ML21307A187

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2 to Updated Safety Analysis Report, Chapter 11, Radioactive Waste Management
ML21307A187
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Issue date: 10/28/2021
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TABLE OF CONTENTS Section Title Page 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1-1 11.1 SOURCE TERMS 11.1-1 11.1.1 FISSION PRODUCTS 11.1-2 11.1.1.1 Noble Radiogas Fission Products 11.1-2 11.1.1.2 Radiohalogen Fission Products 11.1-7 11.1.1.3 Other Fission Products 11.1-9 11.1.1.4 Nomenclature 11.1-10 11.1.2 ACTIVATION PRODUCTS 11.1-10 11.1.2.1 Coolant Activation Products 11.1-10 11.1.2.2 Non-coolant Activation Products 11.1-11 11.1.2.3 Steam and Power Conversion System N-16 Inventory 11.1-11 11.1.3 TRITIUM 11.1-11 11.1.4 FUEL FISSION PRODUCTION INVENTORY AND FUEL EXPERIENCE 11.1-15 11.1.4.1 Fuel Fission Product Inventory 11.1-15 11.1.4.2 Fuel Experience 11.1-15 11.1.5 PROCESS LEAKAGE SOURCES 11.1-16 11.1.6 LIQUID RADWASTE SYSTEM 11.1-17 11.1.7 RADIOACTIVE SOURCES IN THE GAS TREATMENT SYSTEM 11.1-17 11.1.8 SOURCE TERMS FOR COMPONENT FAILURES 11.1-17 11.1.8.1 Offgas System Failure 11.1-17 11.1.8.2 Liquid Radwaste System 11.1-17 11.

1.9 REFERENCES

FOR SECTION 11.1 11.1-18 11.2 LIQUID WASTE MANAGEMENT SYSTEMS 11.2-1 11.2.1 DESIGN BASES 11.2-1 11.2.1.1 Power Generation Design Objectives 11.2-1 11.2.1.2 Radiological Design Objectives 11.2-1 Revision 12 11-i January, 2003

TABLE OF CONTENTS (Continued)

Section Title Page 11.2.1.3 Design Criteria 11.2-1 11.2.1.4 Cost-Benefit Analysis 11.2-2 11.2.1.5 Accident Analysis 11.2-2 11.2.1.6 Component Design Parameters 11.2-3 11.2.1.7 Surge Input Collection Capabilities 11.2-3 11.2.1.8 Control of Tank Leakage and Overflows 11.2-4 11.2.1.9 ALARA Design Features 11.2-4 11.2.1.10 Control of Inadvertent Releases 11.2-6 11.2.2 SYSTEM DESCRIPTION 11.2-7 11.2.2.1 Input Streams 11.2-7 11.2.2.2 Separation of Inputs 11.2-7 11.2.2.3 Previous Experience 11.2-8 11.2.2.4 Treatment of High Purity/Low Conductivity Wastes 11.2-8 11.2.2.5 Treatment of Medium-to-Low Purity/Medium Conductivity Wastes 11.2-9 11.2.2.6 (Deleted) 11.2-10 11.2.2.7 Treatment of Detergent Drains 11.2-11 11.2.2.8 Treatment of Spent Resins 11.2-11 11.2.2.9 Treatment of Filter/Demineralizer Backwash 11.2-12 11.2.2.10 Detailed Component Design 11.2-12 11.2.2.11 Field Routed Pipe 11.2-17 11.2.2.12 System Control and Operating Procedures 11.2-18 11.2.2.13 Selection of Normal and Alternate Flow Paths 11.2-20 11.2.2.14 Performance Tests 11.2-20 11.2.3 RADIOACTIVE RELEASES 11.2-21 11.2.3.1 Description 11.2-21 11.2.3.2 Dilution Factors 11.2-21 11.2.3.3 Release Points 11.2-21 11.2.3.4 Estimated Releases 11.2-21a 11.

2.4 REFERENCES

FOR SECTION 11.2 11.2-22 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 11.3-1 11.3.1 DESIGN BASES 11.3-1 11.3.1.1 Design Objective 11.3-1 11.3.1.2 Design Criteria 11.3-1 11.3.1.3 Equipment Design Criteria 11.3-3 Revision 17 11-ii October, 2011

TABLE OF CONTENTS (Continued)

Section Title Page 11.3.2 SYSTEM DESCRIPTION 11.3-3 11.3.2.1 Main Condenser Steam Jet Air Ejector Low-Temp RECHAR System 11.3-4 11.3.2.2 System Design Description 11.3-14 11.3.2.3 Operating Procedure 11.3-20 11.3.2.4 Performance Tests 11.3-21 11.3.3 RADIOACTIVE RELEASES 11.3-23 11.3.3.1 Release Points 11.3-23 11.3.3.2 Dilution Factors 11.3-24 11.3.3.3 Estimated Releases and Dose Rates 11.3-24 11.

3.4 REFERENCES

FOR SECTION 11.3 11.3-25 11.4 SOLID RADIOACTIVE WASTE MANAGEMENT SYSTEM 11.4-1 11.4.1 DESIGN BASES 11.4-1 11.4.1.1 Power Generation Design Objectives 11.4-1 11.4.1.2 Radiological Design Objectives 11.4-2 11.4.1.3 Design Criteria 11.4-2 11.4.1.4 Component Design Parameters 11.4-3 11.4.1.5 ALARA Design Features 11.4-3 11.4.1.6 Safety Precautions 11.4-4 11.4.2 SYSTEM DESCRIPTION 11.4-4 11.4.2.1 Treatment of Wet Solid Radioactive Waste 11.4-4 11.4.2.2 Treatment of Dry Solid Radioactive Waste 11.4-6 11.4.2.3 Detailed Component Design 11.4-8 11.4.2.4 Instrumentation, Controls, Alarms, and Protective Devices 11.4-14 11.

4.3 REFERENCES

FOR SECTION 11.4 11.4-16 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5-1 11.5.1 DESIGN BASES 11.5-1 11.5.1.1 Design Objectives 11.5-1 11.5.1.2 Design Criteria 11.5-3 Revision 17 11-iii October, 2011

TABLE OF CONTENTS (Continued)

Section Title Page 11.5.2 SYSTEM DESCRIPTION 11.5-6 11.5.2.1 Systems Required for Safety 11.5-6 11.5.2.2 Systems Required for Plant Operation 11.5-8 11.5.2.3 Inspection, Calibration and Maintenance 11.5-18 11.5.3 EFFLUENT MONITORING AND SAMPLING 11.5-21 11.5.3.1 Implementation of General Design Criterion 64 11.5-21 11.5.4 PROCESS MONITORING AND SAMPLING 11.5-22 11.5.4.1 Implementation of General Design Criterion 60 11.5-22 11.5.4.2 Implementation of General Design Criterion 64 11.5-23 Revision 12 11-iv January, 2003

LIST OF TABLES Table Title Page 11.1-1 Noble Radiogas Source Terms 11.1-19 11.1-2 Power Isolation Event - Anticipated Occurrence 11.1-21 11.1-3 Halogen Radioisotopes In Reactor Water 11.1-22 11.1-4 Other Fission Product Radioisotopes In Reactor Water 11.1-23 11.1-5 Coolant Activation Products In Reactor Water and Steam 11.1-25 11.1-6 Non-coolant Activation Products In Reactor Water 11.1-26 11.2-1 Single Equipment Item Malfunction Evaluation 11.2-26 11.2-2 Design Data For Liquid Radwaste System Sumps 11.2-28 11.2-3 Design Data For Liquid Radwaste System Sump Pumps 11.2-30 11.2-4 Design Data For Liquid Radwaste System Tanks 11.2-32 11.2-5 Design Data For Liquid Radwaste System Pumps 11.2-34 11.2-6 Design Data For Liquid Radwaste System Piping 11.2-37 11.2-7 Design Data For Liquid Radwaste System Process Equipment 11.2-38 11.2-8 Summary of Liquid Radwaste System Equipment Redundancy 11.2-39 11.2-9 Evaluation of Liquid Radwaste System Capacity for Handling Large Waste Input Volumes 11.2-42 11.2-10 Process Flow Data For Liquid Radwaste System 11.2-45 11.2-11 Quality Requirements For Condensate Makeup 11.2-50 11.2-12 Criteria For Selection of Process Flow Path For Liquid Radwaste System Inputs 11.2-51 11.2-13 Significant Nuclide Annual Release To Discharge Tunnel 11.2-52 Revision 12 11-v January, 2003

LIST OF TABLES (Continued)

Table Title Page 11.2-14 Annual Release By Stream To Discharge Tunnel 11.2-55 11.2-15 Input Parameters For Calculating Liquid Releases (GALE) 11.2-58 11.2-16 Tanks Located Outside the Containment Which Contain Potentially Radioactive Fluid 11.2-61 11.3-1a Estimated Air Ejector Offgas Release Rates 11.3-26 11.3-1b Estimated Air Ejector Offgas Release Rates 11.3-28 11.3-1c Estimated Air Ejector Offgas Release Rates 11.3-30 11.3-1d Estimated Air Ejector Offgas Release Rates 11.3-32 11.3-2 Offgas System Major Equipment Items 11.3-34 11.3-3 (Deleted) 11.3-36 11.3-4 Equipment Malfunction Analysis 11.3-37 11.3-5 Frequency and Quantity of Steam Discharged to Suppression Pool 11.3-44 11.3-6 Gaseous Radwaste Equipment Design Requirements 11.3-46 11.3-7 Offgas System Alarmed Process Parameters 11.3-47 11.3-8a Input Parameters Used For Calculating Gaseous Releases 11.3-48 11.3-8b Input Parameters Used For Calculating Gaseous Releases 11.3-50 11.3-8c Input Parameters Used For Calculating Gaseous Releases 11.3-52 11.3-8d Input Parameters Used For Calculating Gaseous Releases 11.3-54 Revision 19 11-vi October, 2015

LIST OF TABLES (Continued)

Table Title Page 11.3-9a Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 1 11.3-56 11.3-9b Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 1 11.3-57 11.3-9c Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 1 11.3-58 11.3-9d Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 1 11.3-59 11.3-10a Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 2 11.3-60 11.3-10b Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 2 11.3-61 11.3-10c Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 2 11.3-62 11.3-10d Calculated Release of Radioactive Materials in Gaseous Effluents - Unit 2 11.3-63 11.3-11a Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary 11.3-64 11.3-11b Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary 11.3-65 11.3-11c Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary 11.3-66 11.3-11d Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary 11.3-67 11.4-1 Maximum Monthly Radioactive Waste Inputs to Solid Radioactive Waste System 11.4-17 11.4-2 Solid Radwaste System Influent Nuclide Activities 11.4-19 11.4-3 (Deleted) 11.4-20 11.4-4 Solid Radwaste System Demineralizer Activities 11.4-21 Revision 17 11-vii October, 2011

LIST OF TABLES (Continued)

Table Title Page 11.5-1 Gaseous and Airborne Process and Effluent Radiation Monitor 11.5-24 11.5-2 Process and Effluent Radiation Monitoring System Characteristics 11.5-26 11.5-3 Liquid Process and Effluent Radiation Monitors 11.5-28 11.5-4 Radiological Analysis Summary of Liquid Process Samples 11.5-30 11.5-5 Radiological Analysis Summary of Gaseous Process Samples 11.5-33 11.5-6 Radiological Analysis Summary of Liquid Effluent Samples 11.5-34 11.5-7 Radiological Analysis Summary of Gaseous Effluent Samples 11.5-37 Revision 12 11-viii January, 2003

11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS General Electric has evaluated radioactive material sources (activation and fission product releases from fuel) 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 operating BWRs has resulted in doses to offsite persons which have been only a small fraction of <10 CFR 20>, or of natural background dose. (Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the <10 CFR 20>

regulations prior to October 4, 1993. Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised <10 CFR 20> dated October 4, 1993.)

The information provided in this section defines the design basis radioactive material levels in the reactor water, steam and offgas. The various radioisotopes listed have been grouped as coolant activation products, non-coolant activation products and fission products. The fission product levels are based on measurements of BWR reactor water and offgas at several stations through mid-1971. Emphasis was placed on observations made at KRB (in the Republic of Germany) and Dresden 2.

The design basis radioactive material levels do not necessarily include all the radioisotopes observed or predicted theoretically to be present.

The radioisotopes included are considered significant to one or more of the following criteria:

a. Plant equipment design.
b. Shielding design.

Revision 12 11.1-1 January, 2003

c. Understanding system operation and performance.
d. Measurement practicability.
e. Evaluation of radioactive material releases to the environment.

For halogens, radioisotopes with half-lives of less than three minutes were omitted. For other fission product radioisotopes in reactor water, radioisotopes with half-lives of less than 10 minutes were not considered.

11.1.1 FISSION PRODUCTS 11.1.1.1 Noble Radiogas Fission Products The noble radiogas fission product source terms observed in operating BWRs are generally complex mixtures. Their sources vary from miniscule defects in cladding to tramp uranium on external cladding surfaces.

The relative concentrations or amounts of noble radiogas isotopes can be described as follows:

Equilibrium: Rg ~ k1y (11.1-1)

Recoil: Rg ~ k2y (11.1-2)

The nomenclature in <Section 11.1.1.4> defines the terms in these and succeeding equations. The constants k1 and k2 describe the fractions of the total fissions that are involved in each of the releases. The equilibrium and recoil mixtures are the two extremes of the mixture spectrum that are physically possible. When a sufficient time delay occurs between the fission event and the time of release of the radiogases from the fuel to the coolant, the radiogases approach equilibrium levels in the fuel and the equilibrium mixture results.

Revision 12 11.1-2 January, 2003

When there is no delay or impedance between the fission event and the release of the radiogases, the recoil mixture is observed.

Prior to Vallecitos Boiling Water Reactor (VBWR) and Dresden 1 experience, it was assumed that noble radiogas leakage from the fuel would be the equilibrium mixture of the noble radiogases present in the fuel.

VBWR and early Dresden 1 experience indicated that the actual mixture most often observed approached a distribution which was intermediate in character to the two extremes (Reference 1). This intermediate decay mixture was termed the diffusion mixture. It must be emphasized that this diffusion mixture is merely one possible point on the mixture spectrum ranging from the equilibrium to the recoil mixture and does not have the absolute mathematical and mechanistic basis for the calculational methods possible for equilibrium and recoil mixtures.

However, the diffusion distribution pattern which has been described is as follows:

Diffusion: Rg ~ k3y0.5 (11.1-3)

The constant, k3, describes the fraction of total fissions that are involved in the release. The value of the exponent of the decay constant, , is midway between the values for equilibrium, 0, and recoil, 1. The diffusion pattern value of 0.5 was originally derived from diffusion theory.

Although the previously described diffusion mixture has been used by GE as a basis for design since 1963, the design basis release magnitude used has varied from 0.5 Ci/sec to 0.1 Ci/sec as measured after 30 minute decay (t = 30 min). 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 conveniently measurable and was Revision 12 11.1-3 January, 2003

consistent with the nominal design basis 30 minute offgas holdup system used on a number of plants. Since approximately 1967, the design basis release magnitude used (including the 1971 source terms) has been established at an annual average of 0.1 Ci/sec (t = 30 min). This design basis is considered as an annual average with some time above and some time below this value. This design value was selected on the basis of operating experience rather than predictive assumptions. Several judgment factors, including the significance of environmental release, reactor water radioisotope concentrations, liquid waste handling and effluent disposal criteria, building air contamination, shielding design, and other component contamination affecting maintenance, have been considered in establishing this level.

Noble radiogas source terms from fuel above 0.1 Ci/sec (t = 30 min) can be tolerated for reasonable periods of time. Continual assessment of these values is made on the basis of actual operating experience in BWRs (Reference 2).

While the noble radiogas source term magnitude was established at 0.1 Ci/sec (t = 30 min), it was recognized 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 accurately characterize the noble radiogas mixture pattern for an operating BWR.

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

Rg Kgym (1 et) (et) (11.1-4)

With the exception of Kr-85 with a half-life of 10.74 years, the noble radiogas fission products in the fuel are essentially at an equilibrium condition after an irradiation period of several months (rate of Revision 12 11.1-4 January, 2003

formation is equal to the rate of decay). So for practical purposes the

-T term (1 - e ) approaches 1 and can be neglected when the reactor has been operating at a steady-state for long periods of time. The term

-t (e ) is used to adjust the releases from the fuel (t = 0) to the decay time for which values are needed. Historically, t equal to 30 minutes has been used. When discussing long steady-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:

Rg Kgym (11.1-5)

The constant, Kg, describes the magnitude of leakage. The relative rates of leakage of the different noble radiogas isotopes are accounted for by the variable, m, the exponent of the decay constant, .

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

log (Rg/y) m log log (Kg) (11.1-6)

Equation 11.1-6 represents a straight line when log Rg/y is plotted versus log (); m is the slope of the line. This straight line is obtained by plotting (Rg/y) versus () on logarithmic graph paper.

By fitting actual data from KRB and Dresden 2 (using least squares techniques) to the equation, the slope, m, can be obtained. This can be estimated on the plotted graph. With radiogas leakage at KRB over the nearly 5 year period varying from 0.001 to 0.056 Ci/sec (t = 30 min) and with radiogas leakage at Dresden 2 varying from 0.001 to 0.169 Ci/sec (t = 30 min), the average value of m was determined. The value 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 were observed in the range m equal to 0.1 to m equal to 0.6. After establishing the value of m equal to 0.4, the value Revision 12 11.1-5 January, 2003

of Kg can be calculated by selecting a value for Rg, or as has been done historically, the design basis is set by the total design basis source term magnitude at t equal to 30 minutes. With Rg at 30 minutes equal to 100,000 µCi/sec, Kg can be calculated as being 2.6 x 107 and Equation 11.1-4 becomes:

R g 2.6 x 107 y0.4 (1 e T ) (e t ) (11.1-7)

This updated noble radiogas source term mixture has been termed the 1971 Mixture to differentiate it from the diffusion mixture. The noble gas source term for each radioisotope can be calculated from Equation 11.1-7. The resultant source terms are presented in

as leakage from fuel (t = 0) and after 30 minute decay. 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 Kr-85 based on a few actual measurements. Out of the thirteen commonly considered noble gases, normal operational releases to the primary coolant are expected to be approximately 25,000 Ci/sec as evaluated at 30 minutes, and 100 Ci/sec of I-131. These values can be compared to the design base value of 100,000 Ci/sec for the summation of the same thirteen noble gases, and 700 Ci/sec for I-131.
presents the source terms released to the reactor pressure vessel as a consequence of a power isolation event, which is the only anticipated operational occurrence in which significant activity is expected to be released. Revision 12 11.1-6 January, 2003 11.1.1.2 Radiohalogen Fission Products Historically, the radiohalogen design basis source term was established by the same equation as that used for noble radiogases. In a similar fashion, a simplified equation can be shown to describe the release of each halogen radioisotope: R h Khyn (11.1-8) The constant, Kh, describes the magnitude of leakage from fuel. The relative rates of halogen radioisotope leakage are expressed in terms of n, the exponent of the decay constant, . As was done with the noble radiogases, the 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 observed in the range of n equal to 0.1 to n equal to 0.9. It appeared that the use of the previous method of calculating radiohalogen leakage from fuel was overly conservative. <Figure 11.1-3> relates KRB and Dresden 2 noble radiogas versus I-131 leakage. While it can be seen from Dresden 2 data during the period of August 1970 to January 1971 that there is a relationship between noble radiogas and I-131 leakage under one fuel condition, there was no simple relationship for all fuel conditions 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 Revision 12 11.1-7 January, 2003 0.1 Ci/sec (t = 30 min). This may be partially explained by inherent limitations due to internal 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 would be indicated by I-131 leakage from the fuel in excess of 700 Ci/sec. When high radiohalogen leakages are observed, other fission products will be present in greater amounts. Using these judgment factors and experience to date, the design basis radiohalogen source terms from fuel were established based on I-131 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 min). With the I-131 design basis source term established, Kh can be calculated as being 2.4 x 107 and halogen radioisotope release can be expressed by the following equation: R h 2.4 x 107 y0.5 (1 e T) (e t ) (11.1-9) The concentrations of the radiohalogens in the reactor water are calculated by modeling the plants piping network as line sections or streams which connect flow junctions. Each line section is described as to the mass flow rate, steam quality, gas and liquid specific volumes, mass inventory, type of unit operations occurring, and the nature of flow for radioactive decay calculations. With this information, along with the plant nuclear and chemical data, the radioactive material transport performance of each line can be expressed mathematically. The resulting system of linear equations is then solved to obtain the required concentrations. Although carryover of most soluble radioisotopes from reactor water to steam is observed to be less than 0.1 percent (<0.001 fraction), the Revision 12 11.1-8 January, 2003 observed carryover for radiohalogens has varied from 0.1 percent to about 2 percent on newer plants. The average of observed radiohalogen carryover measurements has been 1.2 percent by weight of reactor water in steam, with a standard deviation of 0.9. In the present source term definition, a radiohalogen carryover of 2 percent (0.02 fraction) was used. The halogen release rate from the fuel can be calculated from Equation 11.1-9. The resultant concentrations are presented in
. 11.1.1.3 Other Fission Products The observations of other fission products (and transuranic nuclides, including Np-239) 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 and are presented in
. Carryover of these radioisotopes from the reactor water to the steam is estimated to be less than 0.1 percent (<0.001 fraction). In addition to carryover, however, decay of noble radiogases in the steam leaving the reactor results in production of noble gas daughter radioisotopes in the steam and condensate systems. Some daughter radioisotopes (e.g., yttrium and lanthanum), were not listed as being in reactor water. Their independent leakage to the coolant is negligible; however, these radioisotopes may be observed in some samples in equilibrium or approaching equilibrium with the parent radioisotope. Except for Np-239, trace concentrations of transuranic isotopes have been observed in only a few samples where extensive and complex analyses were carried out. The predominant alpha emitter present in reactor Revision 12 11.1-9 January, 2003 water is Cm-242 at an estimated concentration of 10-6 Ci/g or less. The concentration of alpha emitting plutonium radioisotopes is more than one order of magnitude lower than that of Cm-242. Plutonium-241 (a beta emitter) may also be present in concentrations comparable to the Cm-242 level. 11.1.1.4 Nomenclature The following list of nomenclature defines the terms used in equations for source term calculations: Rg = Leakage rate of a noble gas radioisotope (Ci/sec). Rh = Leakage rate of a halogen radioisotope (Ci/sec). y = Fission yield of a radioisotope (atoms/fission). = Decay constant of a radioisotope (sec-1). T = Fuel irradiation time (sec). t = Decay time following leakage from fuel (sec). m = Noble radiogas decay constant exponent (dimensionless). n = Radiohalogen decay constant exponent (dimensionless). Kg = A constant establishing the level of noble radiogas leakage from fuel. Kh = A constant establishing the level of radiohalogen leakage from fuel. 11.1.2 ACTIVATION PRODUCTS 11.1.2.1 Coolant Activation Products The coolant activation products are not adequately correlated by simple equations. Design basis concentrations in reactor water and steam have been estimated conservatively from experience data. The resultant concentrations are presented in
. Revision 12 11.1-10 January, 2003 11.1.2.2 Non-coolant Activation Products The activation products formed by activation of impurities in the coolant or by corrosion of irradiated system materials are not adequately correlated by simple equations. The design basis source terms of non-coolant activation products have been estimated conservatively from experience data. The resultant concentrations are presented in
. Carry-over of these isotopes from the reactor water to the steam is estimated to be less than 0.1 percent (<0.001 fraction). The effect of operating above the design basis zinc source term due to zinc injection was evaluated. The evaluation determined the effects to be negligible; therefore, the references and tables concerning the zinc concentrations, calculated MPC levels, and doses were not updated as a result of implementing the zinc injection process. 11.1.2.3 Steam and Power Conversion System N-16 Inventory Steam and power conversion system N-16 inventories are given in <Section 12.2.1>. 11.1.3 TRITIUM In a BWR, tritium is produced by three principal methods:
a. Activation of naturally occurring deuterium in the primary coolant.
b. Nuclear fission of UO2 fuel.
c. Neutron reactions with boron used in reactivity control rods.
The tritium, formed in control rods, which may be released from a BWR in liquid or gaseous effluents, is believed to be negligible. A prime Revision 12 11.1-11 January, 2003 source of tritium available for release from a BWR is that produced from activation of deuterium in the primary coolant. Some fission product tritium may also transfer from fuel to primary coolant. This discussion 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 activation of deuterium in the primary coolant is available for release in liquid or gaseous effluents. The tritium formed in a BWR from deuterium activation can be calculated using the equation: V Ract (11.1-11) 3.7 104P Where: Ract = Tritium formation rate by deuterium activation (Ci/sec/MWt) = Macroscopic thermal neutron cross section (cm-1) = Thermal neutron flux (neutrons/(cm2) (sec)) V = Coolant volume in core (cm3) = Tritium radioactive decay constant (1.78 x 10-9 sec-1) P = Reactor power level (MWt) For recent BWR designs, Ract is calculated to be 1.30.4 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 density in the core, abundance of deuterium in light water (some additional deuterium is present due to the H(n,) D reaction), thermal neutron flux, and microscopic cross section for deuterium. The fraction of tritium produced by fission which may transfer from fuel to the coolant (which is then available for release in liquid and gaseous effluents) is more difficult to estimate. However, since Revision 12 11.1-12 January, 2003 zircaloy-clad fuel rods are used in BWRs, essentially all fission product tritium remains in the fuel rods unless defects are present in the cladding material (Reference 3). The study made at Dresden 1 in 1968 by the U.S. Public Health Service (USPHS) suggests that essentially all of the tritium released from the plant could be accounted for by the deuterium activation source (Reference 4). For purposes of estimating the leakage of tritium from defective fuel, it can be assumed that it leaks in a manner similar to the leakage of noble radiogases. Thus, use can be made of the empirical relationship described as the diffusion mixture, used for predicting the source term of individual noble gas radioisotopes as a function of the total noble gas source term. The equation which describes this relationship is: R dif K y 0.5 (11.1-12) Where: Rdif = Leakage rate of tritium from fuel (µCi/sec) y = Fission yield fraction (atoms/fission) = Radioactive decay constant (sec-1) K = A constant related to total tritium leakage rate If the total noble radiogas source term is 105 Ci/sec after 30 minute decay, leakage from fuel can be calculated to be about 0.24 Ci/sec of tritium. To place this value in perspective in the USPHS study, the observed rate of Kr-85 (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 average noble radiogas leakage from a BWR is expected to be less than Revision 12 11.1-13 January, 2003 0.1 Ci/sec (t = 30 min), the annual average tritium release rate from the fission source can be conservatively estimated at 0.120.12 Ci/sec, or 0.0 to 0.24 Ci/sec. Based on this approach, the estimated total tritium appearance rate in reactor coolant and release rate in the effluent is about 20 Ci/yr. Tritium formed in the reactor is generally present as tritiated oxide (HTO) and to a lesser degree as tritiated gas (HT). Tritium concentration (on a weight basis) in the steam formed in the reactor is the same as in the reactor water at any given time. This tritium concentration is also present in condensate and feedwater. Since radioactive effluents generally originate from the reactor and power cycle equipment, radioactive effluents also have this tritium concentration. The condensate storage tanks receive treated water from the liquid waste management system and reject water from the condensate system. Thus, all plant process water has a common tritium concentration. Offgases released from the plant contain tritium, which is present as tritiated gas (HT) resulting from reactor water radiolysis, as well as HTO. In addition, water vapor from the turbine gland seal steam packing exhauster and a lesser amount present in ventilation air due to process steam leaks or evaporation from sumps, tanks and spills on floors also contain tritium. The remainder of the tritium leaves the plant in liquid effluents or with solid wastes. Recombination of radiolytic gases in the air ejector offgas system forms water, which is condensed and returned to the main condenser. This tends to reduce the amount of tritium leaving in gaseous effluents. Revision 12 11.1-14 January, 2003 Reducing the gaseous tritium release results in a slightly higher tritium concentration in the plant process water. Reducing the amount of liquid effluent discharged also results in a higher process coolant equilibrium 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 the plant discharge or as solid waste. Reduction due to radioactive decay is negligible due to the 12 year half-life of tritium. The USPHS study at Dresden 1 estimated that approximately 90 percent of the tritium release was observed in liquid effluent, with the remaining 10 percent leaving as gaseous effluent (Reference 4). Efforts to reduce the volume of liquid effluent discharges may change this distribution so that a greater amount of tritium leaves as gaseous effluent. From a practical standpoint, the fraction of tritium leaving as liquid effluent may vary between 60 and 90 percent, with the remainder leaving in gaseous effluent. 11.1.4 FUEL FISSION PRODUCTION INVENTORY AND FUEL EXPERIENCE 11.1.4.1 Fuel Fission Product Inventory Fuel fission product inventory information is used in establishing 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 experience, including fuel failure experience, burnup experience and thermal conditions under which the experience was gained, is presented in (Reference 5), (Reference 6), (Reference 7), and (Reference 8). Revision 12 11.1-15 January, 2003 11.1.5 PROCESS LEAKAGE SOURCES Process leakage results in potential release paths for noble gases and other volatile fission products through ventilation systems. Liquid from process leaks is collected and routed to the liquid-solid radwaste system. Radionuclide releases through 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 relatively significant, GE has conducted measurements to identify and qualify these low level release paths. GE has maintained an awareness of other measurements by the Electric Power Research Institute and other organizations and routine measurements by utilities with operating BWRs. Leakage of fluids from the process system results in the release of radionuclides into plant buildings. In general, the noble radiogases remain airborne and are released to the atmosphere with little delay through the building ventilation exhaust ducts. The radionuclides partition between air and water, and airborne radioiodines may plate out on metal surfaces, concrete and paint. A significant amount of radioiodine remains in air or is desorbed from surfaces. Radioiodines are found in ventilation air as methyl iodide and as inorganic iodine which is here defined as particulate, elemental and hypoiodous acid forms of iodine. Particulates are also present in the ventilation exhaust air. The estimated release rate of radioactive materials in gaseous effluents is presented in <Section 11.3.3>. Revision 12 11.1-16 January, 2003 11.1.6 LIQUID RADWASTE SYSTEM Radioactive sources for the liquid radwaste system are described in <Section 11.2.3> and are based on information contained in <NUREG-0016> (Reference 9). 11.1.7 RADIOACTIVE SOURCES IN THE GAS TREATMENT SYSTEM Radioactive sources for the gas treatment system are described in <Section 11.3.2.1.2>. 11.1.8 SOURCE TERMS FOR COMPONENT FAILURES 11.1.8.1 Offgas System Failure Source terms for evaluation of the radiological consequences of component failures within the offgas system are contained in
for the Design Basis and Normal (realistic) operating conditions. 11.1.8.2 Liquid Radwaste System Radiation sources used for component failures are consistent with an offgas release rate of 100,000 Ci/sec after 30 minutes decay. This results in maximum inventories of radioisotopes in the system and is not anticipated to occur during operation of the plant. The isotopic breakdown of the inventory in each significant component of the liquid radwaste system is presented in
. Revision 12 11.1-17 January, 2003 11.

1.9 REFERENCES

FOR SECTION 11.1

1. Brutschy, F. J., A Comparison of Fission Product Release Studies in Loops and VBWR, Paper presented at the Tripartite Conference on Transport of Materials in Water Systems, Chalk River, Canada, February 1961.
2. Skarpelos, J. M. and R. S. Gilbert, Technical Derivation of BWR 1971 Design Basis Radioactive Material Source Terms, General Electric Company, NEDO-10871, March 1975.
3. Ray, J. W., Tritium in Power Reactors, Reactor and Fuel-Processing Technology, 12 (1), pp. 19-26, Winter 1968-1969.
4. Kahn, B., et al, Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor, BRH/DER 70-1, March 1970.
5. Williamson, H. E., Ditmore, D. C., Experience with BWR Fuel Through September 1971, General Electric Company, NEDO-10505, May 1972. (Update)
6. Elkins, R. B., Experience with BWR Fuel Through September 1974, General Electric Company, NEDO-20922, June 1975.
7. Williamson, H. E., Ditmore, D. C., Current State of Knowledge of High Performance BWR Zircaloy Clad UO2 Fuel, General Electric Company, NEDO-10173, May 1970.
8. Elkins, R. B., Experience with BWR Fuel Through December 1976, General Electric Company, NEDO-21660, July 1977.
9. U.S. Nuclear Regulatory Commission, Calculation of Releases of Radioactive Material in Gaseous and Liquid Effluents from Boiling Water Reactors (BWR-GALE Code), <NUREG-0016>, April 1976.

Revision 12 11.1-18 January, 2003

TABLE 11.1-1 NOBLE RADIOGAS SOURCE TERMS Source Term Source Term

= t 0 = t 30 Isotope Half-Life _(Ci/sec)_ _(Ci/sec)_

Kr-83m 1.86 Hr 3.4 +3 2.9 +3 85m 4.4 Hr 6.1 +3 5.6 +3 85 10.74 Yr 10 to 20(1) 10 to 20(1) 87 76 Min 2.0 +4 1.5 +4 88 2.79 Hr 2.0 +4 1.8 +4 89 3.18 Min 1.3 +5 1.8 +2 90 32.3 Sec 2.8 +5 -

91 8.6 Sec 3.3 +5 -

92 1.84 Sec 3.3 +5 -

93 1.29 Sec 9.3 +4 -

94 1.0 Sec 2.3 +4 -

95 0.5 Sec 2.1 +3 -

97 1.0 Sec 1.4 +1 -

Xe-131m 11.96 Day 1.5 +1 1.5 +1 133m 2.26 Day 2.9 +2 2.8 +2 133 5.27 Day 8.2 +3 8.2 +3 135 9.16 Hr 2.2 +4 2.2 +4 135m 15.7 Min 2.6 +4 6.9 +3 137 3.82 Min 1.5 +5 6.7 +2 138 14.2 Min 8.9 +4 2.1 +4 139 40 Sec 2.8 +5 -

Revision 12 11.1-19 January, 2003

TABLE 11.1-1 (Continued)

Source Term Source Term

= t 0 = t 30 Isotope Half-Life _(Ci/sec)_ _(Ci/sec)_

Xe-140 13.6 Sec 3.0 +5 -

141 1.72 Sec 2.4 +5 -

142 1.22 Sec 7.3 +4 -

143 0.96 Sec 1.2 +4 -

144 9.0 Sec 5.6 +2 -

Total Approx. 2.5 +6 Approx. 1.0 +5 NOTE:

(1)

Estimated from experimental observations.

