ML12011A165

Amendment 61 to Final Safety Analysis Report, Chapter 11, Radioactive Waste Management

ML12011A165
Person / Time
Site:	Columbia
Issue date:	12/14/2011
From:	Energy Northwest
To:	Office of Nuclear Reactor Regulation
References
GO2-11-201
Download: ML12011A165 (179)

Text

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

TABLE OF CONTENTS

Section Page LDC N-0 2-0 0 5 11-i 11.1 SOURCE TERMS........................................................................11.

1-1 11.1.1 FISSION PRODUCTS.................................................................11.1-1 11.1.1.1 Noble Radionuclide Fission Products............................................11.1-1 11.1.1.2 Radiohal ogen Fission Products....................................................11.1-4 11.1.1.3 Other Fission Products..............................................................11.1-6 11.1.1.4 Nomenclature.........................................................................11.1-6 11.1.2 ACTIVATIO N PRODUCTS...........................................................11.1-7 11.1.2.1 Coolant Activation Products.......................................................11.1-7 11.1.2.2 Noncoolant Activation Products...................................................11.1-7 11.1.2.3 Steam and Power Conversion System N-16 Inventory.........................11.1-8 11.1.3 TRITIUM.................................................................................

11.1-8 11.1.4 FUEL FISSION PRODUCT INVENTORY AND FUEL EXPERIENCE.......11.1-11 11.1.4.1 Fuel Fission Product Inventory....................................................11.

1-11 11.1.4.2 Fuel Experience......................................................................11.

1-11 11.1.5 RADIOACTIVITY LEAKAGE AND EFFLUENT SOURCES....................

11.1-11 11.

1.6 REFERENCES

..........................................................................

11.1-12 11.2 LIQUID WASTE MANAGEMENT SYSTEM.....................................11.2-1 11.2.1 DESIGN BASIS........................................................................11.

2-1 11.2.2 SYSTEM DE SCRIPTION............................................................11.2-3 11.2.2.1 Process Description..................................................................11.2-3 11.2.2.2 Subsys tems Description.............................................................11.2-4 11.2.2.2.1 Equipment Drain Subsystem....................................................11.2-4 11.2.2.2.2 Floor Dr ain Subsystem...........................................................11.2-5 11.2.2.2.3 Chemical Waste Subsystem......................................................11.2-6 11.2.2.2.4 Shared Equipment.................................................................11.2-7 11.2.2.2.5 Surge Capacities...................................................................11.

2-7 11.2.2.2.6 Desi gn Features....................................................................11.

2-8 11.2.3 RADIOACTIVE RELEASES........................................................11.2-9 11.2.3.1 Release Point and Dilution.........................................................11.2-9 11.2.3.2 Calculation of Rele ases of Radioactive Materials..............................11.2-9 11.2.3.3 Exposure of Persons at or Beyond the Site Boundary.........................11.2-10 11.2.3.4 Cost-Benefit Analysis...............................................................11.2-10

11.3 GASEOUS WASTE MANAGEMENT SYSTEMS................................11.3-1 11.3.1 DESIGN BASES.......................................................................11.

3-1 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

TABLE OF CONTENTS (Continued)

Section Page LDCN-03-069 11-ii 11.3.2 SYSTEM DESCRIPTION............................................................11.3-2 11.3.2.1 Main Condenser Steam Jet Air Ejector RECHAR System...................11.3-2 11.3.2.2 Other Radioactive Gas Sources....................................................11.3-6 11.3.2.3 Cost-Benefit Analysis...............................................................11.3-8 11.3.2.4 Design Features of the Offgas System...........................................11.3-8 11.3.2.4.1 Maintainability.....................................................................11.

3-8 11.3.2.4.2 Pressure Boundaries..............................................................11.3-8 11.3.2.4.3 Building Seismic Design.........................................................11.3-9 11.3.2.4.4 Cons truction of Proces s Systems...............................................11.3-9 11.3.2.4.5 Instrumenta tion and Control.....................................................

11.3-9 11.3.2.4.6 Detonati on Resistance............................................................11.

3-10 11.3.2.4.7 Operator Exposur e Criteria and Controls.....................................

11.3-10 11.3.2.4.8 Equipmen t Malfunction..........................................................11.

3-10 11.3.2.5 Offgas System Operating Procedure..............................................11.3-11 11.3.2.5.1 Prestart up Preparations...........................................................

11.3-11 11.3.2.5.2 Startup...............................................................................

11.3-11 11.3.2.5.3 Normal Operation.................................................................11.3-11 11.3.2.6 Offgas System Performance Tests................................................11.3-11 11.3.2.6.1 R ecombiner.........................................................................11.

3-11 11.3.2.6.2 Prefilter.............................................................................11.

3-12 11.3.2.6.3 Desiccan t Gas Dryer..............................................................11.3-12 11.3.2.6.4 Charcoal Performance............................................................11.

3-12 11.3.2.6.5 Post Filter...........................................................................11.

3-12 11.3.3 RADIOACTIVE RELEASES........................................................11.

3-12 11.3.3.1 Release Points........................................................................11.

3-12 11.3.3.2 Dilution Factors......................................................................11.

3-13 11.3.3.3 Estimated Releases...................................................................11.

3-13 11.

3.4 REFERENCES

.........................................................................

11.3-14 11.4 SOLID WASTE MANAGEMENT SYSTEM.......................................11.4-1 11.4.1 DESIGN BASIS........................................................................11.

4-1 11.4.2 SYSTEM DE SCRIPTION............................................................11.4-2 11.4.2.1 General................................................................................11.4-3 11.4.2.2 Radwaste Disposal Syst em For Reactor Water Cleanup Resin..............11.4-3 11.4.2.3 Radwaste Disposal System For Condensate Demineralizer Resin...........11.4-4 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

TABLE OF CONTENTS (Continued)

Section Page LDCN-04-052 11-iii 11.4.2.4 Radwaste Disposal System For Fuel Pool, Floor Drain, and Waste Collector Filter Resin...............................................................11.4-4 11.4.2.5 Radwaste Disposal System For Spent Resin....................................11.4-4 11.4.2.6 Resin Contai ner Handling and Storage...........................................11.4-4 11.4.2.7 Miscellaneous Dry Solid Waste System.........................................11.4-5 11.4.2.8 Expected Volumes...................................................................11.

4-5 11.4.2.9 Packaging.............................................................................11.4-6 11.4.2.10 Storage Facilities...................................................................11.

4-6 11.4.2.11 Shipment.............................................................................11.4-6 11.4.2.12 Process Monitoring................................................................11.4-7 11.4.3 PROCESS CO NTROL PROGRAM................................................11.4-7 11.4.3.1 Objective..............................................................................11.4-7 11.4.3.2 Process Control Program...........................................................11.4-8 11.4.3.3 Process Control Systems............................................................11.4-9 11.4.3.4 Waste Characterization.............................................................11.4-9 11.4.3.5 Processing Methods (Wet Wastes)................................................11.4-10 11.4.3.6 Control Instrume ntation and Sampling Program...............................11.4-11 11.4.3.7 Maintena nce and Calibration.......................................................11.

4-11 11.4.3.8 Waste Proc essing System Capacity...............................................11.4-12 11.4.3.9 Waste Storage Capacity.............................................................11.4-12 11.4.3.10 Compliance With ALARA Principles...........................................11.

4-12 11.4.3.11 Unanticipated Wastes..............................................................11.4-14 11.4.3.12 Waste Classification...............................................................11.4-14 11.4.3.13 Waste Packaging and Shipping...................................................11.

4-14 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS..........................................................11.5-1 11.5.1 DESIGN BASIS........................................................................11.

5-1 11.5.1.1 Design Objectives....................................................................11.

5-1 11.5.1.1.1 Systems Re quired for Safety....................................................

11.5-1 11.5.1.1.2 Systems Required for Plant Operation.........................................11.5-1 11.5.1.2 Design Criteria.......................................................................11.

5-3 11.5.1.2.1 Systems Re quired for Safety....................................................

11.5-3 11.5.1.2.2 Systems Required for Plant Operation.........................................11.5-4

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

TABLE OF CONTENTS (Continued)

Section Page LDCN-08-014 11-iv 11.5.2 SYSTEM DE SCRIPTION

............................................................ 11.5-4 11.5.2.1 Systems Requir ed for Safety ....................................................... 11.5-4 11.5.2.1.1 Main Steam Line Radiation Monitoring System ............................. 11.5-4 11.5.2.1.2 Reactor Building Exhaust Plenum Radiation Monitoring System ......... 11.5-5 11.5.2.1.3 Control Room Fresh Air Intake Radiation Monitoring System ........... 11.5-6 11.5.2.1.4 Standby Serv ice Water Radiation Monitoring System ...................... 11.5-7 11.5.2.2 Systems Required for Plant Operation ........................................... 11.5-7 11.5.2.2.1 Gaseous Process and Effluent Radiation Monitoring System .............. 11.5-8 11.5.2.2.1.1 Offgas Pretreatment Radiation Monitoring System ....................... 11.5-8 11.5.2.2.1.2 Offgas Posttreatment Radiation Monitoring System ...................... 11.5-8 11.5.2.2.1.3 Offgas Charcoal Bed Vault Radiation Monitoring System .............. 11.5-9 11.5.2.2.1.4 Mechanical Vacuum Pump Exhaust Radiation Monitoring System .... 11.5-10 11.5.2.2.1.5 Reactor Building Elevated Release Duct Radiation Monitoring System ............................................................................ 11.5-10 11.5.2.2.1.6 Turbine-Generator Building Ventilation Release Duct Radiation Monitoring System ............................................................. 11.5-13 11.5.2.2.1.7 Radwaste Building Ve ntilation Release Ducts Radiation Monitoring System ............................................................. 11.5-14 11.5.2.2.1.8 NRC Safety Evaluation Report, NUREG-0892 Acceptance ............ 11.5-14 11.5.2.2.2 Liquid Process and Effluent Radiation Monitoring System ................ 11.5-15 11.5.2.2.2.1 Standby Service Water Radiation Monitoring System .................... 11.5-16 11.5.2.2.2.2 Reactor Building Closed Cooling Water Radi ation Monitoring System ............................................................................ 11.5-16 11.5.2.2.2.3 Radwaste Effluent Radiation Mon itoring System ......................... 11.5-16 11.5.2.2.2.4 Circulating Water and Plant Service Wa ter Radiation Monitoring Systems ...........................................................................

11.5-17 11.5.2.2.3 Primary Containment Ra diation Monitoring System ........................ 11.5-17 11.5.2.2.3.1 Leak Detection Monitors ...................................................... 11.5-17 11.5.2.2.3.2 Loss-of-Cool ant Accident Tracking Radiation Monitoring Systems (Containment Drywell) ........................................................ 11.5-18 11.5.2.3 Sampling

............................................................................... 11.5-19 11.5.2.3.1 Process Sampling

.................................................................. 11.5-19 11.5.2.3.2 Effluent Sampling

................................................................. 11.5-19 11.5.2.3.3 Analytical Procedures ............................................................ 11.5-20 11.5.2.3.4 Inservice Inspection, Calibration, and Maintenance ........................ 11.5-20 11.5.3 EFFLUENT MONITORI NG AND SAMPLING ................................ 11.5-21 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

TABLE OF CONTENTS (Continued)

Section Page 11-v 11.5.4 PROCESS MONITORI NG AND SAMPLING...................................

11.5-21 11.6 POSTACCIDENT SAMPLING SYSTEM...........................................11.6-1 11.6.1 DESIGN BASIS........................................................................11.

6-1 11.6.2 SYSTEM DE SCRIPTION............................................................11.6-1 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

LIST OF TABLES

Number Page LDC N-0 2-0 0 5 11-vi 11.1-1 Noble Radiogas Source Ter m s ....................................................

11.1-13 11.1-2 Halogen Radioisotopes in Reactor Water.......................................11.1-15 11.1-3 Other Fission Product Radioiso topes in Reactor Water......................11.1-16

11.1-4 Coolant Activation Products in Reactor Water and Steam...................11.1-18

11.1-5 Noncoolant Activation Produc ts in Reactor Water.............................

11.1-19 11.2-1 Liquid Waste Management System Radioisotope I nventory Equipment Drain Subsystem.....................................................................

11.2-11 11.2-2 Liquid Waste Management System Radioisotope I nventory Chemical Waste Subsystem....................................................................

11.2-13 11.2-3 Liquid Waste Management System Radioisotope I nventory RWCU and Condensate Filter Demi neralizers................................................11.2-15

11.2-4 Liquid Waste Manageme nt System Radioisotope Inventory Spent Resin and Condensate Backwash Receiving Tanks....................................

11.2-17 11.2-5 Liquid Waste Management System Radioisotope Inventory Phase Separators.............................................................................11.

2-19 11.2-6 Annual Average Concentration of Radionuclides in Liquid Effluent.......11.2-21

11.2-7 Tank Design Features...............................................................11.2-23

11.2-8 Equipment Drain S ubsystem Sources............................................

11.2-24 11.2-9 Floor Drain Subs ystem Sources...................................................11.

2-25 11.2-10 Chemical Waste Subsystem Sources.............................................

11.2-26 11.2-11 Radwaste System Proc ess Flow Diagram Data................................

11.2-27 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

LIST OF TABLES (Continued)

Number Title Page LDC N-0 2-0 0 0, 0 2-005 11-vii 11.2-12 Radwaste Process Equipment Design Basis.....................................11.2-37 11.2-13 Liquid Radwaste Equipment.......................................................11.

2-38 11.2-14 Annual Releases of Radioactive Materi al as Liquid...........................11.2-41 11.3-1 Design Air Ejector Offgas Release Rates.......................................11.3-15

11.3-2 Offgas System Majo r Equipment Items..........................................

11.3-16 11.3-3 Process Data for the Of fgas (RECHAR) System..............................

11.3-18 11.3-4 Offgas System Alarmed Process Parameters...................................

11.3-19 11.3-5 Equipment Malfunc tion Analysis.................................................11.3-20

11.3-6 Release Point Data..................................................................11.3-24

11.3-7 Gaseous Waste System Release...................................................11.

3-25 11.3-8 Building Volume and Ventilation Rates.........................................

11.3-27 11.3-9 Maximum Sector Annual Average Concentrations of Gaseous Radioactive Materials at the Original Restricted Area Boundary...........11.3-28 11.3-10 Frequency and Quantity of Steam Discharged to Suppression Pool........11.3-29

11.4-1 Waste Processing Sy stems Capacities............................................

11.4-15 11.4-2 Solid Waste Management Syst em Major Equipment Items..................11.4-16

11.4-3 Significant Isotope Activity in Dewatered Waste..............................

11.4-18 11.4-4 Expected Annual Pro duction of Solids...........................................

11.4-21 C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

LIST OF TABLES (Continued)

Number Title Page 11-viii 11.5-1 Process and Effluent Radiation Monitoring System (Gaseous and Airborne Monitors).................................................................

11.5-23 11.5-2 Process and Effluent Radiation Monitoring System (Liquid Monitors)....11.5-27

11.5-3 Radiological Analysis Summary of Gaseous Effluent Samples..............11.5-28

11.5-4 Radiological Analysis Summary of Liquid Process Samples................11.5-29

11.5-5 Radiological Analysis Summary of Gaseous Process Samples..............11.5-30

11.5-6 Radiological Analysis Summary of Liquid Effluent Samples................11.5-31

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

LIST OF FIGURES Number Title LDC N-0 2-0 0 0, 0 2-005 11-ix 11.1-1 Noble Radiogas Decay Constant Exponent Frequency Histogram 11.1-2 Radiohalogen Decay Constant Exponent Freque ncy Histogram 11.1-3 Noble Radiogas Leakage Versus Iodine-131 Leakage

11.2-1 Radwaste System Process Diagram 11.2-2 Flow Diagram Radioactive Waste System Equipment Drain Processing

11.2-3 Flow Diagram Radioactive Waste System Floor Drain Processing

11.2-4 Flow Diagram Chemical Wast e Processing (Sheets 1 through 3) 11.3-1 Offgas System - Low Temperature

11.3-2 Offgas System P&

ID (Sheets 1 and 2)

11.4-1 Flow Diagram Radioactive Waste Disposal Solids Handling System

11.5-1 Main Steam Line Monitors

11.5-2 Reactor Building Exhaust Plenum Monitors and Charcoal Bed Vault Monitor

11.5-3 Control Room Fresh Air Intake Monitors

11.5-4 Process and Effluent Liquid Radiation Monitors 11.5-5 Offgas Pretreatment and Po sttreatment Radiation Monitors 11.5-6 Mechanical Vacuum Pumps Exhaust Monitor and Reactor Building Elevated Release Stack Monitor

11.5-7 Turbine and Radwaste Building Ventilation Release Duct Monitors

11.5-8 DELETED C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 Chapter 11 RADIOACTIVE WASTE MANAGEMENT

LIST OF FIGURES (Continued)

Number Title 11-x 11.5-9 Primary Containment Leak Detection Monitor

11.5-10 Primary Containment and Elevated Release Stack LOCA Tracking

11.5-11 Elevated Release Stack LOCA Monitoring

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-1 Chapter 11 RADIOACTIVE WA S TE MANAGEMENT

11.1 SOURCE TERMS The italicized information is historical and was provided to support the application for an operating license.

Radioactive material sources (activation products and fission product release from fuel) have been evaluated in operating boiling water reactors (BWRs) over the past decade. These source terms are reviewed and periodically revised to incorporate up-to-date information. Release of radioactive material from operatin g BWRs has resulted in doses to offsite persons which have been only a small fraction of 10 CFR 20 pe rmissible or natural background doses.

The information provided in this section defines the design-basis radioactive material levels in the reactor water, steam, and o ffgas. The various radioisot opes listed have been grouped as coolant activation products, noncoolant activation products, and fission products. The fission product levels are based on measur ements of BWR reactor water and offgas at several stations through mid-1971. Emphasis was placed on observations made at Kernkraftwerk RWE-Bayernwerk GmbH (KRB) and Dr esden-2. The design-basis radioactive material levels do not necessarily include all the radioisotopes observed or predicted theoretically to be present. The radioisotopes incl uded are considered significant to one or more of the following criteria:

a. Plant equipment design,

b. Shielding design,

c. Understanding system oper ation and performance, d. Measurement practicability, and

e. Evaluating radioactive material releases to the environment.

The inventory of radionuclides used to determ ine shielding requirement s of system components are discussed in Chapter 12.

11.1.1 FISSION PRODUCTS 11.1.1.1 Noble Radi ogas Fission Products The noble gas radionuclide fission product source terms observed in operating BWRs are generally complex mixtures whos e sources vary from miniscule defects in cladding to "tramp" uranium on external cladding su rfaces. The relative concentr ations or amounts of noble radiogas can be desc ribed as follows.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-2 Equilibrium:

R g,i = K 1 Y 1 (11.1-1)

Recoil: R g,i = K 2 Y 1 i (11.1-2)

The nomen c l ature in Section 11.1.1.4 defines t h e terms in t h ese and succeeding equations. The constants K 1 and K 2 describe the fractions of the total fiss i ons that are involved in each of the releases.

T h e equilibrium and recoil mixtures a r e the two extremes of the mixture spectrum that are physically possible. When a sufficient time delay o ccurs between the fissi o n event and the time of release of the radiogases from t h e fuel to the coolant, the radiogases app r oach equilibrium levels in the fuel and the equilibrium mixture res u lts. Where there is no delay between the fission event and the rel e ase of the radiogases, the r ecoil mixture is observed.

Prior to Vallecitos Boiling Wat e r Reactor (VBWR) and Dresden

-1 experience, it was assumed that noble radiogas leakage f r om the fuel would be the equilibrium mixture of the noble radiogases present in the fuel.

The VBWR and early Dresden-1 experience (Reference 11.1-1) indicated that the actual mixture most often observed approa c hed a distribution w h ich was i n termediate in character to the two extremes. This interme d iate decay mixture w a s termed the "diffusion" mixture. It must be emphasized that this "diff u sion" mixture is m e rely one possible point on the mixture spectrum ranging from t h e equilibrium to t h e recoil mixt u r e and does not have the absolute mathematical and mechanistic bas is for the calculational m e thods possible for equilibrium and recoil mixtures. However, the "diffusion" distribution pattern w h ich has been described is as follows:

Diffusion:

R g,i = K 3 Y i i 0.5 (11.1-3)

The constant K 3 describes the fraction of total fissions that are involved in the release. The value of the exponent of the decay constant, i, is midway between the values for the equilibrium case, 0, and recoil case 1. The "diffusion" pattern value of 0.5 was originally derived from diffusion theory.

Although the previously described "diffusion" mixture was used by GE as a basis for design since 1963, the design-basis releas e magnitude used has varied fr om 0.5 Ci/sec to 0.1 Ci/sec as measured after 30-minute decay (t=30 minutes).

• Since about 1967, the design-basis release magnitude used (incl uding the 1971 source terms) wa s established at an annual average of 0.1 Ci/sec (t = 30 minutes). This design basis is c onsidered as an annual average with some time above and some tim e below this value. This de sign value was selected on the basis of operating experience ra ther than predictive assumptions. Several judgment factors,

• The noble radiogas source term rate after 30-minute decay has been used as a conventional measure of the design-basis fuel leakage rate since it is c onveniently measurable and was consistent with the nominal design-basis 30-minute offgas hol dup system used on a number of previous plants.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-3 including the significance of envi ronmental release, reactor wate r radioisotope concentrations, liquid waste handling and effluen t disposal criteria, building air contamination, shielding design, and turbine and other component contamination affec ting maintenance, have been considered in estab lishing this level.

Noble radiogas source terms from fuel above 0.1 Ci/sec (t = 30 minutes) can be tolerated for reasonable periods of time. Continual assessmen t of these values is made on the basis of actual operating experience in BWRs (References 11.1-2 and 11.1-3). While the noble radiogas source-term magnitude was established at 0.1 Ci/sec (t = 30 minutes), it was rec ognized that there may be a more statistically applicable distribution for the noble radiogas mixture. Sufficient data were available from KRB operations from 1967 to mid-1971 along with Dresden-2 data from operation in 1970 and several months in 1971 to more a ccurately characterize the nobl e radiogas mixture pattern for an operating BWR.

The basic equation for each radioisotope used to analyze the collected data is:

RKYeieig,i g ii m Tt 1 (11.1-4)

With the exception of Kr-85 with a half-life of 10.74 years, the noble r adiogas fission products in the fuel are essentially at an equilibrium condition after an irradi ation period of several months (rate of formation is equal to the rate of decay). So for prac tical purposes the term (1 - e - i T) approaches 1 and can be neglected when th e reactor has been operating steady state for long periods of time. The term (1 - e

- i T) is used to adjust the releases from the fuel (t = 0) to the decay time for which values are needed. Histor ically, t = 30 minutes has been used. When discussing long ste ady-state operation and leakage from the fuel (t = 0), the following simplified form of Equation 11.1-4 can be used to describe the leakage of each noble radiogas:

RKYg,igii m (11.1-5)

The constant, K g , describes the magnitude of leakage from the fuel. The relative rates of leakage of the different noble r adiogas isotopes is accounted for by the variable, m, the exponent of the decay constant, i.

Dividing both sides of Equation 11.1-5 by y i , the fission yield, and taki ng the logarithm of both sides results in the following equation:

log/logRYm log Kg,iii g (11.1-6)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-4 Equation 11.1-6 represents a straight line when log (R g,i/y i) is plotted vs. log ( i); m is the slope of the line. This strai ght line is obtained by plotting (R g,i/y i) vs. ( i) on logarithmic graph paper. By fitting ac tual data from KRB a nd Dresden-2 (using leas t squares techniques) to the equation, the slope, m, c an be obtained. This can be estimated on the plotted graph.

With radiogas leakage at KRB over the nearly 5-year period va rying from 0.001 to 0.056 Ci/sec (t = 30 minutes) and with radiogas le akage at Dresden-2 va rying from 0.001 to 0.169 Ci/sec (t = 30 minutes), the average value of m was de termined. The va lue for m is 0.4 with a standard deviation of +/-0.07. This is illustrated in Figure 11.1-1 as a frequency histogram. As can be seen from this figure, variations in m we re observed in the ra nge of m = 0.1 to m = 0.6. After establishing the value of m = 0.4, the value of K g can be calculated by selecting a value for R g , or as has been done historic ally, the design basis is set by the total design-basis source-term magnitude at t = 30 minutes. With SR g at 30 minutes = 100,000 mCi/sec, K g can be calculated as being 2.6 x 10 7 and Equation 11.1-4 becomes: RxYeieig,iii tt26101 7. 0.4 (11.1-7)

This updated noble radiogas s ource-term mixture has been te rmed the "1971 mixture" to differentiate it from th e "diffusion" mixture. The noble gas s ource term for each radioisotope can be calculated from Equation 11.1-7.

The resultant source terms are presented in Table 11.1-1 as leakage from fuel (t

= 0) and after 30-minute deca

y. While Kr-85 can be calculated using Equation 11.1-7, the number of confirming experimental observations was limited by the difficulty of measuring very low release rates of this isotope. Therefore, the table provides an estimated range for 85Kr based on a few actual measurements.

11.1.1.2 Radiohalogen Fission Products Historically, the radioha logen design-basis source term was establis hed by the same equation as that used for noble radiogases. In a fashion similar to t hat used with gases, a simplified equation can be shown to describe the release of each radiohalogen:

RKYhhii n ,i (11.1-8)

The constant, K h , describes the magnitude of leakage from fuel. The relative rates of radiohalogen leakage is expressed in terms of n, the exponent of the decay constant, i. As was done with the noble radiogases, th e average value was determined for n.

The value for n is 0.5 with a standard deviation of

+/-0.19. This is illustrated in Figure 11.1-2 as a frequency histogram. As can be seen from this figure, variations in n were obs erved in the range of n = 0.1 to n = 0.9.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-5 It appeared that the use of the previous method of calc ulating radiohalogen leakage from fuel was overly conservative.

Figure 11.1-3 relates KRB and Dresde n-2 noble radiogas versus I-131 leakage. While it can be seen from Dr esden-2 data, during the period August 1970 to January 1971, that there is a relationship between noble radiogas and I-131 leakage under one fuel condition, there was no simp le relationship for all fuel c onditions experienced. Also, it can be seen that during this period, high radiogas leakages were not accompanied by high radioiodine leakage from the fuel. Except for one KRB datum point, all steady-state I-131 leakages observed at KRB or Dresden-2 were equal to or less than 505 Ci/sec. Even at Dresden-1 in March 1965, when severe defects were experienced in stainless steel clad fuel, I-131 leakages greater than 500 Ci/sec were not experienced.

Figure 11.1-3 shows that these higher radioiodine leakages from the fuel were related to noble radiogas source terms of less than the design-basis value of 0.1 Ci/sec (t = 30 minutes). This may be partially explained by inherent limitations due to inte rnal plant operational problems that caused plant derating.

In general, it would not be anticipated that operation at full power would continue for any significant time period with fuel cladding defects, which woul d be indicated by 131 I leakage from the fuel in excess of 700 Ci/sec. When high radiohaloge n leakages are observed, other fission products will be pres ent in greater amounts.

Using these judgment factors and experience to date, the design-basis radiohalogen source terms from fuel were established based on 131 I leakage of 700 Ci/sec. This value, as seen in Figure 11.1-3 , accommodates the experience data and the design-basis noble radiogas source term of 0.1 Ci/sec (t = 30 mi nutes). With the I-131 design-bas is source-term established, K h can be calculated as being 2.4 x 10 7 and radiohalogen release c an be expressed by the following equation:

RxYeieihii Tt ,i..24101 7 05 (11.1-9) Concentrations of radiohaloge ns in reactor water can be calculated using the following equation:

C R M h h i ,i ,i (11.1-10) The observed "carryover" for r adiohalogens has varied from 0.1% to about 2% on newer plants. The average of obser ved radiohalogen carryover meas urements has been 1.2% by weight of reactor water in stea m with a standard deviation of +/-0.9%. In the present source term definition, a radiohalogen carryover of 2% (0.02 fraction) was used.

The radiohalogen release rate from the fuel was calculated from equation 11.1-9.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-6 Concentrations in reactor water were calculated from equation 11.1-10.

The resultant concentrations are presented in Table 11.1-2. Radiohalogens with half-lives less than 3 minutes were omitted.

11.1.1.3 Other Fission Products The observations of other fission products (and transuranic nuclides, including 239 Np) in operating BWRs are not adequately correlated by simple equations. For these radioisotopes, design-basis concentrations in reactor water have been estimated conservatively from experience data (Reference 11.1-8) and are presented in Table 11.1-3. Radioisotopes with half-lives less than 10 minutes were not considered. Carryover of these radioisotopes from the reactor water to the stea m is estimated to be 0.1% ( 0.001 fraction) (Reference 11.1-8). In addition to carryover, however, deca y of noble radiogases in the st eam leaving the reactor will result in production of noble gas daughter radioisotopes in the steam and condensate systems.

Some daughter radioisotope s (e.g., yttrium and lant hanum) were not listed as being in reactor water. Their independent leak age to the coolant is negligib le; however, these radioisotopes may be observed in some samples in equilibri um or approaching equilibrium with the parent radioisotope.

Except for 239 Np, trace concentrations of transuranic isotopes have b een observed in only a few samples where extensive and complex analyses were carried out. The predominant alpha emitter present in reactor water is 242Cm at an estimated concentration of 10

-6 Ci/g or less, which is below the maximum permissible c oncentration in drinking water applicable to continuous use by the general public. The concentration of al pha-emitting plutonium radioisotopes is more than one order of magnitu de lower than that of 242 Cm.

Plutonium-241 (a beta emitter) may also be present in conc entrations comparable to the 242 Cm level.

11.1.1.4 Nomenclature The following list defines the terms used in equations fo r source term calculations:

R g,i = leakage rate of nobl e gas radioisotope i (Ci/sec)

R h,i = leakage rate of a halogen radioisotope i ( Ci/sec) Y i = fission yield of a radioi sotope i (atoms/fission)

I = decay constant of a radioisotope i (sec

-1)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-7 T = fuel irradiation time (sec)

t = decay time following leakage from fuel (sec)

m = noble radiogas decay con s tant exponent (dimensionless)

n = radioha l ogen decay cons t ant exponent (dimensionless)

K g = a constant establishing the level of noble radiogas leakage from fuel

K h = a constant establishing the level of radiohalogen leakage from fuel C h,i = concentration of a radiohal ogen i in reactor water ( Ci/g) M = mass of water in the operating reactor (g)

= cleanup system removal constant (se c-1)

= cleanup system flow rate (g/sec)

M(g) g = grams mass

= halogen steam carryover removal constant (sec

-1) = concentration of rad Ci g xiohalogen isotope in steam (Ci/g)

C steam f l ow (g / s ec)M (g)h , i (/) 11.1.2 ACTIVATI ON PRODUCTS

11.1.2.1 Coolant Ac tivation Products The coolant activation products are not adequatel y correlated by simp le equations. Design basis concentrations in reactor water and st eam have been estimated conservatively from experience data. The resultant c oncentrations are presented in Table 11.1-4.

11.1.2.2 Noncoolant Activation Products The activation products formed by activation of impurities in th e coolant or by corrosion of irradiated system material s are not adequately correlated by simple equations. The design-basis source terms of noncool ant activation products have b een estimated conservatively C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-8 from experience data (Reference 11.1-8). The resultant concentr ations are presented in Table 11.1

-5. Carryover of thes e isotopes fr o m the react o r water to t h e steam is estimated to be <0.1% (<0.001 fraction) (Reference 11.1-8). 11.1.2.3 Steam and Power Conv ersion System N-16 Inventory The main steam and reactor feedwater syst ems sources are discussed in Section 12.2.1.2.2.7. This section discusses the N-16 source strength in the moisture separators and reheaters, main condenser and hotwell, feedwater heaters, and associated piping.

11.1.3 TRITIUM

In a BWR, tritium is produced by three principal methods:

a. Activation of naturally occurri ng deuterium in the primary coolant, b. Nuclear fission of UO 2 fuel, and c. Neutron reactions w ith boron used in reac tivity control rods.

The tritium formed in control r ods, which may be released from a BWR in liquid or gaseous effluents, is believed to be negligible. The prime source of tritiu m available for release from a BWR is that produced from activation of de uterium in the primary coolant. Some fission product tritium may also transfer from fuel to prim ary coolant. This discus sion is limited to the uncertainties associated with estimating the amounts of tritium generated in a BWR which are available for release. All of the tritium produced by activati on of deuterium in the primary coolant is available for release in liquid or gaseous effluents.

The tritium formed in a BWR from deuterium activation is calculated using the equation R V xP ac t 37 1 0 4.(11.1-11) where:

R act = tritium formation rate by deuterium activation act (Ci/sec/MWt)

= macroscopic thermal neutron cross section (cm

-1)

= thermal neutron flux (neutrons/cm 2-sec) V = coolant volume in core (cm

3)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-9 = tritium radioactive decay constant (1.78 x 1 0-9 se c-1) P = reactor power level (MWt)

For recent BWR designs, R act is calculated to be (1.3 +/- 0.4) x 10

-4 Ci/sec/MWt. The uncertainty indicated is derived from the estimated errors in selecting values for the coolant volume in the core, coolant de nsity in the core, abundance of deuterium in light water [some additional deuterium will be pr esent because of the H(n,) D reaction], thermal neutron flux, and microscopic cross s ection for deuterium.

The fraction of tritium produced by fission which may transfer from the fuel to the coolant (which will then be available for release in li quid and gaseous effluents) is more difficult to estimate. However, since zircaloy-clad fuel rods are used in BWRs, essentially all fission product tritium will remain in the fuel rods unle ss defects are present in the cladding material (Reference 11.1-4).

The study made at Dresden-1 in 1968 by the U.S. Public Health Service (USPHS) suggests that essentially all of the tr itium released from the plant could be accounted for by the deuterium activation source (Reference 11.1-3). For purposes of estimating the leakage of tritium from defective fuel, it is assumed that it leaks in a manner similar to the leakage of noble radiogases. Thus, use is made of the empi rical relationship desc ribed as the "diffusion" mixture used for predicting the source term of individual noble radiogas isotopes as a function of the total noble gas source term. The equa tion which describes th is relationship is:

RKydif 05. (11.1-12)

where:

R dif = leakage rate of tritium from fuel (Ci/sec) y = fission yield fraction (atoms/fission)

= radioactive decay constant (sec

-1) K = a constant related to total tritium leakage rate.

