ML21133A108

0 to Updated Final Safety Analysis Report, Chapter 11, Sections 11.1 Through 11.5

ML21133A108
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LGS UFSAR CHAPTER 11 - RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS The radioactivity source terms, reactor coolant concentrations, and offgas release rates used in the design and evaluation of LGS vary depending upon the specific application. This section gives nominal noble gas source terms based on an offgas release rate of 100,000 Ci/sec at 30 minutes decay. The corresponding halogen and particulate/corrosion product coolant concentrations are also given. From these source terms, the shielding design basis source terms given in Section 12.2.1 were derived. The shielding design basis fission product source terms correspond to an offgas release rate of 350,000 Ci/sec at 30 minute decay. The expected source terms, as used in Chapter 11 for 10CFR50, Appendix I evaluations and in Section 12.2.2 for in-plant airborne activity and exposure estimates, are based on the source terms in ANSI N237 (1976), as incorporated into NUREG-0016 and in accordance with Regulatory Guide 1.112.

These expected source terms correspond to an offgas release rate of 60,000 Ci/sec at 30 minutes decay. The source terms used for accident analysis in Chapter 15 also vary with the application. The source terms are given for each accident in Chapter 15.

The information in this section defines the design basis radioactive material levels in the reactor water, steam, and offgas. The various radioisotopes listed are grouped as coolant activation products, noncoolant 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 and Dresden 2. The design basis radioactive material levels do not necessarily include all the radioisotopes observed or theoretically predicted to be present. The radioisotopes included are considered significant to one or more of the following criteria:

a. Plant equipment design

b. Shielding design

c. Understanding system operation and performance

d. Measurement practicability

e. Evaluating radioactive material releases to the environment For halogens, radioisotopes with half-lives of less than 3 minutes have been omitted. For other fission product radioisotopes in reactor water, radioisotopes with half-lives less than 10 minutes are 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 whose sources vary from minuscule defects in cladding to "tramp" uranium on external cladding surfaces. The relative concentrations or amounts of noble radiogas isotopes can be described as follows:

CHAPTER 11 11.1-1 REV. 14, SEPTEMBER 2008

LGS UFSAR Equilibrium: R ~ k1Y (EQ. 11.1-1)

Recoil: R ~ k2Y (EQ. 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 release of the radiogases from the fuel to the coolant, the radiogases approach equilibrium levels in the fuel and the equilibrium mixture results. 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 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.

The VBWR and the 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 11.1-1). 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 ~ k3Y.5 (EQ. 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 was 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). Since about 1967, the design basis release magnitude established and used (including the 1971 source terms) is 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, structure air contamination, shielding design, and turbine 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 11.1-2).

While the noble radiogas source term magnitude had been established at 0.1 Ci/sec (t = 30 min),

it was recognized that there may be a more statistically applicable distribution for the noble CHAPTER 11 11.1-2 REV. 14, SEPTEMBER 2008

LGS UFSAR 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-e-t) (e-t) (EQ. 11.1-4)

With the exception of Kr-85 (with a half-life of 10.74 year), the noble radiogas fission products in the fuel are essentially at an equilibrium condition after an irradiation period of several months (rate of formation is equal to the rate of decay). So for practical purposes the term (1-e-t) approaches 1 and can be neglected when the reactor has been operating at steady-state for long periods of time. The term (e-t) is used to adjust the releases from the fuel (t = 0) to the decay time for which values are needed. Historically t = 30 min 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 describes the leakage of each noble radiogas:

Rg = KgYm (EQ. 11.1-5)

The constant (Kg) describes the magnitude of leakage. The relative rates of leakage for the different noble radiogas isotopes is 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) (EQ. 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 minutes),

and with radiogas leakage at Dresden 2 varying from 0.001-0.169 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 = 0.1 to m = 0.6. After establishing the value of m = 0.4, the value of (K) can be calculated by selecting a value for (R), or as has been done historically, the design basis is set by the total design basis source term magnitude at t = 30 min. With Rg at 30 min = 0.1 Ci/sec, (K) can be calculated as being 2.6x107 and Equation 11.1-4 becomes:

Rg = 2.6x107 Y0.4 (1-eT) (e-t) (EQ. 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 Table 11.1-1 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.

CHAPTER 11 11.1-3 REV. 14, SEPTEMBER 2008

LGS UFSAR 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 fashion similar to that used with gases, a simplified equation can be shown to describe the release of each halogen radioisotope:

Rh = KhYn (EQ. 11.1-8)

The constant (K) describes the magnitude of leakage from fuel. The relative rates of halogen radioisotope leakage is 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 = 0.1 to n = 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 Dresden 2 data, during the period August 1970 to January 1971, indicates 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. The data also indicates that during this period, high radiogas leakages were not accompanied by high radioiodine leakage from the fuel. Except for one KRB datum point, all steady-state I-131 leakages observed at KRB or Dresden 2 were equal to or less than 505 Ci/sec. Even at Dresden 1 in March 1965, when severe defects were experienced in stainless steel clad fuel, I-131 leakages greater than 500 Ci/sec were not experienced. Figure 11.1-3 shows that these higher radioiodine leakages from the fuel were related to noble radiogas source terms of less than the design basis value of 0.1 Ci/sec (t = 30 min). This may be partially explained by inherent limitations due to internal plant operational problems that caused plant derating.

In general, it is not 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 radiohalogen design basis source terms from fuel were established based on I-131 leakage of 700 Ci/sec. This value, as shown in Figure 11.1-3, accommodates the experience data and the noble radiogas design basis source term of 0.1 Ci/sec (t = 30 min). With the I-131 design basis source term established, (Kh) is calculated as being 2.4x107; and halogen radioisotope release can be expressed by the following equation:

Rh = 2.4x107 Y0.5 (1-e-T) (e-t) (EQ. 11.1-9)

Concentrations of radiohalogens in reactor water can be calculated using the following equation:

Ch = Rh (EQ. 11.1-10)

( + + )M Although carryover of most soluble radioisotopes from reactor water to steam is observed to be

<0.1% (<0.001 fraction), the observed "carryover" for radiohalogens has varied from 0.1% to CHAPTER 11 11.1-4 REV. 14, SEPTEMBER 2008

LGS UFSAR about 2% on newer plants. The average of observed radiohalogen carryover measurements has been 1.2% 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% (0.02 fraction) is used.

The halogen release rate from the fuel can be calculated from Equation 11.1-9. Concentrations in reactor water can be calculated from Equation 11.1-10. The resultant concentrations are presented in Table 11.1-2.

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 Table 11.1-3. Carryover of these radioisotopes from the reactor water to the steam is estimated to be <0.1% (<0.001 fraction). However, in addition to carryover, decay of noble radiogases in the steam leaving the reactor results in the production of noble gas daughter radioisotopes in the steam and condensate systems.

Some daughter radioisotopes (e.g., yttrium and lanthanum) are 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 water is Cm-242 at an estimated concentration of 10-6 Ci/g or less, which is below the maximum permissible concentration (pre-1994 10CFR20) in drinking water applicable to continuous use by the general public. 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)

CHAPTER 11 11.1-5 REV. 14, SEPTEMBER 2008

LGS UFSAR 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 Ch = concentration of a halogen radioisotope in reactor water (Ci/g)

M = mass of water in the operating reactor (g)

= cleanup system removal constant (sec-1)

= cleanup system flow rate (g/sec)

M(g) g = grams mass

= halogen steam carryover removal constant (sec-1)

Cs (Ci/ g) steam flow (g/sec)

=

Ch (Ci/ g)

M(g)

Cs = concentration of halogen radioisotopes in steam (Ci/g) 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 Table 11.1-4.

11.1.2.2 Noncoolant 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 noncoolant activation products have been estimated conservatively from experience data (Reference 11.1-3). The resultant concentrations are presented in Table 11.1-5.

Carryover of these isotopes from the reactor water to the steam is estimated to be <0.1% (<0.001 fraction).

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

a. Activation of naturally occurring deuterium in the primary coolant CHAPTER 11 11.1-6 REV. 14, SEPTEMBER 2008

LGS UFSAR

b. Nuclear fission of UO2 fuel

c. Neutron reactions with boron used in reactivity control rods The tritium formed in control rods may be released if through-wall cracking of the control rod jacketing occurs. High exposure original equipment (OE) type control rods in particular have been observed to experience stress corrosion cracking of the jacketing material due to swelling of the contained boron carbide from neutron absorption. Propagation of through-wall cracking can result in boron leaching into the reactor coolant and release of accumulated tritium. Such tritium releases occur intermittently and have proven to be not quantifiable, with no clear trend having been established throughout the industry. The safety function of the control rod is not adversely affected since stress corrosion cracking and boron leaching are anticipated and accounted for, and a program exists to monitor and replace control rods as required. Following the routine, cyclic replacement of the high exposure and potentially cracked control rods with advanced designs less susceptible to stress corrosion cracking, tritium release into the reactor coolant from control rods is expected to be negligible .

A prime 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:

Ract = V (EQ. 11.1-11) 3.7x104 P where:

Ract = tritium formation rate by deuterium activation (Ci/sec/MW t)

= macroscopic thermal neutron cross-section (cm-1)

= thermal neutron flux (neutrons/cm2-sec)

V = coolant volume in the core (cm3)

= tritium radioactive decay constant (1.78x10-9 sec-1)

P = reactor power level (MWt)

For recent BWR designs (Ract) is calculated to be (1.3 +/- 0.4)x104 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 because of 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, CHAPTER 11 11.1-7 REV. 14, SEPTEMBER 2008

LGS UFSAR since 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 11.1-4).

The study made at Dresden 1 in 1968 by the U.S. Public Health Service suggests that essentially all of the tritium released from the plant could be accounted for by the deuterium activation source (Reference 11.1-5). 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:

Rdif = KY0.5 (EQ. 11.1-12) where:

Rdif = leakage rate of tritium from fuel (Ci/sec)

Y = fission yield fraction (atoms/fission)

= radioactive decay constant (sec-1)

K = a constant related to total tritium leakage rate If the total noble radiogas source term is 0.1 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 U.S.

Public Health Service 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 0.015-0.10 Ci/sec. Since the annual average noble radiogas leakage from a BWR is expected to be less than 0.1 Ci/sec (t = 30 min), the annual average tritium release rate from the fission source can be conservatively estimated at 0.12 +/- 0.12 Ci/sec, or 0.0-0.24 Ci/sec.

Based on this approach, the estimated total tritium appearance rate in reactor coolant and the 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 tritiated water vapor (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.

CHAPTER 11 11.1-8 REV. 14, SEPTEMBER 2008

LGS UFSAR Recombination of radiolysis 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. 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, 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 U.S. Public Health Service study at Dresden 1 estimated that approximately 90% of the tritium release was observed in liquid effluent, with the remaining 10% leaving as gaseous effluent (Reference 11.1-5). Efforts to reduce the volume of liquid effluent discharges may change this distribution so that a greater amount of tritium will leave as gaseous effluent. From a practical standpoint, the fraction of tritium leaving as liquid effluent may vary between 60% and 90%, with the remainder leaving in gaseous effluent.

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 References 11.1-6, 11.1-7 and 11.1-8.

11.1.5 PROCESS LEAKAGE SOURCES Process leakage results in potential release paths for noble gases and other volatile fission products via ventilation systems. Liquid from process leaks is collected and routed to the liquid waste management system. Radioactive releases from the liquid waste management system are discussed in Section 11.2. Radionuclide releases via ventilation paths are at extremely low levels and have been insignificant compared to process offgas from operating BWR plants. However, because the implementation of improved process offgas treatment systems make the ventilation release relatively significant, GE has conducted measurements to identify and qualify these low level release paths. GE maintains an awareness of other measurements by the EPRI 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 structures. In general, the noble radiogases remain airborne and are released to the atmosphere with little delay via the structure's ventilation exhaust ducts. The radionuclides are partitioned between air and water, and airborne radioiodines may plateout 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 in the ventilation exhaust air.

CHAPTER 11 11.1-9 REV. 14, SEPTEMBER 2008

LGS UFSAR Experience with airborne radiological releases from the BWR enclosure HVAC and the main condenser mechanical vacuum pump have been compiled and evaluated in NEDO-21159 (Reference 11.1-9). This report is periodically updated to incorporate the most recent data on airborne emission. The results of these evaluations are based on data obtained by utility personnel and special in-plant studies of operating BWR plants by independent organizations and GE. An evaluation of the radioactive releases from ventilation systems for compliance with 10CFR50, Appendix I is given in Section 11.3. An evaluation of exposures due to airborne radioactivity is given in Section 12.2.2.

11.

1.6 REFERENCES

11.1-1 F.J. Brutschy, "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).

11.1-2 J.M. Skarpelos and R.S. Gilbert, "Technical Derivation of BWR 1971 Design Basis Radioactive Source Tems", NEDO-10871, (March 1975).

11.1-3 R.B. Elkins, "Experience with BWR Fuel Through December 1976", NEDO-21660, (July 1977).

11.1-4 J.W. Ray, "Tritium in Power Reactor," "Reactor and Fuel Processing Technology",

12 (1), pp. 19-26, (Winter 1968-1969).

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

11.1-6 H.E. Williamson and D.C. Ditmore, "Experience with BWR Fuel Through September 1971", NEDO-10505, (May 1972).

11.1-7 R.B. Elkins, "Experience with BWR Fuel Through September 1974", NEDO-20922, (June 1975).

11.1-8 H.E. Williamson and D.C. Ditmore, "Current State of Knowledge of High Performance BWR Zircaloy Clad UO2 Fuel", NEDO-10173, (May 1970).

11.1-9 T.R. Marrero, "Airborne Releases From BWRs for Environmental Impact Evaluations", NEDO-21159, (March 1976).

CHAPTER 11 11.1-10 REV. 14, SEPTEMBER 2008

LGS UFSAR Table 11.1-1 NOBLE RADIOGAS SOURCE TERMS SOURCE TERMS SOURCE TERM

@t=0 @ t = 30 min ISOTOPE HALF-LIFE (Ci/sec) (Ci/sec)

Kr-83m 1.86 hr 3.4x103 2.9x103 Kr-85m 4.4 hr 6.1x103 5.6x103 Kr-85 10.74 yr 10 to 20(1) 10 to 20(1)

Kr-87 76.0 min 2.0x104 1.5x104 Kr-88 2.79 hr 2.0x104 1.8x104 Kr-89 3.18 min 1.3x105 1.8x102 Kr-90 32.3 sec 2.8x105 -

Kr-91 8.6 sec 3.3x105 -

Kr-92 1.84 sec 3.3x105 -

Kr-93 1.29 sec 9.3x104 -

Kr-94 1.0 sec 2.3x104 -

Kr-95 0.5 sec 2.1x103 -

Kr-97 1.0 sec 1.4x101 -

Xe-131m 11.96 day 1.5x101 1.5x101 Xe-133m 2.26 day 2.9x102 2.8x102 Xe-133 5.27 day 8.2x103 8.2x103 Xe-135m 15.7 min 2.6x104 6.9x103 Xe-135 9.16 hr 2.2x104 2.2x104 Xe-137 3.82 min 1.5x105 6.7x102 Xe-138 14.2 min 8.9x104 2.1x104 Xe-139 40.0 sec 2.8x105 -

Xe-140 13.6 sec 3.0x105 -

Xe-141 1.72 sec 2.4x105 -

Xe-142 1.22 sec 7.3x104 -

Xe-143 0.96 sec 1.2x104 -

Xe-144 9.0 sec 5.6x102 -

TOTALS ~2.5x106 ~1.0x105 (1)

Estimated from experimental observations.

CHAPTER 11 11.1-11 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.1-2 HALOGEN RADIOISOTOPES IN REACTOR WATER CONCENTRATION ISOTOPE HALF-LIFE (Ci/g)

Br-83 2.40 hr 1.5x10-2 Br-84 31.8 min 2.7x10-2 Br-85 3.0 min 1.7x10-2 I-131 8.065 day 1.3x10-2 I-132 2.284 hr 1.2x10-2 I-133 20.8 hr 8.9x10-2 I-134 52.3 min 2.4x10-2 I-135 6.7 hr 1.3x10-2 CHAPTER 11 11.1-12 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.1-3 OTHER FISSION PRODUCT RADIOISOTOPES IN REACTOR WATER CONCENTRATION ISOTOPE HALF-LIFE (Ci/g)

Sr-89 50.8 day 3.1x10-3 Sr-90 28.9 yr 2.3x10-4 Sr-91 9.67 hr 6.9x10-2 Sr-92 2.69 hr 1.1x10-1 Zr-95 65.5 day 4.0x10-5 Zr-97 16.8 hr 3.2x10-5 Nb-95 35.1 day 4.2x10-5 Mo-99 66.6 hr 2.2x10-2 Tc-99m 6.007 hr 2.8x10-1 Tc-101 14.2 min 1.4x10-1 Ru-103 39.8 day 1.9x10-5 Ru-106 368.0 day 2.6x10-6 Te-129m 34.1 day 4.0x10-5 Te-132 78.0 hr 4.9x10-2 Cs-134 2.06 yr 1.6x10-4 Cs-136 13.0 day 1.1x10-4 Cs-137 30.2 yr 2.4x10-4 Cs-138 32.3 min 1.9x10-1 Ba-139 83.2 min 1.6x10-1 Ba-140 12.8 day 9.0x10-3 Ba-141 18.3 min 1.7x10-1 Ba-142 10.7 min 1.7x10-1 Ce-141 32.53 day 3.9x10-5 Ce-143 33.0 hr 3.5x10-5 Ce-144 284.4 day 3.5x10-5 Pr-143 13.58 day 3.8x10-5 Nd-147 11.06 day 1.4x10-5 Pu-239 2.35 day 2.4x10-1 CHAPTER 11 11.1-13 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.1-4 COOLANT ACTIVATION PRODUCTS IN REACTOR WATER AND STEAM REACTOR STEAM WATER CONCENTRATION CONCENTRATION ISOTOPE HALF-LIFE (Ci/g) (Ci/g)

N-13 9.99 min 7x10-3 4x10-2 N-16 7.13 sec 2.50x10+2 4.80x10+1 N-17 4.14 sec 2x10-2 6x10-3 O-19 26.8 sec 8x10-1 7x10-1 F-18 109.8 min 4x10-3 4x10-3 CHAPTER 11 11.1-14 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.1-5 NONCOOLANT ACTIVATION PRODUCTS IN REACTOR WATER CONCENTRATION ISOTOPE HALF-LIFE (Ci/g)

Na-24 15.0 hr 2.0x10-3 P-32 14.31 day 2.0x10-5 Cr-51 27.8 day 5.0x10-4 Mn-54 313.0 day 4.0x10-5 Mn-56 2.582 hr 5.0x10-2 Co-58 71.4 day 5.0x10-3 Co-60 5.258 yr 5.0x10-4 Fe-59 45.0 day 8.0x10-5 Ni-65 2.55 hr 3.0x10-4 Zn-65 243.7 day 2.0x10-6 Zn-69m 13.7 hr 3.0x10-5 Ag-110m 253.0 day 6.0x10-5 W-187 23.9 hr 3.0x10-3 CHAPTER 11 11.1-15 REV. 13, SEPTEMBER 2006

LGS UFSAR 11.2 LIQUID WASTE MANAGEMENT SYSTEM The liquid waste management system is designed to process and dispose of, or recycle all of the radioactive, or potentially radioactive liquid wastes generated in the operation of the plant. The liquid waste management system consists of equipment drain (low conductivity), floor drain (high conductivity), chemical waste, and laundry drain subsystems. These subsystems are shown on drawings M-62, M-63 and M-64. Equipment locations are shown on drawings provided in Section 1.2.