Revision 12 11.1-20 January, 2003

TABLE 11.1-2 POWER ISOLATION EVENT - ANTICIPATED OCCURRENCE Isotopic Spiking Isotope Activity (Ci)/Bundle I-131 2.1 132 3.3 133 5.1 134 5.5 135 4.9 Kr-83m 0.9 85m 2.2 85 0.5 87 4.4 88 5.2 89 8.2 Xe-131m 0.1 133m 0.3 133 11.8 135m 1.8 135 11.2 137 10.7 138 10.8 Revision 12 11.1-21 January, 2003

TABLE 11.1-3 HALOGEN RADIOISOTOPES IN REACTOR WATER Concentration Isotope _Half-Life (Ci/g)

Br-83 2.40 hr 1.4 x 10-2 84 31.8 min 3.0 x 10-2 85 3.0 min 1.9 x 10-2 I-131 8.065 day 1.2 x 10-2 132 2.284 hr 1.2 x 10-1 133 20.8 hr 8.3 x 10-2 134 52.3 min 2.6 x 10-1 135 6.7 hr 1.2 x 10-1 Revision 12 11.1-22 January, 2003

TABLE 11.1-4 OTHER FISSION PRODUCT RADIOISOTOPES IN REACTOR WATER Concentration Isotope _Half-Life ___(Ci/g)___

Sr-89 50.8 day 2.82 x 10-3 Sr-90 28.9 yr 2.09 x 10-4 Sr-91 9.67 hr 6.90 x 10-2 Sr-92 2.69 hr 1.15 x 10-1 Zr-95 65.5 day 3.13 x 10-5 Zr-97 16.8 hr 3.03 x 10-5 Nb-95 35.1 day 3.76 x 10-5 Mo-99 66.6 hr 2.09 x 10-2 Tc-99m 6.007 hr 7.95 x 10-2 Tc-101 14.2 min 1.11 x 10-1 Ru-103 39.8 day 1.78 x 10-5 Ru-106 368 day 2.40 x 10-6 Te-129m 34.1 day 6.27 x 10-5 Te-132 78.0 hr 1.25 x 10-2 Cs-134 2.06 yr 1.46 x 10-4 Cs-136 13.0 day 9.61 x 10-5 Cs-137 30.2 yr 2.19 x 10-4 Cs-138 32.3 min 2.20 x 10-1 Ba-139 83.2 min 1.78 x 10-1 Ba-140 12.8 day 8.04 x 10-3 Revision 12 11.1-23 January, 2003

TABLE 11.1-4 (Continued)

Concentration Isotope _Half-Life (Ci/g)

Ba-141 18.3 min 2.10 x 10-1 Ba-142 10.7 min 1.99 x 10-1 Ce-141 32.53 day 3.55 x 10-5 Ce-143 33.0 hr 3.24 x 10-5 Ce-144 284.4 day 3.13 x 10-5 Pr-143 13.58 day 3.45 x 10-5 Nd-147 11.06 day 1.25 x 10-5 Np-239 2.35 day 2.19 x 10-1 Revision 12 11.1-24 January, 2003

TABLE 11.1-5 COOLANT ACTIVATION PRODUCTS IN REACTOR WATER AND STEAM Reactor Water Steam Concentration Isotope Half-Life Concentration (Ci/g) (Ci/g)

N-13 9.99 Min 1.5 -3 1.0 -1 N-16 7.13 Sec 5.0 +1 4.0 +1 N-17 4.14 Sec 4.0 -2 2.0 -2 O-19 26.8 Sec 7.7 -1 1.8 +0 F-18 109.8 Min 4.4 -4 4.2 -2 Revision 12 11.1-25 January, 2003

TABLE 11.1-6 NON-COOLANT ACTIVATION PRODUCTS IN REACTOR WATER Concentration Isotope _Half-Life (Ci/g)

Na-24 15.0 Hr 2.1 x 10-3 P-32 14.31 Day 2.1 x 10-5 Cr-51 27.8 Day 5.2 x 10-4 Mn-54 313.0 Day 4.2 x 10-5 Mn-56 2.582 Hr 5.2 x 10-2 Co-58 71.4 Day 5.2 x 10-3 Co-60 5.258 Yr 5.2 x 10-4 Fe-59 45.0 Day 8.4 x 10-5 Ni-65 2.55 Hr 3.1 x 10-4 Zn-65 243.7 Day 2.1 x 10-5 Zn-69m 13.7 Hr 3.1 x 10-5 Ag-110m 253.0 Day 6.3 x 10-5 W-187 23.9 Hr 3.1 x 10-3 Revision 12 11.1-26 January, 2003

11.2 LIQUID WASTE MANAGEMENT SYSTEMS 11.2.1 DESIGN BASES 11.2.1.1 Power Generation Design Objectives The liquid radioactive waste (LRW) system is designed to collect and treat, for reuse or disposal, all radioactive (or potentially radioactive) liquid wastes produced in the plant. This is done in such a manner that, for all anticipated quantities of waste produced, the availability of the plant for power generation is not adversely affected.

11.2.1.2 Radiological Design Objectives The LRW system is designed to restrict releases of radioactive material to the environment and exposures to both operating personnel and the general public to as low as reasonably achievable (ALARA) in accordance with the guidelines given in <10 CFR 50, Appendix I>.

11.2.1.3 Design Criteria The original LRW system is designed in accordance with the following design criteria:

a. For each reactor at the site, the estimated annual total quantity of radioactive material (excluding tritium) above background in the liquid effluents released to unrestricted areas is less than 5 curies.
b. For the total radioactive liquid effluents, the resultant whole body dose to any individual offsite is less than 5 mrem/yr.

Revision 19 11.2-1 October, 2015

c. Design and construction of all LRW system components satisfies or exceeds the intent of all applicable criteria set forth in

<Regulatory Guide 1.143> (as detailed in <Section 1.8>) and ANSI N197-1976.

d. All LRW system components and the structure in which they are housed are designed and constructed in accordance with the codes, standards, seismic classifications, and safety classifications listed in
.
e. LRW components are located in areas of sufficient size and accessibility to facilitate efficient maintenance.

11.2.1.4 Cost-Benefit Analysis

<Section 11.2.3> includes an analysis that shows that the LRW system, as designed, is capable of controlling releases of radioactive material within the numerical design objectives of <10 CFR 50, Appendix I>.

Under the rules of Section II, Paragraph D of <10 CFR 50, Appendix I> a cost-benefit analysis is not required for this system because the design satisfies the Guides on Design Objectives for Light-Water-Cooled Nuclear Power Reactors, proposed in the Concluding Statement of Position of the Regulatory Staff, in Docket-RM-50-2.

11.2.1.5 Accident Analysis An analysis is included in <Chapter 15> to determine the radiological consequences for case situations in which equipment malfunctions and/or operator errors are hypothesized during periods of operation at design basis fuel leakage. Design provisions are included to prevent the uncontrolled release of radioactive material to the environment as a Revision 19 11.2-2 October, 2015

result of any single equipment malfunction or operator error. An evaluation of failures of single pieces of equipment is provided by

. Revision 16 11.2-2a October, 2009 11.2.1.6 Component Design Parameters With the exception of normal wearing parts, such as seals and bearings, all pumps, valves, piping, tanks, pressure vessels, and other components in the LRW system are fabricated from materials which are intended to provide a minimum service life of 40 years without replacement. In selecting materials to satisfy this criterion, due consideration is given to the following:
a. The corrosive nature of both the process fluid and the external environment.
b. Ease with which the material may be decontaminated.
c. Wall thickness requirements dictated by design pressures, temperatures, flow rates, and corrosion rates.
Tabulations of LRW system components and design parameters are presented by
,
,
,
,
, and
. 11.2.1.7 Surge Input Collection Capabilities Redundancy of equipment for collecting and processing inputs to the LRW system is described in detail in <Section 11.2.2> and summarized by
. Considerable excess capacity is built into the collecting and processing equipment of each subsystem to handle all anticipated normal and maximum input quantities. An evaluation of this capability is presented by
which shows that only when the suppression pool is drained for maintenance does the LRW system fall short of needed capacity. This occurrence is satisfactorily handled by reducing the rate at which the suppression pool is drained. This method adds, at most, three days to the outage and results in no increase in radiation exposures to operating personnel or the general public. Revision 16 11.2-3 October, 2009 11.2.1.8 Control of Tank Leakage and Overflows With the exception of the condensate storage tank, all tanks containing radioactive material are housed inside reinforced, concrete structures with floor drains for routing tank leakage to the LRW system. A seismically qualified retaining structure (dike) surrounds the condensate storage tank to contain any leakage from this source. Further information on these dikes is provided in <Section 9.2.6>. All tanks in the LRW and solid radioactive waste (SRW) systems have overflow lines piped up solid to embedded drain piping that is routed to sumps. Any water collected in these sumps is pumped back to the LRW collection tanks. Any vents and manways for the tanks are located above these overflow lines. Should an operator error result in an overflow through a vent or manway, the release would also be pumped back to the LRW collection tanks, since, as noted above, the tanks are housed in structures with floor drains that route tank leakage to the LRW system. For each tank in the LRW and SRW systems, level indication and high level alarms (with the exception of those associated with the radwaste evaporators and concentrated waste tanks) are provided in the radwaste building control room.
lists tanks outside containment which may contain potentially radioactive fluids. 11.2.1.9 ALARA Design Features Numerous features have been incorporated into the design of both the LRW system and the building housing this system to ensure that exposures of operating personnel to radiation will be kept within ALARA guidelines. The following is a listing of the most significant ALARA design features:
a. All floors and wall areas subject to contamination with radioactive material are coated with nuclear grade epoxy coatings to aid in decontamination.
Revision 19 11.2-4 October, 2015
b. With the exception of the detergent drains tanks and chemical waste distillate tanks, which are low in activity content, all redundant tanks are located in separate, shielded cubicles. This allows one tank to be repaired or inspected with minimal personnel exposure while the second tank is being used to process waste.
c. Most redundant pumps and process equipment are located in separate, shielded cubicles similar to those for the tanks as described above.
d. All normal operations are performed remotely from the centralized radwaste building control room. This eliminates exposures to operating personnel during normal operation and minimizes operating errors that could indirectly result in greater exposures to both operating and maintenance personnel.
e. Pipe lines containing radioactive fluids are routed through shielded chases. There is no instrumentation, valves or other equipment located in these chases, eliminating the need to enter them for any reason other than to maintain the piping itself.
f. As much as possible, pipes containing filter backwash slurries or spent resins make use of bends rather than standard elbow fittings to reduce the chances of plugging. As another precaution against sludge buildup, these lines are backflushed after every use.
g. Backflush connections are provided on all process piping in pump cubicles to permit this piping to be decontaminated before entering the cubicles for maintenance purposes.
Revision 19 11.2-5 October, 2015
h. All pump seals are mechanical type to minimize seal leakage and to eliminate the need for periodic adjustment of the seals. Rotating element facing material is chosen to maximize seal operating life.
i. The majority of valves are top-entry, diaphragm type with ethylene propylene terpolymer (EPT) elastomer diaphragms. This type of valve has the following advantages: a) no crud traps; b) no leakage unless the diaphragm fails; and c) quick, simple procedures for replacement of worn seals (diaphragm).
j. Materials of construction for pumps, valves, piping, tanks, and process equipment are selected to provide long-term corrosion resistance and improved decontamination capability. Most pumps, tanks and other process equipment are constructed of austenitic stainless steel. Where protection against chloride stress corrosion is needed, materials such as Alloy 20 stainless steel, and Incoloy are used. Piping and valves are constructed of materials such as austenitic stainless steel, Alloy 20 stainless steel, Incoloy, Yoloy or suitably lined with materials such as polypropylene or nuclear grade epoxy coatings. Other corrosion resistant materials may also be used.
11.2.1.10 Control of Inadvertent Releases Releases as a result of equipment failures or malfunctions are discussed in <Section 11.2.1.5>. Another way in which unintentional releases could occur would be as a result of operator errors either allowing a tank to overflow or pumping the contents of the wrong tank to the discharge tunnel entrance structure. Provisions for control of tank overflows are discussed in <Section 11.2.1.8>. Provisions for preventing the contents of the wrong tank from being discharged are discussed below. Revision 13 11.2-6 December, 2003 All LRW system discharges are directed to the Unit 1 emergency service water discharge pipe. All pipe lines going to the discharge point are routed through one central discharge flow control station, where the liquid can be directed through a flow control station. The flow control valve is remote-manually adjusted from the radwaste building control room. Between the control valve station and each sample tank that can be discharged is a power operated shutoff valve that must be opened before a tank can actually be drained to the discharge point. As protection against inadvertent discharges, an administratively controlled, manual, normally locked closed valve with position indicating limit switches is provided in series with each discharge isolation valve. 11.2.2 SYSTEM DESCRIPTION 11.2.2.1 Input Streams The LRW system is designed and sized to handle all radioactive liquid wastes for the Perry Nuclear Power Plant, based on having a condensate polishing treatment system as discussed in <Section 10.4.6>. The input streams for the system are shown on the detailed process flow diagram in <Figure 11.2-1 (1)>, <Figure 11.2-1 (2)>, <Figure 11.2-1(3)>, and <Figure 11.2-1 (4)>. For these streams, normal and expected maximum quantities of significant radioactive nuclides and total flow quantities are given in
. 11.2.2.2 Separation of Inputs Incoming streams of liquid waste are collected and treated in one of four separate process streams according to their composition. These four subdivisions are high purity/low conductivity wastes (primarily Revision 19 11.2-7 October, 2015 equipment drains), medium-to-low purity/medium conductivity wastes (primarily floor drains), high conductivity chemical wastes, and detergent drains. In addition to handling these four categories of liquid waste, the LRW system collects spent resin slurries and filter backwash slurries prior to being sent to the SRW disposal system. 11.2.2.3 Previous Experience The type of process equipment used in the system described herein has been used effectively in many previous BWR units, including Dresden Units 1, 2 and 3, Quad Cities Units 1 and 2, Oyster Creek, and Nine Mile Point. Justification for the decontamination factors used for this equipment is based on available data from several operating units, equipment manufacturers data, topical reports, and standards given in (Reference 1), (Reference 2), (Reference 3), (Reference 4), (Reference 5), (Reference 6), (Reference 7), (Reference 8), (Reference 9), (Reference 10), (Reference 11), (Reference 12), (Reference 13), (Reference 14), (Reference 15), (Reference 16), (Reference 17), (Reference 18), (Reference 19), (Reference 20), (Reference 21), (Reference 22), (Reference 23), (Reference 24), and (Reference 25). 11.2.2.4 Treatment of High Purity/Low Conductivity Wastes Input streams to this subsystem consist of equipment drains, cask pit drawdown, suppression pool water (normally diverted to suppression pool cleanup system), blowdown of reactor water (normally directed to hotwell), rinse water from condensate demineralizers, and residual heat removal system flush/test. These inputs are collected in one of two waste collector tanks, each sized to hold one days maximum normal input. With the exception of equipment drains, these waste streams can be diverted to the floor drain collector tanks if water quality or flow Revision 12 11.2-8 January, 2003 conditions warrant. After a batch of waste is collected, it is sent through a traveling belt filter to remove suspended solids, and then a mixed-bed demineralizer to remove dissolved solids. Alternate flow paths for treatment of these wastes are discussed in <Section 11.2.2.13>. Two waste sample tanks, each sized to hold one batch of waste, are provided for sampling, mixing and temporary storage of the treated effluent. After a batch is sampled, it may be recycled to the waste collector tank for further treatment, sent to the condenser Hotwell (normal path), the condensate storage system or discharged. The system is completely redundant, either through backup equipment or cross-ties with identical equipment in one of the other subsystems. The major inputs to the high purity subsystem are equipment drains. The embedded drainage piping system for collecting this waste water is described in <Section 9.3.3>. The equipment drain piping in each structure housing radioactive (or potentially radioactive) fluid systems is routed to a sump located at the lowest elevation of the building. After one of these sumps is filled, one of two redundant, vertical sump pumps automatically pumps the contents to the waste collector tanks in the LRW system. 11.2.2.5 Treatment of Medium-to-Low Purity/Medium Conductivity Wastes Input streams to this subsystem consist of floor drains, decantate from the backwash settling tanks, decantate from the solid radwaste disposal system and backwash from the radwaste and condensate demineralizers. These inputs are collected in one of two floor drain collector tanks, each sized to hold approximately three days maximum normal input. With the exception of floor drains, these waste streams can be diverted to the waste collector tanks if water quality or flow conditions warrant. After a batch is collected, it is normally filtered, demineralized and re-used. Alternate flow paths for treatment of these wastes are discussed in <Section 11.2.2.13>. Two floor drain sample tanks, each Revision 16 11.2-9 October, 2009 sized to hold one batch of waste, are provided for sampling and temporary storage of treated effluent. After sampling, a batch is either recycled for further treatment, sent to condenser hotwell (normal path), the condensate storage system or discharged. The system is completely redundant, either through backup equipment or cross-ties with identical equipment in one of the other subsystems. The major inputs to the medium-to-low purity subsystem are floor drains, which consist of miscellaneous unidentified equipment leakage and floor washdown. The embedded drainage piping system for collecting this waste water is described in <Section 9.3.3>. The floor drain piping in each structure housing radioactive (or potentially radioactive) fluid systems is routed to a sump located at the lowest elevation of the building. After one of these sumps is filled, one of two redundant, vertical sump pumps automatically sends the contents to the floor drain collector tanks in the LRW system. 11.2.2.6 (Deleted) Revision 17 11.2-10 October, 2011 Flow paths for treatment of these wastes are discussed in <Section 11.2.2.13>. 11.2.2.7 Treatment of Detergent Drains Inputs to this subsystem consist of personnel decontamination solutions and floor drains from nonradioactive areas of the control complex. Cleaning of protective clothing will be performed onsite and/or contracted offsite. All waste inputs are collected in the laundry and floor drains sump located at the lowest elevation of the control complex. When this sump is filled, one of two redundant sump pumps automatically transfers the contents to the LRW system detergent drains tanks. This waste is then collected and manually drained to the Radwaste Floor Drain so that it can be processed. 11.2.2.8 Treatment of Spent Resins Spent resins from the mixed-bed condensate demineralizers, waste demineralizer, floor drains demineralizer, and suppression pool Revision 17 11.2-11 October, 2011 demineralizer are collected in two spent resin storage tanks. Each tank is sized to hold the resins for six months. The spent resins are transferred to the SRW disposal system as a water slurry. 11.2.2.9 Treatment of Filter/Demineralizer Backwash Backwash slurries from the condensate filter, fuel pool filter/demineralizer and RWCU filter/demineralizer backwash receiving tanks are pumped to settling tanks located in the radwaste building. The sludge is allowed to settle to the bottom of these tanks while relatively clean water is drawn off the top and pumped to the floor drain collector tanks or waste collector tanks for further treatment. Periodically, the sludge is transferred to the SRW disposal system as a water slurry. 11.2.2.10 Detailed Component Design Piping and instrumentation for the LRW system are shown in <Figure 11.2-1>. For a definition of symbols used on this system diagram, see <Figure 1.2-22>. Design data for all LRW system components is given in
,
,
,
,
, and
. The safety class for equipment and piping in the system is given in
. Also shown in this table are the seismic classifications and principal construction codes for LRW system components and for the radwaste building.
a. Collection Tank Design All collection tanks are atmospheric, cylindrical, stainless steel tanks and are either horizontal or vertical. Vertical tanks have closed tops and dished bottoms for easy drainage. Vent, overflow, recycle, and drain lines are provided for each tank. A level Revision 12 11.2-12 January, 2003
sensor is provided on each tank for remote level indication, level recording and alarm/control functions.
b. Pump Design LRW pumps other than sump pumps are horizontal, centrifugal type, driven by 460 volt drip proof motors. Each pump is provided with inlet and outlet shutoff valves for maintenance and a discharge pressure sensor with readout in the radwaste building control room (RWBCR). All pump seals are single or double mechanical type.
In addition, filter aid pumps (waste collector and floor drain collector) are positive displacement.
c. Waste Collector Filter/Floor Drains Filter Each filter is a flatbed, continuous belt type, precoat filter unit rated at 100 to 150 gpm when used to filter waste water. Each unit can also be used to dewater resin or filter backwash slurries, for which case the process rate is 50 gpm.
For improved filtration efficiency, provisions are made for body feed of precoat material to the filter influent. During periods of non-use, water is continuously recirculated through the filter to prevent deterioration of the filter precoat. Upon completion of a filtration run, the precoat material and accumulated crud is partially dried by air and the filter belt is indexed, causing the semi-dry cake to fall off the end of the belt and down a stainless steel chute into a waste mixing/dewatering tank in the SRW disposal system. Each filter has a filtration surface area of 68 square feet. Operating differential pressure varies from 2 to 13 psi. Design differential pressure is 15 psi. Revision 12 11.2-13 January, 2003
d. Waste Demineralizer/Floor Drains Demineralizer These demineralizers are identical 200 gpm mixed bed units, using a mixture of cation and anion resins. Each demineralizer is designed for a process flow rate of 6 to 8 gpm per square foot. They are cross-tied by manual valves to achieve redundancy in both subsystems.
Maximum pressure differential at rated flow is 22.5 psi. A demineralizer run may be terminated on a high differential pressure or a high conductivity signal. The spent resin is then transferred to one of the spent resin tanks.
e. (Deleted)
f. (Deleted)
Revision 17 11.2-14 October, 2011
g. Settling Tanks All settling tanks are vertical, atmospheric, cylindrical, stainless steel tanks with closed tops and dished bottoms. Each tank is provided with vent, overflow, drain, recycle, and decant lines. Manways are provided on the settling tanks, which are located above the overflow line. The manways may be left open to support operational practices. Connections for flushwater and sparging air or condensate are also provided.
Four ultrasonic level indicators are provided on each tank to indicate in the RWBCR when the sludge level is at 25, 50, 75, or 100 percent of the maximum permissible level. The tanks are designed so that this maximum level is below the elevation of the decant lines. Each tank is also provided with a liquid level sensor for remote level indication and alarm/control functions.
h. Spent Resin Tanks Two vertical, atmospheric, cylindrical, stainless steel spent resin tanks are provided. Each tank has a closed top and dished bottom and is provided with vent, drain, overflow, recycle, and flush lines. The entrance to the overflow line is provided with a wire mesh screen to prevent resins from entering the overflow.
Revision 16 11.2-15 October, 2009 Each tank is provided with a liquid level sensor for remote level indication and alarm/control functions.
i. Concentrated Waste Tanks Two vertical, atmospheric, cylindrical, Incoloy concentrated waste tanks are provided. Each tank has a closed top, dished bottom and vent, overflow, drain, and recycle lines. All lines normally containing concentrated waste are heat traced and insulated to prevent solidification of the concentrate. Each tank is provided with a heating element to maintain the tank temperature between 120F and 150F. A level sensor is provided for remote indication and control functions. Temperature elements are provided to monitor and record temperature in the tanks.
j. Sample Tanks The waste sample tanks and floor drains sample tanks are vertical, atmospheric, cylindrical, stainless steel tanks. The chemical waste distillate tanks are horizontal, atmospheric, cylindrical, stainless steel tanks. All tanks have vent, overflow, drain, and recycle lines. Each tank has a level sensor for remote level indication, level recording and alarm/control functions.
k. Sumps Radioactive floor and equipment drains are collected in sumps located in the basement of all structures housing radioactive fluid systems. These sumps range in size from 50 to 1,000 gallons. With the exception of those sumps that are normally nonradioactive, all sumps are lined with stainless steel for leakage control and to facilitate decontamination.
Revision 19 11.2-16 October, 2015 Many sumps are provided with a small recessed boot in the area of the bottom from which the sump pump takes suction. This ensures that the pump suction is submerged at all times while allowing most of the sump to be drained completely to minimize buildup of radioactive sludge and to facilitate decontamination. All sumps are covered with grating or solid plates. Solid plates are used where shielding is needed, or if the sump is in an open area where litter could end up in the sump. Each sump is provided with level switches for alarm and control functions. Sumps inside containment have additional instrumentation for leak rate detection as discussed in <Section 7.6>. The quantity of waste water sent to the LRW system from each sump is monitored in the RWBCR.
l. Sump Pumps Except for the annulus sump, which is expected to be used very infrequently, all sumps have redundant, duplex sump pumps.
Vertical turbine pumps are used in sumps containing relatively clean water. Standard vertical, open impeller, centrifugal sump pumps are used in sumps where trash could accumulate. All sump pumps are provided with suction strainers to prevent refuse from clogging or damaging the pump impeller. Pump motors are totally enclosed and fan cooled to prevent contamination of the motor internals. 11.2.2.11 Field Routed Pipe Routing of piping and tubing in the LRW system that normally carry radioactive fluids is shown on piping drawings to ensure proper Revision 19 11.2-17 October, 2015 protection of operating personnel against exposure to radiation. Therefore, there will be no field routed radioactive piping or tubing for which shielding design criteria or controls will be necessary. 11.2.2.12 System Control and Operating Procedures
a. General Control and indication of equipment associated with the radwaste evaporators is on control panel H51-P031 in the RWBCR. Control and indication of all other radwaste sub-systems is via the distributed control system (DCS) in the RWBCR. The DCS operators console (1H51P0510) consists of operator workstations used for control and indication of the LRW system and a maintenance/engineering workstation. Various display screens shown on monitors at the operators console provide the operating status (off/on) of system pumps and the position (open/closed) of power operated valves.
Important system parameters such as tank levels, pump discharge pressures, etc. are also indicated and/or recorded on this operators console. An alarm sounds at the console if abnormal conditions such as high tank level or high discharge activity should occur. Additional control panels are located near the radwaste filters and demineralizers for use when reconditioning this equipment. After a filter has received a fresh precoat or a demineralizer has been refilled with new resin, control of this equipment is returned to the radwaste building control room. Revision 19 11.2-18 October, 2015
b. Distributed Control System The control logic for the LRW system is controlled by a distributed control system. Redundant power sources are provided to the DCS.
Separate logic is provided for train A and train B components to the extent possible. The DCS has two pairs of redundant processors. The DCS features an automatic and a manual mode, either of which can be used to control the LRW system. Revision 19 11.2-18a October, 2015
c. Normal Control of Discharges Except for detergent wastes, all liquid effluents from the LRW system are normally routed to the condensate storage system or main condenser for reuse in the plant. This is done on a batch basis after a sample of the effluent is taken to determine if it is suitable for reuse. If the sample does not meet the water quality standards for condensate makeup given in
the batch is either recycled for further treatment or discharged through the discharge tunnel entrance structure, depending on the chemical content and activity level. All streams to be discharged, with the exception of the atmospheric drain line from the Turbine Building supply plenums are routed through one central flow control station, where a flow control valve is used. These valves are modulated remote-manually from the RWBCR to achieve the desired flow rate. The stream is then monitored for gross gamma activity and routed to the discharge tunnel entrance structure, which discharges to the environment at the point shown in <Figure 1.2-18>. For each batch discharged, the activity monitor is set to actuate an alarm in the RWBCR if the activity level exceeds a preselected value. This value is calculated for each batch based on the activity level of a sample taken from the batch and on the flow rate of the dilution flow at the time that it is desired to discharge the batch. The value is set so that after dilution, the concentration will be substantially below the limits as defined by <10 CFR 20>. (Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the <10 CFR 20> regulations prior to October 4, 1993. Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised <10 CFR 20> dated October 4, 1993.) Revision 13 11.2-19 December, 2003 11.2.2.13 Selection of Normal and Alternate Flow Paths Normal flow paths for all input streams to the liquid radwaste system are described in <Section 11.2.2.4>, <Section 11.2.2.5>, <Section 11.2.2.6>, <Section 11.2.2.7>, <Section 11.2.2.8>, and <Section 11.2.2.9>. However, because of the variable nature of these input streams, alternate flow paths for their treatment may sometimes be necessary. In <Figure 11.2-2>, the normal and alternate flow paths for each input stream are summarized. For each flow path, the percentages of total flow are given for expected normal operation, design and sizing of equipment and calculation of quantities of radioactivity discharged. Explanation of each flow path used is given in
. Water discharged from the atmospheric drain line from the Turbine Building supply plenums will be periodically monitored with grab samples. This source is from a radiologically clean area but does have the potential to condense radioactive tritium that has been recycled back into this plenum from the plant gaseous vents. Detectable tritium in this pathway can also be from naturally occurring tritium production from cosmic radiation. The fluids condensed in this pathway have already been evaluated for compliance with limits defined by <10 CFR 20> via the pathway analysis for gaseous vents. 11.2.2.14 Performance Tests Prior to plant startup, all equipment in the radwaste system will be tested for operability. Reports in the literature on performance tests for this equipment are given in (Reference 1), (Reference 2), (Reference 3), (Reference 4), (Reference 5), (Reference 6), (Reference 7), (Reference 8), (Reference 9), (Reference 10), (Reference 11), (Reference 12), (Reference 13), (Reference 14), (Reference 15), (Reference 16), (Reference 17), (Reference 18), (Reference 19), (Reference 20), Revision 13 11.2-20 December, 2003 (Reference 21), (Reference 22), (Reference 23), (Reference 24), and (Reference 25). 11.2.3 RADIOACTIVE RELEASES 11.2.3.1 Description The criteria for recycle, treatment and discharge of radioactive wastes is discussed in <Section 11.2.2>. In calculating the radioactive releases to the environment it was assumed that 10 percent of the high purity, chemical waste streams and 25 percent of the low purity waste stream are discharged. 11.2.3.2 Dilution Factors The liquid waste discharged to the environment from LRW systems is diluted by the service water and/or ESW of Unit 1. Values in
were calculated using the normal minimum dilution flow of 30,000 gpm. During certain operating conditions or certain seasons, flows may be less than the normal minimum flows indicated. However, flows will exceed the normal minimum flow a substantial portion of the year, thus the values in
are a valid conservative estimate of annual discharges. 11.2.3.3 Release Points Releases to the environment are by way of the discharge tunnel entrance structure or to the plant storm drains for the discharge from the atmospheric drain line on the Turbine Building supply plenums. The discharge tunnel entrance structure is shown on the process flow diagram in <Figure 11.2-2> and the site plot plan in <Figure 1.2-18>. Revision 16 11.2-21 October, 2009 11.2.3.4 Estimated Releases The release rate of radioactive materials in liquid effluents is presented in
and
. These values were calculated with the GALE Code and are based on the assumptions and parameters provided in <NUREG-0016> (BWR-GALE Code) and
. As shown in
, the estimated releases are a small fraction of the limits as defined by <10 CFR 20>, and are considered as low as reasonably achievable. (Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the <10 CFR 20> regulations prior to October 4, 1993. Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised <10 CFR 20> dated October 4, 1993.) The estimated offsite doses for the Perry site and a comparison with the design objectives of <10 CFR 50, Appendix I> and the dose limits of <40 CFR 190> are presented in <Section 5.2.4> of the PNPP Environmental Report. Revision 16 11.2-21a October, 2009 Subsequent to the original evaluation discussed above, modifications have been made to the condensate cleanup and liquid radwaste systems which could result in liquid effluents which are different in quantity or activity concentration than those predicted in the original analysis. The control of the liquid radwaste system effluents, as described in <Section 11.2.2.12>, remains unchanged and ensures that liquid effluent releases remain within the design objectives of <10 CFR 50, Appendix I> and substantially below the limits as defined by <10 CFR 20>. The release of condensate from the atmospheric drain line from the Turbine Building supply plenums may also contain radioactive tritium due to the recycle of released air from the gaseous effluent vents. The release of radioactive tritium in the gaseous vents is analyzed in <Section 11.3>. 11.

2.4 REFERENCES

FOR SECTION 11.2

1. BWR Radwaste System Performance, H. L. Loy and W. F. Dietrich, Symposium on Waste Management at Nuclear Reactors, AiChE National Meeting, Cincinnati, Ohio, May 16, 1971, NEDM-10346.
2. Design and Project of Radioactive Waste Treatment and Disposal System at Large BWR Nuclear Power Stations, A. Shimozoto, M. Takeshima, and K. Osano, Hitachi Review, Volume 19, No. 12, pp 426-434, UDC 621.039.7.
3. Management of Radioactive Wastes at Nuclear Power Stations: J. O.

Blomeke, et. al., ORNL-4070.

Revision 13 11.2-22 December, 2003

4. Experience With Precoat Filters in Nuclear Power Plants, C. H.

Becker and J. S. Brown, presented at 34th Annual American Power Conference, Chicago, Illinois, April 18 - 20, 1972.

5. Test of AMF 2 gpm Radioactive Waste Evaporator at Rochester Ginna Power Station - AMF-BEAIRD Engineering Calculation, Dated October 8, 1971.
6. Removal of Radionuclides from Water by Water Treatment Processes, Ray J. Morton and Conrad P. Stroub, American Water Work Association Journal, Volume 48, May 1956, p. 545.
7. Disposal of Radioactive Wastes, Abel Walman, Ibid., Volume 49, May 1957, p. 505.
8. Mixed-Bed Ion Exchange for the Removal of Radioactivity, H. Gladys Swope, Op. Cit., Volume 49, August 1957, p. 1085.
9. Chemical Coagulation Studies on Removal of Radioactivity in Waters, Lloyd R. Setter and Helen H. Russell, Op. Cit., Volume 50, May 1958, p. 590.
10. Decontamination of Radioactive Sea Water, Minorce Honma and Allen E. Greendale, Op. Cit., Volume 50, November 1958, p. 1490.
11. Removal of Radionuclides from the Pasco Supply by Conventional Treatment, Robert L. Junkins, Op. Cit., Volume 52, July 1960,
p. 834.
12. Behavior of Radionuclides on Ion Exchange Resins, Dade W.

Moellor, George W. Leddicatte and Sam A. Reynolds, Op. Cit.,

Volume 53, July 1961, p. 862.

Revision 12 11.2-23 January, 2003

13. Effect of Detergents on the Decontamination of Radioactive Waste Water by Precipitation and Coagulation, Colin R. Frost, Op. Cit.,

Volume 54, September 1962, p. 1082.

14. Reducing Radioisotope Concentrations in Reactor Effluent by High Coagulant Feed, Wyatt B. Silker, Op. Cit., Volume 55, March 1963,
p. 355.
15. Decontamination of Radioactive Aqueous Effluent by the Calcium-Ferric Phosphate Process, Colin R. Frost, Op. Cit.,

Volume 55, October 1963, p. 1237.

16. Studies on Radioisotope Removal by Water Treatment Processes, Rolf Eliassen, Warren J. Kaufman, John B. Nesbitt and Morton I.

Goldman, Op. Cit., Volume 43, August 1951, p. 615.

17. Studies on Radioisotope Removal by Water Treatment Processes; Discussion, Arthur E. Gorman, Op. Cit., Volume 43, August 1951,
p. 630.
18. Studies on Radioisotope Removal by Water Treatment Processes; Discussion, Arthur E. Gorman, Op. Cit., Volume 43, August 1951,
p. 633.
19. Studies on Radioisotope Removal by Water Treatment Processes; Authors Closure; Op. Cit., Volume 43, August 1951, p. 635.
20. Removal of Radioactive Material from Water By Slurrying With Powdered Metal, William J. Lacy, Op. Cit., Volume 44, September 1952, p. 824.
21. Reduction of Radioactivity in Water, Sol Goodgal, Earnest F.

Gloyna and Dayton E. Carritt, Op. Cit., Volume 46, January 1954,

p. 66.

Revision 12 11.2-24 January, 2003

22. Accelerating Calcium Carbonate Precipitation of Softening Plants, Robert F. McCauley and Rolf Eliassen, Op. Cit., Volume 47, May 1955, p. 487.
23. Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor, U.S. Department of Health, Education and Welfare, BRH/DER 70-1, March 1970.
24. Ibid., BRH/DER 70-2, March 1970.
25. Management of Radioactive Wastes at Nuclear Power Plant, IAEA, Vienna, 1968, STI/PUB/208.

Revision 12 11.2-25 January, 2003

TABLE 11.2-1 SINGLE EQUIPMENT ITEM MALFUNCTION EVALUATION Equipment Design Item Malfunction Consequences Precautions Discharge Does not respond Radioactive isotope Radiation monitor flow to signal to concentration in downstream of the control throttle flow discharge could flow control valve valve exceed limits of signals flow

<10 CFR 20>. control valve to close. Monitor has high radiation alarm to alert the control room operator. Remote manual isolation of discharge can be initiated using redundant valves.

Discharge Improperly Activity of liquid Monitor has down radiation calibrated or discharged to scale alarm to monitor power failure environment is not warn operator in monitored or control room of recorded. of loss of power.

