastes 240 ft3 1,000 gallons per year chemically contaminated reactor coolant sample and two decontamination tank changeouts per year.
Rev. 32 I WOLF CREEK
TABLE 11.4-3 (Sheet 3)
(This Table is considered historical)
Influent Volume to Solid Source Radwaste System Comments
Dry and Compacted
Waste 10,000 ft3 Volume is based on data from operating plants and NRC Question 360.1(11.4).
- Normally does not require disposal as solid radwaste
Rev. 32 I WOLF CREEK
TABLE 11.4-4
ESTIMATED GENERATION OF EXPECTED AND MAXIMUM ANNUAL ACTIVITIES OF SOLID RADWASTE (CURIES)
(This Table is considerld Annua Ra 1oact1ve Ef uent Re ease Report h~$tori~al. trtual cufies release~ are documented in
Spent Resins and Spent Resins And Evaporator Evaporator Charcoal Dry and
---<-J?i*Tmary) Filter Cartridges __________ Filter Cartridges Bottoms Bottoms --- Chemical Compacted Isotope (Secondary) -(Primary) ---(Secondary) Wastes waste
cr-51 1.4E+1 9.4E-3 4.7E-1 1.6E-4 1.1E-4 NEG Mn-54 2.7E+1 5.6E-3 4.2E-1 3.0E-4 1.4E-4 NEG Fe-55 1. 9E+2 2.4E-2 2.5E+O l.4E-3 8.3E-4 NEG Fe-59 1.6E+1 9.5E-3 4.7E-1 2.3E-4 1.2E-4 NEG Co-58 4.6E+2 1.6E-1 1.2E+1 5.3E-3 3.2E-4 NEG Co-60 2.5E+2 2.8E-2 3.2E+O 1.7E-3 1.1E-3 NEG Br-83 (1) NEG NEG NEG NEG NEG NEG Br-84 (1) NEG NEG NEG NEG NEG NEG Rb-86 (1) 2.6E-1 2.7E-4 l.OE-2 NEG NEG NEG Rb-88 (1) NEG NEG NEG NEG NEG NEG Sr-89 (1) 6.5E+0 3.4E-3 1.9E-1 9.0E-5 4.9E-5 NEG Sr-90 (1) 1.4E+O 1.2E-4 1. 7E-2 NEG NEG NEG Sr-91 (1) NEG NEG NEG NEG NEG NEG Y-90 (1) 1.3E+O 1.2E-4 1.6E-2 NEG NEG NEG Y-91m (1) NEG NEG NEG NEG NEG NEG Y-91 (1) 1.5E+0 S.BE-4 4.2E-2 1. 7E-5 1.1E-5 NEG zr-95 (1) 1.5E+O 7.8E-4 1.9E-2 2.4E-5 1. 1E- 5 NEG Nb-95 (1) 3.4E+0 1.5E-3 4.9E-2 6.7E-5 2.3E-5 NEG Nb-95m (1) 1.6E+0 8.3E-4 2.0E-2 9.5E-4 1.2E-5 NEG Mo-99 (1) 7.4E-2 NEG NEG NEG NEG NEG Ru-103 (1) 5.9E-1 2.9E-4 7.7E-3 NEG NEG NEG Ru-106 (1) 9.4E-1 1.1E-4 6.7E-3 NEG NEG NEG Te-125m (1) 6.4E-1 1.8E-4 8.4E-3 NEG NEG NEG Te-127m (1) 1.2E+1 2. 4E-3 1.3E-1 9.3E-5 7.8E-5 NEG Te-127 (1) 1.2E+1 2.4E-3 1.3E-1 9.4E-5 7.8E-5 NEG Te-129m (1) 1.5E+1 7.6E-3 2.2E-1 1.SE-4 1.1E-4 NEG Te-129 (1) 9.4E+O 4.9E-3 1.4E-1 9.7E-5 7.3E-5 NEG Te-131m (1) NEG NEG NEG NEG NEG NEG Te-131 (1) NEG NEG NEG NEG NEG NEG Te-132 (1) 8.6E-2 NEG NEG NEG NEG NEG I-130 (1) NEG NEG NEG NEG NEG NEG I-131 (1) 8.9E+1 7.6E-2 3.3E+O S.SE-3 7.4E-4 NEG I-132 (1) 8.7E-2 NEG NEG NEG NEG NEG I-133 (1) NEG NEG NEG NEG NEG NEG I-134 (1) NEG NEG NEG NEG NEG NEG I-135 (1) NEG NEG NEG NEG NEG NEG
Rev. 32 I WOLF CREEK
TABLE 11.4-4 (Sheet 2)
(This Table is considered hi$torical. Actual curies released are documented in Annual Rad1oact1ve Effluent Release Report)
Spent Resins and Spent Resins And Evaporator Evaporator Dry and Filter Cartridges Filter Cartridges Bottoms Bottoms Chemical Charcoal Compacted Isotope Primary>---~ -- (Secondary) (Primary) (Secondary) wastes Filters waste
Cs-134 (1) 1.7E+3 3.8E-1 3.8E+1 1. 9E-2 1.3E-2 NEG cs-136 (1) 1.BE+1 2.1E-2 6.7E-1 1.7E-4 1.5E-4 NEG Cs-137 (1) 1.5E+3 2.9E-1 3.0E+1 1.6E-2 1.0E-2 NEG Ba-l37m (1) 1. 4E+3 2.7E-1 2.8E+1 4.0E-2 9.6E-3 NEG Ba-140 (1) 3.2E-1 3.0E-4 1.1E-2 NEG NEG NEG La-140 (1) 3.7E-1 3.5E-4 1. 3E- 2 NEG NEG NEG Ce-141 (1) 6.8E-1 4.9E-4 2.2E-2 1.0E-5 NEG NEG Ce-144 (1) 2.8E+0 5.6E-4 4.4E-2 2.9E-5 1.4E-5 NEG Pr-143 (1) 9.2E-0 NEG NEG NEG NEG NEG Pr-144 (1) 2.8E+0 5.6E-4 4.4E-2 4.8E-3 1.4E-5 NEG
Total 5.8E+3 1.3E+O 1.2E+2 9.9E-2 3.9E-2 NEG <5.0E+O
(1) Consistent with Section 11.1, the maximum activities would be obtained by multiplying the Curie value given for the indicated isotopes by a factor of 2.
(2) The demineralizer skid resins, which discharge to the solid radwaste system, consists of activities from evaporator bottoms (primary), evaporator bottoms (secondary) and non hazardous chemical wastes.
Rev. 32 I WOLF CREEK
TABLE 11.4-5
SOLID RADWASTE SYSTEM - COMPONENT DESCRIPTION
Evaporator Bottoms Tank (Primary)
Quantity l Capacity (usable), gal 1,000 Design pressure, psig 15 Design temperature,°F 250 Material SB-424, Incoloy 825 Design Code ASME Sec. VIII
Evaporator Bottoms Tank (Secondary)
Quantity 1 Capacity (usable), gal 2,500 Design pressure, psig 15 Design temperature,°F 250 Material SB-424, Incoloy 825 Design code ASME Sec. VIII Spent Resin Storage Tank (Primary)
Quantity l Capacity (usable), ft3 350 Design pressure, psig 150 Design temperature,°F 200 Material Austenitic stainless steel Design code(1) ASME Sec. VIII
Spent Resin Storage Tank (Secondary)
Quantity l Capacity (usable), gal 4,200 Design pressure, psig 150 Design temperature,°F 200 Material Austenitic stainless steel Design code ASME Sec. VIII
Spent Resin Sluice Pump (Primary)
Quantity 1 Type Canned centrifugal Design pressure psig 150 Design temperature,°F 200 Design flow, gpm Rated 140 Runout 250
Rev. 0 WOLF CREEK
TABLE 11.4-5 (Sheet 2)
Design head, ft Rated 250 Runout 210 Material Austenitic stainless steel Design code(1) Manufacturers standard (MS)
Spent Resin Sluice Pump (Secondary)
Quantity 1 Type Vertical inline centrifugal Design pressure, psig 300 Design temperature,°F 140 Design flow, gpm 225 Design head, ft 250 Material Austenitic stainless steel Design code MS
Evaporator Bottoms Tank Pump (Primary)
Quantity 1 Type Vertical inline centrifugal Design pressure, psig 300 Design temperature,°F 220 Design flow, gpm 225 Design head, ft 50 Material Alloy 20 Design code MS
Evaporator Bottoms Tank Pump (Secondary)
Quantity 1 Type Vertical inline centrifugal Design pressure, psig 300 Design temperature,°F 220 Design flow, gpm 225 Design head, ft 50 Material Alloy 20 Design code MS
Acid Addition Tank
Quantity 1 Capacity (usable), gal 250 Design pressure, psig 10 Design temperature,°F 150 Material Carbon steel Design code ASME Sec. VIII
Rev. 0 WOLF CREEK
TABLE 11.4-5 (Sheet 3)
Caustic Addition Tank
Quantity 1 Capacity (usable), gal 550 Design pressure, psig 10 Design temperature,°F 150 Material Austenitic stainless steel Design code ASME Sec. VIII
Acid Addition Metering Pump
Quantity l Type Positive displacement diaphragm Design pressure, psig 220 Design temperature,°F 104 Design flow, gph 25 Design head, psi 45 Material Alloy 20 S.S.