The plant drainage system, a major input source to the liquid waste management system, is described in Section 9.3.3. The equipment and floor drainage collection system is shown in drawing M-61. Sufficient treatment is available to process all waste water to condensate storage water quality or to preset quality limits acceptable for release to the environment. The liquid waste management system has no safety-related function.

11.2.1 DESIGN BASES

a. The system has the capability to process the maximum anticipated quantities of liquid wastes without impairing the operation or availability of the plant during both normal and expected occurrence conditions, to satisfy the requirements of 10CFR20 and 10CFR50.

b. Alternate process routes, subsystem cross-ties, and adequate volumes are designed into the system to provide for operational and anticipated surge waste volumes which could occur during refueling, abnormal leakage, decontamination activities, and equipment maintenance.

c. The system is designed so that no potentially radioactive liquids can be discharged to the environment unless they have been processed, monitored, and mixed with cooling tower blowdown. This results in an offsite release and radiation exposures to individuals and the general population (on an annual average basis) within the limits of 10CFR20 and 10CFR50.

d. The liquid waste management system design meets the requirements of GDC 60 and GDC 61.

e. The liquid waste management system is designed to keep the exposure to plant personnel as low as reasonably achievable (ALARA) during normal operation and plant maintenance, in accordance with Regulatory Guide 8.8.

f. The expected radionuclide activity concentrations in the liquid waste management process equipment are based on 60,000 Ci/sec noble gas release rate for one reactor unit after a 30 minute delay. The maximum expected radionuclide activity concentrations in the liquid waste management process equipment are based on 350,000 Ci/sec noble gas release rate for one reactor unit after a 30 minute delay.

g. All piping and equipment in the liquid waste management system are designed to seismic Category II requirements. The seismic category, quality group CHAPTER 11 11.2-1 REV. 16, SEPTEMBER 2012

LGS UFSAR classification, and corresponding codes and standards that apply to the design of the liquid waste management system are discussed in Section 3.2.

h. Design features provided to reduce maintenance, equipment downtime, liquid leakage, or gaseous releases of radioactive materials to the enclosure atmosphere, to facilitate cleaning, or otherwise improve radwaste operations are discussed in Section 12.3.

i. All atmospheric liquid radwaste tanks are provided with an overflow connection adequately sized to accommodate expected tank inlet flow conditions. The overflow is connected below the tank vent and at least one inch above the high level, alarm trip point. Overflow liquid is routed to a redundant tank or to the nearest atmospheric drainage point, as shown in drawings M-62, M-63 and M-64.

j. Processed wastes are collected in sample tanks prior to their reuse in the condensate transfer system or are discharged in a controlled manner through the monitored discharge pipe into the cooling tower blowdown pipe for dilution, before entering the Schuylkill River.

k. The expected and maximum radionuclide activity inventories of liquid waste management system components containing significant amounts of radioactive liquids are shown in Tables 11.2-8 and 11.2-9, and are based upon the assumptions given in Table 11.2-1 and those listed below:

1. Expected flow rates for streams (Figure 11.2-4) are given in Table 11.2-2.

2. Reactor water radionuclide activity concentrations are as listed in Tables 11.1-1 through 11.1-5 for design conditions, and Table 11.2-3 for expected conditions.

3. Decontamination factors within the liquid waste management system are as given in Table 11.2-4.

4. While a process stream is collecting in a collection tank, the isotopes already in the tank are undergoing radioactive decay (see Table 11.2-5 for expected holdup times).

5. One refueling shutdown per year per unit, not occurring simultaneously is assumed.

The expected daily inputs and activities for each of the subsystems is shown in Tables 11.2-6 and 11.2-7. An evaluation of the causes for the maximum expected inputs for each subsystem shows that operational modes exclude, and the unlikely occurrence of the same failure in both units minimizes, the potential for coincidental maximum input from both units into the same subsystem.

Control and monitoring of radioactive release in accordance with GDC 60 and GDC 64 is discussed in Sections 11.2.3 and 11.5.

11.2.2 SYSTEM DESCRIPTION CHAPTER 11 11.2-2 REV. 16, SEPTEMBER 2012

LGS UFSAR The liquid waste management system collects, monitors, processes, stores, and disposes of radioactive liquid wastes. The liquid waste management system piping, equipment, instrumentation, and flow paths are shown in drawings M-62, M-63 and M-64. Included in the system are:

a. Piping and equipment carrying potentially radioactive wastes

b. Floor drain systems in controlled access areas which may contain potentially radioactive wastes

c. Tanks and sumps used to collect potentially radioactive wastes The liquid wastes are collected in sumps located in all the enclosures housing radioactive equipment and are pumped to collection tanks located in the radwaste enclosure. Plant drainage systems, including measures taken to limit, control, and remove oil and other organic compounds, are discussed in Section 9.3.3.

The incoming wastes are classified, collected, and treated as floor drain, equipment drain, chemical, and laundry drain wastes.

Connections for tie-ins are provided, so that temporary equipment can draw suction from either the equipment drain surge tank (00T304) or floor drain surge tank (00T309) in order to process liquid radwaste when high demand is placed on the existing liquid radwaste processing equipment. After processing through the temporary equipment, the effluent stream will be recirculated back to either the equipment drain/floor drain surge tanks or routed to the equipment drain/floor drain sample tanks.

Equipment is selected, arranged, and shielded to minimize exposures to plant personnel during operation, inspection and maintenance. For example, sumps, pumps, valves, and instruments which may contain radioactivity are located in controlled access areas. Tanks and processing equipment that may contain significant quantities of radioactive material are shielded. Operation of the liquid waste management system is normally on a batch basis, as dictated by the waste generation rate from the plant. Protection against accidental discharge is provided by instrumentation for detection and alarm of abnormal conditions and by procedural controls.

The liquid waste management system is divided into several subsystems (Section 11.2.2.1), so that the liquid wastes from various sources can be kept segregated and processed separately.

Cross-connections between the subsystems provide additional flexibility for processing of the wastes by alternate methods.

11.2.2.1 System Operation The liquid waste management system consists of four process subsystems:

a. Equipment drain subsystem

b. Floor drain subsystem

c. Chemical waste subsystem

d. Laundry drain subsystem CHAPTER 11 11.2-3 REV. 16, SEPTEMBER 2012

LGS UFSAR 11.2.2.1.1 Equipment Drain Subsystem Low conductivity wastes from piping and equipment drains are collected in the equipment drain collection tank and/or the equipment drain surge tank. Other inputs received in the collection tank are listed in Table 11.2-7. Wastes collected in the equipment drain collection or surge tanks are processed on a batch basis through a precoat filter and mixed bed demineralizer and are then collected in one of two sample tanks.

From an equipment drain sample tank, wastes are normally returned to a CST for plant reuse. A recycle routing allows water which exceeds preset quality limits to be: recycled to the equipment drain collection tank for additional processing through the filter and demineralizer or recycled to the floor drain collection tank for additional processing. If the water inventory of the plant does not permit wastes to be returned to the CST, or if the wastes are unacceptable for plant reuse, then the wastes are routed to the floor drain subsystem and discharged from the plant after mixing with cooling tower blowdown.

11.2.2.1.2 Floor Drain Subsystem Wastes originating mainly from the drywell, reactor, turbine, radwaste, and offgas enclosure floor drains are collected in the floor drain collection tank. In addition, high conductivity piping and equipment drains are also collected and treated with these wastes.

The wastes collected in the floor drain collection tank are processed through a precoat filter and, on an as needed basis, through the mixed bed demineralizer, bypassing floor drain sample tank No. 1, and discharged to floor drain sample tank No. 2 for final sampling and analysis. Provision is available to discharge directly to sample tank No. 1 for use when warranted. The basis for selecting the treatment path consists of water quality, equipment availability, and economic considerations. If the quality of the filtered waste in the sample tank No. 2 is unsuitable for plant reuse, then the flexibility exists to reprocess the batch through the floor drain system and floor drain demineralizer or to recycle it to the equipment drain collection tank.

Treated floor drain wastes are routinely discharged from the plant after mixing with cooling tower blowdown. If the treated wastes meet the specifications of water quality used in the plant (including reactor water chemistry parameters), and if the water inventory of the plant permits their recycle, then they may be returned to the CST for reuse.

11.2.2.1.3 Chemical Waste Subsystem Chemical wastes collected in the chemical waste tank consist of laboratory wastes, decontamination solutions and other corrosive wastes. After being accumulated in the chemical waste tank, these wastes are chemically neutralized, if required, and transferred to the floor drain collection tank or floor drain surge tank for batch processing through the floor drain subsystem.

The chemical waste subsystem is designed to permit addition of an evaporation system as an alternate means of waste processing. However, installation of this equipment was not completed for initial plant operation and the equipment has been abandoned.

11.2.2.1.4 Laundry Drain Subsystem Laundry wastes consist of detergent-containing water from personnel decontamination facilities in the plant. These wastes are routed to two laundry drain tanks interconnected by an overflow line.

In addition, rain water from within the yard dikes for Condensate Storage Tank (CST) tanks (Units CHAPTER 11 11.2-4 REV. 16, SEPTEMBER 2012

LGS UFSAR 1 & 2) and Refueling Water Storage Tank may be routed to the two laundry drain tanks. From the tank, the wastes are processed through the laundry drain filter and collected in the sample tank for sampling and analysis.

Effluent from the sample tank is normally discharged through the monitored discharge pipe into the cooling tower blowdown pipe. High conductivity filtrate can be recycled to the laundry drain tanks or to the floor drain system.

The laundry drain system is no longer in service to process and discharge liquid waste. However, a valve remains open to a laundry drain tank for inputs from personnel decontamination facilities.

11.2.2.2 Process Equipment Description Major components of the liquid waste management system are described below. A summary of the design parameters for major components is provided in Table 11.2-10.

11.2.2.2.1 Pumps

a. Sump Pumps Liquid waste management system sump pumps are vertical, centrifugal-type pumps of standard design, cast iron construction. The top column closures on the pump support plates utilize vapor-tight gland-type seals. Each sump is provided with two pumps except the offgas enclosure floor drain sump which utilizes one pump.

Sump pumps start automatically when a predetermined high level in the sump is reached; the sump pump stops at a predetermined low water level.

b. Process Pumps Liquid waste management system process pumps are vertical, in-line centrifugal pumps of ASME Section III, Class 3 design, stainless steel or carbon steel construction, and are provided with a mechanical-type seal.

11.2.2.2.2 Tanks Tanks are sized to accommodate the anticipated volumes generated from the four liquid waste management subsystems. The tanks are constructed of carbon steel, stainless steel, or aluminum.

The tanks are provided with a mixing eductor and vented to the radwaste enclosure ventilation system. Refer to Section 9.4.3 for a discussion of the radwaste enclosure ventilation system.

11.2.2.2.3 Filters

a. Equipment and Floor Drain Filters Equipment and floor drain filters are a precoat, removable, single element bundle-type design, using a powdered ion exchange resin or other filter media.

CHAPTER 11 11.2-5 REV. 16, SEPTEMBER 2012

LGS UFSAR Both filters are equipped with a common precoating tank and pump, and individual filter holding pumps.

Filter media lifetime is based on a pressure drop across the filter and not specific radioactive content. Both filters are backwashed to the waste sludge tank. The filter vessels are constructed of carbon steel and are designed to meet the requirements of ASME Section III, Class 3. A corrosion-resistant lining is provided to minimize corrosion of the carbon steel vessels.

Each filter is located in a separate shielded room to minimize exposures to personnel during routine maintenance. Use of condensate water for backwashing and design of the filter internals ensure a minimum accumulation of radioactive material in the filters.

b. Laundry Drain Filter The laundry drain filter is of a removable cartridge- type design. The filter vessel is constructed of stainless steel and designed to meet the requirements of ASME Section III, Class 3. Filter construction also permits "quick change" replacement of disposable cartridges to minimize personnel exposure to radiation.

11.2.2.2.4 Demineralizers The equipment and floor drain demineralizers are deep-bed, mixed cation and anion-type, with flow rate capacities consistent with their associated filters.

A resin bed is replaced when effluent quality exceeds preset limits, as well as on high differential pressure.

The demineralizer vessels are constructed of carbon steel and are designed to meet the requirements of ASME Section III, Class 3. The vessels are equipped with a rubber lining. Fine mesh strainers are provided in the vessel discharge and in the piping downstream to prevent resin fines from being transferred to other portions of the system.

Both demineralizers were designed to be backwashed to their respective spent resin tanks.

However, for normal plant operation, the demineralizers will be backwashed directly to the waste sludge tank via a bypass line (Section 11.4.2.1.1). New mixed dry resins are added to the demineralizer through a resin addition funnel and flow by gravity into the demineralizer to provide an even distribution. If desired, the capability exists for mixing the resin bed inside the vessel with air. Resins can also be supplied from the Deep Bed Condensate Demineralizer (DBCD) system.

These resins are hydropneumatically transferred into the radwaste demineralizers (Section 10.4.6.2 ).

Each vessel is located in a separate shielded room to minimize exposures to personnel during routine maintenance of other equipment.

11.2.2.2.5 Radwaste Evaporators Radwaste evaporators were not made operational for initial plant operation and have been abandoned (Section 11.2.2.1.3).

11.2.2.2.6 Piping CHAPTER 11 11.2-6 REV. 16, SEPTEMBER 2012

LGS UFSAR Liquid waste management system piping is routed to minimize crud traps. Radioactive piping is located in shielded areas to minimize personnel exposures. The floor drain, equipment drain, and laundry drain subsystem piping is constructed of carbon steel. The chemical waste subsystem piping is constructed of stainless steel. Liquid waste management system piping is designed to meet the requirements of ANSI B31.1.

11.2.3 RADIOACTIVE RELEASES During processing of liquid wastes, radioactivity is removed so that the bulk of the liquid is restored to clean water, which is either recycled in the plant or discharged to the environment.

The radioactivity removed from the liquids is concentrated in filters and ion exchange resins.

These concentrated wastes are sent to the solid waste management system for treatment and eventual shipment to a licensed burial ground. If the liquid is to be discharged, the activity concentration is consistent with the discharge criteria of 10CFR20. Normally, most of the liquid passing through the liquid waste management system is recycled in the plant. However, the treatment in this system is such that these liquids can be discharged from the plant after monitoring, if required by plant water balance considerations or if the treated wastes do not meet the specifications of water quality used in the plant (including reactor water chemistry parameters). Activity concentration of tritiated water discharged from the system is consistent with the discharge criteria of 10CFR20. Normally, the liquid passing through the laundry drain subsystem is discharged directly, in accordance with 10CFR20 and 10CFR50 guidelines; however, it may be processed through the floor drain subsystem if necessary.

The expected annual activity releases for each waste stream are given in Table 11.2-11.

Design and administrative controls are incorporated into the liquid waste management system to prevent inadvertent releases to the environment. Controls include administrative procedures, operator training, redundant discharge valves, and discharge radiation monitors that trip alarms and initiate automatic discharge valve closure (Section 11.5). Prior to any discharging, activity concentrations are measured in samples taken from the various sample tanks. A single line is provided for plant discharges to minimize the potential for operator error. This line includes a loop seal with a siphon breaker to prevent inadvertent siphoning of the sample tanks.

The processed liquid waste that is not recycled in the plant is discharged into the cooling tower blowdown pipe on a batch basis. Flow rate measurement devices are provided in both the radwaste effluent line and the cooling tower blowdown line as shown on drawings M-63 and M-09, respectively. Processed liquid wastes are discharged at up to 280 gpm from the equipment and floor drain subsystems, and 10 gpm from the laundry drain subsystem. The discharges are mixed with the cooling tower blowdown flow to maintain the concentration of radionuclides at the release point below the limits of 10CFR20 (UFSAR Table 11.2-12). This mixing occurs within the site boundary and is used in determining specific activity concentrations for the releases.

Expected average annual radionuclide concentrations are compared to 10CFR20 limits in Table 11.2-12.

The doses resulting from liquid effluents are a small fraction of the 10CFR20 dose limit of 500 mR/yr and well within 10CFR50, Appendix I design objectives as shown in Table 11.2-13.

CHAPTER 11 11.2-7 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.2-1 ASSUMPTIONS AND PARAMETERS USED FOR EVALUATION OF RADIOACTIVITY RELEASES ITEM VALUE OR REFERENCE

a. Nuclear Steam Supply System

1. Maximum core thermal power 3527 MWt

2. Mass of reactor coolant in 5.5x105 lb vessel at full power

3. Total steam flow for 102% 1.53x107 lb/hr rated power

b. Reactor Water Cleanup System

1. Average process flow rate 1.33x105 lb/hr for 2 vessels

2. Number of demineralizers 2

3. Powdex resin quantity per 32 lb demineralizer (dry)

4. Backwash frequency 3.3 days (6.6 day run for each demineralizer)

5. Backwash volume 1100 gal/backwash

c. Condensate Filter/Demineralizers

1. Average total flow rate for 1.53x107 lb/hr 8 vessels 102% of rated power condition)

2. Number of filter/demineralizers 8

3. Backwash frequency 1.43 days (10 day run for each filter/demin)

4. Backwash volume 9000 gal/backwash

d. Liquid Waste Management System

1. Sources, flow rates, and Table 11.2-2 expected activities in process streams CHAPTER 11 11.2-8 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-1 (Cont'd)

ITEM VALUE OR REFERENCE

2. Holdup times for collection, Table 11.2-5 processing, and discharge

3. Capacities of tanks and Table 11.2-10 processing equipment

4. Decontamination factors Table 11.2-4

5. Liquid source terms for Table 11.2-6 normal operation

6. Quantity of tritium - 11 Ci/yr released in liquid effluents (2 units)

7. Average fraction of process streams discharged to environment after processing(1)

(a) Floor drain subsystem 0.10 (b) Equipment drain subsystem 0.01 (c) Chemical drain subsystem 0.10 (d) Laundry drain subsystem 1.00

8. Liquid waste management system demineralizer resin replacement frequency (a) Floor drain 25.8 days (b) Equipment drain 125.8 days

9. Liquid waste management system demineralizer backwash volume (a) Floor drain 1500 gal (b) Equipment drain 1500 gal (1)

These values are taken from NUREG-0016 (April 1976), section 1.5.2.9.5.