Recycle line is provided on sample tank to permit each batch of waste being dis-charged to be sampled for iso-topic content prior to release.

Discharge Improperly Quantity of liquid Level recorder on flow calibrated or being discharged to sample tank pro-monitor power failure environment is not vides indirect monitored or record of quantity recorded. of liquid released.

Revision 12 11.2-26 January, 2003

TABLE 11.2-1 (Continued)

Equipment Design Item Malfunction Consequences Precautions Service Flow through Radioactive isotope Flow sensor on water line is blocked concentration in weir of discharge discharge or lost. discharge could structure signals header flow exceed limits of power operated

<10 CFR 20>. valve in LRW discharge line to close on loss of dilution water flow.

Revision 16 11.2-27 October, 2009

TABLE 11.2-2 DESIGN DATA FOR LIQUID RADWASTE SYSTEM SUMPS Stainless Operating Steel Capacity Type of Liner Sump Description Number (gal) Cover Plate Provided Drywell Equipment 1 500 Shielding Yes Drains Plate Containment 1 440 Checkered Yes Equipment Drains Plate Radwaste Building 1 500 Checkered Yes Equipment Drains Plate Intermediate 1 500 Grating Yes Building Equipment Drains Turbine Power 1 1,000 Checkered Yes Complex Equipment Plate Drains Control Complex 1 500 Checkered Yes Equipment Drains Plate Drywell Floor Drains 1 125 Shielding Yes Plate Containment Floor 1 390 Checkered Yes Drains Plate Annulus Floor Drains 1 50 Checkered No Plate Intermediate 1 500 Grating Yes Building Floor Drains Auxiliary Building 1 345/675 Checkered Yes Floor/Equipment Plate Drains Turbine Power 1 1,000 Checkered Yes Complex Floor Drains Plate Revision 19 11.2-28 October, 2015

TABLE 11.2-2 (Continued)

Stainless Operating Steel Capacity Type of Liner Sump Description Number (gal) Cover Plate Provided Turbine Laydown 1 450 Checkered No Area Floor Drains Plate Radwaste Building 1 500 Checkered Yes Floor Drains Plate Turbine Lube Oil 1 750 Checkered No Area Floor Drains Plate Heater Bay Floor 1 270 Checkered Yes Drains Plate Turbine Power 1 270 Grating Yes Complex Chemical Drains Control Complex 1 500 Checkered Yes Laundry and Floor Plate Drains NOTE:

(Deleted)

Revision 19 11.2-29 October, 2015

TABLE 11.2-3 DESIGN DATA FOR LIQUID RADWASTE SYSTEM SUMP PUMPS Max.

Design Design Design TDH @ Shutoff Fluid Press. Temp. Flow Rate Design Head Temp. Tag Sump Pump Type Quantity (psig) (F) (gpm) Pt. (ft) (ft) (F) Material Number Drywell Equipment Drain Duplex 2 100 200 50 60 80 150 CI 1G61C001A Sump Pumps V.T.(1) 1G61C001B Containment Equipment Drain Duplex 2 100 200 50 60 80 150 CI 1G61C002A Sump Pumps V.T.(1) 1G61C002B Radwaste Bldg. Equipment Duplex 2 100 150 50 65 82 100 CI G61C003A Drain Sump Pumps V.T.(1) G61C003B Intermediate Bldg. Duplex 2 100 150 50 65 82 100 CI G61C004A Equipment Drain Sump Pumps V.T.(1) G61C004B Turbine Power Complex Duplex 2 100 150 100 110 140 100 CI 1G61C007A Equipment Drain Sump Pumps V.T.(1) 1G61C007B Drywell Floor Drain Duplex 2 100 150 50 65 82 100 CI 1G61C008A Sump Pumps V.T.(1) 1G61C008B Containment Floor Drain Duplex 2 100 150 50 65 88 100 CI 1G61C009A Sump Pumps 1G61C009B Annulus Floor Drain Single 2 100 150 25 55 58 100 CI 1G61C010 Sump Pumps 2G61C010 Intermediate Bldg. Floor Duplex 2 100 150 50 60 80 100 CI G61C011A Drain Sump Pumps G61C011B Auxiliary Bldg. Floor Duplex 2 100 150 100 70 82 100 CI 1G61C012A Drain Sump Pumps 1G61C012B Turbine Power Complex Duplex 2 100 150 100 85 109 100 CI 1G61C014A Floor Drain Sump Pumps 1G61C014B Turbine Power Complex Single 1 100 150 750 130 170 100 CI 1G61C014C Floor Drain Sump Pumps V.T.(1)

Revision 12 11.2-30 January, 2003

TABLE 11.2-3 (Continued)

Max.

Design Design Design TDH @ Shutoff Fluid Press. Temp. Flow Rate Design Head Temp. Tag Sump Pump Type Quantity (psig) (F) (gpm) Pt. (ft) (ft) (F) Material Number Turbine Laydown Area Duplex 2 100 150 50 25 32 100 CI 1G61C015A Floor Drain Sump Pumps 1G61C015B Radwaste Bldg. Floor Duplex 1 100 150 50 60 82 100 CI 0G61C016A Drain Sump Pumps Duplex 1 100 150 50 65 88 100 CI 0G61C016B Turbine Lube Oil Area Duplex 2 100 150 50 50 75 100 CI 1G61C005A Floor Drain Sump Pumps 1G61C005B Heater Bay Floor Drain Duplex 2 100 150 25 95 98 100 CI 1G61C019A Sump Pumps 1G61C019B Control Complex Equipment Duplex 2 100 150 50 65 82 100 CI G61C013A Drain Sump Pumps V.T.(1) G61C013B Turbine Power Complex Duplex 2 100 150 25 60 64 100 SS 1G61C017A Chemical Drain Sump 1G61C017B Pumps Control Complex Laundry Duplex 2 100 175 50 75 96 150 CI G61C018A and Floor Drain Sump G61C018B Pumps Auxiliary Bldg. Equipment Duplex 2 100 150 250 70 104 100 CI 1G61C020A Drain Sump Pumps V.T.(1) 1G61C020B NOTE:

(1)

V.T. - Abbreviation for vertical turbine.

Revision 14 11.2-31 October, 2005

TABLE 11.2-4 DESIGN DATA FOR LIQUID RADWASTE SYSTEM TANKS Design Design Operating Quantity Head Press. Temp. Capacity(1) Tag Tank (2 Units) Type Design (Psig) (F) Matl (Gal.) Number Waste Collector Tanks 2 Vertical Flat Top, Atmos. 150 304 SS 36,500 G50-A001A Dish. Bot. G50-A001B Waste Sample Tanks 2 Vertical Flat Top, Atmos. 150 304 SS 34,000 G50-A002A Dish. Bot. G50-A002B Floor Drain Collector Tanks 2 Vertical Flat Top, Atmos. 150 304 SS 36,500 G50-A003A Dish. Bot. G50-A003B Floor Drain Sample Tanks 2 Vertical Flat Top, Atmos. 150 304 SS 34,000 G50-A004A Dish. Bot. G50-A004B Chemical Waste Tanks 2 Vertical Flat Top, Atmos. 150 316 SS 19,650 G50-A005A Dish. Bot. G50-A005B Concentrated Waste Tanks 2 Vertical Flat Top, Atmos. 200 Incoloy 4,900 G50-A006A Dish. Bot. 825 G50-A006B Chemical Waste Distillate 2 Horiz. Shallow Atmos. 150 304 SS 19,100 G50-A007A Tank Dished G50-A007B Detergent Drains Tanks 2 Horiz. Shallow Atmos. 150 304 SS 1,550 G50-A008A Dished G50-A008B Spent Resin Tanks 2 Vertical Flat Top, Atmos. 150 304 SS 9,500 G50-A009A Elip. Bot. G50-A009B Condensate Filter Backwash 2 Horiz. Shallow Atmos. 150 304 SS 9,900 G50-A010 Receiving Tank Dished G50-A010 Condensate Filter Backwash 2 Vertical Flat Top, Atmos. 150 304 SS 17,600 G50-A011A Settling Tanks Dish. Bot. G50-A011B RWCU F/D Backwash 2 Vertical Flat Top, Atmos. 150 304 SS 4,400 G50-A013A Settling Tanks Dish. Bot. G50-A013B Revision 12 11.2-32 January, 2003

TABLE 11.2-4 (Continued)

Design Design Operating Quantity Head Press. Temp. Capacity(1) Tag Tank (2 Units) Type Design (Psig) (°F) Matl (Gal.) Number Fuel Pool F/D Backwash 2 Vertical Flat Top, Atmos. 150 304 SS 17,600 G50-A014A Settling Tanks Dish. Bot. G50-A014B LRW Filter Precoat Tank 2 Vertical Flat Top, Atmos. 150 304 SS 1,475 G50-A015 Dish. Bot. (Plasite lined)

LRW Demineralizer Resin 1 Vertical Flat Top, Atmos. 150 CS 825 G50-A016 Feed Tank Cone Bot. (Koroseal lined)

LRW Filter Aid Tank 1 Vertical Flat Top, Atmos. 150 304 SS 1,000 G50-A017 Dish. Bot.

Fuel Pool F/D Backwash 1 Horiz. Shallow Atmos. 150 304 SS 9,400 G50-A022 Receiving Tank Dished LRW Phosphate Tank 1 Vertical Flat Top, Atmos. 100 304 SS 250 G50-A023 Flat Bot.

LRW Hot Water Heater 1 Vertical ASME 130 200 CS 500 G50-B003 Dished (Phenolic Lining)

Waste Collector 1 Vertical Flat Top, Atmos. 150 304 SS 400 G50-A024 Filtrate Tank Dish Bot.

Floor Drains 1 Vertical Flat Top, Atmos. 150 304 SS 400 G50-A025 Filtrate Tank Dish Bot.

NOTE:

(1)

Operating capacity is arbitrarily defined herein to be the volume of that portion of the tank up to 6 inches below the lowest point in the overflow line.

Revision 12 11.2-33 January, 2003

TABLE 11.2-5 DESIGN DATA FOR LIQUID RADWASTE SYSTEM PUMPS Max.

Design Design Design TDH @ Shutoff Fluid Press. Temp. Flow Rate Design Head Temp. Tag Pump Type Quantity (psig) (F) (gpm) Pt. (ft) (ft) (F) Material Number Waste Collector Horz. 2 125 150 150 175 205 140 316 SS G50-C001A Transfer Pumps Cent. G50-C001B Waste Sample Pumps Horz. 2 125 150 200 110 122 100 316 SS G50-C002A Cent. G50-C002B Floor Drains Collector Horz. 2 125 150 150 175 205 100 316 SS G50-C003A Transfer Pumps Cent. G50-C003B Floor Drains Sample Horz. 2 125 150 200 110 122 100 316 SS G50-C004A Pumps Cent. G50-C004B Chemical Waste Pumps Horz. 2 125 150 120 110 125 100 Alloy G50-C005A Cent. 20 SS G50-C005B Chemical Waste Horz. 2 125 150 200 110 122 120 316 SS G50-C006A Distillate Pumps Cent. G50-C006B Detergent Drains Pumps Horz. 2 125 150 50 110 114 140 316 SS G50-C007A Cent. G50-C007B Spent Resin Pumps Horz. 2 125 150 400 135 160 100 316 SS G50-C008A Cent. G50-C008B Condensate Backwash Horz. 2 125 150 450 80 90 100 316 SS 1G50-C009A Transfer Pumps Cent. 1G50-C009B Cond. Sludge Discharge Horz. 2 125 150 400 180 208 100 316 SS G50-C010A Mixing Pumps Cent. G50-C010B Condensate Sludge Horz. 2 125 150 450 75 81 100 316 SS G50-C011A Decant Pumps Cent. G50-C011B RWCU Backwash Horz. 1 125 150 350 95 105 120 316 SS 1G50-C012 Transfer Pumps Cent.

Revision 12 11.2-34 January, 2003

TABLE 11.2-5 (Continued)

Max.

Design Design Design TDH @ Shutoff Fluid Press. Temp. Flow Rate Design Head Temp. Tag Pump Type Quantity (psig) (F) (gpm) Pt. (ft) (ft) (F) Material Number RWCU Sludge Discharge Horz. 2 125 150 200 175 182 120 316 SS G50-C013A Mixing Pumps Cent. G50-C013B RWCU Sludge Decant Pumps Horz. 2 125 150 50 55 60 120 316 SS G50-C014A Cent. G50-C014B Fuel Pool Sludge Horz. 2 125 150 400 180 208 100 316 SS G50-C015A Discharge Mixing Pumps Cent. G50-C015B Fuel Pool Sludge Horz. 2 125 150 450 75 86 100 316 SS G50-C016A Decant Pumps Cent. G50-C016B Waste Collector Horz. 1 125 150 150 100 120 140 CI G50-C017 Filtrate Pump Cent.

Floor Drains Horz. 1 125 150 150 100 120 100 CI G50-C018 Filtrate Pump Cent.

Waste Collector Filter Positive 1 125 150 0.12 to 134 - 100 316 SS G50-C019 Aid Pump Displnt. 1.2 Floor Drains Filter Positive 1 125 150 0.12 to 134 - 100 316 SS G50-C020 Aid Pump Displnt. 1.2 Radwaste Precoat Pumps Horz. 2 150 150 350 55 65 100 CI G50-C021A Cent. G50-C021B Spent Resin Transfer Horz. 1 125 150 200 175 182 100 316 SS 1G50-C022 Pumps Cent.

WEC Concentrate Pumps Horz. 4 125 250 45 220 Alloy G50-C023A Cent. 20 SS G50-C023B G50-C023C G50-C023D WEC Distillate Pumps Canned 2 125 150 31.2 120 316 SS G50-C024A G50-C024B Revision 12 11.2-35 January, 2003

TABLE 11.2-5 (Continued)

Max.

Design Design Design TDH @ Shutoff Fluid Press. Temp. Flow Rate Design Head Temp. Tag Pump Type Quantity (psig) (F) (gpm) Pt. (ft) (ft) (F) Material Number Concentrated Waste Horz. 2 125 200 150 150 180 150 Alloy G50-C026A Transfer Pumps Cent. 20 SS G50-C026B Fuel Pool F/D Backwash Horz. 1 125 150 450 120 135 100 316 SS G50-C027 Transfer Pump Cent.

Revision 12 11.2-36 January, 2003

TABLE 11.2-6 DESIGN DATA FOR LIQUID RADWASTE SYSTEM PIPING Pipe Line Specification Design Code Material Schedule or Wall G18-4 ANSI B31.1 Stainless steel 0.065 wall tubing; ASTM A213, Gr TP316 L1-3 ASME Code, Carbon steel; Sch. 80 (2 and Section III, ASME SA 160 Gr B less)

Class 3 Sch. 40 (2-1/2 thru 10)

L1-4 ANSI B31.1 Carbon steel; Sch. 80 (2 and ASTM A106 Gr B less)

Sch. 40 (2-1/2 thru 10)

L2-3 ASME Code Stainless steel; Sch. 40S Section III, ASME SA 312 or Class 3 SA 376 TP 304 L2-4 ANSI B31.1 Stainless steel; Sch. 40S ASTM A 312 or A 376 TP 304 L6-4 ANSI B31.1 Yoloy (nickel/ Sch. 80 (2 and copper alloy less) steel); Sch. 40 (2-1/2 ASTM A53 to 20)

L7-3 ASME Code, Alloy 20 stain- Sch. 40S Section III, less steel, ASME Class 3 SB 464 L7-4 ANSI B31.1 Alloy 20 stain- Sch. 40S less steel; ASTM B-464 N13-4 ANSI B31.1 Polypropylene Std. wall lined carbon steel pipe per ASTM A53 Revision 12 11.2-37 January, 2003

TABLE 11.2-7 DESIGN DATA FOR LIQUID RADWASTE SYSTEM PROCESS EQUIPMENT Design Design Design Quantity Press. Temp. Flow Rate Tag Equipment Type (2 Units) (psig) (F) (gpm) Material Number Waste Demineralizer Mixed bed 1 125 150 150 to 200 304 and 316 SS G50-D003 Floor Drains Mixed bed 1 125 150 150 to 200 304 and 316 SS G50-D004 Demineralizer Waste Collector Precoat; 1 See Note(2) 150 100 to 150 304 SS G50-D001 Filter Flat bed Floor Drains Filter Precoat 1 See Note(2) 150 100 to 150 304 SS G50-D002 Waste Evaporator/ Horizontal; 2 50 300 30 Incoloy 825 G50-B001A Condensers(3) bowed tube; (evaporator G50-B001B waste on section); 304 SS (evaporator) shell side; condenser section) G50-B002A forced G50-B002B recirculation (condenser)

Detergent(1) Cartridge 2 125 175 50 304 SS (vessel) G50-D005A Drains Epoxy impregnated G50-D005B Filters cellulose (element)

NOTES:

(1)

Detergent drain filters are abandoned in place.

(2)

Enclosure designed to operate at atmospheric pressure; filter shells are designed to cycle up to 15 psig during periodic blowdown/cleaning.

(3)

Evaporators steam supply has been cut and capped, rendering the evaporators inoperable for the waste processing function. Condensers in service.

Revision 17 11.2-38 October, 2011

TABLE 11.2-8

SUMMARY

OF LIQUID RADWASTE SYSTEM EQUIPMENT REDUNDANCY Maximum Normal Collecting Collecting Degree of or Processing or Processing Equipment Redundancy Capacity Capacity Equipment Drain None Varies (440 to Varies (440 to Sump 1,000 gal.) 1,000 gal.)

Equipment Drain 100% Varies (50 to Varies (100 to Sump Pump 250 gpm) 500 gpm)

Waste Collector 100% 36,500 gal. 73,000 gal.

Tank Waste Sample 100% 34,000 gal. 68,000 gal.

Tank Waste Collector 100% 150 gpm 150 gpm Transfer Pump Waste Sample 100% 200 gpm 200 gpm Pump Waste 100%(1) Varies (150 to Varies (150 to Demineralizer 200 gpm) 200 gpm)

Waste Collector 100%(2) 150 gpm 150 gpm Filter Floor Drains None Varies (50 to Varies (50 to Sump 1,000 gal.) 1,000 gal.)

Floor Drains 100%(3) Varies (25 to Varies (50 to Sump Pump 750 gpm) 750 gpm)

Floor Drains 100% 36,500 gal. 73,000 gal.

Collector Tank Floor Drains 100% 34,000 gal. 68,000 gal.

Sample Tank Floor Drain 100% 150 gpm 150 gpm Collector Transfer Pump Revision 12 11.2-39 January, 2003

TABLE 11.2-8 (Continued)

Maximum Normal Collecting Collecting Degree of or Processing or Processing Equipment Redundancy Capacity Capacity Floor Drains 100% 200 gpm 200 gpm Sample Pump Floor Drains 100%(1) Varies (150 to Varies (150 to Demineralizer 200 gpm) 200 gpm)

Floor Drains 100%(2) 150 gpm 150 gpm Filter Chemical Drains None 270 gal. 270 gal.

Sump Chemical Drains 100% 25 gpm 50 gpm Sump Pump Chemical Waste 100% 19,650 gal. 39,300 gal.

Tank Waste Evaporator(5)

Chemical Waste 100% 19,100 gal. 38,200 gal.

Distillate Tank Chemical Waste 100% 200 gpm 400 gpm Distillate Pump Laundry and None 500 gal. 500 gal.

Floor Drains Sump Laundry and 100% 50 gpm 100 gpm Floor Drains Sump Pump Detergent Drains 100% 1,550 gal. 3,100 gal.

Tank Detergent Drains(4) 100% 50 gpm 100 gpm Pump Revision 17 11.2-40 October, 2011

TABLE 11.2-8 (Continued)

Maximum Normal Collecting Collecting Degree of or Processing or Processing Equipment Redundancy Capacity Capacity Detergent Drains(4) 100% 50 gpm 100 gpm Filter Chemical Waste 100% 120 gpm 240 gpm Pump NOTES:

(1)

Waste demineralizer and floor drains demineralizer can be cross tied.

(2)

Waste collector filter and floor drains filter can be cross tied.

(3)

Except for annulus floor drain sump pump.

(4)

Detergent drain filters are abandoned in place. Detergent drain pump can only be used to recycle or transfer waste between detergent drain tanks. Detergent drain tank waste is drained and processed via radwaste floor drain system.

(5)

Steam supply to the Waste Evaporators has been permanently severed.

Revision 16 11.2-41 October, 2009

TABLE 11.2-9 EVALUATION OF LIQUID RADWASTE SYSTEM CAPACITY FOR HANDLING LARGE WASTE INPUT VOLUMES(5)

Input Input Total per Frequency Rate, Duration, Occurrence of Waste Input Description (gpm) (Minutes) (Gal./day) Occurrence Disposition of Waste Input

1. High purity (equipment drains) subsystem:
a. Max. normal quantity of 50 to Inter- 33,000 158 days Collect in one waste collector tank miscellaneous equipment 500 mittent per year and process. Total process time is leakage approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per occurrence.
b. Maximum quantity of 50 to Inter- 73,100 14 days Collect in both waste collector miscellaneous equipment 500 mittent per year tanks and process. Total process time leakage (estimated is approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per quantity taken from occurrence.

ANSI N197-1976)(1)

c. Condensate polishing 200 207 41,500 60 days Collect in two waste collector tanks demineralizer rinse per year (or FDCT) and process. Total process during condenser tube time is approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per leak period or plant occurrence.

startup.(3)

d. Reactor blowdown via 300 180 54,000 Rare Collect in both waste collector reactor water cleanup tanks and process. Total process system during startup time is approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per (normally directed to occurrence.

main condenser)(3)

e. Suppression pool drain 1,000 1,000 1,000,000 Once every Collect in one waste collector (for decontamination, (Inter- 10 years tank and one floor drain inspection and mittently) collector tank in 34,000 gallon maintenance of pool)(2)(3) batches as these tanks become available. Total process time is approximately 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> per occurrence.
f. Spent fuel shipping 200 235 47,000 18 days Collect in two waste collector cask pit drawdown(3) per year tanks and process. Total process time is approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
g. RHR flush/test(3) 2,000 20.4 40,800 24 days Collect in two waste collector per year tanks and process. Total process time is approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

Revision 19 11.2-42 October, 2015

TABLE 11.2-9 (Continued)

Input Input Total per Frequency Rate, Duration, Occurrence of Waste Input Description (gpm) (Minutes) (Gal./day) Occurrence Disposition of Waste Input

2. Low purity (floor drains) subsystem:
a. Max. normal quantity of 25 to Inter- 13,000 158 days Collect in one floor drain collector miscellaneous floor 750 mittent per year tank and process. Total process drainage time is approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> per occurrence.
b. Maximum quantity of 25 to Inter- 67,600 14 days Collect in both floor drain miscellaneous drainage 750 mittent per year collector tanks and process. Total (Estimated quantity taken process time is approximately from ANSI N197-1976)(1) 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.
c. Decant from backwash 450 90 39,200 60 days Collect in both floor drains settling tanks for reactor (Inter- per year collector tanks in 8,000 gallon water cleanup filter/ mittently) batches and process. Total process demineralizers and time is approximately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

condensate polishing filters during startup or condenser tube leak period.

3. Chemical waste subsystem:
a. Condensate polishing 200 65 13,000 60 days Collect in one chemical waste tank demineralizer per year and process. Total process time regeneration solutions is approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

during startup(4) or condenser tube leak period.

NOTES:

(1)

The maximum leak rate used here is for the drywell. It is assumed to occur in both drywells simultaneously, even though the probability of this happening is very low. This maximum leak rate could also occur in the containment, turbine power complex, auxiliary bldg., radwaste bldg., control complex, or intermediate bldg. However, the probability of simultaneous leakage in these areas while the maximum leakage rate is assumed in both drywells is extremely low. Since these areas are accessible, it is assumed that repairs could be made quickly enough to avoid such multiple failures.

Revision 16 11.2-43 October, 2009

TABLE 11.2-9 (Continued)

NOTES: (Continued)

(2)

The total volume of water in the suppression pool is approximately 1,000,000 gallons. Since the reactor will be completely shut down while the pool is being inspected, the condenser hotwell can be used to store a portion of this volume (approximately 500,000 gallons). The remaining portion of the pool inventory will be pumped to the LRW system as tankage in this system becomes available to collect and process this waste. (It is assumed that one reactor will still be operating, requiring half of the LRW system processing capacity to be available for handling waste from the operating unit.)

(3)

These inputs can be diverted to the low purity (floor drains) subsystem if processing conditions warrant.

(4)

Condensate demineralizers are no longer regenerated.

(5)

Table 11.2-9, Evaluation of Liquid Radwaste System Capacity for Handling Large Waste Input Volumes, was developed with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because single unit operation will generate less waste input volume than that attained by dual unit operation.

Revision 19 11.2-44 October, 2015

TABLE 11.2-10 PROCESS FLOW DATA FOR LIQUID RADWASTE SYSTEM(10)