When the total noble radiogas source term is 100,000 Ci/sec after 30-minutes decay, leakage from fuel is calculat ed to be about 0.24 Ci/sec of tritium. To pla ce this value in perspective, in the USPHS study (Reference 11.1-3), the observed rate of 85 Kr (which has a half-life similar to that of tritium) was 0.06 to 0.4 times that calculated using the "diffusion" mixture relationship. This would suggest that the actual tritium leakage rate might range from 0.015 to 0.10 Ci/sec. Since the annual a verage noble radiogas leakage from a BWR is expected to C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-10 be less than 100,000 Ci/sec (t = 30-minutes), the annual a verage tritium release rate from the fission source is conservativ ely estimated at 0.12 +/- 0.12 Ci/sec, or 0.0 to 0.24 Ci/sec. Based on this approach, the estimated total tr itium appearance rate in reactor coolant and release rate in the efflu ent is about 19 Ci/yr.

Tritium formed in the reactor is present as tritiated oxide (HTO) and to a lesser degree as tritiated gas (HT). Tritium concentration on a weight basis in th e steam formed in the reactor is the same as in the reactor water at any given time. This tritiu m concentration is also present in condensate and feedwater. Si nce radioactive effluents generall y originate from the reactor and power cycle equipment, radioac tive effluents also have this tritium concentration. The condensate storage tanks recei ve treated water from the liqui d waste management systems and rejected water from the condensate system.

Thus, all plant process water approaches a common tritium concentration.

Offgases released from the pl ant contain tritium, which is pr esent as tritiated gas (HT) resulting from reactor water radiolysis as well as tritiated water vapor (HTO). In addition, water vapor from the turbine gland seal steam packing exhauster and some water vapor present in ventilation air due to process steam leaks and evaporation from sumps and tanks also contain tritium. The remainder of the trit ium leaves the plant in liquid effluents or with solid wastes.

Recombination of radiolysis gases in the off gas system forms water, which is condensed and returned to the main condenser. This tends to reduce the amount of tritium leaving in gaseous effluents. Reducing the gaseous tritium rele ase results in a sli ghtly higher tritium concentration in the plant process water. Re ducing the amount of li quid effluent discharged will also result in a higher proc ess coolant tritium concentration.

Essentially all tritium in the primary coolant is eventually released to the environs, either as water vapor and gas to the atmosphere or as liquid effluent to th e plant discharge or as solid waste. Reduction due to radioactive decay is negligible due to the 12.3-year half-life of tritium.

The USPHS study at Dresden-1 es timated that approximately 90%

of the tritium release was observed in liquid effluent, with the re maining 10% leaving as gaseous effluent (Reference 11.1-5).

Efforts to reduce the volume of liquid effluent discharges may change this distribution so that a greater amount of tritium will leave as gaseous effluent. From a practical standpoint, the fraction of tritium leaving as liquid effluent may vary between 60% and 90% with the remainder leaving in gaseous effluent.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 1 7 11.1-11 The amount of tritium released to the environment in liquid and gaseous effluents is based on the draft Regulatory Guide 1.CC, BWR-GALE code analyses. This is discussed in Sections 11.2 and 11.3 respectively.

11.1.4 FUEL FISSION PROD U CT INVE N TORY A N D FUEL EXPERIENCE

11.1.4.1 Fuel Fission Product Inventory Fuel fission product inventory in formation is used in estab lishing fission product source terms for accident analysis and is, therefore, discussed in Chapter 15.

11.1.4.2 Fuel Experience A discussion of BWR fuel experienced including fuel failure, burn-up and thermal conditions under which the experience was gained, is presented in References 11.1-2 , 11.1-3 , and 11.1-6.

11.1.5 RADIOACTIVITY LEAKAGE AND EFFLUENT SOURCES

Process leakage results in pot ential release paths for noble ga ses and other volatile fission products via ventilation systems.

Liquids from process leaks are collected and routed to radioactive equipment and floor drain systems. Radioisotope releases via ventilation paths are at extremely low levels and have been insignificant compared to process offgas from operating BWR plants. However, because the implementation of improved process offgas treatment systems makes the ventilation release compara tively significant, meas urements have been conducted to identify and quantify these low-level release paths.

In addition an awareness of measurements by the Electric Power Research Institute, other organizations, and routine measurements by other utilities with operati ng BWRs has been maintained. Design basis estimates of the various liquid, gaseous, and solids effluents are discussed in Sections 11.2 , 11.3 , and 11.4 which follow.

Concurrently, analytical and mathematical model studies are being performed to provide a description of the transport, residence, and re lease of various radio nuclides in and from an operating BWR.

Process leakage measurement, detection, and control methods are further discussed in Sections 5.2.5 , 11.2.2 , 11.3.2 , and 12.1.2.

The effect of process leakage sources on the in-plant airborne radionuclide concentrations and the adequacy of plant ventilation systems is discussed in Section 12.2.2. Liquid radioactive sources are discus sed in Section 11.2 and gaseous radioactive s ources are discussed in Section 11.3.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 11.1-12 11.1.6 REFERENC E S 11.1-1 Brutschy, F. J., "A comparison of Fission Product Re lease Studies in Loops and VBWR," Paper presented at the Tripartite Conference on Transport of Materials in Water Systems, Chalk Ri ver, Canada (February 1961).

11.1-2 Williamson, H. E., Ditmore, D. C., "Experience with BWR Fuel Through September 1971," NEDO-10505, May 1972. (Update) 11.1-3 Elkins, R. B., "Experience w ith BWR Fuel Through September 1974,"

NEDO-20922, June 1975.

11.1-4 Ray, J. W., "Tritium in Power Reactors," Reactor and Fuel Processing Technology, 12 (1), pp. 19-26, Winter 1968-1969.

11.1-5 Kahn, B., et al, "Radiological Surve illance Studies at a Boiling Water Nuclear Power Reactor," BRH/DER 70-1, March 1970.

11.1-6 Williamson, H. E., Ditmore, D. C., "Current State of Knowledge of High Performance BWR Zircaloy Clad UO Fuel," NEDO-10173, May 1970.

11.1-7 Marrero, T. R., "Airborne Releases From BWRs for Environmental Impact Evaluation," NEDO-21 159, March 1976.

11.1-8 Gilbert, R. S., Skar pelos, J. M., "Technical Derivation of 1971 BWR Design Basis Radioactive Material Sour ce Terms," NEDO-10871, March 1973.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Table 11.1-1

Noble Radiogas Source Terms

Isotope Half-Life Source Term t = 0 ( Ci/sec) Source Term t = 30 minutes

( Ci/sec) 11.1-13 83m Kr 1.86 hr 3.4 x 10 3 2.9 x 1 0 3 85m Kr 4.4 hr 6.1 x 10 3 5.6 x 1 0 3 85Kr 10.74 years 10 to 2 0 a 10 to 20 a 87 Kr 76 minutes 2.0 x 10 4 1.5 x 1 0 4 88 Kr 2.79 hr 2.0 x 10 4 1.8 x 1 0 4 89 Kr 3.18 minutes 1.3 x 1 0 5 1.8 x 1 0 2 90 Kr 32.3 sec 2.8 x 1 0 5 --- 91 Kr 8.6 sec 3.3 x 1 0 5 --- 92 Kr 1.84 sec 3.3 x 1 0 5 --- 93 Kr 1.29 sec 9.9 x 1 0 4 --- 94 Kr 1.0 sec 2.3 x 1 0 4 --- 95 Kr 0.5 sec 2.1 x 10 3 --- 97 Kr 1.0 sec 1.4 x 1 0 1 --- 131m Xe 11.96 days 1.5 x 1 0 1 1.5 x 1 0 1 133m Xe 2.26 days 2.9 x 1 0 2 2.8 x 1 0 2 133 Xe 5.27 days 8.2 x 1 0 3 8.2 x 1 0 3 135m Xe 15.7 minutes 2.6 x 10 4 6.9 x 1 0 3 135 Xe 9.16 hr 2.2 x 10 4 2.2 x 1 0 4 137 Xe 3.82 minutes 1.5 x 10 5 6.7 x 1 0 2 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Table 11.1-1 Noble Radiogas Source Terms (Continued)

Isotope Half-Life Source Term t = 0 ( Ci/sec) Source Term t = 30 minutes

( Ci/sec) 11.1-14 138 Xe 14.2 minutes 8.9 x 10 4 2.1 x 1 0 4 139 Xe 40 sec 2.8 x 1 0 5 --- 140 Xe 13.6 sec 3.0 x 1 0 5 --- 141 Xe 1.72 sec 2.4 x 1 0 5 --- 142 Xe 1.22 sec 7.3 x 1 0 4 --- 143 Xe 0.96 sec 1.2 x 1 0 4 --- 144 Xe 9.0 sec 5.6 x 1 0 2 --- Totals 2.5 x 1 0 6 1.0 x 1 0 5 a Estimated from experimental observations.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-15 Table 11.1-2 Halogen Radioisotopes in Reactor Water Isotope Half-Life Concentration

( Ci/g) 83 Br 2.40 hr 1.5 x 10-2 84 Br 31.8 minutes 2.7 x 10-2 85 Br 3.0 minutes 1.7 x 10-2 131 I 8.07 days 1.3 x 1 0-2 132 I 2.28 hr 1.2 x 1 0-1 133 I 20.8 hr 8.8 x 1 0-2 134 I 52.3 minutes 2.4 x 1 0-1 135 I 6.7 hr 1.3 x 1 0-1 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Table 11.1-3 Other Fission Product Radioisotopes in Reactor Water Isotope Half-Life Concentration

( Ci/g) 11.1-16 89 Sr 50.8 days 3.1 x 1 0-3 90 Sr 28.9 years 2.3 x 1 0-4 91 Sr 9.67 hr 6.9 x 10-2 92 Sr 2.69 hr 1.1 x 10-1 95 Zr 65.5 days 4.0 x 1 0-5 97 Zr 16.8 hr 3.2 x10-5 95 Nb 35.1 days 4.2 x 1 0-5 99 Mo 66.6 hr 2.2 x1 0-2 99m Tc 6.007 hr 2.8 x 10-1 101 Tc 14.2 minutes 1.4 x 10-1 103 Ru 39.8 days 1.9 x 1 0-5 106 Ru 368 days 2.6 x 1 0-6 129m Te 34.1 days 4.0 x 1 0-5 132 Te 78.0 hr 4.9 x 10-2 134 Cs 2.06 years 1.6 x- 1 0-4 136 Cs 13.0 days 1.1 x 1 0-4 137 Cs 30.2 years 2.4 x 1 0-4 138 Cs 32.3 minutes 1.9 x 1 0-1 139 Ba 83.2 minutes 1.6 x 10-1 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Table 11.1-3 Other Fission Product Radioisotopes in Reactor Water (Continued)

Isotope Half-Life Concentration

( Ci/g) 11.1-17 140 Ba 12.8 days 9.0 x 1 0-3 141 Ba 18.3 minutes 1.7 x 10-1 142 Ba 10.7 minutes 1.7 x 10-1 141 Ce 32.53 days 3.9 x 1 0-5 143 Ce 33.0 hr 3.5 x 1 0-5 144 Ce 384.4 days 3.5 x 1 0-5 143 Pr 13.58 days 3.8 x 1 0-5 147 Nd 11.06 days 1.4 x 1 0-5 239 Np 2.35 days 2.4 x 1 0-1 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-18 Table 11.1-4 Coolant Activation Products in Reactor Water and Steam

Isotope Half-Life Steam Concentration

( Ci/g) Reactor Water Concentration

( Ci/g) 13 N 9.99 minutes 7 x10-3 4 x 1 0-2 16 N 7.13 sec 5 x 1 0 1 4 x 1 0 1 17 N 4.14 sec 2 x 1 0-2 6 x 1 0-3 19 O 26.8 sec 8 x 1 0-1 7 x 1 0-1 18 F 109.8 minutes 4 x 1 0-3 4 x 1 0-3 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.1-19 Table 11.1-5 Noncoolant Activation Pr oducts in Reactor Water Isotope Half-Life Concentration

( Ci/g) 24 Na 15 hr 2 x 1 0-3 32 P 14.31 days 2 x 1 0-5 51 Cr 27.8 days 5 x 1 0-4 54 Mn 313 days 4 x 1 0-5 56 Mn 2.58 hr 5 x 10-2 58 Co 71.4 days 5 x 1 0-3 60 Co 5.25 years 5 x 1 0-4 59 Fe 45 days 8 x 1 0-5 65 Ni 2.55 hr 3 x 10-4 65 Zn 243.7 days 2 x 1 0-6 69m Zn 13.7 hr 3 x 10-5 110m Ag 253 days 6 x 1 0-5 187 W 23.9 hr 3 x 10-3

Noble Radiogas Decay Constant Exponent Frequency Histogram 900547.45 11.1-1 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.Equilibrium Frequency of Measurements Diffusion Recoil m = KRB and Dresden 2 Average 000.10.20.30.40.50.60.70.80.91.0 Noble Radiogas Decay Constant Exponent (m) 5 10 15 20 25-1+1 Columbia Generating Station Final Safety Analysis Report Radiohalogen Decay Constant Exponent Frequency Histogram 900547.46 11.1-2 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.Equilibrium Diffusion Recoil n = KRB and Dresden 2 Average 15 10 5 000.10.20.30.40.50.60.70.80.91.0 Radiohalogen Decay Constant Exponent (n)

Frequency of Measurements

-1+1 Columbia Generating Station Final Safety Analysis Report Noble Radiogas Leakage Versus Iodine-131 Leakage 900547.47 11.1-3 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.KRB Dresden 2 50,000 100,000 150,000 0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 Noble Radiogas Leakage at +30 min ( Ci/sec)Dresden 1 March 1965 Dresden 2 (August 1970 - January 1971)

Iodine-131 Leakage From Fuel (µCi/sec)Iodine-131 Design Basis

1971 Source Terms Iodine-131 Design Basis

Pre-1971 Source Terms 1690 Columbia Generating Station Final Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.2-1 11.2 LIQUID WASTE MANAGEMENT SYSTEM 11.2.1 DESIGN BASIS

The liquid waste management system is designed to collect, segregate, store, and process potentially radioactive liquids generated during normal plan t operation and anticipated operational occurrences. The design objective is to keep the radiation dose in unrestricted areas as low as is reasonably achievable (ALARA) within the guidelines of Appendix I to 10 CFR 50. The design incorporates the objectives of maximum recycle and minimum release of radioactive liquids without limiting plant operations or availability.

The criteria considered in the design of this system include volume, radioac tivity, operational exposure, and required quality for recycle of the processed li quid. Radioisotopic inventories of the components used for the design are listed in Tables 11.2-1 through 11.2-5. These values are based on the reactor water source term associated with the design-bas is fuel leakage rate.

Allowance is made for concentration, d ecay and daughter product buildup in filters, demineralizers, and tanks.

Equipment locations and arrange ments are shown in Section 1.2.

The system is designed to treat pr ocess liquids with radioisotope concentrations associated with the design-basis fuel leakage and produce a quality of water which allows its recycle for plant reuse. Water inventory will occasionally require the discharge of processed liquids to the environs, in which case c oncentrations of radioiso topes in the effluent (Table 11.2-6) will be significantly less than the va lues specified in 10 CFR 20 and within the release limits established in the Technical Spec ifications. Radiation exposure to persons in unrestricted areas resulting from liquid waste discharged during normal operation and anticipated operational occurrences is less than the guidelines specified in 10 CFR 50, Appendix I.

Tanks that hold radioactive liquid, including the condensate storage tanks, are monitored for level and alarm primarily in the radwaste control room. Table 11.2-7 lists the design features of the tanks in the radwaste bu ilding used to prevent uncontrolle d releases due to spillage and shows overflow alarms and drai nage paths. The radwaste systems are operated from the radwaste control room; hence, additional local alarms are not required for radwaste tank levels.

All liquid waste management system tank overflows and drains are routed to the radwaste building equipment and floor drain sumps (see Figure 9.3-11

). Radioactive liquid samples are primarily routed to sampling sinks. Those samp les, which are local, drain into floor drain trenches, equipment or floor drain funnels, and pump beds. Th e above sample receivers are routed to various radioactive sumps, all of which are processed by the liquid waste management system.

Indoor tanks that hold radioactive liquid are not enclosed by individual curbs or elevated thresholds. The portion of the ra dwaste building that houses radi oactive liquid tanks is Seismic C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 2-0 0 5 11.2-2 Category I, as discussed in Section 3.8.4.1.2. The radwaste building can retain the normal operating capacity of all liquid radwaste tanks, which are nonseism ic. Analysis of a postulated radioactive release due to a liquid radwaste tank failure is presented in Section 15.7.3.

The radioactive and nonradioactive equipment and floor drains within the plant are segregated. Equipment and floor drains within the reactor bu ilding and radwaste building are routed to the liquid waste management system. The turbine building equipment and floor drains are segregated for radioactivity by component source and ar ea (see Section 9.3.3). Although three of the turbine building sumps are designated as collectors of nonradioactiv e waste water, there is a possibility of low level c ontamination of the effluent wate

r. One contamination source is steam leaks inside the building which condense on interior surfaces and are routed to floor drains. Because of this possibility, the discharge of these three su mps is routed to the radwaste system for processing.

The design of the system was accomplished prior to the issuan ce of Regulatory Guide 8.8.

However, the system does incorporate substantially the guidance provided in this regulatory guide.

Like the nonradioactive sump s, the storm water draina ge system (see Section 9.3.3.2.3.1) is not intended to collect radioactiv e materials. Nonethel ess, radionuclides have been detected in the pond water and sediments. These concentra tions are attributed to unanticipated and unavoidable occurrences. For example, special system draindowns during maintenance activities may have contributed minor amounts of tritium and corrosion products. Tritium that leaves the plant as a vapor can condense on building roofs and exterior wa lls and be carried to the pond in storm drainage. Also , water treatment filter backwa shes that are routed to the pond can contribute radionuc lides which were withdrawn from the river.

The degree of compliance with Regulatory Guide 1.143 is described in Section 1.8.

The liquid waste management system is desi gned to the requirement s of General Design Criteria, Appendix A to 10 CFR 50, as follows:

General Design Criterion 60 The system capacity as require d by General Design Cr iterion 60 is sufficient for the volume of liquid waste expected from norm al operation and anticipated ope rational occurrences such as condenser leakage, maintenance activities, and process equipment down time. Flow rates are listed in Tables 11.2-8 through 11.2-10.

General Design Criterion 64 Radioactivity monitoring in the sample tanks and in the effluent discharge path ensures that excess liquid discharged to the environs does not exceed the limits of 10 CFR 20. Sampled

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-3 fluids exceeding these limits are returned to an appropriate collector tank for reprocessing. The radwaste effluent radiation monito ring system is described in Section 11.5.2.2.2.3. 11.2.2 SYSTEM DESCRIPTION

11.2.2.1 Process Description Radioactive liquid wastes are collected and segregated into three categories: high purity waste, low purity waste, and chemical waste. Wastes thus classified are treated in subsystems designated as equipment drain, floor drain, and chemical waste, respectively.

High purity wastes, treated in the equipment drain subsystem, ha ve low conductivity and relatively high radioactivity concentrations. Radioactive material is removed from these wastes by filtration and ion exchange. Following treatment and batch sampling, the processed waste is normally returned to condensat e storage for reuse in plant.

Low purity wastes, treated in the floor drain subsystem, have moderate conductivity and generally low radioactivity. As with high purity wastes, radioactive material is removed by filtration and ion exchange. Following treatment and batch sampling, the processed waste is normally returned to condensate storage for reuse in the plant.

The high conductivity and organic content of chem ical wastes preclude nor mal treatment in the system demineralizers by ion exchange. Th ese wastes can be ne utralized by adding appropriate neutralizing agent and thorough mixing. If necess ary, following neutralization, these wastes are routed to a backwash tank or phase separator where unexpended ion-exchange capacity of the resins is used to remove contaminants from the chemical waste. After thorough mixing and a period of quiescence to allow the ion-exchange process to proceed, the excess liquid is decanted to the floor drain system fo r further processing. The spent resins in the separators are then proces sed as described below.

The installed chemical waste conc entrators are currently not used. Their preoperational testing and use has been deferred until a need is identified. There are currently no plans to activate the concentrators.

The installed detergent drain tanks are not norma lly used. Any wastes requiring disposal are handled on a case-by-case basi s and routed for processing a nd disposal in accordance with plant procedures and regulator y requirements and guidelines.

All liquid radwaste process stream s terminate in one of the samp le or distillate tanks. Since the liquid waste management system is operated on a batc h basis, this arrangement allows each treated batch to be sampled to ensure that the treatment was effect ive. If the sample indicates that the processed liquid is st ill above acceptable radioactivity limits or substandard in purity, equipment is provided to either recycle the batch through the same treatment or through a C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-4 subsystem providing a higher degree of treatment. If the sample indicates that the level of activity is within limits required to discharge and the water is in excess of inventory capacity or is substandard in purity, the processed liquid may be discharged. The actual release of effluents from any processed and sampled batch tank requires the opening of a key-locked valve in accordance with written operating procedures. All required information regarding the batch release must be documented. These procedures are established to prevent inadvertent release of liquids that have not been suitably processed and analyzed.

Expended ion exchange resins ar e removed by backwashing to the spent resin tank and phase separators. Excess backwash wa ter is removed from the phase separators by decantation and routed to either the floor drain or equipment drain collector tank for treatment. The powdered resin sludge is accumulated for radioactive decay. Following accumulation of successive layers in the phase sepa rator, the sludge is then transferre d to the radwaste processing system for dewatering or solidification. The deep bed resins in the spent resin tank, after a decay period, are also transferred to the solid radwaste processing system. Wa ter separated from the wastes is returned via the wa ste sludge phase separator to th e floor drain collector tank for treatment. The solid waste management system is described in Section 11.4.

Noncondensable gases from the liquid waste processing vessels are vented through the radwaste building exhaust system. Th is system is described in Section 9.4.3.2.

A process flow diagram, Figure 11.2-1 , together with process flow diagram data, Table 11.2-11 , show the volumes, flow ra tes, and radioactivity c oncentration used in the design of the liquid and solid waste management system.

The radionuclide distribution of liquid waste management system influents is based on the reactor coolant concentrations, as discussed in Section 11.1 , taking into consideration mixing and dilution sources.

Decontamination factors which we re used for evaluations of the system are those values specified in draft Re gulatory Guide 1.CC.

Table 11.2-12 shows the process equipment design basis decontamination factors which were used to generate the radi oactivity values in Table 11.2-11.

11.2.2.2 Subsystems Description 11.2.2.2.1 Equipment Drain Subsystem

The equipment drain subsystem consists of a waste collector ta nk, waste surge tank, pressure precoat filter, deep bed demineralizer, two waste sample tanks, and auxiliary equipment necessary to operate the subsystem. Sizes and capacities of the equipment are listed in Table 11.2-13. The waste surge tank prin cipally serves as a recept acle for reactor hydrotest and thermal expansion water a nd residual heat removal (RHR) system flush water during

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-5 startup and testing of these sy stems. The waste surge tank also serves as backup during equipment downtime. The piping and instrumenta tion drawing for this s ubsystem is shown in Figure 11.2-2. High purity (low solids content) liquid wastes are collected in the waste collector tank from the following sources:

a. Drywell equipment drain sump, b. Reactor building equipment drain sump, c. Radwaste building equipment drain sump, d. Turbine building equipment drain sump,

e. Reactor water cleanup system,

f. RHR system,

g. Cleanup phase separators (decant water),

h. Condensate phase separa tors (decant water), and i. Fuel pool seal rupture dr ains (to waste surge tank).

The quantities of these wastes are summarized in Table 11.2-8. Since these wastes can contain a high percentage of primary reactor water, the radioactive concentration could be relatively high (on the order of 2.4 Ci/ml).

In the event of a compon ent malfunction within the equipment drain subsystem, sufficient cross ties are provided to the floor drain subsystem to permit continued proce ssing of the wastes.

Sufficient capacity and treatment capability are provided to handle such conditions.

Normally, the equipment drain subsystem treated effluents are recycled to the condensate storage tanks for reuse w ithin the plant. When condensate st orage capacity is not available or the water is substandard in purity, the purified liquid from this subsystem ma y be sampled, analyzed and, if acceptable for release as de scribed previously, routed to the blowdown line for discharge. Liquid waste that is unacceptable for discharge is reprocessed.

11.2.2.2.2 Floor Drain Subsystem

The floor drain subsystem consists of a floor drain collector tank, pressure precoat filter, deep bed demineralizer, sample tank, and auxiliary equipment necessary to operate the subsystem. Sizes and capacities of the equipment are listed in Table 11.2-13. The flow diagram for this subsystem is shown in Figure 11.2-3.

Intermediate purity liquid wastes are collected in the floor drain co llector tank from the following sources:

a. Drywell floor drain sump, b. Reactor building floor drain sumps, C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-6 c. Radwaste building floor drain sumps, d. Turbine building floor drain sump, and

e. Waste sludge phase separator (decant water).

The quantities of these wastes are summarized in Table 11.2-9. These wastes are normally of intermediate purity (50 mho/cm and higher) and have radioactiv e concentrations on the order of 0.1 Ci/ml.

Similar to the equipment drain subsystem, the floor drain subsystem nor mally functions as an independent process stream.

Equipment redundancy and inters ystem cross ties have been provided to allow substituti on for any failed component.

High purity effluent is routed to condensate storage for reuse in the plant. When condensate storage capacity is exceeded or the processed liquid is substandard in purity, it may be discharged via the blowdown line, if it meets acceptable limits.

11.2.2.2.3 Chemical Waste Subsystem

The chemical waste subsystem cons ists of two of each of the fo llowing: detergent drain tanks, chemical waste tanks, decontamination soluti on concentrators, dec ontamination solution concentrated waste tanks, and distillate tanks. It contains also a polishing (deep bed) demineralizer and auxiliary equipment necessary to operate the subsystem. Sizes and capacities of the equipment are listed in Table 11.2-13. The flow diagram for this subsystem is shown in Figure 11.2-4. The decontamination solution concentrators, con centrated waste tanks, distillate tanks, and polishing demineralizer have been inst alled but will not be used until plant operating experience indicates a need and system testing is accomplished. Chemical wastes are currently processed by routing to a backwash tank or phase separator for use of unexpended ion-exchange capacity of the resins to clean th e water prior to decanting to the floor drain subsystem for further processing.

Chemical wastes collected in the chemical waste ta nk are from the following sources:

a. Detergent drains,

b. Shop decontamination solutions,

c. Reactor and turbine build ing decontamination drains, d. Low purity wastes from either the equipment or floor drain subsystems, e. Filter demineralizer elemen t chemical cleaning solutions, f. Battery room drains,

g. Chemical system overf lows and tank drains, and h. Laboratory drains.

The quantities of these wastes are summarized in Table 11.2-10. These chemical wastes are of such high conductivity and organic content as to preclude normal treatment by ion exchange, C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-7 and the radioactivity concentrations are variable. These wastes are processed by routing to a backwash tank or phase separa tor and to the floor drain sy stem for further processing.

If the concentrators were ever activated, the evaporator concentrates would be processed by the solid waste management sy stem and the distillate would be routed to the distillate tank. After analysis, the distillate could be routed through a polishing demineraliz er to further reduce impurities, recycled through the ev aporator, or sent directly to condensate storage for plant reuse. As with the other subs ystems, when high purity water storage capacity is exceeded or the processed liquid is of substandard purity, liquid within 10 CFR 20 release limits could be discharged via the blowdown line.

11.2.2.2.4 Shared Equipment

Other than serving as mutual backup, main pr ocess equipment normally is not shared between subsystems. Auxiliary equipment not in the direct process stream is shared between subsystems. Shared equipment includes the following:

a. The waste precoat tank and waste preco at pump are shared between the waste collector filter and the floor drain filter,

b. The waste filter aid tank is shared between the waste collector filter and floor drain filter,

c. The resin addition tank is shared between the waste demineralizer, floor drain demineralizer, and polishing demineralizer, and

d. The chemical addition tanks, caustic and acid, and associated pumps are shared between the waste collector tank, floor drain collector tank, detergent tanks, and chemical waste tanks.

11.2.2.2.5 Surge Capacities

The radwaste system process data is th e basis for sizing of the equipment.

Tables 11.2-8 , 11.2-9 , and 11.2-10 list startup flows, daily flows, and maximum flows for the equipment drain subsystem, floor drain subsystem and chemical waste subsystem, respectively. Anticipated operational occurrences such as startup operations, equipment malfunction, and shutdown operations are accounted for in these tabulations. The ba ses for these types of values are presented in NEDO-10951, "Releases fr om BWR Radwaste Management Systems,"

July 1973.

The surge storage and proce ss capacities can be envisioned by comparing the normal and maximum daily volumes, listed in Table 11.2-11 , with the design flow rates of pumps and tank volumes listed in Table 11.2-13. Alternate processing rather than bypass operations are used C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-8 during equipment downtime. The equipment and floor drain subsystems are sized such that with either subsystem inoperative, the remaining subsystem is capable of processing the maximum expected volume of both subsystems. Additionally, the waste surge tank provides reserve storage capacity. The chemical waste subsystem incorporates two parallel processing paths. Cross connections allow individual components from either process path to serve as a substitute in the other process pa th. The parallel path processing and adequate storage capacity ensures that inoperability of a ny component in this subsystem will not limit plant operation.

11.2.2.2.6 Design Features The design pressures and temperatures for individual components are listed in Table 11.2-13. Collection and storage tanks are designed fo r atmospheric pressu re. The mixed-bed demineralizer units, precoat filter units, and concentrators are pressure vessels. The quality classification for the system is Qua lity Group D+ as defined in Section

3.2. Chapter

12 discusses the design features incorporated in the system to maintain occupational exposure ALARA. As show n in the general arrangeme nt drawings in Section 1.2, the radwaste processing equipment is located in shielded rooms and cells. Process lines that penetrate shield walls are rout ed to prevent a direct radia tion path from the tanks and equipment to normally occupied areas.

Control of the liquid waste management system is from a shield ed radwaste control room and from shielded local operating galleries. Flanges are provided wh ere required for maintenance in an otherwise all-welded syst em. The cells and concrete room s provide secondary enclosures which facilitate collection of spills or leaks from the system for processing. Vessels that may have their contents mixed with air or that may have batches transferred into them by means of pneumatic transfer are closed and vented to the radwaste bu ilding ventilation system. Piping and tubing 2 in. and under is fiel d routed but required to be in specified space envelopes for shielding and in-plant exposure consideration.

The liquid waste management system is desi gned to minimize the effects of equipment malfunction and operator error. After initiation by the opera tor, valve positioning, equipment startup, and system operation can be automatically performed by a process controller. Failure of any valve to properly align stops the sequence and prevents a pump start. Once on line, processing continues until the feed volume is processed or the capacity of a piece of process equipment is reached. In either case, automati c shutdown occurs with va lves returning to the shutdown position. During initial system startup or in the event of controller failure, the processes are manually controlled. System variables, such as tank levels, flow rates, pressures, and conductivity are indicated and alarmed in the radwaste control room.

The discharge from the liquid waste subsystems to the blowdown line is normally isolated by a closed manual valve and a key-lock ed closed manual valve separate d by a telltale drain. This path is further protected by air-operated isolation valves, flow control valves, flow indication

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-9 and radiation monitoring inst rumentation. These design features are complimented by administrative procedures which prevent inadverten t radioactive liquid rele ases and releases of liquids that exceed the Technical Specifications release limit.

11.2.3 RADIOACTIVE RELEASES

The decision to recycle or to release a particular batch of processed liquid is based on plant water inventory and the type and concentration of chemical impur ities being processed. It is not expected that tritium buildup in the plant will determine the frequency of releases.

Concentration of radionuclides released to the environment will be significantly less than those values specified in 10 CFR 20. The quantities of radioactive materials released during normal operation will not result in dose ra tes to persons in unrestricted areas in excess of 10 CFR 50, Appendix I, values as implemented by the Technical Specifications release limits.

11.2.3.1 Release Point and Dilution Excess liquid effluent is discharged to the circul ating water blowdown line at a variable rate up to about 190 gpm. Dilution is furnished by ci rculating water system blowdown when required or when normal blowdown is in progress. Th e circulating water blow down line terminates in the Columbia River. Applicable concentration limits specified in 10 CFR 20 apply at the point of discharge to the river.

The storm water drainage pond described in Section 9.3.3.2.3.1 is a point of release of detectable radionuclides. For the reasons discussed in Section 11.2.1 , most of the activity is believed to be either co ndensed gaseous effluent or material of external origin. The pond is within a plant restricted area (see Section 2.1.1.3), and public access is restricted by a fence which surrounds the pond a nd discharge channel.

11.2.3.2 Calculation of Releases of Radioactive Materials Quantities of radioactive materi als released with liquids we re calculated for initial plant licensing using the GALE code presented in dr aft Regulatory Guide 1.CC to show compliance with Appendix I to 10 CFR 50 for normal operati on plus anticipated operational occurrences.

The plant operational parameters, including source terms, were t hose presented in Appendix B of the guide with waste stream flow rates adjusted to the Columbia Generating Station plant design. The calculated quantities of radioactive material s released with liquids used for initial plant licensing are presented in Table 11.2-14 in terms of curies per year. The radionuclide concentrations in the effluent are presented in Table 11.2-6 and are compared with the values of 10 CFR 20, Appendix B, Table II, Column 2.

Since becoming operational, releases of radioactive materials in liquid effluents have been determined using actual flow volumes and quantitati ve and qualitative laboratory analyses. Doses due to radioactive materials released in liquid effluents are determined to be in

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-10 compliance with Technical Specific ations at specified intervals. Compliance is reported in the Annual Radioactive Effluent Rele ase Report using the NRC LADT AP II computer code along with parameters outlined in the Offsite Dose Calculation Manual.

11.2.3.3 Exposure of Persons at or Beyond the Site Boundary Estimated annual exposure of pe rsons in unrestricted areas re sulting from liquid effluents (Table 11.2-6) is discussed in Section 5.2 of the Environmental Report, Operating License stage. The estimated total body dose of 2.3 mrem per year and largest calculated single organ dose of 1.6 mrem per year to the bone are well within the guidelines of 10 CFR 50, Appendix I.

11.2.3.4 Cost-Benefit Analysis It has been determined that a cost-benefit analysis, as described in Appendix I to 10 CFR 50 Section II.D is not required for Columbia Generating Station.

The question of eligibility of Columbia Genera ting Station to dispense with the Appendix I cost-benefit analysis as to "per site" limitations was reviewed by the NRC in connection with their review of WPPSS Nuclear Project No.