CHAPTER 11 11.2-9 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-2 LIQUID WASTE MANAGEMENT SYSTEM FLOWS AVERAGE BATCH AVERAGE VOLUME INTERVAL FOR PER DAY FOR NORMAL MAXIMUM NORMAL OPERATION NORMAL OPERATION ACTIVITY ACTIVITY STREAM OF BOTH UNITS VOLUME PER NOMINAL FLOW OF BOTH UNITS CONCENTRATION CONCENTRATION NO.(1) (days) BATCH (gal) RATE (gpm) (gal) (Ci/cc) (Ci/cc)

Floor Drain Subsystem 1 - - - 10400 6.2x10-2 1.1 2 1.6 16800 280 10400 1.1x10-2 2.9x10-1 3 1.6 16800 280 10400 1.1x10-3 3.0x10-2 4 1.6 16800 280 10400 2.1x10-5 8.2x10-4 5 1.6 16800 280 10400 1.4x10-5 3.6x10-4 Equipment Drain Subsystem 6 - - - 38780 7.8x10-2 1.4 7 0.53 20000 280 38780 2.6x10-2 5.8x10-1 8 0.53 20000 280 38780 2.6x10-3 6.1x10-2 9 0.53 20000 280 38780 3.2x10-5 9.6x10-4 10 0.53 20000 280 38780 2.1x10-5 5.1x10-4 Chemical Drain Subsystem (2) 11 - - - 1200 8.7x10-3 1.6x10-1 12 5 6000 200 1200 7.3x10-4 2.3x10-2 13 5 6000 280 1200 7.1x10-4 2.2x10-2 14 5 6000 280 1200 7.2x10-5 2.2x10-3 15 5 6000 280 1200 1.4x10-6 4.2x10-5 16 5 6000 280 1200 1.2x10-6 3.0x10-5 Laundry Drain Subsystem 17 - - - 900 1.0x10-4 -

18 1 900 25 900 1.0x10-4 -

19 1 900 25 900 1.0x10-4 -

20 2 1800 10 900 1.0x10-4 -

(1) Refer to Figure 11.2-4 for locations of stream nodal numbers.

(2) Batch frequencies, volumes, and activity concentrations are based on chemical waste processing via the floor drain subsystem (Section 11.2.2.1.3).

CHAPTER 11 11.2-10 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-3 EXPECTED RADIONUCLIDE ACTIVITY CONCENTRATIONS IN REACTOR COOLANT AND MAIN STEAM USED FOR EVALUATION OF RADIOACTIVITY RELEASES(1)

REACTOR REACTOR WATER STEAM ISOTOPE (Ci/g) (Ci/g)

Noble gases Kr-83m - 1.12E-03 Kr-85m - 1.94E-03 Kr-85 - 6.12E-06 Kr-87 - 6.73E-03 Kr-88 - 6.73E-03 Kr-89 - 4.18E-02 Kr-90 - 9.18E-02 Kr-91 - 1.12E-01 Kr-92 - 1.12E-01 Kr-93 - 2.96E-02 Kr-94 - 7.34E-03 Kr-95 - 6.73E-04 Kr-97 - 4.49E-06 Xe-131m - 4.79E-06 Xe-133m - 9.18E-05 Xe-133 - 2.65E-03 Xe-135m - 8.57E-03 Xe-135 - 7.34E-03 Xe-137 - 4.79E-02 Xe-138 - 2.86E-02 Xe-139 - 9.18E-02 Xe-140 - 9.79E-02 Xe-141 - 7.96E-02 Xe-142 - 2.35E-02 Xe-143 - 3.88E-03 Xe-144 - 1.84E-04 Halogens Br-83 2.43E-03 4.86E-05 Br-84 3.70E-03 7.41E-05 Br-85 2.15E-03 4.30E-05 I-131 5.03E-03 1.01E-04 CHAPTER 11 11.2-11 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.2-3 (Cont'd)

REACTOR REACTOR WATER STEAM ISOTOPE (Ci/g) (/Ci/g)

I-132 2.42E-02 4.83E-04 I-133 1.91E-02 3.81E-04 I-134 5.29E-02 1.06E-03 I-135 1.76E-02 3.53E-04 Fission products Rb-89 3.63E-03 3.63E-06 Sr-89 1.01E-04 1.01E-07 Sr-90 6.08E-06 6.08E-09 Sr-91 3.77E-03 3.77E-06 Sr-92 9.61E-03 9.61E-06 Y-91 4.05E-05 4.05E-08 Y-92 5.27E-03 5.27E-06 Y-93 3.78E-03 3.78E-06 Zr-95 7.08E-06 7.08E-09 Zr-97 4.85E-06 4.85E-09 Nb-95 7.08E-06 7.08E-09 Nb-98 3.12E-03 3.12E-06 Mo-99 2.00E-03 2.00E-06 Tc-99m 1.84E-02 1.84E-05 Tc-101 6.62E-02 6.62E-05 Tc-104 5.93E-02 5.93E-05 Ru-103 2.02E-05 2.02E-08 Ru-105 1.79E-03 1.79E-06 Ru-106 3.04E-06 3.04E-09 Ag-110m 1.01E-06 1.01E-09 Te-129m 4.05E-05 4.05E-08 Te-131m 9.86E-05 9.86E-08 Te-132 1.00E-05 1.00E-08 Cs-134 3.04E-05 3.04E-08 Cs-136 2.02E-05 2.02E-08 Cs-137 7.09E-05 7.09E-08 Cs-138 7.42E-03 7.42E-06 Ba-139 8.10E-03 8.10E-06 Ba-140 4.04E-04 4.04E-07 Ba-141 7.42E-03 7.42E-06 Ba-142 4.39E-03 4.39E-06 La-142 4.09E-03 4.09E-06 Ce-141 3.03E-05 3.03E-08 Ce-143 2.97E-05 2.97E-08 CHAPTER 11 11.2-12 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.2-3 (Cont'd)

REACTOR REACTOR WATER STEAM ISOTOPE (Ci/g) (Ci/g)

Ce-144 3.04E-06 3.04E-09 Pr-143 4.04E-05 4.04E-08 Nd-147 3.03E-06 3.03E-09 W-187 2.94E-04 2.94E-07 Np-239 6.99E-03 6.99E-06 Coolant activation products N-13 5E-02 7E-03 N-16* 6E+01 5E+01 N-17 9E-03 2E-02 O-19 7E-01 2E-01 F-18 4E-03 4E-03 Noncoolant activation products Na-24 8.68E-03 8.68E-06 P-32 2.02E-04 2.02E-07 Cr-51 5.06E-03 5.06E-06 Mn-54 6.07E-05 6.07E-08 Mn-56 4.27E-02 4.27E-05 Fe-55 1.01E-03 1.01E-06 Fe-59 3.04E-05 3.04E-08 Co-58 2.02E-04 2.02E-07 Co-60 4.05E-04 4.05E-07 Ni-63 1.01E-06 1.01E-09 Ni-65 2.56E-04 2.56E-07 Cu-64 2.88E-02 2.88E-05 Zn-65 2.02E-04 2.02E-07 Zn-69 1.92E-03 1.92E-06 Tritium H-3 1E-02 1E-02 (1) The values in this table are calculated based on the GALE code in NUREG-0016 (April 1976).

• The N-16 values listed in the table are without Hydrogen Water Chemistry (HWC). Values with HWC are 4.80E+01 (water) and 2.50E+02 (steam).

CHAPTER 11 11.2-13 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.2-4 DECONTAMINATION FACTORS USED FOR EVALUATION OF RADIOACTIVITY RELEASES(1)

CESIUM AND EQUIPMENT IODINE RUBIDIUM OTHERS Equipment drain filter/

demineralizer(2) 10 2 10 Equipment drain 100 10 100 demineralizer Floor drain filter/

demineralizer(2) 10 2 10 Floor drain 100 2 100 demineralizer RWCU 10 2 10 filter/demineralizer Condensate 10 2 10 filter/demineralizer Laundry drain 1 1 1 cartridge filter (1)

The values are taken from NUREG-0016 (April 1976), table 1-3.

(2)

Powdered resin is used to precoat the filter/demineralizer; decontamination factor value of Powdex is used.

CHAPTER 11 11.2-14 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-5 EXPECTED HOLDUP TIMES FOR COLLECTION, PROCESSING, AND DISCHARGE USED FOR EVALUATION OF RADIOACTIVITY RELEASES HOLDUP TIME PROCESS SUBDIVISION (days)

Floor Drain Subsystem Collection 1.616 Processing 0.042 Sampling 0.042 Total 1.700 Equipment Drain Subsystem Collection 0.519 Processing 0.050 Sampling 0.042 Total 0.611 Chemical Drain Subsystem(1)

Collection 5.000 Processing 0.063 Sampling 0.042 Total 5.105 Laundry Drain Subsystem Collection 1.000 Processing 0.025 Sampling 0.042 Total 1.067 (1)

Holdup times shown for the chemical drain subsystem are based on processing via the floor drain subsystem (Section 11.2.2.1.3)

CHAPTER 11 11.2-15 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-6 AVERAGE DAILY INPUTS AND ACTIVITIES TO THE LIQUID WASTE MANAGEMENT SYSTEM FROM TWO UNITS AVERAGE DAILY INPUT FROM TWO PRIMARY COOLANT UNITS IN NORMAL ACTIVITY FRACTION SOURCE OPERATION(1)(gal) (PCA)

Floor drains Drywell 1400 1.0000 Reactor enclosure 4000 0.0100 Turbine enclosure - condensate pump area 1000 0.0100

- backwash area 3000 0.0100 Radwaste enclosure 1000 0.0100 TOTAL 10400 0.1430 Equipment drains Drywell 6800 1.0000 Reactor enclosure 7440 0.0100 Turbine enclosure - condensate pump area 2000 0.0100

- backwash area 3920 0.0100 Radwaste enclosure 1060 0.0100 TOTAL 21220 0.3270 Decant water RWCU phase separator 620 0.0020 Condensate phase separator 11983 0.0002 External processing effluent 5537 0.0020 TOTAL 18139 0.0008 Chemical wastes Lab drains 1000 0.0200 Chemical lab drains 200 0.0200 TOTAL 1200 0.0200 Laundry drains 900 -

(1) These values are taken directly from NUREG-0016 (April 1976).

CHAPTER 11 11.2-16 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-7 BATCHED INPUTS TO THE LIQUID WASTE MANAGEMENT SYSTEM FROM THE SOLID WASTE MANAGEMENT SYSTEM FOR NORMAL OPERATION OF TWO UNITS LIQUID RADWASTE INPUT TO EQUIPMENT DRAIN COLLECTION TANK FROM EACH SECOND INTERMEDIATE FIRST INTERMEDIATE COLLECTOR INPUT SECOND INTERMEDIATE COLLECTOR INPUT COLLECTOR BATCH BATCH FREQUENCY FREQUENCY PER PER BATCH BATCH SIZE COLLECTOR BATCH SIZE COLLECTOR BATCH SIZE FREQUENCY SOURCE COLLECTOR (gal) (days) COLLECTOR (gal) (days) (gal) (days)

RWCU filter/demineralizers RWCU backwash 1100 3.3 RWCU phase 1100 3.4 1000 1.7 (total of 4) receiving tanks separators (total of 2) (total of 2)

Condensate Condensate 9000 1.43 Condensate 9000 1.43 8300 0.72 filter/demineralizers backwash receiving phase (total of 16) tanks separators (total of 2) (total of 4)

Deep bed condensate Spend resin tanks 3675 TDD None 3675 TDD demineralizers (Total of 4) (w/320 cu. ft.

(Total of 8) of resin)

Spent resin tanks Resin measuring 375 TBD None - - 375 TBD (Total of 4) tank (w/85 cu. ft.

of resin)

CHAPTER 11 11.2-17 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-8 EXPECTED RADIONUCLIDE ACTIVITY INVENTORIES OF LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS (1)

FLOOR FLOOR FLOOR DRAIN DRAIN FLOOR EQUIPMENT EQUIPMENT EQUIPMENT LAUNDRY LAUNDRY DRAIN SAMPLE SAMPLE DRAIN DRAIN DRAIN DRAIN CHEMICAL DRAIN DRAIN COLLECTION TANK TANK SURGE COLLECTION SAMPLE SURGE WASTE COLLECTION SAMPLE ISOTOPE TANK NO. 1 NO. 2 TANK TANK TANK TANK TANK TANK TANK Br-83 1.98E-03 1.49E-04 1.49E-06 1.98E-03 8.99E-03 6.36E-06 9.24E-03 3.19E-05 none none Br-84 6.64E-04 1.80E-05 1.80E-05 6.64E-04 3.10E-03 6.46E-07 3.10E-03 1.07E-05 none none Br-85 3.64E-05 none none 3.64E-05 1.70E-04 none 1.70E-04 5.88E-07 none none I-131 4.27E-02 4.25E-03 4.25E-05 1.29E-01 6.71E-02 6.68E-05 1.93E-01 1.86E-03 3.21E-06 6.17E-06 I-132 1.86E-02 1.37E-03 1.37E-05 1.87E-02 8.46E-02 5.85E-05 8.66E-02 3.01E-04 none none I-133 9.62E-02 9.29E-03 9.29E-05 1.30E-01 2.12E-01 2.03E-04 4.41E-01 2.08E-03 none none I-134 1.55E-02 6.98E-04 6.98E-06 1.55E-02 7.24E-02 2.77E-05 7.25E-02 2.51E-04 none none I-135 3.92E-02 3.53E-03 3.53E-05 3.99E-02 1.35E-01 1.19E-04 1.83E-01 6.44E-04 none none Rb-89 3.15E-04 1.06E-05 5.30E-04 3.15E-04 1.47E-03 2.89E-06 1.47E-03 5.09E-06 none none Cs-134 2.76E-04 1.38E-04 7.65E-06 9.84E-04 4.14E-04 2.07E-05 1.24E-03 1.38E-05 7.26E-05 1.45E-04 Cs-136 1.76E-04 8.80E-05 4.41E-05 5.68E-04 2.71E-04 1.36E-05 7.96E-04 8.10E-06 none none Cs-137 6.45E-04 3.22E-04 1.61E-04 2.31E-03 9.67E-04 4.83E-05 2.91E-03 3.21E-04 1.34E-04 2.68E-04 Cs-138 1.35E-03 1.86E-04 9.28E-05 1.35E-03 6.28E-03 6.67E-05 6.28E-03 2.17E-05 none none Na-24 3.66E-02 3.50E-03 3.50E-05 4.39E-02 9.00E-02 8.51E-05 1.69E-01 7.06E-04 none none P-32 1.76E-03 1.76E-04 1.76E-06 5.72E-03 2.72E-03 2.71E-06 7.98E-03 8.14E-05 none none Cr-51 4.51E-02 4.51E-03 4.51E-05 1.53E-01 6.85E-02 6.84E-05 2.03E-01 2.16E-03 none none Mn-54 5.51E-04 5.51E-05 5.51E-07 1.96E-03 8.27E-04 8.27E-07 2.48E-03 2.74E-05 5.59E-06 1.11E-05 Mn-56 3.72E-02 2.85E-03 2.85E-05 3.72E-02 1.68E-01 1.21E-04 1.73E-01 6.01E-04 none none Fe-55 9.21E-03 9.21E-04 9.21E-06 3.08E-02 1.29E-02 1.29E-05 3.86E-02 4.30E-04 none none Fe-59 2.73E-04 2.73E-04 2.73E-06 9.45E-04 4.13E-04 4.12E-07 1.23E-03 1.33E-05 none none Co-58 1.83E-03 1.83E-04 1.83E-06 6.38E-03 2.74E-03 2.74E-06 8.21E-03 8.95E-05 2.22E-05 4.44E-05 Co-60 3.68E-03 3.68E-06 3.68E-06 1.32E-02 5.52E-03 5.52E-06 1.66E-02 1.84E-04 5.04E-05 1.01E-04 Ni-63 9.21E-06 9.21E-07 9.81E-09 3.28E-05 1.38E-05 1.38E-08 4.15E-05 4.60E-07 none none Ni-65 2.22E-04 1.69E-05 1.69E-07 2.22E-04 1.00E-03 7.42E-07 10.4E-03 3.58E-06 none none Cu-64 1.09E-01 1.03E-02 1.03E-04 1.09E-01 2.86E-01 2.65E-04 2.86E-01 2.05E-03 none none Zn-65 1.83E-03 1.83E-04 1.83E-04 6.48E-03 2.75E-03 2.75E-06 8.26E-03 9.11E-05 none none Zn-69 6.16E-04 5.73E-05 5.73E-07 6.16E-04 2.88E-03 1.20E-06 2.88E-03 9.95E-06 none none Sr-89 9.17E-04 9.17E-05 9.17E-07 3.18E-03 1.29E-03 1.39E-06 4.13E-03 4.48E-05 none none Sr-90 5.53E-05 5.53E-06 5.53E-08 1.97E-04 8.29E-05 8.29E-08 2.49E-04 2.76E-06 none none Sr-91 1.16E-02 1.08E-03 1.08E-05 1.23E-02 3.41E-02 3.12E-05 5.38E-02 1.99E-04 none none Sr-92 7.87E-03 6.10E-04 6.10E-06 7.87E-03 3.53E-02 2.59E-05 3.67E-02 1.28E-04 none none Y-91 5.15E-04 5.21E-05 5.21E-07 2.00E-03 6.59E-04 6.77E-07 2.32E-03 2.78E-05 none none Y-92 1.42E-02 1.29E-03 1.29E-05 1.42E-02 5.50E-02 4.98E-05 6.61E-02 2.28E-04 none none Y-93 1.22E-02 1.14E-03 1.14E-05 1.32E-02 3.50E-02 3.22E-05 5.66E-02 2.13E-04 none none Zr-95 6.39E-05 6.38E-06 6.38E-08 2.23E-04 9.63E-05 9.62E-08 2.88E-04 3.13E-06 7.77E-06 1.54E-05 Zr-97 2.21E-05 2.12E-06 2.12E-08 2.77E-05 5.18E-05 4.94E-08 1.02E-04 4.47E-07 none none CHAPTER 11 11.2-18 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-8 (Cont'd)