Solids/ Normal Stream Stream Normal Maximum Gallons/ Batch Gal./Yr Isotopic Number Description Batches/Day Batches/Day Batch (Pounds) (Both Units) Activity(1) 1-a Drywell Floor Drains 2.82 113.3 255 N/A 525,000 M (each unit) 1-b Containment Floor Drains 2.56 38.5 390 N/A 729,000 M (each unit) 1-c Turbine Building Floor 2.0 2.0 1,000 N/A 1,460,000 S Drains (each unit) 1-d Radwaste Building Floor 2.0 2.0 500 N/A 365,000 R Drains (common) 1-e Auxiliary Building Floor 1/1.72 43.5 345 N/A 146,000 M Drains (each unit) 1-f Heater Bay Floor Drains - 1.0 270 N/A - S (each unit) 1-g Annulus Floor Drains - 1/5.0 50 N/A - (S+M)/2 (each unit) 1-h Intermediate Building 3.2 20.0 500 N/A 584,000 M Floor Drains (common) 2 Decantate from SRW 1.0 - 250 N/A 91,000 S/4 Disposal System 3 RHR Flush/Test (each unit) 1/30 2.0 40,800 N/A 993,000 Negligible 7 Floor Drains Effluent 1/2.65 - 35,000 N/A 4,821,000 to Condenser 8 Floor Drains Effluent - 1/5.3 35,000 N/A 2,410,000
Design Discharge (max.) 9-a Recirc. Pumps & Valves in 8.6 57.8 360 N/A 2,260,000 M Drywell (each unit) Revision 19 11.2-45 October, 2015 TABLE 11.2-10 (Continued) Solids/ Normal Stream Stream Normal Maximum Gallons/ Batch Gal./Yr Isotopic Number Description Batches/Day Batches/Day Batch (Pounds) (Both Units) Activity(1) 9-b Drywell Steam Valves and 8.6 57.8 140 N/A 879,000 S Coolers in Drywell (each unit) 9-c Misc. Pumps, Valves and 9.64 35.86 265 N/A 1,865,000 M RCIC Equip. in Containment (each unit) 9-d Steam Valves in Containment 9.64 35.86 50 N/A 352,000 S (each unit) 9-e RWCU Sample Drains in 9.64 35.86 125 N/A 880,000 M Containment (each unit) 9-f Radwaste Building 1.0 1.0 500 N/A 182,000 R Equipment Drains (common) 9-g Turbine Building 5.76 5.76 1,000 N/A 4,205,000 S Equipment Drains (each unit) 9-h Auxiliary Building 1/11.2 1/3.0 675 N/A 44,000 M x 102 Equipment Drains (each unit) 9-j Intermediate Building 1/10.0 1/5.0 500 N/A 18,000 R Equipment Drains (common) 9-k Control Complex Equipment 1/5.0 1.0 500 N/A 37,000 Negligible Drains (common) 10 Cond. Demin. Rinse (each 1/14.6 1/2.0 41,500 N/A 2,075,000 S/4 unit) 11 Reactor Blowdown via RWCU - Rare - - - - (each unit) 13-g W.D. Effluent Design - 1/4.5 35,000 N/A 2,839,000
Discharge (max.) 13-h W.D. Effluent to Condenser 1.12 - 35,000 N/A 14,308,000 Revision 12 11.2-46 January, 2003 TABLE 11.2-10 (Continued) Solids/ Normal Stream Stream Normal Maximum Gallons/ Batch Gal./Yr Isotopic Number Description Batches/Day Batches/Day Batch (Pounds) (Both Units) Activity(1) 14 Cond. Mixed Bed Demin. 1/14.6 1/2.0 13,000 N/A 650,000 See Note(3) Regeneration Solutions 15 Chemical Drains (each unit) 1.1 1.1 500 N/A 401,000 M/4 24 Radioactive Chemical Waste - 1/36.5 15,360 N/A 154,000
Effluent Design Discharge (max.) 25 Hot Shower and Detergent 3.0 3.0 500 - 547,000 Negligible Drains 27 Detergent Waste Effluent 1.48 2.65 1,600 - 864,000 Negligible 30 Floor Drains Demin. 1/30.5 1/22.35 1,455 1,970 17,000 Buildup on Spent Resins Transfer F.D. Demin. 31 Waste Demineralizer Spent 1/35 1/15.35 1,455 1,970 15,000 Buildup on Waste Resins Transfer Demineralizer 34-a Cond. Demin. Spent Resins 6/3.6 yrs 6/3.6 yrs 9,970 12,090 17,000
to SRW Disposal (both units) Revision 17 11.2-47 October, 2011 TABLE 11.2-10 (Continued) Solids/ Normal Stream Stream Normal Maximum Gallons/ Batch Gal./Yr Isotopic Number Description Batches/Day Batches/Day Batch (Pounds) (Both Units) Activity(1) 34-b W.D., F.D. and S.P.D. 1/83.5 1/31.6 9,980 11,700 44,000
Spent Resins to SRW Disposal 35 Condensate Filter Backwash 1/3.0 8.0 5,200 360 1,265,000 See Note(5) 40 Cond. Filter Sludge to 2/36.0 1/2 7,000 4,350 142,000
SRW Disposal 43 Avg. CBST Decantate 1/1.5 8.0 4,610 N/A 1,122,000 S/6 45 RWCU F/D Backwash (each unit) 1/6.5 1.0 2,400 70 270,000 See Note(7) 48 RWCU F/D Sludge to SRW 2/97.5 5/30 2,150 1,040 16,000
Disposal 51 Avg. RBST Decantate 1/3.25 1.0 2,300 N/A 258,000 M/4 53 Fuel Pool F/D Backwash 1/5.2 1/5.2 2,160 65 152,000 See Note(2) 62 Fuel Pool F/D Sludge to 1/348 1/30 7,000 4,350 7,000
SRW Disposal 65 Decantate from Fuel Pool 1/5.2 1/5.2 2,055 N/A 144,000 Negligible Filter Backwash 81 Cask Pit Drawdown 1/20.3 - 47,000 N/A 845,000 Negligible 84 Suppression Pool - 1/10 yrs 1,000,000 N/A - Negligible Maintenance Drain 86 Cond. Mixed Bed Demin. 6/3.6 yrs 6/3.6 yrs 4,950 5,750 16,000 See Note(4) Spent Resins Transfer 87 Suppression Pool Cleanup 1/30 1/15 1,750 2,365 21,000 Buildup on Demin. Spent Resins Suppression Pool Transfer Cleanup Demin. Revision 16 11.2-48 October, 2009 TABLE 11.2-10 (Continued) NOTES: (1) M = maximum concentration in reactor water. S = maximum concentration in condensate. R = maximum concentration in radwaste sump. (2) Activity for this stream is 1.0 Curie/year, based on operating data from Nine Mile Point Nuclear Station, Unit No. 1. (3) Activity is calculated on basis of 1.38 x 108 gallons of water treated by the condensate demineralizers every 90 days. (4) Activity is calculated on basis of 1.38 x 108 gallons of water treated by the condensate demineralizers every 90 days plus M/4 times the demineralizer backwash volume. (5) Activity is calculated on the basis of the filter buildup per batch (0.98 curies) plus S/6 times the backwash volume. (6) Activity is calculated on the basis of the filter buildup for 8 days at the normal condensate flow rate (6.5 curies), a fill time of 4 days, and a decay time of 2 days. (7) Activity is calculated on the basis of the filter buildup per batch (355 curies) plus M/4 times the backwash volume. (8) Activity is calculated on the basis of the RWCU buildup per batch (355 curies) plus M/4 times the backwash volume, a fill time of 60 days, and a decay time of 60 days. (9) Activity is calculated on the basis of the filter buildup per batch, a fill time of 100 days and a decay time of 100 days. (10) Table 11.2-10, Process Flow Data for Liquid Radwaste System, was calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because single unit operation will generate less waste input volume than that attained by dual unit operation. Revision 19 11.2-49 October, 2015 TABLE 11.2-11 QUALITY REQUIREMENTS FOR CONDENSATE MAKEUP
a. Specific Conductivity at 25C <1.0 µmho/cm
b. pH at 25C 5.3 to 7.5
c. Chloride (as Cl-) <0.05 ppm Revision 12 11.2-50 January, 2003
TABLE 11.2-12 CRITERIA FOR SELECTION OF PROCESS FLOW PATH FOR LIQUID RADWASTE SYSTEM INPUTS Description Selecting Process Subsystem Process Flow Path(1) Flow Path(1) High purity/ a. Collect, Sample and a. Batch is within low conductivity Reuse. the limit for con-densate makeup given in
.
b. Collect, Sample, b. Batch is above Body Feed, Filter, any limit given in Demineralize, and
. Reuse. Conductivity <100 mho/cm
c. Collect, Sample, c. Batch is above any Body Feed, Filter, limit given in Demineralize, and
. Discharge or Reuse. Conductivity >100 mho/cm Medium-to-low a. Collect, Sample, a. Conductivity purity/medium Body Feed, Filter, <100 mho/cm conductivity Demineralize, and Reuse.
b. Collect, Sample, b. Conductivity Body Feed, Filter, >100 mho/cm Demineralize, and Reuse or Discharge.
Detergent a. Collect drain to a. Always Drains floor drains. NOTE: (1) Depending on actual processing conditions, these flow paths and processing criteria may change. Revision 16 11.2-51 October, 2009 TABLE 11.2-13 SIGNIFICANT NUCLIDE ANNUAL RELEASE TO DISCHARGE TUNNEL <10 CFR 20, Annual Release Concentration Appendix B> to Discharge In Plant Effluent Tunnel Discharge MPC Fraction Concentration Fraction of Nuclide (Ci/yr/unit) (Ci/cc) (Ci/cc) of MPC (Ci/cc) Effluent Conc. Na-24 .023 3.9-10 3-5 1.3-5 5E-5 7.8E-6 P-32 .0022 3.7-11 2-5 1.9-6 9E-6 4.1E-6 Cr-51 .057 9.5-10 2-3 4.7-7 5E-4 1.9E-6 Mn-54 .00071 1.2-11 1-4 1.2-7 3E-5 4.0E-7 Mn-56 .00075 1.3-11 1-4 1.3-7 7E-5 1.9E-7 Co-58 .0023 3.9-11 9-5 4.3-7 2E-5 1.9E-6 Co-60 .0047 7.9-11 3-5 2.6-6 3E-6 2.6E-5 Fe-55 .012 2.1-10 2-3 1.0-7 1E-4 2.1E-6 Fe-59 .00035 5.9-12 5-5 1.2-7 1E-5 5.9E-7 Ni-63 .00001 1.6-13 7-4 2.3-10 1E-4 1.6E-9 Cu-64 .06 1.0-9 2-4 5.0-6 2E-4 5.0E-6 Zn-65 .01221 2.0-10 1-4 2.0-6 5E-6 4.0E-5 Zn-69m .0045 7.6-11 6-5 1.3-6 6E-5 1.3E-6 Zn-69 .0048 8.0-11 2-3 4.0-8 8E-4 1.0E-7 Np-239 .053 8.9-10 1-4 8.9-6 2E-5 4.4E-5 Br-83 .00003 5.0-13 3-6 1.7-7 9E-4 5.5E-10 Sr-89 .0012 2.1-11 3-6 6.9-6 8E-6 2.6E-6 Sr-90 .00007 1.2-12 3-7 3.9-6 5E-7 2.4E-6 Y-90 .00002 3.4-13 2-5 1.7-8 7E-6 4.9E-8 Sr-91 .0048 8.0-11 5-5 1.6-6 2E-5 4.0E-6 Y-91m .0031 5.2-11 3-3 1.7-8 2E-3 2.6E-8 Y-91 .00074 1.2-11 3-5 4.2-7 8E-6 1.5E-6 Y-92 .0019 3.1-11 6-5 5.2-7 4E-5 7.7E-7 Revision 19 11.2-52 October, 2015 TABLE 11.2-13 (Continued) <10 CFR 20, Annual Release Concentration Appendix B> to Discharge In Plant Effluent Tunnel Discharge MPC Fraction Concentration Fraction of Nuclide (Ci/yr/unit) (µCi/cc) (µCi/cc) of MPC (µCi/cc) Effluent Conc. Y-93 .0054 9.0-11 3-5 3.0-6 2E-5 4.5E-6 Zr-95 .00008 1.3-12 6-5 2.2-8 2E-5 6.5E-8 Nb-95 .00008 1.3-12 1-4 1.3-8 3E-5 4.3E-8 Zr-97 .00001 1.6-13 2-5 8.2-9 9E-6 1.8E-8 Nb-97 .00001 1.6-13 9-4 1.8-10 3E-4 5.3E-10 Mo-99 .016 2.7-10 4-5 6.7-6 2E-5 1.3E-5 Tc-99m .023 3.9-10 3-3 1.3-7 1E-3 3.9E-7 Ru-103 .00023 3.9-12 8-5 4.8-8 3E-5 1.3E-7 Rh-103m .00023 3.9-12 1-2 3.9-10 6E-3 6.5E-10 Ru-105 .00031 5.2-12 1-4 5.2-8 7E-5 7.4E-8 Rh-105 .0016 2.7-11 1-4 2.7-7 5E-5 5.4E-7 Ru-106 .00003 5.0-13 1-5 5.0-8 3E-6 1.7E-7 Rh-106 .00003 5.0-13 3-6 1.7-7 1E-4 5.0E-9 Ag-110m .00001 1.6-13 3-5 5.4-9 6E-6 2.7E-8 W-187 .0013 2.2-11 6-5 3.7-7 3E-5 7.3E-7 Te-129m .00046 7.7-12 2-5 3.9-7 7E-6 1.1E-6 Te-129 .00029 4.9-12 8-4 6.1-9 4E-4 1.2E-8 Te-131m .00053 8.9-12 4-5 2.2-7 8E-6 1.1E-6 I-131 .06 1.0-9 3-7 3.4-3 1E-6 1.0E-3 Te-132 .00009 1.5-12 2-5 7.4-8 9E-6 1.7E-7 I-132 .00035 5.9-12 8-6 7.4-7 1E-4 5.9E-8 I-133 .083 1.4-9 1-6 1.4-3 7E-6 2.0E-4 Cs-134 .0077 1.3-10 9-6 1.4-5 9E-7 1.4E-4 I-135 .012 2.1-10 4-6 5.2-5 3E-5 7.0E-6 Cs-136 .0047 7.9-11 6-5 1.3-6 6E-6 1.3E-5 Cs-137 .018 3.0-10 2-5 1.5-5 1E-6 3.0E-4 Ba-137m .017 2.8-10 3-6 9.4-5 NA NA Ba-140 .0043 7.3-11 2-5 3.6-6 8E-6 9.1E-6 Revision 19 11.2-53 October, 2015 TABLE 11.2-13 (Continued) <10 CFR 20, Annual Release Concentration Appendix B> to Discharge In Plant Effluent Tunnel Discharge MPC Fraction Concentration Fraction of Nuclide (Ci/yr/unit) (µCi/cc) (µCi/cc) of MPC (µCi/cc) Effluent Conc. La-140 .002 3.4-11 2-5 1.7-6 9E-6 3.8E-6 La-141 .00007 1.2-12 3-6 3.9-7 5E-5 2.4E-8 Ce-141 .00039 6.5-12 9-5 7.3-8 3E-5 2.2E-7 Ce-143 .00017 2.8-12 4-5 7.0-8 2E-5 1.4E-7 Pr-143 .00045 7.6-12 5-5 1.5-7 2E-5 3.8E-7 Ce-144 .00003 5.0-13 1-5 5.0-8 3E-6 1.7E-7 Pr-144 .00003 5.0-13 3-6 1.7-7 6E-4 8.3E-10 Nd-147 .00003 5.0-13 6-5 8.4-9 2E-5 2.5E-8 All others .00001 1.6-13 - - - - (except H-3) Total .5098 8.3-9 - 5.0-3 - 1.8E-3 (except H-3) H-3 47 7.9-7 3-3 2.6-4 1E-3 7.9E-4 NA - Not Applicable Revision 12 11.2-54 January, 2003 TABLE 11.2-14 ANNUAL RELEASE BY STREAM TO DISCHARGE TUNNEL(1) Releases to Discharge Tunnel Adjusted High Purity Low Purity Chemical Total LWS Total Total Nuclide (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) Na24 0.01122 0.00468 See Note(1) 0.01589 0.02273 0.02300 P32 0.00108 0.00045 See Note(1) 0.00153 0.00219 0.00220 Cr51 0.02804 0.01169 0.00002 0.03975 0.05684 0.05700 Mn54 0.00035 0.00015 See Note(1) 0.00049 0.00071 0.00071 Mn56 0.00037 0.00015 See Note(1) 0.00053 0.00075 0.00075 Fe55 0.00582 0.00243 0.00001 0.00825 0.01180 0.01200 Fe59 0.00017 0.00007 See Note(1) 0.00024 0.00035 0.00035 Co58 0.00115 0.00048 See Note(1) 0.00163 0.00233 0.00230 Co60 0.00233 0.00097 See Note(1) 0.00330 0.00472 0.00470 Ni63 0.00001 See Note(1) See Note(1) 0.00001 0.00001 0.00001 Cu64 0.02949 0.01230 See Note(1) 0.04178 0.05975 0.06000 Zn65 0.00605 0.00262 See Note(1) 0.00867 0.01221 0.01221 Zn69m 0.00221 0.00092 See Note(1) 0.00313 0.00447 0.00450 Zn69 0.00237 0.00099 See Note(1) 0.00336 0.00481 0.00480 W187 0.00064 0.00027 See Note(1) 0.00091 0.00130 0.00130 Np239 0.02630 0.01096 See Note(1) 0.03726 0.05328 0.05300 Br83 0.00002 0.00001 See Note(1) 0.00002 0.00003 0.00003 Sr89 0.00058 0.00024 See Note(1) 0.00082 0.00117 0.00120 Sr90 0.00003 0.00001 See Note(1) 0.00005 0.00007 0.00007 Y90 0.00001 See Note(1) See Note(1) 0.00002 0.00002 0.00002 Sr91 0.00239 0.00099 See Note(1) 0.00338 0.00483 0.00480 Y91m 0.00154 0.00064 See Note(1) 0.00218 0.00312 0.00310 Y91 0.00037 0.00015 See Note(1) 0.00052 0.00074 0.00074 Sr92 0.00010 0.00004 See Note(1) 0.00014 0.00019 0.00019 Y92 0.00091 0.00038 See Note(1) 0.00130 0.00185 0.00190 Y93 0.00265 0.00110 See Note(1) 0.00375 0.00536 0.00540 Revision 12 11.2-55 January, 2003 TABLE 11.2-14 (Continued) Releases to Discharge Tunnel Adjusted High Purity Low Purity Chemical Total LWS Total Total Nuclide (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) Zr95 0.00004 0.00002 See Note(1) 0.00006 0.00008 0.00008 Nb95 0.00004 0.00002 See Note(1) 0.00006 0.00008 0.00008 Zr97 0.00001 See Note(1) See Note(1) 0.00001 0.00001 0.00001 Nb97m 0.00001 See Note(1) See Note(1) 0.00001 0.00001 0.00001 Nb97 0.00001 See Note(1) See Note(1) 0.00001 0.00001 0.00001 Mo99 0.00804 0.00335 See Note(1) 0.01140 0.01629 0.01600 Tc99m 0.01116 0.00465 See Note(1) 0.01581 0.02261 0.02300 Ru103 0.00011 0.00005 See Note(1) 0.00016 0.00023 0.00023 Rh103m 0.00011 0.00005 See Note(1) 0.00016 0.00023 0.00023 Ru105 0.00015 0.00006 See Note(1) 0.00021 0.00031 0.00031 Rh105m 0.00015 0.00006 See Note(1) 0.00021 0.00031 0.00031 Rh105 0.00077 0.00032 See Note(1) 0.00110 0.00157 0.00160 Ru106 0.00002 0.00001 See Note(1) 0.00002 0.00004 0.00003 Rh106 0.00002 0.00001 See Note(1) 0.00002 0.00004 0.00003 Ag110m 0.00001 See Note(1) See Note(1) 0.00001 0.00001 0.00001 Te129m 0.00023 0.00009 See Note(1) 0.00032 0.00046 0.00046 Te129 0.00014 0.00006 See Note(1) 0.00021 0.00029 0.00029 Te131m 0.00026 0.00011 See Note(1) 0.00037 0.00053 0.00053 Te131 0.00005 0.00002 See Note(1) 0.00007 0.00010 0.00010 I131 0.02786 0.01162 0.00214 0.04162 0.05951 0.06000 Te132 0.00004 0.00002 See Note(1) 0.00006 0.00009 0.00009 I132 0.00018 0.00007 See Note(1) 0.00025 0.00035 0.00035 I133 0.04119 0.01717 0.00001 0.05838 0.08348 0.08300 Cs134 0.00174 0.00364 0.00001 0.00539 0.00771 0.00770 I135 0.00573 0.00239 See Note(1) 0.00812 0.01161 0.01200 Cs136 0.00107 0.00224 See Note(1) 0.00331 0.00474 0.00470 Cs137 0.00408 0.00850 0.00001 0.01259 0.01800 0.01800 Revision 12 11.2-56 January, 2003 TABLE 11.2-14 (Continued) Releases to Discharge Tunnel Adjusted High Purity Low Purity Chemical Total LWS Total Total Nuclide (Curies) (Curies) (Curies) (Curies) (Ci/yr) (Ci/yr) Ba137m 0.00381 0.00795 0.00001 0.01177 0.01683 0.01700 Ba140 0.00215 0.00089 See Note(1) 0.00304 0.00435 0.00430 La140 0.00100 0.00042 See Note(1) 0.00142 0.00203 0.00200 La141 0.00004 0.00001 See Note(1) 0.00005 0.00007 0.00007 Ce141 0.00019 0.00008 See Note(1) 0.00027 0.00039 0.00039 Ce143 0.00008 0.00003 See Note(1) 0.00012 0.00017 0.00017 Pr143 0.00022 0.00009 See Note(1) 0.00032 0.00045 0.00045 Ce144 0.00002 0.00001 See Note(1) 0.00002 0.00004 0.00003 Pr144 0.00002 0.00001 See Note(1) 0.00002 0.00004 0.00003 Nd147 0.00002 0.00001 See Note(1) 0.00002 0.00003 0.00003 All Others 0.00001 See Note(1) See Note(1) 0.00001 0.00001 0.00001 Total (Except 0.23694 0.11675 0.00224 0.35593 0.50876 0.50981 Tritium) Tritium 47 Curies per year Release NOTE: (1) Less than .00001 Ci. Revision 12 11.2-57 January, 2003 TABLE 11.2-15 INPUT PARAMETERS FOR CALCULATING LIQUID RELEASES (GALE) MAXIMUM CORE THERMAL POWER - 3758 MWt REACTOR COOLANT CLEANUP SYSTEM Average flow rate - 1.54 x 105 lb/hr Demineralizer type - powdered resin CONDENSATE DEMINERALIZERS Average flow rate - 10.5 x 106 lb/hr Demineralizer type - deep bed Number and size (ft3) of demineralizers - six condensate demineralizers each containing 260 cubic feet of mixed resin Regeneration frequency - 3.5 days per demineralizer for a total regeneration time of 21 days(5) Regenerant volume - 12,000 gallons/batch(5) Revision 16 11.2-58 October, 2009 TABLE 11.2-15 (Continued) LIQUID WASTE PROCESSING SYSTEMS Holdup Fraction of Times Flow(1) Primary Collection/ Fraction Rates Coolant Discharge Assumed Name Sources (gpd) Activity (days) Discharge High Equipment Drains Purity Drywell 4300 1.0 1.7/.65 0.1 Waste Containment 2550 .01 1.7/.65 0.1 Radwaste Building 500 .01 1.7/.65 0.1 Turbine Building 5760 .01 1.7/.65 0.1 Auxiliary Building 60 .01 1.7/.65 0.1 Intermediate Building 25 .01 1.7/.65 0.1 Control Complex 50 Negligible 1.7/.65 0.1 Drywell and Containment Steam Valves 1685 .01 1.7/.65 0.1 Cond. Demin. Rinse 1230 .002 1.7/.65 0.1 RHR Flush/Test 340 Negligible 1.7/.65 0.1 Low Floor Drains Purity Drywell 720 1.0 1.7/.65 0.25 Waste Containment 1000 .01 1.7/.65 0.25 Turbine Building 2000 .01 1.7/.65 0.25 Radwaste Building 500 .01 1.7/.65 0.25 Auxiliary Building 200 .01 1.7/.65 0.25 Intermediate Building 800 .01 1.7/.65 0.25 Decantate 2210 .002 1.7/.65 0.25 Chemical Chemical Drains 275 .02 6.1/.37(2) 0.1 Waste Regene- Cond. Mixed Bed rant Demin. Reg. Sol. 820 (3) 6.1/.37(2) 0.1 Waste(5) Revision 16 11.2-59 October, 2009 TABLE 11.2-15 (Continued) Decontamination Factors Halogens/Cs, Rb/Other Name Component Capacity Nuclides High Waste Collector Tank 35,000 gallons N/A Purity Waste Sample Tank 35,000 gallons N/A Waste Waste Collector Filter 144,000 gpd 1/1/1 Waste Demineralizer 288,000 gpd 102/10/102 Low Floor Drains Collector Purity Tank 35,000 gallons N/A Waste Floor Drains Sample Tank 35,000 gallons N/A Floor Drains Filter 144,000 gpd 1/1/1 Floor Drains Demineralizer 288,000 gpd 102/2/102 Chemical Chemical Waste Tank 20,000 gallons N/A Waste Chemical Waste Distillate Tank 20,000 gallons N/A Waste or Floor Drains Demineralizer 288,000 gpd 102/2/102 Regenerant(4)(5) Waste NOTES: (1) Values based on one-half of the total flow for two units. (2) Collection time is based on total flow for chemical waste and regenerant waste since they utilize a common tank. (3) Value calculated internally in BWR-GALE Code. (4) Part of chemical waste system. (5) Waste no longer generated. Revision 17 11.2-60 October, 2011 TABLE 11.2-16 TANKS LOCATED OUTSIDE THE CONTAINMENT WHICH CONTAIN POTENTIALLY RADIOACTIVE FLUID Tank Level(2) High Level(2) Tank Quantity Location Monitoring Annunication Overflow Control(3) Waste Mixing 2 Radwaste Bldg. SRW control panel SWR and LRW Local Level switch Dewatering tanks control panels discontinues flow into (G51-A001 A+B) tanks on high level Fuel Pool Surge 2 Intermediate Control Room Control Room CRW Tank Bldg. RWCU Filter/Demin 2 Radwaste Bldg. RWBCR RWBCR DRW Backwash settling tank Floor Drain Collector 2 Radwaste Bldg. RWBCR RWBCR DRW Tanks Waste Collector Tanks 2 Radwaste Bldg. RWBCR RWBCR CRW LRW Filter Aid Tank 1 Radwaste Bldg. RWBCR LRW Control Panel(4) DRW TBF Control Panel TBF Control Panel LRW Precoat Tank 1 Radwaste Bldg. TBF Control Panel LRW Control Panel(4) DRW TBF Control Panel Floor Drains Filtrate 1 Radwaste Bldg. TBF Control Panel LRW Control Panel(4) DRW Tank TBF Control Panel Waste Collector Filtrate 1 Radwaste Bldg. TBF Control Panel LWR Control Panel(4) Waste Collector Tank Tank TBF Control Panel (CRW) Backwash Rinse Receiving 1 Turbine Bldg. Local Panel Control Room DRW Tank Regen. Chemical Receiving 1 Turbine Bldg. Local Panel Control room DRW Tank Cond. Demin. Hot 1 Turbine Bldg. None None Closed System Water Tank Revision 19 11.2-61 October, 2015 TABLE 11.2-16 (Continued) Tank Level(2) High Level(2) Tank Quantity Location Monitoring Annunciation Overflow Control(3) Moisture Separator 4 Turbine Bldg. Local Panel display monitor Bypass to H.P. Drain Tank Condenser 1st Stage Drain 4 Turbine Bldg. None Control Room Bypass to H.P. Tank & display monitor Condenser 2nd Stage Drain 4 Turbine Bldg. None Control Room Bypass to H.P. Tank & display monitor Condenser Condensate Storage 1 Yard Control Room display monitor Catch Basin Tank Fuel Transfer Tube 1 Intermediate None Control Room Contents Pumped to Drain Tank Bldg. Surge Tank on Hi Level Floor Drain Sample 2 Radwaste Bldg. RWBCR RWBCR DRW Tanks Waste Sample Tanks 2 Radwaste Bldg. RWBCR RWBCR CRW Chemical Waste Distillate 2 Radwaste Bldg. RWBCR RWBCR display monitor Tanks Detergent Drain Tanks 2 Radwaste Bldg. RWBCR RWBCR CWT B Chemical Waste Tanks 2 Radwaste Bldg. RWBCR RWBCR DRW Concentrated Waste Tanks 2 Radwaste Bldg. RWBCR None CWT B Spent Resin Tanks 2 Radwaste Bldg. RWBCR RWBCR DRW Condenstate Filter 1 Turbine Bldg. RWBCR RWBCR DRW Backwash Receiving Tanks Condensate Filter 2 Radwaste Bldg. RWBCR RWBCR DRW Backwash Settling Tanks Revision 20 11.2-62 October, 2017 TABLE 11.2-16 (Continued) Tank Level(2) High Level(2) Tank Quantity Location Monitoring Annunication Overflow Control(3) Fuel Pool Filter 1 Intermediate Local Panel RWBCR DRW Demineralizer Bldg. RWBCR Backwash Receiving Tank Fuel Pool F/D 2 Radwaste Bldg. RWBCR RWBCR DRW Backwash Settling Tanks Condensate Return 2 Radwaste Bldg. On Tanks None Start 2nd drain pump Tanks Blowdown Tank 1 Auxiliary Bldg. None None DRW Deaerator 1 Auxiliary Bldg. On Tank - Local Local Panel Directed to Blowdown Panel Tank RF Pumps Seal 1 Heater Bay Local Panel Control Room CRW Leakoff Drain Tanks Precoat Slurry Tank 1 Turbine Bldg. None None Close inlet valve on Hi Signal (DRW) Mix and Hold Tank 1 Turbine Bldg. None None Closed System Anion Regeneration Tank 1 Turbine Bldg. None None Closed System Cation Regen. Tank 1 Turbine Bldg. None None Closed System NOTES: (1) (Deleted) (2) Local panels are located in the same area (building) as the tank unless specified otherwise. (3) CRW and DRW indicate clean and dirty radwaste collection systems. (4) High Level Alarm located on LRW Traveling Belt Filter Control Panel 0H51P0133. General trouble alarm for the TBF Panel can be found in the LRW Control Room. Revision 19 11.2-63 October, 2015 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 11.3.1 DESIGN BASES 11.3.1.1 Design Objective The objective of the gaseous waste management system is to process and control the release of gaseous radioactive effluents to the site environs to maintain as low as reasonably achievable, the exposure of persons in unrestricted areas to radioactive gaseous effluents to <10 CFR 50, Appendix I>. This is to be accomplished while maintaining occupational exposure as low as reasonably achievable and without limiting plant operation or availability. 11.3.1.2 Design Criteria The gaseous effluent treatment 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 established in the Offsite Dose Calculation Manual. In addition, the Offgas Treatment System limits the dose to offsite persons from a Control Rod Drop Accident <Section 15.4.9> to significantly less than the limits specified in <10 CFR 50.67>. As a design basis for this system, an annual average noble radiogas source term (based on 30 minute decay) of 100,000 Ci/sec of the 1971 Mixture will be used.
indicates the design basis noble radiogas source terms referenced to 30 minute decay with the charcoal temperature at 0F.
,
, and
indicate source terms referenced to 30 minute decay with the charcoal temperature at temperatures, 20F, 40F and 70F, respectively. Revision 19 11.3-1 October, 2015 The annual average exposure at the site boundary during normal operation from gaseous sources is not expected to exceed the dose objectives of to <10 CFR 50, Appendix I> in terms of actual doses to actual persons. The radiation dose design basis for the treated offgas is to delay the gas until the required fraction of the radionuclides has decayed and the daughter products are retained by the charcoal and the HEPA filters. The gaseous radwaste equipment is selected, arranged and shielded to maintain occupational exposure as low as reasonably achievable in accordance with <Regulatory Guide 8.8>, and <10 CFR 20>. The gaseous effluent treatment system is designed to the requirements of the General Design Criteria that follow. General Design Criterion 60 The system has sufficient capacity to reduce the offgas activity to permissible levels for release during normal operation, including anticipated operational occurrences, and to avoid any termination of releases or limitation of plant operation due to unfavorable site environmental conditions. General Design Criterion 64 Continuous monitoring of activity levels in the system upstream of the delay line provides advance notice of any potentially significant increase in releases. Continuous monitoring of the system effluent, with automatic isolation at activity levels corresponding to administrative release limits and annunciation at lower levels, along with continuous monitoring of the plant vent release, provide assurance that activity releases to the environment will in all events be maintained within established limits. Revision 12 11.3-2 January, 2003 11.3.1.3 Equipment Design Criteria A list of the offgas system major equipment items which includes materials, rates, process conditions, number of units supplied, and the design codes is provided in
. These equipment items are designed such that they provide no ignition source. Equipment and piping are also designed and constructed in accordance with the requirements of the applicable codes as given in
and
. The quality group classifications of the various systems are shown in
. Seismic category, safety class, quality assurance requirements, and principal construction codes information is contained in <Section 3.2>. The system is designed to Quality Group Classification D, with additional quality requirements as recommended in <Regulatory Guide 1.143>. The failure of the offgas system is analyzed in <Section 15.7.1>. The related failure of the steam jet air ejector lines and the gland seal offgas lines are also analyzed in <Section 15.7.1>. The reactor building, turbine building and radwaste building contain radioactive sources. The design bases for the ventilation systems for these buildings are discussed in <Section 9.4>. 11.3.2 SYSTEM DESCRIPTION The offgas from the main condensersteam jet air ejector is treated by a system using catalytic recombination and low temperature charcoal adsorption (RECHAR system). Descriptions of the major process components including design temperature and pressure are given in
and in the sections that follow. Revision 12 11.3-3 January, 2003 11.3.2.1 Main CondenserSteam Jet Air Ejector Low-Temp RECHAR System Noncondensible radioactive offgas is continuously removed from the main condenser by the air ejector during plant operation. The air ejector offgas will normally contain activation gases, principally N-16, O-19 and N-13. The N-16 and O-19 have short half-lives and are readily decayed. The 10 minute half-life N-13 is present in small amounts that are further reduced by decay. The air ejector offgas will also contain radioactive noble gases including parents of biologically significant Sr-89, Sr-90, Ba-140, and Cs-137. The concentration of these noble gases depends on the amount of tramp uranium in the coolant and on the cladding surfaces (usually extremely small) and the number and size of fuel cladding leaks. 11.3.2.1.1 Process Description A main condenser offgas treatment system has been incorporated in the plant design to reduce the gaseous radwaste emission from the station. The offgas system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. After cooling (to approximately 130F) to strip the condensibles and reduce the volume, the remaining noncondensibles (principally air with traces of krypton and xenon) will be delayed in the nominal 10 minute holdup system. The gas is cooled to 45F and filtered through a HEPA filter. The gas is then passed through a desiccant dryer that reduces the dewpoint between 0F and -40F and is then further chilled prior to entering the charcoal adsorption beds. Charcoal adsorption beds, normally operating in a refrigerated vault between 40F and 0F, selectively adsorb and delay the xenons and kryptons from the bulk carrier gas (principally dry air). After the delay, the gas is again passed through a HEPA filter and discharged to the environment through the offgas building vent. Revision 19 11.3-4 October, 2015 11.3.2.1.1.1 Process Flow Diagram <Figure 11.3-1> is the process flow diagram for the system. The process data for startup and normal operating conditions are on <Figure 11.3-1 (2)>. Information supporting the process data is presented in (Reference 1). The vent is the single release point for this system and is located on the offgas building. The vent is indicated on <Figure 11.3-1> and <Figure 11.3-2>. 11.3.2.1.2 Noble Gas Radionuclide Source Term and Decay The design basis isotopic source terms for the annual average activity input of the main condenser offgas treatment system are given in
at t = 30 minutes. The system is mechanically capable of processing three times the source terms of
without affecting delay time of the noble gases. 11.3.2.1.3 Piping and Instrumentation Diagram (P&ID) The P&ID is shown in <Figure 11.3-2>. The main process routing is indicated by a heavy line. 11.3.2.1.4 Recombiner Sizing The basis for sizing the recombiner is to maintain the hydrogen concentration by volume, below 4 percent (including steam) at the inlet and below 1 percent at the outlet on a dry basis. The exit hydrogen concentration is normally well below the 1 percent maximum allowed. The hydrogen generation rate of the reactor is based on data from nine BWRs. The hydrogen generation rate is given in the process flow diagram, <Figure 11.3-1 (2)>. Revision 12 11.3-5 January, 2003 11.3.2.1.5 Process Design Parameters The Kr and Xe holdup time is closely approximated by the following equation: KDM T V where: T = holdup time of a given gas KD = dynamic adsorption coefficient for the given gas M = weight of charcoal V = flow rate of the carrier gas in consistent units. Dynamic adsorption coefficient values for xenon and krypton were reported by Browning (Reference 2). General Electric has performed pilot plant tests at their Vallecitos Laboratory and the results were reported at the 12th AEC Air Cleaning Conference (Reference 3). Moisture has a detrimental effect on adsorption coefficients. The fully redundant -90F dewpoint, adsorbent air dryers are supplied to prevent moisture from reaching the charcoal. There are redundant moisture analyzers that will alarm on breakthrough of the drier beds; however, breakthrough is not expected since the drier beds will be regenerated on a time basis. The system is slightly pressurized which, together with very stringent leak rate requirements, prevents leakage of moist air into the charcoal. Carrier gas is the air inleakage from the main condenser after the radiolytic hydrogen and oxygen are removed by the recombiner. The air inleakage is nominally sized at 30 scfm total. The Sixth Edition of Heat Exchange Institute Standards for Steam Surface Condensers (Reference 4), Par. Sl(c) (2), indicates that with certain conditions of stable operation and suitable construction, Revision 22 11.3-6 October, 2021 noncondensibles (not including radiological decomposition products) should not exceed 6 scfm for large condensers. Dresden 2, Monticello, Fukushima 1, Tsuruga, and KRB have all operated at 6 scfm or below after initial startup. Dilution air is not added to the system unless the air inleakage is less than 6 scfm. In that event, 6 scfm is added to provide for dilution of residual hydrogen from the recombiner. An initial bleed of oil-free air is added on startup until the recombiner comes up to temperature. 11.3.2.1.6 Charcoal Adsorbers 11.3.2.1.6.1 Charcoal Temperature The charcoal adsorbers normally operate between 40F and 0F. 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 are maintained at a constant temperature by a redundant vault refrigeration system. Failure of the refrigeration system will actuate an alarm in the control room. In addition, a radiation monitor is provided to monitor the radiation level in the charcoal bed vault. High radiation will actuate an alarm in the control room. Limited operations of the charcoal adsorbers above 40F may occur during maintenance activities on the refrigeration system, provided gaseous effluents are verified to be within the limits of the Offsite Dose Calculation Manual. The following analysis was calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because single unit data is less than that attained by dual unit operation. Revision 19 11.3-7 October, 2015 When Unit 1 and Unit 2 are operating simultaneously, temperatures of the refrigeration system must be reduced to +20F, which will comply with the concluding statement of the regulatory staff to <10 CFR 50, Appendix I>. The calculated annual air dose at the reactors site, will not exceed 10 mrads for gamma, and 20 mrads for beta. Revision 19 11.3-7a October, 2015 11.3.2.1.6.2 Gas Channeling in the Charcoal Adsorber 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 channeling or wall effects may reduce efficiency of the holdup bed if this ratio is not greater than 12 (Reference 5). During transfer of the charcoal into the charcoal 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. 11.3.2.1.6.3 Charcoal Bypass Mode A bypass line, isolated with double block and bleed valves, is provided to bypass the charcoal adsorbers. The main purpose of this bypass is to protect the charcoal during preoperation 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 normal operation unless some unforeseen system malfunction would necessitate shutting down the power plant or operating in the bypass mode and remaining within limits. The activity release is controlled by a process monitor upstream of the vent isolation valve that will cause the bypass valve to close on a high radiation alarm. This interlock can be defeated only by a keylock switch. The alarm setting is set below the Offsite Dose Calculation Manual limit. In addition, there is a high high-high alarm on the same monitor that will cause the offgas system to be isolated from the vent if established release limits are exceeded. Revision 12 11.3-8 January, 2003 11.3.2.1.7 Leakage of Radioactive Gases Leakage of radioactive gases from the system is limited by welding piping connections where possible and using bellows stem seals or equivalent 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. 11.3.2.1.8 Hydrogen Concentration Hydrogen concentration or 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. Furthermore, hydrogen concentration is monitored by the redundant hydrogen analyzers prior to the holdup process. If the hydrogen concentration exceeds 2 percent, an alarm is set to annunciate in the control room. 11.3.2.1.9 Field Run Piping No piping in this system is field routed. This includes major process piping, drain lines, steam lines, and sample lines which are shown on <Figure 11.3-2>. 11.3.2.1.10 Liquid Seals There are several liquid loop seals to prevent gas escape through drains shown on <Figure 11.3-2>. These seals are protected against loss of liquid by automatic shutoff valves downstream of the loop seals. If liquid level drops due to leakage in the cooler condenser, prefilter, and holdup line loop seals, the level sensors will initiate closure of Revision 12 11.3-9 January, 2003 these automatic shutoff valves which will mitigate the consequences of the loss of loop seals, and provide a Control Room low level alarm. Each seal has a manual valve that can be used to fill the loop. Seals are also equipped with solenoid valves that close if radioactive release from this system exceeds established limits. 11.3.2.1.11 System Performance Noble gas activity release is about 28-510 Ci/sec from the exit of the steam jet air ejector offgas system based on a nominal 30 scfm air inleakage and an input of 100,000 Ci of 30 minute old 1971 Mixture. The isotopic composition is given in
,
,
, and
in units of Ci/sec and Ci/yr for charcoal temperatures of 0F, 20F, 40F, and 70F. Iodine input into the offgas system 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 charcoal filters remove 99.9 percent of the iodine in 2 inches of charcoal, whereas this system has approximately 74 feet of charcoal in the flow path. Particulates are removed with a 99.95 percent efficiency by a HEPA filter as gas exits the nominal 10 minute holdup. The noble gas decays within the interstices of the activated charcoal and daughters are entrapped there. The charcoal serves as an excellent filter for other particulates and essentially no particulates exit from the charcoal. The charcoal is followed with a HEPA filter which is a safeguard against escape of charcoal dust. Particulate activity discharged from this system is essentially zero. Revision 22 11.3-10 October, 2021 11.3.2.1.12 Isotopic Inventory The isotopic inventory of each equipment piece is given in
for the Design Basis and Normal (Realistic) operating conditions. 11.3.2.1.13 Previous Experience Performance of a similar system operating at ambient temperatures and the results of experimental testing performed by General Electric have been submitted in the General Electric Company proprietary topical report, (Reference 1). Non-proprietary portions of this information are reported in (Reference 3). 11.3.2.1.14 Single Failures and Operator Errors Design provisions are incorporated which preclude the uncontrolled release of radioactivity to the environment as a result of any single operator error or of any single equipment malfunction short of the catastrophic equipment failures described in <Chapter 15>. An analysis of single equipment piece malfunctions is provided in
. Design precautions taken to prevent uncontrolled releases of activity include the following:
a. The system design minimizes ignition sources so that a hydrogen detonation is highly unlikely even in the event of a recombiner failure.
b. Even though measures are taken to avoid a possible detonation, the system pressure boundary is designed to be detonation-resistant.
Revision 12 11.3-11 January, 2003
c. All discharge paths to the environment are monitored.
d. Dilution steam flow to the steam jet air ejector is monitored and alarmed, and valving is such that loss of dilution steam cannot occur without coincident loss of motive steam, so that the process gas is sufficiently diluted if it is flowing at all.
11.3.2.1.15 Other Radioactive Gas Sources Radioactive gases are present in the power plant buildings as a result of process leakage and steam discharges. The process leakage is the source of the radioactive gases in the air discharged through the ventilation system. The design of the ventilation system is discussed in <Section 9.4>. The building volumes and ventilation flow rates are discussed in <Section 12.2.2>. The activity released to the suppression pool from steam discharges is discussed in <Section 12.2.2>. A tabulation of the expected frequency and the quantity of steam discharged to the suppression pool is provided in
. 11.3.2.1.16 Cost-Benefit Ratio In accordance with <10 CFR 50, Appendix I>,Section II, Paragraph D, a cost-benefit analysis is not required for this system because it satisfies the Guides on Design Objectives for Light-Water-Cooled Nuclear Power Reactors proposed in the Concluding Statement of Position of the Regulatory Staff in Docket RM-50-2. Revision 12 11.3-12 January, 2003 11.3.2.1.17 Maintainability of Gaseous Radwaste System Design features which reduce or ease required maintenance include the following:
a. Redundant components for all active, in-process equipment pieces.
b. No rotating equipment in the process stream.
c. Rotating equipment is located in the system only where maintenance can be performed while the system is in operation.
Design features which reduce leakage and releases of radioactive material include the following:
a. Extremely stringent leak rate requirements placed upon all equipment, piping and instruments, and enforced by system integrity testing as discussed in <Section 11.3.2.2.1.7>.
b. Use of welded joints wherever practicable.
c. Specification of valve types with extremely low leak rate characteristics, i.e., bellows seal, double stem seal or equal.
d. Use of loop seals with automatic shutoff valves to prevent loss of liquid due to siphoning.
e. Specification of stringent seat leakage characteristics for valves and lines discharging to the environment through other systems.
Revision 12 11.3-13 January, 2003 11.3.2.2 System Design Description 11.3.2.2.1 Main CondenserSteam Jet Air Ejector Offgas Low-Temp System 11.3.2.2.1.1 Quality Classification and Construction and Testing Requirements Equipment and piping are designed and constructed in accordance with the requirements of the applicable codes as given in
and will comply with the welding and material requirements and the system construction and testing requirements as follows. 11.3.2.2.1.2 Seismic Design 11.3.2.2.1.2.1 Equipment Equipment and components used to collect, process or store gaseous radioactive waste are not designed as Seismic Category I. 11.3.2.2.1.2.2 Buildings Housing Gaseous Radioactive Waste Processing Systems That portion of the offgas system upstream of the gas dryer prefilters is housed in the turbine building, which is a non-seismic, nonsafety class, reinforced concrete structure. The remaining portion of this system is located in the offgas building, which is a Safety Class 3, Seismic Category I, reinforced concrete structure. A detailed discussion of the seismic design for this building is given in <Section 3.7>. Revision 12 11.3-14 January, 2003 11.3.2.2.1.3 Quality Control A program is established that is sufficient to assure that the design, construction and testing requirements are met. The following areas are included in the program:
a. Design and Procurement Document Control - Procedures are established to ensure that requirements are specified and included in design and procurement documents and that deviations from these documents are controlled.
b. Control of Purchased Material, Equipment and Services - Procedures are established to ensure that purchased material, equipment and construction services conform to the procurement documents.
c. Inspection - A program for inspection of activities affecting quality is established and executed by or for the organization performing the activity to verify conformance with the applicable documented instructions, procedures and drawings.
d. Handling, Storage and Shipping - Procedures are established to control the handling, storage, shipping, cleaning, and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration.
e. Inspection, Test and Operating Status - Procedures are established to provide for the identifications of items which have satisfactorily passed required inspections and tests.
f. Corrective Action - Procedures are established to ensure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Revision 12 11.3-15 January, 2003 11.3.2.2.1.4 Welding All welding constituting the pressure boundary of pressure retaining components is performed by qualified welders employing qualified welding procedures per
. 11.3.2.2.1.5 Materials Materials for pressure retaining components of process systems were selected from those covered by the material specifications listed in Section II, Part A, of the ASME Boiler and Pressure Vessel Code, except that malleable, wrought or cast-iron materials are not used. The components satisfy all of the mandatory requirements of the material specifications with regard to manufacture, examination, repair, testing, identification, and certification. A description of the major process equipment including the design temperature and pressure and the materials of construction is given in
. Impact testing of carbon steel equipment, piping and valves operating at cold temperatures is in accordance with Paragraph UG84,Section VIII, of ASME Boiler and Pressure Vessel Code, Division 1. However, Code exceptions are not permitted for equipment. 11.3.2.2.1.6 Construction of Process Systems Pressure retaining components of process systems utilize welded construction to the maximum practicable extent. Process piping systems include the first root valve on sample and instrument lines. Process lines are not less than 3/4 inch nominal pipe size. Sample and instrument lines are not considered as portions of the process systems. Flanged joints or suitable rapid disconnect fittings are not used, except where maintenance requirements clearly indicate that such Revision 12 11.3-16 January, 2003 construction is preferable. Screwed connections in which threads provide the only seal are not used. Screwed connections backed up by seal welding or mechanical joints are used only on lines of 3/4 inch nominal pipe size. However, seal welding is not possible on the 3/4 inch connections on the level sensors for the hold-up line, prefilter, and cooler condenser moisture separator loop seals. Thread sealant is used on these joints. In lines 3/4 inch or greater, but less than two and one-half inch nominal pipe size, socket type welds are used. In lines of two and one-half inch nominal pipe size and larger, pipe welds are of the butt joint type. However, the Dryer Chiller drain line strainers are 3 inch nominal size to maximize strainer screen surface area resulting in socket weld connections larger than two and one half inches. 11.3.2.2.1.7 System Integrity Testing Completed process systems are pressure tested to the maximum practicable extent. Piping systems are hydrostatically tested during the construction phase in their entirety, using available valves or temporary plugs at atmospheric tank connections. Hydrostatic testing of piping systems is performed at a pressure 1.5 times the design pressure, but in no case at less than 75 psig. The test pressure is held for a minimum of 30 minutes with no leakage indicated. Pneumatic testing may be substituted for hydrostatic testing in accordance with the applicable codes. A helium leak test is performed on the entire, as-installed after construction, gaseous radwaste process system. After the initial pressure test and helium leak test (i.e., during the Operations phase) helium leak tests shall be performed on modifications/repairs whenever practicable. All welds performed during such modifications/repairs shall be subject to non-destructive examination (e.g., radiography or liquid penetrant exam). Revision 17 11.3-17 October, 2011 11.3.2.2.1.8 Instrumentation and Control This system is monitored by flow, temperature, pressure, and humidity instrumentation, and by hydrogen analyzers to ensure correct operation Revision 17 11.3-17a October, 2011 and control.
lists the process parameters that are instrumented to alarm in the control room. It also indicates whether the parameters are recorded or indicated. The operator is in control of the system at all times. This system has redundant hydrogen analyzers to monitor the hydrogen concentrations in the offgas system prior to the holdup process. If the hydrogen concentration exceeds 2 percent, an alarm will be annunciated in the control room. These hydrogen analyzers also support operation of the Hydrogen Water Chemistry (HWC) System. During operation of the Hydrogen Water Chemistry (HWC) System, a stoichiometric amount of oxygen is added upstream of the recombiner to recombine the hydrogen in the offgas system. These redundant hydrogen analyzers monitor the hydrogen concentration to allow HWC operation and shutdown HWC if % hydrogen gets too high. A radiation monitor after the offgas condenser continuously monitors 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 radiation monitor is also provided at the outlet of the charcoal adsorbers to continuously monitor the rate from the adsorber beds. This radiation monitor is used to isolate the offgas system on high radioactivity to prevent treated gas of unacceptably high activity from entering the vent. The activity of the gas entering and leaving the offgas treatment system is continuously monitored. Thus, system performance is known to the operator at all times. Provision is made for sampling and periodic analysis of the influent and effluent gases for purposes of determining their compositions. Revision 18 11.3-18 October, 2013 Environmental monitoring is used; however, at the estimated low dose levels, it is doubtful that the measurements can distinguish doses from the plant from normal variation in background radiation. Revision 14 11.3-18a October, 2005 11.3.2.2.1.9 Detonation Resistance Even though the system is designed to be free of ignition sources, the process pressure boundary of the system is detonation resistant. The pressure vessels are designed to withstand 350 psig static pressure, and piping and valving are designed to resist dynamic pressures encountered in long runs of piping at the design temperature (Reference 6). By the procedure described in (Reference 6), a designer can obtain the required wall thickness of a specific equipment design, which normally or possibly contains a detonable mixture of hydrogen and oxygen. This wall thickness is then translated to the corresponding detonation-containing, static equipment pressure rating by using an appropriate code calculation. 11.3.2.2.1.10 Operator Exposure Criteria and Controls The system is normally operated from the main control room. Equipment and process valves containing radioactive fluid are placed in shielded cells maintained at a pressure less than that of normally occupied areas. 11.3.2.2.1.11 Equipment Malfunction Malfunction analysis, indicating consequences and design precautions taken to accommodate failure of various components of the system, is given in
. 11.3.2.2.1.11.1 Previous Experience A system with similar equipment is in service at the KRB plant in Germany. Its performance is reviewed in (Reference 1). The Tsuruga and Fukushima I plants in Japan have similar recombiners in service. Revision 14 11.3-19 October, 2005 Similar systems (ambient temperature charcoal) are in service at Dresden 2 and 3, Pilgrim, Quad Cities 1 and 2, Nuclenor, Hatch, Browns Ferry 1, 2 and 3, and Duane Arnold. 11.3.2.3 Operating Procedure 11.3.2.3.1 Treated (Delayed) Radioactive Gas Sources 11.3.2.3.1.1 Main CondenserSteam Jet Air Ejector Offgas Low-Temp RECHAR System 11.3.2.3.1.1.1 Prestartup Preparations Prior to starting the main steam jet air ejectors (SJAE), the charcoal vault is cooled between 40F and 0F, the glycol cooler is chilled to near 35F and glycol is circulated through the cooler condenser, a desiccant dryer is regenerated and valved in, the offgas condenser cooling water is valved in, and the recombiner heaters are turned on. 11.3.2.3.1.1.2 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 225F with this air bleed and/or by admitting steam to the final stage of the SJAE. With the recombiners preheated, and the desiccant drier 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 with a fixed steam supply to less than four percent hydrogen by volume and the offgas condenser outlet is maintained at less than one percent hydrogen by volume. Revision 12 11.3-20 January, 2003 11.3.2.3.1.1.3 Normal Operation After startup, the noncondensibles 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. 11.3.2.3.1.1.4 Previous Experience Previous experience is reviewed in <Section 11.3.2.2.1.12>. 11.3.2.4 Performance Tests 11.3.2.4.1 Treated (Delayed) Radioactive Gas Sources 11.3.2.4.1.1 Main CondenserSteam Jet Air Ejector Offgas Low-Temp RECHAR System This system is used on a routine basis and does not require specific testing to assure operability. Monitoring equipment will be calibrated and maintained on a specific schedule and on indication of malfunction. 11.3.2.4.1.1.1 Recombiner Recombiner performance is continuously monitored and recorded by thermocouples that monitor the catalyst bed temperature profile and by a hydrogen analyzer that measures the hydrogen concentration of the effluent. Revision 12 11.3-21 January, 2003 During operation of the Hydrogen Water Chemistry (HWC) System, a stoichiometric amount of oxygen must be added upstream of the recombiner to recombine the hydrogen in the offgas system. Redundant hydrogen analyzers monitor hydrogen concentration and allow HWC operation or cause a HWC shutdown if % hydrogen gets too high. Revision 18 11.3-21a October, 2013 11.3.2.4.1.1.2 Prefilter These particulate filters are tested at the time of initial filter installation using DOP (dioctylphthalate) aerosol to determine whether an installed filter meets the minimum inplace efficiency of 99.95 percent retention. The DOP from filter testing is not allowed into the desiccant or the activated charcoal. This equipment is isolated during filter DOP testing and is bypassed until the process lines have been purged clear of test material. Because the DOP would have a detrimental effect on the desiccant and charcoal, this filter is not periodically tested. This is justified because the main function of this prefilter is to prevent the long-lived daughters of the radioactive xenons generated in the holdup pipe, from depositing in the downstream equipment, thereby minimizing contamination. Leakage through the filter has no effect on environmental release. 11.3.2.4.1.1.3 Desiccant Gas Drier Desiccant gas drier performance is continuously monitored by an onstream humidity analyzer. 11.3.2.4.1.1.4 Charcoal Performance The ability of the charcoal to delay the noble gases can be continuously evaluated by comparing activity measured and recorded by the process activity monitors at the exit of the offgas condenser and at the exit of the charcoal adsorbers. Experience with boiling water reactors has shown that the calibration of the offgas and vent effluent monitors changes with isotopic content. Revision 12 11.3-22 January, 2003 Isotopic content can change depending on the presence or absence of fuel cladding leaks in the reactor and the nature of the leaks. Because of this possible variation, the monitors are periodically calibrated against grab samples, and whenever the radiation monitor after the offgas condenser shows significant variation in noble gas activity indicating a significant change in plant operations. Grab sample points are located upstream and downstream of the first charcoal bed and downstream of the last charcoal bed. They can be used for periodic sampling if the monitoring equipment indicates degradation of system delay performance. 11.3.2.4.1.1.5 Post Filter On installation and replacements, these particulate filters will be tested using a DOP smoke test or equivalent. 11.3.2.4.1.1.6 Previous Experience Previous experience is reviewed in <Section 11.3.2.2.1.11.1>. 11.3.3 RADIOACTIVE RELEASES 11.3.3.1 Release Points A simplified flow diagram of the radioactive gas flow and treatment for the containment, the control complex, the auxiliary building, the fuel handling building, the radwaste building, the intermediate building, the turbine building, and the offgas building is presented in <Figure 11.3-3>. The physical location and elevation of the release points are shown on <Figure 1.2-18>. Revision 12 11.3-23 January, 2003 The discharge from the condenser steam jet air ejector is processed by the low TEMP RECHAR System prior to release through the offgas building vent. <Section 11.3.2> discusses the low TEMP RECHAR System.
provides the vent dimensions, effluent velocity and effluent gas temperature for each of the release points.
,
, and
provide parameters for charcoal temperatures at 20F, 40F and 70F. 11.3.3.2 Dilution Factors The atomospheric dilution factors associated with normal plant releases are based upon the average annual meteorological conditions applicable to the site as well as the effective release height of the effluent discharge pathway. The site meteorological conditions are defined in <Section 2.3.5>. Also included in
are the average annual long term dilution factors (x/Q). 11.3.3.3 Estimated Releases and Dose Rates The release rates of radioactive materials in gaseous effluents are presented in
,
,
,
,
,
,
,
,
,
,
and
. These values were calculated with the GALE computer code and are based on the assumptions and parameters provided in <NUREG-0016> and
,
,
, and
. As shown in
,
,
, and
, the estimated releases are a small fraction of the limits of <10 CFR 20>, and are as low as reasonably achievable. (Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the <10 CFR 20> regulations prior to October 4, 1993. Radiological assessments for plant design bases Revision 19 11.3-24 October, 2015 modifications that are performed after October 4, 1993 will be evaluated using the revised <10 CFR 20> dated October 4, 1993). The estimated offsite doses for the Perry site and a comparison with the design objectives of <10 CFR 50, Appendix I> and the dose limits of <40 CFR 190> are presented in <Section 5.2.4> and <Section 5.2.5> of the PNPP Environmental Report. 11.