1 (WNP-1) and WPPSS Nuclear Project No. 4 (WNP-4), Docket Nos. 50-460 and 50-513. The Staff concluded in its testimony at the Atomic Safety and Licensing Board (ASLB) hearing for WNP-1 and WNP-4 held on November 11, 1975, that

The aggregate doses associated with WPPSS Nuclear Project No. 1, WPPSS Project No. 2, and WPPSS Nuclear Proj ect No. 4 operation meet the RM-50-2 (i.e., Annex) design obj ectives [Tr. 724-727].

These conclusions were ratified by the ASLB on its decision pertaining to WNP-1 and WNP-4 of December 22, 1975, RAI-75/12 922, 934.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-11 Table 11.2-1 Liquid Waste Management Sys t em Radioisotope Inventory Equipment Drain Subsystem Radioisotope Inventory ( C i) Radioisotope Waste Collector Tank a Waste Filter a Waste Demineralizer a Waste Sam p le Tank a 83 Br 4.53E4 3.15E4 3.18E2 84 Br 4.67E4 9.19E3 9.28E1 85 Br 1.17E4 4.17E-4 4.21E-6 89 Sr 4.35E4 1.88E6 4.35E2 90 Sr 3.25E3 2.12E5 3.25E1 91 Sr 7.12E5 1.10E6 9.13E3 92 Sr 3.40E5 2.48E5 2.50E3 90 Y 3.53E2 2.00E5 3.92E0 91 Y 4.21E3 5.10E5 4.69E1 91M Y 4.12E5 7.31E5 5.82E3 92 Y 2.15E5 2.37E5 2.33E3 95Zr 5.63E2 2.81 E 2 1.31E4 5.62E0 97 Zr 6.94E1 1.73E1 1.70E1 3.46E-1 95Nb 5.93E2 2.97 E 2 1.69E4 5.93E0 99Mo 2.79E5 1.38 E 5 6.13E5 2.75E3 99MTc 1.70E6 8.18 E 3 2.02E6 1.51E4 101 Tc 2.05E5 5.15E3 5.20E1 103Ru 2.66E2 1.33 E 2 5.17E3 2.66E0 106 Ru 3.67E1 1.84E1 1.12E3 3.67E-1 129M Te 5.60E2 2.00E4 5.59E0 129 Te 3.03E2 1.27E4 3.31E0 132 Te 6.30E5 3.21E6 6.23E3 129 I 2.83E-8 9.76E-5 3.16E-10 131I 1.77E5 2.12E6 1.76E3 132I 8.53E5 3.45E6 7.83E3 133I 9.03E5 1.57E6 8.67E3 134I 4.66E5 1.75E5 1.77E3 135I 7.73E5 7.35E5 6.80E3 134 Cs 2.26E3 1.44E5 2.26E1 135 Cs 4.19E-5 1.97E-2 5.95E-7 136 Cs 1.52E3 2.80E4 1.51E1 137 Cs 3.39E3 2.21E5 3.39E1 138 Cs 3.29E5 6.54E4 6.60E2 137M Ba 2.99E3 2.07E5 3.17E1 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-12 Table 11.2-1 Liquid Waste Management Sys t em Radioisotope Inventory Equipment Drain S ubsystem (Continued)

Radioisotope Inventory ( C i) Radioisotope Waste Collector Tank a Waste Filter a Waste Demineralizer a Waste Sam p le Tank a 139 Ba 3.67E5 1.98E5 2.00E3 140 Ba 1.24E5 2.29E6 1.24E3 141 Ba 2.64E5 1.50E4 1.52E2 142 Ba 2.32E5 2.14E3 2.16E1 141Ce 1.18E3 5.99 E 2 3.04E4 1.22E1 143Ce 3.98E2 1.94 E 2 4.80E2 3.88E0 144Ce 4.94E2 2.47 E 2 1.48E4 4.94E0 140 La 2.07E4 2.28E6 2.29E2 141 La 3.66E4 4.61E4 4.60E2 142 La 2.34E4 3.08E4 3.11E2 143Pr 5.37E2 2.69 E 2 1.71E4 5.37E0 144Pr 4.40E2 2.46 E 2 1.48E4 4.91E0 147Nd 1.92E2 9.59 E 1 1.52E3 1.92E0 239 NP 2.98E6 1.13E7 2.93E4 24 Na 1.80E4 2.52E4 1.70E2 32P 2.77E2 5.54E3 2.76E0 51Cr 6.99E3 3.49 E 3 1.10E5 6.98E1 54Mn 5.65E2 2.82 E 2 1.70E4 5.65E0 56Mn 1.58E5 5.67 E 4 5.57E4 1.13E3 58Co 7.04E4 3.52 E 4 1.68E6 7.03E2 60Co 7.07E3 3.53 E 4 2.26E5 7.07E1 59Fe 1.12E3 5.61 E 2 2.30E4 1.12E1 65Ni 9.40E2 3.36 E 2 3.30E2 6.73E0 65 Zn 2.82E1 1.67E3 2.82E-1 69M Zn 2.63E2 3.52E2 2.47E0 110M Ag 8.47E2 4.24 E 2 2.52E4 8.47E0 187W 3.16E4 1.53E4 2.99E4 3.05E2 a The radioisotope inventory lis ted above for the waste collec tor tank, waste filter, waste demineralizer, and waste sample tank is assumed to be app licable to the floor drain collector tank, floor drain filter, floor drain demineralizer , and floor drain sample tank, respectively. This is done for the purpose of shielding analys is. In actuality, the radionuclide inventory in any component of the floor drain sy stem is discussed in Sections 11.2.2.2.1 and 11.2.2.2.2. The values from the equipment drain subsystem are maximum due to cross-ties for alternate processing.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-13 Table 11.2-2

Liquid Waste Management Sys t em Radioisotope Inventory Chemical W a s t e Subsys t e m Radioisotope Inventory ( C i) Radioisotope Chemical Waste Tank DSC Waste Tank a DSC Waste Measuring Tank b Distillate Tank Decon. Solution Concentrat o r 83Br 84Br 85Br 89 Sr 1.11E3 2.82E3 1

.74E3 1.11E0 1.48E3 90 Sr 7.87E1 2.00E2 1

.23E2 7.87E-2 1.05E2 91Sr 92Sr 90 Y 91 Y 91M Y 92 Y 95 Zr 1.43E1 3.65E1 2

.25E1 1.43E-2 1.92E1 97Zr 95Bv 99 Mo 7.87E3 2.00E4 1

.23E4 7.87E0 1.05E4 99MTc 101Tc 103Ru 106Ru 129M Te 1.43E1 3.65E1 2

.25E1 1.43E-2 1.92E1 129Te 132 Te 1.75E4 4.45E4 2

.75E4 1.75E1 2.34E4 129 I 131 I 4.65E3 1.13E4 6

.95E3 4.65E0 5.92E3 132 I 133 I 134 I 135 I 134 Cs 5.75E1 1.46E2 9

.02E1 5.75E-2 7.68E1 135Cs 136 Cs 3.94E1 1.00E2 6

.19E1 3.94E-2 5.27E1 137 Cs 8.60E1 2.19E2 1

.35E2 8.60E-2 1.15E2 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-14 Table 11.2-2 Liquid Waste Management Sys t em Radioisotope Inventory Chemical Waste Subsystem (Continued)

Radioisotope Inventory ( C i) Radioisotope Chemical Waste Tank DSC Waste Tank a DSC Waste Measuring Tank b Distillate Tank Decon. Solution Concentrat o r 138Cs 137M Ba 139 Ba 140 Ba 3.22E3 8.18E3 5

.05E3 3.22E0 4.30E3 141 Ba 142 Ba 141 Ce 1.40E1 3.56E1 2

.20E1 1.40E-2 1.87E1 143 Ce 1.25E1 3.17E1 1

.96E1 1.25E-2 1.67E1 144 Ce 1.25E1 3.17E1 1

.96E1 1.25E-2 1.67E1 140 La 141 La 142 La 143 Pr 1.36E1 3.44E1 2

.12E1 1.36E-2 1.81E1 144 Pr 147 Nd 5.01E0 1.27E1 7

.85E0 5.01E-3 6.69E0 239 Np 8.60E4 2.19E5 1

.35E5 8.60E1 1.15E5 24 Na 32 P 7.18E0 1.82E1 1.13 E 1 7.18E-3 9.58E0 51 Cr 1.79E2 4.54E2 2

.80E2 1.79E-1 2.38E2 54 Mn 1.43E1 3.65E1 2

.25E1 1.43E-2 1.92E1 56 Mn 58 Co 1.79E3 4.54E3 2

.80E3 1.79E0 2.38E3 60 Co 1.79E2 4.54E2 2

.80E2 1.79E-1 2.38E2 59 Fe 2.87E1 7.26E1 4

.48E1 2.87E-2 3.82E1 65 Ni 65Zn 7.18E-1 1.82E0 1.13E0 7.18E-4 9.58E-1 69M Zn 110M Ag 2.15E1 5.46E1 3

.37E1 2.15E-2 2.87E1 187W a Decontamination solution concentrator waste tank.

b Decontamination solution concentrator waste measuring tank.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-15 Table 11.2-3 Liquid Waste Management Sys t em Radioisotope Inventory RWCU and Condensate Filter Demineralizers Radioisotope Inventory ( C i) Radioisotope RWCU Filter Demineralizer Condensate Filter Demineralizer 83Br 2.50E6 2.30E5 84Br 9.00E5 9.12E4 85Br 3.00E4 5.42E3 89Sr 1.40E7 3.00E5 90Sr 1.10E6 2.56E4 91Sr 3.20E7 2.13E5 92Sr 1.60E7 9.12E4 90Y 1.10E6 2.09E4 91 Y 4.54E4 91M Y 2.20E7 1.28E5 92 Y 9.12E4 95 Zr 1.80E5 2.59E4 97 Zr 2.40E4 8.19E1 95 Nb 5.60E5 2.97E4 99 Mo 5.50E7 3.04E6 99M Tc 2.40E7 3.18E6 101 Tc 1.90E6 1.04E4 103 Ru 9.00E4 1.16E4 106 Ru 1.20E4 1.85E3 129M Te 1.50E6 3.60E3 129 Te 2.29E3 132 Te 3.70E7 1.20E6 129 I 4.89E-6 131I 8.50E7 1.34E7 132I 5.80E7 2.96E6 133I 1.40E8 1.19E7 134I 1.40E7 1.35E6 135I 6.50E7 5.55E6 134 Cs 7.50E5 1.77E4 135 Cs 2.48E-2 136 Cs 4.20E5 7.37E3 137 Cs 1.20E6 2.67E4 138 Cs 5.50E6 3.23E4 127M Ba 1.20E6 2.50E4 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-16 Table 11.2-3 Liquid Waste Management Sys t em Radioisotope Inventory RWCU and Condensate Filter Demineralizers (Continued)

Radioisotope Inventory ( C i) Radioisotope RWCU Filter Deminer alizer Condensate Filter Demineralizer 139 Ba 1.20E7 7.23E4 140 Ba 3.60E7 6.03E5 141 Ba 3.00E6 1.63E4 142 Ba 1.60E6 9.94E3 141 Ce 1.80E5 2.86E4 143 Ce 5.00E4 2.40E3 144 Ce 1.60E5 2.47E4 140 La 4.10E7 5.60E5 141 La 1.63E4 142 La 9.94E3 143 Pr 1.60E5 2.74E4 144 Pr 2.47E4 147 Nd 5.50E4 5.65E3 239 Np 5.00E8 4.27E6 24 Na 1.50E6 9.56E3 32P 9.50E4 1.40E3 51 Cr 2.60E6 2.80E5 54 Mn 2.20E5 2.83E4 56 Mn 6.00E6 2.69E5 58 Co 2.60E7 3.27E6 60 Co 2.80E6 3.61E5 59 Fe 4.30E5 4.96E4 65 Ni 3.80E4 1.59E3 65 Zn 1.10E4 2.16E2 69M Zn 2.00E4 1.34E2 110M Ag 3.40E5 4.24E4 187 W 3.60E6 1.49E5 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-17 Table 11.2-4 Liquid Waste Management Sys t em Radioisotope Inventory Spent Resin and Condensate Backwash Receiving Tanks Radioisotope Inventory ( C i) Radioisotope Spent Resin Tank C ondensate Backwash Receiving Tank 83 Br 3.15E4 2.30E5 84 Br 9.19E3 9.12E4 85 Br 4.17E-4 5.42E3 89 Sr 1.88E6 3.00E5 90 Sr 2.12E5 2.56E4 91Sr 1.10E6 2.13E5 92 Sr 2.48E5 9.12E4 90 Y 2.00E5 2.09E4 91 Y 5.10E5 4.54E4 91M Y 7.31E5 1.28E5 92 Y 2.37E5 9.12E4 95 Zr 1.31E4 2.59E4 97 Zr 1.70E1 8.19E1 95 Nb 1.69E4 2.97E4 99 Mo 6.13E5 3.04E6 99M Tc 2.02E6 3.18E6 101 Tc 5.15E3 1.04E4 103 Ru 5.17E3 1.16E4 106 Ru 1.12E3 1.85E3 129M Te 2.00E4 3.60E3 129 Te 1.27E4 2.29E3 132 Te 3.21E6 1.20E6 129 I 9.76E-5 4.89E-6 131 I 2.12E6 1.34E7 132 I 3.45E6 2.96E6 133 I 1.57E6 1.19E7 134 I 1.75E4 1.35E6 135 I 7.35E4 5.55E6 134 Cs 1.44E4 1.77E4 135 Cs 1.97E-2 2.48E-2 136 Cs 2.80E4 7.37E3 137 Cs 2.21E5 2.67E4 138 Cs 6.54E4 3.23E4 137M Ba 2.07E5 2.50E4 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-18 Table 11.2-4 Liquid Waste Management Sys t em Radioisotope Inventory Spent Resin and Condensate Backw a sh Receiving Tanks (Continued)

Radioisotope Inventory ( C i) Radioisotope Spent Resin Tank Cond e nsate Backwash Receiving Tank 139 Ba 1.98E5 7.23E4 140 Ba 2.29E6 6.03E5 141 Ba 1.50E4 1.63E4 142 Ba 2.14E3 9.94E3 141 Ce 3.04E4 2.86E4 143 Ce 4.80E2 2.40E3 140 Ce 1.48E4 2.47E4 141 La 2.28E6 5.60E5 141 La 4.61E4 1.63E4 142 La 3.08E4 9.94E3 143 Pr 1.71E4 2.74E4 144 Pr 1.48E4 2.47E4 147 Nd 1.52E3 5.65E3 239 Np 1.13E7 4.27E6 24 Na 2.52E4 9.56E3 32 P 5.54E3 1.40E3 51 Cr 1.10E5 2.80E5 54 Mn 1.70E4 2.83E4 56 Mn 5.57E4 2.69E5 58 Co 1.68E6 3.27E6 60 Co 2.26E5 3.61E5 59 Fe 2.30E4 4.96E4 65 Ni 3.30E2 1.59E3 65 Zn 1.67E3 2.16E2 69M Zn 3.52E2 1.34E2 110M Ag 2.52E4 4.24E4 187 W 2.99E4 1.49E5 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-19 Table 11.2-5 Liquid Waste Management Sys t em Radioisotope Inventory Phase S e pa r a tors Radioisotope Inventory ( C i) Radioisotope RWCU Phase Separator Condensate Phase Separator Waste Sludge Phase S e pa r a tor 83 Br 4.95E6 2.26E5 84 Br 1.71E6 7.04E4 85 Br 3.73E4 8.21E2 89 Sr 3.48E8 2.94E6 90 Sr 3.95E7 3.23E5 91 Sr 6.40E7 2.21E5 92 Sr 3.17E7 8.98E4 90 Y 3.95E7 3.15E5 91 Y 5.66E6 4.76E5 91M Y 4.40E7 1.35E5 92 Y 2.33E5 9.57E4 95 Zr 4.83E6 2.55E5 3.70E1 97 Zr 4.70E4 7.14E1 4.53E-1 95 Nb 1.41E7 3.29E5 3.96E1 99 Mo 1.93E8 5.23E6 1.35E4 99M Tc 1.27E8 2.17E6 8.76E3 101Tc 3.41E6 7.67E-2 103 Ru 2.06E6 1.36E5 1.74E1 106 Ru 4.09E5 2.14E4 2.45E0 129MTe 3.15E7 3.48E-4 129 Te 1.82E7 2.07E3 132 Te 1.43E8 2.40E6 129 I 1.29E-1 6.20E-5 131 I 6.70E8 5.31E7 132 I 1.87E8 4.16E6 133 I 3.00E8 1.33E7 134 I 2.72E7 1.18E6 135 I 1.30E8 5.69E6 134 Cs 2.63E7 2.19E5 135 Cs 1.98E-1 3.48E-1 136 Cs 4.00E6 4.07E4 137 Cs 4.31E7 3.37E5 138 Cs 1.05E7 2.50E4

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-20 Table 11.2-5 Liquid Waste Management Sys t em Radioisotope Inventory Phase Separators (Continued)

Radioisotope Inventory ( C i) Radioisotope RWCU Phase Separator Condensate Phase Separator Waste Sludge Phase S e pa r a tor 137M Ba 4.00E7 3.14E5 139 Ba 2.36E7 6.74E4 140 Ba 4.18E8 3.33E6 141 Ba 5.51E6 1.01E4 142 Ba 2.79E6 4.69E3 141 Ce 3.80E6 2.37E5 7.78E1 143 Ce 1.22E5 2.88E3 1.46E1 144 Ce 5.37E6 2.83E5 3.29E1 140 La 4.78E8 3.64E6 141 La 3.86E4 5.32E2 142 La 4.89E4 6.33E2 143 Pr 5.68E6 3.26E5 3.59E1 144 Pr 5.08E6 2.83E5 3.29E1 147 Nd 5.61E5 2.66E4 1.18E1 239 Np 1.57E9 6.92E6 24 Na 3.07E6 1.02E4 32 P 1.19E6 8.19E3 51 Cr 4.93E7 2.15E6 4.50E2 54 Mn 7.43E6 3.26E5 3.76E1 56 Mn 1.19E7 2.52E5 1.52E3 58 Co 7.14E8 3.27E7 4.64E3 60 Co 9.98E7 4.30E6 4.72E2 59 Fe 1.03E7 4.50E5 7.35E1 65 Ni 7.52E4 1.49E3 9.01E0 65 Zn 3.63E5 2.57E3 69M Zn 4.06E4 1.41E2 110M Ag 1.14E7 4.84E5 5.64E1 187 W 7.95E6 1.63E5 9.81E2

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-21 Table 11.2-6 Annual Av e r age Concentration of Radionuclides in Liquid Effluent Nuclide Effluent Concentration (Ci/ml) 10 CFR 20 Table II (Ci/ml) Fraction of 10 CFR 20 Effl u e nt/10 CFR 20 24 Na 1.3E-9 5E-5 0.000026 32P 5.2E-11 9E-6 0.00000575 51Cr 1.3E-9 5E-4 0.0000026 54 Mn 1.6E-11 3E-5 0.000000533 56 Mn 1.4E-9 7E-5 0.00002 55Fe 2.8E-10 1E-4 0.0000028 59Fe 8E-12 1E-5 0.00000008 58 Co 5.4E-11 2E-5 0.0000027 60 Co 1.1E-10 3E-6 0.0000367 65Ni 8E-12 1E-4 0.00000008 64 Cu 4E-9 2E-4 0.00002 65 Zn 5.4E-11 5E-6 0.000011 69m Zn 2.8E-10 6E-5 0.00000467 69 Zn 3E-10 8E-4 0.000000375 187 W 5.4E-11 3E-5 0.0000018 239 Np 1.6E-9 2E-5 0.00008 83Br 7.2E-11 9E-4 0.00000008 84Br 6E-12 4E-4 0.000000015 89 Rb 4.2E-11 9E-4 0.0000000467 89Sr 2.8E-11 8E-6 0.0000035 91Sr 4.4E-10 2E-5 0.000022 91m Y 2.8E-10 2E-3 0.00000014 91Y 1.4E-11 8E-6 0.00000175 92Sr 3E-10 4E-5 0.0000075 92Y 6.2E-10 4E-5 0.0000155 93Y 4.6E-10 2E-5 0.000023 98 Nb 1.6E-11 2E-4 0.00000008 99 Mo 4.6E-10 2E-5 0.000023 99m Tc 1.8E-9 1E-3 0.0000018 101 Tc 4E-12 2E-3 0.000000002 103 Ru 6E-12 3E-5 0.0000002

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-22 Table 11.2-6 Annual Av e r age Concentration of Radionuclides in Liquid Effluent (Con t inued) Nuclide Effluent Concentration (Ci/ml) 10 CFR 20 Table II (Ci/ml) Fraction of 10 CFR 20 Effl u e nt/10 CFR 20 103m Rh 6E-12 6E-3 0.000000001 104 Tc 1.2E-11 4E-4 0.00000003 105 Ru 1.1E-10 7E-5 0.00000165 105m Rh 1.2E-10

-- -- 105 Rh 3.8E-11 5E-5 0.00000076 129m Te 1E-11 7E-6 0.00000150 129 Te 6E-12 4E-4 0.00000016 131m Te 2E-11 8E-6 0.0000026 131 Te 4E-12 8E-5 0.00000005 131I 1.4E-9 1E-6 0.0014 132 Te 2E-12 9E-6 0.000000233 132I 7E-10 1E-4 0.000007 133I 3.6E-9 7E-6 0.000510 134I 2.9E-10 4E-4 0.0000007 134 Cs 1.6E-9 9E-7 0.00175 135I 1.7E-9 3E-5 0.0000580 136 Cs 9.9E-10 6E-6 0.000165 137 Cs 3.6E-9 1E-6 0.0036 137m Ba 3.4E-9 -- -- 138 Cs 1.5E-9 4E-4 0.0000037 139 Ba 1E-10 2E-4 0.0000005 140 Ba 1E-10 8E-6 0.0000131 140 La 2.3E-11 9E-6 0.00000256 141 La 3.6E-11 5E-5 0.00000071 141 Ce 8E-12 3E-5 0.000000280 142 La 7.6E-11 1E-4 0.00000076 143 Ce 6E-12 2E-5 0.0000003 143 Pr 1E-11 2E-5 0.0000005 All others 1.5E-11 1E-8 0.0015 Total 0.009020094

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-000 11.2-23 Table 11.2-7

Tank Design Features Component High Level Alarm Overflows and Drains Floor drain collector tank Yes (RCR) Floor drain sump, W-2 Waste sludge phase separator Yes (RCR) Floor drain sump, W-2 Floor drain sample tank Yes (RCR) Floor drain sump, W-2 Waste collector tank Yes (RCR) Floor drain sump, W-2 Spent resin tank Yes (RCR) Floor drain sump, W-2 Waste surge tank Yes (RCR) Equipment drain sump, W-3 Waste sample tanks Yes (RCR) Equipment drain sump, W-3

Detergent drain tanks b No a Chemical waste sump, W-4 Chemical waste tanks Yes (RCR) Chemical waste sump, W-4 Distillate tanks b No a Chemical waste sump, W-4 Nonoperational decontamination solution concentrator waste tanks b No Chemical waste sump, W-4 Condensate backwash receiving tank Yes (RCR) Equipment drain sump, W-3 Condensate phase separators Yes (RCR) Equipment drain sump, W-3 Decontamination solution concentrator waste measuring tank c No a Decontamination solution concentrator waste tanks RWCU phase separators Yes (RCR) Equipment drain sump, W-3 Condensate storage tanks Yes (MCR) Floor drain sump, T4 a Alarms not installed. LEGEND b Nonoperational. RCR - radwaste control room c Not used. MCR - main control room

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-24 Table 11.2-8

Equipment Drain Subsystem Sources Source Startup Flows (gpd)

Regular Daily Flows (gpd) Irregular Flows (gpd) Maximum Daily Flows (gpd) Equipment drains Drywell 3,860 3,860 28,800 Reactor building 3,755 3,755 14,400 Turbine building 5,726 5,726 5,726 Radwaste building 1,000 1,000 1,000

Reactor hydrotest and water level reduction to operating state 56,720 0 0 RHR system flush water 4,000 a 0 Condensate demineralizer backwash 27,000 13,500 b 40,500 c Cleanup demineralizer backwash 2,430 1,215 d Water inleakage to condenser 0 0 14,400 Total 100,491 14,341 104,826 a Occurs every shutdown prior to placing the RHR system in operation for shutdown cooling.

b Under normal operating conditions, one condensate filter demineralizer would be precoated every 4 days. c The maximum daily flow is based on a main condenser inleakage of 10 gpm, which corresponds to 3 condensate demineralizer precoatings daily and maximum leak and drain inflows. Up to 36 gpm of condenser inleakage can be accommodated. This requires precoating of one condensate demineralizer every 3 hr. This inleakage rate would result in overloading the equipment drain subsystem but could be tolerated for short periods of time during location and repair of the leak.

d Under normal operating conditions, each cleanup demineralizer may be precoated every 6.8 days.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-25 Table 11.2-9

Floor Drain Subsystem Sources Source Regular Daily Flows (gpd)

Irregular Flows (gpd) Maximum Daily Flows (gpd)

Floor drains Drywell 700 28,800 Reactor building 2,000 15,000 Radwaste building 1,000 1,000 Turbine bui l d ing 2,000 2,000 Waste sludge phase separator decant 0 8,48 9 a 8,489 Total 5,700 8,489 55,289 a Under normal operating conditions, the waste sludge phase separator tank is decanted every 3-4 days.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-26 Table 11.2-10

Chemical W a s t e Subsys t e m Sources Source Regular Daily Flow (gpd)

Irregular Flow (gpd) Maximum Daily Flow (gpd)

Detergents drains/shop decontamination solutions 1,000 2,000 Laboratory drains 500 500 Decontamination drains reactor and turbine bui l d ings 1,000 1,000 From floor drain or equipment

drain subsystem 15,000 15,000 Filter demi n e ralizer chemical cleaning solutions Infrequent 2,000 2,000 Battery room drains Infrequent 100 100 Total 1,500 20,600 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-27 Table 11.2-11 Radwaste System Process Flow Diagram Data Equipment Drain Subsystem Flow path 1 2 3 4 5 6 8 Batches/day (normal) 8.5 4.1 1.1 6.3 1.0

/6.8 4.0/7.4 1.0/30.0 Batches/day (maximum) 63.4 15.8 1.1 6.3 1.0 4.0/2.0

- Volume/batch (gal) 455 909 909 909 2430 13,500 4000 Normal daily volume (ga l) 3860 3755 1000 5726

- - - Normal act i v ity ( Ci/c m 3) 5.16E-2 1.30E-2 7.18E-5 1.91 E-5 1.59E-2 7.18E-6 9.22E-7 Maximum daily volume (gal) 28,8 0 0 14,400 1000 5726 2430 27,000 4000 Maximum activity

( Ci/c m 3) 1.72E0 4.32E-1 2.39E-3 1.26E-2 5.32E-1 3.59E-4 3.08E-5 Flow rate (gal/minu t e) 50 50 50 50 53 450 Batch Daily activity ( Ci/day) Normal 7.19E5 1.76E5 2.59E2 3.94E2 - - -

Maximum 2.51E7 6.14 E 6 9.05E3 2.73E3 4

.89E6 3.67E4 4.66E2

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-28 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Equipment Drain S ubsystem (Continued)

Flow path 9 33 12 14 17 18 19 Batches/day (norm a l) 1.0 1.0 1.0 1.0 1.5 2.2 1.1 Batches/day (maximum) 7.0 7.0 7.0 7.0 63.4 16.5 1.1 Volume/batch (gal) 15,000 15,000 15,000 15,000 455 909 909 Normal daily volume (gal) 15,0 0 0 15,000 15,000 15

, 000 700 2000 1000 Normal act i v ity ( Ci/c m 3) 1.16E-2 5.81E-3 1.16E-4 1.16 E-4 7.18E-6 1.05E-6 1.44E-5 Maximum daily volume (gal) 104,826 104 , 826 104,826 104,826 28,800 15,000 1000 Maximum activity

( Ci/c m 3) 3.59E-1 3.53E-1 3.53E-3 3.53 E-3 6.84E-2 1.00E-3 4.78E-4 Flow rate (gal/m i nute) 190 190 190 190 50 100 50 Daily activity ( Ci/day) Normal 6.30E5 3.14E5 6.30E3 6.30E3 1.81E1 7.57E0 5.18E1 Maximum 2.04E7 2.00 E 7 2.03E5 2.04E5 1

.81E5 7.57E3 1.81E3

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-29 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Equipment Drain S ubsystem (Continued)

Flow path 20 117 21 23 107 108 Batches/day (normal) 2.2 10./3.4 1.0/2.6 1

.0/2.6 1.0/2.6 1.0/2.6 Batches/day (maximum) 2.2 1.0 3.7 3.74 3.7 3.7 Volume/batch (gal) 909 8489 15,000 15,000 15,000 15,000 Normal daily volume (ga l) 2000 5700 5700 5700 5700 Normal act i v ity ( Ci/c m 3) 7.18E-7 7.18E-6 1.00E-6 4.99E-5 1.00E-6 1.00E-6 Maximum daily volume (gal) 2000 8489 55,489 55,4 8 9 55,489 55,489 Maximum activity

( Ci/c m 3) 2.39E-5 3.59E-4 1.08E-2 1.08E-2 1.08E-4 1.08E-4 Flow rate (gal/minute) 50 50 190 190 190 190 Daily activity ( Ci/day) Normal 5.18E0 - 2.56 E 3 1.03E3 2.06E1 2.06E1 Maximum 1.81E2 - 2.33E 5 2.33E5 2.33E3 2.33E3 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-30 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Waste Surge Subsystem Flow path 104 37 Batches/day (normal) 1.0/yr 1.0/yr Batches/day (maximum) 1.0 1.0 Volume/batch (gal) 56,720 56,720 Normal daily volume (gal) Normal act i v ity ( Ci/c m 3) 7.18E-6 7.18E-6 Maximum daily volume (gal) 56,720 56,720 Maximum activity

( Ci/c m 3) 3.59E-4 3.59E-4 Flow rate (gal/minute)

Batch 190 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-31 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Chemical W a s t e Subsys t e m Flow path 27 109 120 121 122 Batches/day (norm a l) 1.0 Batches/day (maximum) 2.0 2.0 2.0 1.0 3.7 Volume/batch solids(lb) 581 1162 314 liquids (gal) 1000 12,153 12,153 615 168 Normal daily volume (gal) 1000 Normal activity solids

( Ci/batch) liquids ( C i/c m 3) 1.05E-5 2.76E-3 2.71E-3 1.07E-1 1.07E-1 Maximum daily volume (ga l) 2000 24,306 24,306 Maximum activity soli d s ( Ci/ba t c h) liquids ( C i/c m 3) 1.05E 2.71E-3 1.07E-1 1.07E-1 Flow rate (gal/minute) 25 190 10 31 20 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-32 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Waste Sludge Subsyst e m (Condensate Backwash)

Flow path 56 58 60 6 Batches/day (normal) 4.0/7.4 4.0/7

.4 1/18.5 4/7.4 Batches/day (maximum) 4.0 4.0 1 4.0/2.0 Volume/batch solids (lb) 330 330 3300 - liquids (gal) 13,5 0 0 13,500 7527 13,500 Normal daily volume (gal) 7300 7300 - - Normal activity solids

( Ci/batch) 2.20E6 2.20E6 1.26E7

- liquids ( C i/c m 3) 7.18E-6 7.18E-6 7.18E-6 7.18E-6 Maximum daily volume (ga l) 54,000 54,0 0 0 7527 27,000 Maximum activity soli d s ( Ci/ba t c h) 5.26E7 5.26E7 2.67E7

- liquids ( C i/c m 3) 3.59E-4 3.59E-4 3.59E-4 3.59E-4 Flow rate (gal/

minute) 2500 450 20 450 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-33 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Waste Sludge Subsyst e m (Cleanup Backwash)

Flow path 54 59 5 Batches/day (normal) 2.0/6.8 1.0

/60 1.0/6.8 Batches/day (maximum) 2.0 1.0/60 1.0 Volume/batch solids (lb) 29.7 524 - liquids (gal) 1215 1196 2430 Normal daily volume (gal) 360 - - Normal activity solids

( Ci/ba t ch) 2.55E7 1.19E8

- liquids ( C i/c m 3) 1.59E-2 1.59E-2 1.59E-2 Maximum daily volume (gal) 2430 1196 2430 Maximum activity soli d s ( Ci/ba t c h) 7.68E8 1.80E8

- liquids ( C i/c m 3) 5.32E-1 5.32E-1 5.32E-1 Flow rate (gal/minute) 270 20 53 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-34 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Waste Sludge Subsys tem (Spent Resin)

Flow path 64 119 69 71 Batches/day (normal) 1.0/66 1.0/67 1.0/165

- Batches/day (maximum) 1.0/29 1.0/49 1.0/100

- Volume/batch solids (lb) 1539 15 3 9 1539 1539 liquids (gal) 746 746 746 3510 Normal activity solids

( Ci/batch) 2.39E6 6.15E3 1.16E2

- liquids ( C i/c m 3) 7.18E-6 7.18E-6 7.18E-6 7.18E-6 Maximum activity soli d s ( Ci/ba t c h) 2.74E7 2.34E5 1.26E2

- liquids ( C i/c m 3) 3.59E-4 3.59E-4 3.59E-4 3.59E-4 Flow rate (gal/minute) 37 37 37 20 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2 003 11.2-35 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

Equipment Drain Subsystem Flow path 61 63 62 - 65 Batches/day (normal) 1.0 1.0/3.4 1

.0/5.2 - 1.0/3.4 Batches/day (maximum) 2.

9 1.1/1.0 1.0/5.2 - 1.0 Volume/batch solids (lb) 41.36 41.36 59.4 - 220 liquids (gal) 1692 1692 2430

- 501 Normal daily volume (gal) 1692 - - - - Normal activity solids

( Ci/batch) 3.30E4 8.42E1 1

.47E4 - 4.87E4 liquids ( C i/c m 3) 7.18E-6 7.18E-6 7.18E-6

- 7.18E-6 Maximum daily volume (gal) 4906 1861 - - 501 Maximum activity soli d s ( Ci/ba t c h) 3.30E4 8.42E1 1

.47E6 - 7.33E5 liquids ( C i/c m 3) 3.59E-4 3.59E-4 3.59E-4

- 9.88E-4 Flow rate (gal/minute) 376 376 540 - 20 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-36 Table 11.2-11 Radwaste System Process Flow Diagram Data (Continued)

NOTES:

Process Diagram, Figu r e 11.2-1 , forms part of this data.