FLOOR FLOOR FLOOR DRAIN DRAIN FLOOR EQUIPMENT EQUIPMENT EQUIPMENT LAUNDRY LAUNDRY DRAIN SAMPLE SAMPLE DRAIN DRAIN DRAIN DRAIN CHEMICAL DRAIN DRAIN COLLECTION TANK TANK SURGE COLLECTION SAMPLE SURGE WASTE COLLECTION SAMPLE ISOTOPE TANK NO. 1 NO. 2 TANK TANK TANK TANK TANK TANK TANK Nb-95 6.44E-05 6.44E-06 6.44E-08 2.30E-04 9.65E-05 9.65E-08 2.90E-04 3.21E-06 1.11E-05 2.22E-05 Nb-98 8.98E-04 3.97E-05 3.97E-05 8.98E-04 4.19E-03 1.57E-06 4.19E-03 1.45E-05 none none Mo-99 1.50E-02 1.48E-03 3.44E-02 3.44E-02 2.56E-02 2.53E-05 6.78E-02 5.19E-04 none none Tc-99m 4.75E-02 4.39E-03 4.39E-05 6.66E-02 1.44E-04 1.31E-04 2.20E-01 1.04E-03 none none Tc-101 5.22E-03 2.68E-05 2.68E-07 5.22E-03 2.44E-02 6.92E-07 2.44E-02 8.44E-05 none none Tc-104 6.02E-03 5.97E-05 5.97E-05 6.02E-03 2.81E-02 1.75E-06 2.81E-02 9.71E-05 none none Ru-103 1.82E-04 1.81E-05 1.81E-07 6.23E-04 2.74E-04 2.74E-07 8.17E-04 8.78E-06 7.75E-07 1.54E-06 Ru-105 2.67E-03 2.28E-04 2.28E-06 2.68E-06 1.07E-02 8.89E-06 1.24E-02 4.32E-05 none none Ru-106 2.76E-05 2.76E-06 2.76E-08 9.81E-05 4.14E-05 4.14E-08 1.24E-04 1.38E-06 1.34E-05 2.67E-05 Ag-110m 9.18E-06 9.18E-07 9.18E-09 3.62E-05 1.24E-03 1.38E-05 4.14E-05 4.56E-07 2.46E-06 4.91E-06 Te-129m 3.62E-04 3.62E-05 3.62E-07 1.24E-03 5.50E-04 5.49E-07 1.63E-03 1.74E-05 none none Te-131m 5.93E-04 5.79E-05 5.79E-07 9.60E-04 1.17E-03 1.13E-06 2.70E-03 1.51E-05 none none Te-132 7.70E-05 7.63E-06 7.63E-08 1.87E-04 1.30E-04 1.28E-07 3.49E-04 2.79E-06 none none Ba-139 3.78E-03 2.30E-04 2.30E-06 3.78E-03 1.76E-02 9.67E-06 1.76E-02 6.11E-05 none none Ba-140 3.52E-03 3.51E-04 3.51E-06 1.12E-02 5.43E-03 5.42E-06 1.59E-02 1.61E-04 none none Ba-141 7.53E-04 7.47E-06 7.47E-08 7.53E-04 3.52E-03 2.19E-07 3.52E-03 1.21E-05 none none Ba-142 2.72E-04 6.21E-07 6.21E-09 2.72E-04 1.28E-03 1.36E-08 1.28E-03 4.39E-06 none none La-142 2.41E-03 1.56E-04 1.56E-06 2.41E-03 1.12E-02 6.62E-06 1.12E-02 3.89E-05 none none Ce-141 2.93E-04 2.93E-05 2.93E-07 1.01E-03 4.32E-04 4.36E-07 1.32E-03 1.42E-05 none none Ce-143 1.85E-04 1.81E-05 1.81E-07 3.13E-04 3.56E-04 3.47E-07 8.43E-04 4.92E-06 none none Ce-144 2.75E-05 2.75E-06 2.75E-08 9.80E-05 4.14E-05 4.14E-08 1.24E-04 1.37E-06 2.79E-05 5.58E-05 Pr-143 3.61E-04 3.61E-05 3.61E-07 1.19E-03 5.49E-04 5.48E-07 1.63E-03 1.69E-05 none none Nd-147 2.62E-05 2.61E-06 2.61E-08 8.25E-05 4.06E-05 4.05E-08 1.18E-04 1.18E-06 none none W-187 1.60E-03 1.56E-04 1.56E-04 2.34E-03 3.37E-03 3.24E-06 7.43E-03 3.71E-05 none none Np-239 5.05E-02 4.99E-03 4.99E-05 1.09E-01 8.83E-02 8.70E-05 2.30E-01 1.65E-03 none none OTHERS 2.75E-02 2.04E-04 2.04E-06 2.02E-02 2.75E-02 2.71E-05 5.06E-02 3.01E-04 4.65E-05 8.43E-05 TOTAL 6.83E-01 6.02E-02 1.06E-03 1.08E+00 1.80E+00 1.66E-03 2.84E-00 1.67E-02 3.98E-04 7.86E-04 (1) Activity inventories are given in Curies.

(2) Activity of daughter products resulting from radioactive decay of the influent isotopes during the accumulation period.

CHAPTER 11 11.2-19 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-9 MAXIMUM RADIONUCLIDE ACTIVITY INVENTORIES OF LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS(1)

FLOOR FLOOR FLOOR DRAIN DRAIN FLOOR EQUIPMENT EQUIPMENT EQUIPMENT DRAIN SAMPLE SAMPLE DRAIN DRAIN DRAIN DRAIN CHEMICAL COLLECTION TANK TANK SURGE COLLECTION SAMPLE SURGE WASTE ISOTOPE TANK NO. 1 NO. 2 TANK TANK TANK TANK TANK Br-83 4.28x10-2 3.21x10-3 3.21x10-5 4.28x10-2 1.94x10-1 1.38x10-4 2.00x10-1 6.91x10-4 Br-84 1.69x10-2 4.58x10-4 4.58x10-6 1.69x10-2 7.91x10-2 1.65x10-5 7.91x10-2 2.73x10-4 Br-85 1.01x10-3 - - 1.01x10-3 4.70x10-3 - 4.70x10-3 1.62x10-5 I-131 3.86x10-1 3.85x10-2 3.85x10-4 1.16 6.03x10-1 6.04x10-4 1.73 1.68x10-2 I-132 1.55 1.49x10-1 1.49x10-3 3.49 3.12 2.85x10-3 7.00 5.24x10-2 I-133 1.57 1.52x10-1 1.52x10-3 2.12 3.45 3.33x10-3 7.19 3.40x10-2 I-134 2.46x10-1 1.11x10-2 1.11x10-4 2.46x10-1 1.15 4.40x10-4 1.15 3.97x10-3 I-135 1.01 9.10x10-2 9.10x10-4 1.03 3.47 3.07x10-3 4.70 1.66x10-2 Rb-89 - - - - - - - -

Cs-134 5.09x10-3 2.54x10-3 1.27x10-3 1.81x10-2 7.59x10-3 3.80x10-4 2.27x10-2 2.54x10-4 Cs-136 3.36x10-3 1.68x10-3 8.40x10-4 1.08x10-2 5.15x10-3 2.58x10-4 1.51x10-2 1.54x10-4 Cs-137 7.64x10-3 3.82x10-3 1.91x10-3 2.73x10-2 1.14x10-2 5.70x10-4 3.41x10-2 3.81x10-4 Cs-138 1.21x10-1 1.66x10-2 8.30x10-3 1.21x10-1 5.64x10-1 6.00x10-3 5.64x10-1 1.95x10-3 Na-24 8.44x10-3 8.06x10-4 8.06x10-6 1.01x10-2 2.06x10-2 1.96x10-5 3.88x10-2 1.63x10-4 P-32 1.75x10-4 1.75x10-5 1.75x10-7 5.66x10-4 2.68x10-4 2.68x10-7 7.83x10-4 8.06x10-6 Cr-51 4.46x10-3 4.45x10-4 4.45x10-6 1.51x10-2 6.73x10-3 6.73x10-6 1.99x10-2 2.13x10-4 Mn-54 3.63x10-4 3.63x10-5 3.63x10-7 1.29x10-3 5.42x10-4 5.42x10-7 1.62x10-3 1.81x10-5 Mn-56 4.36x10-2 3.33x10-3 3.33x10-5 4.36x10-2 1.96x10-1 1.42x10-4 2.03x10-1 7.03x10-4 Fe-55 - - - - - - - -

Fe-59 7.19x10-4 7.18x10-5 7.18x10-7 2.49x10-3 1.08x10-3 1.08x10-6 3.21x10-3 3.50x10-5 Co-58 4.51x10-2 4.51x10-3 4.51x10-5 1.58x10-1 6.76x10-2 6.76x10-5 2.02x10-1 2.22x10-3 Co-60 4.55x10-3 4.55x10-4 4.55x10-6 1.62x10-2 6.77x10-3 6.77x10-6 2.03x10-2 2.27x10-4 Ni-63 - - - - - - - -

Ni-65 2.60x10-4 1.99x10-5 1.99x10-7 2.60x10-4 1.17x10-3 8.48x10-7 1.22x10-3 4.20x10-6 Cu-64 - - - - - - - -

Zn-65 1.81x10-5 1.81x10-6 1.81x10-8 6.44x10-5 2.71x10-5 2.71x10-8 8.11x10-5 9.02x10-7 Zn-69 1.20x10-4 1.14x10-5 1.14x10-7 1.40x10-4 3.03x10-4 2.86x10-7 5.53x10-4 2.25x10-6 Sr-89 9.81x10-2 9.80x10-3 9.80x10-5 3.41x10-1 1.47x10-1 1.47x10-4 4.39x10-1 4.79x10-3 Sr-90 7.32x10-3 7.32x10-4 7.32x10-6 2.61x10-3 1.09x10-2 1.09x10-5 3.27x10-2 3.65x10-4 Sr-91 7.43x10-1 6.91x10-2 6.91x10-4 7.92x10-1 2.17 2.00x10-3 3.44 1.28x10-2 Sr-92 3.53x10-1 2.73x10-2 2.73x10-4 3.53x10-1 1.58 1.16x10-3 1.65 5.70x10-3 Y-91 - - - - - - - -

Y-92 - - - - - - - -

Y-93 - - - - - - - -

Zr-95 1.26x10-3 1.26x10-4 1.26x10-6 4.41x10-3 1.89x10-3 1.89x10-6 5.64x10-3 6.19x10-5 CHAPTER 11 11.2-20 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-9 FLOOR FLOOR FLOOR DRAIN DRAIN FLOOR EQUIPMENT EQUIPMENT EQUIPMENT DRAIN SAMPLE SAMPLE DRAIN DRAIN DRAIN DRAIN CHEMICAL COLLECTION TANK TANK SURGE COLLECTION SAMPLE SURGE WASTE ISOTOPE TANK NO. 1 NO. 2 TANK TANK TANK TANK TANK Zr-97 5.12x10-4 4.91x10-5 4.91x10-7 6.42x10-4 1.19x10-3 1.14x10-6 2.35x10-3 1.03x10-5 Nb-95 1.34x10-3 1.34x10-4 1.34x10-6 4.75x10-3 1.99x10-3 1.99x10-6 5.97x10-3 6.65x10-5 Nb-98 - - - - - - - -

Mo-99 5.76x10-1 5.70x10-2 5.70x10-4 1.33 9.79x10-1 9.72x10-4 2.60 2.00x10-2 Tc-99m 2.38 2.18x10-1 2.18x10-3 3.12 7.50 6.67x10-3 1.10x10+1 4.90x10-2 Tc-101 3.87x10-2 1.98x10-4 1.98x10-6 3.87x10-2 1.81x10-1 5.12x10-6 1.81x10-1 6.24x10-4 Tc-104 - - - - - - - -

Ru-103 5.96x10-4 5.96x10-5 5.96x10-7 2.05x10-3 8.97x10-4 8.97x10-7 2.67x10-3 2.89x10-5 Ru-105 - - - - - - - -

Ru-106 8.26x10-5 8.26x10-6 8.26x10-8 2.94x10-4 1.23x10-4 1.23x10-7 3.69x10-4 4.11x10-6 Ag-110m 5.44x10-4 5.44x10-5 5.44x10-7 1.93x10-3 8.12x10-4 8.12x10-7 2.43x10-3 2.71x10-5 Te-129m 1.25x10-3 1.25x10-4 1.25x10-6 4.29x10-3 1.89x10-3 1.89x10-6 5.60x10-3 6.04x10-5 Te-131m - - - - - - - -

Te-132 1.32 1.31x10-1 1.31x10-3 3.21 2.21 2.20x10-3 5.95 4.79x10-2 Ba-139 2.62x10-1 1.59x10-2 1.59x10-4 2.62x10-1 1.22 6.68x10-4 1.22 4.22x10-3 Ba-140 2.74x10-1 2.74x10-2 2.74x10-4 8.78x10-1 4.21x10-1 4.21x10-4 1.23 1.25x10-2 Ba-141 6.04x10-2 5.99x10-4 5.99x10-6 6.04x10-2 2.82x10-1 1.76x10-5 2.82x10-1 9.75x10-4 Ba-142 3.69x10-2 8.42x10-5 8.42x10-7 3.69x10-2 1.72x10-1 1.85x10-6 1.72x10-1 5.96x10-4 La-142 - - - - - - - -

Ce-141 2.96x10-3 3.01x10-4 3.01x10-6 1.09x10-2 3.60x10-3 3.87x10-6 1.32x10-2 1.53x10-4 Ce-143 7.65x10-4 7.49x10-5 7.49x10-7 1.30x10-3 1.47x10-3 1.44x10-6 3.47x10-3 2.04x10-5 Ce-144 1.12x10-3 1.12x10-4 1.12x10-6 3.97x10-3 1.67x10-3 1.67x10-6 4.99x10-3 5.55x10-5 Pr-143 1.20x10-3 1.19x10-4 1.19x10-6 3.98x10-3 1.80x10-3 1.80x10-6 5.35x10-3 5.66x10-5 Nd-147 4.24x10-3 4.23x10-4 4.23x10-6 1.33x10-2 6.53x10-3 6.53x10-6 1.90x10-2 1.91x10-4 W-187 1.64x10-2 1.59x10-3 1.59x10-5 2.38x10-2 3.42x10-2 3.32x10-5 7.47x10-2 3.79x10-4 Np-239 6.07 6.00x10-1 6.00x10-3 1.31x10+1 1.06x10+1 1.05x10-2 2.74x10+1 1.99x10-1 OTHERS(2) 9.64x10-1 9.52x10-2 9.52x10-4 1.54 2.91 2.87x10-3 4.48 2.31x10-2 TOTAL 1.83x10+1 1.73 2.95x10-2 3.37x10+1 4.34x10+1 4.56x10-2 8.38x10+1 5.14x10-1 (1)

Activity inventories are given in Curies.

(2)

Activity of daughter products resulting from radioactive decay of the influent isotopes during the accumulation period.

CHAPTER 11 11.2-21 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-10 LIQUID WASTE MANAGEMENT SYSTEM COMPONENT PARAMETERS A. TANKS DESIGN PRESSURE/TEMP CAPACITY, EACH DIAMETER QUANTITY (psig/ F) TYPE MATERIAL (gal) HEIGHT Equipment drain collection tank 1 Atmos/212 Vert cyl CS 25,000 20'/11' Equipment drain sample tanks 2 Atmos/212 Vert cyl Aluminum 25,000 16'/17' Equipment drain surge tank 1 Atmos/212 Vert cyl CS 75,000 32'/13' Floor drain collection tank 1 Atmos/212 Vert cyl CS 21,000 15'/16' Floor drain sample tank #1 1 Atmos/212 Vert cyl CS 21,000 15'/16' Floor drain sample tank #2 1 Atmos/212 Vert cyl Aluminum 21,000 15'/16' Floor drain surge tank 1 Atmos/212 Vert cyl CS 75,000 32'/13' Chemical waste tank 1 Atmos/212 Vert cyl SS 7,500 10'/13' (1)

Evaporator feed tank 1 Atmos/212 Vert cyl CS 7,500 11'/10.5' Evaporator distillate sample tank(1) 1 Atmos/212 Vert cyl Aluminum 7,500 11'/10' Laundry drain tanks 2 Atmos/212 Vert cyl CS 1,000 5.5'/6' Laundry drain sample tank 1 Atmos/212 Vert cyl CS 2,000 7'/7' 3

Backwash air accumulator 1 125/110 Vert cyl CS 90 ft 4'/6.75' Precoat tank 1 Atm/Ambient Cyl CS 800 6'/4' Resin funnel 2 Atm/Ambient Con cyl CS 3 ft3 1.5'/3.5' CHAPTER 11 11.2-22 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-10 (Cont'd)

B. PUMPS DESIGN RATED FLOW RATED HEAD, RATED POWER PRESSURE/TEMP.

TDH QUANTITY TYPE (gpm) (ft) (hp) (psig/ F)

Equipment drain 1 Vert in-line 280 250 40 150/140 collection tank pump centrifugal Equipment drain 2 Vert in-line 280 180 25 150/140 tanks pumps centrifugal Equipment drain surge 1 Vert in-line 280 220 30 150/140 tank pump centrifugal Floor drain collection 1 Vert in-line 280 250 40 150/140 tank pump centrifugal Floor drain sample 1 Vert in-line 280 110 15 150/140 tank #1 pump centrifugal Floor drain sample 1 Vert in-line 280 180 25 150/140 tank #2 pump centrifugal Floor drain surge tank 1 Vert in-line 280 220 30 150/140 pump centrifugal Chemical waste tank 1 Vert in-line 200 70 7.5 150/140 pump centrifugal Evaporator feed tank 2 Vert in-line 20 25 1 150/140 pump(1) centrifugal Evaporator distillate 1 Vert in-line 50 130 7.5 150/140 sample tank pump(1) centrifugal Laundry drain tanks pumps 2 Vert in-line 25 105 5 150/140 centrifugal Laundry drain sample 2 Vert in-line 10 65 2 150/140 tank pump centrifugal Equipment drain filter 1 Horiz centrif 27 60 3 245/155 holding pump Floor drain filter 1 Horiz centrif 27 60 3 245/155 holding pump Precoat pump 1 Horiz centrif 688 75 20 150/155 CHAPTER 11 11.2-23 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-10 (Cont'd)

C. PROCESSING EQUIPMENT RATED FLOW, DESIGN TYPE DIAMETER MATERIAL EACH EQUIPMENT PRESSURE/TEMP.

QUANTITY HEIGHT TYPE/NUMBER (gpm) PARAMETER (psig/F)

Equipment and floor 2 Precoat type SS wire mesh 280 Filter area: 150/235 drain filters 3'/7' element/90 275 ft2 Equipment and floor 2 Mixed bed Effective resin 280 Resin bed depth: 150/235 drain demineralizers 6'/6' volume of each 3' min Laundry drain filter 1 Shell: Vert Shell: SS 25 Filter area: 75/250 cyl 8.6"/ 48 ft2 46" Cartridge: Cartridge:

"Epocel-30" Epoxy impregnated 6"/32" cellulose of 49 microns CHAPTER 11 11.2-24 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-11 EXPECTED YEARLY ACTIVITY RELEASED FROM LIQUID WASTE MANAGEMENT SYSTEMS (1)