Design code MS Contained solution 3% H2SO4
Caustic Addition Metering Pump
Quantity 1 Type Positive displacement diaphragm Design pressure, psig 110 Design temperature,°F 104 Design flow, gph 60 Design head, psi 45 Material Alloy 20 S.S Design code MS Contained solution 50% NaOH
Resin Charging Tank (CVCS)
Quantity 1 Type Vertical, conical bottom, on wheels Capacity (usable), gal 325 Design pressure, psig ATM Design temperature,°F 120 Material Austenitic stainless steel Design code ASME Sec. VIII
Rev. 0 WOLF CREEK
TABLE 11.4-5 (Sheet 4)
Resin Charging Tank (Radwaste)
Quantity l Type Vertical, conical bottom, on wheels Capacity (usable), gal 325 Design pressure, psig Atmospheric Design temperature,°F 120 Material Austenitic stainless steel Design code ASME Sec. VIII
Spent Resin Sluice Filter (Primary) (FHC01)
- Quantity 1 Design pressure, psig 300 Design temperature,°F 250 Design flow, gpm 250 P @ design flow, psi 5 Particle Retention (See Note 2 of Table 9.3-13)
Material Austenitic stainless steel Design code(1) ASME Sec. VIII
Spent Resin Sluice Filter (Secondary) (FHC02)
- Quantity 1 Design pressure, psig 150 Design temperature,°F 250 Design flow, gpm 225 P @ design flow, psi 5 Particle Retention (See Note 2 of Table 9.3-13)
Material Austenitic stainless steel Design code ASME Section VIII
- See comments on Sheet 2 of Table 9.3-13.
Dry Waste Compactors
Quantity 2 Type Hydraulic press Design code MS
Rev. 10 WOLF CREEK
TABLE 11.4-5 (Sheet 5)
Solid Radwaste Bridge Crane Quantity 1 Capacity, tons 9.3 TV cameras, quantity 4
(1) Table indicates the required code based on its safety-related importance as dictated by service and functional requirements and by the consequences of their failure.
Note that the actual equipment may be supplied to a higher principal construction code than required.
Rev. 8 This figure has been deleted
Rev. 8 W OLF CREEK UPDATED SAFETY ANALYSES REPORT
FIGURE 11.4-2 DRUMMING PROCESS OPERATION SCHEMATIC WOLF CREEK
APPENDIX 11.4A
INTERIM ON-SITE STORAGE FACILITY
11.4A.1 Introduction
In order to permit plant operation in the event that a permanent disposal site is unavailable, it is necessary to store waste on-site. This supplemental storage is provided by the Interim On-Site Storage (IOS) Facility. The existing waste bale drum structure, which is South of the Radwaste Building, will be used as the IOS facility.
Supplemental storage, on a concrete slab or within a building addition located West of the IOS facility and South of the Radwaste Building, provides indoor/outdoor storage of equipment and/or waste which becomes activated during reactor operation.
In addition to the radwaste addition building, supplemental storage of items is permitted in the RCA yard, north laydown area, and the Owens Corning building.
In all supplemental storage locations additional restrictions limiting the radioactive content are provided in station procedures to prevent exceeding the limits of 10 CFR 20 and 10 CFR 50 Appendix I during normal operation, including anticipated operational occurrences.
11.4A.2 Design Objectives
The design of the IOS facility provides storage for solid waste produced at WCGS based on five years of processed waste (i.e. resins and sludges, including filter cartridges) and, due to storage capacity limitations, three and one half years of Dry Active Waste (DAW) generated as a result of normal operation of WCGS. The values contained in Table 11.4A-4, "Estimated Capacity and Radwaste Container Distribution for the IOS Facility", serve as the basis for the design storage capacity.
11.4A.3 Description of Containers
Containers used for packaging of radioactive material, and stored in the IOS, shall meet the applicable DOT requirements for quantity and form or the current burial site regulations for disposal (HIC) when placed in storage. Typical containers expected to be stored in the IOS facility are detailed in Table 11.4A-4. All containers are designed to reduce the occurrence of uncontrolled releases of radioactive materials due to handling, transportation, and storage.
All containers are designed with materials compatible with the stored waste to prevent significant container corrosion.
11.4A.4 Description of Stored Wastes
Solidified radwaste, or waste meeting the no free standing water criteria of Branch Technical Position ETSB 11-3 (i.e. dewatered), shall be stored in the IOS facility. These wastes satisfy all applicable transportation and disposal requirements.
Wet radioactive waste, defined as any waste which does not meet receiving burial site free liquid requirements may be temporarily stored in the IOS facility.
11.4A.4.1 Dry Active Waste (DAW)
This includes contaminated trash (paper, cloth, plastic, etc.) super compacted into drums, typically by an off-site vendor. The exposure rate from these containers is low (2 mrem/hr to about 100 mrem/hr with a majority less than 10 mrem/hr).
11.4A-1 Rev. 32 WOLF CREEK
11.4A.4.2 Solidified/Dewatered Wastes
Resin, filter cartridges and filter sludges will be transferred into HICs, and dewatered to less than 1% free standing water. Tables 11.1-6 (Sheet 1) to 11.1-6 (Sheet 4) and 11.4-4 provide normal activity concentrations in the input streams.
11.4A.4.3 Uncompactible Contaminated Waste
Other wastes not suitable for packaging in drums or HICs may be packaged in LSA boxes (B-25 or equivalent) and stored in the IOS facility, or packaged in modular storage containers and stored on the temporary outdoor storage slab.
11.4A.5 Design Concepts
11.4A.5.1 Storage areas
The wastes will be stored in four separate storage areas as identified in Table 11.4A-4 and Figures 11.4A-1 and 2.
- a. High and Low Level Storage Areas Two separate areas containing all three forms of waste (i.e. super compacted DAW in drums, solidifed/dewatered waste in HICs, and uncompactible waste in LSA boxes).
- b. DAW Storage Areas Two separate areas, adjacent to the high and low level storage areas, containing super compacted DAW in drums.
The storage areas act as a protective barrier to:
- a. Protect the waste containers from weather effects.
- b. Prevent an uncontrolled release of radioactive material to the environment.
- c. Provide shielding for radiation emitted by the waste.
11.4A.5.2 Handling and Storage Operations
Inventory data including batch number, container number, date of storage, and other necessary data shall be maintained. The design includes an index system that allows specific identification of container locations so that administrative controls may be used to effectively inventory stored wastes.
Containers to be stored in the IOS facility are first visually inspected and checked for surface contamination. No damaged containers will be sent to the IOS facility.
Details of the IOS facility layout are shown in Figures 11.4A-1 and 2. The actual waste container configuration may deviate from the above description based on changing waste processing/storage needs. Upon retrieval of containers from storage for transport and permanent disposal, each container is swipe tested.
11.4A.5.3 Personnel Exposure
As required by 10CFR20, occupational exposures shall be kept as low as reasonably achievable (ALARA). During waste handling operations, only employees required to handle the shipment, perform maintenance activities, or perform inspections are allowed in the areas of the IOS facility for the time needed to perform their task.
All operations in the IOS facility are controlled by plant radiation protection personnel to assure that all employees are monitored, confirm that dose limits are not exceeded, and ensure that good working practices are being followed. All operations are conducted in accordance with written procedures.
11.4A-2 Rev. 34 WOLF CREEK
To reduce the possible exposure of personnel during inspection and maintenance, the following concepts have been incorporated in the design of the IOS facility:
- a. The IOS facility and equipment are designed to require minimum maintenance activities in high radiation storage areas.
- b. Containers are handled by a remote-controlled crane carrying CCTV cameras and lights.
- c. Inspection of the storage areas in the IOS facility is to be accomplished using CCTV from the solidification control panel room.
- d. Access to the bridge crane and its cables is provided over the truck bay area to reduce exposure to maintenance personnel. Additional portable shielding may be used as necessary.
- e. Additional portable shields may be used as necessary.
11.4A.5.4 Provision for Liquid Drainage
The IOS facility is provided with an internal drainage system consisting of trenches and stainless steel piping which route potentially contaminated water to a radwaste sump. The drainage is then pumped to the liquid radwaste system, and processed prior to discharging. Walls and curbs are utilized to confine any potentially contaminated water inside the building. The IOS facility is also provided with exterior storm drains to prevent water from entering the storage areas. (see Section 9.3)
11.4A.5.5 Structural and Architectural
The IOS facility is a non-nuclear safety non-seismic Category I structure. The finished floors in the storage areas are constructed with minimal slope in order to accommodate drum stacking, and covered with an easily decontaminable material. The roof of the storage building consists of built up roofing and rigid insulation on a metal deck.