The following definitions ar e used for this data:

Normal volume - Expected flow during steady state normal operation

Maximum volume - Maximum expected flow during unste ady state operation such as startup, shutdown, etc.

Normal activity - Activity level expe cted during operation with no fuel leak s and corrosion product reactor water activity concentration of 0.1 Ci/cm 3 Maximum activity - Activity level expected during operation with fuel leak rate equivalent to reactor water activity concentration of 2.3 Ci/cm 3 and design basis noble radi ogas release rate of 100,000 Ci/sec (corrosion and fission products present)

Maximum volume and maximum activity are not concurrent.

For activity values: E-1 = number x 10

-1; E1 = number x 10 1 E-4 = number x 10

-4; E4 = number x 10 4 Fractional values on tables denote the number of items per occurrence divided by the number of days between each occurrence (i.e., 1/30 batches/day mean s one batch processed every 30 days).

Waste system input activities are based on a reactor water-to-steam partition coefficient of 1.0E-3.

Values for design purposes only. Actu al flows and activities may differ.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.2-37 Table 11.2-12 Radwaste Process Equipment Desi g n Bas i s Equipment Radioactivity Decontamination Factor Deep bed demineralizers Soluble Insoluble 100 50 Precoat fi l t ers Soluble Insoluble 1 2 Evaporators (influent/distillate

)a 1000 a Concentration factor (i nfluent/bottoms) = 0.6/32.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-38 Table 11.2-13 Liquid Radwaste Equipment Material of Construction Quantity Size or Capacity Remarks EQUIPMENT DRAIN SUBSYSTEM Waste collector tank Carb on steel 1 20,000 gal Note 11 Waste collector p u mp Stainless steel 1 190 gal/min u te @ 155 ft T DH Waste sample tanks Carb on steel 2 20,000 gal Note 11 Waste sample pumps Stainless steel 2 190 gal/min u te @ 100 ft T DH Waste collector filter Carbon steel 1 188 f t 2 of filter area Notes 3, 4 Waste filter hold pump Stainless ste e l 1 75 gal/minu t e @ 50 ft TDH Note 3 Waste demineralizer Carbon steel 1 65 f t 3 resin b e d Note 2 Waste surge tank Carb on steel 1 75,000 gal Note 11 Waste surge pump Stainless steel 1 190 gal/min u te @ 155 ft T DH FLOOR DRAIN SUBSYSTEM Floor drain collector tank Carbon steel 1 20,000 gal Note 11 Floor drain collector pump Stainless steel 1 190 gal/minute @ 155 ft TDH Floor drain sample tank Carbon steel 1 20,000 gal Note 11 Floor drain sample pump Stainless steel 1 190 gal/minute @ 100 ft TDH Floor drain filter Carbon steel 1 188 ft 2 of filter area Notes 3, 4 Floor drain filter hold pump Stainless steel 1 75 gal/minute @ 50 ft TDH Note 3 Floor drain demineralizer Carbon steel 1 65 ft 3 resin bed Note 2

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-39 Table 11.2-13 Liquid Radwaste Equ i pment (Continued)

Material of Construction Quantity Size or Capacity Remarks CHEMICAL WASTE SUBSYSTEM Detergent d r ain tank Carb on steel 2 1600 gal Note 12 Detergent d r ain pumps Stainless steel 2 27 gal/minu t e @ 90 ft T D H Detergent drain filter Carbon steel 1 50 gal/minu t e Chemical waste tanks Stainless steel 2 15,000 gal Note 12 Chemical waste pumps Stainless s t eel 2 285 gal/min u te @ 200 ft T DH Concentrator feed pumps Stainless steel 2 30 gal/minu t e @ 95 ft T D H Concentrator (evap o rator) Stainless steel 2 10 gal/minu t e Note 6 Heating element Stainless steel/carbon steel 2 Note 10 Concentrator recycle p u m p Stainless s t eel 2 2300 gal/minute @ 40 ft T DH Note 5 Concentrator conden s er Stainless steel/carbon steel 2 Note 7 Distill a te tanks Stainless ste e l 2 15,000 gal Note 12 Distill a te pumps Stainless ste e l 2 142 gal/minute @ 230 ft T DH Polishing demineralizer Carb o n steel 1 65 f t 3 resin b e d Note 2 AUXILIARY EQUIPMENT Waste pre c o a t tank Carb on steel 1 210 gal Notes 4, 8 Waste precoat pump Ductile iron 1 325 gal/minute @ 50 ft TDH Note 3 Waste filter aid tank Carbon steel 1 630 gal Notes 4, 8 Waste filter aid metering pump Stainless ste e l 2 0 to 154 gal/hr @ 145 psi Resin addition tank Stainless steel 1 1200 gal Note 12 Chemical ad d ition tanks Carb o n steel 2 200 gal Note 9 Chemical ad d ition pumps Carpent e r 20 steel 2 79 gal/h r C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 11.2-40 Table 11.2-13 Liquid Radwaste Equ i pment (Continued)

NOTES: 1. Unless otherwise noted, the d e sign pressure of the equipment is 150 psig, design temperature is 15 0 F. 2. Vessel is rubber lined, internals are s t ainless steel.

3. Design temperature is 22 0 F. 4. Vessel is lined with phenolic/

e poxy cross linked, alkaline polymerized c o ating, internals are stainless steel.

5. Design temperature is 30 0 F. 6. Design pressure is 30 psig, design temperature is 27 4 F. 7. Design flow is:

coolant inlet-750 gpm @ 10 5 F, outlet -12 0 F; design pressure: shell side - 30 psig, tube side - 150 psig; design temperature - 27 4 F. 8. Design pressure is atmospheric.

9. Ves s el i s li n ed with aci d-r e sistant high-bake ph e nolic coating.

10. Design flows: shell side - 6300 lb/hr @ 27 4 F, tube side

- 2300 gpm @ 21 8 F; design pressures: shell side -

60 psig, tube side - 75 p s ig; design temperature - 30 7 F. 11. Tank design pressure is a t mospheric plus static head p r essure, design temperature is 14 0 F. 12. Tank design pressure is a t mospheric plus static head p r essure, design temperature is 15 0 F.

11.2-41 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 Table 11.2-14 Annual Releases of Radi o active Material as Liquid Nuclide Half-Life (days) Concentrat i on in Primary Coolant

( C i/ml) High Purity (Ci) Low Purity (Ci) Total Laws (Ci) Adjusted Total (Ci/y r)a Total (Ci/y r) CORROSION AND ACTIVATION PRODUCTS 24 Na 6.25E-1 8.79E-3 0.00036 0.00043 0.00079 0.00688 0.00692 32P 1.43E1 2.06E-4 0.00001 0

.00002 0.00003 0.00027 0.00027 51 Cr 2.78E1 5.14E-3 0.00027 0.00052 0.00081 0.00704 0.00703 54Mn 3.03E2 6.18E-5 0.00000 0

.00001 0.00001 0.00008 0.00008 56Mn 1.07E-1 4.28E-2 0.00052 0.00034 0.00085 0.00747 0.00745 55Fe 9.50E2 1.03E-3 0.00005 0

.00010 0.00017 0.00143 0.00147 59 Fe 4.50E1 3.08E-5 0.00000 0.00000 0.00000 0.00004 0.00004 58Co 7.13E1 2.06E-4 0.00001 0

.00002 0.00003 0.00028 0.00028 60Co 1.92E2 4.12E-0 0.00002 0

.00004 0.00006 0.00058 0.00058 65 Ni 1.07E-1 2.57E-4 0.00000 0.00000 0.00000 0.00004 0.00004 64 Cu 5.33E-1 2.91E-2 0.00111 0.00128 0.00239 0.02098 0.02098 65 Zn 2.45E2 2.06E-4 0.00001 0.00002 0.00003 0.00028 0.00028 69mZn 5.57E-1 1.94E-3 0.00007 0.00009 0.00017 0.00146 0.00147 69Zn 3.96E-2 0.0 0.00008 0

.00009 0.00018 0.00153 0.00157 187W 9.96E-1 2.98E-4 0.00001 0

.00002 0.00003 0.00028 0.00028 239Np 2.35E0 7.10E-3 0.00036 0

.00060 0.00094 0.00831 0.00790 FISSION PRODUCTS 83Br 1.00E-1 2.42E-3 0.00003 0.00002 0.00004 0.00039 0.00039 84Br 2.21E-2 3.67E-3 0.00000 0.00000 0.00000 0.00003 0.00003 89Rb 1.07E-2 3.60E-3 0.00001 0.00002 0.00002 0.00022 0.00022 89 Sr 5.20E1 1.03E-5 0.00001 0.00 0 01 0.00002 0.00015 0.00015 91Sr 4.03E-1 3.82E-3 0.00013 0.00014 0.00026 0.00233 0.00231

11.2-42 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORTDecember 2003 Table 11.2-14 Annual Releases of Radioactive Material as Li quid (Continued)

Nuclide Half-Life (days) Concentration in Primary Coolant (Ci/ml) High Purity (Ci) Low Purity (Ci) Total Laws (Ci) Adjusted Total (Ci/y r)a Total (Ci/yr) 91m Y 3.47E-2 0.0 0.00008 0.00008 0.00017 0.00146 0.00147 91 Y 5.88E1 4.12E-5 0.00000 0.00001 0.00001 0.00007 0.00007 92 Sr 1.13E-1 8.60E-3 0.00012 0.00007 0.00018 0.00159 0.00157 92 Y 1.47E-1 5.29E-3 0.00021 0.00016 0.00038 0.00328 0.00325 93 Y 4.25E-1 3.83E-3 0.00014 0.00014 0.00027 0.00241 0.00241 98 Nb 3.54E-2 3.11E-3 0.00001 0.00000 0.00001 0.00008 0.00008 99 Mo 2.798E0 2.04E-3 0.00010 0.00018 0.00028 0.00244 0.00241 99m Tc 2.5E-1 1.85E-2 0.00054 0.00054 0.00107 0.00893 0.00937 101Tc 9.72E-3 6.56E-2 0.00000 0.00000 0.00000 0.00002 0.00002 103Ru 3.96E1 2.06E-5 0.00000 0.00000 0.00 0 00 0.00003 0.00003 103mRh 3.96E-2 0.0 0.00000 0

.00000 0.00000 0.00003 0.00003 104Tc 1.25E-2 5.87E-2 0.00000 0.00000 0.00001 0.00006 0.00006 105Ru 1.85E-1 1.80E-3 0.00004 0.00003 0.00006 0.00055 0.00058 105mRh 5.21E-4 0.0 0.00004 0

.00003 0.00006 0.00056 0.00059 105Rh 1.50E0 0.0 0.00001 0.00 0 01 0.00002 0

.00018 0.00019 129mTe 3.40E1 4.11E-5 0.00000 0

.00000 0.00001 0.00005 0.00005 129Te 4.79E-2 0.0 0.00000 0

.00000 0.00000 0.00003 0.00003 131mTe 1.25E0 1.00E-4 0.00000 0

.00001 0.00001 0.00010 0.00010 131Te 1.74E-2 0.0 0.00000 0

.00000 0.00000 0.00002 0.00002 131I 8.05E0 5.11E-3 0.00027 0

.00049 0.00077 0.00643 0.00671 132Te 3.25E0 1.02E-5 0.00000 0

.00000 0.00000 0.00001 0.00001 132I 9.58E-2 2.41E-2 0.00026 0

.00016 0.00041 0.00345 0.00367 133I 8.75E-1 1.93E-2 0.00084 0

.00115 0.00199 0.01667 0.01783 134I 3.67E-2 5.26E-2 0.00010 0

.00006 0.00017 0.00144 0.00147

11.2-43 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 Table 11.2-14 Annual Releases of Radioactive Material as L i quid (Continued)

Nuclide Half-Life (days) Concentrat i on in Primary Coolant

( C i/ml) High Purity (Ci) Low Purity (Ci) Total Laws (Ci) Adjusted Total (Ci/y r)a Total (Ci/y r) 134 Cs 7.49E2 3.09E-5 0.00008 0.00081 0.00089 0.00741 0.00776 135I 2.79E-1 1.77E-2 0.00050 0

.00044 0.00094 0.00788 0.00829 136Cs 1.30E1 2.05E-5 0.00005 0

.00051 0.00057 0.00473 0.00493 137Cs 1.10E4 7.21E-5 0.00020 0

.00188 0.00206 0.01730 0.01783 137M Ba 1.77E-3 0.0 0.00018 0.00175 0.00193 0.01618 0.01678 138Cs 2.24E-2 7.34E-3 0.00022 0.00065 0.00087 0.00724 0.00755 139Ba 5.76E-2 8.09E-3 0.00004 0.00002 0.00006 0.00053 0.00056 140Ba 1.28E1 4.11E-4 0.00002 0

.00004 0.00006 0.00053 0.00056 140 La 1.67E0 0.0 0.00000 0.00001 0.00001 0.00011 0.00011 141La 1.62E-1 0.0 0.00001 0

.00001 0.00002 0.00017 0.00018 141Ce 3.24E1 3.08E-5 0.00000 0

.00000 0.00000 0.00004 0.00004 142La 6.39E-2 4.08E-3 0.00003 0.00002 0.00004 0.00036 0.00038 143Ce 1.38E0 3.01E-5 0.00000 0

.00000 0.00000 0.00003 0.00003 143Pr 1.37E1 4.11E-5 0.00000 0

.00000 0.00001 0.00005 0.00005 All others 1.32E-2 1.38E-2 0.00000 0.00001 0

.00007 0.00007 Total (exce p t tritium) 4.31E-01 0.00719 0.01307 0.02026 0.17962 0.17834 Total release 13.0 13.0 a Adjusted total includes an additional 0.15 ci/yr with the same isotopic distribution as the calculated sour ce term to account for anticipated occurrences such as operator errors resulting in unplanned releases.

Draw. No.

Rev. F i gure 11.2-1 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report11.2-2 74 M532Flow Diagram Radioactive Waste System Equipment Drain ProcessingRev.FigureDraw. No.Amendment 61December 2011 Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report11.2-3 75 M531Flow Diagram Radioactive Waste System Floor Drain ProcessingRev.FigureDraw. No.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-069 11.3-1 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS

11.3.1 DESIGN BASES The objective of the gaseous waste management system is to pro cess and control the release of gaseous radioactive effluents to the site environs so as to maintain as low as reasonably achievable the exposure of persons in unrestric ted areas to radioactiv e gaseous effluents, during normal and anticipated operational occurrences. This is to be accomplished while maintaining occupational exposure as low as is reasonably achievable (ALARA) and without limiting plant operation or availability.

The gaseous waste management systems are designed to limit the dose to offsite persons from routine station releases to significantly less than the limits specified in 10 CFR 20 and to operate within the emission rate limits es tablished in the Technical Specifications.

To evaluate the offgas system design compliance with the li mits of 10 CFR 20, an annual average noble radiogas source term (bas ed on 30-minute de cay) of 100,000

µCi/sec of the "1971 Mixture" as di scussed in Section 11.1 is used. The noble radi ogas effluent release rate from the charcoal ad sorbers is about 49-59

µCi/sec based upon 30 scfm air inleakage and injection. The isotopic composition is given in Table 11.3-1 in units of

µCi/sec and Ci/yr.

To evaluate that the annual average exposure at the site boundary during normal operation and anticipated operational occurrences from gaseous effluents doe s not exceed the dose objectives of Appendix I to 10 CFR 50, an average sour ce term release rate derived from field measurements (1993-200

3) is used.

Equipment and components used to collect, process, or store gaseous radioactive waste are not designed as Seismic Category I. To evaluate that equipmen t failure will not result in offsite whole body doses exceedi ng 0.5 rem in two hours, the Technical Specification release rate limit was used, consistent with the guidance presented in Reference 11.3-7.

The gaseous radwaste equipment is selected, arranged, and shie lded to maintain occupational exposure ALARA. The system was designed prior to the issuance of Regulatory Guide 8.8. However, the system incorporates substantially the guidance provide d in this regulatory guide.

The gaseous effluent treatment system conf orms to the requireme nts of General Design Criteria 60 and 64 as specified in Section 3.1. A list of the major offgas system compone nts and design featur es is provided in Table 11.3-2. Equipment and piping is designe d and constructed in accordance with the requirements of the applicable codes as given in Tables 3.2-1 and 3.2-2. The quality group classifications of the various systems are shown in Table 3.2-1. Seismic category, safety class, qua lity assurance requirements, a nd principal construction codes

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-069 11.3-2 information are contained in Section 3.2. The system is designed to Quality Group Classification D+.

11.3.2 SYSTEM DESCRIPTION

11.3.2.1 Main Condenser Steam Jet Air Ejector RECHAR System

The offgas from the main condenser steam je t air ejector (SJAE) is treated by means of a system utilizing catalytic recombination and charcoal adsorption (RECHAR system) (see Figure 11.3-1). Descriptions of the major process components including design temperature and pressure are given in Table 11.3-2 and in the following.

Noncondensable radioactive offg as is continuously removed from the main condenser by the air ejector during plant operati on. The air ejector offgas will normally contain activation gases, principally 16 N, 19 O, and 13 N. The 16 N and 19 O have short half-lives and are readily decayed.

13 N with a 10-minute half-life is present in small amounts that are further reduced by decay. Activation gas source terms are presented in Table 11.1-4. The air ejector offgas will also contain radioac tive noble gases including parents of biologically significant 89 Sr, 90 Sr, 140 Ba, and 137 Cs. The concentration of th ese noble gases depends on the amount of tramp uranium in the coolant and on th e cladding surfaces (usu ally extremely small) and the number and size of fuel cladding defects.

A main condenser offgas treatmen t system has been in corporated in the plant design to reduce the radioactive gaseous effluents from the station. The offgas system uses a catalytic recombiner to recombine hydrog en and oxygen. After cooling (to approximately 130°F) to strip the condensables and redu ce the volume, the remaining nonc ondensables (principally air with traces of krypton and xenon) are delayed in a holdup line. Th e gas is cooled to 45°F and processed through a HEPA filter. The gas is then passed th rough a desiccant dryer that reduces the dewpoint to approxima tely -90°F. Charcoal adsorpti on beds selectively adsorb and delay the xenon and krypton from the bulk carrier gas (principally dry air). With an air inleakage of 30 scfm, this treatment system results in a delay of 15 hr for krypton and 9.5 days for xenon. After the delay, the gas is again passed through a HE PA filter and discharged to the environment through the reactor building elevated release duct.

Figure 11.3-1 is the process flow diagram for the system. The process da ta for startup and normal operating conditions are submitted as proprietary data under separate cover as Table 11.3-3. The information supporting the pro cess data is presented in Reference 11.3-2. The system is mechan ically capable of processing th ree times the source terms of Table 11.3-1 without affecting delay time of the noble gases.

Table 11.3-1 also lists isotopic activitie s at the discharge of the system, from which the decontamination factor for each nobl e gas isotope can be determined.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-069 11.3-3 The flow diagram is shown in Figure 11.3-2. The main process routing is indicated by a heavy line.

The basis for sizing the recomb iner is to maintain the hyd rogen concentration below 4% (including steam) at the inlet and below 1% at the outlet on a dry basis. The exit hydrogen concentration is normally well below the 1% ma ximum allowed. The hydrogen generation rate of the reactor is based on data from ni ne boiling water reactors (BWR). The hydrogen generation rate is given in Table 11.3-3.

The krypton and xenon holdup time is closely approximated by the following equation:

T KM V D= where:

T = holdup time of a given gas, seconds.

K D = dynamic adsorption coefficient for the given gas, cmgram 3 M = weight of charcoal, grams V = flow rate of the carrier gas, cmsec 3 Dynamic adsorption coefficient values for xenon and krypton used to determine gaseous effluent releases are discussed in Reference 11.3-1. Moisture has a detrimental effect on adsorption coefficients. To pr event moisture from reaching the charcoal, fully redundant, adsorbent (-90°F dewpoint) air dryers are suppl ied. There are redundant moisture analyzers that will alarm on breakthrough of the dryer beds

however, breakthrough is not expected since the dryer beds will be regenerated on a time basis. The system is slightly pressurized, which together with very stringent leak rate require ments, prevents leakage of moist air into the charcoal.

After the hydrogen and oxygen are removed by the recombiner, th e remaining carrier gas is primarily nitrogen with a small percentage of oxygen due to the air inleakage from the main condenser and air injection from the HWC system. Analyses have show n that the system can maintain condenser vacuum at combined air injection and air inleak age rates as high as 93 scfm. Reference 11.3-3, Par. S1 (c) (2) indicates that with certain conditions of stable operation and suitable constructi on, noncondensables (not including radiological decomposition products) should not exceed 6 scfm per condenser shell for large condensers. Dilution air is not added to the system unless the air inleakage is less than 6 scfm. In that event, 6 scfm is C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-4 added to provide for dilution of residual hydrogen from the reco mbiner. An initial bleed of oil-free air is added on startup until the recombiner reaches operating temperature.

The charcoal adsorbers design proc ess flow sheet is for an ambient temperature. Operation at ambient temperature is sufficient to reduce gaseous radioactivity levels to a fraction of that allowed by 10 CFR 50 Appendix I. The decay heat is sufficiently small that , even in the no-flow condition, there is no significant loss of adsorbed noble gases due to temperature rise in the adsorbers. The adsorbers are located in a shielded room and maintained at ambient room temperature. A radiation monitor is prov ided to monitor the radi ation level in the air handling room of the charcoal bed vault for high gas activity from a system leak. High radiation will cause an al arm in the control room.

Channeling in the charcoal adsorbers is prevented by supplying an effective flow distributor on the inlet, having long columns and having a high bed-to-particle diameter ratio of approximately 500. Underhill has stated that ch anneling or wall effects may reduce efficiency of the holdup bed if this ratio is not greater than 12 (Reference 11.3-4). During transfer of the charcoal into the charco al adsorber vessels, radial sizing of the charcoal will be minimized by pouring the charcoal (by gravity or pneumatically) over a cone or other instrument to spread the granules over the surface.

A valve is provided to bypass the charcoal adsorb ers. The main purpose of this bypass is to protect the charcoal during preoperational and startup testing when gas activity is zero or very low.

It may be desirable to use the bypass for short periods during startup or normal operations. This bypass mode would not be used for norma l operation unless some unforeseen system malfunction would necessitate shutting down the power plant or operating in the bypass mode and remaining within the technical specification radioactivity release limits. The radioactivity released is controlled by a pro cess monitor upstream of the vent isolation valve that will cause the bypass valve to close on a high radiation alarm. This interl ock can be defeated only by a key lock switch. In addition, there is a high-high-high alarm setpoint on the same monitor that will cause isolation of the offg as system if established release rate limits are reached.

Leakage of radioactive gases from the system is limited by welding piping connections where possible and using bellows stem seals or equivalent leakage c ontrol valving. The system operates at a maximum of 7 psig during startup and less than 2 psig during normal operation so that the differential pressure to cause leakage is small.

Hydrogen concentration of gases from the air ejector is kept below the flammable limit by maintaining adequate process steam flow for dilution at all times. This steam flow rate is monitored and alarmed in the main control room.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-069 11.3-5 Two parallel independent hydrogen analyzers are used to measure the hydr ogen content of the offgas process flow downstream of the o ffgas condenser. The hydrogen concentration percentage output from the analyzers is indicated and recorded in the main control room along with independent alarm annunciation for a high hydrogen concentration. Each hydrogen analyzer continuously withdraw s a sample of the process o ffgas, conditions the gas to a constant pressure, analyzes the hydrogen content, and returns the sample gas to the main condenser. The main condenser vacuum provide s the pumping force to withdraw the sample gas from the offgas process line and through the hydrogen anal yzer system. The analyzer element is a thermal conductivity cell type unit and does not serv e as an ignition source to a detonable hydrogen-oxygen mixture.

Piping and tubing 2 in. a nd under is field routed, but required to be in specified space envelopes for shielding and in-pla nt exposure considerati on. In the offgas system this includes drain lines, steam lines, and sample lines which are shown in Figure 11.3-2.

There are several liquid se als to prevent gas escape through drains shown in Figure 11.3-2. These seals are protected against permanent loss of liquid by an enlarged section downstream of the seal that can hold the seal volume and w ill drain by gravity back into the loop after a momentary pressure surge has passed. Each seal has a manual valve that is used to fill the loop with condensate after receivi ng a loop seal low level alarm.

Iodine input into the offgas sy stem is small by virtue of its retention in reactor water and condensate. The iodine remaining is essentially removed by adsorption in the charcoal. This is supported by the fact that 2-in.

charcoal filters remove more than 90% of the influent iodine, whereas this system has a pproximately 76 ft of charcoal in the flow path.

Particulates are removed with a 99.95% efficien cy by a HEPA filter prior to the gas entering the charcoal adsorb ers. The noble gas decays within the interstices of the activated charcoal and daughters are entrapped there.

The charcoal serves as an excellent filt er for particulates and essentially no particulates ex it from the charcoal.

The charcoal is fo llowed with a HEPA filter which is a safeguard against escape of charcoal dust. Particulate activity discharged from this system is essentially zero.

With an airflow of about 30 sc fm and the charcoal adsorbers bypassed, the delay time of the system decreases to approximate ly 10 minutes. This bypass line is only intended to be used during preoperational te sting, and initial system startup operation until prope r functioning of upstream equipment is established.

This prevents possible degradation of the charcoal due to the introduction of excessive mois ture or other contaminants.

In the unlikely event it is necessary to bypass the ch arcoal adsorbers to c ontinue operation, the bypassing operation shall be allowed only if the radioactive effluents ar e within the release limits.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-069 11.3-6 The isotopic inventory of each equipment piece, based on th e source term discussed in Section 11.1 , is given in Reference 11.3-5.

Performance of a similar system operating at ambient temperatur es and the resu lts of related experimental testing are discussed in Reference 11.3-2. The Tsuruga and Fukushima 1 plants in Japan have similar recombiners in service. Similar systems (ambient temperature charcoal) are in service at Dresden 2 a nd 3, Pilgrim, Quad Cities 1 a nd 2, Nuclenor, Hatch, Browns Ferry 1, 2, and 3, and Duane Arnold.

Design provisions are incorporated which preclude the uncontrolled release of radioactivity to the environment as a result of any single ope rator error or of a ny single failure. A comprehensive discussion of si ngle failures is provided in Table 11.3-5. Design precautions taken to prevent uncontrolled releases of radioactivity include the following:

a. The system design eliminates ignition sources so that a hydrogen detonation is highly unlikely even in the ev ent of a recombiner failure;

b. The system pressure boundary is detonation resistant;

c. All discharge paths to th e environment are monitored--the normal effluent path by the process radiation monitoring sy stem and equipment areas by the area radiation monitoring system; and

d. Dilution steam flow to the SJAE is monitored and alarmed. Valve control logic causes the air ejector suction valves to close on low steam flow.

11.3.2.2 Other Radioactive Gas Sources There are three build ings that contain radioactive gas sources. They are the reactor building, the turbine generator building, and the radwaste building. The ventila tion systems for these three buildings are described in Section

9.4. Building

volumes and ventilation flow rates are shown in Table 11.3-8. The sources of gaseous radioactiv ity in these buildings are discussed below. In-plant airborne radioactivity concentrations are discussed in Section 12.2.2. The primary containment is divided into two s ections which are designa ted as the drywell and suppression chamber. These are separated by the drywell floor which serves as a pressure barrier between the drywell and suppression chamber.

Radioactive halogens a nd noble gases can be introduced in to the drywell atmosphere from two sources. One source is the le akage that occurs from the valves, especially the main steam isolation valves (MSIVs), the inner refueling bellows seal support drains and the main C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-7 recirculation pumps. This leakage is collected by means of leak-off lines which are directed to the drywell equipment drain sump.

The other source of activity results when a pr essure transient causes the main steam relief valves to open with the resulting flow of steam into the suppression pool. When this occurs, there is a pressure buildup in the suppression chamber atmos phere. A tabulation of the expected relative frequency of pressure relief valve venting to the suppression pool is provided in Table 11.3-10. If the pressure differential across the drywell floor is greater than 0.5 psi, the vacuum breaker valves relieve this pre ssure into the drywell atmosphere. Thus, radioactivity in the suppression chamber atmosphere is introdu ced into the dr ywell. The drywell atmosphere is purged to the environment via the reactor building elevated release duct when access is required, eith er directly or through the standby gas treatment system, considering airborne radiati on levels and release limits.

The reactor building ventilation system supplies fresh air to the secondary containment and exhausts air through the reactor building elevated release duct. There is a small amount of activity in the secondary containment atmosphere that emanates from the various reactor support systems. The reactor bu ilding ventilation exhaust is monitored and a radiation level, set by administrative control to ensure Technical Specifications compliance, will cause automatic heating, ventilating, and air-conditioning (HVAC) system isolation, startup of the standby gas treatment system (SGTS), and an alarm in the control room.

There are three sources of nontr eated radioactive gas sources in the turbine generator building which are as follows:

a. Gland seal steam leakag e and condenser exhaust, b. Equipment leakage, and

c. Main condenser o ffgas during startup.

The activity from these sources is presented in Table 11.3-7.

Gland seal leakage and condenser exhaust cont ribute essentially no airborne radioactivity releases due to use of "clean" steam as discus sed in Section 10.4. During startup, the mechanical vacuum pumps remove the gas pres ent inside the main c ondenser. The exhaust from these pumps is discharged through the react or building elevated re lease duct. Due to radioactive decay during shutdown, only a small amount of activity is exhausted by the vacuum pumps during startup.

Sources of gaseous radioactivity in the radwaste building include,

a. Offgas system leakage,

b. Liquid leakage to the radwaste building, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-8 c. Liquid waste management system tank vents, and d. Hydropneumatic transfer of resins.

Measures taken to minimize leakage from the offgas system are described in Section 11.3.2.1. Liquid leakage from equipment and floor drains in the radwaste building is collected in sumps.

The sumps and tanks containing liq uid radwaste are vented to the radwaste building exhaust system described in Section 9.4.3.2. The air blow during the backwash of the spent powdered resins to the phase sepa rators generates airborne radioactivity which is vented to the radwaste building filtered exhaust system.

11.3.2.3 Cost-Benefit Analysis The cost-benefit analysis is discussed in Section 11.2.3.4. 11.3.2.4 Design Features of the Offgas System Design features of other ga seous waste management syst ems may be found in Section 9.4. 11.3.2.4.1 Maintainability

Design features which reduce or ease requi red maintenance incl ude the following:

a. Redundant components for all activ e, in-process equi pment pieces, and

b. No rotating equipment in the process stream and elsewhere in the system only where maintenance can be performed while the sy stem is in operation.

11.3.2.4.2 Pressure Boundaries Design features and requirements which reduce leakage and releases of radioactive material include the following:

a. Extremely stringent leak rate requireme nts placed on all equi pment, piping, and instruments, and enforced by requiring as-installed leak tests of the entire process system,

b. Use of welded joints except wher e servicing access to equipment and instrumentation access dictates use of flanged joints,

c. Valve types with extremely low leak rate characteristics, i.e., bellows seals double stem seal, or diaphragm, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-9 d. Use of loop seals with en larged discharge section to avoid siphoning and to be self-refilling by gravity follo wing a pressure surge, and

e. Stringent seat-leak characteristics for valves in lines discharging to the environment via other systems.

11.3.2.4.3 Building Seismic Design

The offgas system is located in the turbine generator and radw aste buildings. The portion of the turbine generator building housing the offgas system is a modified Seismic Category II structure designed to withstand the effects of a safe shutdown earthquake (SSE) and maintain its structural integrity. The portion of the radwaste building housing the offgas system is a Seismic Category I structure. The seismic classi fication of the turbine generator and radwaste buildings are discussed in Sections 3.8.4.1.3 and 3.8.4.1.2, respectively.

11.3.2.4.4 Construction of Process Systems

Pressure retaining components of process systems utilize welded construction to the maximum practicable extent. Process piping systems include the firs t root valve on sample and instrument lines. Process lines are not less than 0.

75 in. nominal pipe size. Sample and instrument lines are not consider ed as portions of the process systems. Flanged joints or suitable rapid disconnect fittings are not used except where ma intenance requirements clearly indicate that such construction is preferable. Screwed connections in which threads provide the only seal are not used. Screwed connecti ons backed up by seal we lding or mechanical joints are used only on lines of 0.75 in. nominal pi pe size or less. In lines 0.75 in. or greater, but less than 2.5 in. nominal pipe size, socket type welds are us ed. In lines 2.5 in. nominal pipe size and larger, pipe we lds are of the butt joint type.

11.3.2.4.5 Instrumentation and Control

The offgas system is monitored by flow, temperature, pressure, and humidity instrumentation, and by hydrogen analyzers to ensure correct operation and control.

Table 11.3-4 lists the process parameters that are instrumented to alarm in the control room and indicates whether the parameters are record ed or just indicated.

The radioactivity of the gas ente ring and leaving the offgas system is continuously monitored. Thus, system performance is known to the operator at all times. A radiation monitor after the offgas condenser continuously moni tors radioactivity release from the reactor and input to the charcoal adsorbers. This radiation monitor is used to provide an alarm on high radiation in the offgas. A sample rack with two radiation monitors is also provided at the outlet of the charcoal adsorbers to continuous ly monitor the radio activity from the adso rber beds. These radiation monitors are used to isolate the offgas system on high radioactivity to prevent gas of C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-10 unacceptably high activity from entering the reactor building elevated release duct. Only one monitor is required to be operable.

The offgas system at the SJAE is sampled periodically in accordance with the Technical Specifications. Provision is ma de for sampling and periodic an alysis of the influent and effluent gases for purposes of determining their compositions.

This information is used in calibrating the monitors and in relating the release to calculated offsite doses. Process radiation instrumenta tion is described in detail in Sections 11.5 and 7.6.1.1. 11.3.2.4.6 Detona tion Resistance The pressure boundary of the sy stem is designed to be detona tion resistant. The pressure vessels are designed to withsta nd 350 psig static pressure, and piping and valves are designed to resist dynamic pressures encountered in long runs of piping at the design temperature. This analysis is covered in a proprietary report submitted to the NRC (Reference 11.3-6). By this procedure a designer obtains the required wall thickness for specific equipment to sustain a hydrogen and oxyge n detonation. The wall thickness is then translated by using the appropriate code calculation to the corresponding equipment that must contain the detonation static pressure. The method a ssumes the absence of simultaneous secondary events such as earthquakes.