(curies/year; totals are for 2 units)

EQUIPMENT CHEMICAL LAUNDRY FLOOR DRAIN DRAIN DRAIN LWS ADJUSTED DRAIN ISOTOPE SUBSYSTEM SUBSYSTEM SUBSYSTEM SUBTOTAL TOTAL(2) SUBSYSTEM TOTAL Br-83 2.52E-05 3.35E-05 1.13E-07 5.88E-05 6.70E-04 none 6.70E-04 Br-84 none 1.21E-06 none 1.21E-06 1.38E-05 none 1.38E-05 Br-85 none none none none none none none I-131 9.59E-04 4.67E-04 1.35E-05 1.44E-03 1.64E-02 1.22E-03 1.76E-02 I-132 2.27E-04 3.02E-04 1.03E-06 5.30E-04 6.05E-02 none 6.05E-03 I-133 2.02E-03 1.38E-03 1.40E-05 3.41E-03 3.89E-02 none 3.89E-02 I-134 7.09E-05 8.74E-05 2.47E-07 1.59E-04 1.81E-03 none 1.81E-03 I-135 7.19E-04 7.55E-04 3.63E-06 1.48E-03 1.68E-02 none 1.68E-02 Rb-89 8.06E-06 1.34E-06 none 9.39E-06 1.07E-04 none 1.07E-04 Cs-134 1.56E-04 1.46E-04 2.51E-05 3.27E-04 3.73E-03 2.65E-02 3.02E-02 Cs-136 9.89E-04 9.50E-05 1.47E-05 1.10E-03 1.25E-02 none 1.25E-02 Cs-137 3.64E-03 3.40E-04 5.88E-05 4.04E-03 4.61E-02 4.90E-02 9.50E-02 Cs-138 5.68E-04 1.28E-04 1.57E-06 6.97E-04 7.95E-03 none 7.95E-03 Na-24 7.55E-04 5.70E-04 4.58E-06 1.33E-03 1.52E-02 none 1.52E-02 P-32 3.97E-05 1.91E-05 5.92E-07 5.93E-05 6.77E-04 none 6.77E-04 Cr-51 1.02E-03 4.81E-04 1.57E-05 1.51E-03 1.73E-02 none 1.73E-02 Mn-54 1.24E-05 5.80E-06 1.59E-07 1.84E-05 2.10E-04 2.04E-03 2.25E-03 Mn-56 4.91E-04 6.52E-04 2.23E-06 1.14E-03 1.31E-02 none 1.31E-02 Fe-55 2.08E-04 9.05E-04 3.14E-06 1.12E-03 1.27E-02 none 1.27E-02 Fe-59 6.15E-06 2.90E-06 9.69E-08 9.14E-06 1.04E-04 none 1.04E-04 Co-58 4.12E-05 1.93E-05 6.53E-08 6.06E-05 6.09E-04 8.16E-03 8.85E-03 Co-60 8.32E-05 3.89E-05 1.34E-06 1.23E-04 1.41E-03 1.84E-02 1.98E-02 Ni-63 2.08E-07 9.72E-08 none 3.05E-07 3.48E-06 none 3.48E-06 Ni-65 2.92E-06 3.88E-06 1.33E-08 6.81E-06 7.76E-05 none 7.76E-05 Cu-64 2.20E-03 1.85E-03 1.49E-05 4.06E-03 4.63E-02 none 4.63E-02 Zn-65 4.13E-05 1.93E-05 6.65E-07 6.13E-05 6.98E-04 none 6.98E-04 Zn-69 1.30E-05 4.04E-06 none 1.70E-05 1.94E-04 none 1.94E-04 Sr-89 2.07E-05 9.72E-06 3.26E-07 3.08E-05 3.15E-04 none 3.51E-04 Sr-90 1.24E-06 5.80E-07 2.01E-08 1.84E-06 2.10E-05 none 2.10E-05 Sr-91 2.27E-04 2.04E-04 1.21E-06 4.33E-04 4.93E-03 none 4.93E-03 Sr-92 1.07E-04 1.41E-04 4.89E-07 2.48E-04 2.83E-03 none 2.83E-03 Y-91 1.19E-05 4.87E-06 2.08E-07 1.70E-05 1.94E-04 none 1.94E-04 Y-92 2.61E-04 3.15E-04 1.30E-06 5.78E-04 6.59E-03 none 6.59E-03 Y-93 2.42E-04 2.12E-04 1.32E-06 4.55E-04 5.19E-03 none 5.19E-03 Zr-95 1.44E-06 6.78E-07 2.28E-08 2.16E-06 2.46E-05 2.86E-04 2.88E-03 Zr-97 4.60E-07 3.33E-07 none 7.93E-07 9.04E-06 none 9.04E-06 Nb-95 1.46E-06 6.78E-07 2.35E-08 2.16E-06 2.46E-05 4.08E-03 4.10E-03 CHAPTER 11 11.2-25 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-11 (Cont'd)

EQUIPMENT CHEMICAL LAUNDRY FLOOR DRAIN DRAIN DRAIN LWS ADJUSTED DRAIN ISOTOPE SUBSYSTEM SUBSYSTEM SUBSYSTEM SUBTOTAL TOTAL(2) SUBSYSTEM TOTAL Nb-98 5.70E-06 4.87E-06 1.38E-08 1.06E-05 1.21E-04 none 1.21E-04 Mo-99 3.32E-04 1.75E-04 3.68E-06 5.11E-04 5.82E-03 none 5.82E-03 Tc-99m 9.15E-04 8.20E-04 6.25E-06 1.74E-03 1.99E-02 none 1.99E-02 Tc-101 3.57E-07 2.43E-07 none 6.00E-07 6.84E-06 none 6.84E-06 Tc-104 1.34E-06 1.20E-06 none 2.54E-06 2.90E-05 none 2.90E-05 Ru-103 4.09E-06 1.92E-06 6.40E-08 6.07E-06 6.92E-05 2.86E-04 3.55E-04 Ru-105 4.43E-05 5.34E-05 2.13E-07 9.79E-05 1.12E-03 none 1.12E-03 Ru-106 6.23E-07 2.91E-07 none 9.14E-07 1.04E-05 4.90E-03 4.91E-03 Ag-110m 2.08E-07 9.72E-08 none 3.05E-07 3.48E-06 8.98E-04 9.01E-04 Te-129m 8.18E-06 3.86E-06 1.28E-07 1.22E-05 1.39E-04 none 1.39E-04 Te-131m 1.28E-05 7.81E-06 1.04E-07 2.07E-05 2.36E-04 none 2.36E-04 Te-132 1.71E-06 8.90E-07 2.00E-08 2.62E-06 2.99E-05 none 2.99E-05 Ba-139 3.13E-05 4.10E-05 1.28E-07 7.24E-05 8.26E-04 none 8.26E-04 Ba-140 7.92E-05 3.80E-05 1.16E-06 1.18E-04 1.35E-03 none 1.35E-03 Ba-141 1.67E-07 1.50E-07 none 3.17E-07 3.62E-06 none 3.62E-06 Ba-142 3.19E-09 2.12E-09 none 5.31E-09 6.06E-08 none 6.06E-08 La-142 2.24E-05 2.96E-05 9.47E-08 5.21E-05 5.94E-04 none 5.94E-04 Ce-141 6.62E-06 3.07E-06 1.04E-07 9.79E-06 1.12E-04 none 1.12E-04 Ce-143 4.00E-06 2.39E-06 3.40E-08 6.42E-06 7.32E-05 none 7.32E-05 Ce-144 6.23E-07 2.91E-07 none 9.14E-07 1.04E-05 none 1.04E-05 Pr-143 8.13E-06 3.85E-06 1.23E-07 1.21E-05 1.38E-04 none 1.38E-04 Nd-147 5.90E-07 2.84E-07 none 8.73E-07 9.96E-06 none 9.96E-06 W-187 3.42E-05 2.21E-05 2.52E-07 5.66E-05 6.45E-04 none 6.45E-04 Np-239 1.11E-03 6.03E-04 1.17E-05 1.73E-03 1.97E-02 none 1.97E-02 OTHERS(3) 2.28E-04 1.80E-04 2.16E-06 4.10E-04 4.68E-03 1.02E-02 1.49E-02 TOTAL 1.80E-02 1.12E-02 2.07E-04 2.94E-02 3.35E-01 1.28E-01 4.64E-01 H-3 1.12E+01 (1)

Estimated releases are based on NUREG-0016, Revision 0, GALE Code evaluation.

(2)

Increased the calculated LWS release by 0.15 Ci/yr per reactor using the same isotopic distribution as the calculated LWS releases to account for anticipated operational occurrences that result in unplanned releases.

(3)

Activity of daughter products resulting from radioactive decay of the influent isotopes during the accumulation period.

CHAPTER 11 11.2-26 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.2-12 COMPARISON OF MAXIMUM CALCULATED RADIONUCLIDE CONCENTRATIONS IN THE ENVIRONMENT FROM ROUTINE LIQUID RELEASES TO 10CFR20 LIMITS(1) 10CFR20 AVERAGE ANNUAL TABLE II CONCENTRATION COLUMN 2  % OF IN WATER LIMIT(2) 10CFR20 ISOTOPE (Ci/ml) (Ci/ml) LIMIT Br-83 1.2x10-12 3x10-6 4.1x10-5 Br-84 1.3x10-14 - -

Br-85 - - -

I-131 3.4x10-11 3x10-7 1.1x10-2 I-132 1.1x10-11 8x10-6 1.4x10-4 I-133 2.2x10-11 1x10-6 2.2x10-3 I-134 3.4x10-12 2x10-5 1.7x10-5 I-135 3.2x10-11 4x10-6 7.9x10-4 R-89 2.0x10-13 - -

Cs-134 2.8x10-11 9x10-6 3.2x10-4 Cs-136 2.3x10-11 9x10-5 2.6x10-5 Cs-137 1.8x10-10 2x10-5 9.2x10-4 Cs-138 1.5x10-11 - -

Na-24 2.9x10-11 3x10-5 9.5x10-5 P-32 1.2x10-12 2x10-5 6.1x10-6 Cr-51 3.3x10-11 - -

Mn-54 4.5x10-12 1x10-4 4.5x10-6 Mn-56 2.4x10-11 1x10-4 2.4x10-5 Fe-55 2.3x10-11 8x10-4 2.9x10-6 Fe-59 1.9x10-13 6x10-5 3.2x10-7 Co-58 1.7x10-11 9x10-5 1.9x10-5 Co-60 4.0x10-11 3x10-5 1.3x10-4 Ni-63 6.5x10-15 3x10-5 2.2x10-8 Ni-65 1.4x10-13 1x10-4 1.4x10-7 Cu-64 8.8x10-11 2x10-4 4.4x10-5 Zn-65 1.3x10-12 1x10-4 1.3x10-6 Zn-69 3.7x10-13 2x10-3 1.8x10-8 Sr-89 6.5x10-13 3x10-6 2.2x10-5 Sr-90 4.0x10-14 3x10-7 1.3x10-5 Sr-91 9.3x10-12 7x10-5 1.3x10-5 Sr-92 5.3x10-12 6x10-5 8.8x10-6 Y-91 3.7x10-13 3x10-5 1.2x10-6 Y-92 1.2x10-11 6x10-5 2.0x10-5 Y-93 9.8x10-12 3x10-5 3.3x10-5 Zr-95 5.8x10-12 6x10-5 9.7x10-6 Zr-97 1.7x10-14 2x10-5 8.7x10-8 CHAPTER 11 11.2-27 REV. 14, SEPTEMBER 2008

LGS UFSAR Table 11.2-12 (Cont'd) 10CFR20 AVERAGE ANNUAL TABLE II CONCENTRATION COLUMN 2  % OF IN WATER LIMIT(2) 10CFR20 ISOTOPE (Ci/ml) (Ci/ml) LIMIT Nb-95 8.2x10-12 1x10-4 8.2x10-6 Nb-98 2.2x10-13 - -

Mo-99 1.1x10-11 4x10-5 2.8x10-5 Tc-99m 3.8x10-11 3x10-3 1.3x10-6 Tc-101 1.2x10-14 - -

Tc-104 6.0x10-14 - -

Ru-103 7.0x10-13 8x10-5 8.8x10-7 Ru-105 1.1x10-12 1x10-4 1.1x10-6 Ru-106 9.8x10-12 1x10-5 9.8x10-5 Ag-110m 1.8x10-12 3x10-5 6.1x10-6 Te-129m 2.6x10-13 2x10-5 1.3x10-6 Te-131m 4.5x10-13 4x10-5 1.1x10-6 Te-132 5.6x10-14 2x10-5 2.8x10-7 Ba-139 1.5x10-12 - -

Ba-140 2.6x10-12 2x10-5 1.3x10-5 Ba-141 6.8x10-15 - -

Ba-142 1.1x10-16 - -

La-142 1.1x10-12 - -

Ce-142 2.0x10-13 9x10-5 2.3x10-7 Ce-143 1.3x10-13 4x10-5 3.3x10-7 Ce-144 2.0x10-14 1x10-5 2.0x10-7 Pr-143 2.6x10-13 5x10-5 5.1x10-7 Nd-147 1.8x10-14 6x10-5 3.1x10-8 W-187 1.2x10-12 6x10-5 2.0x10-6 Np-239 3.7x10-11 1x10-4 3.7x10-5 H-3 2.2x10-8 3x10-3 7.5x10-4 (1)

Location: Liquid Discharge Outfall Area (2)

The concentration limits shown are reflective of the 10CFR20 Appendix B values prior to the 1994 revision of 10CFR20, and are consistent with those used in the original licensing of the plant. Current effluent release limits are based on the post-1994 10CFR20 regulation and are controlled by the Radioactive Effluent Controls Program defined by the Technical Specifications. New 10CFR20 limits went into effect on Jan. 5, 1994. As part of the power rerate evaluation, a comparison of the uprated liquid release concentration with the new 10CFR20 limits was made. All the uprated release concentrations were found to be less than 1% of their new limit.

CHAPTER 11 11.2-28 REV. 14, SEPTEMBER 2008

LGS UFSAR Table 11.2-13 COMPARISON OF MAXIMUM INDIVIDUAL DOSES RESULTING FROM LGS UNITS 1 AND 2 WITH 10CFR50, APPENDIX I DESIGN OBJECTIVES 10CFR50,  % OF LGS LIQUID APPENDIX I DESIGN DESIGN DOSE UNITS 1 & 2 OBJECTIVES/2 OBJECTIVE UNITS Total body 1.04 mRem/yr 6 mRem/yr 17.34%

Maximum organ 1.82 mRem/yr 20 mRem/yr 9.08%

(bone)

CHAPTER 11 11.2-29 REV. 13, SEPTEMBER 2006

LGS UFSAR 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS The gaseous waste management systems include all systems that process potential sources of airborne releases of radioactive materials during normal operation and anticipated operational occurrences. Included are the offgas system and various ventilation systems. These reduce radioactive gaseous releases from the plant by filtration or delay, which allows decay of radioisotopes prior to release.

The function of the offgas system is to collect and delay release of noncondensable radioactive gases removed from the main condenser by the air ejectors during normal plant operation. Plant ventilation systems process airborne radioactive releases from other plant sources, such as equipment leakage, maintenance activities, the mechanical vacuum pump, and the steam seal system.

The offgas system is described in detail in this section. The ventilation systems are discussed briefly in this section and in greater detail in Section 9.4.

11.3.1 DESIGN BASES

a. The gaseous waste management systems are designed to control and monitor the release of radioactive materials in gaseous effluents in accordance with GDC 60 and GDC 64.

b. The offgas system design basis and maximum expected source terms correspond respectively to 100,000 Ci/sec and 60,000 Ci/sec of radioactive noble gases after 30 minute delay.

c. The gaseous waste systems are designed to limit offsite doses from routine station releases to significantly less than the limits specified in 10CFR20 and to operate within the dose objectives established in 10CFR50, Appendix I.

d. The gaseous waste management systems are designed with sufficient capacity and redundancy to accommodate all anticipated processing requirements of the plant during normal operation, including anticipated operational occurrences.

e. Continuous monitoring is provided for all potential pathways of airborne radioactive releases, with annunciation at levels lower than normal release limits.

f. Design provisions are incorporated which preclude the uncontrolled release of radioactivity to the environment as a result of any single operator error or any single active component failure.

g. The gaseous waste management system is designed to keep the exposure to plant personnel as low as reasonably achievable (ALARA) during normal operation and plant maintenance, in accordance with Regulatory Guide 8.8.

h. Design features are provided in the offgas system to control leakage and to facilitate operation and maintenance in accordance with the guidelines of Regulatory Guide 1.143.

CHAPTER 11 11.3-1 REV. 18, SEPTEMBER 2016

LGS UFSAR

i. The offgas system is designed in accordance with the guidelines of Regulatory Guide 1.143, with exception as discussed in Section 3.2.

j. Filtration units in the ventilation systems are designed, operated, and maintained in accordance with the guidelines of Regulatory Guide 1.140. with exceptions as discussed in Section 9.4.

k. The offgas system is designed to maintain the concentration of hydrogen in the gases exhausted from the main condenser below flammable limits.

l. Instrumentation is provided in the offgas system to detect abnormal concentrations of hydrogen and other system malfunctions.

m. The offgas system is designed to withstand the effects of a hydrogen explosion without breach of the pressure boundary.

11.3.2 SYSTEM DESCRIPTION 11.3.2.1 Offgas System 11.3.2.1.1 General Noncondensable gases are continuously removed from the main condenser by the SJAE during plant operation. The offgas will consist of activation gases, fission product gases, radiolytic hydrogen and oxygen, and condenser air inleakage.

The offgas system is designed to reduce offgas radioactivity to permissible levels for release under all site atmospheric conditions.

The offgas system uses catalytic recombination for volume reduction and control of hydrogen concentration. Selective adsorption of fission product gases on charcoal is used to provide time for delay before release.

The location of the offgas system components is shown in general arrangement drawings in Section 1.2.

The seismic categories, quality group classifications, and corresponding codes and standards that apply to the design of the gaseous waste management system are discussed in Section 3.2.

11.3.2.1.2 Process Flow Description Figure 11.3-1 is the process flow diagram for the offgas system and contains process design data for startup and normal operating conditions.

Drawings M-69 and M-70 are the piping and instrumentation diagrams for the offgas system.

During startup the mechanical vacuum pump is used to draw a vacuum in the main condenser as described in Section 10.4.2. Once condenser vacuum has been established by the mechanical vacuum pump and reactor steam is available, one of the two-stage SJAE trains are placed in CHAPTER 11 11.3-2 REV. 18, SEPTEMBER 2016

LGS UFSAR service. As an alternative, during startup, shutdown and power operation, auxiliary steam can be used for SJAE operation when available.