3.4 REFERENCES

FOR SECTION 11.3

1. Miller, C. W., Experimental and Operational Confirmation of Offgas System Design Parameters, NEDO-10751, January 1973. (Proprietary)
2. Browning, W. E., et al., Removal of Fission Product Gases from Reactor Offgas Streams by Adsorption, (ORNL) CF59-6-47, June 11, 1959.
3. Siegwarth, D. P., Measurement of Dynamic Adsorption Coefficients for Noble Gases on Activated Carbon, 12th AEC Air Cleaning Conference.
4. Standards for Steam Surface Condensers, Sixth Edition, Heat Exchange Institute, New York, NY, 1970.
5. Underhill, Dwight, et al., Design of Fission Gas Holdup Systems, Proceedings of the Eleventh AEC Air Cleaning Conference, 1970,
p. 217.
6. Head, R. A., et al., Releases From BWR Radwaste Management Systems, NEDO-10951, July 1973.

Revision 19 11.3-25 October, 2015

TABLE 11.3-1a ESTIMATED AIR EJECTOR OFFGAS RELEASE RATES(1)

(Charcoal Temperature = 0F)

T=30 Normal Discharge from Minutes Charcoal Adsorbers(2)

Isotope Half-Life = T 0 Ci/Sec Ci/sec Ci/sec Ci/yr(3)

Kr-83m 1.86 hr 3.4E+03 2.9E+03 - -

Kr-85m 4.4 hr 6.1E+03 5.6E+03 5.3E-01 1.5E+01 Kr-85(4) 10.74 yr 2.0E+01 2.0E+01 2.0E+01 5.7E+02 Kr-87 76 min 2.0E+04 1.5E+04 - -

Kr-88 2.79 hr 2.0E+04 1.8E+04 - -

Kr-89 3.18 min 1.3E+05 1.8E+02 - -

Kr-90 32.3 sec 2.8E+05 - - -

Kr-91 8.6 sec 3.3E+05 - - -

Kr-92 1.84 sec 3.3E+05 - - -

Kr-93 1.29 sec 9.3E+04 - - -

Kr-94 1.0 sec 2.3E+04 - - -

Kr-95 0.5 sec 2.1E+03 - - -

Kr-97 1 sec 1.4E+01 - - -

Xe-131m 11.96 day 1.5E+01 1.5E+01 6.5E-01 1.8E+01 Xe-133m 2.26 day 2.9E+02 2.8E+02 - -

Xe-133 5.27 day 8.2E+03 8.2E+03 6.6E+00 1.9E+02 Xe-135m 15.7 min 2.6E+04 6.9E+03 - -

Xe-135 9.16 hr 2.2E+04 2.2E+04 - -

Xe-137 3.82 min 1.5E+05 6.7E+02 - -

Xe-138 14.2 min 8.9E+04 2.1E+04 - -

Xe-139 40 sec 2.8E+05 - - -

Xe-140 13.6 sec 3.0E+05 - - -

Revision 19 11.3-26 October, 2015

TABLE 11.3-1a (Continued)

T=30 Normal Discharge from Minutes Charcoal Adsorbers(2)

Isotope Half-Life T=0 Ci/Sec Ci/sec Ci/sec Ci/yr(3)_

Xe-141 1.72 sec 2.4E+05 - - -

Xe-142 1.22 sec 7.3E+04 - - -

Xe-143 0.96 sec 1.2E+04 - - -

Xe-144 9 sec 5.6E+02 - - -

TOTALS 2.4E+06 1.0E+05 2.8E+01 7.9E+02 NOTES:

(1)

Release rates are based on the 1971 mixture.

(2) 30 scfm in-leakage.

(3)

Plant Capacity Factor = 0.9.

(4) 10 to 20 Ci/sec estimated from experimental observations.

Revision 12 11.3-27 January, 2003

TABLE 11.3-1b ESTIMATED AIR EJECTOR OFFGAS RELEASE RATES(1)

(Charcoal Temperature = 20F)

T=30 Normal Discharge from min Charcoal Adsorbers(2)

Isotope Half-Life = T 0 Ci/sec Ci/sec Ci/sec Ci/yr(3)_

Kr-83m 1.86 hr 3.4E+03 2.9E+03 - -

Kr-85m 4.4 hr 6.1E+03 5.6E+03 1.0E+00 2.9E+01 Kr-85(4) 10.74 yr 2.0E+01 2.0E+01 2.0E+01 5.7E+02 Kr-87 76 min 2.0E+04 1.5E+04 - -

Kr-88 2.79 hr 2.0E+04 1.8E+04 - -

Kr-89 3.18 min 1.3E+05 1.8E+02 - -

Kr-90 32.3 sec 2.8E+05 - - -

Kr-91 8.6 sec 3.3E+05 - - -

Kr-92 1.84 sec 3.3E+05 - - -

Kr-93 1.29 sec 9.3E+04 - - -

Kr-94 1.0 sec 2.3E+04 - - -

Kr-95 0.5 sec 2.1E+03 - - -

Kr-97 1 sec 1.4E+01 - - -

Xe-131m 11.96 day 1.5E+01 1.5E+01 7.8E-01 2.2E+01 Xe-133m 2.26 day 2.9E+02 2.8E+02 - -

Xe-133 5.27 day 8.2E+03 8.2E+03 9.9E+00 2.8E+02 Xe-135m 15.7 min 2.6E+04 6.9E+03 - -

Xe-135 9.16 hr 2.2E+04 2.2E+04 - -

Xe-137 3.82 min 1.5E+05 6.7E+02 - -

Xe-138 14.2 min 8.9E+04 2.1E+04 - -

Xe-139 40 sec 2.8E+05 - - -

Xe-140 13.6 sec 3.0E+05 - - -

Revision 19 11.3-28 October, 2015

TABLE 11.3-1b (Continued)

T=30 Normal Discharge from min Charcoal Adsorbers(2)

Isotope Half-Life T=0 Ci/sec Ci/sec Ci/sec Ci/yr(3)_

Xe-141 1.72 sec 2.4E+05 - - -

Xe-142 1.22 sec 7.3E+04 - - -

Xe-143 0.96 sec 1.2E+04 - - -

Xe-144 9 sec 5.6E+02 - - -___

TOTALS 2.4E+06 1.0E+05 3.2E+01 9.0E+02 NOTES:

(1)

Release rates are based on the 1971 mixture.

(2) 30 scfm in-leakage.

(3)

Plant Capacity Factor = 0.9.

(4) 10 to 20 Ci/sec estimated from experimental observations.

Revision 12 11.3-29 January, 2003

TABLE 11.3-1c ESTIMATED AIR EJECTOR OFFGAS RELEASE RATES(1)

(Charcoal Temperature = 40F)

T=30 Normal Discharge from min Charcoal Adsorbers(2)

Isotope Half-Life = T 0 Ci/sec Ci/sec Ci/sec Ci/yr(3)_

Kr-83m 1.86 hr 3.4E+03 2.9E+03 - -

Kr-85m 4.4 hr 6.1E+03 5.6E+03 9.3E+00 2.6E+02 Kr-85(4) 10.74 yr 2.0E+01 2.0E+01 2.0E+01 5.7E+02 Kr-87 76 min 2.0E+04 1.5E+04 - -

Kr-88 2.79 hr 2.0E+04 1.8E+04 7.2E-01 2.0E+01 Kr-89 3.18 min 1.3E+05 1.8E+02 - -

Kr-90 32.3 sec 2.8E+05 - - -

Kr-91 8.6 sec 3.3E+05 - - -

Kr-92 1.84 sec 3.3E+05 - - -

Kr-93 1.29 sec 9.3E+04 - - -

Kr-94 1.0 sec 2.3E+04 - - -

Kr-95 0.5 sec 2.1E+03 - - -

Kr-97 1 sec 1.4E+01 - - -

Xe-131m 11.96 day 1.5E+01 1.5E+01 1.8E+00 5.0E+01 Xe-133m 2.26 day 2.9E+02 2.8E+02 - -

Xe-133 5.27 day 8.2E+03 8.2E+03 6.5E+01 1.8E+03 Xe-135m 15.7 min 2.6E+04 6.9E+03 - -

Xe-135 9.16 hr 2.2E+04 2.2E+04 - -

Xe-137 3.82 min 1.5E+05 6.7E+02 - -

Xe-138 14.2 min 8.9E+04 2.1E+04 - -

Xe-139 40 sec 2.8E+05 - - -

Xe-140 13.6 sec 3.0E+05 - - -

Revision 19 11.3-30 October, 2015

TABLE 11.3-1c (Continued)

T=30 Normal Discharge from min Charcoal Adsorbers(2)

Isotope Half-Life T=0 Ci/sec Ci/sec Ci/sec Ci/yr(3)_

Xe-141 1.72 sec 2.4E+05 - - -

Xe-142 1.22 sec 7.3E+04 - - -

Xe-143 0.96 sec 1.2E+04 - - -

Xe-144 9 sec 5.6E+02 _______

TOTALS 2.4E+06 1.0E+05 9.7E+01 2.7E+03 NOTES:

(1)

Release rates are based on the 1971 mixture.

(2) 30 scfm in-leakage.

(3)

Plant Capacity Factor = 0.9.

(4) 10 to 20 Ci/sec estimated from experimental observations.

Revision 12 11.3-31 January, 2003

TABLE 11.3-1d ESTIMATED AIR EJECTOR OFFGAS RELEASE RATES(1)

(Charcoal Temperature = 70F)

T=30 Normal Discharge from min Charcoal Adsorbers(2)

Isotope Half-Life = T 0 Ci/sec Ci/sec Ci/sec Ci/yr(3)_

Kr-83m 1.86 hr 3.4E+03 2.9E+03 1.1E-01 3.1E+00 Kr-85m 4.4 hr 6.1E+03 5.6E+03 7.6E+01 2.2E+03 Kr-85(4) 10.74 yr 2.0E+01 2.0E+01 2.0E+01 5.7E+02 Kr-87 76 min 2.0E+04 1.5E+04 - -

Kr-88 2.79 hr 2.0E+04 1.8E+04 2.0E+01 5.7E+02 Kr-89 3.18 min 1.3E+05 1.8E+02 - -

Kr-90 32.3 sec 2.8E+05 - - -

Kr-91 8.6 sec 3.3E+05 - - -

Kr-92 1.84 sec 3.3E+05 - - -

Kr-93 1.29 sec 9.3E+04 - - -

Kr-94 1.0 sec 2.3E+04 - - -

Kr-95 0.5 sec 2.1E+03 - - -

Kr-97 1 sec 1.4E+01 - - -

Xe-131m 11.96 day 1.5E+01 1.5E+01 3.9E+00 1.1E+02 Xe-133m 2.26 day 2.9E+02 2.8E+02 2.4E-01 6.9E+00 Xe-133 5.27 day 8.2E+03 8.2E+03 3.9E+02 1.1E+04 Xe-135m 15.7 min 2.6E+04 6.9E+03 - -

Xe-135 9.16 hr 2.2E+04 2.2E+04 - -

Xe-137 3.82 min 1.5E+05 6.7E+02 - -

Xe-138 14.2 min 8.9E+04 2.1E+04 - -

Xe-139 40 sec 2.8E+05 - - -

Xe-140 13.6 sec 3.0E+05 - - -

Revision 19 11.3-32 October, 2015

TABLE 11.3-1d (Continued)

T=30 Normal Discharge from min Charcoal Adsorbers(2)

Isotope Half-Life T=0 Ci/sec Ci/sec Ci/sec Ci/yr(3)_

Xe-141 1.72 sec 2.4E+05 - - -

Xe-142 1.22 sec 7.3E+04 - - -

Xe-143 0.96 sec 1.2E+04 - - -

Xe-144 9 sec 5.6E+02 _______

TOTALS 2.4E+06 1.0E+05 5.1E+02 1.4E+04 NOTES:

(1)

Release rates are based on the 1971 mixture.

(2) 30 scfm in-leakage.

(3)

Plant Capacity Factor = 0.9.

(4) 10 to 20 Ci/sec estimated from experimental observations.

Revision 12 11.3-33 January, 2003

TABLE 11.3-2 OFFGAS SYSTEM MAJOR EQUIPMENT ITEMS Offgas Preheaters - 2 required.

Construction: Stainless steel tubes and carbon steel shell. 350 psig design pressure, 1,000 psig tube design pressure 40F/450F shell design temperature, 40F/575F tube design temperature.

Catalytic Recombiners - 2 required.

Construction: Carbon steel cartridge, carbon steel shell. Catalyst cartridge containing a precious metal catalyst on metal base or porous non-dusting ceramic. Catalyst cartridge to be replaceable without removing vessel. 350 psig design pressure. 900F design temperature.

Offgas Condenser - 1 required.

Construction: Low alloy steel shell. Stainless steel tubes. 350 psig shell design pressure. 250 psig tube design pressure. 900F shell design temperature. 150F tube design temperature.

Water Separator - 1 required.

Construction: Carbon steel shell, stainless steel wire mesh. 350 psig design pressure. 250F design temperature.

Cooler-Condenser - 2 required.

Construction: Carbon or stainless steel shell. Stainless steel tubes.

100 psig tube design pressure. 350 psig shell design pressure. 150°F tube design temperature 32F/150F shell design temperature.

Moisture Separators (Downstream of cooler-condenser) - 2 required.

Construction: Carbon steel shell, stainless steel wire mesh. 350 psig design pressure 32F/150F design temperature.

Desiccant Dryer - 4 required.

Construction: Carbon steel shell packed with Linde Mol Sieve or equivalent. 350 psig design pressure, 32F/500F design temperature.

Desiccant Regeneration Skid - 2 required.(1)

Dryer Chiller - 2 required.(1)

Construction: Carbon steel shell, stainless steel tubes, design temperature 32F/500F. Design pressure 50 psig.

Revision 12 11.3-34 January, 2003

TABLE 11.3-2 (Continued)

Regenerator Blower - 2 required.(1)

Construction: Cast iron, design pressure 50 psig, design temperature 32F/150F. Sellers standard.

Dryer Heater - 2 required.

Construction: Carbon steel, design temperature 32F/500F, design pressure 50 psig.

Gas Cooler - 2 required.

Construction: Carbon or stainless steel material. 1,050 psig tube design temperature. -50F/150F design temperature.

Glycol Cooler Skid - 1 required.(1)

Glycol Storage Tank - 1 required.(1)

Construction: Carbon steel 3,000 gallon. Water-filled hydrostat static design pressure. 32F design temperature. API-650.

Glycol Solution Refrigerators and Motor Drives - 3 required.(1)

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

Glycol Pumps and Motor Drives - 3 required.(1)

Construction: Cast iron, 3 inch connections 0°F design temperature.

Sellers standard.

Prefilters and After Filters - 2 required of each type.

Construction: Carbon steel shell. High-efficiency, moisture-resistant filter element. Flanged shell. 350 psig design pressure. -50F/150F design temperature.

Charcoal Adsorbers - 8 beds.

Construction: Carbon steel. Approximately 4-ft od x 21-ft vessels each containing approximately 4 tons of activated carbon. Design pressure 350 psig. Design temperature -50F/250F.

NOTE:

(1)

Not located within the boundary of the portion of the N64 system that actively processes radioactive materials and which is required to be detonation resistant.