11.3.2.4.7 Operator Exposur e Criteria and Controls This system is normally operated from the main control room. Equipment and process valves containing radioactive fluid are placed in shielded cel ls maintained at a pressure negative to normally occupied areas. Ventilation air flows from areas of low airborne contamination to areas of higher airborne contamination. Operating offgas process e quipment does not require personnel access. Redundant equi pment is located in separate cells, minimizing exposure of maintenance personnel. No pro cess fluid is passing through in strumentation pa nels. Signals from the process streams are tr ansmitted to the instrumentation panels by means of electrical signal converters. Design f eatures minimizing occupationa l exposure are discussed in Sections 11.3.2.4.1 and 12.3.1.3.

11.3.2.4.8 Equipment Malfunction Malfunction analyses indicati ng consequences and design precautions taken to accommodate failure of various components of the system are given in Table 11.3-5.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-11 11.3.2.5 Offgas System Operating Procedure 11.3.2.5.1 Prestartup Preparations Prior to starting the main condens er SJAEs, the glycol coolant is chilled to near 35°F and is circulated through the cooler condenser, a desiccant dryer is regenerated and valved in, the offgas condenser cooling water is valved in, and the recomb iner heaters ar e turned on.

11.3.2.5.2 Startup As the reactor is pressurized, preheater steam is supplied and air is bled through the preheater and recombiner. The recombiner is preheated to at least 300

°F with this air bleed and/or by admitting steam to the final SJAE. With the recombiners preh eated and the desiccant dryer and charcoal adsorbers valved in, the SJAE string is started. The bleed air is terminated. As the condenser is pumped down and the reactor power increases, the recombiner inlet stream is diluted to less than 4% hydrogen by volume by a fixed steam supply, and the offgas condenser outlet is maintained at less than 1% hydrogen by volume.

11.3.2.5.3 Normal Operation

After startup, the noncondensables pumped by the SJAE will stabilize. Recombiner performance is closely followed by the recorded temperature profile in the recombiner catalyst bed. The hydrogen effluent concentration is measured by a hydrogen analyzer.

Normal operation is terminated following a normal reactor shutdown or a scram by terminating steam to the SJAEs and the preheater.

11.3.2.6 Offgas System Performance Tests This system is used on a r outine basis and does not require specific testing to ensure operability. Monitoring equipment will be calibrated and mainta ined on a specific schedule and on indication of malfunction.

11.3.2.6.1 Recombiner

Recombiner performance is continuously monitored and recorded by catalyst bed thermocouples that monitor the bed temperat ure profile and by a hy drogen analyzer that measures the hydrogen concen tration of the effluent.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-12 11.3.2.6.2 Prefilter

These particulate filters were tested at the time of filter in stallation using di octylphthalate (DOP) aerosol to determine whether an installed filter meets the minimum in-place efficiency 99.95% particle retention.

The DOP from filter testing is not allowed to ente r into the desiccant or the activated charcoal.

This equipment is isolated dur ing filter DOP testing and is bypassed until the process lines have been purged clear of test material. Because the DOP ma y have a detrimental effect on the desiccant and charco al, the prefilter will not be periodi cally tested. Th is is justified because the main function of this prefilter is to prevent the long-lived daughters of the radioactive xenons genera ted in the holdup pipe from depositing in the downstream equipment. Leakage through the filter would be uni mportant to environmental release.

11.3.2.6.3 Desiccant Gas Dryer

Desiccant gas dryer performance is continuously monitore d by an on-stream humidity analyzer.

11.3.2.6.4 Charcoal Performance

The ability of the charcoal to delay the noble gases can be continuously evaluated by comparing radioactivity measured and recorded by the process ac tivity monitors at the exit of the offgas condenser and at the exit of the charco al adsorbers.

Grab sample points are located upstream and downstream of the first charcoal bed and downstream of the last charcoal bed and can be used for periodic sampling if the monitoring equipment indicates degradation of system delay performance.

11.3.2.6.5 Post Filter

On installation these particulate filters were tested using a DOP aerosol test or equivalent, as described in Section 11.3.2.6.2.

11.3.3 RADIOACTIVE RELEASES 11.3.3.1 Release Points

The reactor building elevated release duct serves the following systems:

Offgas system, Mechanical vacuum pump and gl and seal condens er exhaust, Reactor building ventilation exhaust, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.3-13 Containment purge, and Standby gas treatment system.

The reactor building elevated release duct location is shown in reactor building general arrangement drawings in Section 1.2. The radwaste building ventilation exhausts through three louver hous es 67 ft above plant grade.

The location is shown on Figure 1.2-17. The exhaust from the turbine gene rator building ventilation system is through four exhaust fans located on the radwaste buildi ng, 119 ft above plant grade.

Their location is shown on Figure 1.2-16 (part plan of roof at el. 542 ft-0 in.) and in Figure 1.2-17. The height, flow rate, heat c ontent, and dimensions of the three release points are shown on Table 11.3-6.

11.3.3.2 Dilution Factors The dispersion and dilution of ga seous radioactive effluents rele ased from the plant depends on the meteorology of the site a nd its environs. To determ ine these parameters, onsite meteorological data has been obtained and analyzed as described in Section

2.3. Annual

atmospheric dilution factors have been calcula ted to determine resu ltant annual doses and concentrations of radionuclides from normal operation.

The building ventilation exhaust duc ts do not rise above the buildings; therefore, atmospheric releases for dose analys is purposes were consid ered ground level.

11.3.3.3 Estimated Releases Releases of radioactive material in gaseous effluents for initia l plant licensing were calculated using the GALE code presented in NUREG-0 016 to show compliance with 10 CFR 20, Appendix B, and 10 CFR 50, Appendix I, for no rmal operation plus an ticipated operational occurrences. The operational pa rameters including s ource terms were those presented in Appendix B. The values obtained are presented in Table 11.3-7. These values were used with the meteorology data from Section 2.3.5 to calculate maximum concen trations at the restricted area boundary and maximum individual dose offsit

e. These data pr ovide maximum annual average (/Q) values and does not incl ude a building wake factor.

Restricted area boundary concentrations used for this initial licensing anal ysis are tabulated by radionuclide and compared with 10 CFR 20 limits in Table 11.3-9. The estimated annual dose to persons offsite is presented in Section 5.2 of the Environmental Report and is well within the numerical guidelines of 10 CFR 50, Appendix I.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-069 11.3-14 Since becoming operational, releases of radioactive materials in gaseous effluents have been determined using actual flow rates and quantitative and qualita tive analyses. Doses due to radioactive materials in gaseous effluents are determined to show Technical Specifications compliance at specified intervals. Compliance is reported in the Annual Radioactive Effluent Release Report. Doses are as certained using the NRC GASPAR II computer code, current meteorology (or historical me teorology if current data are unavailable), and parameters outlined in the Offsite Dose Calculation Manual.

Tritium contamination of the auxiliary boiler and associated components ha s resulted in release paths which were not intended (see Section 9.4.16.2). The released radioactivity attributable to the worst case tritium concentration (2E + 06 Ci/liter) in the bo iler water is an insignificant portion of the total activity released in liquid and gaseous effluents and has been analyzed to result in a correspondingly insignificant radiation dose. The tritium concentration levels and makeup water volume ar e monitored and evaluated to ensure that tritium effluent releases from the plant ar e adequately quantified.

11.

3.4 REFERENCES

11.3-1 NUREG-0016 (BWR-GA LE Code), January 1979

11.3-2 Miller, C. W., Experi mental and Operational Confirmation of Off-Gas System Design Parameters , NEDO-10751, January 1973 (Proprietary).

11.3-3 Standards for Steam Surface Conde nsers, Sixth Edition, Heat Exchange Institute, New York, NY, 1970.

11.3-4 Underhill, Dwight, et al., "Design of Fission Gas Holdup Systems," Processing of the Eleventh AEC Air Cleani ng Conference, 1979, p. 217.

11.3-5 Miller, C. W., et al., A General Justification for Classification of Effluent Treatment System Equipment as Group D, NEDO-10734, February 1973.

11.3-6 Nesbitt, L. B., Design Ba sis for New Gas Systems, NEDE-11146, July 1971 (Proprietary).

11.3-7 NUREG-0800 USNRC Sta ndard Review Plan, Revisi on 2, July 1981, Branch Technical Position ETSB 11-5.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.3-15 Table 11.3-1 Design Air Ejector Offgas Re l e ase Rates (30 c f m inleakage) a T=0 T=30 Minutes Normal Disc harge from C h a r coal Adsorb e r s Additional Discharge from Charc o al Ads o rbers During Startup Isotope Half-life Ci/sec Ci/sec Ci/sec Ci/y r b Ci/sec Ci/startup 83m Kr 1.86 hr 3.4 x 10 3 2.9 x 10 3 - - - - 85m Kr 4.4 hr 6.1 x 10 3 5.6 x 10 3 4.3 1.2 x 10 2 1.1 x 10 1 1.4 85 K r c 10.74 yr 10-20 10-20 10-20 280-560 0 0 87 Kr 76 minutes 2.0 x 1 0 4 1.5 x 10 4 - - - - 88 Kr 2.79 hr 2.0 x 10 4 1.8 x 10 4 2.1 x 10-1 6.0 1.4 1.7 x 1 0-1 89 Kr 3.18 minutes 1.3 x 1 0 5 1.8 x 10 2 - - - - 90 Kr 32.3 sec 2.8 x 10 5 - - - - - 91 Kr 8.6 sec 3.3 x 10 5 - - - - - 92 Kr 1.84 sec 3.3 x 10 5 - - - - - 93 Kr 1.29 sec 9.9 x 10 4 - - - - - 94 Kr 1.0 sec 2.3 x 10 4 - - - - - 95 Kr 0.5 sec 2.1 x 10 3 - - - - - 97 Kr 1 sec 1.4 x 10 1 - - - - - 131m Xe 11.96 days 1.5 x 10 1 1.5 x 10 1 1.3 3.7 x 10 1 3.0 x 10-2 1.07 x 10-1 133m Xe 2.26 days 2.9 x 1 0 2 2.8 x 10 2 - - - - 133Xe 5.27 days 8.2 x 10 3 8.2 x 10 3 3.3 x 10 1 9.4 x 10 2 1.9 6.8 135m Xe 15.7 minutes 2.6 x 1 0 4 6.9 x 10 3 - - - - 135 Xe 9.16 hr 2.2 x 10 4 2.2 x 10 4 - - - - 137 Xe 3.82 minutes 1.5 x 1 0 5 6.7 x 10 2 - - - - 138 Xe 14.2 minutes 8.9 x 1 0 4 2.1 x 10 4 - - - - 139 Xe 40 sec 2.8 x 1 0 5 - - - - - 140 Xe 13.6 sec 3.0 x 10 5 - - - - - 141 Xe 1.72 sec 2.4 x 10 5 - - - - - 142 Xe 1.22 sec 7.3 x 10 4 - - - - - 143 Xe 0.96 sec 1.2 x 10 4 - - - - - 144 Xe 9 sec 5.6 x 1 0 2 - - - - - Totals 2.5 x 1 0 6 1.0 x 10 5 49-59 1383-1663 14.3 8.5 a Based on the 1971 mixture.

b This is based on curies present at time of release. No decay in environment is included.

c Estimated from experimental observations.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 3 9 11.3-16 Table 11.3-2 Offgas Sys t em Major Equipment Items Offgas Preheaters - two required Construction: Stainless steel tubes and carbon steel shell.

350 psig shell design pressure, 1000 psig tube design pressure. 400°F shell design temperature, 575°F tube design temperature.

Catalytic Recombiners - two required Construction: Stainless steel cartridge, carbon steel shell. Ca talyst cartridge containing a precious metal catalyst on metal base. Catalyst cartridge to be replaceable without removing vessel. 350 psig design pressu re, 900°F design temperature.

Offgas Condenser - one required Construction: Low alloy steel sh ell. Stainless stee l tubes. 350 psig shell design pressure.

250 psig tube design pressu re, 900°F shell design temperature, 150°F tube design temperature.

Water Separator - one required Construction: Carbon steel shel l, stainless steel wire mesh.

350 psig design pressure, 250°F design temperature.

Cooler-Condenser - two required Construction: Stainless steel shell. Stainless steel tubes. 100 psig tube design pressure, 350 psig shell design pressure. 32/150°F tube design temper ature, 32/150°F shell design temperature.

Moisture Separators (downstream of cooler-condenser) - two required Construction: Carbon steel sh ell, stainless steel wire me sh. 350 psig design pressure, 32/150°F design temperature.

Desiccant Dryer - four required Construction: Carbon steel shel l packed with Type 3A mol ecular sieve or equivalent.

350 psig design pressure, 32°F/

500°F design temperature.

Desiccant Regeneration Skid - two required

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-98-139 11.3-17 Table 11.3-2

Offgas System Major Equipment Items (Continued)

Dryer Chiller - two required

Construction: Carbon steel shel l, stainless steel tubes, desi gn temperature 32°F/500°F, design pressure 50 psig.

Regenerator Blower - two required

Construction: Electrical, design pressure 50 psig design temperature 32°F/150°F.

Dryer Heater - two required

Construction: Electrical, design temperature 32°F/500°F, design pressure 50 psig.

Glycol Cooler Skid - one required

Glycol Storage Tank - one required

Construction: Carbon steel 3000 gal. Water-filled hydrostatic design pressure.

32°F design temperat ure. API-650.

Glycol Solution Refrigerators and Motor Drives - three required

Construction: Conventional refrigeration units. Glycol solution exit temperature 35°F.

Glycol Pumps and Motor Drives - three required

Construction: Cast iron, 0°F design temperature.

Prefilters and After Filters - two required of each type

Construction: Carbon steel shell.

High-efficiency, moisture-resis tant filter element. Flanged shell. 350 psig design pressure. -50/150°F desi gn temperature.

Charcoal Adsorbers - eight beds

Construction: Carbon steel. Approximately 4-ft. O.D. x 21-ft vesse ls each containing approximately 3 tons of activ ated carbon. Design pressure 350 psig, design temperature

-50/250°F.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-041,09-009 11.3-18 Table 11.3-3

Process Data For The O ffgas (RECHAR) System

The Information On This Page Is Proprietary And Was Submitted Under Separate Cover

(CVI DWG 02N64-04,11,1, Rev. 4 or 02N64-04,11, 2, Rev. 2)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.3-19 Table 11.3-4 Offgas Sys t em Alarme d Process Pa r a me t e rs Control Room Parameters Indica t ed Recorded Air ejector discharge pressure - high X Prehea t er discharge t e mperature - l o w X Recombiner ca t alyst tem p erature - high/low X Offgas condenser wa t e r l e vel (dual) - high/low X Offgas condenser gas discharge t e mperature - h i gh (local)

X H 2 analysis (offgas condenser discharge) - dual - high X Offgas condenser disc h a rge radiation - high X Gas flow -

h i gh/low X Cooler - condenser dischar g e temperature - high/low X Glycol solution temperature - high/low X Glycol solution level - low Gas drier discharge humidity - high (local)

X Prefilter dP - high X Charcoal adsorber te mperature - hi g h X Charcoal train flow - high/low X After filter dP - high X Offgas (cha r c oal bed discharge) r a diation - high X Steam flow - low X Desiccant dryer outlet t e mperature - high/low X Dryer chiller outlet te m p erature - high (local)

X Dryer heater temperature - high Dryer heater outlet te m p erature - h i gh (local)

X Loop seals water level - low C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.3-20 Table 11.3-5 Equipment Malfunction Analysis Equipment Item Malfunction Consequences Design Precautions Steam jet air ejectors Low flow of motive high-pressure steam When the hydrogen and oxygen concentrations exceed 4 and 5 vol %,

respectively, the process gas becomes

flammable.

Alarm provided on steam for low steam flow. Recombiner temperature alarm. Inadequate steam flow will cause overheating and deterioration of the catalyst.

Steam flow to be held at constant maximum flow regardless of plant level during

operation. Wear of steam supply nozzle of ejector Increased steam flow to recombiner. This could reduce degree of recombination at low power levels. Low temperature alarms on preheater exit (recombiner inlet). Recombiner outlet H 2 analyzers. Preheaters Steam leak Would further dilute process offgas. Steam consumption would increase. Spare preheater. Low pressure steam supply Recombiner performance would fall off at low power level and hydrogen content of recombiner gas discharge may increase, eventually to a combustible mixture. Low-temperature alarms on preheater exit (recombiner inlet). Recombiner outlet H 2 analyzers. Recombiners Catalyst gradually deactivates Temperature profile changes through catalyst. Eventually excess H 2 would be detected by H 2 analyzer or by gas flowmeter. Eventually the stripped gas could become combustible. Temperature probes in recombiner and H 2 analyzer provided. Spare recombiner. Catalyst gets wet at start H 2 conversion falls off and H 2 is detected by downstream analyzers. Eventually the gas could become combustible. Condensate drains, temperature probes in recombiner. Air bleed system at startup.

Recombiner thermal blanket, spare recombiner, and heater. Hydrogen analyzer.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.3-21 Table 11.3-5 Equipment Malfunction Analysis (Continued)

Equipment Item Malfunction Consequences Design Precautions Offgas condenser Cooling water leak The coolant (reactor condensate) would leak to the process gas (shell) side. This would be detected if drain well liquid level increases. Moderate leakage would be of no concern from a process standpoint. (The process condensate drains to the hotwell).

None Liquid level instruments fail If both drain valves fail to open, water will build up in the condenser and pressure drop will increase. Two independent drain systems, each provided with high and low-level alarms.

The high P, if not detected by instrumentation, could cause pressure

buildup in the main condenser and eventually initiate a reactor scram. If a drain valve fails to close, gas will recycle to

the main condenser, increase the load on the SJAE, and increase operating pressure of the

main condenser.

Water separator Corrosion of wire mesh element Higher quantity of water collected in holdup line and routed to radwaste. Stainless steel mesh specified. Holdup line Corrosion of line Leakage to soil of gaseous and liquid fission products.

Outside of pipe dipped and wrapped. 0.25-in. corrosion allowance. Cooler-condenser Corrosion of tubes Glycol-water solution would leak into process (shell) side and be discharged to clean radwaste. If not detected at radwaste, the glycol solution would discharge to the reactor condensate system. Stainless-steel tubes specified. Low level

alarm glycol tank level. Spare cooler condenser provided.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.3-22 Table 11.3-5 Equipment Malfunction Analysis (Continued)

Equipment Item Malfunction Consequences Design Precautions Cooler-condenser (continued) Icing up of the tubes Shell side of cooler could plug up with ice, gradually building up pressure drop. If this happens, the spare unit could be activated.

Complete blockage of both units would increase P and lead to a reactor scram.

Design glycol-H 2O solution temperature well above freezing point. Spare unit provided. Temperature indication and low alarms on glycol temperature and process

gas temperature. Glycol refrigeration machines Mechanical failure If both spare units fail to operate, the glycol solution temperature will rise and the

dehumidification system performance will deteriorate. This will require rapid regeneration cycles for the desiccant beds and may raise the gas dewpoint as it is discharged from the drier. Two spare refrigerators during normal operation are provided. Glycol solution temperature alarms provided. Gas moisture detectors provided downstream of gas

driers. Moisture separators Corrosion of wire mesh Increased moisture would be retained in process gas routed to gas driers element.

Over a long period, the desiccant drier cycle period would deteriorate as a result of moisture pickup. Pressure drop across prefilter may increase if filter media is

wetted. Stainless steel mesh specified. Spare unit provided. High P alarm on prefilter. Prefilters Loss of integrity of filter media More radioactivity would deposit on the drier desiccant. This would increase the radiation level in the drier vault and make maintenance more difficult, but would not affect releases to the environment. Spare unit provided in separate vault. P instrumentation provided.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORTDecember 2005LDCN-03-009 11.3-23 Table 11.3-5 Equipment Malfunction An alysis (Continued)

Equipment Item Malfunction Consequences Design Precautions Desiccant drier Moisture breakthrough Increased moisture in air entering charcoal adsorbers would decrease adsorption effectiveness, thus reducing radioisotope retention time. Drier cycles on timer. Redundant gas humidity analyzers and alarms supplied.

Redundant gas drier system supplied. Gas drier and first charcoal bed can be bypassed through alternate drier to second charcoal

bed. Desiccant regeneration equipment Mechanical failure Inability to regenerate desiccant. Redundant, shielded desiccant beds and drier equipment is supplied. Charcoal adsorbers Charcoal accumulates moisture Charcoal performance will deteriorate gradually as moisture deposits. Holdup times for krypton and xenon would decrease, and plant emissions would increase. Provisions made for drying charcoal as required during

annual outage. Highly instrumented, mechanically simple gas dehumidification system with redundant equipment. After filters Loss of integrity of filter mediaProbably of no real consequence. The charcoal media itself should be a good filter at the low air velocity. P instrumentation provided. Spare unit provided. System Internal detonation Release of radioactivity if pressure boundary fails. Main process equipment and piping are designed to contain a detonation. Earthquake damage Release of radioactivity. Dose consequences are within 10 CFR 20 limits. Analysis is included in Reference 11.3-5.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 3 9 11.3-24 Table 11.3-6 Release Point Data Reactor Building Radwaste Building Turbine Building Height of release point above grade 230 ft 6 in.

67 ft 119 ft Annual average rate of air flow from release point (cfm) 80,000 83,000 360,000 Annual average heat flow from release point (Btu/hr) 15.09x10 6 41.46x10 6 13.02x10 6 Type and size of release

point (in.)

Duct 45 x 120 3 louver houses

54 x 96 x 30 each 4 exhaust fans

57 x 79 each

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.3-25 Table 11.3-7 Gaseous Waste Sys t em Release Gaseo u s Relea s e Rat e a (Ci/yr) Nuclide Coolant Conc.

( Ci/g) Containment Building Turbine Building React o r Building Radwas te Building Gland Seal Air Eject o r Mechani c al Vac Pump Total 83mKr 1.200E-03 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 85m Kr 2.000E-03 3.1E 00 7.1E 01 3.1E 00 0.0 0.0 2.1E 00 0.0 8.0E 01 85Kr 6.300E-06 0.0 0.0 0.0 0.0 0.0 2.8E 02 0.0 2.8E 02 87 Kr 6.900E-03 3.1E 00 2.0E 02 3.1E 00 0.0 0.0 0.0 0.0 2.1E 02 88 Kr 6.900E-03 3.1E 00 2.4E 02 3.1E 00 0.0 0.0 0.0 0.0 2.4E 02 89Kr 4.300E-02 0.0 0.0 0.0 0.0 0.0 0.0 0.0

0.0 131mXe

4.900E-06 0.0 0.0 0.0 0.0 0.0 5.2E 00 0.0 5.2E 00 133mXe 9.400E-05 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 133 Xe 2.700E-03 6.9E 01 2.9E 02 6.9E 01 1.0E 01 0.0 2.3E 01 2.3E 03 2.8E 03 135m Xe 8.800E-04 4.8E 01 6.8E 02 4.8E 01 0.0 0.0 0.0 0.0 7.8E 02 135 Xe 7.600E-03 3.6E 01 6.6E 02 3.6E 01 4.5E 01 0.0 0.0 3.5E 02 1.1E 03 137Xe 4.900E-02 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 138 Xe 2.900E-02 7.3E 00 1.5E 03 7.3E 00 0.0 0.0 0.0 0.0 1.5E 03 Total noble gases 7.0E 03 131I 3.618E-03 1.8E-02 2.0E-01 1.8E-01 5.0E-02 0.0 0.0 3.0E-02 4.8E-01 133I 1.549E-02 7.1E-02 8.0E-01 7.1E

-01 1.8E-01 0.0 0.0 0.0 1.8E-00 Tritium gaseous release 7.1E 01 NOTE: 0.0 appearing in the table indicates release is less than 1.0 Ci/yr for noble gas, 0.0001 Ci/yr for iodine.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.3-26 Table 11.3-7 Gaseous Waste Sys t em Release (Continued)

Airborne Part iculate Release Rate (Ci/yr)

Nuclide Containment Building Turbine Building React o r Build i ng Radwas te Bui l ding Mechani c al V a c Building Total 51 Cr 3.1E-06 1.4E-02 3.1E-04 9.4E-05 0.0 1.4E-02 54 Mn 3.1E-05 5.2E-04 3.1E-03 4.7E-04 0.0 4.3E-03 59Fe 4.2E-06 5.2E-04 4.2E-04 1.6E-04 0.0 1.1E-03 58Co 6.3E-06 6.3E-04 6.3E-04 4.7E-05 0.0 1.4E-03 60Co 1.0E-04 2.1E-03 1.0E-02 9.4E-04 0.0 1.5E-02 65Zn 2.1E-05 2.1E-04 2.1E-03 1.6E-05 0.0 2.3E-03 89Sr 9.4E-07 6.3E-03 9.4E-05 4.7E-06 0.0 6.4E-03 90Sr 5.2E-08 2.1E-05 5.2E-06 3.1E-06 0.0 2.9E-05 95Zr 4.2E-06 1.0E-04 4.2E-04 5.2E-07 0.0 5.2E-04 124Sb 2.1E-06 3.1E-04 2.1E-04 5.2E-07 0.0 5.2E-04 134Cs 4.2E-05 3.1E-04 4.2E-03 4.7E-05 3.1E-06 4.6E-03 136Cs 3.1E-06 5.2E-05 3.1E-04 4.7E-06 2.1E-06 3.8E-04 137Cs 5.8E-05 6.3E-04 5.8E-03 9.4E-05 1.0E-05 6.6E-03 140Ba 4.2E-06 1.1E-02 4.2E-04 1.0E-06 1.1E-05 1.1E-02 141 Ce 1.0E-06 6.3E-04 1.0E-04 6.3E-05 0.0 8.0E-04 a Estimated release based on GALE code evaluation.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 3 9 11.3-27 Table 11.3-8 Building Volume and Ventilation Rates Building Free Air Volume (f t 3) Ventilation Rate (cf m) Secondary containment (reactor building) 3.5 x 10 6 80,000 Radwaste building 2.0 x 1 0 6 83,000 Turbine bui l d ing 5.7 x 1 0 6 360,000 Primary containment (drywell) 2.0 x 1 0 5 10,50 0 a Primary containment (wetwell) 1.4 x 1 0 5 7,50 0 a a During primary containment pu r ge only.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-98-139 11.3-28 Table 11.3-9 Maximum S ector Annual Average Concentrations of Gaseous Radioactive M a ter i als at t h e Original Restricted Area Boundary Nuclide Annual Average Release Rate (Ci/sec) Boundary Concentration a ( Ci/cm 3) Derived Air Concentration (DAC)b (Ci/cm 3) Concentration/

DAC 3 H 2.3 E-6 1.8 E-11 1 E-7 0.000 1 8 85m Kr 2.5 E-6 1.9 E-11 1 E-7 0.0002 85 Kr 9.0 E-6 6.9 E-11 7 E-7 0.000 0 99 87 Kr 6.6 E-6 5.1 E-11 2 E-8 0.002 88 Kr 8.0 E-6 6.2 E-11 9 E-9 0.000 6 9 131m Xe 1.7 E-7 1.3 E-12 2 E-6 0.000 0 006 133 Xe 9.0 E-5 6.9 E-10 6 E-7 0.0011 135m Xe 2.4 E-5 1.9 E-10 4 E-8 0.0047 135 Xe 3.7 E-5 2.8 E-10 7 E-8 0.0041 138 Xe 4.6 E-5 3.6 E-10 2 E-8 0.018 131 I 1.6 E-8 1.3 E-13 2 E-10 0.0006 133 I 5.7 E-8 4.4 E-13 1 E-9 0.000 4 4 51 Cr 4.3 E-10 3.4 E-15 3 E-8 0.000 0 00 1 1 54 Mn 1.3 E-10 9.7 E-16 1 E-9 0.000 0 00 9 7 59 Fe 3.7 E-11 2.8 E-16 5 E-10 0.000 0 00 5 7 58 Co 4.3 E-11 3.4 E-16 1 E-9 0.000 0 00 3 4 60 Co 4.3 E-10 3.4 E-15 5 E-11 0.000 0 67 65 Zn 7.3 E-11 5.7 E-16 4 E-10 0.000 0 015 89 Sr 2.0 E-10 1.6 E-15 1 E-9 0.000 0 016 90 Sr 9.3 E-13 7.2 E-18 6 E-12 0.000 0 013 95 Zr 1.7 E-11 1.3 E-16 4 E-10 0.000 0 003 124 Sb 1.7 E-11 1.3 E-16 3 E-10 0.000 0 004 134 Ca 1.5 E-10 1.1 E-15 2 E-10 0.000 0 058 136 Cs 1.1 E-11 8.9 E-17 9 E-10 0.000 0 00 0 99 137 Cs 2.1 E-10 1.6 E-15 2 E-10 0.000 0 079 140 Ba 3.7 E-10 2.8 E-15 2 E-9 0.000 0 015 141 Ce 2.5 E-11 1.9 E-16 1 E-9 0.000 0 00 1 9 MPC 0.03 a /Q factor of 7.7 x 10

-6 at 0.5 miles distance in SE sector used. No building wake factor included; meteorological data from 4/74 through 3/75.

b 10 CFR 20, Appendix B, to 20.1001-20.2401, Table II Column I.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 3 9 11.3-29 Table 11.3-10 Frequency and Quantity of Steam Discharged to Suppression Pool Event Frequency Category Quantity of Steam (lb/even t) 1. RCIC test (mon t hly) Moderate 29,000 2. RCIC test (vessel inject i on at startup)

Moderate 116,000

3. Inadvertent RCIC in jection Moderate 5,000 4. SRV test Moderate 118,000 5. Inadvertent SRV o p ening Moderate 118,000 6. Trip of both recircu l ation pumps Moderate 260,00 0 a 7. Turbine trip Moderate 18,000 8. Generator load r e jection Moderate 18,000 9. Pressure regulator f a ilure - open Moderate 256,00 0 a 10. Recirculation flow control failure - dec r easing Moderate 260,00 0 a 11. Loss of all feedwater flow Moderate 267,00 0 a 12. Inadve rtent closure -

all MSIV Moderate 271,00 0 a 13. Loss of condenser vacuum Moderate 291,00 0 a 14. Feedwater control failure -

maximum demand Moderate 270,00 0 a 15. Loss of auxiliary t r ansformer Moderate 251,00 0 a 16. Loss of all grid c onnections Moderate 280,00 0 a 17. Turbine trip w/o bypass Moderate 126,00 0 a 18. Generator load rejection w/o bypass Moderate 127,00 0 a 19. Stuck open SRV Moderate 817,000 Notes: Events 1, 2, and 3 based on steam flow quantity during test m ode per RCIC system process diagram.

Events 4 and 5 assuming test and i n adverte n t opening at 1000 psi reactor pressure for 10 minutes.

Events 6 through 18 based on event description from Chapter 15. Event 19 based on results from 251 St a ndard Plant Suppression Pool Response Analyse s. a Isolation event. It i s assumed that SRV cyc ling is termina t ed at T =

3 0 minutes and reactor depressurization.

Draw. No.

Rev.Figure Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report11.3-2.1 77 M535-1 Offgas System P&IDRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report11.3-2.2 67 M535-2 Offgas System P&IDRev.FigureDraw. No.Amendment 61December 2011 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 2-0 0 0 11.4-1 11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4.1 DESIGN BASIS

Power plant operation results in various solid radioactive wastes that require disposal. These wastes can be in the form of wet solids, such as powdered ion exchange resins from filter demineralizers, e xpended bead resins from deep bed demineralizers, small quantities of miscellaneous liquid, and miscel laneous dry materials such as paper, rags, plastic, and laboratory wastes.

The objective of the solid waste management system is to collect, m onitor, process, and package these waste products in a suitable form for o ffsite shipment and bur ial. In designing the system to meet the stated objective, the following criteria were applied:

The system has the capacity to handle the volumes of waste from normal operations and anticipated operational occurren ces. The expected annual volum es of wet solid wastes are shown in Table 11.4-1.

The system is designed to pro cess the quantity of waste and concentration of radionuclides listed in Table 11.4-3 while maintaining occup a tional e xposure as low as is reasonably achievable (ALARA). This is done by controlli ng the pipe run locatio ns for shielding and exposure considerations, placing th e process equipment in shielded areas, by providing remote operating stations, and by performing dewatering operations in radiation shields if required.

The radwaste building shielding is designed for the highest radi oactivity source, reactor water cleanup (RWCU) resins. The equi pment layout is shown in gene ral arrangement drawings in Section 1.2. Table 11.4

-2 lists capacities, design press u re, and design temperature of the major equipment.

In keeping with the ALARA philosophy and Appendix I to 10 CFR 50, the solid waste management system's contributi on to offsite doses is minimize d by filtration a nd by directing the ventilation air flow from areas of low airborne contaminati on to areas of higher airborne contamination. The filtered ventilation radi oactive releases are discussed in Section 11.3.

The solid waste management syst em operations and procedures are designed to limit the dose to offsite persons from station operations to significantly less than the limits specified in 10 CFR Part 20. Water separated in processing is returned to the liquid waste management system for treatment as described in Section 11.2.2.1 and shown in Figure 11.2-1.

The system can accommodate a va riety of shipping container size s and shapes with and without shields. Provisions are made for the detection and removal of loose surface contamination on the waste containers. The radi ation levels of the waste containers are monitored so that C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-026 11.4-2 provisions can be made to ensu re that shipping regulation radiation levels are not exceeded.

Compliance with applicable regulations, e.

g., 10 CFR Parts 61 and 71 and 49 CFR is discussed in Sections 11.4.2.9 and 11.4.3.

The safety class, quality group classification, quality class, and seismi c category of radwaste systems are specified in Table 3.2-1. The solid waste management system is not designed to Seismic Category I. It is located in the Seismi c Category I portion of th e radwaste building. The seismic classification of the radwas te building is discussed in Section 3.8.4.1.2. See Section 3.1.2.6.4 for a discussion of systems provided to meet General Design Criterion 63.

11.4.2 SYSTEM DESCRIPTION

A portable solid waste management system is used by Columbia Generating Station (CGS), as described in the following subsections.

It is required by 10 CFR 61 that, if certain activity criteria are exceeded, wet solid wastes be stabilized by solid ification or processed to remove free standing liquids with containment in high integrity containers (HICs). The HICs may provide stability alone or in conjunction with an engineered barrier at the disposal site. The presently installed plan t system is designed to interface with the portable dewatering/drying sy stem which, when coupled with HICs, meet the requirements for stabilization in compliance with 10 CFR 61.

There are two resin dewatering systems used at CGS: the Resin Drying System (RDS) and the Self Engaging Dewatering System (SEDS). The resin dewatering systems are vendor provided with NRC approved topical reports (TP P-A, Revision 1 for the RDS, and CNSI-DW-1118-01-P-A for the SEDS), and the systems are operated according to the technical requirements of this report. The types and quantities of waste to be processed are described below. The NRC licensed shipping casks and associated liners are used for

processing, transporting, and disposing of wet solid wastes when required.