The first-stage SJAEs continuously remove noncondensable gases and some steam from the condenser and discharge them to the SJAE condenser where the steam is condensed and returned to the main condenser. The gases are removed from the SJAE condenser by the second-stage ejector and discharged to the gaseous radwaste recombination system together with the second-stage ejector motive steam. This steam provides sufficient dilution to maintain hydrogen concentrations below combustible concentrations. A complete description of the SJAE is given in Section 10.4.2.

The offgas stream from the second-stage ejector is treated first in the recombiner portion of the offgas system. The purpose of the recombiner is to reduce the offgas volume and reduce the hydrogen concentration to less than 4% concentration by volume on a dry basis. The offgas first passes through the preheater in order to vaporize any water droplets and to achieve a sufficient temperature to start the catalytic reaction. Steam used for preheating is provided from the reactor feed pump-turbine steam supply line or from auxiliary steam when available. The recombination process takes place inside a recombiner vessel. The temperature of the gases leaving the recombiner rises as a function of the influent hydrogen concentration. This temperature rise is due to the heat of reaction of the recombination process. The reaction temperature rises approximately 120F for each 1% of H2 recombined. The catalyst is a metal mat-type. The elements are made of crimped high nickel alloy ribbons coated with platinum/palladium. Each recombiner has two independent electric heater assemblies.

Steam flow leaving the recombiner vessel is condensed in the aftercondenser and the offgas stream is cooled. The remaining noncondensable gas (principally air with traces of krypton and xenon) is delayed in a 26 inch diameter, 125 foot long holdup pipe. At a flow rate of 75 scfm, this pipe provides approximately 6.3 minutes of delay for the offgas prior to entering the charcoal adsorption train. Piping from the delay pipe to the charcoal treatment system is heat traced to reduce the possibility of moisture collecting in the line.

Further offgas cooling and condensing of water vapor takes place in the shell side of the cooler condenser heat exchanger. A chilled glycol solution is circulated on the tube side of this heat exchanger. The offgas is cooled to a dew point below 45F before leaving the cooler condenser.

Water that has condensed is drained from the cooler condenser into the condensate accumulator.

The temperature of the gas stream increases to ambient (approximately 65F) as it flows from the cooler condenser to the guard bed, thus preventing condensation.

The glycol solution circulated on the tube side of the cooler condenser heat exchanger is cooled in a closed-loop pumped refrigeration circuit. Dual brine pumps and redundant refrigeration units are provided in the glycol system.

Before entering the main charcoal vessels, the offgas stream passes through a guard bed. The function of the guard bed is to protect the main charcoal adsorbers from moisture if a malfunction of moisture removal features occurs as well as to adsorb impurities in the process gas that might adversely affect performance of the main charcoal vessels.

After passing through the guard bed, the gas enters the main charcoal adsorption beds. The charcoal adsorption beds, maintained at about 65F by redundant room air conditioning units, selectively adsorb and delay the xenon and krypton from the bulk carrier gas. This delay permits CHAPTER 11 11.3-3 REV. 18, SEPTEMBER 2016

LGS UFSAR the xenon and krypton to decay in place. The offgas stream then passes through a HEPA afterfilter where radioactive particulate matter and any charcoal fines are retained.

During offgas system operation, the offgas stream must pass through the charcoal adsorbers.

Bypass piping around the entire adsorber train does not exist. The offgas stream is directed to the turbine enclosure vent stack where it is diluted with a minimum of 183,000 scfm of air before being released from the north stack.

Table 11.3-1 indicates the estimated annual release rate from the offgas system.

All moisture removed from the process stream is returned to the main condenser hot well or clean radwaste.

11.3.2.1.3 System Design Considerations 11.3.2.1.3.1 Charcoal Holdup Time The krypton and xenon holdup times are closely approximated by the following equation:

T = 0.26 KM (EQ. 11.3-1)

V where:

T= hold-up time, hours K= dynamic adsorption coefficient, cm3/g M= mass of charcoal adsorber, thousands of pounds V= gas flow rate, scfm Dynamic adsorption coefficients for krypton and xenon used to determine gaseous effluent releases are discussed in NUREG-0016 (Reference 11.3-1). The charcoal adsorber beds are designed for a delay time of 35 days for xenon under both of the following conditions:

a. 75 scfm flow rate using manufacturer's guaranteed adsorption coefficients (733 cm3/g for xenon and 31.8 cm3/g krypton) b.. BWR GALE code assumptions (NUREG-0016, Revision 0)

The offgas system is capable of handling changes in noncondensable flow rates up to 215 scfm (Unit 1) and 205 scfm (Unit 2) with both units running, or 225 scfm (Unit 1) and 215 scfm (Unit 2) with only one unit running, without operator attention. Even though the offgas system is capable of handling these limits, the air in-leakage should not exceed 150 scfm since it is the maximum design capability of the steam jet air ejectors.

With a condenser air inleakage rate of 30 scfm, the charcoal treatment system provides a design holdup time of 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> for krypton and 38.6 days for xenon based upon NUREG-0016 assumptions. Since it is expected that the condenser air inleakage will be below the design value CHAPTER 11 11.3-4 REV. 18, SEPTEMBER 2016

LGS UFSAR and that the charcoal adsorption coefficients will be higher than the values in NUREG-0016 (References 11.3-3 and 11.3-4), the actual charcoal holdup time should be considerably longer than the design holdup time. Additionally, experience with newer fuel designs (8x8 assemblies) indicates that substantially lower source terms than those used for system design may be expected.

11.3.2.1.3.2 Detonation Resistance All portions of the LGS offgas system are designed to withstand the effects of a hydrogen detonation or are provided with protective features to preclude the existence of a detonable mixture of gases.

Design for extremely short duration (microsecond) loadings which accompany hydrogen detonations is outside the scope of normal industry design codes (i.e., ANSI B31.1, ASME B&PV Code, etc). The industry has developed a methodology for detonation-resistant design of offgas system piping, pressure vessels, and other components based on extensive theoretical and experimental work.

The magnitude of pressure pulses accompanying a hydrogen detonation have been shown to be a function of component geometry (i.e., length/diameter ratio), initial system pressure, and proximity to reflection points (i.e., pipe elbows, etc). The basic methodology used in the design of detonation-resistant BWR offgas systems is described in appendix C of ANSI/ANS 55.4 (1979).

That methodology, with slight variations between the AEs and industry equipment suppliers, has been followed since the early 1970s and was used in the LGS design. The analytical methods used are best described as "static analyses using dynamic material properties". An appropriate wall thickness is determined using peak dynamic pressure and dynamic material properties. For convenience, this wall thickness is often expressed in terms of its code static pressure equivalent.

Application of this methodology provides a conservative design without the need for detailed and laborious analysis of the gas dynamics of the system. The methodology has been demonstrated to be adequately conservative by theoretical analysis and operating experience:

a. A true dynamic analysis of system pressures will typically require a wall thickness one-half that indicated by this approach.

b. No BWR offgas system pressure boundary failures had been observed despite the occurrence of considerably more than 100 system detonations.

An ASME Code Committee (Committee on Air and Gas Treatment, Gas Processing Subcommittee) is currently working towards the codification of the above described industry methodology. The guidance provided in SRP 11.3 (i.e., approximately 20 times operating pressure) has been demonstrated to be nonconservative in many applications. (The contents of this paragraph are historical in nature and do not reflect present day activities.)

All offgas system detonation-resistant piping and components upstream of the charcoal treatment system have been analyzed using the method of analysis employed by Bechtel. All charcoal treatment system components have been analyzed using the method of analysis employed by the system supplier (Reference 11.3-5). Both analytical methods closely parallel that described in ANSI/ANS 55.4 and give approximately equal results.

CHAPTER 11 11.3-5 REV. 18, SEPTEMBER 2016

LGS UFSAR The following is a discussion of factors relevant to the detonation resistance of the LGS offgas treatment system:

a. The SJAEs are designed to withstand a hydrogen detonation occurring at normal system operating pressure (3.5 psia). Because leakage into the standby SJAE train could cause a detonation at higher initial pressures, provisions have been made to maintain the standby SJAE train at main condenser vacuum.

b. The SJAEs that are used do not employ a second-stage condenser. Thus, the driving steam from the SJAE second-stage provides dilution steam such that a detonable mixture of gases will not exist between the SJAE discharge and the offgas aftercondenser. Protective circuits are provided such the offgas system is automatically shut down when loss of dilution steam is detected (low system flow or high recombiner outlet temperature).

c. All system valves use spark-resistant trim.

d. All system piping, valves, vessels, instruments, and other components are designed to withstand the effects of a hydrogen detonation except portions of the piping between the SJAE discharge and the preheater. Detonable mixtures of gases in this piping are precluded as discussed in Item (b) above.

11.3.2.1.4 Component Description The recombiner and associated equipment are located in the lowest level of the control structure.

Each recombiner system consists of a preheater, recombiner vessel, and aftercondenser. The materials of construction, design temperatures, and pressures are listed in Table 11.3-3.

11.3.2.1.4.1 Preheater The preheater is a U-tube parallel heat exchanger. Main steam is used to heat process gas before entering the recombiner. The process gas enters at approximately 280F and is heated as required to vaporize water droplets and achieve sufficient temperature for catalytic reaction.

Auxiliary steam may also be available for heating the process gas flow, should main steam be unavailable. Condensate from the tube side of the heat exchanger is collected in a drain pot underneath the preheater and is routed back to the condenser.

11.3.2.1.4.2 Recombiner The hydrogen and oxygen in the gas stream are recombined in the recombiner vessel by a catalyst of platinum/palladium. Electric heaters with automatic temperature control are provided on the shell of each recombiner. The heaters are used for preheating the recombiner during startup and maintaining it in a dry condition during shutdowns.

11.3.2.1.4.3 Aftercondenser The aftercondenser is a straight tube heat exchanger. Service water is circulated through the aftercondenser tubes to condense the steam in the offgas flow. Noncondensable gases are collected in the aftercondenser, cooled in the air cooler section and vented to the holdup pipe.

Control valves are used to automatically maintain proper condensate level in the aftercondenser CHAPTER 11 11.3-6 REV. 18, SEPTEMBER 2016

LGS UFSAR shell. Condensate from the air cooler section is collected in a drain tank. Condensate from both sources is returned to the main condenser hotwell, or diverted to clean radwaste if main condenser vacuum is not available or the conductivity instrumentation indicates that a tube leak exists.

11.3.2.1.4.4 Holdup Pipe The holdup pipe is approximately 125 feet in length and 26 inches in diameter. The purpose of this holdup pipe is to provide delay time to allow N-16 decay before entering the charcoal adsorption portion of the offgas system. Baffles are provided in each holdup pipe to assure adequate mixing and delay.

11.3.2.1.4.5 Charcoal Adsorption System 11.3.2.1.4.5.1 Cooler Condenser The cooler condenser is a straight tube heat exchanger with a glycol/water solution flowing through the tube side. The glycol system is provided with redundant circulation pumps and redundant refrigeration units that are cooled by service water.

Moisture formed in the cooling process is collected in an accumulator. This condensate is normally routed to clean radwaste but is automatically diverted to chemical waste if the measured conductivity indicates that a glycol leak could exist.

11.3.2.1.4.5.2 Guard Bed The function of the guard bed is to protect the main charcoal adsorbers from moisture in the event of a malfunction of the moisture removal equipment and to remove contaminants that may be in the process stream which could be detrimental to the main adsorber beds. The guard bed is designed with a removable cover and an internal basket which contains the charcoal adsorbent.

The basket allows easy charcoal replacement if necessary. The guard bed contains approximately 190 pounds of charcoal.

11.3.2.1.4.5.3 Main Charcoal Adsorber Bed The Unit 1 adsorber train consists of two 11 foot diameter tanks and five 10 foot diameter tanks.

The Unit 2 adsorber train consists of one 11 foot diameter tank and eight 10 foot diameter tanks.

The different configurations for Units 1 and 2 are the result of space limitations within the enclosure, which was constructed before final system design.

Each adsorber train contains approximately 321,000 lbs of charcoal. The tanks are connected at the top and bottom by 4 inch piping. Portions of each unit's offgas system use a parallel flow path through the main adsorber beds. Appropriate valves are provided to facilitate flow balancing as required .

The charcoal adsorber tanks are maintained at a temperature of 65F or below by redundant air conditioning systems. The last adsorber tank in the Unit 1 train is not located in an air conditioned vault and will operate at room ambient temperature. In the unlikely event that both air conditioning units are unable to function, the radioactive emissions from the offgas system might increase slightly. However, since substantial margin exists in the offgas system design, the releases would CHAPTER 11 11.3-7 REV. 18, SEPTEMBER 2016

LGS UFSAR still be well below acceptable limits for expected air inleakage rates, radioactivity source terms, and adsorption coefficients.

11.3.2.1.4.5.4 Outlet HEPA Filter A HEPA filter is provided to collect any entrained particulates or charcoal fines prior to release.

These filters were individually tested during the startup test program using the DOP method. The filters are equipped with hinged covers to facilitate removal and replacement of the filter element.

11.3.2.1.4.5.5 Leakage of Radioactive Gases The offgas system operates at approximately 7 psig during startup and at approximately 2 psig during normal operation. The differential pressure between the system and atmosphere is small, thus limiting the potential for leakage of radioactive gases.

Leakage of radioactive gases from offgas containing lines in the offgas system is further limited by the use of welded construction, by plug valves with plug and shank seals, globe valves with back-seats and packless design, and by using double stem packed valves with a bleed-off connection that is pressurized by instrument air to slightly higher than the system pressure, wherever practical.

All drains in the offgas system with the exception of the condensate accumulator drain are directed either to the main condenser or clean radwaste during normal operation. The condensate accumulator drains are directed to clean radwaste or chemical waste, as discussed in Section 11.3.2.1.4.3.1. In order to eliminate the possibility of gas leakage to the waste collector tanks should a level controller fail, loop seals or drain pots are provided.

All loop seals are the self-resealing type.

11.3.2.1.4.6 Instrumentation and Control The offgas system is monitored at appropriate locations for flow, temperature, pressure, humidity, conductivity, radiation, and hydrogen concentration to verify specified operation and control as well as to ensure that the hydrogen concentration is maintained below the flammable limit.

Sufficient instrumentation is provided to permit system operation and monitoring from the main control room. Drawings M-69 and M-70 indicate the process instrumentation.

A radiation monitor located upstream of the holdup pipe continuously monitors gaseous radioactivity input to the charcoal adsorption system. This is representative of gaseous radioactivity released from the reactor and therefore indicates the condition of the fuel cladding.

Provision is made for grab sampling of the influent gases for the purpose of determining isotopic composition.

A radiation monitor is also provided at the outlet of the charcoal adsorbers to continuously monitor activity released from the system. The offgas system process radiation instrumentation is further discussed in Section 11.5.

Dilution steam flow (second-stage SJAE motive steam) is monitored upstream of the preheater and recorded in the control room. Low dilution steam flow automatically shuts the first-stage CHAPTER 11 11.3-8 REV. 18, SEPTEMBER 2016

LGS UFSAR SJAE gas suction valves to prevent high concentrations of hydrogen in the offgas piping between the second-stage SJAEs and the preheater.

The HWC control system requires an interface with the Gaseous Radwaste/Recombination System. An SJAE train/recombiner trip digital output is provided to trip the HWC system.

The steam supply to the gas preheater is controlled by the offgas outlet temperature.

The temperature of the recombiner vessel and the inlet and outlet gases are monitored by thermocouples. A high temperature alarm is provided for the recombiner vessel, and high and low temperature alarms are provided on the inlet and outlet lines to alarm a temperature outside of the design range.

Aftercondenser outlet temperature is recorded, and high temperature is alarmed in the main control room. This temperature indicates if adequate cooling flow exists in the aftercondenser.

Three thermal conductivity-type hydrogen analyzers and two oxygen analyzers are used to measure the hydrogen and oxygen content of the offgas process stream at the discharge of the aftercondenser. The system is designed to ensure that there is sufficient oxygen available to combine with the hydrogen and yet ensure that the oxygen level does not get too high so as to impose an increased fire hazard for the offgas charcoal beds. One of the hydrogen analyzers can be switched to measure the hydrogen content of the offgas stream at the inlet to the recombiner.

The sample gas is returned to the main condenser. The hydrogen analyzers are designed to withstand the effects of a hydrogen detonation. The analyzer cell is not capable of causing a detonation, and the analyzer is designed with a flame arrestor on the sample side to inhibit such detonation. The hydrogen concentration is annunciated both locally and in the control room for high and high-high hydrogen concentration (2% and 4%, respectively). A common trouble alarm for cell failure, analyzer cell low flow and low bypass flow, is also provided. Because the offgas system is designed to withstand the effects of hydrogen detonation, no automatic control functions are required. Each hydrogen analyzer is independently calibrated. Condensate from the analyzers is routed to the aftercondenser air cooler drain.

The condensate level in the preheater and aftercondenser is maintained at appropriate levels by level control systems. These level control systems provide drainage either to the main condenser or clean radwaste. During normal operation condensate is directed to the main condenser. A conductivity element is provided in the drain to detect aftercondenser tube leakage if high conductivity is detected. Because the condensate from the air cooler portion of the aftercondenser already has a high conductivity reading due to dissolved air, this portion of aftercondenser drainage is routed to bypass the conductivity cell to avoid a false high reading. As a result, tube leakage in the aftercondenser air cooler will not be detected. Tube leakage in the air cooler section is not anticipated due to the mild service environment that this section of tubes experiences.

Cooler condenser performance is monitored by outlet temperature and moisture. High process outlet temperature is alarmed in the main control room. A moisture element is located downstream of the cooler condenser. The moisture element senses the process offgas dew point temperature. This temperature is indicated both at the local panel and in the main control room.

High moisture in the process stream is alarmed both locally and in the main control room.

CHAPTER 11 11.3-9 REV. 18, SEPTEMBER 2016

LGS UFSAR Moisture condensed from the process offgas in the cooler condenser is collected in the condensate accumulator. The condensate level is maintained at appropriate limits by a level control system. Condensate is normally drained to clean radwaste. A conductivity element is provided in the drain to detect glycol inleakage. If glycol inleakage is detected, drainage is routed to the chemical waste tank.

The guard bed is provided with both differential pressure and temperature indication. High differential pressure and high temperature are alarmed in the control room.

The charcoal beds in the adsorber vessels are monitored by thermocouples. High temperature is alarmed in the control room and at the local panel. Individual adsorber charcoal temperature is indicated at the local control panel and recorded in the control room. Each vessel is provided with two spare thermocouples which can be used if the first thermocouple fails.

Differential pressure is measured across the HEPA filter. High differential pressure is alarmed in the main control room. Flow from the discharge of the second-stage SJAE can be recycled to the main condenser if transient flow conditions approach system maximum. Recycle line operation is described in Section 10.4.2.5 .