11.4A.5.6 Shielding
Shielding evaluations were performed utilizing the waste stream distribution, historical generation, isotopic activities, and storage configurations as described in Tables 11.4A-1, 2, 3A - 3D, and 4, and Figures 11.4A-1 and 2. The storage configuration provides adequate shielding for five years of radioactive waste. The concrete walls provide shielding primarily for the outer layers of containers. Consideration was given for a self-shielding effect due to the large number of containers in the storage areas (i.e. containers with high exposure rates will, to the extent possible, be placed in the center of the storage areas using containers with lower exposure rates for shielding). The roof, made of built up roofing and rigid insulation on a metal deck, provides shielding equivalent to approximately 0.25 inches of steel. Additional portable container shields may be used as necessary.
Maximum anticipated dose rates outside of the IOS are shown in Tables 11.4A-5A and 5B. Maximum anticipated dose rates along the south RCA boundary are shown in Table 11.4A-6. The dose rates are also shown in Figures 11.4A-3, 11.4A-3A and 11.4A-3B.
11.4A-3 Rev. 18 WOLF CREEK
11.4A.5.7 Design Basis Events
11.4A.5.7.1 Fire Protection
Fire protection is accomplished through the use of non-combustible construction materials, local fire extinguishers, and local hose stations. Fire/smoke detection devices, which alarm locally, and in the main control room, are provided throughout the IOS facility. The only combustible material in the IOS facility is DAW and HIC liner material (high density, cross linked polyethylene).
11.4A.5.7.2 Flood Protection
The topography of the site is such that flooding from natural causes is not a design basis event for above grade buildings. (see Section 2.0)
11.4A.5.7.3 Wind Protection
The IOS is a reinforced structure designed for a wind velocity of 100 miles/hr.
This velocity corresponds to a recurrence time of 100 years.
11.4A.5.7.4 Tornado Protection
The storage areas and stored waste have been evaluated with respect to a tornado, and it has been determined that the design is such that there will be no adverse affects from a tornado for the following reasons.
a) All waste is stored in a form that is resistant to the release and spread of radioactivity.
b) Waste with high activity levels will be stored in tornado resistant rooms (i.e. rooms that have three foot thick reinforced walls which are 16'-9" high) in containers that, due to their weight, will remain in place during a tornado.
c) Waste with low activity levels will be stored in non-tornado resistant rooms (i.e. rooms that have only one foot thick reinforced masonry block walls which are 14' high). However, the waste that will be stored in the non-tornado resistant rooms will have low activity levels (i.e., 2 mrem/hr to 100 mrem/hr, with the majority less than 10 mrem/hr).
d) The non-tornado resistant rooms, although they themselves do not provide resistance to a tornado, are protected from a tornado by surrounding structures. The rooms are located in the Waste Bale Drumming Area which is designed to withstand 100 mph winds. Also, most tornadoes come from the southwest, and the rooms will be shielded by three foot thick 16'-9" high walls on the west, a concrete segmented shield on the south, and the Radwaste Building on the north.
e) If, in the unlikely event that most of the waste stored in the non-tornado resistant rooms were dispersed during a tornado, the released activity levels would remain below the 2.5 rem whole body or 30 rem thyroid dose limit allowed by GL 81-38.
f) In the unlikely event a tornado missile were to enter one of these rooms, and penetrate a container, the missile would tend to plug its own hole, minimizing any potential for release of radioactivity.
Liquid waste will be contained by the curbs and floor drain system.
11.4A-4 Rev. 18 WOLF CREEK
Based on these reasons, the storage of radwaste as allowed per this modification does not present a radiation hazard with respect to a tornado. In the unlikely event of waste container failure or dispersal due to a tornado, plant procedures will provide instructions on handling and repackaging/reprocessing of the waste on a case by case basis. In case of a unique failure not anticipated in plant procedures, WCGS Engineering and Technical personnel would evaluate the situation and determine the best course of action based on the specific conditions.
11.4A.5.7.5 Seismic Event
In the unlikely event of waste container failure due to a seismic event, plant procedures will provide instructions on handling and repackaging/reprocessing of the waste on a case by case basis. A failure due to a seismic event would in all likelihood result in the failed container remaining within the IOS facility. In case of a unique failure not anticipated in plant procedures, WCGS Engineering and Technical personnel would evaluate the situation and determine the best course of action based on the specific conditions. In no case would the method of resolution fail to meet shipping and burial criteria, or result in any radioactive release to the environment.
11.4A.5.7.6 Waste Container Failure
In the unlikely event of waste container failure after final packaging, during storage, or prior to shipment, plant procedures will provide instructions on handling and repackaging/ reprocessing of the waste on a case by case basis. A failure within the IOS facility would in all likelihood result in the failed container remaining within the IOS facility. In case of a unique failure not anticipated in plant procedures, WCGS Engineering and Technical personnel would evaluate the situation and determine the best course of action based on the specific conditions. In no case would the method of resolution fail to meet shipping and burial criteria, or result in any radioactive release to the environment.
11.4A.5.8 HVAC Systems
The IOS facility is maintained at a negative pressure by the Radwaste Building ventilation system. This is accomplished by an interlock that requires an exhaust fan in operation, prior to starting a supply fan. Also, two exhaust fans are provided with interlocks to ensure that upon the loss of one fan, the other will automatically start. All exhaust air is monitored and filtered prior to release. (see Section 9.4.5)
11.4A.5.9 Bridge Crane
11.4A.5.9.1 Crane Description
The bridge crane has a rated capacity of 9-1/3 tons. The crane has the capability to handle all containers (i.e. HICs, LSA boxes, and drums). The drum grab has the capability to recover fallen drums. The crane carries TV cameras and lighting for storage, handling and inspection of containers, and may perform other tasks in the storage and truck bay areas as required.
There are two motors on the crane, one high speed and one low speed for bridge, trolley and hoist movement. The redundant motors can be used to move the crane in the event one motor fails. In the event of other problems, a cable can be manually attached for crane retrieval.
11.4A-5 Rev.13 WOLF CREEK
11.4A.5.9.2 Crane Control
The solid radwaste control console is equipped so that radwaste movements may be accomplished by remotely controlling the bridge crane. The crane system is designed for precise placement of drums, HICs or LSA boxes, and for lifting and placement of the cask transportation lid. The bridge and trolley are accurately positioned by the use of a CCTV monitoring system and an overhead index system. It will have sufficient range to move HICs from the solid radwaste disposal station to the storage areas, and unload drums and boxes from the trucks and move them to their storage areas.
11.4A.5.9.3 CCTV System
The CCTV includes cameras mounted on the bridge crane. Monitors are installed in the solidification control panel room. They are equipped with manual control capabilities to adjust the pan and tilt for the cameras. The cameras on the crane are fixed focus and adjusted locally to get a close view of any container for inspection purposes, the two surveillance cameras have pan and tilt capabilities.
11.4A.5.10 Lighting
Fixed lights are provided throughout the IOS facility. These lights provide illumination for all IOS activities, including inspections.
11.4A.5.11 Security The IOS facility is surrounded by a chain link fence bounding the RCA. Access to the IOS facility is controlled to minimize personnel exposure.
11.4A.6 Monitoring Operations
11.4A.6.1 Containers
Before the radioactive waste containers are placed in storage, the activity level of each container is determined. Radiological monitoring of the storage containers is performed using portable equipment. Swipe testing and analysis capability is provided in the truck bay area.
11.4A.6.2 Storage Areas
The IOS facility includes provision for remote monitoring of the storage areas through closed circuit television (CCTV) so that the condition of any stored container can be observed. In order to maximize visual inspection in the storage areas for the longest period of time, drums will initially be stacked in every other row, to the extent practicable.
Area radiation monitors are installed, one in the corridor across from the radwaste control room and another in a truck bay area near the personnel entrance. If predetermined radiation setpoints are exceeded, alarms sound both locally and in the main control room. Additional radiation monitoring is performed by the plant radiation protection group using portable equipment as necessary.
11.4A-6 Rev.19 WOLF CREEK
11.4A.6.3 Offsite
The IOS facility is designed to ensure that the annual dose to the public is a small fraction of the 25 mrem/yr allowed from all sources of the Uranium cycle, as per 40CFR190. Exposure levels are monitored at the RCA boundary fence using RDD dosimeters. Table 11.4A-7 details anticipated dose rates at the restricted area boundary.
All potential pathways for the release of radioactivity to the environment are controlled and monitored. In particular, water from potentially contaminated drains is processed in the liquid radwaste system, and air from the IOS facility is processed in the Radwaste Building exhaust system. Both systems sample and analyze for radioactivity prior to release to the environment. (see Section 11.5)
Since the normal operation of the IOS facility is not expected to produce any radioactive discharge or otherwise hazardous effluents, no significant effects on environmental air or water quality are expected. Offsite environmental surveillance is implemented through the environmental monitoring program.