Revision 12 11.3-35 January, 2003

(DELETED) Revision 12 11.3-36 January, 2003 TABLE 11.3-4 EQUIPMENT MALFUNCTION ANALYSIS Equipment Design Item Malfunction Consequences Precaution Steam jet air Low flow of motive When the hydrogen and oxygen Alarm provided on steam for ejector high pressure steam concentration exceed 4 and low steam flow. Recombiner 5 vol %, respectively, the temperature alarm. process gas may become flammable, if insufficient steam is supplied. Inadequate steam flow will Steam flow to be held at cause overheating and constant maximum flow deterioration of the regardless to plant level. catalyst. Recombiner temperature alarm. Wear of supply Increased steam flow to Low temperature alarms on steam nozzle of recombiner. This would preheater exit (recombiner ejector reduce degree of recombination inlet). Recombiner H2 at low power levels. analyzers. Preheaters Steam leak Would further dilute process Spare preheater. offgas. Steam consumption would increase. Low pressure steam Recombiner performance would Low-temperature alarms on supply fall off at low power level, preheater exit (recombiner and hydrogen content of inlet). Recombiner outlet recombiner gas discharge may H2 analyzers. increase, eventually to a combustible mixture. Revision 12 11.3-37 January, 2003 TABLE 11.3-4 (Continued) Equipment Design Item Malfunction Consequences Precaution Recombiners Catalyst gradually Temperature profile changes Temperature probes in deactivates through catalyst. Eventually recombiner H2 analyzer excess H2 would be detected by provided. Spare recombiner. H2 analyzer or by a flowmeter. Eventually the stripped gas could become combustible. Catalyst gets wet at H2 conversion falls off and H2 Condensate drains, start is detected by downstream temperature probes in analyzers. Eventually the recombiner. Air bleed system gas could become combustible. at startup. Recombiner thermal blanket, spare recombiner and heater, hydrogen analyzer. Offgas Cooling water leak The coolant (reactor None. condenser 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.) Revision 19 11.3-38 October, 2015 TABLE 11.3-4 (Continued) Equipment Design Item Malfunction Consequences Precaution Offgas Liquid level If both drain valves fail to Two independent drain condenser instruments fail open, water will build up in systems, each provided with (Cont.) the condenser and pressure high and low level alarms. drop will increase. The high delta 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 Corrosion of wire Higher quantity of water Stainless steel mesh separator mesh element collected in holdup line and specified. routed to radwaste. Revision 12 11.3-39 January, 2003 TABLE 11.3-4 (Continued) Equipment Design Item Malfunction Consequences Precaution Cooler Corrosion of tubes Glycol-water solution would Stainless-steel tubes condensers leak into process (shell) specified. Low level alarm side and be discharged to glycol tank level. Spare clean radwaste. If not cooler condenser provided. detected as radwaste, the glycol solution would discharge to reactor condensate system. Icing up of tubes Shell side of cooler could Design glycol-H2 solution plug up with ice, gradually temperature well above building up pressure drop. freezing point. Spare unit If this happens, the spare provided. Temperature unit could be activated. indication and low alarms Complete blockage of both on glycol temperature and units. process gas temperature. Glycol Mechanical If both spare units fail to Two spare refrigerators refrigeration operate, the glycol solution during normal operation machines temperature will rise and the are provided. Glycol dehumidification system solution temperature performance will deteriorate. alarms provided. Gas This will require rapid moisture detectors regeneration cycles for the provided downstream of desiccant beds and may gas driers. raise the gas dewpoint as it is discharged from the drier. Revision 12 11.3-40 January, 2003 TABLE 11.3-4 (Continued) Equipment Design Item Malfunction Consequences Precaution Moisture Corrosion wire mesh Increased moisture would be Stainless steel mesh separators element retained in process gas routed specified. Spare unit to gas driers. Over a long provided. High delta P period, the desiccant drier alarm on prefilter. cycle period would deteriorate as result of moisture pickup. Pressure drop across prefilter may increase if filter media is wetted. Prefilters Loss of integrity More radioactivity would Spare unit provided in of filter deposite in the drier desiccant. separate vault. Delta P This would increase radiation instrumentation provided. level in the drier vault and make maintenance more difficult, but would not affect releases to the environment. Desiccant Moisture breakthrough Moisture would freezeout in Drier cycles on time. drier gas cooler and would result Redundant gas humidity in increased system pressure analyzers and alarms drop. Gas with a high supplied. Redundant drier dewpoint temperature would system supplied gas drier reach charcoal bed. and first charcoal bed can be bypassed through alternate drier to second charcoal bed. Revision 14 11.3-41 October, 2005 TABLE 11.3-4 (Continued) Equipment Design Item Malfunction Consequences Precaution Desiccant Mechanical failure Inability to regenerate Redundant, shielded desiccant regeneration desiccant. beds and drier equipment equipment supplied. Charcoal Charcoal accumulates Charcoal performance will Highly instrumented, absorbers moisture deteriorate gradually as mechanically simple gas moisture deposits. Holdup dehumidification system times for krypton and xenon with redundant equipment. would decrease, and plant emissions would increase. Provisions made for drying charcoal. Vault Mechanical failure If temperature exceeds Spare refrigeration unit refrigeration approximately 40F, increased provided. Charcoal adsorber units emission could occur. vaults A and B temperature instrumentation provided. After filter Loss of integrity of Probably of no real Delta P instrumentation filter media consequence, the charcoal provided. Spare unit media itself should be a provided. good filter at the low air velocity. Revision 19 11.3-42 October, 2015 TABLE 11.3-4 (Continued) Equipment Design Item Malfunction Consequences Precaution System Internal detonation Release of radioactivity if Main process equipment and pressure boundary fails. piping are designed to contain a detonation. System Earthquake damage Release of radioactivity. Dose consequences are within <10 CFR 20> limits. Analysis is included in (Reference 6). Revision 12 11.3-43 January, 2003 TABLE 11.3-5 FREQUENCY AND QUANTITY OF STEAM DISCHARGED TO SUPPRESSION POOL Frequency Quantity of Steam Event(1) Category_ ___lbs/event)____
1. RCIC Test (Monthly) Moderate 27,600
2. Inadvertent RCIC Injection Moderate 4,600
3. SRV Test (each valve) Moderate 3,900
4. SRV Flow Capacity Test (each valve) Infrequent 15,300
5. Total SRV Leakage (19 valve max.) Continuous 380/Hr
6. Trip of Both Recirc. Pump Motor Moderate 30,000
7. Turbine Trip Moderate 30,000
8. Generator Load Rejection Moderate 30,000
9. Pressure Regulator Failure, Open Moderate 374,000(2)
10. Recirc. Controller Failure Moderate 30,000
11. Loss of All Feedwater Flow Moderate 30,000
12. Inadvertent MSIV Closure Moderate 374,000(2)
13. Loss of Condenser Vacuum Moderate 374,000(2)
14. Feedwater Control Failure, Max. Demand Moderate 30,000
15. Loss of Auxiliary Transformer Moderate 934,000(2)
16. Loss of All Grid Connections Moderate 934,000(2)
17. Turbine Trip w/o Bypass Infrequent 374,000(2)
18. Generator Load Rejection w/o Bypass Infrequent 374,000(2)
19. Stuck Open SRV Moderate 641,000 NOTES:
(1) Bases and assumptions for the listed events are as follows: (a) Events 1 and 2 are based on steam flow rate during test mode per RCIC System Process Diagram 762E421C, for 60 and 10 minutes, respectively. (b) Event 3 assumes test SRV opened 30 seconds maximum at 300-500 psig vessel pressure. (c) Event 4 assumes tested SRV opened 30-60 seconds at 1,000 psig vessel pressure. (d) Event 5 is based on maximum average SRV leakage rate of 20 lb/hr/valve. Revision 12 11.3-44 January, 2003 TABLE 11.3-5 (Continued) NOTES: (Continued) (e) Events 6 through 18 are based on event descriptions from <Chapter 15>. (f) Event 19 is based on vessel depressurized to 100 psia with two additional SRVs opened 10 minutes following scram. (2) Isolation event. Except for Events 15 and 16, it is assumed that SRV actuation is terminated 30 minutes into the event whereupon the reactor is depressurized at 100°F/hr via RHR steam condensing mode. For Events 15 and 16, it is assumed that loss of plant air prevents availability of RHR steam condensing mode and normal SRV opening, vessel depressurized via ADSSRVs. RHR steam condensing mode is not used at PNPP. Revision 19 11.3-45 October, 2015 TABLE 11.3-6 GASEOUS RADWASTE EQUIPMENT DESIGN REQUIREMENTS Welder Design and Qualification Inspection Fabrication Materials(1)_ and Procedure and Testing Pressure ASME Code ASME Code ASME Code ASME Code Vessels Section VIIISection IISection IXSection VIII Div 1 Div 1 Atmos- ASME Code(2) ASME Code ASME Code ASME Code(2) pheric Section IIISection IISection IXSection III or 0-15 Class 3, API Class 3, API psig 620;650, AWWA 620;650, Tanks D-100 AWWA D-100 Heat ASME Code ASME Code ASME Code ASME Code Ex- Section VIIISection IISection IXSection VIII changers Div 1; and Div 1 TEMA Piping ANSI B 31.1 ASTMORASMEASME Code ASME Code(2) and Code Section IX Valves Section II Pumps Manufacturers(3) ASME Code ASME Code ASME Code(2) Standards Section IISection IXSection III or Manufac- (as required) Class 3; and turers Hydraulic Standard Institute NOTES: (1) Material manufacturers certified test reports should be obtained whenever possible. (2) ASME Code stamp and material traceability not required. (3) Manufacturers standard for the intended service. Hydrotesting should be 1.5 times the design pressure. Revision 12 11.3-46 January, 2003 TABLE 11.3-7 OFFGAS SYSTEM ALARMED PROCESS PARAMETERS Control Room Parameters Indicated Recorded Air ejector discharge pressure - high X Preheater discharge temperature - low X Recombiner catalyst temperature - high/low X Offgas condenser water level (dual) - high/low (LOCAL) Offgas condenser gas discharge temperature - high (LOCAL) H2 analysis (offgas condenser discharge) - dual - high X Offgas condenser discharge radiation - high X X Gas flow - high/low X Moisture separator discharge temperature - high/low X Glycol solution temperature - high/low X Glycol solution level - low Gas drier discharge humidity - high (LOCAL) Prefilter dP - high X Charcoal adsorber temperature - high X Carbon vault A & B temperature - high/low X X Carbon train flow - high/low X After filter dP - high X Offgas (carbon bed discharge) radiation - high X X Steam flow - low Carbon train dP - high X Revision 12 11.3-47 January, 2003 TABLE 11.3-8a INPUT PARAMETERS USED FOR CALCULATING GASEOUS RELEASES(5) (charcoal temperature = 0F) Maximum core thermal power - 3,758 MWt Total main steam flow rate - 16.3x106 lb/hr Mass of reactor coolant in the reactor vessel - 5.28x105 lb Mass of steam in the reactor vessel - 1.93x104 lb Holdup Times Charcoal delay (krypton) - 2.47 days Charcoal delay (xenon) - 54.2 days Mass of charcoal in the offgas system - approximately 32 tons Operating and dew point temperatures of offgas system - 0F and -20F, respectively Dynamic adsorption coefficient for Xe and Kr is proprietary (General Electric) Ventilation and Exhaust Systems Purge Rate and Decontamination DF Frequency (Reactor Building __Factors (DF) Bases Building Only) Reactor(1) 100 HEPA 5,000 cfm, continuous(4) 10 Charcoal approximately 25,000 cfm, refueling Auxiliary(1) 100 HEPA 10 Charcoal Radwaste(1) 100 HEPA 10 Charcoal Turbine(2) 1 - 1 - Offgas(3) 100 HEPA 10 Charcoal Revision 19 11.3-48 October, 2015 TABLE 11.3-8a (Continued) Effluent Gas Effluent Velocity Vent Temperature Release Points ____(ft/min)_____ Dimensions __(maximum)_ Unit 1 Plant Vent 4,100 48x90 105F Turbine Building/ 4,000 120x120 115F Heater Bay Vent Offgas Vent 2,100 34x34 105F Unit 2 Plant Vent 3,500 48x90 105F Turbine Building/ 4,000 120x120 115F Heater Bay Vent Offgas Vent 1,900 34x34 105F NOTES: (1) The reactor building, auxiliary building and radwaste building releases are through the plant vent. (2) The turbine building releases are through the turbine building/heater bay vent. (3) The offgas building releases are through the offgas vent. (4) Assuming a continuous reactor building purge provides an enveloping dose estimate for routine gaseous releases. (5) The information in this table was used with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-49 October, 2015 TABLE 11.3-8b INPUT PARAMETERS USED FOR CALCULATING GASEOUS RELEASES(5) (charcoal temperature = 20F) Maximum core thermal power - 3,758 MWt Total main steam flow rate - 16.3x106 lb/hr Mass of reactor coolant in the reactor vessel - 5.28x105 lb Mass of steam in the reactor vessel - 1.93x104 lb Holdup Times Charcoal delay (krypton) - 2.30 days Charcoal delay (xenon) - 51.1 days Mass of charcoal in the offgas system - approximately 32 tons Operating and dew point temperatures of offgas system +20F and -20F, respectively Dynamic adsorption coefficient for Xe and Kr is proprietary (General Electric) Ventilation and Exhaust Systems Purge Rate and Decontamination DF Frequency (Reactor Building __Factors (DF) Bases Building Only) Reactor(1) 100 HEPA 5,000 cfm, continuous(4) 10 Charcoal 30,000 cfm, refueling Auxiliary(1) 100 HEPA 10 Charcoal Radwaste(1) 100 HEPA 10 Charcoal Turbine(2) 1 - 1 - Offgas(3) 100 HEPA 10 Charcoal Revision 19 11.3-50 October, 2015 TABLE 11.3-8b (Continued) NOTES: (1) The reactor building, auxiliary building and radwaste building releases are through the plant vent. (2) The turbine building releases are through the turbine building/heater bay vent. (3) The offgas building releases are through the offgas vent. (4) Assuming a continuous reactor building purge provides an enveloping dose estimate for routine gaseous releases. (5) This historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-51 October, 2015 TABLE 11.3-8c INPUT PARAMETERS USED FOR CALCULATING GASEOUS RELEASES(5) (charcoal temperature = 40F) Maximum core thermal power - 3,758 MWt Total main steam flow rate - 16.3x106 lb/hr Mass of reactor coolant in the reactor vessel - 5.28x105 lb Mass of steam in the reactor vessel - 1.93x104 lb Holdup Times Charcoal delay (krypton) - 1.72 days Charcoal delay (xenon) - 36.8 days Mass of charcoal in the offgas system - approximately 32 tons Operating and dew point temperatures of offgas system +40F and -20F, respectively Dynamic adsorption coefficient for Xe and Kr is proprietary (General Electric) Ventilation and Exhaust Systems Purge Rate and Decontamination DF Frequency (Reactor Building __Factors (DF) Bases Building Only) Reactor(1) 100 HEPA 5,000 cfm, continuous(4) 10 Charcoal 30,000 cfm, refueling Auxiliary(1) 100 HEPA 10 Charcoal Radwaste(1) 100 HEPA 10 Charcoal Turbine(2) 1 - 1 - Offgas(3) 100 HEPA 10 Charcoal Revision 19 11.3-52 October, 2015 TABLE 11.3-8c (Continued) NOTES: (1) The reactor building, auxiliary building and radwaste building releases are through the plant vent. (2) The turbine building releases are through the turbine building/heater bay vent. (3) The offgas building releases are through the offgas vent. (4) Assuming a continuous reactor building purge provides an enveloping dose estimate for routine gaseous releases. (5) This historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-53 October, 2015 TABLE 11.3-8d INPUT PARAMETERS USED FOR CALCULATING GASEOUS RELEASES(5) (charcoal temperature = 70F) Maximum core thermal power - 3,758 MWt Total main steam flow rate - 16.3x106 lb/hr Mass of reactor coolant in the reactor vessel - 5.28x105 lb Mass of steam in the reactor vessel - 1.93x104 lb Holdup Times Charcoal delay (krypton) - 1.16 days Charcoal delay (xenon) - 23.1 days Mass of charcoal in the offgas system - approximately 32 tons Operating and dew point temperatures of offgas system +70F and -20F, respectively Dynamic adsorption coefficient for Xe and Kr is proprietary (General Electric) Ventilation and Exhaust Systems Purge Rate and Decontamination DF Frequency (Reactor Building __Factors (DF) Bases Building Only) Reactor(1) 100 HEPA 5,000 cfm, continuous(4) 10 Charcoal 30,000 cfm, refueling Auxiliary(1) 100 HEPA 10 Charcoal Radwaste(1) 100 HEPA 10 Charcoal Turbine(2) 1 - 1 - Offgas(3) 100 HEPA 10 Charcoal Revision 19 11.3-54 October, 2015 TABLE 11.3-8d (Continued) NOTES: (1) The reactor building, auxiliary building and radwaste building releases are through the plant vent. (2) The turbine building releases are through the turbine building/heater bay vent. (3) The offgas building releases are through the offgas vent. (4) Assuming a continuous reactor building purge provides an enveloping dose estimate for routine gaseous releases. (5) This historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-55 October, 2015 TABLE 11.3-9a CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 1(4) (Ci/year) Unit 1 Unit 1 Unit 1 Unit 1 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 7 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 See Note(1) See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 8 See Note(1) Xe-133m See Note(1) See Note(1) See Note(1) See Note(1) Xe-133 142 250 85 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 113 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.9-2 1.9-1 See Note(1) 3.0-2 I-133 1.5-1 7.6-1 See Note(1) See Note(1) Cr-51 9.6-5 1.3-2 - See Note(1) Mn-54 3.6-4 6.0-4 - See Note(1) Fe-59 1.6-4 5.0-4 - See Note(1) Co-58 5.7-5 6.0-4 - See Note(1) Co-60 1.1-3 2.0-3 - See Note(1) Zn-65 5.5-5 2.0-4 - See Note(1) Sr-89 6.3-6 6.0-3 - See Note(1) Sr-90 3.1-6 2.0-5 - See Note(1) Zr-95 8.5-6 1.0-4 - See Note(1) Sb-124 4.7-6 3.0-4 - See Note(1) Cs-134 1.3-4 3.0-4 - 3.0-6 Cs-136 1.1-5 5.0-5 - 2.0-6 Cs-137 2.0-4 6.0-4 - 1.0-5 Ba-140 9.0-6 1.1-2 - 1.1-5 Ce-141 2.8-5 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 13 - H-3 47 - - - NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 0F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (4) This historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-56 October, 2015 TABLE 11.3-9b CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 1(4) (Ci/year) Unit 1 Unit 1 Unit 1 Unit 1 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 13 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 See Note(1) See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 10 See Note(1) Xe-133m See Note(1) See Note(1) See Note(1) See Note(1) Xe-133 142 250 130 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 113 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.9-2 1.9-1 See Note(1) 3.0-2 I-133 1.5-1 7.6-1 See Note(1) See Note(1) Cr-51 9.6-5 1.3-2 - See Note(1) Mn-54 3.6-4 6.0-4 - See Note(1) Fe-59 1.6-4 5.0-4 - See Note(1) Co-58 5.7-5 6.0-4 - See Note(1) Co-60 1.1-3 2.0-3 - See Note(1) Zn-65 5.5-5 2.0-4 - See Note(1) Sr-89 6.3-6 6.0-3 - See Note(1) Sr-90 3.1-6 2.0-5 - See Note(1) Zr-95 8.5-6 1.0-4 - See Note(1) Sb-124 4.7-6 3.0-4 - See Note(1) Cs-134 1.3-4 3.0-4 - 3.0-6 Cs-136 1.1-5 5.0-5 - 2.0-6 Cs-137 2.0-4 6.0-4 - 1.0-5 Ba-140 9.0-6 1.1-2 - 1.1-5 Ce-141 2.8-5 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 43 - H-3 47 - - - NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 20F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (4) This historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-57 October, 2015 TABLE 11.3-9c CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 1(4) (Ci/year) Unit 1 Unit 1 Unit 1 Unit 1 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 120 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 9 See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 23 See Note(1) Xe-133m See Note(1) See Note(1) See Note(1) See Note(1) Xe-133 142 250 840 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 113 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.9-2 1.9-1 See Note(1) 3.0-2 I-133 1.5-1 7.6-1 See Note(1) See Note(1) Cr-51 9.6-5 1.3-2 - See Note(1) Mn-54 3.6-4 6.0-4 - See Note(1) Fe-59 1.6-4 5.0-4 - See Note(1) Co-58 5.7-5 6.0-4 - See Note(1) Co-60 1.1-3 2.0-3 - See Note(1) Zn-65 5.5-5 2.0-4 - See Note(1) Sr-89 6.3-6 6.0-3 - See Note(1) Sr-90 3.1-6 2.0-5 - See Note(1) Zr-95 8.5-6 1.0-4 - See Note(1) Sb-124 4.7-6 3.0-4 - See Note(1) Cs-134 1.3-4 3.0-4 - 3.0-6 Cs-136 1.1-5 5.0-5 - 2.0-6 Cs-137 2.0-4 6.0-4 - 1.0-5 Ba-140 9.0-6 1.1-2 - 1.1-5 Ce-141 2.8-5 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 83 - H-3 47 - - - NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 40F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (4) This historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-58 October, 2015 TABLE 11.3-9d CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 1(4) (Ci/year) Unit 1 Unit 1 Unit 1 Unit 1 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 990 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 260 See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 51 See Note(1) Xe-133m See Note(1) See Note(1) 3 See Note(1) Xe-133 142 250 5,100 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 113 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.9-2 1.9-1 See Note(1) 3.0-2 I-133 1.5-1 7.6-1 See Note(1) See Note(1) Cr-51 9.6-5 1.3-2 - See Note(1) Mn-54 3.6-4 6.0-4 - See Note(1) Fe-59 1.6-4 5.0-4 - See Note(1) Co-58 5.7-5 6.0-4 - See Note(1) Co-60 1.1-3 2.0-3 - See Note(1) Zn-65 5.5-5 2.0-4 - See Note(1) Sr-89 6.3-6 6.0-3 - See Note(1) Sr-90 3.1-6 2.0-5 - See Note(1) Zr-95 8.5-6 1.0-4 - See Note(1) Sb-124 4.7-6 3.0-4 - See Note(1) Cs-134 1.3-4 3.0-4 - 3.0-6 Cs-136 1.1-5 5.0-5 - 2.0-6 Cs-137 2.0-4 6.0-4 - 1.0-5 Ba-140 9.0-6 1.1-2 - 1.1-5 Ce-141 2.8-5 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 160 - H-3 47 - - - NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 70F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (4) This historical analysis remains bounding based on including Unit 2 operation because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. Revision 19 11.3-59 October, 2015 TABLE 11.3-10a CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 2(4) (Ci/year) Unit 2 Unit 2 Unit 2(5) Unit 2 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 7 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 See Note(1) See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 8 See Note(1) Xe-133m See Note(1) See Note(1) See Note(1) See Note(1) Xe-133 132 250 85 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 68 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.4-2 1.9-1 See Note(1) 3.0-2 I-133 1.4-1 7.6-1 See Note(1) See Note(1) Cr-51 6.0-6 1.3-2 - See Note(1) Mn-54 6.0-5 6.0-4 - See Note(1) Fe-59 8.0-6 5.0-4 - See Note(1) Co-58 1.2-5 6.0-4 - See Note(1) Co-60 2.0-4 2.0-3 - See Note(1) Zn-65 4.0-5 2.0-4 - See Note(1) Sr-89 1.8-6 6.0-3 - See Note(1) Sr-90 1.0-7 2.0-5 - See Note(1) Zr-95 8.0-6 1.0-4 - See Note(1) Sb-124 4.0-6 3.0-4 - See Note(1) Cs-134 8.0-5 3.0-4 - 3.0-6 Cs-136 6.0-6 5.0-5 - 2.0-6 Cs-137 1.1-4 6.0-4 - 1.0-5 Ba-140 8.0-6 1.1-2 - 1.1-5 Ce-141 2.0-6 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 13 - H-3 47 - - - Revision 19 11.3-60 October, 2015 TABLE 11.3-10a (Continued) NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 0F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% <NUREG-0016>. (4) The information in this table was calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains valid because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. (5) Unit 2 plant vent has active inputs from Unit 1. Revision 19 11.3-60a October, 2015 TABLE 11.3-10b CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 2(4) (Ci/year) Unit 2 Unit 2 Unit 2(5) Unit 2 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 13 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 See Note(1) See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 10 See Note(1) Xe-133m See Note(1) See Note(1) See Note(1) See Note(1) Xe-133 132 250 130 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 68 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.4-2 1.9-1 See Note(1) 3.0-2 I-133 1.4-1 7.6-1 See Note(1) See Note(1) Cr-51 6.0-6 1.3-2 - See Note(1) Mn-54 6.0-5 6.0-4 - See Note(1) Fe-59 8.0-6 5.0-4 - See Note(1) Co-58 1.2-5 6.0-4 - See Note(1) Co-60 2.0-4 2.0-3 - See Note(1) Zn-65 4.0-5 2.0-4 - See Note(1) Sr-89 1.8-6 6.0-3 - See Note(1) Sr-90 1.0-7 2.0-5 - See Note(1) Zr-95 8.0-6 1.0-4 - See Note(1) Sb-124 4.0-6 3.0-4 - See Note(1) Cs-134 8.0-5 3.0-4 - 3.0-6 Cs-136 6.0-6 5.0-5 - 2.0-6 Cs-137 1.1-4 6.0-4 - 1.0-5 Ba-140 8.0-6 1.1-2 - 1.1-5 Ce-141 2.0-6 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 43 - H-3 47 - - - Revision 19 11.3-61 October, 2015 TABLE 11.3-10b (Continued) NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 20F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% <NUREG-0016>. (4) The information in this table was calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains valid because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. (5) Unit 2 plant vent has active inputs from Unit 1. Revision 19 11.3-61a October, 2015 TABLE 11.3-10c CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 2(4) (Ci/year) Unit 2 Unit 2 Unit 2(5) Unit 2 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 120 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 9 See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 23 See Note(1) Xe-133m See Note(1) See Note(1) See Note(1) See Note(1) Xe-133 132 250 840 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 68 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.4-2 1.9-1 See Note(1) 3.0-2 I-133 1.4-1 7.6-1 See Note(1) See Note(1) Cr-51 6.0-6 1.3-2 - See Note(1) Mn-54 6.0-5 6.0-4 - See Note(1) Fe-59 8.0-6 5.0-4 - See Note(1) Co-58 1.2-5 6.0-4 - See Note(1) Co-60 2.0-4 2.0-3 - See Note(1) Zn-65 4.0-5 2.0-4 - See Note(1) Sr-89 1.8-6 6.0-3 - See Note(1) Sr-90 1.0-7 2.0-5 - See Note(1) Zr-95 8.0-6 1.0-4 - See Note(1) Sb-124 4.0-6 3.0-4 - See Note(1) Cs-134 8.0-5 3.0-4 - 3.0-6 Cs-136 6.0-6 5.0-5 - 2.0-6 Cs-137 1.1-4 6.0-4 - 1.0-5 Ba-140 8.0-6 1.1-2 - 1.1-5 Ce-141 2.0-6 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 83 - H-3 47 - - - Revision 19 11.3-62 October, 2015 TABLE 11.3-10c (Continued) NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 40F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% <NUREG-0016>. (4) The information in this table was calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains valid because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. (5) Unit 2 plant vent has active inputs from Unit 1. Revision 19 11.3-62a October, 2015 TABLE 11.3-10d CALCULATED RELEASE OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS - UNIT 2(4) (Ci/year) Unit 2 Unit 2 Unit 2(5) Unit 2 Offgas Mech. Vac. Nuclide Plant Vent Turbine Bldg. Bldg. Vent(2)(3) Pump Discharge Kr-83m See Note(1) See Note(1) See Note(1) See Note(1) Kr-85m 6 68 990 See Note(1) Kr-85 See Note(1) See Note(1) 260 See Note(1) Kr-87 6 130 See Note(1) See Note(1) Kr-88 6 230 260 See Note(1) Kr-89 See Note(1) See Note(1) See Note(1) See Note(1) Xe-131m See Note(1) See Note(1) 51 See Note(1) Xe-133m See Note(1) See Note(1) 3 See Note(1) Xe-133 132 250 5,100 2,300 Xe-135m 92 650 See Note(1) See Note(1) Xe-135 68 630 See Note(1) 350 Xe-137 See Note(1) See Note(1) See Note(1) See Note(1) Xe-138 14 1,400 See Note(1) See Note(1) I-131 3.4-2 1.9-1 See Note(1) 3.0-2 I-133 1.4-1 7.6-1 See Note(1) See Note(1) Cr-51 6.0-6 1.3-2 - See Note(1) Mn-54 6.0-5 6.0-4 - See Note(1) Fe-59 8.0-6 5.0-4 - See Note(1) Co-58 1.2-5 6.0-4 - See Note(1) Co-60 2.0-4 2.0-3 - See Note(1) Zn-65 4.0-5 2.0-4 - See Note(1) Sr-89 1.8-6 6.0-3 - See Note(1) Sr-90 1.0-7 2.0-5 - See Note(1) Zr-95 8.0-6 1.0-4 - See Note(1) Sb-124 4.0-6 3.0-4 - See Note(1) Cs-134 8.0-5 3.0-4 - 3.0-6 Cs-136 6.0-6 5.0-5 - 2.0-6 Cs-137 1.1-4 6.0-4 - 1.0-5 Ba-140 8.0-6 1.1-2 - 1.1-5 Ce-141 2.0-6 6.0-4 - See Note(1) C-14 - - 9.5 - Ar-41 25 - 160 - H-3 47 - - - Revision 19 11.3-63 October, 2015 TABLE 11.3-10d (Continued) NOTES: (1) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (2) Charcoal temperature = 70F. (3) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% <NUREG-0016>. (4) The information in this table was calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made not to complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains valid because input parameters used for calculating gaseous releases for a single unit are less than that attained by dual unit operation. (5) Unit 2 plant vent has active inputs from Unit 1. Revision 19 11.3-63a October, 2015 TABLE 11.3-11a AVERAGE ANNUAL CONCENTRATIONS OF GASEOUS EFFLUENTS AT EXCLUSION BOUNDARY(5) Fraction of Annual Release MPC <10 CFR 20, Appendix B> Effluent Nuclide (Ci/yr - two units)(3)(4) (Ci/cc) Fraction of MPC(1) Effluent Concentrations (Ci/cc) Concentrations(1) Kr-83m See Note(2) 3.-8 - 5E-5 - Kr-85m 162 1.-7 1.4-4 1.-7 1.4-4 Kr-85 520 3.-7 1.5-4 7E-7 6.4E-5 Kr-87 272 2.-8 1.2-3 2.-8 1.2-3 Kr-88 472 2.-8 2.0-3 9E-9 4.4E-3 Kr-89 See Note(2) 3.-8 - 1E-9 - Xe-131m 16 4.-7 3.4-6 2E-6 6.8E-7 Xe-133m See Note(2) 3.-7 - 6E-7 - Xe-133 5,544 3.-7 1.6-3 5E-7 9.6E-4 Xe-135m 1,484 3.-8 4.2-3 4E-8 3.1E-3 Xe-135 2,141 1.-7 1.8-3 7E-8 2.6E-3 Xe-137 See Note(2) 3.-8 - 1E-9 - Xe-138 2,828 3.-8 8.1-3 2E-8 1.2E-2 I-131 .51 1.-8 4.4-6 2E-10 2.2E-4 I-133 1.8 7.-9 2.2-5 1E-9 1.5E-4 Cr-51 1.3-2 8.-8 1.4-8 3E-8 3.7E-8 Mn-54 1.6-3 1.-9 1.4-7 1.-9 1.4-7 Fe-59 1.2-3 2.-9 5.1-8 5E-10 2.0E-7 Co-58 1.3-3 2.-9 5.6-8 1E-9 1.1E-7 Co-60 5.3-3 3.-10 1.5-6 5E-11 9.0E-6 Zn-65 5.0-4 2.-9 2.1-8 4E-10 1.0E-7 Sr-89 1.2-2 1.-9 1.0-6 2E-10 5.0E-6 Sr-90 4.3-5 2.-10 1.8-8 6E-12 6.0E-7 Zr-95 2.2-4 1.-9 1.9-8 4E-10 4.7E-8 Sb-124 6.1-4 3.-8 1.7-9 3E-10 1.7E-7 Cs-134 8.1-4 4.-10 1.7-7 2E-10 3.4E-7 Cs-136 1.2-4 6.-9 1.7-9 9E-10 1.1E-8 Cs-137 1.5-3 5.-10 2.6-7 2E-10 6.5E-7 Ba-140 2.2-2 1.-9 1.9-6 2E-9 9.5E-7 Ce-141 1.2-3 5.-9 2.1-8 8E-10 1.3E-7 C-14 19 1.-7 1.6-5 3E-9 5.3E-4 Ar-41 76 4.-8 1.6-4 1E-8 6.4E-4 H-3 94 2.-7 4.0-5 1E-7 8.0E-5 NOTES: (1) Based on an average annual /Q of 2.7-6 sec/m3. (2) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (3) Offgas system charcoal temperature = 0F. (4) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (5) The Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary data provided in Table 11.3-11a through 11.3-11d were calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made to not complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because single unit annual releases are less than those attained by dual unit operation. Revision 19 11.3-64 October, 2015 TABLE 11.3-11b AVERAGE ANNUAL CONCENTRATIONS OF GASEOUS EFFLUENTS AT EXCLUSION BOUNDARY(5) Fraction of Annual Release MPC <10 CFR 20, Appendix B> Effluent Nuclide (Ci/yr - two units)(3)(4) (Ci/cc) Fraction of MPC(1) Effluent Concentrations (Ci/cc) Concentrations(1) Kr-83m See Note(2) 3.-8 - 5E-5 - Kr-85m 174 1.-7 1.5-4 1.-7 1.5-4 Kr-85 520 3.-7 1.5-4 7E-7 6.4E-5 Kr-87 272 2.-8 1.2-3 2.-8 1.2-3 Kr-88 472 2.-8 2.0-3 9E-9 4.4E-3 Kr-89 See Note(2) 3.-8 - 1E-9 - Xe-131m 20 4.-7 4.3-6 2E-6 8.6E-7 Xe-133m See Note(2) 3.-7 - 6E-7 - Xe-133 5,634 3.-7 1.6-3 5E-7 9.6E-4 Xe-135m 1,484 3.-8 4.2-3 4E-8 3.1E-3 Xe-135 2,141 1.-7 1.8-3 7E-8 2.6E-3 Xe-137 See Note(2) 3.-8 - 1E-9 - Xe-138 2,828 3.-8 8.1-3 2E-8 1.2E-2 I-131 .51 1.-8 4.4-6 2E-10 2.2E-4 I-133 1.8 7.-9 2.2-5 1E-9 1.5E-4 Cr-51 1.3-2 8.-8 1.4-8 3E-8 3.7E-8 Mn-54 1.6-3 1.-9 1.4-7 1.-9 1.4-7 Fe-59 1.2-3 2.-9 5.1-8 5E-10 2.0E-7 Co-58 1.3-3 2.-9 5.6-8 1E-9 1.1E-7 Co-60 5.3-3 3.-10 1.5-6 5E-11 9.0E-6 Zn-65 5.0-4 2.-9 2.1-8 4E-10 1.0E-7 Sr-89 1.2-2 1.-9 1.0-6 2E-10 5.0E-6 Sr-90 4.3-5 2.-10 1.8-8 6E-12 6.0E-7 Zr-95 2.2-4 1.-9 1.9-8 4E-10 4.7E-8 Sb-124 6.1-4 3.-8 1.7-9 3E-10 1.7E-7 Cs-134 8.1-4 4.-10 1.7-7 2E-10 3.4E-7 Cs-136 1.2-4 6.-9 1.7-9 9E-10 1.1E-8 Cs-137 1.5-3 5.-10 2.6-7 2E-9 6.5E-7 Ba-140 2.2-2 1.-9 1.9-6 8E-10 9.5E-7 Ce-141 1.2-3 5.-9 2.1-8 3E-9 1.3E-7 C-14 19 1.-7 1.6-5 1E-8 5.3E-4 Ar-41 136 4.-8 2.9-4 1E-8 1.2E-3 H-3 94 2.-7 4.0-5 1E-7 8.0E-5 NOTES: (1) Based on an average annual /Q of 2.7-6 sec/m3. (2) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (3) Offgas system charcoal temperature = 20F. (4) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (5) The Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary data provided in Table 11.3-11a through 11.3-11d were calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made to not complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because single unit annual releases are less than those attained by dual unit operation. Revision 19 11.3-65 October, 2015 TABLE 11.3-11c AVERAGE ANNUAL CONCENTRATIONS OF GASEOUS EFFLUENTS AT EXCLUSION BOUNDARY(5) Fraction of Annual Release MPC <10 CFR 20, Appendix B> Effluent Nuclide (Ci/yr - two units)(3)(4) (Ci/cc) Fraction of MPC(1) Effluent Concentrations (Ci/cc) Concentrations(1) Kr-83m See Note(2) 3.-8 - 5E-5 - Kr-85m 388 1.-7 3.3-4 1.-7 3.3-4 Kr-85 520 3.-7 1.5-4 7E-7 6.4E-5 Kr-87 272 2.-8 1.2-3 2.-8 1.2-3 Kr-88 490 2.-8 2.1-3 9E-9 4.7E-3 Kr-89 See Note(2) 3.-8 - 1E-9 - Xe-131m 46 4.-7 9.8-6 2E-6 2.0E-6 Xe-133m See Note(2) 3.-7 - 6E-7 - Xe-133 7,054 3.-7 2.0-3 5E-7 1.2E-3 Xe-135m 1,484 3.-8 4.2-3 4E-8 3.1E-3 Xe-135 2,141 1.-7 1.8-3 7E-8 2.6E-3 Xe-137 See Note(2) 3.-8 - 1E-9 - Xe-138 2,828 3.-8 8.1-3 2E-8 1.2E-2 I-131 .51 1.-8 4.4-6 2E-10 2.2E-4 I-133 1.8 7.-9 2.2-5 1E-9 1.5E-4 Cr-51 1.3-2 8.-8 1.4-8 3E-8 3.7E-8 Mn-54 1.6-3 1.-9 1.4-7 1.-9 1.4-7 Fe-59 1.2-3 2.-9 5.1-8 5E-10 2.0E-7 Co-58 1.3-3 2.-9 5.6-8 1E-9 1.1E-7 Co-60 5.3-3 3.-10 1.5-6 5E-11 9.0E-6 Zn-65 5.0-4 2.-9 2.1-8 4E-10 1.0E-7 Sr-89 1.2-2 1.-9 1.0-6 2E-10 5.0E-6 Sr-90 4.3-5 2.-10 1.8-8 6E-12 6.0E-7 Zr-95 2.2-4 1.-9 1.9-8 4E-10 4.7E-8 Sb-124 6.1-4 3.-8 1.7-9 3E-10 1.7E-7 Cs-134 8.1-4 4.-10 1.7-7 2E-10 3.4E-7 Cs-136 1.2-4 6.-9 1.7-9 9E-10 1.1E-8 Cs-137 1.5-3 5.-10 2.6-7 2E-10 6.5E-7 Ba-140 2.2-2 1.-9 1.9-6 2E-9 9.5E-7 Ce-141 1.2-3 5.-9 2.1-8 8E-10 1.3E-7 C-14 19 1.-7 1.6-5 3E-9 5.3E-4 Ar-41 216 4.-8 4.6-4 1E-8 1.8E-3 H-3 94 2.-7 4.0-5 1E-7 8.0E-5 NOTES: (1) Based on an average annual /Q of 2.7-6 sec/m3. (2) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (3) Offgas system charcoal temperature = 40F. (4) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (5) The Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary data provided in Table 11.3-11a through 11.3-11d were calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made to not complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because single unit annual releases are less than those attained by dual unit operation. Revision 19 11.3-66 October, 2015 TABLE 11.3-11d AVERAGE ANNUAL CONCENTRATIONS OF GASEOUS EFFLUENTS AT EXCLUSION BOUNDARY(5) Fraction of Annual Release MPC <10 CFR 20, Appendix B> Effluent Nuclide (Ci/yr - two units)(3)(4) (Ci/cc) Fraction of MPC(1) Effluent Concentrations (Ci/cc) Concentrations(1) Kr-83m See Note(2) 3.-8 5.7-6 5E-5 3.4E-9 Kr-85m 2,128 1.-7 1.8-3 1.-7 1.8-3 Kr-85 520 3.-7 1.5-4 7E-7 6.4E-5 Kr-87 272 2.-8 1.2-3 2.-8 1.2-3 Kr-88 992 2.-8 4.2-3 9E-9 9.3E-3 Kr-89 See Note(2) 3.-8 - 1E-9 - Xe-131m 102 4.-7 2.2-5 2E-6 4.4E-6 Xe-133m 6 3.-7 1.7-6 6E-7 8.5E-7 Xe-133 15,574 3.-7 4.4-3 5E-7 2.6E-3 Xe-135m 1,484 3.-8 4.2-3 4E-8 3.1E-3 Xe-135 2,141 1.-7 1.8-3 7E-8 2.6E-3 Xe-137 See Note(2) 3.-8 - 1E-9 - Xe-138 2,828 3.-8 8.1-3 2E-8 1.2E-2 I-131 .51 1.-8 4.4-6 2E-10 2.2E-4 I-133 1.8 7.-9 2.2-5 1E-9 1.5E-4 Cr-51 1.3-2 8.-8 1.4-8 3E-8 3.7E-8 Mn-54 1.6-3 1.-9 1.4-7 1.-9 1.4-7 Fe-59 1.2-3 2.-9 5.1-8 5E-10 2.0E-7 Co-58 1.3-3 2.-9 5.6-8 1E-9 1.1E-7 Co-60 5.3-3 3.-10 1.5-6 5E-11 9.0E-6 Zn-65 5.0-4 2.-9 2.1-8 4E-10 1.0E-7 Sr-89 1.2-2 1.-9 1.0-6 2E-10 5.0E-6 Sr-90 4.3-5 2.-10 1.8-8 6E-12 6.0E-7 Zr-95 2.2-4 1.-9 1.9-8 4E-10 4.7E-8 Sb-124 6.1-4 3.-8 1.7-9 3E-10 1.7E-7 Cs-134 8.1-4 4.-10 1.7-7 2E-10 3.4E-7 Cs-136 1.2-4 6.-9 1.7-9 9E-10 1.1E-8 Cs-137 1.5-3 5.-10 2.6-7 2E-10 6.5E-7 Ba-140 2.2-2 1.-9 1.9-6 2E-9 9.5E-7 Ce-141 1.2-3 5.-9 2.1-8 8E-10 1.3E-7 C-14 19 1.-7 1.6-5 3E-9 5.3E-4 Ar-41 370 4.-8 7.9-4 1E-8 3.2E-3 H-3 94 2.-7 4.0-5 1E-7 8.0E-5 NOTES: (1) Based on an average annual /Q of 2.7-6 sec/m3. (2) Less than 1 Ci/yr noble gas, less than 10-4 Ci/yr iodine. (3) Offgas system charcoal temperature = 70F. (4) Design based on 51,300 Ci/sec at T = 30 min, plant capacity factor, 80% per <NUREG-0016>. (5) The Average Annual Concentrations of Gaseous Effluents at Exclusion Boundary data provided in Table 11.3-11a through 11.3-11d were calculated with the expectation that Unit 2 would achieve commercial operation. A later decision was made to not complete Unit 2 and subsequently, the unit was abandoned. However, this historical analysis remains bounding based on including Unit 2 operation because single unit annual releases are less than those attained by dual unit operation. Revision 19 11.3-67 October, 2015 11.4 SOLID RADIOACTIVE WASTE MANAGEMENT SYSTEM 11.4.1 DESIGN BASES 11.4.1.1 Power Generation Design Objectives The primary design objective of the solid radioactive waste (SRW) system is to control, collect, handle, process, and package all wet and dry solid radioactive waste generated by PNPP as a result of normal operation, and to store these wastes until they are shipped to authorized receiving and storage areas offsite. This will be done in such a manner that, for all anticipated quantities of waste produced, the availability of the power plant for power generation will not be adversely affected. The types of solid radioactive waste to be processed, anticipated quantities and curie content are given in
for the original design basis of the solid radioactive waste system. Waste quantities and curie content by isotope are given in
,
, and
. Subsequent to the original design of the SRW system, modifications have been made to the condensate cleanup and liquid radwaste systems which could result in quantities or activities of solid radioactive waste which are different than those in the original design basis. The control, processing and packaging of the solid radioactive waste remains unchanged. The ALARA design features, as discussed in <Section 11.4.1.5>, and the safety precautions, as discussed in <Section 11.4.1.6> are unaffected by the changes to the quantities or activities of the waste to be processed. Revision 16 11.4-1 October, 2009 11.4.1.2 Radiological Design Objectives Packaging of solid radioactive material is accomplished in a manner which ensures that no radioactive material will be released to the environment during shipment of the waste to offsite burial or storage facilities. The SRW system is designed to limit exposures to both operating personnel and the general public to as low as reasonably achievable. 11.4.1.3 Design Criteria
a. The SRW system components, piping and the structure that houses the system are designed and fabricated in accordance with the codes, standards, seismic categories, and quality group classifications given in
.
b. The SRW system design is in compliance with the guidance provided by <Regulatory Guide 1.143> and Branch Technical Position ETSB 11-3.
c. All wet radioactive waste (filter backwash slurries and spent resins) are processed per the Process Control Program (Reference 1) prior to shipment offsite. Packaging and transporting of radioactive wastes is performed in conformance with <10 CFR 71> and applicable ICC and DOT regulations.
d. The SRW system design and shielding provisions ensure that (during all phases of processing, handling and shipment of radioactive waste) exposure to operating personnel and the general public is within the applicable limits of <10 CFR 20>, <49 CFR 173> and as low as is reasonably achievable in accordance with <Regulatory Guide 8.8>.
Revision 17 11.4-2 October, 2011
e. The Process Control Program (Reference 1) provides a means to verify the absence of free liquid in the containers in accordance with Branch Technical Position ETSB 11-3.
f. The SRW system design, equipment sizing and equipment redundancy ensure that the maximum expected quantities of all radioactive waste inputs during any 30 day period can be prepared for shipping via the Process Control Program (Reference 1) and temporarily stored onsite without affecting plant availability. Design quantities of radioactive waste inputs to the SRW system are presented in
. 11.4.1.4 Component Design Parameters With the exception of normal wearing parts, such as seals and bearings, all pumps, valves, piping, tanks, and other components in the SRW system are fabricated from materials which are intended to provide a minimum service life of 40 years without replacement. In selecting materials to meet this criterion, due consideration is given to: a) the corrosive nature of both the process medium and the external environment, b) decontaminability of the material, and c) wall thickness requirements dictated by design pressures, flow rates and corrosion rates. The design classifications of SRW system equipment items are given in
. 11.4.1.5 ALARA Design Features Numerous features have been incorporated into the design of both the SRW system and the building housing this system to ensure that exposures of operating personnel to radiation will be kept within ALARA guidelines. See <Section 11.2.1.9> for a listing of the most significant ALARA design features. Revision 13 11.4-3 December, 2003 11.4.1.6 Safety Precautions All tanks, pumps and other equipment containing radioactive liquids are located in shielded cubicles or pipe chases. All access to these areas is strictly controlled by administrative procedures. 11.4.2 SYSTEM DESCRIPTION 11.4.2.1 Treatment of Wet Solid Radioactive Waste NOTE: Mobile radioactive waste processing is used in combination with portions of the SRW system described in this Chapter. The details of the mobile SRW package and its interface with the SRW processing described in this chapter are contained in the Perry Process Control Program (Reference 1). The types, anticipated quantities and expected activity levels of wet solid radioactive waste to be processed are identified in
. The system diagram is presented in <Figure 11.4-1>. This diagram shows the process flow routes, process flow conditions, equipment capacities, instrumentation, and system design data. Instrumentation, controls, alarms, and protection devices are discussed under <Section 11.4.2.4>. The SRW system is designed to process spent resin slurry, precoat-type filter backwash slurry, and traveling belt discharge cake. These waste streams are transferred from the LRW system Revision 16 11.4-4 October, 2009 collection tanks to a vendors dewatering system. After this transfer, the fill isolation valve is closed and the fill line is backflushed to the tank from which the waste stream originated. Processing of the waste is controlled from the SRW control panel and vendor system control panel. Using selector switches on this panel, the operator selects which waste storage tank to take waste from filling the waste container is controlled from the vendor system control panel. The method of waste processing is detailed in the Process Control Program. Once onsite processing of the waste is complete, the overhead bridge crane picks up the container and takes it either to a short term storage area or to a truck bay where it is loaded for transfer to an authorized receiving, reprocessing, or storage area. 11.4.2.1.1 Component Failure and System Malfunctions The SRW system is designed to preclude the accidental release of radioactive waste into the solid waste packaging area due to component failure or system malfunctions. Instrumentation and controls monitor each phase of the packaging operation, serving to detect possible system malfunctions and terminate the packaging operation as required to prevent inadvertent releases of radioactive waste into the solid waste packaging area. Full operator surveillance is maintained during the entire packaging operation through CCTV monitors located on and adjacent to the control panel. Means are provided for the operator to terminate the Revision 16 11.4-5 October, 2009 packaging operation in instances of component failures which may cause the release of radioactive materials from the SRW system. The possibility of component failures is considered very low because of the low pressures at which the packaging operation occurs. The interface between the vendors system and permanent plant equipment is evaluated for accidental releases of radioactivity before the mobile system is approved for use. The air flow patterns in the drumming station are such that any radioactive gases released would pass into the radwaste ventilation system, and be treated by a series of roughing, HEPA and charcoal filters prior to release to the environs <Figure 9.4-7>. 11.4.2.2 Treatment of Dry Solid Radioactive Waste A dry solid radwaste subsystem is provided for processing dry filter media (ventilation filters), contaminated clothing, equipment, tools and glassware, and miscellaneous radioactive wastes that are not amenable to solidification prior to packaging. Potentially radioactively contaminated waste and radioactive material such as tooling, components and equipment are collected throughout the RRA and brought to one of two main areas: the Waste Abatement and Reclamation Facility (WARF) or the DAW handling area on the 623-6 elevation of the radwaste building. Other areas may be established temporarily based on operational needs as determined by the Radiation Protection Manager (RPM). The types, anticipated quantities and expected activity levels of dry solid radioactive wastes are identified in
. These numbers are based on operating plant data. Revision 18 11.4-6 October, 2013 11.4.2.2.1 Compressible Dry Solid Radioactive Waste Radioactive Material, contaminated cloth, paper, glass, floor sweepings, and similar low-level activity wastes are accumulated in designated storage areas. The waste is stored until a sufficient amount accumulates to warrant shipment to an authorized processing and storage area located offsite. All radioactive material and stored DAW are kept in metal containers or areas protected by fire suppression systems while onsite. 11.4.2.2.2 Incompressible Dry Solid Radioactive Waste Spent filter cartridges, air filter elements, contaminated tools, and similar incompressible solid wastes are packaged in various size shipping containers depending on their size. Shielding is provided around the shipping container as required. Highly radioactive material is centered in the shipping container and solidification agent is added, thus providing additional shielding. 11.4.2.2.3 Segregation of Clean and Contaminated Loose Wastes Normally segregation of clean and contaminated loose wastes is contracted to a licensed offsite vendor. Potentially contaminated waste will be monitored for radioactivity levels above background before disposal as clean. Material exhibiting any level of radioactivity above background, as demonstrated by the use of this equipment (or other equipment utilizing the same type of sensitive monitors) will either be decontaminated or disposed of as radioactive waste. An aggregate of this sorted clean waste and other clean waste from the RRA will be monitored before disposal as clean. This program is in compliance with <NRC Notice 85-92>. Revision 18 11.4-7 October, 2013 11.4.2.3 Detailed Component Design All items under this Section address the permanent plant equipment that will interface with a vendors mobile system.
a. Collection Tank Design These tanks are treated as a part of the LRW system; refer to
<Section 11.2.2.10.a>, for this information.
b. General Pump Design All pumps, whether centrifugal or positive displacement, are designed to the requirements of the Hydraulic Institute Standards for rating, testing, application, and materials. For pumps handling radioactive fluids, shafts are sealed with mechanical seals which are balanced, single (or double if process fluid necessitates) seals with a carbon stationary insert, ceramic rotating seal ring, silicone or EPR elastomer O-rings, 316L SS metal parts, flushing connection, vent and drain connection, and throttle bushing (for single mechanical seals only). The vent and drain connections and the throttle bushings are provided to permit installation of a drain for the fluid that leaks from a worn seal.
The bearing lubrication that may leak out of the lubrication system will be allowed to accumulate on the pump base separate from the pump shaft seal drain piping. A solenoid operated shutoff valve is provided for control of seal water to each pump with mechanical seals. This valve is designed to open when the pump is started, to close when the pump is stopped, and to fail open on loss of power.
c. Waste Mixing/Dewatering Tanks Two redundant mixing/dewatering tanks are provided in shielded cubicles at Elevation 630-0. Each tank is an atmospheric, Revision 13 11.4-8 December, 2003
750 gallon, vertical, cone bottomed, 316L stainless steel vessel mounted off the floor on carbon steel support legs. Connections are provided for vent/overflow, concentrate and slurry waste feeds, flushwater, level monitors, traveling belt filter chute discharge, dewatering, and drain. A tank mixer is mounted on top of the tank and is controlled from the SRW control panel. A manway with hinged cover is also located on top of the tank. Inside the tank are the tank washdown nozzles, mixer blades all constructed of 316 or 316L stainless steel.
d. Waste Dewatering Pumps The dewatering pump is mounted on a base plate attached to the legs of the mixing/dewatering tank. It is a 10 gpm, motor driven centrifugal pump, controlled from the SRW control panel. The pump has two suction connections. The upper connection is not used.
When used to drain the tank, it takes suction from a connection near the bottom of the tank. The dewatering pump is constructed of 316 stainless steel. Pump seals are single mechanical type. Revision 16 11.4-9 October, 2009
e. Waste Feed Pumps The waste feed pump is mounted on a skid plate attached to the legs of the mixing/dewatering tank. It is a progressing cavity, positive displacement, metering pump built to food industry standards. It is driven by an SCR variable speed, dc motor and is controlled from the SRW control panel. The SCR controller can be reset to adjust the pump flow rate from 15 to 40 gpm. Portions of the feed pump in contact with radioactive fluids are constructed of 316L stainless steel. Seals are double mechanical type.
f. Overhead Bridge Crane The bridge crane has a rated capacity of 15 tons and a span of 34-3. It is mounted on 60 pound ASCE rails that allow full travel of the crane in the north-south direction between column lines RW-A and RW-D, permitting full access to the truck bay, temporary storage facility and processing gallery.
The unit is controlled entirely from the SRW control panel. A 3-position digital indexing and readout system on the control panel indicates where the bridge, trolley and hoist are at all times. In addition to this system, the operator can view all movements of the crane on a closed circuit TV monitor. For maintenance purposes, a local control station is provided, with controls for bridge, hoist and trolley. The bridge, trolley and hoist have both high and low speeds; the former is for rough positioning and the latter is for accurate final positioning. High/low speeds for the bridge, trolley and hoist are approximately 58/5.8, 50/5.0 and 22.5/2.25 fpm, respectively. The bridge and trolley drives have full magnetic soft start electric starting controls to minimize drive wear. The Revision 16 11.4-10 October, 2009 crane travel controls are such that when the load is not fully up, the bridge and trolley are permitted to move only when the hoist override control switch is turned On. Bridge rail end stops are provided to limit travel of the bridge so that the load cannot hit the end walls. All necessary controls, relays, etc., for controlling a power-operated container uprighting mechanism are wired into the bridge crane and control panel for use in the event that one is purchased for future use.
g. Shipping Containers Normally, large containers will be used as shipping containers for processed waste. The exact size varies from vendor to vendor.
Standard DOT 17H steel drums and steel boxes that comply with DOT Industrial Packaging requirements are used as shipping containers for compacted or non-compacted waste. Revision 18 11.4-11 October, 2013 (INTENTIONALLY BLANK) Revision 18 11.4-12 October, 2013
h. Solidified Waste Storage Vault A shielded area located adjacent to the radwaste truck bay on the 620 elevation of the radwaste building, measuring 50-6 long by 25-6 wide by 13-4 high (usable height) is used to provide temporary onsite storage of packaged waste. This allows for further decay time and lessens the effect on plant operations of such events as a truckers strike or temporary shutdown of a burial site.
i. Interim/Temporary Storage of Radioactive Wastes The interim/temporary storage of radioactive wastes were evaluated for compliance with <Generic Letter 81-38>. Radioactive waste collected onsite awaiting disposal off-site is called temporary storage or staged waste. Any radioactive waste remaining on-site greater than 90 days will have waste form and container selection considered for impact and must comply with <Generic Letter 81-38>.
Radioactive waste may be stored anywhere within the generating facility, under the direction and approval of the Radiation Revision 18 11.4-13 October, 2013 Protection Manager. A Radiation Protection Instruction (RPI) outlines the methods and protocol for storing and inspecting radioactive waste outside the RRA and/or the generating facility. 11.4.2.4 Instrumentation, Controls, Alarms, and Protective Devices 11.4.2.4.1 Controls The SRW system is controlled entirely from the SRW panel and an adjacent control panel during all normal operations. The control panel is equipped with the following: a semi-graphic display of the processing system; control switches for all normally used valves; control switches for all motor driven equipment; status lights for all power operated valves, pumps, indexing controls and readout for the bridge crane; readout of certain process parameters (levels); add a CCTV monitoring system of the processing gallery, storage vault and truck bay. The control panel contains a solid state programmable controller to control the processing system. In general terms, the process is controlled as explained in the following paragraphs: To begin filling a waste container the operator verifies that the vendor system is installed properly and all necessary connections have been made to the permanent plant systems. An empty waste container is placed in the waste process area and the vendor dewatering system fill head is placed on the waste container. Valves are aligned to the LRW system necessary to select one of the following waste streams; spent resin, filter/demineralizer sludge or RWCU sludge. Filling of the waste container is controlled from the vendor system control panel. After filling the waste container the process lines are backflushed to the LRW tank from which the waste container was filled. Revision 16 11.4-14 October, 2009
a. (Deleted)
b. (Deleted)
c. (Deleted)
Revision 16 11.4-15 October, 2009 11.4.2.4.2 Instrumentation
a. Waste Mixing/Dewatering Tanks Each of these tanks has an ultrasonic level transmitter for level readout and high/low alarms on the control panel, and for control interlocking functions. A back-up high level probe is provided for control interlocking.
b. Waste Feed Pumps Each discharge line has a magnetic flowmeter for flow recording on the control panel.
11.