Process offgas flow rate is monitored downstream of the HEPA filter. High flow rate is alarmed in the main control room and at the local panel. Flow rate is indicated locally and recorded in the main control room.

Thermocouples are provided in each charcoal adsorber tank room. The thermocouples automatically operate the room air conditioning units to maintain room temperature at or below 65F. Room temperature is indicated and alarmed at the local panel. Room temperature is also alarmed and recorded in the main control room.

11.3.2.1.4.7 Offgas System Operating Procedure 11.3.2.1.4.7.1 Startup Before starting the offgas system, the following conditions are necessary: main steam or auxiliary steam is available for the SJAEs and the preheater; cooling water (condensate) is supplied to the SJAE intercondenser; the recombiner is preheated to approximately 300F by the electric heaters; service water is supplied to the aftercondenser, glycol refrigeration units, and compartment refrigeration units; the glycol coolant is chilled, one glycol pump is running and circulating chilled glycol through the cooler condenser; the adsorber rooms are at 65F or below, and air circulation fans are operating .

The system can be started either with the main condenser at 20 in Hg vac with vacuum having been drawn by the mechanical vacuum pump or at atmospheric pressure with the initial vacuum drawn by the 2nd stage SJAE using auxiliary steam. With main condenser vacuum established, the system is prepared for startup by admitting main or auxiliary heating steam to the preheater, and selecting and setting various instrument control stations at the main control room system panels. The SJAE train discharge and main steam or auxiliary steam supply valves are opened.

The 2nd stage SJAE is placed into service and a vacuum is drawn in the SJAE intercondenser.

The discharge flow heats the process piping down through the preheater, recombiner, aftercondenser, and holdup pipe before being recirculated to the main condenser. The 1st stage SJAE air suction valves are opened to allow the 2nd stage SJAE to draw condenser vacuum down CHAPTER 11 11.3-10 REV. 18, SEPTEMBER 2016

LGS UFSAR to approximately 25 to 26 in Hg vac. The mechanical vacuum pump can then be secured. Full condenser vacuum is achieved by performing the following: close the 1st stage air suction valves, open the 1st stage steam supply valves, re-open the 1st stage air suction valves. Once the 1st stage air suction valves are opened (with either the 1st stage SJAE in or out of service) the recirculation line downstream of the holdup pipe to the main condenser is closed and flow is directed to the charcoal treatment system. Recycle lines at the discharge of the 2nd stage SJAE will return some flow to the main condenser should system back pressure exceed approximately 7 psig at startup flows.

If the 2nd stage SJAE is used to draw the initial vacuum, the startup sequence is the same as above. The only difference is that auxiliary steam must be used.

11.3.2.1.4.7.2 Normal Operation and Shutdown After startup, the flow rate of noncondensables exhausted by the SJAE should stabilize, primarily as a function of reactor power level and condenser inleakage. The instrumentation discussed in Section 11.3.2.1.4.4 can be used to monitor system performance. Operator action is not required for steady-state system operation, including changes in offgas noncondensable flow rates up to 215 scfm (Unit 1) and 205 scfm (Unit 2) with both units running, or 225 scfm (Unit 1) and 215 scfm (Unit 2) with only one unit running. Even though the offgas system is capable of handling these limits, the air in-leakage should not exceed 150 scfm since it is the maximum design capability of the steam jet air ejectors.

Normal operation is terminated following a normal reactor shutdown or a scram by shutting the gas suction valves to the SJAEs and then stopping steam flow to the SJAEs. Steam is secured to the first-stage SJAEs and then to the second-stage. Normal practice is to run the offgas system throughout reactor pressure vessel depressurization. Main steam can support SJAE operation down to 205 psig reactor pressure. Auxiliary steam may be used at any time during shutdown operations.

11.3.2.1.4.7.3 Equipment Malfunction An equipment malfunction analysis, indicating the consequences and design precautions taken to accommodate failure of various components of the offgas system, is presented in Table 11.3-5.

11.3.2.1.4.8 Operator Error The potential for operator error resulting in high offsite releases while operating the offgas system is small. Offgas flow from the SJAEs must be directed through the recombiner and ambient charcoal systems. There are no system bypasses other than the mechanical vacuum pump. The systems are instrumented such that the operators can detect and correct failures of various components in the systems (Table 11.3-5).

11.3.2.1.4.9 Serviceability and Reliability Reliability of the offgas system is accomplished through redundancy and the use of passive components.

Two redundant SJAE systems, including intercondensers, are provided. Redundant level control valves, flow control valves, and instrumentation are furnished for the preheater. Redundant level CHAPTER 11 11.3-11 REV. 18, SEPTEMBER 2016

LGS UFSAR control valves and instrumentation are supplied for the aftercondenser, holdup pipe, and condensate accumulator drains. Redundant glycol cooling pumps, glycol refrigeration units, charcoal adsorber vessels' compartment refrigeration units, and compartment fan coil units are provided. Redundant control valves may be operated or isolated from the control room; redundant equipment can be placed in or out-of-service from the control room. The recombiner systems of the two units may also be cross connected into one charcoal train through a spool piece interconnection. Each charcoal train can process up to 300 scfm of offgas noncondensable flow.

The guard beds and first charcoal adsorber vessels in each train can be bypassed if they become contaminated with moisture. The HEPA filter may also be bypassed if a high differential pressure condition exists.

Redundant equipment for the preheater, recombiner, and aftercondenser is not installed. Piping connections are installed that allow for a second recombiner system to be connected to the existing system of each unit in the future.

Serviceability is accomplished wherever possible, by using passive components and by locating in accessible areas equipment and instruments which are not passive or which require periodic maintenance.

Instruments for the recombiner system that require periodic maintenance are located outside the recombiner rooms. Instruments for the ambient charcoal system are either located outside rooms containing radioactive equipment or are shielded from radioactive equipment. Instrumentation requiring calibration are accessible during system operation.

The guard beds and HEPA filters are located in individual rooms and are provided with system bypasses. The vessels are designed with removable covers. This arrangement allows replacement of the charcoal basket or filter element when the offgas system is operating.

11.3.2.2 Other Radioactive Gas Release Paths There are three general areas that contain sources of radioactive gas: the primary and secondary containment, the turbine enclosure, the radwaste enclosure and Chemical Laboratory Expansion .

The description of the ventilation systems for these enclosures is presented in Section 9.4. The enclosure volumes, flow rates, sources, and other information required to calculate the airborne concentrations of radioactive materials are contained in Sections 12.2.2, 12.3.3, and 12.4.

11.3.2.2.1 Primary and Secondary Containment As indicated in Section 9.4, the two reactor enclosures and the common refueling area have been designated as HVAC Zones I, II, and III. The primary containment and secondary containment HVAC systems that are important for the treatment of potentially radioactive air are the following:

Zone I and Zone II equipment compartment exhaust systems, the SGTS, and the RERS.

The equipment compartment exhaust systems process ventilation exhaust from the areas in Zones I and II which are most likely to have sources of airborne radioactivity. These areas are listed on drawing M-76. The exhaust system contains HEPA and 2 inch charcoal filters.

Discharge air from other areas in HVAC Zones I and II is released unfiltered through the reactor enclosure air exhaust system. Radiation monitors in the exhaust duct-work cause HVAC isolation CHAPTER 11 11.3-12 REV. 18, SEPTEMBER 2016

LGS UFSAR on high-high radiation and initiate the RERS and SGTS. These systems provide increased filtration and delay of airborne radioactivity before release. Equipment compartment exhaust systems are more completely described in Section 9.4.

Air from HVAC Zone III (the common refueling area) is normally exhausted unfiltered through the refueling area air exhaust fans. Radiation monitors are provided in the exhaust duct-work to isolate normal HVAC and initiate the SGTS on high-high radiation. There may be small quantities of radioactivity released unfiltered from the refueling area and the spent fuel pool, especially during the early stages of refueling. However, the quantities of iodine and particulates released from this unfiltered source are expected to be much less than the releases from equipment leakage, equipment maintenance, and drywell purge, all of which are filtered. Considering the uncertainties in the calculation of the reactor enclosure releases and the conservative use of a minimum 90% efficiency for the charcoal filtration systems, it is expected that the actual releases from the reactor enclosure to the atmosphere should be lower than the estimates used in this evaluation.

During power operation, radioactivity released from minor system leakage inside the primary containment is contained, except for minor releases necessary to control containment pressure.

Pressure is controlled by use of the low volume purge. In this purge mode, gas is supplied by the containment atmosphere control system through 1 inch lines and is exhausted from the primary containment through 2 inch lines connecting to the high volume purge exhaust lines. Low volume purge air is processed through the equipment compartment exhaust system prior to release to the environment. The containment isolation valves on the high volume and low volume purge lines are closed on a containment isolation signal.

Before a shutdown requiring containment entry, the containment is purged of nitrogen and airborne radioactivity by the drywell purge system, which uses the SGTS filters to reduce releases. When required, high volume purging of the containment continues while maintenance activities are conducted inside primary containment. The MSRVs are vented to the suppression pool. The activity released from the actuation of these relief valves is contained in the primary containment until its atmosphere is purged through the SGTS when preparing for personnel access.

11.3.2.2.2 Radwaste Enclosure and Chemistry Laboratory Expansion The supply system delivers filtered and tempered air that is distributed throughout the enclosure in quantities sufficient to maintain required temperatures. The equipment compartment exhaust system consists of two 100% capacity fans and two 100% capacity filter housings. The Chemistry Laboratory Expansion air exhaust system consists of two 100% capacity fans and two 100%

capacity filter housings. The fume hood exhaust air system consists of two 100% capacity fans and two 100% capacity filter housings. Each filter housing has a bank of prefilters and a bank of HEPA filters. This exhaust system is balanced to assure that the flow of air within the enclosure is into areas with higher potential for airborne radioactivity contamination. The tank exhaust system provides a means of filtering and venting air from tanks and equipment housed in the radwaste enclosure. A single fan and filter train are employed for this purpose. There are HEPA filters in this system. Since the flow of air from tanks and equipment varies, space air is admitted as required to maintain system volume. Noncontaminated areas of the radwaste enclosure are exhausted by the unfiltered radwaste enclosure air exhaust system.

CHAPTER 11 11.3-13 REV. 18, SEPTEMBER 2016

LGS UFSAR A radiation monitor is provided on the combined radwaste enclosure air and equipment compartment exhaust systems. All radwaste enclosure and Chemistry Laboratory Expansion exhaust systems discharge to the north stack.

Each of the above exhaust systems and the respective supply system are interlocked so that failure of the exhaust system shuts down the supply system. This condition is alarmed in the radwaste control room.

11.3.2.2.3 Turbine Enclosure As indicated in Section 9.4.4, the turbine enclosure ventilation system contains a filtration system with HEPA filters and an 8 inch deep charcoal filter. Enclosure air from those areas of the turbine enclosure, where equipment leakage and airborne activity are most likely, is processed through the filtration system before it is released through the north stack to the atmosphere. Air from noncontaminated areas is released through the north stack without filtration.

In the past, discharge from the steam packing exhausters has presented a source of gaseous radioactive releases in some BWR plants. At LGS however, clean steam from the steam seal evaporator is provided for gland seal purposes and, therefore, essentially no activity is released from this system. Section 10.4.3 provides a detailed description of the gland seal steam system.

During the startup of each plant, air is removed from the main condenser by a mechanical vacuum pump. This vacuum pump discharges through the turbine enclosure equipment compartment exhaust air filter assemblies. The vacuum pump exhaust can bypass the turbine enclosure equipment compartment exhaust filters under administrative control when it is determined that filtration is not required to limit offsite doses. A radiation detector continuously monitors the effluent from the turbine enclosure exhaust system via the north stack, and an alarm is actuated upon the detection of a high radiation level.

11.3.3 RADIOACTIVE RELEASES AND ESTIMATED DOSES The activity released from the various vents is monitored to ensure that the airborne concentrations at offsite locations will be below the limits of 10CFR20. In addition, the yearly releases are kept ALARA in order to meet the dose guidelines 10CFR50, Appendix I. The expected annual activity releases are given in Table 11.3-1.

An evaluation of the gaseous radioactive releases to show compliance with the above guidelines has been performed. The assumptions used in this evaluation are summarized in Table 11.3-2.

The fractions of radioiodine and particulates assumed to be released from the reactor enclosure, the turbine enclosure, and the radwaste enclosure are included. These assumptions were based on NUREG-0016, Revision 0. A further discussion of each assumption follows:

a. Reactor Enclosure As indicated in Sections 11.3.2.2.1 and 9.4.2, the reactor enclosure is ventilated during normal operation by two systems: the reactor enclosure exhaust system and the reactor enclosure equipment compartment exhaust system. Only the latter system includes HEPA and charcoal filter systems. The design philosophy is to identify those areas with the potential for airborne contamination and to exhaust them through the filtered exhaust. Clean areas are exhausted without filtration.

CHAPTER 11 11.3-14 REV. 18, SEPTEMBER 2016

LGS UFSAR This minimizes the mixing of air between contaminated and noncontaminated areas and thus maintains the airborne doses to workers in noncontaminated areas to ALARA conditions. Drawing M-76 identifies the areas served by each of the systems. Because the sources of equipment leakage within the reactor enclosure (e.g., RWCU pumps, RHR system, etc) are served by the filtered exhaust, it was assumed that all reactor enclosure releases are filtered.

To use the NUREG-0016, Revision 0, GALE code estimates of releases, it was conservatively assumed that the reactor enclosure (containment/auxiliary building) contributions were all released through 2 inch, 70% efficient charcoal filters and 99% efficient HEPA filters.

b. Turbine Enclosure The areas of the turbine enclosure containing radioactive sources that could leak and result in airborne activity have been identified as shown in Sections 11.3.2.2.3 and 9.4.4.2.5 and on drawing M-75. The turbine enclosure ventilation systems have been set up such that the air flows from the clean areas to the areas of potential contamination and then is exhausted through the 8 inch charcoal and HEPA filters of the equipment compartment exhaust system. In addition, air from the clean areas is exhausted or recirculated by the turbine enclosure ventilation system to keep the airborne doses ALARA to workers in noncontaminated areas and to minimize the cost and doses during maintenance for a larger filter system.

Because all identified sources of activity are served by the filter system, it was assumed that all turbine enclosure releases identified by NUREG-0016, Revision 0, and the GALE code are released through the 8 inch, 99% efficient charcoal filters and the 99% efficient HEPA filters.

c. Radwaste Enclosure and Chemistry Laboratory Expansion As indicated in Sections 11.3.2.2.2 and 9.4.3 and on drawing M-79, HEPA filters are installed on tank vents, HEPA filters are installed on the equipment compartment exhausts and the fume hood exhaust system, and the clean areas are exhausted unfiltered. By directing air flow to contaminated areas from the clean areas rather than using a recirculation-type system, airborne contamination is minimized, and most of the released activity is filtered before it is released. For the analysis indicated in Table 11.3-2, it was assumed that all particulate releases are filtered through either the 99% efficient tank vent HEPA filter or the 99% efficient HEPA filter on the equipment compartments identified as having the greatest potential for leakage. The above assumptions are reflected in the release analyses summarized in Tables 11.3-1 and 11.5-4, and the dose analyses summarized in Tables 11.3-7 and 11.3-4.

Expected average annual radionuclide concentrations are compared to 10CFR20 limits in Table 11.3-7.

The doses resulting from gaseous effluents are a small fraction of the 10CFR20 dose limit of 500 mRem/year and are well within 10CFR50, Appendix I design objectives, as shown in Table 11.3-4.

CHAPTER 11 11.3-15 REV. 18, SEPTEMBER 2016

LGS UFSAR All gaseous releases are through the following three release points: the north stack, the south stack (one per unit), and the hot maintenance shop.

a. Offgas system

b. Mechanical vacuum pump and gland seal condenser exhaust systems

c. Containment purge system

d. SGTS

e. Turbine enclosure ventilation system

f. Radwaste enclosure ventilation system

g. Chemistry Laboratory Expansion air exhaust system

h. Chemistry Laboratory Expansion fume hood air exhaust system The south stack serves the refueling floor and reactor enclosure ventilation exhaust. The locations of the stacks are shown on reactor enclosure general arrangement drawings in Section 1.2.

The hot maintenance shop exhaust serves the hot maintenance ventilation exhaust system.

The height, flow rate, heat content, and dimensions of the three release points are shown on Table 11.3-6.

11.

3.4 REFERENCES

11.3-1 NUREG-0016, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors (BWR-GALE Code)", NRC, (April 1976).

11.3-2 "Standards for Steam Surface Condensers", 6th Edition, Heat Exchange Institute, New York, NY, (1970).

11.3-3 D.P. Siegworth, "Measurement of Dynamic Adsorption Coefficients for Noble Gases on Activated Carbon", 12th AEC Air Cleaning Conference, (1971).

11.3-4 D. Underhill, "Design of Fission Gas Holdup Systems," 11th AEC Air Cleaning Conference, (1970).

11.3-5 "Analysis of Detonation Design Pressure for Hydrogen LGS Charcoal Offgas Treatment Systems", Helix Process Systems - PROPRIETARY.