11.4A-7 Rev. 25 TABLE 11.4A-1
ISOTOPIC DISTRIBUTION OF RADWASTE (PERCENT ABUNDANCE)
NUCLIDE HALF-LIFE *** RESINS, FILTERS & EVAP *** DAW (DAYS) CLASS A CLASS B CLASS C -------
Mn-54 312.7 1.43 3.94 1.80 1.45 Fe-55 2.7* 57.35 21.70 41.00 59.70 Co-57 270.9 0.00 0.43 0.00 0.11 Co-58 70.8 2.28 22.70 25.60 1.69 Co-60 5.27* 12.67 11.70 6.60 24.90 Ni-59 75000* 0.00 0.17 0.00 0.00 Ni-63 100.1* 15.07 16.20 12.60 7.37 Ag-110m 249.85 0.00 0.00 1.90 0.24 H-3 12.28* 0.88 0.00 3.40 0.02 C-14 5730* 0.17 0.54 0.50 0.00 Nb-95 35.06 0.00 0.10 1.60 1.73 Cs-134 2.062* 4.02 9.07 0.30 1.02 Cs-137 30.17* 6.07 12.50 0.50 1.46 Ce-144 284.3 0.00 0.00 0.10 0.34 Sb-125 2.77* 0.00 0.76 0.00 0.00 Cm243/44 28.5* 0.002 0.00 0.00 0.00 Sr-95 24.4** 0.00 0.00 0.00 0.00 Zr-95 64.02 0.00 0.15 2.40 0.00 SR-90 28.6* 0.00 0.01 0.00 0.00 Cr-51 27.7 0.00 0.00 1.70 0.00
BASED ON CHARACTERIZATION OF WASTE SAMPLES FROM PLANT OPERATIONS DURING 1988 TO 1991 AND RADMAN COMPUTER CODE.
DAW ISOTOPIC DISTRIBUTION IS BASED ON RADMAN COMPUTER CODE.
- HALF-LIFE IN YEARS
- HALF-LIFE IN SECONDS
Rev. 8 TABLE 11.4A-2
AVERAGE ANNUAL ACTIVITY OF RADWASTE (RESINS/FILTERS)
(1988 TO 1991)
- WASTE CLASS **********
TYPE/ ***** CLASS A ***** ***** CLASS B ***** ***** CLASS C *****
PERIOD VOLUME ACTIVITY VOLUME ACTIVITY VOLUME ACTIVITY (ft3) (mCi) (ft3) (mCi) (ft3) (mCi)
====== ======== ====== ======== ====== ==
120.3 5.31E+05 120.3 5.05E+05 84.3 1.72E+04 120.3 6.29E+05 120.3 2.56E+05 120.3 1.11E+05 205.8 2.31E+05 1988 TO 411.6 3.46E+01 120.3 1.94E+05 1991 411.6 1.06E+02 120.3 2.07E+05 205.8 2.64E+05 205.8 8.00E+03 388.2 1.30E+00 83.4 8.91E+02 83.4 8.53E+02 83.4 7.80E+03 83.4 2.12E+03 205.8 4.65E-02
TOTAL 2523.3 1.55E+06 687 1.39E+06 84.3 1.72E+04 ANNUAL AVE 630.8 3.89E+05 171.8 3.48E+05 21.1 4.30E+03 mCi/Cuft 6.16E+02 2.03E+03 2.04E+02
PROJECTED VOL. CUFT. 710 200 80
EST'D Ci/Yr 4.37E+02 4.06E+02 1.63E+01
Rev. 8
TABLE 11.4A-4
Estimated Capacity and Radwaste Container Distribution for the IOS Facility
AREA DIMENSIONS WASTE 5 Yr CAPACITY*
ACTIVITY (Inside) TYPE CONTAINERS (cuft)
(CURIE)
HLSA 30' x 20'9" Primary 20 PL8-120 2,406 2,480 Resin
LLSA 30' x 46'
SECTION 5 Resin/ 5 PL6-80 417 228 Filter
SECTION 6 Sec Resin 6 PL14-215 1,235 182
SECTION 7 DAW 36 B25-boxes 3,456 6
SECTION 8 DAW 162 Drums 1,782 5
DRUM AREA 'A' 5'5" x 32'11" DAW 147 Drums 1,617 4
DRUM AREA 'B' 15'5" x 17'10" DAW 72 Drums 792 2
TOTAL 60' x 100' 31 HICS 11,705 2,907 including Truck Bay: 36 boxes and 381 drums
- Volume is based on the Following anticipated usage and waste configuration as shown in Figure 11.4A-1.
WASTE CONTAINER CONTAINER STREAM TYPE VOLUME
PRIMARY RESIN PL8-120 120.3 cuft SECONDARY RESIN PL14-215 205.8 cuft FILTERS PL6-80 83.4 cuft DAW 85 Gal.Drum 11 cuft 79 Gal.Drum 11 cuft 55 Gal.Drum 7.5 cuft B-25 Box 96 cuft
Rev. 8
TABLE 11.4A-7 TOTAL OFFSITE DOSE AT THE UNRESTRICTED AREA (Exclusion Area Boundary, EAB, is 1200 meters from Center of Containment)
Dose at the EAB Source/EAB direction Hourly Dose Annual Dose (mrem/hr) (mrem/yr) 5 Year Storage
West Side
High Level Storage 1.5160E-06 Low Level Storage 5 1.1515E-07 Low Level Storage 6 5.4731E-08
5 yr Storage Dose At West EAB 1.6859E-06 0.0148
South Side
High Level Storage 7.7607E-07 Low Level Storage 5 8.0440E-07 Low Level Storage 6 1.7227E-07
5 yr Storage Dose At South EAB 1.7527E-06 0.0154
East Side
High Level Storage 1.7690E-06 Low Level Storage 5 1.1512E-07 Low Level Storage 6 6.1442E-08
5 yr Storage Dose At East EAB 1.9456E-06 0.0171
3 Year Storage
West Side
High Level Storage 4.9940E-07 Low Level Storage 5 4.2856E-08 Low Level Storage 6 3.8725E-08
3 yr Storage Dose At West EAB 5.8097E-07 0.0051
South Side
High Level Storage 5.4989E-07 Low Level Storage 5 5.3937E-07 Low Level Storage 6 5.9423E-08
3 yr Storage Dose At South EAB 1.1486E-06 0.0101 Rev. 13 TABLE 11.4A-7 (Sheet 2)
TOTAL OFFSITE DOSE AT THE UNRESTRICTED AREA (Exclusion Area Boundary, EAB, is 1200 meters from Center of Containment - Continued)
Dose at the EAB Source/EAB direction Hourly Dose Annual Dose (mrem/hr) (mrem/yr)
East Side
High Level Storage 8.9247E-07 Low Level Storage 5 9.6958E-08 Low Level Storage 6 4.3531E-08
3 yr Storage Dose At East EAB 1.0330E-06 0.0091
2 Year Storage
High Level Storage 8.3118E-07 Low Level Storage 5 2.3780E-08 Low Level Storage 6 2.7916E-08
2 yr Storage Dose At West EAB 8.8288E-07 .0077
South Side
High Level Storage 4.0898E-07 Low Level Storage 5 4.1932E-07 Low Level Storage 6 2.7141E-08
3 yr Storage Dose At South EAB 8.5544E-07 .0075
East Side High Level Storage 4.5226E-07 Low Level Storage 5 8.8748E-08 Low Level Storage 6 3.1376E-08
2 yr Storage Dose At East EAB 5.7238E-07 0.0050
NOTE: Low Level Storage Sections are described by Figure 11.4A-1.
Rev. 13 c c
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WOLF CREEK
11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS
The function of the process and effluent radiological monitoring systems is to monitor, record, and control the release of radioactive materials that may be generated during normal operation, anticipated operational occurrences, and postulated accidents.
The process and effluent radioactivity monitoring systems furnish information to operations personnel concerning radioactivity levels in principal plant process streams and atmospheres. The monitoring systems indicate and alarm excessive radioactivity levels (GDC-63). They initiate operation of standby systems, provide inputs to the ventilation and liquid discharge isolation systems, and record the rate of release of radioactive materials to the environs, as outlined in Regulatory Guide 1.21 and GDCs 60 and 64. The systems consist of permanently installed, continuous-monitoring devices together with a program and provisions for specific sample collections and laboratory analyses.
11.5.1 DESIGN BASES
The principal objectives and criteria of the process and effluent radiological monitoring systems are provided below.
11.5.1.1 Safety Design Bases
SAFETY DESIGN BASES - The control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, and the fuel building exhaust monitors are designed to activate engineered safety features systems in the event that airborne radioactivity in excess of allowable limits exists.
Additional design bases are stated in the following sections:
- a. Containment purge isolation system, Sections 6.2.4, 7.3.2, 9.4.6, and 12.3.4.
- b. Fuel building ventilation isolation, Sections 7.3.3, 9.4.2, and 12.3.4.