System operation is closely monitored by Ener gy Northwest personnel.

If a vendor operates the dewatering system, the vendor will be requi red to submit their ope rating procedures for Energy Northwest review and approval. The ve ndor procedures will either be incorporated into a CGS procedure or will be a pproved "as is" by CGS prior to use.

Dewatering of resins to meet applicable dewatering criteria is conducted in accordance with approved procedures inside the ra dwaste building in the liner st orage area or in shipping casks on trucks, where any spills are routed to existi ng floor drain sumps and the building ventilation filtration system ensures no unf iltered airborne releases occu r from dewatering activities.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 9-0 0 0 11.4-3 11.4.2.1 General The sources of the various radioactive wet re sin waste inputs to the system are shown in Figure 11.2-1. Table 11.2-11 shows the design basis expect ed frequency of input, the quantities of solids generated, th e radioactivity level of the solids after accumulation, and the volume of liquid used in sluici ng accumulated solids to the processing equipment. The excess liquid is subsequently returned to the liquid waste management system.

These values are based on experience from operational BWR nuclear power stations.

Figure 11.4-1 shows the solid waste management sy stem up to and including the portable portion of the system describe d in the vendor's topical repor ts. The phase separation and concentration portions of the sys t em are shown on Figures 5.4-22 , 10.4-5 , 11.2-2 , and 11.2-3. Tanks containing radioactive waste are provided with overflow connections which direct any overflow to drain sumps.

The solid waste processing areas are located in the radwaste building, where wet and dry solid wastes may be processed. We t solid wastes include backwash resin from the RWCU system, the condensate filter demineralizer system, the fuel pool filter demineralizer s, the floor drain and equipment drain filter demineralizers, and spent resin from the floor drain demineralizer and the waste demineralizer. Dry solid wastes include items such as rags, paper, plastics, small equipment parts, and laboratory wastes.

11.4.2.2 Radwaste Disposal System For Reactor Water Cleanup Resin The backwash discharge from the cleanup filter deminerali zers is collected a nd concentrated in two 4500-gal cleanup phas e separators which are located be low the cleanup demineralizers in the radwaste building. After several backwashes are accumulated, the waste is transferred to the portable dewatering system.

The cleanup phase separators are designed to concentrate the resin from 0.5% by weight solids to approximately 5% by weight solids by sedime ntation and decantation of the slurry. While the working separator is filling, th e other previously filled tank is held isolated to the extent practicable to allow for additional decay of resin activity.

After each backwash batch is re ceived by the working separator, the batch is allowed to settle for a period of time and the decantate is then transferred by pumping to the waste collector tank. When sufficient resin has accumulated, the working separato r is isolated and allowed to stand for a period to permit radioa ctive decay. At the end of this decay period the sludge is fluidized to approximately 5% weight solids and transferred by pumping to the portable dewatering system.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.4-4 11.4.2.3 Radwaste Disposal System For Condensate De mineralizer Resin The backwash discharge from the condensate filter demineralizers is collected in the condensate backwash receivi ng tank which is located be low the condensate filter demineralizers in the radwaste building. After collection, the waste is transferred by pumping to one of the two condensate pha se separators for processing.

Operation of the condensate phase separators is similar to that for the clean up phase separators. Backwash resin is received at 0.5% by weight solids and concen trated to approximately 5% by weight solids, allowed to stand for a period of radioactive decay and then decanted and transferred by pumping to th e portable dewatering system.

11.4.2.4 Radwaste Disposal System For Fuel Pool, Floor Drain, and Waste Collector Filter Resin Backwash resin wastes from the fuel pool filter demine ralizers, floor drain, and waste collector filter demineralizers are backwa shed to the waste sludge phase separator tank. The waste sludge phase separator is designed to concentrate the resin from 0.5% by weight solids to approximately 5% by weight solids by sedimentation and decantation.

After each backwash batch is received by the se parator, it is allowed to settle for a period of time and the decantate is then transferred by pumping to the floor drain collector tank.

When an appropriate quantity of resin is accumulated, the resin is flui dized to approximately 5% by weight solids and transferred by pum ping to the portable dewatering system.

11.4.2.5 Radwaste Disposal System For Spent Resin Spent bead resins from th e floor and equipment drain polishing demineralizers are hydropneumatically transferred to the spent resin tank. The ta nk is designed to retain one batch of resins plus resin transfer water plus freeboard.

The decay time of any single batch is governed by the need to make the spent resin tank available to receive a subseque nt batch from an alternate demineralizer. The frequency of spent resin discharge from the floor and equipment drain polishing demineralizers is estimated to be about once every 2 months.

Each batch of the spent resin is transferred at approximately 40% by weight solids to th e portable dewatering system.

11.4.2.6 Resin Container Handling and Storage Filled containers from the dewa tering operation in th e container storage area are lifted by a crane and placed on a track-riding dolly. The dolly moves the containers from the storage area to the loading area where another crane lifts the container onto th e truck for offsite shipment.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 2-0 0 0 11.4-5 Alternatively, the dewatering ope ration for high activity resin wast e can be performed in a cask on or off the truck bed. General lo cations and arrangeme nts are shown in Figure 1.2-13.

11.4.2.7 Miscellaneous Dry Solid Waste System Dry active waste may consist of air filtration medi a, miscellaneous paper, plastic and rags from contaminated areas, contaminat ed clothing, tools and equipment parts which cannot be effectively decontaminated, solid laboratory wastes, and ot her similar materials. The radioactivity of much of this waste is low enough to permit handling by contact. Compressible wastes may be compacted into me tal containers to reduce their vol umes. Alternately, container vans (C-vans) or other containe rs suitable for shipment may be used for dry radioactive waste shipped to a vendor for vol ume reduction services.

A relatively small quantity of high activity compactable and noncom pactable dry active radioactive waste (DAW) may be loaded into the open top or encapsulation liners. Other containers whose geomet ry is compatible with shipment in shielded shipping casks may be used.

A locally controlled compactor system, if used, will include the following features: hydraulic pump with motor, hydraulic oil storage, high-efficiency filters, fan, and accessories. Ventilation air will be pulled across the top of the containers and then through high-efficiency filters by a fan during the compression process a nd exhausted to the radw aste building exhaust system described in Section

9.4. Solid

wastes and other nonliquid radioactive material and C-vans may be stored temporarily near the truck loading area outside the radwaste building, on the outdoor curbed pad adjacent to the radwaste building, or other suitable location. Noncompressible solid wastes are packaged in containers suitable for the waste. Low activity waste can be stored until enough is accumulated to permit economical transportation to an offsite disposal si te or to a vendor for volume reduction of the wa ste prior to disposal.

Irradiated reactor components consisting of spent control rod blades, fuel channels, in-core ion chambers, and other equipment are stored in the spent fuel storage pool to allow for radioactive decay pr ior to shipment.

The packaging of dry solid wastes will be administratively cont rolled to ensure the 10 CFR 61 and/or burial facility free standing liquid criteria are met.

11.4.2.8 Expected Volumes The design basis expected freque ncy of solids input, the quantities of solids generated, and the radioactivity of the solids a f ter accumulation are shown in T a ble 11.2-1

1. Table 11.4-3 shows the expected solids production rate and significant nuclides associated with each batch of

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.4-6 dewatered waste. The distribution of these nuclides (depe ndent on the concentration in the reactor water) corresponds to the design basis offgas noble gas release rate as described in Section 11.3.1. Table 11.4-1 , excluding dry and compacted waste, shows the expected annual container production of solid wastes based on the process diagram data inputs. No decay after container filling operations has been considered.

11.4.2.9 Packaging Radioactive wastes are packag ed and shipped from CGS in containers that meet the requirements established in 49 CFR 171-180 for the Departme nt of Transportation and 10 CFR 71 for the NRC.

Packaging of wastes being dewatered is typically performed remotely behind shielding as described below. Empty contai ners may be brought into the processing area using the dolly and filled with waste for processing. The quan tity of wastes packaged in the container is controlled by opera ting procedures.

Dewatered liners are capped, surveyed for surface contamina tion, and decontaminated as necessary prior to shipment.

11.4.2.10 Storage Facilities The general arrangement drawings in Section 1.2 show the layout of the radwaste handling areas in the radwaste building.

For the liners presently in use (of up to 210 ft 3 capacity), the storage area can accommodate about 15 filled liners. High activity containers can be stored to allow for 6 months decay prior to shipment if necessary. It is expected, however, that most radwaste cont ainers will be shipped within 1 to 3 months of generation.

11.4.2.11 Shipment The following describes a typical loading sequence for liners:

A truck containing the cask is moved into the truck loading area. The dolly is moved to the storage area loading station and a capped container of waste is plac ed on the dolly by the storage area crane. The dolly is moved back to the truck loading area where the liner is lifted into the cask, a nd the cask lid is placed on the cask. The cask is decontaminated if necessary for shipment. Similar operations are performed when loading unshiel ded containers onto the truck. Dewatering of high activity wastes can also be performed in liners "in cask" in place on the shipping truck.

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 0-0 2 6 11.4-7 Any unshielded containers found to have external contamina tion exceeding 49 CFR limits are decontaminated. Further smear tests and cleani ng are carried out as re quired until the activity on the container is within acceptable limits.

Radwaste tank failure and spent fuel cask drop incidents are discussed in Sections 15.7.3 and 15.7.5, respectively.

11.4.2.12 Process Monitoring Process monitoring is performed by the dewate ring system operator a nd the operator in the radwaste control room. They are in communica tion during waste transfer to the dewatering system. The dewatering system processing is monitored by remote closed-circuit TV cameras and other instrumentation as described in the topical reports for the process.

Each RWCU phase separator is eq uipped with one level indicating de vice for total liquid level.

The total liquid level indicator utilizes an air bubbler and a pre ssure sensing level transmitter which drives a 0-100% level gauge and a high-level alarm in the radwaste control room. The level transmitter also drives a level indicator on the local control pane l and provides control functions for the decant pump, the resin discharge pump, and the phase separator inlet selector valve.

The waste sludge phase separator ha s total liquid level indication.

It uses an air bubbler and a pressure-sensing level transmitter. In addition to the level gauge and high-level alarm in the radwaste control room, the le vel transmitter provides control inputs to the decant pump, the stop and flush circuit on the sludge discharge pump, and the disc harge valves from the waste collector and floor drain collector tanks to the waste sludge phase separator.

The condensate phase sepa rators level instrumentation is th e same as that described for the RWCU phase separators. Level indication for the spent resin ta nk is essentially the same as that described for the RWCU phase separators. The concentrated waste measuring and waste mixing tanks are not in service.

11.4.3 PROCESS CONTROL PROGRAM

11.4.3.1 Objective The objectives of the process control program are to characterize and classify radwaste and ensure the complete solidifica tion of all wet wastes being so lidified and to ensure that dewatered wet wastes and disposal packages meet the free standing liquid and stability requirements of 10 CFR 61. To meet these objectives, the process control program has incorporated the recommendations set forth in NUREG-0800 and Branch Technical Position -

ETSB 11-3.

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 0-0 2 6 11.4-8 11.4.3.2 Process Control Program To ensure that acceptable wast e forms are produced for dis posal in compliance with the requirements in 10 CFR 61, the process control program provides for characterization of individual waste streams, classi fication of final waste products, proper disposal packaging, and verification that waste dewatering and/or solidification has been succes sful. In addition, nonstable waste form processing, waste storage, handling, and transportation activities take place under the process control prog ram to ensure compliance with all applicable regulations in 10 CFR Parts 20, 61, and 71.

Stability of waste form at Co lumbia Generating Station is normally achieved through the combined properties of dewatere d media and HICs or HICs pl us engineered barriers. If solidification for stability is performed at Columbia Generati ng Station, it will be done in accordance with approved procedur es to meet applicable dispos al site license conditions and other applicable requirements.

Other processing methods such as portable deionization, soli dification for nonstable waste form, evaporation, filtration, etc., may be used to process nonhazardous radioactively contaminated liquids. If these activities will result in the disposal of waste at a licensed burial site, they will be conducted in accordance with approved procedures to ensure that applicable disposal site license conditions and other requirements ar e met. Solidification agents listed in the burial facility license may be used to ensure free-standing liquid criteria are achieved.

The process control program contains the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wast es will be accomplished in such a way as to ensure compliance with 10 CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements govern ing the disposal of solid radioactive wastes.

Changes to the process control program are documented, and records of reviews performed are

retained for the duration of the operating license. This doc umentation contains sufficient information to support the change together with the appropriate anal yses or evaluations justifying the change and a determination that the change will maintain the overall conformance of the solidified waste product to the existing requirements of federa l, state, and other applicable regulations. Changes become effective after re view and acceptance in accordance with the Operational Quality Assu rance Program Description (OQAPD).

The radioactive waste proce ss control program incorporates the following elements:

a. Waste stream descriptions,

b. Process controls,

c. Characterization, C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-026 11.4-9 d. Processing and stabilization,

e. Sampling,

f. Scaling factors,

g. Classification,

h. Computer code usage,

i. Analytical methods,

j. Packaging,

k. ALARA,

l. Shipping, m. Documentation, n. Equipment maintena nce and calibration, o. Minimization a nd segregation, p. Storage,

q. Trending, and

r. Reporting.

Plant procedures implement th e process control program.

11.4.3.3 Process Control Systems

The transfer of wet solid wastes to the processing system is monitored and controlled from the process control panel in the radwaste control room. The resin transfer system interfaces with the portable dewatering system at the point of connection w ith the dewatering equipment.

At the interface point, the applicable dewatering system is operated per the technical requirements of the NRC-approve d topical report and is designed to ensure that processed waste in conjunction with the de watering HIC or liner is prepar ed for burial and meets the 10 CFR Part 61 criteria. The process control program is describe d and controlled by procedure. The dewatering system s control panels are in a remo te location shielded from the waste processing resin containi ng components. It is designed for automatic operation and provides indication and alarm fo r liner level, temperature, pressure, and other operating parameters.

If a waste processing vendor is used, they are m onitored by Energy Northw est to ensure that applicable procedures, Energy Northwest or vendor, are being followed and that an acceptable end product is formed as described in the procedure used.

11.4.3.4 Waste Characterization

The wet wastes at CGS to be processed are charact erized in individual streams for RWCU resins, equipment drain radioactive (EDR) and floor drain radioactive (FDR) powdered resins, EDR and FDR bead resins, and condensate resins. The four major systems producing wastes are described below.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-026 11.4-10 a. Floor drain system - Wastes from the turbine building, reactor building, and radwaste building floor drain sumps are routinely monitored and collected for processing in the floor drain collector tank. The floor drain filter and demineralizer sludges are combined with equipment drain filte rs and sludges to form a mixture which is sampled prior to processing. Similarly, the EDR/FDR bead (polishing) resins are also sampled prior to processing.

b. Equipment drain system - Wastes from the turbine building, reactor building, and radwaste building equipment drain sumps are routinely monitored and collected in the wa ste collector tank. Sludges a nd resins from the high purity filter demineralizers are sampled prior to processing as described above in the FDR system description.

c. Condensate filter demineralizer sy stem - The condensate polishing filter demineralizers use pressure precoated ion exchange media filters. Resins are sampled prior to processing.

d. Reactor water cleanup filter demineralizer syst em - The RWCU filter demineralizers use pressure precoated ion exchange media filters. Resins are sampled prior to processing.

Accumulations in collection tanks and phase separators will be tracked to aid in determining processing schedules and potential problems in system operation by tracking volumes. Selected coolant isotopic concentrations are trended to provide early indication of changes important to waste classification.

The waste streams in the foregoing systems are characterized by a periodic sampling and analysis program that establishe s plant-specific isotopic correlati on factors and relationships for inferring concentrations (i.e., sc aling factors) of all 10 CFR 61 nuclides from easy to measure gamma-emitting species.

Individual waste stream activities and concentrations are determined for each batch prior to shipment for disposal. The evaluation of historical evidence is used as a screen to determine the appropriate disposal contai ner followed by a formal evalua tion based on actual samples or dose to curie determinat ions as appropriate.

11.4.3.5 Processing Methods (Wet Wastes)

The dewatering units are portable systems contai ning all necessary equipm ent and controls for removing the free water from ion-exchange resins and filter media.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-026 11.4-11 Containers used for de watering are furnished to CGS with factory installed "internals" functionally identical to thos e used during qualification testing. The determination of "functionally identical" may be performed by the vendor in accordance with their NRC approved Quality Assurance program or by CG S in accordance with th e Quality Assurance program. The internals are free-standing and self-supporting, without protuberances which might damage the container. A fill head interfaces with the container and liner dewatering internals. The dewatering equipment is operated using the la beled controls on the control panel.

The resins that are below th e Class A limits of 10 CFR 61 ar e normally dewatered in carbon steel containers. Resins exceeding Class A limit are dewatered in HICs to provide waste form stability.

The RDS process includes use of moisture indicators and the SE DS process uses dewatering verification to verify that free liquid criteria are met.

11.4.3.6 Control Instrument ation and Sampling Program

Processing of radwaste at CGS is conducted using instrumentation and controls for each batch to ensure that (a) suitable, well characterized waste is delivered to the various waste processing subsystems, (b) adequate process control information is provided to system operators to ensure adherence to proper operating parameter limits, such as tank levels, flow rates, release concentrations, etc

., and (c) sufficient information is available to limit personnel exposures in conformance with the ALARA program.

The control instrumentation used at CGS includes in-process instruments, as well as portable radiation monitoring instruments.

The sampling program at CGS is a twofold program. Individual waste streams are characterized and classified by CGS personnel w ith the responsibility for shipping radioactive waste. Additionally, samples are sent to an offsite laboratory on a periodic basis to validate scaling factors.

11.4.3.7 Maintenanc e and Calibration The control of the waste processing system is ensured by a thorough preventative and corrective maintenance program described in plant procedures for scheduled maintenance systems and maintena nce work requests.

The control instrumentation is an integral pa rt of the process system. The instruments providing the controlling functions are calibrated on a predetermi ned schedule and after each maintenance activity as applicable.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-026 11.4-12 The periodic verification of calibration or recalibration assures that the process control program associated instruments are maintained and that conditions within the system are known.

The maintenance and calibration activities are performed in accordance with written procedures.

11.4.3.8 Waste Processing System Capacity Wastes can be dewate red in up to 210 ft 3 liners in approximately 8 to 12 hr. This gives two shift operation a processing capacity of at least 100,000 ft 3 of waste per year. Similar processing capacity is availabl e from portable solidification equipment which could be moved onsite if solidificati on to provide stabilization became necessary.

Dry wastes are segregated and monitored to reduce volumes where practicable. The expected design basis volume of radwaste is approximately 20,000 ft 3/year, however, actual volumes have been reduced to less than 10,000 ft 3/year. Forced outages and refueling outages increase volume but can be limited by preplanning material usage.

The radwaste processing capac ity at CGS is sized to provi de the needed capacity for anticipated occurrences and nor mal operation. This includes wet wastes, liquids, and solid wastes. Table 11.4-1 lists some of the major flowrates and capacities for several of the waste processes.

Table 11.4-2 tabulates the major equipment ite ms in the permanently installed waste processes.

11.4.3.9 Waste Storage Capacity

The storage capacity for DAW is adequate due to cont ainerization allowing outside storage prior to shipping.

For the liners currently in use (up to 210 ft 3 capacity), approximately 15 liners can be readily accommodated. This storage space is available for the portable waste processing system and to store processed liners. Sufficien t casks and transport packages have been factored into the planning to allow shipment of processed radwaste at a rate to preclude storage problems.

11.4.3.10 Compliance With ALARA Principles

a. Facility layout for the portable de watering equipment - Th e portable dewatering systems are designed to meet Regulator y Guide 8.8. The waste container filling, capping, and decont amination operations are ei ther performed remotely from a control area or as quickly as possible to minimize exposures. The system operator can view container f illing and processing operations using a television monitoring system to prevent overfilling the waste container.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 11.4-13 The portable system is locat ed to limit the exposure of personnel to high dose rate piping and is designed to minimize the accumulati on of crud deposits in the system. Piping and pumps are designed to allow complete flushing and, when possible, piping is flushed pr ior to maintenance. The control area is located in a relatively low radiation area away from the dewatering operation. Equipment with clean components is segregated from the areas containing waste except in the processing area. This allows ma intenance on non-contaminated equipment to be performed in low radiation areas.

The placement of the portable proces sing system was determined based on criteria from Regulatory Guide 8.8 on ALARA and quality assurance provisions from Regulatory Guide 1.143. Processing radiation shields, in-cask processing and/or careful planning of processing activities contri bute to efforts to minimize exposures from radwas te system operations.

b. Plant layout for ALARA - The installed radwaste processing system (excluding the portable dewatering system) was desi gned for remote operation from the radwaste control room in the 467-ft elev ation of the radwaste building. The control room is designed with visual aids mimicking the processes being controlled. This allows for visu al indication of processing status.

c. Exposure control - The radwaste operators contro l the processing of radioactive waste from the radwaste control room that is located in a low radiation area.

Most radwaste processing system components and systems can be remotely aligned from the radwaste control room without going into the radiation areas.

This allows flex ibility of operation and ensu res processing capacity is maintained with reduced exposure to personnel. The systems and components can be flushed remotely from the control room reducing ra diation levels for maintenance activities, also helping meet the ALARA concept.

The installed systems at CGS are sized with the capacity to process peak volumes of waste to reduce the effluents offsite, meeting criter ia of 10 CFR 50, Appendix I.

The processing control from a remote cont rol room and the ab ility to process all waste streams meets the intent of ALARA for both onsite and offsite.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-018 11.4-14 11.4.3.11 Unanticipated Wastes

The waste streams at CGS have been characterized for normal expected waste components.

Periodically, there may be wast e produced from operations su ch as decontamination or cleaning that have not been characterized.

These wastes are classified and processed for di sposal based on the prac tical experience of the CGS staff. It is unlikely that any wastes will be produced at CGS that cannot be prepared for disposal in accordance with 10 CF R 61. When changes in the sy stem or process do occur that are out of the ordinary, proce ssing requirements will be determined on a case-by-case basis.

11.4.3.12 Waste Classification

Waste is classified at CGS to determine if wastes meet 10 CF R 61 Class A, B, or C. The individual waste stream will be sampled and an alyzed for nuclides in Tables 1 and 2 of 10 CFR 61 to establish the waste classification.

This analysis may in many instances use scaling factors to tie difficult to measure nuclid es, to more easily identified gamma emitting nuclides. The requirements for the determination of these scali ng factors is administratively controlled by plant procedure.

11.4.3.13 Waste Packaging and Shipping

The radioactive waste shipped fr om CGS for disposal or furthe r processing is containerized, prepared, and shipped in accordance with applicable state, DOT, and NRC regulations.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 Table 11.4-1 Was t e Processing Systems Capaciti e s LDC N-9 8-1 3 9 11.4-15 Was t e Processing Systems Flow Rates and Capac i t i es Liquid Waste EDR subsystem 14,3 4 1 to 104,826 gpd EDR storage 135,000 gal FDR subsy s tem 5700 to 55,289 gpd FDR storage 40,000 gal Chemical waste subsy s tem 1500 to 20,600 gpd Chemical waste storage 65,000 gal Filter sludge and chemical w a ste concentrates Normal 8000 to 16,000 f t 3/year Processed volume for d i sposal 8000 to 16,000 f t 3/year Portable solidification system Normal 200 ft 3/day Maximum 100,000 ft 3/yea r a Storage 72 50-f t 3 liners Dry compactible waste Compactible radwaste 20,000 ft 3/year Dewatering system Normal 200 ft 3/12-hr shift Maximum 100,000 ft 3/yea r a a Figured on two shifts w ith increa s e in personnel.

Table 11.4-2 Solid Waste Management S y stem Major Equipment Items Equipment Number Required Construction Design Pressure Design Temperatu r e Capacity C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.4-16Cleanup phase separators 2 Stainless steel shell and internals Atmospheric 250 F 4500 gal each Cleanup sludge discharge mixing pump 1 Stainless steel 150 psig 150 F 210 gpm at 170 ft TDH Cleanup decant pump 1 Stainless steel 150 psig 150 F 53 gpm at 50 ft TDH Condensate backwash

receiving tank 1 Stainless steel shell and internals Atmospheric 150 F 19,000 gal Condensate backwash

transfer pump 1 Stainless steel 150 F 450 gpm at 50 ft TDH Condensate phase

separator 2 Epoxy-coated carbon steel shell, stainless steel internals Atmospheric 250 F 23,500 gal each Condensate sludge discharge mixing pump 1 Stainless steel 150 psig 150 F 420 gpm at 160 ft TDH Condensate decant pump 1 S t ainless steel 150 psig 150 F 450 gpm at 50 ft TDH Waste sludge phase separator tank 1 Epoxy-coated carbon steel, stainless steel internals Atmospheric 150 F 13,000 gal Waste decant pump 1 Stainless steel 150 psig 150 F 53 gpm at 50 ft TDH Waste sludge discharge mixing pump 1 Stainless steel 150 psig 150 F 210 gpm at 105 ft TDH

Table 11.4-2 Solid Waste Management System Major Equipment Items (Continued)

Equipment Number Required Construction Design Pressure Design Temperatu r e Capacity C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 LDC N-0 2-0 2 9 11.4-17 Spent resin tank 1 Stainless steel shell and internals Atmospheric 150 F 1200 gal Spent resin pump 1 Stainless steel 150 psig 150 F 21 gpm at 105 ft TDH Decontamination solution concentrated waste t a n k a 2 Stainless steel shell and internals Atmospheric 150 F 700 gal each Decontamination solution concentrated waste pum p a 1 Stainless steel 150 psig 150 F 30 gpm at 70 ft TDH Concentrated waste

measuring tank a 1 Stainless steel Atmospheric 150 F 400 gal Transfer dolly b 1 a Not in service.

b Track riding dolly for transfer of waste containers betw een the storage area and truck shipping area.

Table 11.4-3 Significant Isotope Acti v ity in Dewatered Waste Stream Clean up Sludge Waste Sludge Equipment Drain Resin Floor Drain Resin Condensate Sludge Batch Solid Production 524 lb/60 days 220 lb/3.4 days 1539 lb/66 days 1539 lb/67 days 33 0 0 lb/18.5 days Isotopes Ci/f t 3 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.4-18 89 Sr 0.63 -- 9.0 x 10-3 6.5 x 10-5 4.8 x 10-3 90 Sr 0.16 -- 1.0 x 10-3 7.2 x 10-6 8.4 x 10-4 91 Sr -- -- 5.2 x 10-3 3.7 x 10-5 -- 92 Sr 0.46 -- 1.0 x 10-3 3.4 x 10-5 7.1 x 10-6 90 Y 0.16 -- 1.0 x 10-3 3.4 x 10-5 8.4 x 10-4 91 Y 0.023 -- 2.3 x 10-3 -- 1.7 x 10-3 91m Y -- -- 3.6 x 10-3 -- -- 92 Y -- -- 1.0 x 10-3 -- -- 95 Zr 0.29 8.1 x 10-4 -- -- 4.2 x 10-4 95 Nb 0.44 1.1 x 10-3 8.4 x 10-4 -- 6.3 x 10-4 99 Mo -- 2.0 x 10-3 2.9 x 10-3 2.4 x 10-5 1.3 x 10-4 99m Tc 0.0015 1.3 x 10-3 9.7 x 10-3 1.1 x 10-4 1.3 x 10-4 103 Ru -- -- -- --

2.1 x 10-4 122 Sb 0.020 2.0 x 10-4 1.1 x 10-5 -- -- 129m Te 0.038 -- 1.3 x 10-4 6.9 x 10-7 -- 129 Te 0.024 -- -- -- -- 132 Te -- -- 1.6 x 10-2 1.3 x 10-4 1.3 x 10-4 83 Br -- -- 1.3 x 10-4 4.5 x 10-6 --

Table 11.4-3 Significant Isotope Activity in Dewatered Was t e (Continued)

Stream Clean up Sludge Waste Sludge Equipment Drain Resin Floor Drain Resin Condensate Sludge Batch Solid Production 524 lb/60 days 220 lb/3.4 days 1539 lb/66 days 1539 lb/67 days 33 0 0 lb/18.5 days Isotopes Ci/f t 3 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.4-19 131 I 0.077 7.8 x 10-4 1.01 x 1 0-2 7.4 x 10-5 2.3 x 10-2 132 I -- -- 1.7 x 10-3 3.5 x 10-5 1.3 x 10-4 133 I -- -- 7.3 x 10-3 7.7 x 10-5 -- 134 I -- -- 8.4 x 10-4 4.7 x 10-5 -- 135 I -- -- 3.6 x 10-3 5.7 x 10-5 -- 134 Cs 0.17 3.0 x 10-3 6.0 x 10-4 4.9 x 10-6 4.2 x 10-4 136 Cs 0.0078 --

1.3 x 10-4 -- -- 137 Cs 0.18 4.4 x 10-3 1.0 x 10-3 7.6 x 10-6 8.4 x 10-4 138 Cs -- -- 2.1 x 10-4 -- -- 13 7 m Ba 0.17 -- 1.0 x 10-3 -- 6.3 x 10-4 139 Ba -- -- 8.4 x 10-4 3.9 x 10-5 -- 140 Ba 0.076 -- 1.0 x 10-2 2.6 x 10-5 2.7 x 10-3 141 Ce -- -- 1.3 x 10-4 -- 3.8 x 10-3 144 Ce 0.019 -- 1.3 x 10-4 4.7 x 10-6 6.3 x 10-4 140 La 0.087 3.9 x 10-5 1.0 x 10-2 -- 2.9 x 10-3 141 La -- -- 2.1 x 10-4 -- -- 142 La -- -- 1.3 x 10-4 -- -- 143 Pr 0.022 -- 1.3 x 10-4 -- 8.4 x 10-4 Table 11.4-3 Significant Isotope Activity in Dewatered Was t e (Continued)

Stream Clean up Sludge Waste Sludge Equipment Drain Resin Floor Drain Resin Condensate Sludge Batch Solid Production 524 lb/60 days 220 lb/3.4 days 1539 lb/66 days 1539 lb/67 days 33 0 0 lb/18.5 days Isotopes Ci/f t 3 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.4-20 144 Pr 0.019 -- 1.3 x 10-4 -- 6.3 x 10-4 239 Np -- -- 5.4 x 10-2 4.5 x 10-4 -- 51 Cr 1.8 1.5 x 10-2 4.4 x 10-4 4.1 x 10-6 3.1 x 10-3 54Mn 0.089 2.1 x 10-3 1.3 x 10-4 7.0 x 10-5 6.6 x 10-4 56Mn -- 2.3 x 10-4 2.2 x 10-4 7.8 x 10-6 -- 58 Co 1.8 3.3 x 10-3 8.4 x 10-3 6.1 x 10-5 6.2 x 10-2 60 Co 0.65 2.4 x 10-2 1.1 x 10-3 9.2 x 10-6 9.7 x 10-3 59 Fe 0.018 2.6 x 10-4 1.3 x 10-4 8.3 x 10-7 8.8 x 10-4 65 Zn 2.1 7.4 x 10-2 3.5 x 10-4 2.3 x 10-4 4.4 x 10-4 110m Ag 0.14 8.8 x 10-6 1.3 x 10-4 9.0 x 10-6 1.1 x 10-3 187 W -- 1.5 x 10-4 1.3 x 10-4 1.4 x 10-6 -- Total 9.67 1.4 x 10-1 1.7 x 10-1 1.7 x 10-3 1.2 x 10-1 Table 11.4-4 Expected Annual Pro d uction of Solids C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 11.4-21 ft 3/year Normal Activity Ci/Containe r a Maximum Activity Ci/Containe r a Cleanup filter demineral i zer sludge 720 1196 1820 Condensate filter deminer a lizer sludge 7460 19 40 Waste floor drain and fuel pool filter demineralizer sludge 1000 0.2 3.4 Waste demineralizer resin 540 29 330 Total volume (f t 3) 9720 a Based on 164 ft 3 of dewatered resi n per container

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report11.4-1 30 M536Flow Diagram Radioactive Waste Disposal Solids Handling SystemRev.FigureDraw. No.Amendment 61December 2011 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 11.5-1 11.5 PROCESS AND EFFLUE NT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS The process and effluent radiological monitori ng and sampling systems are provided to allow determination of the content of radioactive material in various gaseous and liquid process and effluent streams. The design objective and criteria are primarily determined by the system designation of either

a. Instrumentation systems required for safety, or b. Instrumentation systems required for plant operation.

11.5.1 DESIGN BASIS

11.5.1.1 Design Objectives

The process and effluent radiol ogical monitoring and sampling system is designed to provide for compliance with the requirements of 10 CF R Part 50 including the General Design Criteria (GDC) of Appendix A and provides the monitoring and sampling required to make measurements, evaluations, and reports recomme nded by Regulatory Guid e 1.21, Revision l.

11.5.1.1.1 Systems Required for Safety

The main objective of radiation monitoring syst ems (RMS) required for sa fety is to initiate appropriate protective action to limit the release of radioactive materials from the reactor vessel and reactor building if predetermined radiation levels are exceeded in major process/effluent streams and to limit inflow of airborne radi oactivity to the control room following an accidental release.

Additional objectives are to have these systems available under all operating conditions, including accidents and postaccidents, and to provide control room personnel with an indication of the radiation levels in the major process/effluent streams plus alarm annunciation if high radiation levels are detected.

The RMS provided to meet these objectives are

a. Main steam line RMS,

b. Reactor building ventila tion exhaust plenum RMS, c. Control room fresh air intake RMS, and

d. Standby service water RMS.

11.5.1.1.2 Systems Required for Plant Operation

The main objective of the RMS is to provide operating pers onnel with measurement of the content of radioactive materials in all effluent and important process streams. This allows demonstration of compliance with Technical Specifications by pr oviding gross radiation level C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-052 11.5-2 monitoring and collection of halogens and particulates on cartridges and filters (gaseous effluents). Additional objectives are to initiate discharge valv e isolation on the offgas, liquid radwaste, or drain systems if predetermined release rates are exceeded and to provide for sampling at certain radiation monitor locations to allow determ ination of specific radionuclide content.