4.3 REFERENCES

FOR SECTION 11.4.

1. Letter from M. R. Edelmen, CEI, to B. J. Youngblood, NRC, August 28, 1985 (PY-CEI/NRR-0329L).

Revision 16 11.4-16 October, 2009

TABLE 11.4-1 MAXIMUM MONTHLY RADIOACTIVE WASTE INPUTS TO SOLID RADIOACTIVE WASTE SYSTEM (8)

Quantity Maximum Activity Method Container Vol. per Max. Batches Max. Vol Level of Size to be 3 3 3 3 Waste Inputs Batch (ft ) per Month per Month (ft ) (Ci/ft ) Processing Used (ft )

2. Spent resin slurry from:
a. Condensate demineralizers Dewater 50 or 200 813 (Free Water)(2) 813 (Free Water) (per ft3 resin) 1,333 (Total) 0.14 1,333 (Total)
b. Radwaste 504 (Wet Resin)(1) 504 (Wet Resin) 2 x 10-1 (per Same as for 50 or 200 demineralizers 830 (Free Water)(2) 830 (Free Water) ft3 resin) condensate demin.

1,334 (Total) 1 1,334 (Total) resins.

3. Backwash slurry from:
a. Condensate 333 (Sludge) 5,000 (Sludge) 5.15 x 10-2 Dewater 50 or 200 filters 600 (Free Water)(2) 9,000 (Free Water) per ft3 sludge) 933 (Total) 15 14,000 (Total)
b. Reactor Water 80 (Sludge)(3) 400 (Sludge) 9.9 (per Same as for 50 Cleanup F/D 207 (Free Water)(2) 1,035 (Free Water) ft3 sludge) condensate filter (Powdered resins) 287 (Total) 5 1,435 (Total) sludge.
c. Fuel Pool F/D 333 (Sludge)(3) 333 (Sludge) 1.55 x 10-2 Same as for 50 or 200 (Powered resins) 600 (Free Water)(2) 600 (Free Water) (per ft3 sludge) condensate filter 933 (Total) 1 933 (Total) sludge.

Revision 17 11.4-17 October, 2011

TABLE 11.4-1 (Continued)

(8)

Quantity Maximum Activity Method Container Vol. per Max. Batches Max. Vol Level of Size to be 3 3 3 3 Waste Inputs Batch (ft ) per Month per Month (ft ) (Ci/ft ) Processing Used (ft )

4. Traveling belt 0.133(4) 65 8.65 1.55 x 10-2 Dewater 50 or 200 cake
5. Dry compressible - - 3,000(5) 50 x 10-6 7.3 (55 gal.)

waste (clothing, mCi/ft3(6) paper, general trash)

6. Dry incompressible - - 800(5) 5 x 10-6 Packaged into 7.3 or 50 waste (tools, pipe, mCi/ft3(6)(7) waste drums high filter elements, trash) activity items are surrounded by solidification agent NOTES:

(1)

Density of wet resin mixture is approximately 71.5 lb/ft3, of which, weight of resin is 23.25 lb and weight of absorbed and interstitial water is 48.25 lb.

(2)

The term free water is used here to mean the volume of water in excess of the sum of the amount absorbed in the resins sludge and the amount occupying the void spaces in the settled resin sludge volume.

(3)

Density of powdered resin sludge is approximately 66 lb/ft3, of which, weight of powdered resin is 13 lb, and weight or absorbed and interstitial water is 53 lb.

(4)

Backwash from TBF is a moist cake, containing approximately 15.3 lb of diatomaceous earth and crud, and up to 35.7 lb of water.

(5)

The average expected volumes of dry solid wastes are 1,650 ft3/mo. (compressible) and 500 ft3/mo. (noncompressible).

(6)

The isotopes present are expected to consist primarily of miscellaneous fission products proportional to those found in the reactor coolant.

(7)

Certain filters may attain much higher activity levels depending on the fluid stream being filtered; these will be packaged within the solidified waste.

(8)

Volumes are based on the process flow data for the liquid radwaste system.

(9)

The Evaporators steam supply has been permanently cut.

Revision 18 11.4-18 October, 2013

TABLE 11.4-2 SOLID RADWASTE SYSTEM INFLUENT NUCLIDE ACTIVITIES Condensate Radwaste Filter Filter Sludge(1) Sludge(2)

Isotope ____Ci/cc_____ ____Ci/cc_____

Na-24 2.9-3 Negligible P-32 4.6-3 9.7-6 Cr-51 1.4-1 4.6-3 Mn-54 1.2-2 8.8-3 Co-58 1.4+0 3.7-1 Co-60 1.6-1 1.3-1 Fe-59 2.2-2 2.7-3 Zn-65 6.2-4 4.0-4 Zn-69m 3.1-5 Negligible Ag-110m 1.8-2 1.2-2 Ag-110 1.8-2 1.2-2 W-187 2.0-2 Negligible TOTAL 1.8+0 5.5-1 NOTES:

(1)

Activity based on 4 days accumulation of 8 batches followed by a 2 day decay period.

(2)

Activity based on 100 days accumulation of 149 batches of filter sludge from the waste collector and floor drain systems followed by a 100 day decay period.

Revision 13 11.4-19 December, 2003

TABLE 11.4-3 (Deleted)

Revision 17 11.4-20 October, 2011

TABLE 11.4-4 SOLID RADWASTE SYSTEM DEMINERALIZER ACTIVITIES RWCU Filter/

Demineralizer Condensate Radwaste Sludge Demineralizer Demineralizer Isotope __(Ci/cc)___ __(Ci/cc)___ __(Ci/cc)___

P-32 2.1-2 -- --

Cr-51 3.5+0 -- --

Mn-54 2.1+0 -- --

Co-58 1.4+2 -- --

Co-60 3.1+1 -- --

Fe-59 1.3+0 -- --

Zn-65 1.0-1 -- --

Br-83 -- -- 6.5-4 Br-84 -- -- 1.2-4 I-131 8.7-1 6.6-7 2.9-1 I-134 -- -- 1.8-3 Sr-89 7.0+1 2.2-2 4.1-1 Tc-101 -- -- 2.2-4 Cs-134 1.1+1 1.8-2 5.3-2 Cs-136 1.0-1 3.2-7 4.1-3 Cs-138 -- -- 7.0-4 Ba-139 -- -- 2.5-3 Sr-90 1.7+1 3.2-2 8.1-2 Y-90 1.7+1 3.2-2 6.5-2 Sr-92 -- -- 6.0-3 Y-92 -- -- 9.2-3 Revision 13 11.4-21 December, 2003

TABLE 11.4-4 (Continued)

RWCU Filter/

Demineralizer Condensate Radwaste Sludge Demineralizer Demineralizer Isotope __(Ci/cc)___ __(Ci/cc)___ __(Ci/cc)___

Mo-99 2.3-5 -- 1.6-1 Tc-99m 2.5-5 -- 8.1-2 Ru-103 3.0-1 5.5-5 2.1-3 Rh-103m 2.9-1 5.3-5 1.1-3 Ru-106 1.6-1 2.3-4 7.6-4 Rh-106 1.6-1 2.3-4 6.0-4 Ag-110m 2.9+0 -- --

Ag-110 2.9+0 -- --

Te-132 5.2-4 -- 3.9-1 I-132 5.2-4 -- 2.3-2 I-135 -- -- 3.8-2 Cs-137 1.8+1 3.4-2 8.1-2 Ba-137m 1.8+1 3.4-2 6.5-2 Ba-140 6.6+0 1.2-5 3.2-1 La-140 7.7+0 1.4-5 6.5-2 Ba-142 -- -- 2.1-4 La-142 -- -- 7.6-4 Ce-143 -- -- 1.1-4 Pr-143 3.8-2 -- 1.5-3 Ce-144 2.1+0 2.9-3 9.7-3 Pr-144 2.1+0 2.9-3 7.6-3 Nd-147 5.6-3 -- 4.2-4 Revision 13 11.4-22 December, 2003

TABLE 11.4-4 (Continued)

RWCU Filter/

Demineralizer Condensate Radwaste Sludge Demineralizer Demineralizer Isotope __(Ci/cc)___ __(Ci/cc)___ __(Ci/cc)___

Pm-147 9.4-3 3.4-6 2.5-5 Np-239 1.3-5 -- 1.4+0 Pu-239 2.2-3 4.5-7 6.5-1 Br-85 -- -- 9.2-6 Sr-91 -- -- 3.8-2 Y-91m -- -- 3.6-3 Y-91 1.3+0 1.9-4 2.7-1 Zr-95 1.1+0 5.3-4 6.5-3 Nb-95m 2.4-2 1.1-5 8.7-5 Nb-95 1.9+0 9.8-4 8.1-3 Zr-97 -- -- 4.2-5 Nb-97m -- -- 3.0-6 Nb-97 -- -- 4.9-6 Te-129m 4.9-1 6.1-5 3.8-3 Te-129 -- -- 8.7-6 I-129 -- -- 4.8-6 I-133 -- -- 1.4-1 Ba-141 -- -- 3.5-4 La-141 -- -- 2.8-3 Ce-141 1.2+0 1.4-4 2.4+0 Totals 3.6+2 1.8-1 7.1+0 Revision 13 11.4-23 December, 2003

11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS The process and effluent radiological monitoring 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 BASES 11.5.1.1 Design Objectives 11.5.1.1.1 Systems Required for Safety The main objective of radiation monitoring systems required for safety is to initiate appropriate protective action to limit the potential release of radioactive materials from the reactor vessel and primary and secondary containment, if predetermined radiation levels are exceeded in major process/effluent streams. An additional objective is 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.

Main steam line and containment ventilation exhaust radiation monitoring is provided to meet these objectives.

Revision 12 11.5-1 January, 2003

11.5.1.1.2 Systems Required for Plant Operation The main objective of radiation monitoring systems required for plant operation is to provide operating personnel with measurement of the content of radioactive material in all effluent and important process streams. This complies with plant normal operational limits by providing gross radiation level monitoring and collection of halogens and particulates on filters (gaseous effluents) as required by

<Regulatory Guide 1.21>. Additional objectives are to initiate discharge valve isolation on the offgas or liquid radwaste systems if predetermined release rates are exceeded and to provide for sampling at certain radiation monitor locations to allow determination of specific radionuclide content.

The radiation monitoring provided to meet these objectives are:

a. For gaseous effluent streams
1. Unit Vent
2. Offgas Vent Pipe
3. Turbine Building/Heater Bay Vent
b. For liquid effluent streams
1. Radwaste discharge
2. Emergency service water system (Loops A and B)
3. ADHR heat exchanger service water outlet Revision 18 11.5-2 October, 2013
c. For gaseous process streams
1. Offgas pretreatment
2. Offgas post-treatment
3. Carbon bed vault
4. Annulus exhaust
5. Steam packing exhauster
d. For liquid process streams
1. Underdrain
2. Nuclear closed cooling water 11.5.1.2 Design Criteria 11.5.1.2.1 Systems Required for Safety The design criteria for the safety-related radioactivity monitoring systems are that the systems:
a. Withstand the effect of natural phenomena (e.g., earthquakes) without loss of capability to perform their functions.
b. Perform their intended safety function in the environment resulting from normal and postulated accident conditions.
c. Meet the reliability, testability, independence, and failure mode requirements of engineered safety features.

Revision 12 11.5-3 January, 2003

d. Provide continuous outputs 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. Assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
g. Initiate prompt protective action prior to exceeding plant limits.
h. Provide warning of increasing radiation levels indicative of abnormal conditions by alarm annunciation.
i. Insofar as practical, provide self-monitoring of components to the extent that power failure or component malfunction causes annunciation and channel trip.
j. Maintain full scale output if radiation detection exceeds full scale.
k. Have sensitivities and ranges compatible with anticipated radiation levels.

The applicable General Design Criteria are 63 and 64. The systems meet the design requirements for Safety Class 2, Seismic Category I systems, along with the quality assurance requirements of <10 CFR 50, Appendix B>.

Revision 12 11.5-4 January, 2003

11.5.1.2.2 Systems Required for Plant Operation The design criteria for operational radiation monitoring systems are that the systems:

a. Provide continuous indication of radiation levels in the control room.
b. Provide warning of increasing radiation levels indicative of abnormal conditions by alarm annunciation.
c. Insofar as practical, provide self-monitoring of components to the extent that power failure or component malfunction causes annunciation and, for systems initiating discharge isolation, channel trip.
d. Monitor a sample representative of the bulk stream or volume.
e. Have provisions for calibration, function and instrumentation checks.
f. Have sensitivities and ranges compatible with anticipated radiation levels.
g. Maintain full scale output if radiation detection exceeds full scale.

The instrument channels monitoring discharges from the gaseous and liquid radwaste treatment systems have provisions to alarm and to initiate automatic closure of the waste discharge valve on the affected treatment system prior to exceeding the normal operation limits as required by <Regulatory Guide 1.21>.

The applicable General Design Criteria are 60, 63 and 64.