- c. Control room intake isolation, Sections 6.4.1, 7.3.4, 9.4.1, and 12.3.4.
These radioactivity monitors are protection system elements and are designed in accordance with IEEE Standard 279.
The safety evaluation of these systems is discussed in Section 7.3.
11.5-1 Rev. 0 WOLF CREEK
These monitors also serve for in-plant worker protection, and this function is discussed in Section 12.3.4. Compliance with Regulatory Guide 1.97 is discussed in Appendix 7A.
11.5.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The process and effluent radioactivity monitors operate continuously during both intermittent and continuous discharges of potentially radioactive plant effluents, in compliance with Regulatory Guide 1.21. The monitors verify that the most restrictive anticipated nuclides are at concentrations within the limits specified in 10 CFR 20 and that the concentrations are low enough that 10 CFR 50, Appendix I, dose guidelines are met for unrestricted areas.
POWER GENERATION DESIGN BASIS TWO - The process and effluent radioactivity monitors alarm and automatically terminate the release of effluents when radionuclide concentrations exceed the limits specified (GDC-60). Where termination of releases is not feasible, the monitors provide continuous indication of the magnitude of the activity released.
POWER GENERATION DESIGN BASIS THREE - The radwaste process system monitors measure radioactivity in process streams to aid personnel in the treatment of radioactive fluids prior to recycle or discharge (GDC-63).
POWER GENERATION DESIGN BASIS FOUR - The process and effluent radioactivity monitors monitor the containment atmosphere, spaces containing components for recirculation of LOCA fluids, and effluent discharge paths for radioactivity that may be released from postulated accidents, as required by GDC-64.
POWER GENERATION DESIGN BASIS FIVE - The process and effluent monitors indicate the existence and, to the extent possible, the magnitude of reactor coolant and reactor auxiliary system leakage to the containment atmosphere, cooling water systems, or the secondary side of the steam generators.
POWER GENERATION DESIGN BASIS SIX - The process and effluent radioactivity monitors provide alarm and automatic termination of the transfer of radioactivity fluids to storage facilities in zone A areas, defined in Section 12.4.1.1.
POWER GENERATION DESIGN BASIS SEVEN - Process radioactivity monitors provide alarm and gross indication of the extent of any failed fuel within the primary system.
11.5-2 Rev. 7 WOLF CREEK
POWER GENERATION DESIGN BASIS EIGHT - The effluent radioactivity monitors provide sufficient radioactivity release data to prepare the reports required by Regulatory Guide 1.21.
11.5.1.3 Codes and Standards
Codes and standards applicable to the process and effluent radioactivity monitors are indicated in Table 3.2-1. The monitors listed in Section 11.5.1.1 are designed as protection system elements.
11.5.2 SYSTEM DESCRIPTION
11.5.2.1 General Description
11.5.2.1.1 Data Collection
The process and effluent radiological monitoring systems consist of liquid and airborne radioactivity monitors with the attendant controls, alarms, pumps, valves, and indicators required to meet the design bases. Each monitor consists of the detector assembly and a local microprocessor. The local microprocessor processes the detector assembly signal in digital form, computes average radioactivity levels, stores data, performs alarm or control functions, and transmits the digital signal to the control room microprocessor. Signal transmission is accomplished via redundant data highways. A single fault in either data highway would not prevent the control room microprocessor from receiving the data.
The local microprocessors for monitors which perform safety functions (control room ventilation, fuel building ventilation, containment atmosphere, and containment purge monitors, refer to Section 12.3.4) are wired directly to individual indicators located on the seismic Category I radioactivity monitoring system cabinets in the control room. The input from the safety-related channels to the daisy-chain loop is an isolated signal to ensure that the safety-related signals are not affected by signals or conditions existing in the nonsafety portion of the system.
The control room microprocessor provides controls and indication for the radioactivity monitoring system. Indication is via a CRT located in the control room. The signals from each monitor may also be recorded on a system printer.
11.5.2.1.2 Alarms
Each monitor channel is provided with a three-level alarm system. One alarm setpoint is below the background counting rate and serves as a circuit failure alarm. The other two-alarm setpoints provide sequential alarms on increasing radioactivity levels. Loss
11.5-3 Rev. 0 WOLF CREEK
of power causes an alarm on all three-alarm circuits. The alarms must be manually reset and can be reset only after the alarm condition is corrected.
11.5.2.1.3 Check Sources
Each monitor is provided with a check source, operated from the control room, which simulates a radioactive sample in the detector assembly for operational and gross calibration checks.
11.5.2.1.4 Power Supplies
All Class IE radioactivity monitoring systems are powered from Class IE motor control centers. The power supplies for all of the monitors are given in Table 11.5-5.
11.5.2.1.5 Calibration and Maintenance
The radioactivity monitors are calibrated by the manufacturer for at least the principal radionuclides listed in Tables 11.5-1 through 11.5-4. The manufacturer's calibration standards are traceable to National Institute of Standards and Technology primary calibration standard sources and are accurate to at least 5 percent. The source detector geometry during this primary calibration is identical to the sample detector geometry. Secondary standards counted in reproducible geometry during the primary calibration are supplied with each continuous monitor. Each continuous monitor is calibrated at a frequency established by station procedures.
The count rate response of each continuous monitor to remotely positionable check sources is recorded by the manufacturer after the primary calibration.
This count rate response and background count rate is checked at intervals specified by plant procedures during reactor operation.
Surveillance is performed in accordance with Technical Specifications or the ODCM.
Any fluid released to the environment is analyzed for radioactivity prior to release. If, at any time, a monitor requires maintenance or decontamination, the process flow is terminated or periodic grab sampling with laboratory analysis is implemented.
11.5-4 Rev. 13 WOLF CREEK
This does not impair system integrity since the detector is off-line and not installed in the stream.
11.5.2.1.6 Sensitivities
Each effluent monitoring system is able to detect a minimum concentration within the release limits established in the Technical Specifications.
Due to sensitivity considerations, monitors are located at the effluent release points. Dilution factors between the release point and the site boundary are considered in complying with the limitations of 10 CFR 50, Appendix I. Tables 11.5-1 through 11.5-4 provide the detailed sensitivity selection criteria for the process and effluent monitors.
11.5.2.1.7 Monitor Locations
The location of each process and effluent radioactivity monitor is shown on the radiation zone drawings, Figure 12.3-2. The monitors are located in low background areas, near the systems being monitored, to minimize background and sampling interferences.
11.5.2.1.8 Ranges and Setpoints
The ranges of the various process monitors are based on the expected activity levels in the system being monitored. The bases for their setpoints are determined by the need for process control and to alert the operators of leakage of radioactivity into normally nonradioactive systems.
The ranges of the various effluent monitors are based on the ability to detect radioactivity concentrations at the effluent release point which might result in site boundary doses in excess of 10 CFR 50 Appendix I levels to those from postulated accidents. The Hi alarm is administratively established at a point sufficiently below the Hi-Hi alarm so as to provide additional assurance that Technical Specification limits are not exceeded. The Hi-Hi alarm is established to ensure that Technical Specification limits are not exceeded.
(See Offsite Dose Calculation Manual.)
The ranges and setpoints for the process and effluent monitors are provided in Tables 11.5-1 through 11.5-4.
11.5-5 Rev. 14 WOLF CREEK
11.5.2.1.9 Expected System Parameters
The expected ranges of system parameters, such as flow, composition, and concentrations, are summarized in Tables 11.5-1 through 11.5-4. Detailed information on the individual systems can be found in other sections of the USAR, principally Chapters 9.0 and 11.0.
11.5.2.2 Liquid Monitoring Systems
11.5.2.2.1 Selection Criteria for Liquid Monitors
The liquid monitors consist of fixed-volume, off-line, leadshielded sample chambers through which the liquid samples flow. A NaI(Tl) gamma scintillation detector is located within each sample chamber to detect the activity level.
The detector assemblies monitor gross gamma activity in the range of 10 -7 to 10-2 mCi/ml. These range apply to all liquid monitors except O-SJ-RE-01 The controlling isotope for the liquid monitors is Cs-137. Minimum detectable concentrations are listed in Tables 11.5-1 and 11.5-2.
A manually operated isolation valve at the sample chamber inlet is provided to permit purging of the sample chamber to facilitate background activity checks.
A source of noncontaminated water is provided for decontamination purposes.
Sample chambers in which permanent contamination interferes with measurement can readily be replaced. Liquid monitor alarms are annunciated in the control room on the plant annunciator, the NPIS computer, and the radiation monitoring system CRT (RM-11). The NPIS computer located in the TSC provides a visual display of alarm status. The RM-11 in the control room provides audible and visual alarm indication.
The liquid radioactivity monitors are located to comply with the design bases.
The specific sample points are selected to provide representative samples of the systems monitored, to reduce sample transport times, and to limit the amount of radioactivity released in the event of a high radioactivity signal.