The RMS provided to meet these objectives are

a. For gaseous process streams

1. Offgas pretreatment RMS, 2. Offgas post-treatment RMS,

3. Charcoal bed vault RMS, and

4. Mechanical vacuum pump exhaust RMS.

b. For gaseous effluent streams

1. Reactor building elevated release duct RMS,

2. Turbine generator building ventilation release duct RMS, and 3. Radwaste building ventilation release duct RMS.

c. For liquid process streams

1. Standby servi ce water RMS and 2. Reactor building clos ed cooling water RMS.

d. For liquid effluent streams

1. Radwaste effluent RMS,

2. Circulating water (blowdown) RMS,

3. Standby service wa ter RMS, and

4. Plant service water (TSW) RMS.

e. For primary containment monitoring

1. Leak detection RMS and 2. Loss-of-coolant accident (LOCA) tracking RMS.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.5-3 11.5.1.2 Design Criteria 11.5.1.2.1 Systems Required for Safety The design criteria for the RMS required for safety are that the systems shall:

a. Withstand the effect of natural phenomena (e.g., ear thquakes) without loss of capability to perform their functions, b. Perform its intended safe ty function in the environment resulting from normal and postulated accident conditions, c. Meet the reliability , testability, independe nce, and failure m ode requirements of engineered safety features,

d. Provide continuous output on control room panels,

e. Permit checking of the operational availability of each channel during reactor operation with provision for calibration function and instrument checks,

f. Ensure an extremely high probability of accomplishing its safety function in the event of anticipated operational occurrences,

g. Initiate prompt protective action prior to exceeding Technical Specifications limits,

h. Provide warning of incr easing radiation leve ls indicative of abnormal conditions by alarm annunciation, i. Insofar as practical, provide self-monitoring of compon ents to the extent that power failure or component malfunction causes annunciation and channel trip,

j. Register full scale output if radiation detection exceeds full scale, and

k. Have sensitivities and ra nges compatible with anti cipated radia tion levels.

The applicable GDC are 1, 2, 3, 4, 13, 19, 20, 21, 22, 23, 24, 29, 60, and 64 (see Section 3.1). The systems shall meet the design re quirements for Safety Class 2, Seismic Category I systems, along with the quality assurance requirements of 10 CFR Part 50, Appendix B.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.5-4 11.5.1.2.2 Systems Required for Plant Operation

The design criteria for operational RMS are that the systems shall

a. Provide continuous indication of radiation levels in the main control room,

b. Provide warning of incr easing radiation leve ls indicative of abnormal conditions by alarm annunciation, c. Insofar as practical, provide self-monitoring of compon ents to the extent that power failure or component malfunction causes annunciation and, for systems initiating protective ac tion, channel trip, d. Monitor a sample representativ e of the bulk stream or volume,

e. Have provisions for calibration, function and instrumentation checks,

f. Have sensitivities and ra nges compatible with antici pated radiati on levels and Technical Specifications limits, and

g. Register full scale output if radiation detection exceeds full scale.

The RMS monitors discharges from the gaseous and liquid radwaste treatment systems and nonradioactive sumps have provision s to alarm and to initiate auto matic closure of the effluent discharge valves on the affected treatment systems prior to exceedi ng the normal operation limits specified in th e Technical Specifications. Add itionally, the primary containment monitoring system meets the criteria, except for item g.

The applicable GDC are 13, 60, and 64 (see Section 3.1).

11.5.2 SYSTEM DESCRIPTION

11.5.2.1 Systems Required for Safety

Information on these systems is presented in Tables 11.5-1 and 11.5-2 and the arrangements shown in Figures 11.5-1 through 11.5-4. The equipment is designe d to Quality Class I and Seismic Category I requirements. High reliability is further achieved by the use of redundancy as noted below.

11.5.2.1.1 Main Steam Line Radiation Monitoring System

This system monitors the gamma radiation level exterior to th e main steam lines. The normal radiation level is produced primarily by coolan t activation gases plus smaller quantities of C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.5-5 fission gases being tran sported with the steam. In the event of a gr oss release of fission products from the core, this monitoring system provides signals to the following:

a. Reactor water sample valves, b. Mechanical vacuum pumps, c. Mechanical vacuum pumps isolation valves, d. Gland seal exhausters, and e. Control room annunciators.

The system consists of four redundant instrument channels. E ach channel consists of a local detector (gamma-sensitive ion chamber) and a control room ra diation recorder and a readout module with an auxiliary trip unit. Power for two channels (A and C) is supplied from the reactor protection system (RPS) bus A and for the other two channels (B and D) from RPS bus B. Channels A and C are physically and el ectrically independent of channels B and D.

The detectors are located near th e main steam lines in the steam tunnel as it enters the turbine building. The detectors are geometrically arranged so that this system is capable of detecting significant increases in radia tion level with any number of ma in steam lines in operation.

Each radiation monitor has four trip circuits: two upscale (high-high and high), one downscale (low), and one inoperative. Each trip is visua lly displayed on the aff ected radiation readout module. A high-high or i noperative trip results in a channel tr ip in the auxiliary unit which is an input to the reactor water sample valves, mechanical vacuum pump shutdown, and discharge valve closure. A high trip actuates a main steam line (MSL) high control room annunciator common to all channels. A downscale and inope rative trip actuates a MSL downscale/inoperative control room annunciator co mmon to all channels. High and low trips do not result in a channel trip.

Each radiation monitor displays the measured radiation level.

Arrangement details are shown in Figure 11.5-1.

11.5.2.1.2 Reactor Building Exhaust Pl enum Radiation Monitoring System

This system monitors the radiation level of the reactor building ventilation system exhaust plenum prior to its discharge from the building into the elevated release duct. A high radioactivity level in the exha ust system could be due to fi ssion gases from a leak or an accident.

The system consists of four redundant instrument channels. E ach channel consists of a local detection assembly (a sensor and converter unit containing a Geiger-Mueller (GM) tube and electronics) and a control room radiation readout module. The 120-V ac power for channels (A and B) is provided from Division 1, and fo r channels (C and D) from Division 2 power panels; the multipoint strip char t recorder supplied from the Division 2 uninterruptible power supply (UPS) power panel records the output of all four channels. Th e detection assemblies C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 11.5-6 are located outside the exhaust air plenum upstream of the s econdary containment discharge isolation valves. The distance upstream from the inboard discharg e isolation valve is such that, at the maximum design flow, the transport time from the det ector location to the inboard discharge valve is greater than the total time required to res pond to trip leve l radiation and close the inboard discharge valve before exceeding 10 CFR 50.67 dose limits (see Section 9.4.2.3).

Each radiation monitor has two trip ci rcuits: one upscale (high-high) and one downscale/inoperative (fail safe de sign). Two-out-of-two upscale/

downscale trips (channels A and B) initiates closure of the reactor building ven tilation outboard isolat ion valves and the primary containment outboard purge and vent valves, and initiates startup of standby gas treatment (SGT) system train B. The same c ondition for channels C and D initiates closure of the corresponding inboard valves a nd initiates startup of SGT train A.

An upscale trip is displayed on the affected radiation readout module and actuates a reactor building vent high-high radia tion control room annunciato r common to all channels. A downscale trip is also displayed on the ra diation readout module and actuates a reactor building vent downscale control room annunciator common to all channels.

An additional trip signal for high radiation alarm is provided by the recorder and actuates a reactor building vent high radiation control room annunciator. Each radiation monitor displays the measured radiation level.

Arrangement details are shown in Figure 11.5-2.

11.5.2.1.3 Control Room Fresh Air Intake Radiation Monitoring System This monitoring system measures the radioactivity in the two remo te fresh air intake lines to the main control room. In the event of a release of abnormal gaseous radioactivity from the plant and the transporting of this radioactivity by wind currents to the remote air intakes, the monitoring system provides an alarm in the control room. The system consists of two divisionally separate d channels. Each channel consists of redundant local detectors (beta scintillation type) and redundant control room indicator-trip units, alarms, and output to a recorder.

Required 120-V ac supply for Divi sion 1 and 2 equipment in both the main control room and remote locations is provided on a divisional basis by 120/240-V ac critical (Class 1E) instrumentation power system.

Gas samples are withdrawn from sample probes in a continuously flowing section of the fresh air intake pipelines. These samp les run in stainless steel tubing to local sample racks located on the 441-ft level of the radw aste building. The four divisionally separated local cabinets each have a detector and preamplifier, blower, and sample flow rate measuring equipment.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 11.5-7 The detectors are housed in lead shields to minimize the effects of back ground radiation and enhance response to low level radioactivity.

Associated radiatio n readout modules and recorders are mounted in the main control room.

Each radiation monitor has three trip circuits: one upscale fo r high radiation, one upscale for high-high radiation, and one downscale for instrument inoperative. Al l alarms annunciate in the main control room.

Arrangement details are shown in Figure 11.5-3.

11.5.2.l.4 Standby Service Water Radiation Monitoring System

This system monitors gamma radi ation levels of the service wa ter liquid process and effluent streams.

Each monitor system consists of a gamma scintillation detector inserted into an offline chamber to which a process stream sample is piped. The detector locations are selected to obtain a reasonable geometry and are positioned away from crud trap and associated high background regions. Lead shielding is provided to further reduce background levels.

At each liquid offline detector location, a c ontinuous sample from the liquid process pipe passes through a shielded detec tion assembly for gross radiation monitoring and then is returned to the process pipe. The detection assembly consists of a detector mounted in a shielded sample chamber. Th e local radiation monitor and th e meter in the control room display the measured gross radiat ion level. The sample chambe r and lines can be drained to allow assessment of background buildup. The flow meter at the samp le rack provides local sample line flow indication.

Sample flow for each detector is from the standby service piping downstream of each of the two residual heat removal (RHR) heat exch anger (loops A and B). These monitors are designed to detect any primary coolant leakag e into the standby service water through the RHR heat exchanger, during operation of the RHR heat exchangers in the shutdown heat removal mode.

Additional details are shown in Figure 11.5-4.

11.5.2.2 Systems Required for Plant Operation All systems associated with the plant process cycle provide for indication and recording of radiation levels in the main control room in conjunction w ith alarm annunciation features.

Information on these systems is presented in Tables 11.5-1 and 11.5-2 and the arrangements are shown in Figures 11.5-2 and 11.5-4 through 11.5-10.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-014 11.5-8 11.5.2.2.1 Gaseous Proces s and Effluent Radiation Monitoring System

11.5.2.2.1.1 Offgas Pretreat ment Radiation Monitoring System. This system monitors radioactivity at the outlet of the water separator downstream of the catalytic recombiners. The monitor detects the radiation leve l which is attributable to the fission gases produced in the reactor and transported with steam through the turbine to the condenser.

A continuous sample is extracted from the offgas pipe via a stainl ess steel sample line that is then passed through a sample cham ber and a sample panel before being returned to the suction side of the steam jet air ejector (SJAE). The sample chamber is a steel pipe which is internally polished to minimize plateout. It can be purged with room ai r to check detector response by using a three-way solenoid-operate d valve. The valve is contro lled by a switch located in the main control room. The sample panel me asures and indicates sample line flow.

The detector is a gamma-sensitive ionization chamber mounted extern al to the sample chamber.

The channel has a logarithmic radiation readout module which provides a system alarm output and is provided with a recorder.

The 120-V ac UPS power for count ratemeter and trip auxiliary circ uit is supplied from critical Division 2 panel; strip chart recorder receives reliable (Division B) instrument power, and local 120-V ac for the offgas sample and vial sampler control panel.

The radiation readout module has four trip ci rcuits: two upscale (h igh-high and high), one downscale (low), and one inoperative. The trip outputs are used for alarm function only. Each trip is displayed on the radiation monitor a nd actuates a control room annunciator: offgas high-high, offgas high, a nd offgas downscale/inopera tive. Sample line flow is displayed at the sample panel.

The radiation level output by the monitor can be di rectly correlated to th e concentration of the noble gases by using the semiautoma tic vial sampler panel to obtai n a grab sample. To draw a sample, a serum bottle is inserted into a sample chamber, the sample lines are evacuated, and a solenoid-operated sample valve is opened to allow offgas to enter the bottle. After the sampling bottle is removed, the sample is analyzed in th e counting room with a multichannel gamma analyzer to determine the concentration of the va rious noble gas radionuclides.

A correlation between the observed activity and the monitor reading permits calibration of the monitor.

For arrangement details see Figure 11.5-5.

11.5.2.2.1.2 Offgas Posttreatm ent Radiation Monitoring System. This system monitors radioactivity in the offgas pi ping downstream of the offgas system charcoal vessels and upstream of the offgas system discharge valve. A c ontinuous sample is extracted from the C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 11.5-9 offgas system piping, passed through the offg as posttreatment sample panels for monitoring and sampling, and returned to the offgas system piping. Each sample panel has a cartridge and filter (one for particulate coll ection and one for haloge n collection) in series with filter bypass capabilities (with respect to flow) with two identical con tinuous gross radiation detection assemblies. Each gross radia tion detection assembly consists of a shielded chamber, a radiation detector, and a check source. Two ra diation monitors in the main control room display the measured gr oss radiation level.

The sample panels and shielded chambers can be purged with room air to check detector response by using solenoid valves th at can be operated from the c ontrol room or at the sample panels. The sample panels measure and indicate sample li ne flow. A solenoid operated check source for each detection assembly operated from the control room or locally can be used to check operability of the gr oss radiation channel.

Channel A monitor receives reliable (Division A) 120-V ac power and ch annel B is supplied with Division 2 power; the +/-24-V dc auxiliary tr ip circuits are supplied internally from the

monitor power supply. Control room chart reco rder receives reliable (Division B) instrument power. Offgas posttreatment sample panel 11A is provided 120-V ac (D ivision A) power and offgas posttreatment sample panel 11B is provided 120-V ac (Division B) power.

Each radiation readout module ha s four trip circuits: two ups cale (high-high-high, and high), one downscale (low), and one inoper ative. Each trip is visually displayed on the radiation monitor. The first three trips actuate corresponding control room annunciators: offgas posttreatment high-high-high radiation, offgas posttreatment high radi ation, and offgas posttreatment downscale. A high-high trip from the recorder actuates an offgas posttreatment high radiation annunciator in the main control room. High or low sample flow measured at the sample panel actuates a control room offgas posttr eatment trouble annunciator.

A trip auxiliary unit in the control room takes the high-high-high and downscale trip outputs and, if its logic is satisfied, initiates closure of the offgas sy stem discharge and drain valves.

The logic is satisfied if two high-high-high, one high-high-high and one downscale, or two downscale trips occur. The high-high-high trip setpoints are determined such that valve closure is initiated prior to exceeding Technical Specifications limits. Any one high upscale trip initiates closure of the offgas system bypass line valve and initiates opening of the charcoal adsorber treatment line valve.

Grab sampling functions required for isotopic analysis and gross monitoring calibrations can be performed at each redun dant sample panel.

For arrangement details see Figure 11.5-5.

11.5.2.2.1.3 Offgas Charcoal Bed Vault Radiation Monitori ng System. The charcoal bed vault air handling room is monitored for an in crease in the gross gamma radiation level from

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 11.5-10 leakage of radioactive noble gases out of the treatment system. The channel includes a sensor, indicator, trip unit, and a locally mounted auxiliary unit. The detector is mounted outside of

the air handling room door. The insu lated door only attenua tes 80 keV gamma (133 Xe) by approximately 40%. An indicator and trip unit is located in the main control room. The channel provides for sensing and readout, both local and remote , of gamma radiation over a range of six logarithmic decades (10

+0 to 10+6 mR/hr).

The indicator and trip unit has one adjustable upscale trip circuit for alarm and one downscale trip circuit for instrument inoperative which a nnunciates in the main co ntrol room. The trip circuits are capable of convenien t operational verification by means of test signals or through the use of portable gamma sources. Power is s upplied from the channel A power supply of the reactor handling ventilati on exhaust plenum RMS.

For arrangement details see Figure 11.5-2.

11.5.2.2.1.4 Mechanical Vacuum Pump Exhaust Radiation Monitoring System. The radiation monitor on the mechanical vacuum pump exhaust is designed to alarm, stop, and isolate the mechanical vacuum pumps in the case of high le vel of radioactive gases in air being exhausted to the reactor building elevated release duct. The mechanical vacuum pump is operated during plant startups to remove bulk air from the condenser and is secured at the point where the steam jet air ejection suction is available a nd condenser offgases ar e routed through the recombiner charcoal pro cess treatment system. In addition to monitoring discharges via the mechanical vacuum pumps, th e turbine gland seal air exha uster system is continuously monitored via this process radiation monitoring system. Clean seali ng steam is used on the turbine gland seals to maintain the releases of radionuclides to as low as is reasonably achievable (ALARA) limits. The monitor complie s with GDC 64 and is Quality Class II and Seismic Category II.

The channel includes an energy-co mpensated GM detector, a readout module, and recorder in the main control room. The ch annel provides for sensing and readout, both local and remote, of gamma radiation ove r a range of four logarithmic decades (10

-2 to 10 2 mR/hr).

The indicator and trip unit has two adjustable trip circuits. The upscale tr ip circuit generates a high radiation alarm and stops th e mechanical vacuum pumps. The adjustable downscale trip circuit generates an instrument alarm.

For arrangement details see Figure 11.5-6.

11.5.2.2.1.5 Reactor Building Elevated Release Duct Radiation Monitoring System. This monitoring system measures radi oactivity in the reactor building elevated release duct from the gland seal and mechanical vacuum pumps, the treated offgas effluent, the SGTS exhaust, and the exhaust air from the reactor building ventilation. This system consists of two main subsystems, an offline sampling system and an inline gamma radiati on monitoring system.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 11.5-11 The sampling system is used to comply with Regulatory Guide 1.21, Revision 1, and as such satisfies GDC 64. A continuous representative sample is extract ed from the el evated release duct through an isokinetic probe, passes down a 50-ft vertical tube, through a filter to collect particulates, and through an im pregnated charcoal cartridge to collect iodine. The sample travels through a flow indicator and then a sample pump prior to being re turned downstream to the sampling point. Samples are analyzed at least w eekly to determine th e quantities of the specific radionuclides released. Th e sample flow rate is indicated and tota lized locally and the exhaust duct flow is also indicated locally. Both of these variables are recorded and alarmed in the control room and are input to the transient data acquisition system (TDAS). Arrangement details are shown in Figure 11.5-6. Table 11.5-3 lists sampling freque ncies and required sensitivities.

The gamma radiation monito ring system uses three separate de tectors. The lo w range detector is used to satisfy Regulatory Guide 1.21, Revision 1, and as such, complies with GDC 64.

The intermediate and high range detectors, along with the LOCA detectors of Section 11.5.2.2.3.2 , are used to satisfy NUREG-0737 and Regulatory Guide 1.97, Revision 2, requirements and, as such, comply with GDC 13 and 64. This system is an EG&G ORTEC Gamma Spectroscopy system. It uses inline detectors to provide an isotopic analysis of the reactor building elevated release duct effluents. Tw o cryogenically cooled, high purity germanium coaxial detectors provide a range of 1.8 x 10

-5 to 7.8 x 10 4 Ci/cm 3 with overlap.

a. Intermediate range Approximately 1.8 x 10

-5 to 1.8 x 10 0 Ci/cm 3 (133 Xe) b. High range Approximately 7.8 x 10

-1 to 7.8 x 10 4 Ci/cm 3 (133 Xe) These detectors provide approxi mately 40% efficiency to a 1332.5 keV gamma [c ompared to a 3 in. x 3 in. NaI (T1) detector at 25 cm]. Both detectors monitor activ ity through the release duct and are mounted in the reactor building at el evation 618 ft 7 in. Lead enclosures provide shielding from postaccident background radiation. Collimator design and detector locations ensure representative sampling.

A third high efficiency detector is located insi de the release duct to monitor low level normal operation activity. The usable range for this detector is approximately 3.3 x 10

-8 to 3.3 x10-3 Ci/cm 3. The high efficiency de tector provides approximately120% efficiency to 1332.5 keV gamma

[compared to a 3 in. x 3 in. Na I (T1) detector at 25 cm].

The postaccident system consis ts of a computer controlled acquisition and analysis loop, located in radwaste building el. 525 ft, feeding an additional computer located in the control room. Either computer can control detector, si gnal processing, and spectra l analysis functions.

Commercial application software from ORTEC is used for both the control of the detection process and nuclide identification.

The libraries which are used for nuclide identification are

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 11.5-12 based on Tables 11.3-8 and 15.6-10 and 15.6-11. The control room computer is used for system status monitoring a nd data output. System status is monitored and alarmed.

This system provides two types of information and each has its differen t response time. Gross gamma, in counts/sec, originates at the local acquisition and analysis panel from a log count-rate meter fed directly from the spectroscopic amplifier.

This real time si gnal is then input directly to the trending recorder and TD AS system. The record er updates its digital display and alarm information every 6 sec. This conforms to the re quirement in Regulatory Guide 1.97 to monitor noble gas as a Type C and E, Category 2 variable.

Effluent isotopic information availability is controlled by the detector counting times, which are a function of stack activity. Counting times are increased during periods of low stack activity and will decrease for peri ods of high activity. System re sponses will be the sum of the detector counting and data tran smission times. Accurate response times for both sets of data will be determined via field testing. This conforms to the requi rement in Regulatory Guide 1.97 to monitor particulate and halogen release as a Type E, Ca tegory 3 variable.

This system will satisfy the requirements of Regulatory Guide 1.97, Revision 2, and NUREG-0737 for normal and postaccident mon itoring of noble gases, particulates, and halogens. All equipment is qualified to operate in the required postaccident environment and is supplied by reliable battery-backed power. It is designed to meet the pertinent sections of ANSI N42.18-1980 (formerly ANSI N13.10), "S pecification and Performance of On-Site Instrumentation for Continuous ly Monitoring Radioactivity in Effluents." Additionally, guidelines from ANSI N42.14-1991, "Calibration and Use of Germanium Spectrometers for the Measurement of Gamma-Ray Emission Rates of Radionuclides," ar e used to monitor system performance as part of the surveillance and calibration process. In situ calibration uses National Institute of Standards and Technology (NIST) traceable standards. Additionally, transfer calibrations may be performed on gas samples drawn from the reactor stack and analyzed on NIST referenced equipment when sufficient stack gas activity is present.

Table 11.5-1 lists the sensitivity and range of each detector. Arrangement details are shown in Figure 11.5-11.

These detectors provid e continuous monitoring with overlap from normal plant operation (typically in the mid 10

-8 decade with no failed fuel) to a worst case DBA LOCA with an expected containment activity of approximately 2.92 x 10 4 Ci/cm 3.

The system has built-in electronic test circu its. Calibration curves ha ve been developed from the calibration data equating activity observed on the monitor to the Ci/cm 3 concentration in the effluent. Initial system ca libration parameters were esta blished by comparisons between analysis done via the inline sy stem and an NIST traceable gamma spectroscopy system.

Calibration procedures use transfer and linearity standards. Electrical power for this C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-000 11.5-13 monitoring system is from a reliable power supply. Also see Figure 11.5-11 , and Table 11.5-1.

11.5.2.2.1.6 Turbine Genera tor Building Ventilati on Release Duct Ra diation Monitoring System. This monitoring system measures the radioactivity in the turbine building exhaust prior to its discharge to the environment and in doing so complies with Regulatory Guide 1.21, Revision 1, GDC 64, and NUREG-0737. This monitor detects the fission and activation

products from the steam which may leak from th e turbine or the other primary components in the building. The gaseous activity in the exhaust is expected to normally be below detectable levels. The particulate and iodine activity is typically accumulated on a filter and cartridge respectively for a week to obtain sufficient activity to be detectable. These filters are analyzed to determine the quantities of sp ecific radionuclides pr esent and the results, together with the gaseous activity strip chart recorder, provide a permanent record of radioactivity released to the environment.

A continuous representative samp le is extracted from the exha ust vent through a multi-ported isokinetic probe, down a 30-ft tube to pass through a filter paper to colle ct particulates, and through an impregnated charcoal cartridge to collect iodine. The sample travels through two gas monitors, a local flow indicator, and then a sample pump prior to being exhausted to the radwaste building roof area by the turbine building exhaust fans. The sample flow rate is automatically adjusted to compensate for effluent flow changes.

The gas monitors are mounted in lead-shielded chambers. Each gas ch annel consists of the local detector and preamplifier w ith count rate meter a nd a recorder in the main control room.

Arrangement details are shown on Figure 11.5-7. The normal sample flows thr ough the extended range (10

-2 Ci/cm 3 to 10+3 Ci/cm 3) gas channel and through the low range (10

-6 Ci/cm 3 to 10-1 Ci/cm 3) gas channel, providing an overall range of 10

-6 Ci/cm 3 to 10 3 Ci/cm 3 with one decade of overlap. The normal-range channel is equipped with a radioactive test source while th e extended-range channel has a built-in electronic test circuit. Calibration curves have been developed from the calibration data equating activity obser ved on the monitor to the Ci/cm 3 concentration in the effluent. Initial system calibration was based on factory calibrations using NBS traceable 133 Xe and 85 Kr standards. Current calibration procedures use transfer and linearity standards.

These monitors have no control functions. The low range monitor has two adjustable trip circuits, one high and on e high-high for high radia tion alarm, plus a low trip for instrument inoperative that annunciates in th e main control room. The ex tended range monitor has a low alarm for instrument inoperative that annunciates in the control room. The reliable power for normal/high range radiation monitor(s) and recorder(s) is provided from Division A power panel and backed by standby power.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-000 11.5-14 11.5.2.2.1.7 Radwaste Buildi ng Ventilation Release Ducts Radi ation Monitoring System.

This monitor system measures the radioactivity in the radwaste building ventilation air exhaust as it is being discharged to the environment and in doing so complies with Regulatory Guide 1.21, Revision 1, GDC 64, and NUREG-0737. Radioactivity originates from radwaste tank vents, from primary water processing equipment, and from laboratory sampling hoods, as well as various cubicles havi ng liquid process treatment systems within the building. A continuous sample is drawn from each of the two out of three e xhaust fans that are operating. The sampling uses isoki netic probes with a fixed flow ra te expected to be isokinetic over the normal ventilation operating range. The representative sample is withdrawn through a multi-ported duct probe, through a tube passing through a particulate filter, and through a charcoal cartridge to collect pa rticulate and iodine sa mples, which are rem oved at least weekly for laboratory radiochemical analyses. The filtered air sample streams from the operable exhaust fans are combined to pass through two gas monitors.

The gas monitors are mounted in lead shielded chambers. Each gas channel consists of a local detector and preamplifier with countrate meter and a recorder in the main control room. The normal sample flows through the extended range (10

-2 Ci/cm 3 to 10+3 Ci/cm 3) gas channel and through the low range (10

-6 Ci/cm 3 to 10-1 Ci/cm 3) gas channel.

Arrangement details are shown in Figure 11.5-7. These monitors have no control functions. The low-range monitor has two adjustable trip circuits, one high and on e high-high for high radiation alarming, plus a low for instrument inoperative that annunciates in th e main control room. The ex tended range monitor has a low alarm for instrument inoperative that annunciates in the control room. The reliable power for normal/high range radiation monitor(s) and recorder(s) is provided from Division A, power panel and backed by standby power.

11.5.2.2.1.8 NRC Safety Evaluation Report, NUREG-0892 Acceptance. System comparisons to ANSI 13.10 (redesignated as ANSI N42.18) have been performed.

An inline gamma spectroscopy system provides an isotopic analysis of a ll effluents exiting the reactor building elevated release duct. See Section 11.5.2.2.1.5 , Figure 11.5-11 , and Table 11.5-1.

Particulate/iodine sampling of the other buildings (radwaste a nd turbine) exhausts will be handled by the normal effluent samplers where the postaccident re lease concentration is quite low.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 11.5-15 If there were a reactor accident with a core fission product release, the reactor building (secondary containment) immediately isolates. The atmosphe re is maintained at a 0.25 in. H 2 O vacuum by the SGT system. Th e only potential airborne contam ination that could reach the other buildings is from the SGT system bypass leakage pathwa ys as listed in Table 6.2-16. Shielding needed for the radwas te and turbine building normal effluent sampling systems was evaluated using assumptions cons istent with the discussion in Appendix J (see Sections J.2 and J.3). For the purpose of defining the shielding requirements for the radwaste and turbine building, an iodine concentration of 3.7 x 10

+4 µCi/cm 3 is assumed in the in leakage air. This iodine concentration is based on a 50% core inventory release to the drywell atmosphere and a 50% plate-out factor. This plate-out factor is conservative for determ ining sample shielding requirements. With 0.35 scfh bypass leakage into the radwaste building, the building exhaust (83,000 cfm) concentration will be 2.6 x 10

-3 µCi/cm 3 , ignoring building volu me dilution. The normal effluent sampler operates at 3 cfm; therefore, the ch arcoal cartridge 30 minutes accumulation would be 6.7 mCi. This would resu lt in a dose rate of 21 mR/hr at 1 ft from the cartridge. Doubling the dose rate to account for particulates yi elds 42 mR/hr at 1 ft from the sample assembly. Because of the high exhaust flow rate (260,000 cfm) and less inleakage (0.24 scfh), the turb ine building exhaust is less concentrated.

The radwaste and turbine building normal effluent sampling systems are adequate for postaccident sampling and no shielding is necessary (see Section 11.5.2.2.1.5

). The evaluation for shielding was not repeated for the implementation of the alternative source term (AST).

11.5.2.2.2 Liquid Process and Effluent Radiation Monitoring System

These systems monitor gamma ra diation levels of liquid process and effluent streams.

Each monitor system consists of a gamma scintillation detector inserted into either a well in the process piping or a sump or an offline chamber to which a process stream sample is piped. The detector locations are selected to obtai n a reasonable geometry and are positioned away from crud traps and associated high background regions. Lead sh ielding is provided to further reduce background levels.

At each liquid offline detector location a continuous sample is extracted from the liquid process pipe, passed through a shielded detection assembly for gross radiation monitoring, and then returned to the process pipe. The detection assembly consists of a detector mounted in a shielded sample chamber equipped with a check source. The meter and recorder in the control room displays the measured gross radiation level. The sample chamber and lines can be drained to allow assessment of background buildup. The flow meter at the sample rack provides local sample line flow indication. A solenoid-opera ted check source operated from the control room is used to check operability of the channel response.

The critical +/-24-V dc for monito r(s) is supplied from Division 1 (and 2) power panels, except for the plant service water m onitor which receives 120-V ac from Division A panel, reliable

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 11.5-16 power is provided for control room strip chart recorders, sample control panel receives local 120-V ac.

The detector's local preamplifie r unit is designed to remain fully operational in the expected environment. If exposed to radiation transien ts which exceed the cha nnel range, the channel maintains full scale de flection and returns to normal functioning when th e transient has subsided.

Each radiation monitor, except for the circulatin g water, has four trip circuits: two upscale (high-high and high), one downscal e (low), and one inoperative.

Each trip is visually displayed on the affected radiati on monitor. Two of these trip s actuate corresponding control room annunciators: one upscal e (high radiation) and the downscale for the affected liquid monitoring channel. High or low sample flow measured at the sample panel actuates a control room high-low flow annunciator for the affected liquid channel.

All alarms are annunciated in a control room. Liquid monitoring system details are given in Table 11.5-2 and the monitor arrangements are shown in Figure 11.5-4.

11.5.2.2.2.1 Standby Service Water Radiation Monitoring System. See Section 11.5.2.1.4.

11.5.2.2.2.2 Reactor Building Closed C ooling Water Radiation Monitoring System. The radiation detector is located offline and samples the closed co oling water piping side of the reactor closed cooling water system (RCC) heat exchangers.

The monitor system is a diagnostic tool to veri fy that no inleakage of primary plant water has occurred from the reactor water cleanup system nonregenerative heat exchanger system which uses RCC as coolant. Since the RCC is a closed system, inleakage would be detected by the monitor system.

11.5.2.2.2.3 Radwaste Effluent Radiation Monitoring System. This monitor system measures the radioactivity in the radwaste effluent discharge prior to its entering the cooling tower blowdown line.

Liquid waste can be discharged fr om several radwaste processed water tanks such as the floor drain sample tank, waste sample tanks, or distillate drain tanks.

These tanks contain water that has been processed through one or more treatment systems such as evaporation, filtration, and ion exchange. Prior to the discharge from a ny tank, the liquid in the appropriate tank is sampled and analyzed in the labor atory for radioactivity. Based on this analysis, discharge is permitted as specified in the Offsite Dose Calc ulation Manual (ODCM).

The radiation detector is located offline and samples the common discharge line from the liquid radwaste system through which all liquid radwaste is discharged to the blowdown line. The

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 11.5-17 piping arrangement is designed so that the samp le well can be flushed to lower background levels.

The high-high upscale trip on the ra dwaste effluent radiation monitor is used to initiate closure of the radwaste system discharge valve. The trip point is set such that closure is initiated prior to exceeding ODCM limits for liquid effluents. The high upscale trip actuates an annunciator in the radwaste control room as well as the main control room.

11.5.2.2.2.4 Circulating Water and Plant Service Water Radiation Monitoring Systems. The circulating water monitor is located on the disc harge side of the circ ulating water pumps for the main condenser in the coolant blowdown line to the Columbia River. The location of this monitor permits detection of radioactive material leaking to the circulating water from any source, including the TSW system. If an alarm condition exists, circulating water blowdown to the Columbia River is terminated by automatic closure of the circulating water blowdown valve in the circulating water pump house (see Section 9.2.1.2), and it annunciates in the control room.

During plant outages, water is discharged to the river via temporary pumps installed in the circulating water pump house and connected to the blowdown line. Automatic termination of blowdown is not provided for this temporary arrangement. The only possible source of radiation would be through the TSW system which is monitored for radiati on. There is about a 2-hr holdup time in the circulati ng water pipe giving ample time for blowdown to be manually secured if an alarm is received.

The TSW radiation monitor is located in the TSW return head er to the circulating water system. The monitor is an offline type that continuously meas ures the radioactiv ity level of the TSW returning flow to the circulating water system. The radiation detector is lead shielded to minimize background radiation effe cts. The signal from the dete ctor is displayed on a readout module and recorded in the main control room.

11.5.2.2.3 Primary Containment Radiation Monitoring System

This monitoring system is composed of two parts, a sensitive two-ch annel leak detection system and a four-channel high activity LOCA tracking system.

11.5.2.2.3.1 Leak Detection Monitors. This monitoring system me asures the radioactivity in the drywell and in doing so complies with Regul atory Guide 1.45, Revision 0, and GDC 30.

The radioactivity in the drywell is from coolant and corrosi on activation products plus fission products produced in the reactor and released through leaks.

This monitoring system has two redundant subsystems, each having two de tectors, individually monitoring particulates, and nobl e gas activity. Additionally a charcoal sample cartridge is provided to trap haloge n gases. The detectors are housed in divisionally separated sample

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-06-048 11.5-18 racks located in a reactor building sample room. The sample r acks have incorporated blowers and flow controls to withdraw gas samples from the primary containment atmosphere via

stainless steel sampling lines and vent back to the containment.