Revision 12 11.5-5 January, 2003

11.5.2 SYSTEM DESCRIPTION 11.5.2.1 Systems Required for Safety Information on the system monitors is presented in
and the arrangements shown in <Figure 11.5-1 (1)>, <Figure 11.5-1 (2)>, <Figure 11.5-1 (3)>, <Figure 11.5-1 (4)>, <Figure 11.5-1 (5)>, <Figure 11.5-1 (6)>, <Figure 11.5-1 (7)>, <Figure 11.5-1 (8)>, <Figure 11.5-1 (9)>, <Figure 11.5-1 (10)>, <Figure 11.5-1 (11)>, and <Figure 11.5-1 (12)>. 11.5.2.1.1 Main Steam Line Radiation Monitoring System This system monitors the gamma radiation level exterior to the main steam lines. The normal radiation level is produced primarily by coolant activation gases plus smaller quantities of fission gases being transported with the steam. In the event of a gross release of fission products from the core, this monitoring system provides channel trip signals for isolation of the CRVICS reactor water sample valves. The system consists of four redundant instrument channels. Each channel consists of a local detector (gamma-sensitive ion chamber) and a control room ratemeter with an auxiliary trip unit. Power for the 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 electrically independent of channels B and D. The detectors are physically located in separate pipe wells which extend into the steam tunnel near the main steam lines just downstream of the outboard main steam line isolation valves. The detectors are geometrically arranged so that this system is capable of detecting significant increases in radiation level with any number of main steam lines in operation.
lists the range of the detectors. Revision 12 11.5-6 January, 2003 Each radiation monitor has two upscale (high-high and high), one downscale and one inoperative trip circuits. Each trip is visually displayed on the affected radiation monitor. A high-high or inoperative trip in the radiation monitor results in a channel trip, which is an input to the CRVICS. A logic trip from a one-out-of-two twice MSL channel trip results in initiation of reactor water sample valve closure. A logic trip from one-out-of-two MSL channels A or C results in initiation of condenser air removal pump shutdown, and closure of the condenser air removal pump isolation valve. A high trip actuates a MSL high radiation control room annunciator. A downscale trip actuates a MSL downscale control room annunciator common to all channels. High and low trips do not result in a channel trip. Each radiation monitor visually displays the measured radiation level. 11.5.2.1.2 Containment Ventilation Exhaust Radiation Monitors This system monitors the radiation level exterior to the containment ventilation system exhaust duct. A high activity level in the ductwork could be due to fission gases from a leak or an accident. The system consists of four redundant instrument channels. Each channel consists of a local detection assembly (containing a Geiger-Mueller (GM) tube and electronics) and a control room ratemeter. Power for two channels (A and C) is supplied from RPS bus A and for the other two channels (B and D) from RPS bus B. Channels A and C are physically and electrically independent of channels B and D. One two-pen recorder powered from an inverter on the 125 volt dc non-divisional bus allows the output of two channels to be recorded by the use of selection switches. The detection assemblies are physically located outside and adjacent to the exhaust ducting downstream of the containment discharge isolation valves. Each radiation monitor provides both an analog output signal and a contact which opens on upscale (high-high) radiation or an inoperative Revision 15 11.5-7 October, 2007 circuit. Two-out-of-two upscale/inoperative trips in channels A and D initiate closure of the containment ventilation outboard isolation valves and the drywell outboard isolation valves. The same condition for channels B and C initiates closure of the containment inboard valves and drywell inboard valves. An upscale/inoperative trip is visually displayed on the affected radiation monitor ratemeter and actuates a containment and drywell ventilation exhaust high-high inoperative radiation control room annunciator. A downscale trip is also visually displayed on the radiation monitor ratemeter. Containment and drywell vent high radiation and downscale trip control annunciators common to all channels and are generated from the analog signal. Each radiation monitor ratemeter visually displays the measured radiation level.
lists the range of the detectors. 11.5.2.2 Systems Required for Plant Operation Information on these systems is presented in
,
, and
and the arrangements are shown in <Figure 11.5-1>. 11.5.2.2.1 Offgas Pretreatment Radiation Monitor This system monitors radioactivity in the condenser offgas at the inlet to the holdup piping after it has passed through the offgas condenser and moisture separator. The monitor detects the radiation level 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 sample line. It is then passed through a sample chamber and a sample panel before being returned to the suction side of the steam jet air ejector (SJAE). The sample chamber is a stainless steel pipe which is internally Revision 12 11.5-8 January, 2003 polished to minimize plateout. It can be purged with room air to check detector response to background radiation by using a three-way solenoid operated valve. The valve is controlled by a switch located in the control room. The sample panel measures and indicates sample line flow. A detector (GM tube) is positioned adjacent to the vertical sample chamber and is connected to a ratemeter in the control room. Power is supplied from an inverter on the 125 volt dc non-divisional bus for the radiation monitor, detector and recorder; power is also supplied from a 120 volt ac miscellaneous distribution panel for the sample and vial sampler panels which can be transferred to a diesel bus. The radiation monitor has three trip circuits: one upscale and one downscale in the radiation monitor itself, and one upscale in the recorder. The trip outputs are used for alarm function only. The ratemeter trip functions are visually displayed and all trip outputs actuate control room annunciators for each of the following: offgas high, offgas recorder and offgas downscale/inoperative. High or low sample line flow measured at the sample panel actuates a control room offgas pre-treatment sample high-low flow annunciator. The radiation level output by the monitor can be directly correlated to the concentration of the noble gases by using the semiautomatic vial sampler panel to obtain a grab sample. To draw a sample, a serum bottle is inserted into a sampler chamber, the sample lines are evacuated and a solenoid-operated sample valve is opened to allow offgas to enter the bottle. The bottle is then removed and the sample is analyzed in the counting room with a multichannel gamma pulse height analyzer to determine the concentration of the various noble gas radionuclides. A correlation between the observed activity and the monitor reading permits calibration of the monitor. Revision 19 11.5-9 October, 2015 11.5.2.2.2 Offgas Post-Treatment Radiation Monitor This system monitors radioactivity in the offgas piping downstream of the offgas system charcoal adsorbers and upstream of the offgas system discharge valve. A continuous sample is extracted from the offgas system piping, passed through the offgas post-treatment sample panel for monitoring and sampling, and returned to the offgas system piping. The sample panel has a pair of filters (one for particulate collection and one for halogen collection) in parallel (with respect to flow) with two identical continuous gross radiation detection assemblies. Each gross radiation assembly consists of a shielded chamber, a set of GM tubes and a check source. Two radiation monitor ratemeters in the control room analyze and visually display the measured gross radiation level. The sample panel shielded chambers can be purged with room air to check detector response to background radiation by using a solenoid valve operated from the control room. The sample panel measures and indicates sample line flow. A solenoid operated check source for each detection assembly operated from the control room can be used to check operability of the gross radiation channel. Power is supplied from an inverter on the 125 volt dc non-divisional bus for the radiation monitors and recorders, and from a 120 volt ac miscellaneous distribution panel for the sample panel purge circuit. Each radiation monitor has three trip circuits: two upscale (high-high-high and high), and one downscale (low/inoperative). Each trip is visually displayed on the radiation monitor. These three trips actuate corresponding control room annunciators: offgas post-treatment high-high-high radiation, offgas post-treatment high radiation and offgas post-treatment downscale/inoperative. A trip circuit on the Revision 15 11.5-10 October, 2007 recorder actuates an offgas post-treatment high-high radiation annunciator. High or low sample flow measured at the sample panel actuates a control room offgas post-treatment sample panel high-low flow annunciator. An auxiliary trip unit in the control room takes the high-high-high (HHH) and downscale trip outputs and, if its logic is satisfied, initiates closure of the offgas system discharge and drain valves. The logic is satisfied if two HHH, one HHH and one downscale, or two downscale trips occur. The HHH trip setpoints are determined such that valve closure is initiated prior to exceeding release rate limits. Any one high upscale trip initiates closure of offgas system bypass line valve and permits opening of the treatment line valve. A vial sampler panel similar to the pretreatment sampler panel is provided for grab sample collection to allow isotopic analysis and gross monitor calibration. 11.5.2.2.3 Carbon Bed Vault Radiation Monitor Carbon vault A and B are monitored for gross gamma radiation level. Each channel includes detector, a ratemeter and a locally mounted auxiliary unit. The ratemeter is located in the control room. The channel provides for sensing and readout, both local and remote of gamma radiation over a range of six logarithmic decades (1 to 106 mR/hr). The ratemeter has one adjustable upscale trip circuit for alarm and one downscale trip circuit for instrument trouble. The trip circuits are capable of operational verification by means of test signals or through the use of portable gamma sources. Power is supplied from an inverter on the 125 volt dc non-divisional bus. Revision 15 11.5-11 October, 2007 11.5.2.2.4 Plant Vent Radiation Monitor This unit monitors a sample of the plant vent effluent discharge <Figure 9.4-18> for particulate, iodine and gas radioactivity and also provides samples of the collected particulate and halogen for laboratory analysis. A representative sample is continuously extracted from the plant vent through an isokinetic probe in accordance with ANSI N13.1-1969 with the additional feature of manually regulating the sample flow in proportion to the vent stack flow. The sample is supplied through a 1 inch sample line which is also used to supply a representative sample to the postaccident effluent radiation monitors. This sample line is heat traced to preclude any condensation. A portion of this representative sample is taken by another isokinetic probe and passed through the shielded particulate, iodine and gas detector assemblies which are provided with scintillation detectors and check sources. The ratemeters in the control room analyze and visually display the measured radiation level for the particulate (gross Beta), gas (gross Beta) and Iodine cartridge (gamma). Power is supplied from the non-1E 120 volt ac miscellaneous distribution panel for the radiation monitor ratemeters and recorders. The 480 volt ac 3/non-1E diesel backed bus supplies power for the sample pumps. Power for the isokinetic sample pump is non-1E diesel backed to ensure system availability. Each of the ratemeters has two upscale and one downscale trip circuits which are visually displayed on the ratemeter and annunciated in the control room. The noble gas ratemeter has an alarm detection circuit which activates at full scale and shuts off the high voltage and sample pump. High or low differential pressure across the filters at the sample panel are annunciated in the control room. Revision 12 11.5-12 January, 2003 11.5.2.2.5 Turbine Building/Heater Bay Vent Radiation Monitor This unit monitors a sample of the turbine building/heater bay vent discharge for particulate, iodine and gas radioactivity and also provides samples of the collected particulate and halogen for laboratory analysis. A representative sample is continuously extracted from the turbine building/heater bay discharge vent downstream of the exhaust fans shown in <Figure 9.4-9>. Sampling and monitoring is as described for the plant vent radiation monitor. Power is supplied from the non-1E 120 volt ac miscellaneous distribution panel for the ratemeters and recorders. The 480 volt ac 3/non-1E diesel backed bus supplies power for the sample pumps. Power for the isokinetic sample pump is non-1E diesel backed to ensure system availability. Each of the ratemeters has two upscale and one downscale trip circuits which are displayed on the ratemeters and annunciated in the control room. The noble gas ratemeter has an alarm detection circuit which activates at full scale and shuts off the high voltage and sample pump. High or low differential pressure across the filters at the sample panel are annunciated in the control room. 11.5.2.2.6 Offgas Vent Pipe Monitor This unit monitors a sample of the offgas vent pipe discharge downstream of the exhaust fans <Figure 9.4-10> for particulate, iodine and gas activity and also provides samples of the collected particulate and halogen for laboratory analysis. A representative sample is continuously extracted from the offgas vent pipe and monitored as described for the plant vent radiation monitor. Power is supplied from non-1E 120 volt ac miscellaneous distribution panel for the radiation monitor ratemeters and recorders. 480 volt ac Revision 12 11.5-13 January, 2003 3/non-1E diesel backed bus supplies power for the sample pumps. Power for the isokinetic sample pump is non-1E diesel backed to assure system availability. Each of the ratemeters has two upscale and one downscale trip circuits which are visually displayed on the ratemeters and annunciated in the control room. The noble gas ratemeter has an alarm detection circuit which activates at full scale and shuts off the high voltage and sample pump. High or low differential pressure across the filters, measured at the sample panel, are annunciated in the control room. 11.5.2.2.7 Annulus Exhaust Radiation Monitor These units monitor the annulus exhaust for gas activity (gross Beta) and provides samples of collected particulate and halogen for laboratory analysis. The units are identified as Annulus Exhaust - Train A Radiation Monitor and Annulus Exhaust - Train B Radiation Monitor. A sample is continuously extracted from the annulus exhaust duct downstream of the annulus exhaust filter trains A and B through an isokinetic probe <Figure 6.5-1>. The sample is passed through a fixed particulate sample filter, a fixed halogen collection cartridge, and through a shielded scintillation detector with a check source. For units with digital readout modules, remote LED pulsers are provided with the detector assemblies in place of the radioactive check sources. The LED pulsers are used to excite the photomultiplier tubes within the scintillation detectors for functional testing of the instrument channel. The detector monitors the gross Beta gas activity. Ratemeters in the control room analyze and visually display the measured gas activity. Power is supplied from the non-1E 120 volt ac miscellaneous distribution panel for the ratemeters and recorders while the sample pumps are supplied by the 480 volt ac 3/non-1E diesel backed bus. Revision 19 11.5-14 October, 2015 The ratemeter has two upscale and one downscale trip circuits which are visually displayed on the ratemeter and annunciated in the control room. High or low differential pressure measured across the filters in the sample panel are annunciated in the control room. Revision 19 11.5-14a October, 2015 11.5.2.2.8 Steam Packing Exhauster Radiation Monitor The discharge from the steam packing exhauster is monitored for radioactivity by a shielded inline detector assembly which is provided with a scintillation detector and a check source. To provide check source function, a remote LED pulser is provided with the detector assembly in place of the radioactive check source. The LED pulser is used to excite the photomultiplier tube within the scintillation detector for functional testing of the instrument channel. The detector assembly is located on the steam packing exhauster effluent line which discharges to the offgas vent pipe as shown in <Figure 10.1-10>. A ratemeter in the control room analyzes and visually displays the measured radiation (gross Gamma). Power is supplied from the non-1E 120 volt ac miscellaneous distribution panel for the ratemeter and recorder. The ratemeter has two upscale and one downscale trips which are displayed on the ratemeter and are annunciated in the control room. 11.5.2.2.9 Liquid Process and Effluent Monitoring Systems These systems, listed in
, monitor the gamma radiation levels of liquid process and effluent streams. With the exception of the radwaste system effluent, the streams monitored normally contain only background levels of radioactive materials. Increases in radiation level may be indicative of heat exchanger leakage or equipment malfunction. Power is supplied from an inverter on the 125 volt dc non-divisional buses for the radiation monitors and recorders, and from a 120 volt ac local bus for the sample panels. The underdrain liquid monitors are powered by a non-1E 120 volt bus. Revision 19 11.5-15 October, 2015 Each radiation monitor has three trip circuits: two upscale (high-high and high) and one downscale (low). Each trip is visually displayed on the affected radiation monitor. Two of these trips actuate corresponding control room annunciators: one upscale (high-high Revision 19 11.5-15a October, 2015 radiation) and the downscale for the affected liquid monitoring channel. High or low sample flow measured at the sample panel actuates a control room flow annunciator for the affected liquid channel. For each liquid monitoring location, except for the underdrain system, a continuous sample is extracted from the liquid process pipe, passed through a liquid sample panel which contains a detection assembly for gross gamma radiation monitoring, and returned to the process pipe. The detection assembly consists of a scintillation detector mounted in a shielded sample chamber equipped with a check source. A ratemeter in the control room displays the measured gross radiation level and the analog signal is recorded. The sample panel chamber and lines can be drained to allow assessment of radiation background. The panel measures and indicates sample line flow. A solenoid operated check source operated from the control room can be used to check operability of the channel. 11.5.2.2.9.1 Radwaste Effluent Radiation Monitor This system consists of one channel, the radwaste effluent to ESW discharge pipe, which monitors the radioactivity in the radwaste effluent prior to its discharge. Liquid waste can be discharged from several radwaste processed water tanks such as the floor drain sample tanks, waste sample tanks, chemical waste distillate tanks, or detergent drain tanks. These tanks contain liquids that have been processed through one or more treatment systems such as evaporation, filtration and ion exchange. Prior to discharge from any tank, the liquid in the appropriate tank is sampled and analyzed. Based upon this analysis, discharge is permitted at a specified release rate and dilution rate. Revision 19 11.5-16 October, 2015 The upscale trip on the radwaste 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 limits for liquid effluents. When the channel is inoperable, an override switch is used to allow discharging to continue. Prior to the switch being in OVERRIDE, additional sampling is performed. Annunciation will occur when the switch is in OVERRIDE. The upscale trip also actuates an annunciator in the control room. 11.5.2.2.9.2 Emergency Service Water Radiation Monitoring This system consists of two channels <Figure 9.2-1>: one for monitoring downstream of equipment in emergency service water system Loop A and the other for Loop B. If a high radiation level is detected, the affected emergency service water line can be manually isolated. 11.5.2.2.9.3 Nuclear Closed Cooling System Radiation Monitoring This system has a channel for monitoring downstream of equipment in the nuclear closed cooling water system <Figure 9.2-4>. 11.5.2.2.9.4 Underdrain System Radiation Monitor The condition where amounts of radioactive material resulting in radionuclide concentration in the underdrain system approaching significant levels has been analyzed and is considered highly unlikely <Section 2.4.13.3>. However, radiation monitors will be located inside the gravity discharge system manholes, at the point where the lower subsystem liquid effluent discharges into the gravity drain system, to continuously monitor and detect gross amounts of radioactive concentrations in the groundwater of the underdrain system. These radiation monitors are mounted on the platform inside the manhole <Figure 2.4-71>. One monitor Revision 20 11.5-17 October, 2017 will be located in the east gravity discharge system manhole and one monitor will be located in the west gravity discharge system manhole. Each radiation monitor will transmit a preamplified signal to its associated ratemeter located in the control room. When the level of radioactivity at either radiation monitor exceeds a preset value, the associated channel will alarm in the control room, alerting the operator, and automatically stop the service pumps, and the backup pumps in the underdrain system. Radioactive concentrations of the magnitude as postulated by the failure of a waste collector tank <Section 15.7.2> and <Section 15.7.3>, can be detected and alarmed by these radiation monitors. Continuous monitoring of the liquid effluent discharge from the underdrain system will ensure that the limits of <10 CFR 20, Appendix B> are not exceeded, and that early detection of abnormalities is achieved. 11.5.2.2.9.5 ADHR Heat Exchanger Service Water Radiation Monitor This system has a channel for monitoring leakage from the ADHR system to the Service Water system <Figure 9.2-14>. If a high radiation level is detected, the ADHR system can be manually isolated. 11.5.2.3 Inspection, Calibration and Maintenance 11.5.2.3.1 Inspection and Tests During reactor operation, daily checks of system operability are made by observing channel behavior. At periodic intervals during reactor operation, the detector response (of each monitor provided with a remotely positioned check source) will be recorded together with the instrument background count rate to ensure proper functioning of the Revision 18 11.5-18 October, 2013 monitors. Any detector whose response cannot be verified by observation during normal operation or by using the remotely positioned check source will have its response checked with a portable check source. A record will be maintained showing the background radiation level and the detector response. Revision 18 11.5-18a October, 2013 Appropriate channels are tested in accordance with plant procedures. Verification of valve operation, ventilation diversion or other trip function will be done at the time of testing if it can be done without jeopardizing the plant safety. The tests will be documented. 11.5.2.3.1.1 Detailed Inspection and Tests
a. The following monitors have alarm trip circuits which can be tested by using test signals or portable gamma sources:
1. Main steam line
2. Containment ventilation exhaust
3. Offgas pretreatment
4. Carbon bed vault
b. The following monitors include built-in check sources which can be operated from the control room:
1. Offgas post-treatment
2. Annulus exhaust
3. Offgas Vent Pipe
4. Unit Vent
5. Turbine Building/Heater Bay Vent
6. Steam packing exhauster
7. Radwaste effluent to ESW Revision 13 11.5-19 December, 2003
8. Emergency service water
9. Nuclear closed cooling water
10. ADHR heat exchanger service water outlet 11.5.2.3.2 Calibration The radiation monitors calibration is traceable to certified National Institute of Standards and Technology (NIST) or commercial radionuclide standards. The source-detector geometry during primary calibration is identical to the sample-detector geometry in actual use. Secondary standards which were counted in reproducible geometry during the primary calibration may be used with each monitor for calibration after installation. A calibration can also be performed by using liquid or gaseous radionuclide standards or by analyzing particulate, iodine or gaseous grab samples with laboratory instruments.
11.5.2.3.3 Maintenance The detectors, electronics, recorders, and sample pumps are serviced and maintained to ensure reliable operations. Such maintenance includes cleaning, lubrication and assurance of free movement of the recorder in addition to the replacement or adjustment of components required after performing a test or calibration check. If work is performed which would affect the calibration, a recalibration is performed at the completion of the work. 11.5.2.3.4 Audits and Verifications Independent audits and verifications of test, calibration and maintenance records and procedures are conducted as described in <Section 17.2>. Revision 18 11.5-20 October, 2013 11.5.3 EFFLUENT MONITORING AND SAMPLING 11.5.3.1 Implementation of General Design Criterion 64 All potentially radioactive effluent discharge paths are continuously monitored for gross radiation level. Liquid releases are monitored for gross gamma with the exception of the Turbine Building atmospheric drain line from the Turbine Building supply plenum. The Turbine Building atmospheric drain line will be monitored with grab samples as established in
. Solid waste shipping containers are monitored with gamma sensitive portable survey instruments. Gaseous releases are monitored for gross gamma, gamma, and gross beta. Gaseous batch releases from low radioactivity areas needed to support maintenance activities, plant surveillances, or recovery from a potentially hazardous chemical atmosphere, where discharge through the normal effluent points is not practical, are controlled and monitored in accordance with the methodology of the ODCM for inoperable gaseous effluent monitors. The following gaseous effluent paths are sampled and monitored:
a. Unit Vent
b. Offgas Vent Pipe
c. Turbine Building/Heater Bay Vent The following liquid effluent paths are sampled and monitored:
Liquid Radwaste System Emergency Service Water (Loops A and B) ADHR heat exchanger service water outlet The monitors and ranges are listed in
. Revision 18 11.5-21 October, 2013 For the atmospheric drain line on the Turbine Building supply plenum, periodic samples are taken for isotopic and tritium analysis to monitor this pathway. An isotopic analysis is performed periodically on samples obtained from each effluent release path in order to verify the adequacy of effluent processing to meet the discharge limits to unrestricted areas. This effluent monitoring and sampling program is used to provide the information for the effluent measuring and reporting programs required by <10 CFR 50.36a>, <10 CFR 50, Appendix A> (General Design Criterion 64), <10 CFR 50, Appendix I> and <Regulatory Guide 1.21> in annual reports to the NRC. The frequency of the periodic sampling and analysis described herein is a minimum and will be increased if effluent levels approach limits.
,
,
, and
present the sample schedules. 11.5.4 PROCESS MONITORING AND SAMPLING 11.5.4.1 Implementation of General Design Criterion 60 The potentially significant radioactive discharge paths are equipped with a control system to automatically isolate the discharge on indication of a high radiation level. These include:
a. Offgas post-treatment
b. Containment ventilation exhaust
c. Liquid radwaste effluent The effluent isolation functions for each monitor are given in
and
. Revision 13 11.5-22 December, 2003 11.5.4.2 Implementation of General Design Criterion 64 Radiation levels in radioactive and potentially radioactive process streams are monitored by the following process monitors:
a. Main steam line
b. Offgas pretreatment
c. Offgas post-treatment
d. Carbon bed vault
e. Nuclear closed cooling water
f. Underdrain
g. Steam packing exhauster
h. Annulus exhaust Airborne radioactivity in the containment, drywell, fuel handling building, and other areas are monitored as described in <Section 12.3.4>
as these are used to monitor in-plant airborne radioactivity. The area radiation monitors described in <Section 12.3.4> detect abnormal radiation levels in the various process equipment rooms. Batch releases are sampled and analyzed prior to discharge in addition to the continuous effluent monitoring. The radwaste process monitoring systems are listed in
. Revision 12 11.5-23 January, 2003 TABLE 11.5-1 GASEOUS AND AIRBORNE PROCESS AND EFFLUENT RADIATION MONITOR Radiation Monitor Sample Point Instrument Channels Function Location 1D17K610 A,B,C,D Pipewells in steam Ion chambers - Control Room alarms Steam Tunnel, Main Steam Line tunnel downstream redundant channels and indication. Auxiliary of outer isolation Isolates Reactor Building valve Water Sample Valves 630 and Condenser Air Removal Lines 1D17K612 Sample from the Geiger-Mueller Control Room alarms Turbine Offgas Pretreatment offgas water and indication Building separator effluent 577 1D17K601 A,B Sample from carbonGeiger-Mueller Control Room alarms Offgas Offgas Post- vault discharge Redundant channels and indication. Building treatment Isolates Offgas 584 System 1D17K611 A,B Detectors view CarbonGeiger-Mueller Control Room Offgas Carbon Bed Vault Bed Vaults A and B indication and Building Refrigeration Ductwork alarms 603 1D17K609 A,B,C,D Ventilation duct Geiger-Mueller Control Room Intermediate Containment downstream of Redundant channels indication and Building, Ventilation Exhaust Containment alarms. Close Containment Isolation Valve Containment and Ventl. Exh. Drywell Purge Duct Ventl. System valves 672 Revision 19 11.5-24 October, 2015 TABLE 11.5-1 (Continued) Radiation Monitor Sample Point Instrument Channels Function Location 1D17K690 A,B Isokinetic sample Beta scintillation Local and Control Intermediate Annulus Exhaust downstream of filter channel for gases Room alarms and Building Train A and Train B trains and sample filters indication 620 for particulate and halogen with sample pump 1D17K780 Isokinetic sample 3-Channel, Gas- Local and Control Intermediate 2D17K780(2) from Unit Vent Halogen-Particulate, Room alarms and Building Unit Vent Auto-isokinetic scintillation type indication 682 sampler with sample pump 1D17K850 Isokinetic sample 3-Channel, Gas- Local and Control Heater Bay Turbine Building/ from HB/TB Vent Halogen-Particulate, Room alarms and Equipment Heater Bay Vent Auto-isokinetic scintillation type indication House 667 sampler with sample pump 1D17K830 Isokinetic sample 3-Channel, Gas- Local and Control Turbine Offgas Vent Pipe from Offgas Vent Halogen-Particulate, alarms and Building Pipe Auto-isokinetic scintillation type indication 620 sampler with sample pump 1D17K840 Steam packing Inline gas Control Room alarms Turbine Steam Packing exhauster effluent scintillation and indication Building Exhauster line channel 624 NOTE: (1) (Deleted) (2) The Unit 2 plant vent receives input from the active Unit 1. Revision 19 11.5-25 October, 2015 TABLE 11.5-2 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM CHARACTERISTICS No. of Trips Monitoring Number of Detector Instrument Upscale - High (Trip) Prealarm Systems Units(1) Sensitivity Range (Scale) Downscale Setpoint(2) Setpoint(2) Main Steam Lines 4-IC 3.7 x 10-10 amp/R/hr 1 to 106 mr/hr 2-1 ORM Above Full Power

Background

Offgas Pretreatment 1-GM - 1 to 106 mr/hr 2-1 ORM Variable Offgas Post- 2-GM 1.2 x10-4 Ci/cc 10 to 106 2-1 ORM Variable Treatment (Kr-85) counts/min Carbon Bed Vault 2-GM - 1 to 106 mr/hr 1-1 NA Above

Background

Containment Ventl. 4-GM - .01 to 100 mr/hr 2-1 Technical Above Exhaust (each channel) Specification Background Annulus Exhaust 2-BSD-G 1x10-6 Ci/cc 10 to 106 10 to 107 2-1 Variable Above (Xe-133) counts/min counts/min Background (for digital readout modules)

Unit Vent Exhaust 1-BSD-G 4.7x10-7 Ci/cc 2-1 (Kr-85) 1-BSD-P 2.7 x 10-11 Ci/cc See Note(3) See Note(3)

(Cs-137) 1-GSD-H 1.6 x 10-11 Ci/cc 10 to 107 (I-131) counts/min (for digital readout modules)

Turbine Bldg./ 1-BSD-G 4.7x10-7 Ci/cc 10 to 106 2-1 Heater Bay Vent (Kr-85) counts/min 1-BSD-P 2.7 x 10-11 Ci/cc (each channel) See Note(3) See Note(3)

(Cs-137) 1-GSD-H 1.6 x 10-11 Ci/cc 10 to 107 (I-131) counts/min (for digital readout modules)

Offgas Vent Pipe 1-BSD-G 4.7x10-7 Ci/cc 10 to 107 2-1 (Kr-85) counts/min 1-BSD-P 2.7 x 10-11 Ci/cc for digital readout See Note(3) See Note(3)

(Cs-137) modules 1-GSD-H 1.6 x 10-11 Ci/cc (I-131)

Revision 19 11.5-26 October, 2015

TABLE 11.5-2 (Continued)

No. of Trips Monitoring Number of Detector Instrument Upscale - High (Trip) Prealarm Systems Units(1) Sensitivity Range (Scale) Downscale Setpoint(2) Setpoint(2)

Steam Packing 1-BSD-G 3.8 x 10-6 µCi/cc 10 to 106 10 to 107 2-1 Variable Variable Exhauster (Xe-133) counts/min. counts/min.

(for digital Emergency Service 1-GSD-L 1 x 10-6 µCi/cc 10 to 106 readout 1-1 Variable Variable Water Loop A (Cs-137) counts/min modules)

Emergency Service 1-GSD-L 1 x 10-6 µCi/cc 10 to 106 1-1 Variable Variable Water Loop B (Cs-137) counts/min.

Nuclear Closed 1-GSD-L 1 x 10-6 µCi/cc 10 to 106 1-1 Variable Variable Cooling Water (Cs-137) counts/min.

Plant Radwaste 1-GSD-L 1 x 10-6 µCi/cc 10 to 106 1-1 Variable -

Discharge - ESW (Cs-137) counts/min.

Discharge Underdrain 1-GSD-L 8.5 x 10-6 µCi/cc 10 to 106 2-1 Variable Variable (I-GSD-L) counts/min.

ADHR Heat Exchanger 1-GSD-L 1 x 10-6 µCi/cc 10 to 106 1-1 Variable Variable Service Water Outlet (Cs-137) counts/min.

NOTES:

(1) Types of detectors are designated as follows:

GM - Geiger-Mueller detector IC - Ion chamber detector BSD-G - Beta scintillation detector - Gas Channel BSD-P - Beta scintillation detector - Particulate Filter GSD-A - Gamma scintillation detector - Halogen Cartridge GSD-L - Gamma scintillation detector - Liquid Channel (2) Setpoints to be revised as required to be compatible with limits established and current calibrated sensitivity of the applicable channel.

(3) Basis for setpoint calculations:

a. For noble gas channels, setpoints calculated in accordance with the Offsite Dose Calculation Manual based on compliance with <10 CFR 20> limits for unrestricted areas.
b. For particulate and iodine channels, setpoints are variable.

Revision 21 11.5-27 October, 2019

TABLE 11.5-3 LIQUID PROCESS AND EFFLUENT RADIATION MONITORS Radiation Monitor Sample Point Instrument Channels Function Location 1D17K604 ESW - Loop A Gamma - scint., Control Room Auxiliary Emergency Service downstream of RHR offline indication and alarm Building Water Loop A Heat Exchanger with sample pump 568 - East and West 1D17K605 ESW - Loop B Gamma - scint., Control Room Auxiliary Emergency Service downstream of RHR offline indication and alarm Building Water Loop B Heat Exchanger with sample pump 568 - West and East D17K607 Downstream of Gamma - scint., Control Room Control Nuclear Closed nuclear closed offline indication and alarm Complex Cooling System Cooling Heat with sample pump 599 Exchangers D17K606 Radwaste line Gamma - scint., Control Room and Auxiliary Radwaste Effluent downstream of offline Radwaste Control Room Building to ESW - Discharge discharge valves with sample pump indication and 620 - East G50-F153 and alarm. Close G50-F155 discharge valve on high trip.

D17K820 A Gravity Drain System Gamma - scint., Control Room Gravity Drain Underdrain System discharge lines INLINE indication and System Manhole alarm. Stop No. 23, 608 underdrain pumps on high trip.

Revision 20 11.5-28 October, 2017

TABLE 11.5-3 (Continued)

Radiation Monitor Sample Point Instrument Channels Function Location D17K820 B Gravity Drain System Gamma - scint., Control Room Gravity Drain Underdrain System discharge lines offline indication and System Manhole with sample pump alarm. Stop No. 20, 608 underdrain pumps on high trip.

1D17K608 Downstream of ADHR Gamma - scint., Control Room Auxiliary ADHR Heat Exchanger Heat Exchanger SW offline indication and alarm Building Service Water Outlet outlet with sample pump 568 - East NOTE:

(1) (Deleted)

Revision 20 11.5-29 October, 2017

TABLE 11.5-4 RADIOLOGICAL ANALYSIS

SUMMARY

OF LIQUID PROCESS SAMPLES Grab Sample LLD Sample Description Frequency Analysis (Ci/ml) Purpose

1. Reactor Coolant 31 days Iodine Dose 5 x 10-7 Evaluate fuel cladding

<Figure 5.1-1> Equivalent integrity 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Gamma Spectrum 5 x 10-7 Determine radionuclides present in system (activity referenced in

,
,
,
,
, and
2. Reactor Water Cleanup Periodically Gamma Spectrum 1 x 10-6 Evaluate cleanup efficiency System <Figure 5.4-16>, (typical demineralizer
<Figure 5.4-17> activity referenced in
)
3. Condensate Demineralizer
<Figure 10.1-6> Influent Periodically Gamma Spectrum 1 x 10-6 Evaluate demineralizer performance Effluent Periodically Gamma Spectrum 1 x 10-6 Evaluate demineralizer performance
4. Condensate Storage Tank - Weekly Gamma Spectrum 1 x 10-6 Tank inventory Unit 1 (500,000 gal.)
Revision 12 11.5-30 January, 2003 TABLE 11.5-4 (Continued) Grab SampleLLD Sample Description Frequency Analysis (Ci/ml) Purpose
5. (Deleted)
6. Fuel Pool Filter Demineralizer Inlet and Outlet Periodically Gamma Spectrum 1 x 10-6 Evaluate system performance
(~1,000 gpm)
7. Waste Collector Tank Periodically Gamma Spectrum 1 x 10-6 Evaluate system performance (36,500 gal. nominal cap.)
8. Floor Drain Collector Periodically Gamma Spectrum 1 x 10-6 Evaluate system performance Tank (36,500 gal. nominal cap.)
9. Chemical Waste Tank Periodically Gamma Spectrum 1 x 10-6 Evaluate system performance (19,650 gal. capacity)
Revision 19 11.5-31 October, 2015 TABLE 11.5-4 (Continued) Grab SampleLLD Sample Description Frequency Analysis (Ci/ml) Purpose
11. Evaporator Distillate Periodically Gamma Spectrum 1 x 10-6 Evaluate evaporator Tanks (2; 20,000 gal. performance each)
12. Nuclear Closed Cooling Quarterly Gamma Spectrum 1 x 10-6 Evaluate system integrity
(<Figure 9.2-4> (Normal activity system flow 23,000 gpm) negligible)
13. ADHR Heat Exchanger Periodically Gamma Spectrum 1 x 10-6 Monitor leakage from the ADHR Service Water Outlet system to the Service Water
(<Figure 9.2-14> system system flow 3500 gpm) Revision 18 11.5-32 October, 2013 TABLE 11.5-5 RADIOLOGICAL ANALYSIS

SUMMARY

OF GASEOUS PROCESS SAMPLES Sample LLD Sample Description Frequency Analysis (Ci/ml) Purpose

1. Offgas Pre-Treatment Gamma Isotopic(2) 1 x 10-4 Determine offgas Sample activity(1)
a. Downstream of Offgas Monthly Isotopic composition Condenser and fuel performance
2. Offgas Post-treatment Sample Gamma Isotopic(2) 1 x 10-4 Determine offgas system cleanup performance(1)
a. Upstream of Charcoal Periodically Adsorber
b. Downstream of First Periodically Evaluate Charcoal Charcoal Adsorber Noble Gas Delay
c. Downstream of all Monthly Determine Noble Gas release Charcoal Beds rate from charcoal beds.

NOTES:

(1)

Anticipated process flow is 6-30 cfm; compositions are referenced in
. (2) Principal emitters for LLD are Kr-85m, Kr-87, Kr-88, Xe-133, Xe-135, Xe-138. Revision 12 11.5-33 January, 2003 TABLE 11.5-6 RADIOLOGICAL ANALYSIS

SUMMARY

OF LIQUID EFFLUENT SAMPLES Sample LLD Sample Description Frequency(2) Analysis (Ci/ml) Purpose

1. Floor Drain Sample Tank Batch(1) Gamma Isotopic(5) 5 x 10-7 Effluent discharge record (34,000 gal. capacity) Dissolved Gas(4) 1 x 10-5

I-131 1 x 10-6

2. Waste Sample Tanks Batch(1) Gamma Isotopic(5) 5 x 10-7 Effluent discharge record (34,000 gal. capacity) Dissolved Gas(4) 1 x 10-5

I-131 1 x 10-6

3. Chemical Waste Distillate Batch(1) Gamma Isotopic(5) 5 x 10-7 Effluent discharge record Tanks (19,100 gal. Dissolved Gas(4) 1 x 10-5

capacity) I-131 1 x 10-6

4. Detergent Drain Tank Batch(1) Gamma Isotopic(5) 5 x 10-7 Effluent discharge record (1,550 gal. capacity) Dissolved Gas(4) 1 x 10-5 (normal activity I-131 1 x 10-6 negligible)
5. Liquid Radwaste Effluents
(
for flow and activities) Composite of all Monthly Tritium 1 x 10-5 Effluent discharge record tanks discharged(3) Gross Alpha 1 x 10-7 Quarterly Sr-89/90 5 x 10-8 Quarterly Fe-55 1 x 10-6 Revision 12 11.5-34 January, 2003 TABLE 11.5-6 (Continued) Sample LLD Sample Description Frequency Analysis (Ci/ml) Purpose
6. Circulating Water Quarterly Gamma Isotopic 5 x 10-7 Effluent discharge record Cooling Tower Blowdown (8,200 gpm)
7. Underdrain sumps Quarterly Gamma Isotopic 5 x 10-7 Effluent discharge record
(~80 gpm) (negligible normal activity) <Section 15.7.3.5>
8. Discharge canal Weekly Gamma Isotopic 5 x 10-7 Effluent discharge record
(~67,000 gpm)
9. Emergency Service Water Gamma Isotopic(5) 5 x 10-7 Effluent discharge record (6)
Loops A and B, outlet Weekly Dissolved Gas(4) 1 x 10-5 of RHR Heat Exchangers I-131 1 x 10-6 Composite of all Monthly Tritium 1 x 10-5 Tanks Discharged Gross Alpha 1 x 10-7 Quarterly Sr-89/90 5 x 10-8 Fe-55 1 x 10-6
10. Atmospheric drain line Weekly(7)(10) Gamma Isotopic(5) 5 x 10-7 Effluent discharge record from Turbine Building Monthly(7) Dissolved Gas(8) 1 x 10-5 supply plenums Monthly(9) Tritium 1 x 10-5 Gross Alpha 1 x 10-7 Quarterly(9) Sr-89/90 5 x 10-8 Fe-55 1 x 10-6 Revision 13 11.5-35 December, 2003
TABLE 11.5-6 (Continued) NOTES: (1) If tank is to be discharged, analyses will be performed on each batch. If tank is not to be discharged, analyses will be performed periodically to evaluate equipment performance. (2) If no discharge event occurs during the period frequency shall be so adjusted. (3) Refer to
for equipment flows. (4) Analysis is run on one batch release per month. (5) Principal gamma emitters for which the LLD specification applies: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. (Note: The LLD for Ce-144 is 5 x 10-6) (6) Samples will be drawn every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during operation if activity is greater than LLD. Samples are normally drawn daily and composited weekly for analysis. (7) Sampling is only required when the drains have been directed to storm drains. (8) Analysis is run on one grab sample per month. (9) Composite sample from weekly grab samples. (10) Samples are normally drawn daily when the drains have been directed to storm drains and composited weekly for analysis. Revision 13 11.5-36 December, 2003 TABLE 11.5-7 RADIOLOGICAL ANALYSIS

SUMMARY

OF GASEOUS EFFLUENT SAMPLES Sample LLD Sample Description Frequency Analysis (Ci/ml) Purpose

1. Unit Vents, Heater Bay/ Weekly Principal gamma 1x10-11 Effluent Record(5)

Turbine Building Vents, emitters(1)

Offgas Vent Pipe(4) for at least I-131 and Ba-La-140 I-131(2) 1x10-12 I-133(2) 1x10-10 Monthly Principal gamma 1x10-4 Grab emitters(3)

Monthly Gross Alpha(1) 1x10-11 Composite Quarterly Sr-89, 90 1x10-11 Composite Monthly Tritium 1x10-6 NOTES:

(1)

On particulate filter - Principal gamma emitters are Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144.

(2)

On charcoal cartridge.

(3)

Gas samples - Principal gamma emitters are Kr-87, Kr-88, Xe-133m, Xe-133, Xe-135, and Xe-138.

(4)

Effluent flow rates are referenced in <Figure 11.3-3>.

(5)

Compositions are listed in
and
. Revision 12 11.5-37 January, 2003
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