The continuous liquid radioactivity monitoring systems are discussed in the following sections. A summary of the functions and characteristics of each monitor is presented in Tables 11.5-1 and 11.5-2.
11.5-6 Rev. 14 WOLF CREEK
11.5.2.2.2 Liquid Process Radioactivity Monitors
A detailed listing of liquid process monitor parameters is given in Table 11.5-1.
11.5.2.2.2.1 Component Cooling Water Monitors
The component cooling water system (CCWS) is discussed in Section 9.2.2.
The CCWS radioactivity monitors, 0-EG-RE-9 and 0-EG-RE-10, detect, indicate, and alarm elevated radiation levels in the CCWS. The elevated radiation levels would be indicative of radioactive leakage into the CCWS from systems and components served by the CCWS. Each detector assembly receives a continuous sample flow when an associated CCWS pump is operating. The CCWS pumps provide the motive force for the sample flow. Each detector sample is taken from the CCWS upstream of the CCW heat exchanger and the sample is returned to the CCWS downstream of the heat exchanger. The alert alarm provides indication of radioactive inleakage to the system. A high alarm is provided to indicate increasing radioactivity levels and to close the component cooling water surge tank air vent and makeup water valves.
11.5.2.2.2.2 Steam Generator Liquid Radioactivity Monitor
The steam generator liquid sample system is discussed in Section 9.3.2.
The steam generator liquid radioactivity monitor, 0-SJ-RE-2,continuously monitors the blowdown from the steam generators, either individually or collectively, to detect, indicate, and alarm primary-to-secondary system leaks in the steam generators. This monitor closes the steam generator blowdown isolation valves on high radiation to prevent the discharge of radioactive fluid and to limit radioactive contamination of the blowdown demineralizers.
The monitor also provides backup information and verification of the condenser air removal system gaseous radioactivity monitor (Section 11.5.2.3.2.1). The fixed-volume detector assembly receives a continuous flow from the steam generator liquid sample header which samples the tube sheet area near the minimum water level of the steam generators. The sample point is located downstream of the sample system heat exchanger to provide conditioning and pressure reduction of the radioactivity monitor sample. The radioactivity alarms provide indication of primary-to-secondary leakage in the steam generator.
11.5-7 Rev. 11 WOLF CREEK
11.5.2.2.2.3 Steam Generator Blowdown Processing System Radio-activity Monitor
The steam generator blowdown processing system is discussed in Section 10.4.8.
The steam generator blowdown process radioactivity monitor, 0-BM-RE-25, continuously monitors the fluid entering the steam generator blowdown filters to detect, alarm, and indicate excessive radioactivity levels in the blowdown system. The steam generator blowdown process radioactivity monitor acts to terminate blowdown from the steam generators to prevent discharge of radioactive fluid and to limit radioactive contamination of the blowdown demineralizers. The monitor provides backup information for the steam generator liquid radioactivity monitor (Section 11.5.2.2.2.2) and the condenser air removal gaseous radioactivity monitor (Section 11.5.2.3.2.1) for the detection of a primary-to-secondary leakage in the steam generator. The fixed-volume detector assembly receives a continuous flow from the discharge of the blowdown system heat exchangers and returns the sample to the system. The sample location provides an unfiltered sample at temperatures within the limits of the detector. The high radioactivity alarm closes the steam generator blowdown isolation valves and the blowdown system discharge valve to terminate blowdown and prevent discharge of radioactivity from the steam generators.
11.5.2.2.2.4 Boron Recycle System Distillate Radioactivity Monitor
The boron recycle system is discussed in Section 9.3.6.
The boron recycle radioactivity monitor, 0-HE-RE-16, is permanently out of service and no longer used
11.5-8 Rev. 14 WOLF CREEK
11.5.2.2.2.5 Chemical and Volume Control System Letdown Monitor
The chemical and volume control system (CVCS) is discussed in Section 9.3.4.
The CVCS letdown radioactivity monitor, 0-SJ-RE-01, acts as a gross failed fuel detector. The fixed-volume detector assembly continuously monitors the CVCS letdown sample line which extracts a sample upstream of the CVCS letdown demineralizers. The radiation alarms alert the operator to an abnormal increase in gross gamma activity in the CVCS letdown system. Determination of the cause can be made by laboratory analysis. The sample location provides an unfiltered sample prior to demineralization. The arrangement and location of the sample line provide sufficient delay in transport to allow decay of nitrogen-16, which could cause erroneously high readings.
11.5.2.2.2.6 Auxiliary Steam System Condensate Recovery Monitor
The auxiliary steam system is discussed in Section 9.5.9.
The auxiliary steam condensate recovery radioactivity monitor, 0-FB-RE-50, detects radioactive contamination from the potentially radioactive systems which discharge to the auxiliary steam condensate recovery tank. The fixed-volume detector assembly continuously monitors the discharge of the auxiliary steam condensate transfer pumps. The radioactivity alarms alert the operator to possible contamination, isolates auxiliary steam supply to the radwaste building and trips the auxiliary steam condensate transfer pumps. The source of the contamination can be determined by selective isolation of the potentially radioactive systems. The sample location ensures that all potentially radioactive sources are monitored.
11.5-9 Rev. 14 WOLF CREEK
11.5.2.2.3 Liquid Effluent Radioactivity Monitors
A detailed listing of the liquid effluent monitor parameters is given in Table 11.5-2.
11.5.2.2.3.1 Steam Generator Blowdown Discharge Radioactivity Monitor
The steam generator blowdown system is discussed in Section 10.4.8.
The steam generator blowdown discharge radioactivity monitor, 0-BM-RE-52, continuously monitors the blowdown discharge pump outlet to detect radioactivity due to system demineralizer break-through and to provide backup to the steam generator blowdown process radioactivity monitor (Section 11.5.2.1.2.3) to prevent discharge of radioactive fluid. The sample point is located on the discharge of the pump in order to monitor discharge or recycled blowdown fluid and upstream of the discharge isolation valve to limit the radioactivity released.
The high radioactivity alarm acts to close the blowdown isolationvalves and the blowdown discharge valve.
A weekly laboratory isotopic analysis is made for any liquid discharged, in conformance with Regulatory Guide 1.21.
11.5.2.2.3.2 Liquid Radwaste Discharge Monitor
The liquid radwaste system is discussed in Section 11.2.
The liquid radwaste radiation monitor, 0-HB-RE-18, continuously monitors the discharge of the liquid radwaste processing system to prevent the discharge of radioactive fluid to the environs. The fixed-volume detector assembly continuously monitors the system discharge line upstream of the discharge valve. The high radioactivity alarm closes the liquid radwaste system discharge valve to terminate discharge. The sample point is located to ensure that all potentially radioactive fluids from the liquid radwaste processing system are monitored prior to discharge. Laboratory isotopic analyses are made of each batch prior to discharge, as required by Regulatory Guide 1.21 and the plant Technical Specifications.
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11.5.2.2.3.3 Secondary Liquid Waste System Monitor
The secondary liquid waste system is discussed in Section 10.4.10.
The secondary liquid waste system discharge radioactivity monitor, 0-HF-RE-45, monitors secondary liquid waste system effluents prior to discharge to the environs. The fixed-volume detector assembly continuously monitors the discharge line upstream of the discharge isolation valve. The high radioactivity alarm closes the secondary liquid waste system discharge valve to prevent the discharge of radioactive fluid. The sample location ensures that all potentially radioactive sources from the system are monitored prior to discharge. Laboratory isotopic analyses are made of each batch prior to discharge, in accordance with Regulatory Guide 1.21.
11.5.2.2.3.4 Turbine Building Drain Monitor
The turbine building drain effluent radioactivity monitor, 0-LE-RE-59, is provided to monitor turbine building liquid effluents prior to release to the environs. The fixed-volume detector assembly continuously monitors the drain effluent line upstream of the drain line isolation valve. The high radioactivity alarm closes the drain line isolation valve to prevent the release of radioactive fluids. The sample location ensures that all potentially radioactive turbine building liquid effluents are monitored prior to discharge. A weekly isotopic analysis is made in the laboratory, in conformance with Regulatory Guide 1.21.
11.5.2.2.3.5 Wastewater Treatment System Monitor
Radioactivity monitor HF-RE-95 monitors the discharge from the high and low TDS collection drain tanks to the Wastewater Treatment System. The fixed volume detector assembly continuously monitors the discharge line upstream of the discharge isolation valve. The high radioactivity setpoint will close the discharge isolation valve automatically to terminate the release of radioactive fluid. This discharge is normally not radioactive and would remain so unless a primary to secondary steam generator tube leak would occur. Such a tube leak and resultant radioactivity release from the primary system would first be detected in the steam generator liquid radiation monitor (SJ-RE-02) steam generator blowdown process radiation monitor (BM-RE-25) steam generator discharge radiation monitor (BM-RE-52) and/or condenser air discharge monitor (GE-RE-92).