The environment in which the local cabinets are located is main tained to limit upper temperature excursions that may occur in the reactor building during an accident. Associated radiation readout modules and recorders are mounted in the main control room along with alarm annunciators.

Required 120-V ac supply for Divi sion 1 and 2 equipment in both the main control room and the reactor building sample room is provided on a divisional basis by the 120/240-V ac critical (Class 1E) instrumentation power system.

The two-channel detector assemb lies are provided with lead shielding to minimize the effects of background radiation to ensure high sensitivity. The detectors are of the beta scintillation type and are provided with check sources to verify system operab ility. The particulate detector views a fixed filter collector on which airborne particulates are trapped. The noble gas detector views a fixed volume of gas.

Each radiation monitor has three trip circuits: one upscale Hi to alarm and close sample line valve, one upscale Alert, and one downscale for instrument inope rative. All alarms annunciate in the main control room. This monitoring subsystem provides no cont rol function and is a diagnostic tool which enables the main control room operator to take appropriate action.

Arrangement details are shown in Figure 11.5-9.

11.5.2.2.3.2 Loss-of-Coolant Ac cident Tracking Radiation Monitoring Systems (Containment Drywell). The LOCA monitoring systems, CMS-RE-27E and CMS-RE-27F, monitor the drywell atmosphere from inside the drywell; while CMS-RE-27A a nd CMS-RE-27B monitor the drywell atmosphere through penetrations in the bioshield wall. These four monitors provide LOCA monitoring for abnormal radioactivity levels following an accident condition involving rupture of the reactor coolant boundary.

The in-containment LOCA monitors CMS-RE-27E and CMS-RE-27F, located at approximately 517 ft level Azim uth 291° and 515 ft level Azim uth 51.5° respectively, track long-term decreases in cont ainment radioactivity that take place with decay and decontamination and comply with GDC 13 and 64. The in-containment LOCA monitors each contain a Victoreen Model 877-1 ionizati on chamber with a range of 1 to 10 7 R/hr. These two detectors are qualified and meet the criteria of Table II.F.1-3 of NUREG 0737, including calibration with a high level gamma source. The chambers w ill respond to low energy gamma radiation such as 81 keV from 133 Xe fulfilling the Regulatory Guide 1.97, Revision 2 criteria.

The LOCA monitors, CMS-RE-27A and CMS-R E-27B, are housed in the bioshield wall and are against the containment steel shell. These monitors will m onitor the drywell radiation in conjunction with the in-containment LOCA m onitors CMS-RE-27E and CMS-RE-27F. The

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-06-033 11.5-19 LOCA monitors, while not required for plant opera tion, are used to mon itor the drywell in the post-LOCA situation.

The LOCA monitoring systems (CMS-RE-27A, CMS-RE-27B, CMS-RE-27E, and CMS-RE-27F) are redundant and separately supplied with power from Division 1 and 2 in both the main control room and remote locations.

Each ionization chamber is wired to a local monitor. Output from each local monitor is wired to a monitor located in the main control room. Radia tion levels within the primary containment for each LOCA monitoring system are recorded in the main control room.

Each monitor has alarm circuitry and indica tion for high radiation and for instrument inoperative that annunciate in the main control room. This monitoring subsystem provides no control function and is a diagnostic tool which enables the main control room operator to take appropriate action.

Arrangement details are shown in Figure 11.5-10.

11.5.2.3 Sampling

The following sections present a detailed description of the radiological sampling procedures, frequencies and objectives for al l plant process and effluent sa mpling. This sample program provides the means to show co mpliance with the ODCM for th e process radiation monitoring system and radwaste system.

11.5.2.3.1 Process Sampling

Section 9.3.2 describes the design of sampling facili ties provided for general sampling. The sample frequency, type of anal yses, analytical sensitivity, and the purpose of the sample are summarized in Table 11.5-4 for each liquid process sample location and in Table 11.5-5 for each gas process sample location. The analytical procedures used in sample analysis are presented in Section 11.5.2.3.3. These samples serve to monitor radioactivity levels within various plant systems.

11.5.2.3.2 Effluent Sampling Effluent sampling of all potentially radioactive liquid and gaseous efflue nt paths is conducted on a regular basis to verify the adequacy of effl uent processing to meet the discharge limits to unrestricted areas. This effluent sampling program will be of su ch a comprehensive nature as to provide the information fo r the effluent measuring an d reporting programs required by 10 CFR Part 50.36a in annual re ports to the NRC. The fre quency of the periodic sampling and analysis described herein is normal and will be increased if effluent levels approach C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 11.5-20 Technical Specifications limits.

Tables 11.5-3 and 11.5-6 summarize the sample and analysis schedules which correspond to Regulatory Guide 1.21, Revision 1, guidance.

11.5.2.3.3 Analytical Procedures

Samples of process and effluent gases and liquids will be analyzed in the laboratory by the following techniques: gross beta counting, gross alpha counti ng, gamma spectrometry, liquid scintillation counting, and radiochemical separations.

Instrumentation which is available in the laboratory for the measurem ent of radioactivity includes a single channel analy zer, alpha counter, beta counter , liquid scintillation counter, and multichannel gamma spectrometer.

Samples for beta counting are evaporated to dryness on metal planchets prior to counting.

Sample volume, counting geomet ry, and counting time are chosen to achieve the required measurement sensitivities. Correction factor s are applied for sample-detector geometry, self-absorption and counter-resolving time.

Gross beta and gross alpha anal yses of liquid efflue nt samples may be performed with an internal proportional counter. The samples are prepared for counting by evaporation onto metal planchets. Gross beta counting is also performed usi ng a liquid scintillation counter.

Sample volume and counting times are chosen to achieve the required measurement sensitivity.

When possible, sample volume is selected to maintain a sample residue thickness of less than 1 mg/cm 2. Correction factors are a pplied for self-absorption.

Gamma ray spectrometry is used for isotopic an alyses of gaseous, air borne particulate and iodine, and liquid samples. Hi gh-resolution germanium solid-state detectors are available for this purpose. The detectors are calibrated against NIST traceable gamma ray standards for a variety of sample detector geometries. The gamma spectra ar e resolved using a software program developed fo r computerized spectrum resolution.

Gaseous tritium samples are collected by condensa tions or adsorption, a nd the resultant liquid is analyzed by liquid scin tillation counting techniques.

Radiochemical separatio ns are used for the routine analysis of 89 Sr and 90 Sr. 11.5.2.3.4 Inservice Inspection, Calibration, and Maintenance

During reactor operation, checks of system operability are made by observing channel behavior. At monthly intervals during react or operation, the response of each detector supplied with a remotely activated check source or LED source simulator will be recorded

together with the instrument background count rate to ensure pr oper functioning of the C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 11.5-21 monitor. Any detector whose response is observed to be inconsistent with that expected for power operation and is not supplied with a re motely activated check source or LED source simulator will be checked with a portable sour ce or by comparison of detector response to system activity. An ex ception to the portable source check requirement will be allowed for those detectors mounted in areas that are deemed hazardous to personnel.

The system has electronic testing and calibratin g equipment, which permits channel testing without relocating or dismantling ch annel components. An internal trip test circuit, adjustable over the full range of the readout me ter, is used for tes ting. Each channel is tested prior to performing a calibration check. Verification of valve ope ration, ventilation di version, or other trip functions is done at this time if it can be done without jeopardizing the plant safety. The tests are performed in conformance with the ODCM test frequencies.

The continuous radiation monitors are calib rated to commercial ra dionuclide standards traceable to the NIST.

Each continuous monitor is calibrated during the refueling outage or every 18 months as required by the ODCM, usi ng standard sources with NIST traceability.

A calibration can also be perf ormed by using liquid or gaseous radionuclid e standards or by comparison analysis of particul ate, iodine, liquid, or gaseous grab samples with laboratory instruments.

11.5.3 EFFLUENT MONITORING AND SAMPLING

The implementation of the requirements of GDC 64 concerning monitoring of gaseous effluent discharge paths for radioactiv ity is discussed in Section 11.5.2.2.1. Section 11.5.2.2.2 provides applicable discussions for liquid effluent radiation monitors.

11.5.4 PROCESS MONITORING AND SAMPLING

The implementation of the requirements of GDC 60 concerning automatic closure of isolation valves in gaseous and liqui d effluent discharge paths is discussed in Sections 11.5.2.1 and 11.5.2.2. Section 11.5.2.2.1 provides a discussion of gaseous process radiation monitors.

Section 11.5.2.2.2.2 provides a discussion of liquid process radiation monitors.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORTDecember 2005LDCN-04-043,05-000 11.5-23 Table 11.5-1 Process and Effluent Radiation Monitoring System (Gaseous and Airborne Monitors)

Monitor Detector Location (Number of Channels)

Type Efficiency/

Sensitivity

Range Principal Radionuclides Measured Expected Activity Upscale Setpoints Alarms Trips A. SAFETY-RELATED SYSTEMS Main steam line radiation monitors

MS-RE-3A, 3B, 3C, and 3D Adjacent to steam lines

(4) -ion chamber 3 x 10-10 amp/R/h 10 0 - 10 6 mR/h (6 dec. log)

Coolant activation

gases Steam line

activity defined

in Table 11.1-4 Above full

power background, below trip N/A Reactor building exhaust plenum radiation monitors REA-RE-9A, 9B, 9C, and 9D Inline (4) GM b 10 10 2 mR/hr (4 dec. log) Reactor bldg.

activity defined in Table 11.3-7 Above background, below trip Tech Specs Control room fresh air intakes WOA-RE-31A, 31B, 32A, and 32B Offline (4) -scint 2 x 10

-6 µCi/cm 3 10 1 x 10 7 cpm (6 dec. log) 133 Xe a Below monitor range Above background, below low trip Tech Specs B. SYSTEMS REQUIRED FOR PLANT OPERATION Offgas pretreatment radiation monitor OG-RE-2 Offline, adjacent to

sample chamber (1) -ion chamber 3 x 10-10 amp/R/h 10 0 - 10 6 mR/h (6 dec. log)

Noble gas fission products Offgas activity

defined in Table 11.3-1 Above background Tech Specs Offgas posttreatment radiation monitors

OG-RE-601A and OG-RE-601B Offline (2) -scint, Part. Filter, Iodine Filter 4.08 x 10 5 cpm/µCi/ mL 10 0 - 10 7 cpm (2 x 10-5 to 24 µCi/cc) (7 dec. log) 133 Xe a Offgas activity defined in Table 11.3-1 Above background ODCM LDCN-05-000 11.5-24 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORTDecember 2005 Table 11.5-1 Process and Effluent Radiation Monitoring System (Gaseous and Airborne Monitors) (Continued)

Monitor Detector Location (Number of Channels)

Type Efficiency/

Sensitivity

Range Principal Radionuclides Measured

Expected Activity Upscale Setpoints Alarms Trips Charcoal bed vault radiation monitor

OG-RE-11 Charcoal bed vault (1) GM N/A 10 0 - 10 6 mR/h (6 dec. log) Noble gas Charcoal bed inventory defined

in Section 11.3 Above background N/A Mechanical vacuum pump discharge

AR-RE-21 Inline (1) GM 10 10 6 mR/hr (4 dec. log) 133Xe Within monitor range Above background ODCM Reactor building elevated discharge radiation monitor PRM-RE-1A Inline (1) HP Ge 3.0 x 10 8 cps/µCi/cm 3 3.3 x 10-8 to 3.3 x 10-3 µCi/cm 3 133 Xe a RB activity Table 11.3-7 ODCM N/A PRM-RE-1B Inline (1) HP Ge 5.5 x 10 5 cps/µCi/cm 3 1.8 x 10-5 to 1.8 x 10 0 µCi/cm 3 133 Xe a LOCA Table 15.6-10 and 15.6-11 N/A N/A PRM-RE-1C Inline (1) HP Ge 1.28 x 10 1 cps/µCi/cm 3 7.8 x 10-1 to 7.8 x 10 4 µCi/cm 3 133 Xe a LOCA Table 15.6-10 and 15.6-11 N/A N/A Particulate filter Offline Filter N/A from 10-12 µCi/cm 3 emitters c ODCM Table 6.2.2.1.2-1 N/A N/A Iodine filter Offline Charc.

cart. N/A from 10-12 µCi/cm 3 emitters ODCM Table 6.2.2.1.2-1 N/A N/A LDCN-11-000 11.5-25 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011 Table 11.5-1 Process and Effluent Radiation Monitoring System (Gaseous and Airborne Monitors) (Continued)

Monitor Detector Location (Number of Channels)

Type Efficiency/

Sensitivity

Range Principal Radionuclides Measured

Expected Activity Upscale Setpoints Alarms Trips Turbine bldg. vent. exhaust radiation TEA-RE-13 TEA-RE-13A Offline (2) -scint part. filter c iodine filter 50 cpm/ pCi/cm 3 10 1 -10 7 cpm (6 dec. log) 133 Xe a Turbine bldg.

activity defined in Table 11.3-7 ODCM N/A -scint 10 10-3 Ci/cm 3 133 Xe (5 dec. log) 133 Xe a LOCA mixture of

F.P. activity see

Table 15.6-9 Radwaste bldg. vent. exhaust radiation WEA-RE-14 WEA-RE-14A Offline (2) -scint part. filter c iodine filter 50 cpm/ pCi/cm 3 10 1 -10 7 cpm (6 dec. log) 133 Xe a Radwaste bldg.

activity defined in

Table 11.3-7 ODCM N/A -scint 10-2 -10 3 Ci/cm 3 133 Xe (5 dec. log) 133 Xe a LOCA mixture of F.P. activity see

Table 15.6-9 Primary containment LOCA monitors Adjacent to

containment

steel walls -ion chambers amp/R/h R/h (6 dec. log) Fission product gases Within monitor

range Above background N/A CMS-RE-27A, 27B Outside(2) 1 x 10

-10 amp/R/h 10-2 x 10 4R/h (6 dec. log) b CMS-RE-27E, 27F Inside (2) 7 x 10-11 amp/R/h 10 0 - 10 7R/h (6 dec. log)

LDCN-06-048 11.5-26 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORTDecember 2009 Table 11.5-1 Process And Effluent Radiation Monitoring System (Gaseous And Airborne Monitors) (Continued)

Monitor Location (Number of Channels)

Type Efficiency/

Sensitivity

Range Principal Radionuclides Measured Expected Activity Upscale Setpoints Alarms Trips Primary containment radiation monitor Offline Containment discussed in

Section 12.2 Particulate

CMS-RE-12-1A

CMS-RE-12-1B (2) Part. filter

-scint N/A 10 10-6 µCi/cc (6 dec. log)

Fission gas

daughter and corrosion activation product d Above background, below trip Full scale Gas CMS-RE-12-3A

CMS-RE-12-3B (2) -scint N/A 10 10-1 µCi/cc (6 dec. log)

Above background, below trip Full scale a Sensitivity based on this radionuclide.

b Not required for plant operation.

c Composite of particulate filters analyzed for 89 Sr and 90 Sr. d For particulate, Cs-137 is the fission product radionuclide used for converting counts per minute to µCi/cc. For gas, Kr-85 is the fission product radionuclide used for converting counts per minute to µCi/cc. The monitors display units of µCi/cc.

LDCN-06-015,06-041 11.5-27 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORTDecember 2007 Table 11.5-2 Process and Effluent Radiation Monitoring System (Liquid Monitors)

Upscale Setpoints Monitor Detector Location (Number of Channels)

Type Range Principal Radionuclides Measured Expected Activity

Alarms Trips Residual heat removal

standby service water radiation monitor SW-RE-4, 5 Offline (2) -scint 10 10 6 cps (7 dec. log) 137 Cs a 60 Co Less than minimum detector sensitivity Above background Not applicable Reactor building closed cooling water radiation monitor RCC-RE-7 Offline (1) -scint 10 10 6 cps (7 dec. log) 137 Cs a 60 Co Less than minimum detector sensitivity Above background Not applicable Radwaste effluent radiation monitor FDR-RE-6 Offline (1) -scint 10 10 6 cps (7 dec. log) 137 Cs a 60 Co Section 11.2 Above background ODCM b Circulating water effluent radiation monitor

CBD-RE-8 Inline (1) -scint 10-1 -10 6 cps (7 dec. log) 137 Cs a 60 Co Less than minimum detector sensitivity Above background Section 11.5.2.2.2.4 Plant service water radiation monitor TSW-RE-5 Offline (1) -scint 1.0E-08 to 1.0E-02 µCi/cc (6 dec. log) 137 Cs a 60 Co Less than minimum detector sensitivity Above background Not applicable a Sensitivity based on this radionuclide.

b The alarm point will be set, based upon the activity, radionuclides, and dilution factor so that the concentration in the discharge line is less than 10 CFR 20 Appendix B, Table II, Column 2 limits.

11.5-28 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 Table 11.5-3 Radiological Analysis Summary of Gaseous Effluent Samples Sensitivity Grab Sample Sample Description Frequency Analysis µCi/ml Purpose Reactor building elevated release exhaust Weekly Gamma spectrum a Radioiodines b 10-11 10-10 Effluent record Quarterly 89 Sr and 90 Sr c 10-11 Monthly Gross alpha Tritium 10-11 10-11 Radwaste building exhaust As above As above Effluent record Turbine building exhaust As above As above Effluent record a On particulate filter b On charcoal cartridge c On composite of particulate filters

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000 LDC N-9 8-1 3 9 11.5-29 Table 11.5-4 Radiological Analysis Summary of Liquid Process Samples Sensitivity Grab Sample Sample D e scri ption Frequency Analysis Ci/ml Purpose Reactor coolant In accordance with Techn i cal Specif i cati o ns 131 I, 13 3 I 1 0-6 Evaluate fuel cladding integrity Gamma spectrum 1 0-6 Determine radionuclides present in s y stem Reactor water cleanup system Pe riodically Gamma spectrum 1 0-6 Evaluate cleanup efficiency Condensate storage tanks W eekly Gamma spectrum 1 0-6 Tank inventory Fuel pool filter - demineralizer inlet and outlet Periodically Gamma spectrum 10

-6 Evaluate system performance Waste collector tank Batch a Gamma spectrum 10

-6 Evaluate system performance Floor drain collector tank Batch a Gamma spectrum 10

-6 Evaluate system performance Chemical waste tank Batch a Gamma spectrum 10

-6 Evaluate system performance Detergent drain tank (2) Batch a Gamma spectrum 10

-6 Tank inventory a Analysis performed on an infr equent basis as needed to evaluate equipment performance or unusual water chemistry.

11.5-30 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 Table 11.5-5 Radiological Analysis Summary of G a seous Process Samples Sensitivity Grab Sample Sample Description Frequency Analysis Ci/ml Purpose Containment atmosphere (drywell) Periodically Gamma spectrum a 10-11 Evaluate prior to discharge Prior to entry Gamma spectrum 10

-10 Determine need for respiratory equipment Offgas pretreatment monitor sample Periodically Gamma spectrum 10

-6 Determine offgas activity Offgas posttreatment sample Periodically Gamma spectrum a,b Gamma spectrum c 10-10 10-6 Determine offgas system cleanup performance a On particulate filter b On charcoal cartridge c Noble gas

11.5-31 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 Table 11.5-6 Radiological Analysis Summary of Liquid Effluent Samples Sensitivity Grab Sample Sample Description Frequency Analysis Ci/ml Purpose Floor drain sample tank Batch a Gamma spectrum 10

-6 Effluent discharge record Waste sample tanks (2) Batch a Gamma spectrum 10

-6 Effluent discharge record Liquid radwaste effluent (composite of all tanks

discharged)

Monthly b Quarterly b Gross alpha

Tritium 89 Sr/90 Sr 10-7 10-5 10-8 Effluent discharge record Circulating water discharge

line Weekly grab of continuously

collected proportional

sample Gross or Gamma Spectrum

Tritium 10-6 10-5 Effluent discharge record (backup sample) a If tank is to be discharged, analyses w ill be performed on each batch. If tank is not to be discharged, analyses will be performed periodically to evaluate equipment performance.

b If liquids were discharged during the previous month or quarter, as applicable.

Chamber Main Steam Line Monitors 900547.3211.5-1 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Monitor Recorder Power- Supply Monitor Recorder Power- Supply MG-1 MC-7AReactor Water Sample Valve Annunciator on Panel H13-P602 Ion Chamber"A" Ion Chamber"C"Reactor Water Sample Valve Annunciator on Panel H13-P602 Monitor Recorder Power- Supply Monitor Recorder Power- Supply MG-2 MC-8AReactor Water Sample Valve Annunciator on Panel H13-P602 Ion Chamber"B" Ion"D"Mech. Vac. Pump and Valve InterlockReactor Water Sample Valve Annunciator on Panel H13-P602 Backup Power Supply PP-7AMech. Vac. Pump and Valve InterlockMech. Vac. Pump and Valve InterlockMech. Vac. Pump and Valve Interlock Columbia Generating StationFinal Safety Analysis Report Reactor Building Exhaust Plenum Monitors andCharcoal Bed Vault Monitor 900547.4811.5-2 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.A B C D Annunciators H13-P602 Recorder Annunciators H13-P602 Aux Unit Det CharcoalBed Vault PP-8AA PP-7AA Reactor Building Exhaust Plenum Monitor Monitor Monitor Monitor Power Supply Monitor Interlocks to standby gas treatment system "B" purge and vent valves (outbo), Reactor Bldg vent system (outbo) valves.

Interlocks to standby gas treatment system "A" purge and vent valves (Inbo), Reactor Bldg vent system (Inbo) valves.

Columbia Generating StationFinal Safety Analysis Report Control Room Fresh Air Intake Monitors 900547.49 11.5-3 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.Pre Amp Monitor Recorder Annunciator Pre Amp Monitor Annunciator Pre Amp Monitor Recorder Annunciator Pre Amp Monitor Annunciator"A""B""C""D" Line B Line A Div. I Div. II Trip Fresh Air Samples Detector Columbia Generating Station Final Safety Analysis Report Columbia Generating StationFinal Safety Analysis ReportDraw. No.Rev.Figure Amendment 59December 2007 Form No. 960690FH LDCN-06-015Process and Effluent Liquid Radiation Monitors 900547.5011.5-4 Local Monitor Monitor Annunciator on Panel H13-P602 Scintillation Detector DetectorWellPre Amp Monitor Annunciator on Panel H13-P602 Scintillation

Detector DetectorWell Instrument Bus "A" Recorder on Panel H13-P604Pre Amp Monitor Annunciator on Panel H13-P602 Scintillation Detector Detector Well Instrument Bus "B" Recorder on Panel H13-P600Pre Amp Monitor Annunciator

on Panel H13-P602 Scintillation Detector DetectorWell Instrument Bus "B" Recorder on Panel H13-P600Pre Amp Monitor Annunciator on Panels H13-P602& G11-P001 DetectorWell Instrument Bus "A"Pre Amp Monitor Annunciator on Panel H13-P602 Scintillation Detector DetectorWell Instrument Bus "B" Recorder on Panel H13-P600Effluent Valve Interlock Scintillation

Detector Reactor BuildingClosed Cooling Water Monitor Radwaste Effluent MonitorCirculating Water (Blowdown) MonitorStandby Service Water (RHR Loop "A")

MonitorPlant Service Water MonitorStandby Service Water (RHR Loop "B")

Monitor Recorder on Panel BD-RAD-24 900547.51 Columbia Generating StationFinal Safety Analysis ReportOffgas Pretreatment and Posttreatment Radiation MonitorsDraw. No.Rev.Figure Amendment 58December 200511.5-5 Form No. 960690FH LDCN-04-043To SJAE Inlet After Filter Bypass Microprocessor Recorder Monitor Monitor Detector Filter Collector E-PP-8AA (MC-8A)Monitor Recorder From Recombiner From 2nd Stage SJAE Sample ChamberToVial Sampler Ion Chamber MC-8A Annunciator

Panel H13-P602 Annunciator Panel H13-P602Offgas Treatment System Computers E-PP-US (MC-7A)Sample Chamber Detector Sample Chamber Microprocessor Mechanical Vacuum Pumps Exhaust Monitor andReactor Building Elevated Release Stack Monitor 950021.3211.5-6 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Reactor Building Ventilation Release Duct MonitorMechanical Vacuum Pumps Exhaust MonitorTrip Mechanical PumpsTo Elevated

Release Stack Annunciator From Mechanical Vacuum Pumps14" Ar(2)-1 Aux.Unit Monitor Recorder Detector Control Room Annunciator Filter Collectors FE RB Elevated Release Stack dPT FE FIT Rec FI Isokinetic

Probe Columbia Generating StationFinal Safety Analysis Report Detector Detector Detector Detector Figure Form No. 960690 LDCN-11-000Draw. No.Rev.960690.01Turbine and Radwaste Building Ventilation Release Duct Monitors11.5-7 Filter Collector(Typ)Radwaste Building Ventilation Release Ducts Monitor Preamp Monitor Monitor Preamp Computer Annunciator On Panel H13-P602Turbine Building Exhaust DuctTurbine Building Ventilation Release Duct Monitor Isokinetic

Probe Filter Collector(Typ)Columbia Generating StationFinal Safety Analysis Report Amendment 61 December 2011 Recorder Preamp Monitor Monitor Preamp Computer Annunciator On Panel H13-P602 Recorder 970187.2011.5-8 Figure Amendment 57December 2003Draw. No.Rev.DELETED Form No. 960690 LDCN-02-005 Columbia Generating StationFinal Safety Analysis Report Primary Containment Leak Detection Monitor 970187.2111.5-9 Figure Amendment 60December 2009 Form No. 960690Draw. No.Rev.Drywell Recorder Charcoal Cartridge Particulate Detector Gaseous Detector Annunciator Columbia Generating StationFinal Safety Analysis Report Local Monitor Local Monitor Remote Monitor Remote Monitor LDCN-06-048 Amendment 60 December 200 9 Primary Containment and Elevated Release StackLOCA Tracking 950021.3311.5-10 Figure Form No. 960690Draw. No.Rev.Columbia Generating StationFinal Safety Analysis Report Annunciator Computer Annunciator Primary Containment Steel Wall CMS-RE-27E Inside El. 524'CMS-RE-27F CMS-RE-27A Outside CMS-RE-27BBio-shield WallTo AnnunciatorTo Computer RecorderTo Computer Recorder Local Monitor Ion Chambers ComputerLog Amp HV Power Supply IndicatorTrip Unit RecorderLog Amp HV Power Supply IndicatorTrip Unit Recorder To Annunciator Local Monitor Monitor Monitor LDCN-06-033 Columbia Generating StationFinal Safety Analysis ReportDraw. No.Rev.Figure Amendment 58December 2005 Form No. 960690FH LDCN-05-000 Elevated Release Duct RB 618' 7" RB 611' 10" Flow PRM-COMP-1RW 525'Control Room T Hi Rad PRM-RR-3 PRM-CPL-1 PRM-RE-1A Cryostat Detector T T PRM-RE-1C Cryostat Detector PRM-RE-1B Cryostat Detector RB 618' 7" Intermediate Range Signal Processing High Range Signal Processing Low Range Signal Processing PRM-CP-1 RB 606'PRM-C-1B Compressor PRM-C-1C Compressor PRM-C-1A Compressor PRM-CAB-1Elevated Release Stack LOCA Monitoring 950021.2911.5-11 PRM-COMP-3 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-056 11.6-1 11.6 POSTACCIDENT SAMPLING SYSTEM

11.6.1 DESIGN BASIS Columbia Generating Station is using a General Electric postacc ident sampling system (PASS) capable of sampling the primary containment and reactor buildi ng atmosphere a nd of obtaining liquid samples from the reactor, RHR loops and va rious reactor building sumps. This system is designed to obtain grab samp les which may be analyzed ons ite or transported to offsite facilities for more detailed analysis if necessary. The sample station is located in the radwaste building and is shielded to reduce radiation exposure ra tes to the operator. All remote-operated valves are controlled from this area. Lead pigs are provided for radiation protection when transporting samples eith er to onsite facil ities or offsite.

All valves used are fully qualified for the envi ronment in which they are located inside and outside reactor containment. Power for the po staccident sampling equi pment is supplied from Division 1 and Division 2 critical power sources and will be available during accident conditions.

License Amendment No. 184 removed the re quirement for PASS from the Technical Specifications effective January 27, 2003. While this ame ndment removes the Technical Specification requirements for PASS, it requires a commitment for contingency plans for obtaining and analyzing highly radioactive samples from the Reactor Coolant System, suppression pool, and containment atmosphere. Until further changes are made to the PASS hardware it will be operated as described in this FSAR Secti on. This operation meets the commitment for contingency plans for sampling.

11.6.2 SYSTEM DESCRIPTION

Gas samples will be obtained from locations in the drywell, th e suppression pool atmosphere, and from the secondary containmen t atmosphere. The sample syst em is designed to operate at pressures ranging from subatmos pheric to maximum design pr essures of the primary and secondary containment. Heat-tra ced sample lines are used outsi de the primary containment to prevent precipitation of moisture and resultant loss of particulat es and iodines in the sample lines. The gas samples may be passed through a particulate filter and silver zeo lite cartridge for determination of particulate activity and i odine activity by subsequent analysis of the samples on a gamma spectrometer system. Al ternatively, the samp le flow bypasses the particulate/iodine sampler, is chilled to remove moisture, and a 15-ml grab sample can be taken for determination of gaseous radioactivity a nd for gas composition by gas chromatography. This size sample vial has been adopted for all gas samples to be consistent with present offgas sample vial counting factors.

Reactor coolant samples will be obtained from tw o points in the jet pump pressure instrument system when the reactor is at pressure. The jet pump pressure system has been determined to C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 11.6-2 be an optimum sample point fo r accident conditions. The pressure taps are well protected from damage and debris. If the recirculation pumps are secured, the water level will be raised about 18 in. above normal. This provides natural circulation of th e bulk coolant past the taps. Also, the pressure taps are loca ted sufficiently low to permit sa mpling at a reactor water level even below the lower core support plate.

A single sample line is also connected to both loops in the RHR system. This provides a means of obtaining a reactor coolant sample when the reactor is depressurized and at least one of the RHR loops is operated in the shutdown cooling mode. Si milarly, a suppression pool liquid sample can be obtained from the RHR loop lined up in the suppression pool cooling mode. Samples from the five drain sumps in the reactor building are also available.

The sample system isolation valv es are controlled from the local control panel. The sample system is designed for a purge flow of 1 gpm, whic h is sufficient to main tain turbulent flow in the sample line. Purge flow is returned to the suppression pool. The high flush flow also serves to alleviate cross-contamination of th e samples when switching from one sample point to another.

All liquid samples are taken into septum bottles mounted on sa mpling needles. The sample station is basically a bypass loop on the sample purge line. In the normal lineup, the sample flows through a conductivity cell (readable range 0.1 to 1000

µS/cm) and then through a ball valve bored out to 0.10-ml volume.

Flow through the sample pane l is established, the valve is rotated 90°, and a syringe is us ed to flush the sample plus a measured volume of diluent (generally 10 ml) through the valv e and into the sample bottle.

This provides a dilution of 100:1 to the sample. Alternativ ely, the valve sampling sequence can be repeated 10 times to provide a 1-ml sample diluted 10:1. The sample is transported to the laboratory for further dilution and subsequent analysis. Alternatively, the sample fl ow can be diverted through a 70-ml bomb to obtain a large pressurized volume.

This 70-ml volume can be circulated and depressurized into a known volume gas expansion chamber. The pressure change in this chamber will be used to calculate the total di ssolved gases in the reac tor coolant. A grab sample of these gases may be taken through a septum port for subsequent analysis.

Ten milliliter aliquots of this degassed liquid can also be take n for on or offsite chemical analyses requiring a relatively large sample. A radiation monitor in the liquid sample enclosure monitors liquid flow from the sample station to provide immediate assessment of the sample activity level. This m onitor also provides information as to the effectiveness of the demineralized water flus hing of the sample system following sample operation. The control instrumentation is installed in tw o 2 ft x 2 ft x 6 ft high standa rd cabinet control panels. One panel contains the conductivity a nd radiation level rea douts. Another cont rol panel contains the flow, pressure and temperat ure indicators, and the various control valves and switches.

A graphic display panel, installed directly below the main control panel, shows the status of the pumps and valves at all times.

The panel also indicates the relative position of the pressure C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-056 11.6-3 gauges and other items of concern to the operator. The us e of this panel will improve operator comprehension and assist in trouble-shooting operation.

Appropriate sample handling tools, a gas sampler vial positioner and gas vial cask are available to the operator at the sampling station. The gas vi al is installed and remo ved by use of the vial positioner through the front of the gas sampler. The vial is then ma nually placed down in the cask with the positioner which allows the vial to be maintained about 3 ft from the individual performing the operation.

The small-volume (10 ml) liquid sample is remotely obtained through the bottom of the sample station by use of the small-volume cask and cask positioner. The cask positioner holds the cask and positions the cask directly under the li quid sampler. The samp le vial is manually raised within the cask to engage the hypodermic needles. When the sample vial has been filled, the bottle is manually w ithdrawn into the cask. The sample vial is always contained within lead shielding during this operation. The cask is then lowered and sealed prior to transport to the laboratory.

A large-volume cask and cask positioner is available for transporting large liquid samples.

A 21-ml bottle is contained within a lead shielded cask. This sample bottle is raised from its location in the cask to the samp le station needles for bottle f illing. The sample station will only deliver 10 ml to this sample bottle. When filled, the bottle is withdrawn into the cask. The sample bottle is always shie lded by 5 to 6 in. of lead wh en in position under the sample station and during the fill and withdraw cycles, thus reduci ng operator exposure.

The cask is transported to the required position under the sample station by a dolly cask positioner. When in position th is cask is hydraulically elevat ed approximately 1.5 in. by a small hand pump for contact with the sample station shield ing under the liquid sample enclosure floor. The sample bo ttle is raised, held, and lowere d by a simple push/pull cable.

The cask is sealed by a threaded top plug that inserts above the sample bottle. The weight of this large-volume cask is approximately 700 lb.

The particulate filters and i odine cartridges are removed vi a a drawer arrangement. The quantity of activity which is accumulated on the cartridges is controlled by a combination of flow orificing and time sequence control of the fl ow valve opening. In addition, the deposition of iodine is monitored during sampling using a radiation detect or installed adjacent to the cartridge. These samples will hence be limited to activity levels which will normally not require shielded sample carriers to tr ansport the samples to the laboratory.

Based on information develope d by General Electric, Ener gy Northwest has developed plant-specific procedures for the determination of the extent of co re damage under accident conditions using inplant radia tion and hydrogen monitors. This meets the commitment for ability to determine fuel damage, postaccident.