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11.5.2.3 Airborne Monitoring Systems
11.5.2.3.1 Selection Criteria for Airborne Monitors
11.5.2.3.1.1 Introduction
The type of fixed instrumentation used for monitoring airborne radioactivity is offline. The offline system extracts a sample from the process stream and transports that sample to the radioactivity monitoring system, which contains the specified equipment to detect particulates, halogens, and/or noble gases.
11.5.2.3.1.2 Sampling Criteria
The sampling system for the particulate/halogen/noble gas monitors is designed and installed to meet the intent of ANSI N13.1-1969 . Systems whose sensitivity is dependent upon sample flow employ isokinetic nozzles and suitable control of flow rate.
11.5.2.3.1.3 Detection Criteria
Since both radioactive particulates and radioactive noble gases are beta emitters, beta sensitive scintillation detectors are used to sense radioactivity in order to minimize the effects due to background radiation and, consequently, obtain a lower minimum detectable concentration.
Where spectrometric analysis is required (such as in iodine monitoring) an NaI(Tl), gamma scintillation detector assembly is employed.
11.5.2.3.1.4 Instrumentation Criteria
Instrumentation necessary to indicate, alarm, and perform control functions is provided to complete the monitoring system.
Since radioactive concentrations may vary substantially, wide range instruments are utilized. All airborne radiation monitors include provisions for obtaining a grab sample for laboratory isotopic analysis. The particulate and charcoal filters can readily be removed for periodic isotopic laboratory analyses, as required by the Technical Specifications.
The airborne particulate monitors each consist of a fixed filter upon which radioactive particulate matter is deposited. The fixed filter is located in front of a beta scintillation detector coupled to a photomultiplier tube.
Each airborne iodine monitor consists of a charcoal cartridge upon which iodine is adsorbed. The air sample is prefiltered to remove particulates. The charcoal cartridge is located in front of a gamma scintillation detector coupled to a photomultiplier tube.
11.5-12 Rev. 4 WOLF CREEK
Each airborne noble gas monitor consists of a fixed-volume sample chamber through which prefiltered sample air is passed. A beta scintillation detector is located within the sample chamber to detect the activity level of the air sample.
All of the detectors and sample chambers are enclosed in heavily shielded lead pigs. Two motor-operated valves operated locally are provided to permit air-purging of the sample chamber to facilitate background activity checks.
The sensitivities and alarm setpoints are given in Tables 11.5-3 and 11.5-4.
The alert-alarm points are based on the methodologies presented in the ODCM.
11.5.2.3.2 Airborne Process Radioactivity Monitors
A detailed listing of airborne process monitor parameters is given in Table 11.5-3.
11.5.2.3.2.1 Condenser Air Discharge Monitor
The condenser air discharge monitor, 0-GE-RE-92, is provided to detect, indicate, and alarm gaseous activity in the condenser air removal system exhaust. The condenser air discharge monitor closes the steam generator blowdown isolation valves on high radiation to prevent discharge of radioactive fluid and to limit radioactive contamination of the blowdown demineralizers.
The monitor is also equipped with particulate and iodine filters which are removed and analyzed in the laboratory. This monitor provides backup to the steam generator liquid and the steam generator blowdown processing radiation monitors for detection of primary-to-secondary leaks in the steam generator.
The condenser air removal system removes noncondensable gases which would be present if a primary-to-secondary leak occurred. Particulate and iodines would also be removed by entrainment in the air discharged.
The monitor is provided with a nozzle to extract a representative sample from the exhaust duct. A sample cooler is provided to dry the sample prior to entering the sample filters or the fixed-volume gaseous detector assembly to preclude damage to the filters or to the detector. The sample point is located upstream of the condenser air removal system filters.
The radiation alarms alert the operator to the presence of gaseous activity and the possibility of steam generator tube leakage.
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11.5.2.3.2.2 Containment Atmosphere Radioactivity Monitors
The containment atmosphere radioactivity monitors, 0-GT-RE-31 and 0-GT-RE-32, continuously monitor the containment atmosphere for particulate, iodine, and gaseous radioactivity. They isolate the containment purge system on high gaseous activity via the ESFAS. See Sections 7.3.2 and 9.4.6 for further discussion of this function. These monitors also serve for reactor coolant pressure boundary leakage detection (See Section 5.2.5 for a detailed description of this function) and for personnel protection (see Section 12.3.4 for a detailed description of this function). The containment atmosphere radioactivity monitors provide backup indication for the containment purge monitors. These seismic Category I monitors are completely redundant.
Samples are extracted from the operating deck level (El. 2047'-6") through sample lines which penetrate the containment. The monitors are located as close as possible to the containment penetrations to minimize the length of the sample tubing and the effects of sample plate out. The sample points are located in areas which ensure that representative samples are obtained. Each sample passes through the penetration, then through the fixed filter (particulate), charcoal filter (iodine), and fixed-volume gaseous detector assemblies. After passing through the pumping system, the sample is discharged back to the containment through a separate penetration.
Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the radioactivity monitoring system CRT in the control room.
11.5.2.3.2.3 Containment Purge System Radioactivity Monitors
The containment purge system radioactivity monitors, 0-GT-RE-22 and 0-GT-RE-33, continuously monitor the containment purge exhaust duct during purge operations for particulate, iodine, and gaseous radioactivity. The purpose of these monitors is to isolate the containment purge system on high gaseous activity via the ESFAS. See Sections 7.3.2 and 9.4.6 for additional information concerning this function. These monitors also serve as backup indication for personnel protection (see Section 12.3.4) and reactor coolant pressure boundary leakage detection (see Section 5.2.5) for the containment atmosphere radioactivity monitors.
These seismic Category I monitors are completely redundant.
The sample points are located outside the containment between the containment isolation dampers and the containment purge filter adsorber unit.
11.5-14 Rev. 0 WOLF CREEK
Each monitor is provided with two isokinetic nozzles to ensure that representative samples are obtained for both normal purge and minipurge flow rates. Isokinetic nozzle selection is accomplished by sample selector valves which automatically align the correct nozzle to the monitor based on operation of the minipurge and normal purge exhaust systems. The sample is extracted through the selected nozzle and then passed through the selector valve, the fixed filter (particulate), charcoal filter (iodine), and fixed-volume gaseous detectors. The sample then passes through the pumping system and is discharged back to the duct.
Indication is provided for each monitor on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the radioactivity monitoring system CRT in the control room.
11.5.2.3.2.4 Containment High Range Radiation Monitors
The containment digital high range radiation monitor (DHRRM) system includes two redundant monitors, 0-GT-RE-59 and 0-GT-RE-60, to detect and indicate radiation levels in the containment over a range from 30 rads/hr to 108 rads/hr. The DHRRM also provides an alarm function.
Each DHRRM subsystem consists of a gamma radiation detector, a microprocessor, junction box, and control/display module. The subsystems are safety related and designed and qualified to IEEE 323-1974 for the normal and accident environments for their installed locations. The subsystems are also designed and qualified to be seismic Category I. The detector locations are indicated on Figure 12.3-2, Sheet 4. Detectors are mounted on the inside surface of the containment wall at El. 2052'-0" for GT-RE-60 and at El. 2073'-0" for GT-RE-59.
The DHRRM subsystems are also connected to the process and effluent radiation monitoring system (optically isolated) for readout on the CRT (SPO-56A) in the control room.
11.5.2.3.2.5 Auxiliary/Fuel Building Ventilation Exhaust Radioactivity Monitor
The Auxiliary/Fuel building ventilation exhaust radiation monitors 0-GG-RE-27 and 0-GG-RE-28, continuously monitor for particulate, iodine, and gaseous radioactivity in the Auxiliary/Fuel building ventilation exhaust system. In the event of a fuel handling accident, these monitors function to isolate the normal ventilation and start up the emergency ventilation system on high gaseous activity via the ESFAS. Sections 7.3.3 and 9.4.2 have additional information about this function. These monitors have an additional function to alert workers to high airborne radioactivity in the fuel building. This latter function is discussed in Section 12.3.4.
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These seismic Category I monitors are completely redundant.
During normal operation, each monitor extracts a sample from the normal exhaust duct through individual isokinetic nozzles and sample selector valves. This normal sample point is upstream of the fuel building normal exhaust filter adsorber unit.
When the emergency ventilation system is in use, the capability is provided from the control room to transfer the sample points via sample selector valves to isokinetic nozzles located in the fuel building emergency exhaust system upstream of the emergency exhaust filter adsorber units, with one monitor aligned to each emergency exhaust duct.
Indication is provided by individual indicators on the radioactivity monitoring system control panel and, through isolated signals, by the radioactivity monitoring system CRT in the control room.
11.5.2.3.2.6 Control Room Ventilation Radioactivity Monitor
The control room ventilation radioactivity monitors, 0-GK-RE-04 and 0-GK-RE-05, continuously monitor the supply air of the normal heating, ventilation, and air-conditioning system for particulate, iodine, and gaseous radioactivity to provide protection for the control room operators. These monitors function automatically to switch the control room from the normal to the emergency ventilation system on high gaseous activity via the ESFAS. See Sections 6.4, 7.3.4, and 9.4.1 for more details. These monitors also function to alert the operators to high airborne radioactivity in the control room ventilation supply. This function is described in Section 12.3.4.
These seismic Category I monitors are completely redundant.
Samples are extracted through individual isokinetic nozzles, and flow through the fixed filter (particulate), charcoal filter (iodine), and fixed-volume gaseous detector assemblies prior to passing through the pumping system for discharge.
Indication for these monitors is provided on individual indicators on the radioactivity monitoring system control panel and, through isolated signals, on the radioactivity monitoring system CRT in the control room.
11.5.2.3.3 Airborne Effluent Radioactivity Monitors
A detailed listing of airborne effluent monitor parameters is given in Table 11.5-4.
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11.5.2.3.3.1 Unit Vent Radioactivity Monitor
The unit vent radioactivity monitor, 0-GT-RE-21, continuously monitors the effluent from the unit vent for particulate, iodine (halogen), and gaseous radioactivity. The unit vent, via ventilation exhaust systems, continuously purges various tanks and sumps normally containing low-level radioactive aerated liquids that can potentially generate airborne activity.
The exhaust systems which supply air to the unit vent are from the fuel building, auxiliary building, the access control area, the containment purge, and the condenser air discharge.
All of these systems are filtered before they exhaust to the unit vent. The unit vent monitor measures actual plant effluents and not inplant concentrations. Thus, the system continuously monitors downstream of the last point of potential radioactivity entry. The monitoring system consists of an off-line, three-way airborne radioactivity monitor. An isokinetic sampling probe is located downstream of the last point of potential radioactivity entry for sample collection.
The Alert alarms are set below the High alarms to act as precautionary warnings. The High alarm is set to ensure that Technical Specification limits are not exceeded. (See Offsite Dose Calculation Manual.) Refer to Table 11.5-4 for the alert and high alarm setpoints, the range, and the sensitivity.
Portions of the sample tubing located outside the building are adequately protected and routed to prevent the accumulation and freezing of condensate.
The sample extracted by the isokinetic nozzle is passed through the fixed filter (particulate), charcoal filter (iodine), and fixed-volume (gaseous) detector assemblies and then through the pumping system for discharge back to the unit vent.
Indication is provided on the radioactivity monitoring system CRT in the control room. This monitor provides a signal to the radioactive release report generation system described in Section 11.5.2.1.1.
11.5.2.3.3.2 Radwaste Building Ventilation Effluent Radioactivity Monitor
The radwaste building ventilation effluent radiation monitor, 0-GH-RE-10, continuously monitors for particulate, halogen, and gaseous radioactivity in the effluent duct downstream of the exhaust filter and fans. The sample point is located downstream of the last possible point of radioactive influent, including the
11.5-17 Rev. 0 WOLF CREEK
waste gas decay tank discharge line. The flow path provides ventilation exhaust for all parts of the building structure and components within the building and provides a discharge path for the waste gas decay tank release line. These components represent potential sources for the release of gaseous and air particulate and iodine activities in addition to the drainage sumps, tanks, and equipment purged by the waste processing system.
The monitoring system consists of a fixed filter particulate monitor, an iodine monitor, and gaseous activity monitor.
The sample is extracted through an isokinetic nozzle to ensure that a representative sample of the air is obtained prior to release to the environment. After passing through the fixed filter (particulate), charcoal filter (halogen), and fixed-volume (noble gas) detector assemblies and the pumping system, the sample is discharged back to the exhaust duct.
The sensitivities and alarm setpoints are given in Table 11.5-4. The Alert alarm is set below the High alarm to act as a precautionary warning. The High alarm is set to ensure that Technical Specification limits are not exceeded.
(See Offsite Dose Calculation Manual.)
Indication of this monitor is provided on the radiation monitoring system CRT in the control room. This monitor provides a signal to the NPIS computer in the TSC computer room, (see Section 11.5.2.1.1).
This monitor isolates the waste gas decay tank discharge line if the radioactivity release rate is above the preset limit when the waste gas discharge valve has been deliberately or inadvertently opened.
11.5.2.4 Safety Evaluation
The control room ventilation monitors, the containment atmosphere monitors, the containment purge monitors, the containment LOCA atmosphere monitors, and the fuel building exhaust monitors are redundant, independent, seismic Category I, with Class IE power supplies. The control room and fuel building monitors will automatically switch from the normal to the emergency ventilation systems on high gaseous activity via the ESFAS. The containment atmosphere and containment purge monitors will automatically isolate the containment purge and stop the fans on high gaseous activity via the ESFAS.
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11.5.3 EFFLUENT MONITORING AND SAMPLING
All potentially radioactive effluent discharge paths are continuously monitored for gross radiation level. Liquid releases are monitored for gross gamma.
Airborne releases are monitored for gross beta activity (particulates and noble gases) and gross gamma (iodines).
An isotopic analysis is performed on samples obtained from each continuous effluent release path and per batch for each batch type effluent release path in order to verify the adequacy of effluent processing to meet the discharge limits to unrestricted areas. This effluent sampling program is of such a comprehensive nature as to provide the information for the effluent measuring and reporting programs required by 10 CFR 50 Part 36A and Appendix I and Regulatory Guide 1.21 in annual reports to the NRC. The effluent release data is compiled and the annual effluent report is generated.
By a combination of the installed equipment described previously in Section 11.5 and the installed equipment described in Section 12.3.4, along with portable equipment described in Section 12.5, and the Emergency Plan, the requirements of General Design Criterion 64 to monitor normal operations, anticipated operational occurrences, and postulated accidents are met.
11.5.4 PROCESS MONITORING AND SAMPLING
All potentially significant radioactive systems which lead to effluent discharge paths are equipped with a control system to automatically isolate the discharge on indication of a high radioactivity level. These include the containment purge system, the fuel building ventilation system, and the gaseous and liquid radwaste systems. Batch releases are sampled and analyzed prior to discharge, in addition to the continuous effluent monitoring.
By means of the continuous radioactivity monitors mentioned above and their associated control valves, and due to the extensive sampling program described in the Environmental Report, General Design Criterion 60 and the Radiological Effluent Technical Specifications are met with regard to the control of releases of radioactivity to the environment.
Process monitoring is accomplished by continuous radioactivity monitors discussed in Sections 11.5.2.2.2 and 11.5.2.3.2. By means of the continuous radioactivity monitors, GDC-63 is met with regard to monitoring radioactivity levels in the radioactive waste process systems.
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WOLF CREEK
TABLE 11.5-5
POWER SUPPLIES FOR PROCESS AND EFFLUENT MONITORS
Liquid Process Radioactivity Monitors (non-IE)
Normal Restored After Monitor Name Power Loss of Offsite and Number Supply Power
Component cooling water Non-IE MCCs No 0-EG-RE-9 0-EG-RE-10
Steam generator Non-IE MCCS No liquid radioactivity 0-SJ-RE-2
Steam generator Non-IE MCCs No blowdown processing system 0-BM-RE-25
Boron recycle Non-IE MCCs No system distillate 0-HE-RE-16
CVCS letdown Non-IE MCCs No 0-SJ-RE-01
Auxiliary steam Non-IE MCCs No system liquid condensate recovery 0-FB-RE-50
Rev. 8 WOLF CREEK
TABLE 11.5-5 (Sheet 2)
Liquid Effluent Radioactivity Monitors (Non-IE)
Normal Restored After Monitor Name Power Loss of Offsite and Number Supply Power Secondary liquid Non-IE MCCS No waste system 0-HF-RE-45 Wastewater treatment Non-IE MCCS No system influent 1-HF-RE-95 Liquid radwaste Non-IE MCCs No discharge 0-HB-RE-18 Turbine building Non-IE MCCs No drain 0-LE-RE-59 Steam generator Non-IE MCCs No blowdown discharge 0-BM-RE-52 Airborne Process Radioactivity Monitors (Class IE)
Containment Class IE MCCs Yes atmosphere 0-GT-RE-31 0-GT-RE-32 Containment Class IE MCCs Yes purge system 0-GT-RE-22 0-GT-RE-33 Containment high Class IE MCCs Yes activity monitors 0-GT-RE-59 0-GT-RE-60 Fuel building Class IE MCCs Yes exhaust 0-GG-RE-27 0-GG-RE-28 Control room Class IE MCCs Yes air supply 0-GK-RE-04 0-GK-RE-05 Rev. 4 WOLF CREEK
TABLE 11.5-5 (Sheet 3)
Airborne Process Radioactivity Monitor (Non-IE)
Normal Restored After Monitor Name Power Loss of Offsite and Number Supply Power
Condenser air Non-IE MCC No discharge 0-GE-RE-92
Airborne Effluent Radioactivity Monitors (Non-IE)
Plant unit Non-IE MCCs No vent 0-GT-RE-21
Radwaste building Non-IE MCCs No exhaust 0-GH-RE-10
Rev. 0