CHAPTER 11 11.3-16 REV. 18, SEPTEMBER 2016

LGS UFSAR Table 11.3-1 EXPECTED ANNUAL ACTIVITY RELEASED FROM GASEOUS WASTE MANAGEMENT SYSTEMS (1)

(Curies/year; 2 units)

REACTOR TURBINE RADWASTE GLAND AIR MECHANICAL NUCLIDE ENCLOSURE ENCLOSURE ENCLOSURE SEAL EJECTOR VACUUM PUMP TOTAL Ar-41 5.10e+01 <1.0 <1.0 <1.0 <1.0 <1.0 5.10E+01 Kr-83m <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 Kr-85m 1.22E+01 1.39E+02 <1.0 <1.0 1.22E+01 <1.0 1.63E+02 Kr-85 <1.0 <1.0 <1.0 <1.0 5.71E+02 <1.0 5.71E+02 Kr-87 1.22E+01 2.65E+02 <1.0 <1.0 <1.0 <1.0 2.77E+02 Kr-88 1.22E+01 4.69E+02 <1.0 <1.0 <1.0 <1.0 4.81E+02 Kr-89 <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 Xe-131m <1.0 <1.0 <1.0 <1.0 1.43E+01 <1.0 1.43E+01 Xe-133m <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 Xe-133 2.65E+02 5.10E+02 2.04E+01 <1.0 1.14E+01 4.69E+03 5.60E+03 Xe-135m 1.88E+02 1.33E+03 <1.0 <1.0 <1.0 <1.0 1.51E+03 Xe-135 1.39E+02 1.29E+03 9.18E+01 <1.0 <1.0 7.14E+02 2.23E+03 Xe-137 <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 <1.0 Xe-138 2.86E+01 2.86E+03 <1.0 <1.0 <1.0 <1.0 2.88E+03 TOTAL NOBLE GASES 1.38E+04 I-131 2.04E-01 3.88E-03 1.02E-01 <1.0E-4 <1.0E-4 6.12E-02 3.71E-01 I-133 8.16E-01 1.55E-02E 3.67E-01 <1.0E-4 <1.0E-4 <1.0E-4 1.20E+00 TOTAL HALOGENS(4) 1.57E+00 CHAPTER 11 11.3-17 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-1 REACTOR TURBINE RADWASTE GLAND AIR MECHANICAL NUCLIDE ENCLOSURE ENCLOSURE ENCLOSURE SEAL EJECTOR VACUUM PUMP TOTAL Tritium Gaseous Release 1.44E+02 Carbon-14 <1.0 <1.0 <1.0 <1.0 1.90E+01 <1.0 5.10E+01 Cr-51 1.22E-05 2.65E-04 1.84E-04 0.00E+00 0.00E+01 0.00E+00 4.61E-04 Mn-54 1.22E-04 1.22E-05 6.12E-04 0.00E+00 0.00E+00 0.00E+00 7.47E-04 Co-58 2.45E-05 1.22E-05 9.18E-05 0.00E+00 0.00E+00 0.00E+00 1.29E-04 Fe-59 1.63E-05 1.02E-05 3.06E-04 0.00E+00 0.00E+00 0.00E+00 3.33E-04 Co-60 4.08E-04 4.08E-05 1.84E-03 0.00E+00 0.00E+00 0.00E+00 2.28E-03 Zn-65 8.16E-05 4.08E-06 3.06E-05 0.00E+00 0.00E+00 0.00E+00 1.16E-04 Sr-89 3.67E-06 1.22E-04 9.18E-06 0.00E+00 0.00E+00 0.00E+00 1.35E-04 Sr-90 2.04E-07 4.08E-07 6.12E-06 0.00E+00 0.00E+00 0.00E+00 6.73E-06 Zr-95 1.63E-05 2.04E-06 1.02E-06 0.00E+00 0.00E+00 0.00E+00 1.94E-05 Sb-124 8.16E-06 6.12E-06 1.02E-06 0.00E+00 0.00E+00 0.00E+00 1.53E-05 Cs-134 1.63E-04 6.12E-06 9.18E-05 0.00E+00 0.00E+00 0.00E+00 2.67E-04 Cs-136 1.22E-05 1.02E-06 9.18E-06 0.00E+00 0.00E+00 0.00E+00 2.65E-05 Cs-137 2.24E-04 1.22E-05 1.84E-04 0.00E+00 0.00E+00 0.00E+00 4.41E-04 Ba-140 1.63E-05 2.24E-04 2.04E-06 0.00E+00 0.00E+00 0.00E+00 2.65E-04 Ce-141 4.08E-06 1.22E-05 5.30E-05 0.00E+00 0.00E+00 0.00E+00 6.94E-05 TOTAL AIRBORNE PARTICULATE RELEASE 5.32E-03 (1) Estimated releases based on NUREG-0016, Revision 0, GALE Code evaluation.

(2) * = Less than 1.0 Ci/yr (3) ** = Less than 1.0x10-4 Ci/yr (4) Includes both gaseous and particulate releases.

CHAPTER 11 11.3-18 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-2 ASSUMPTIONS AND PARAMETERS USED FOR EVALUATION OF GASEOUS RELEASES Power 3527 MWt Capacity factor 80%

Total steam flow 14.863x106 lb/hr Mass of water in reactor 3.8x105 lb RWCU filter/demineralizer flow 1.33x105 lb/hr Fraction of feedwater through condensate 1.00 filter/demineralizer Gland seal system uses clean system with no radioactive releases Reactor Enclosure Iodine Release Fraction 0.3 (2" Carbon Filter)

Particulate Release Fraction 0.01 (HEPA Filter)

Turbine Enclosure Iodine Release Fraction 0.01 (8" Deep-Bed Charcoal Filter)

Particulate Release Fraction 0.01 Radwaste Enclosure Iodine Release Fraction 1.0 Particulate Release Fraction 0.01 Mechanical Vacuum Pump Iodine Release Fraction 0.01 Particulate Release Fraction 0.01 Charcoal Delay System designed with vendor coefficients for 35-day Xenon holdup at 75 scfm condenser inleakage. (This corresponds to 58.6 days Xe holdup using the NUREG-0016, Revision 0 methodology).

CHAPTER 11 11.3-19 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-3 OFFGAS SYSTEM MAJOR EQUIPMENT DESCRIPTION(3)

DESIGN PRESSURE/

EQUIPMENT TEMPERATURE EQUIPMENT NUMBERS TYPE QTY MATERIAL CAPACITY SIZE (psig/F)

Preheater 10E/20E-131 Shell and U-tube 2 Shell, Channel: CS 681,800 Btu/hr 697 ft² Shell side: 350/450 Tubes, Sheet: SS Effective area Tube side: 350/450 Aftercondenser 10E/20E-127 Shell and straight 2 Shell, sheet: SS 16.3x106 Btu/hr 1003 ft² Shell side: 350/1100 tube Tubes: Sanicro-28 Tube side: 150/440 (Unit 1), SS (Unit 2)

Chanel: CS Recombiner 10S/20S-125 Vertical Cylinder 2 Shell: SS 1600 lbs Vessel: 76" OD, 350/1100 (Vessel )

Internals: Catalyst support Catalyst 102" high Assembly (SS)

Catalyst: Metal mat coated Catalyst: 51" with precious metals Dia, 35" deep Holdup pipe KBG-106 2 CS 26" dia x 125"long 600/850 Outlet HEPA filter 10F/20F-371 Vertical Cylinder 2 Vessel: CS 300 scfm at 0.4 Vessel: 54" high, 14" 445/150 Internals: C size filter psi dia element Cartr: dia Glycol cooler condenser 10E/20E-377 Shell and straight 2 Shell: SS 86,760 Btu/hr 100.8 ft² 615/150 tube Tubes: SS Guard bed vessel 10S/20S-370 Vertical Cylinder 2 136" dia x 36" high 175(2)/150 Charcoal adsorber 1AS/1BS-371 Vertical Cylinder 2 CS 82,125 lbs 132" dia x 354" high 375/150 vessels Unit 1 1CS-1GS-371 5 CS Charcoal each 31,500 lbs 120" dia x 174" high 250(1)/150 Charcoal lbs CHAPTER 11 11.3-20 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-3 (Cont'd)

DESIGN PRESSURE/

EQUIPMENT TEMPERATURE EQUIPMENT NUMBERS TYPE QTY MATERIAL CAPACITY SIZE (psig/F)

Charcoal adsorber 2AS-371 Vertical Cylinder 1 CS 78,750 lbs 132" dia x 354" high 375/150 vessels Charcoal Unit 2 2BS-2IS-371 8 CS 30,375 lbs 120" dia x 174" high 250(1)/150 Charcoal (each)

Charcoal beds --------------- Sutcliffe Speakman Charcoal Adsorber Mesh size 8x16 --------------

Units 1 and 2 203C coefficient At design temperature Xe: 733, Kr: 31.8 Glycol unit pumps 1AP/BP-700 Centrifugal 4 CS 40 gpm ----------------- 125 psig 2AP/BP-700 -----------

Glycol fill Tank 10/20T-549 Vertical Cylinder 2 CS ----------- ----------------- 125/150 Glycol Expansion tank 10/20T-550 Vertical Cylinder 2 CS ----------- ----------------- 125/150 Glycol Air Separator 10/20S-515 Vertical Cylinder 2 CS ----------- ----------------- 125/150 Refrigeration units 1/2AE-376 Semi-Hermetic 4 Freon 22 Refrigerant 92,135 Btu/hr (glycol) 1/2BE-376 Refrigeration units OAE-378 Semi-Hermetic 2 Freon 22 Refrigerant 125,000 Btu/hr ----------------- -----------

(compartment cooling) OBE-378 (1) Vessels have been hydrostatically tested at 375 psig.

(2) Calculation of code static-equivalent pressure based on actual full wall thickness and material properties yields 359 psig.

(3) Design Codes and Standards are provided in Table 3.2-1.

CHAPTER 11 11.3-21 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-4 COMPARISON OF MAXIMUM INDIVIDUAL DOSES RESULTING FROM LGS UNIT 1 & 2 WITH 10CFR50, APPENDIX I DESIGN OBJECTIVES 10CFR50, APPENDIX I RM-50-2 LGS DESIGN SITE GASEOUS DOSE OBJECTIVE DESIGN UNITS 1 & 2 (2 UNITS) OBJECTIVE(3)

Gamma air dose 0.88 20 -

(mRad/yr)(1)

Beta air dose 0.60 40 -

(mRad/yr)(1)

Total body of 0.47 10 5 individual (mRem/yr)(1)

Skin of 0.92 30 15 individual (mRem/yr)(1)

Any organ all 10.76 30 15 pathways (mRem/yr)(2) (Thyroid)

(1)

Doses from noble gases only.

(2)

Doses from radioiodines and air particulates with half-lives greater than eight days.

(3)

Annex to 10CFR50, Appendix I, "Concluding Statement of Position of the Regulatory Staff, Public Rulemaking Hearing on Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As-Low-As-Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Reactors", AEC Docket No. RM-50-2.

CHAPTER 11 11.3-22 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-5 OFFGAS SYSTEM EQUIPMENT MALFUNCTION ANALYSIS EQUIPMENT ITEM MALFUNCTION CONSEQUENCES DESIGN PRECAUTIONS

1. SJAE pressure Fail closed Insufficient dilution of Low flow signal will shut offgas control valves hydrogen in offgas. suction valve. Low flow and low SJAE supply steam pressure will alarm in control room. Operator can start redundant SJAE train.

2. Piping from SJAE to Pressure boundary Increased local airborne Area monitors in normally and including holdup leakage radioactivity levels and accessible areas will alarm.

Pipe increased releases from Ventilation releases will be plant ventilation. filtered by turbine equipment compartment exhaust system and monitored. Low pressure system, radiographed welds, and system is designed to withstand hydrogen detonation.

3. Recombiner preheater Steam leak from tube 1. Increased preheater 1. Temperature elements at side into shell side. outlet temperature preheater outlet would regulate steam flow to preheater

2. Increased quantity of 2. Additional steam flow would dilution steam be condensed in the after condenser and returned to CRW or the main condenser.

Liquid level instrument 1. Level control valve 1. Redundant level control fails or level control opens. Preheater level instrumentation. Alarm on valve fails drops. Steam recycles low condensate level in through drain line to preheater. Operator closes level main condenser. control valve or drain isolation valve from control room .

Redundant level control valve provided.

2. Level control valve 2. Redundant level control valve closes. provided. Will actuate automatically.

CHAPTER 11 11.3-23 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-5 (Cont'd)

EQUIPMENT ITEM MALFUNCTION CONSEQUENCES DESIGN PRECAUTIONS

4. Recombiners Catalyst deteriorates 1. Recombiner outlet 1. Outlet temperature monitors or is wetted temperature drops. record low temperature.

2. High hydrogen 2. Analyzers record hydrogen level concentrations at and alarm at 2% and 4%. Operator recombiner outlet. isolates gas flow, or adjusts dilution.

5. Aftercondenser Cooling water leak The cooling water would Leakage would be detected by high (tube to shell side) leak to the shell side of conductivity in the drain line.

the heat exchanger. Water High conductivity alarms in flow would increase to the main control room .

condenser. The aftercondenser drain can be swapped manually to clean radwaste.

Excessive leakage would actuate aftercondenser high level alarm.

Liquid level 1. Level control valve 1. Redundant level instrument instrument fails or opens. Aftercondenser alarms on low condensate level control fails shell drain level level in aftercondenser.

drops or air cooler drain tank level Operator closes level drops. Steam or control valve or drain noncondensables recycle isolation valve from the back to the main condenser, control room. Loop seal or clean radwaste. provided on drain to clean radwaste to prevent offgas entering clean radwaste.

Loop seal is self-resealing type.

2. Level control valve 2. Redundant level control valve closes. provided. Redundant level control valve provided.

6. Piping from holdup Pressure boundary Release of offgas mixture Local radiation pipe to HEPA filter leakage to radwaste or offgas monitors in the radwaste exit enclosure enclosure will detect leakage.

Ventilation monitors in ducts from the offgas enclosure will detect leakage from offgas piping.

CHAPTER 11 11.3-24 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-5 (Cont'd)

EQUIPMENT ITEM MALFUNCTION CONSEQUENCES DESIGN PRECAUTIONS

7. Holdup pipe Liquid level 1. Level control valve opens. 1. Redundant level instrumentation.

instrument fails or Holdup pipe drain pot Alarm on low condensate level level control valve liquid level drops. in drain pot.Operator closes fails Offgas recycles through level control valve or drain the drain line to the isolation valve from control main condenser or to room. Loop seal provided on clean radwaste. drain to clean radwaste to prevent offgas entering clean radwaste. Loop seal is of the self-resealing type.

2. Level control valve 2. Redundant level control valve closes provided .

8. Glycol system Mechanical failure If glycol unit fails, The failure of the glycol system (pumps or refrigeration the moisture content of will activate alarms at local panel units) the process stream will and alarms at main control room increase resulting in panel. Redundant glycol decreased adsorption. refrigeration units and glycol pumps are installed. Moisture instrumentation and temperature instrumentation is provided at the cooler condenser outlet.

High outlet temperature and high moisture are alarmed at the local panel and in the main control room.

In addition, system is oversized and glycol loop has sufficient thermal capacity which allows time to change over to redundant equipment.

9. Charcoal Moisture in gas Charcoal adsorption Increasing moisture in inlet stream adsorbers stream performance will will be detected by the moisture deteriorate gradually element. Various design precautions as charcoal gets wet. are included upstream to preclude Holdup times decrease moisture from entering adsorbers.

and plant gaseous Sacrificial guard beds are provided releases will increase. to protect main adsorber vessels.

CHAPTER 11 11.3-25 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-5 (Cont'd)

EQUIPMENT ITEM MALFUNCTION CONSEQUENCES DESIGN PRECAUTIONS

10. Charcoal vault air Mechanical failure If ambient temperature Redundant air conditioning units conditioning units of refrigeration units increases, delay efficiency are provided. Redundant fan and or air coolers. of charcoal beds decreases. cooler for each cell provided.

Increased releases would High ambient temperature occur depending on the alarms in main control fuel leakage rate. room and local panel. Ambient temperature is recorded in the main control room. Refrigeration trouble alarm in main control room. Refrigeration system capacity is oversized for heat load.

11. Outlet HEPA filter Plugging of filter Increasing differential Differential pressure is alarmed media pressure in HEPA will in control room and indicated create higher system locally. DOP test connections back pressure. SJAE are provided. HEPA filter can bypass valve may open be bypassed for servicing.

recycling offgas back to main condenser.

CHAPTER 11 11.3-26 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.3-6 RELEASE POINT DATA TURBINE ENCLOSURE REACTOR ENCLOSURE HOT MAINTENANCE NORTH STACK SOUTH STACK SHOP (one only) (one per unit)

Height of release point above grade, ft 200 200 199 Annual average rate of air 458,000 223,000 7,000 flow from release point, cfm Annual average heat flow from 26.2x106 5.4x106 ---

release point, Btu/hr Type and size of release point Duct Duct Duct 20'x12' 10'x10' CHAPTER 11 11.3-27 REV. 18, SEPTEMBER 2016

LGS UFSAR Table 11.3-7 COMPARISON OF MAXIMUM CALCULATED RADIONUCLIDE CONCENTRATIONS IN THE ENVIRONMENT FROM ROUTINE ATMOSPHERIC RELEASES TO 10CFR20 LIMITS LOCATION: ESE SITE BOUNDARY 10CFR20 AVERAGE ANNUAL TABLE II CONCENTRATION COLUMN 1  % OF IN AIR LIMIT(1) 10CFR20 ISOTOPE (Ci/ml) (Ci/ml) LIMIT Ar-41 1.00E-12 3E-08 3.3E-03 Kr-85M 3.23E-12 1E-07 3.2E-03 Kr-85 1.14E-11 3E-07 3.8E-03 Kr-87 5.57E-12 2E-08 2.8E-02 Kr-88 5.59E-12 2E-08 2.8E-02 Xe-131M 2.85E-13 4E-07 7.1E-05 Xe-133 1.10E-10 3E-07 3.7E-02 Xe-135M 2.69E-11 3E-08 9.0E-02 Xe-135 4.46E-11 1E-07 4.5E-02 Xe-138 5.05E-11 3E-08 1.7E-01 I-131 6.73E-15 1E-10 6.7E-03 I-133 2.24E-14 4E-10 5.6E-03 Cr-51 7.84E-18 8E-08 9.8E-09 Mn-54 1.26E-17 1E-10 6.7E-03 Fe-59 5.46E-18 2E-09 2.7E-07 Co-58 2.15E-18 2E-09 1.1E-07 Co-60 3.75E-17 3E-10 7.5E-07 Zn-65 1.95E-18 2E-09 9.7E-08 Sr-89 2.25E-18 3E-10 7.5E-07 Sr-90 1.12E-19 3E-11 3.7E-07 Zr-95 3.24E-19 1E-09 2.6E-07 Sb-124 2.56E-19 1E-10 2.6E-07 Cs-134 4.44E-18 4E-10 1.1E-06 Cs-136 4.44E-19 6E-09 7.4E-09 Cs-137 7.51E-18 5E-10 1.5E-06 Ba-140 4.44E-18 1E-09 4.4E-07 Ce-141 1.16E-18 5E-09 2.3E-08 H-3 2.39E-12 2E-07 1.5E-03 C-14 3.86E-13 1E-07 3.9E-04 (1)

The concentration limits shown are reflective of the 10CFR20 Appendix B values prior to the 1994 revision of 10CFR20, and are consistent with those used in the original licensing of the plant.

Current effluent release limits are based on the post-1994 10CFR20 regulation and are controlled by the Radioactive Effluent Controls Program defined by the Technical Specifications. New 10CFR20 limits went into effect on Jan. 1, 1994. As part of the power rerated evaluation, a comparison of the uprated atmospheric release concentrations with the new 10CFR20 limits was made. All the uprated released concentrations were found to be less than 1% of their new limit values.

CHAPTER 11 11.3-28 REV. 14, SEPTEMBER 2008

LGS UFSAR 11.4 SOLID WASTE MANAGEMENT SYSTEM The applicant is committed to providing a solid waste management system that complies with the intent of BTP ETSB 11-3, "Design Guidance for Solid Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants.". A process control program that ensures suitability of packaged wastes for shipment and burial in consideration of applicable federal and state regulations and other requirements was submitted in conformance with the guidelines of SRP section 11.4 of NUREG-0800 (Reference 11.4-1). Process parameter tolerances were verified during preoperational testing of the solid radwaste system, and the final Process Control Program was submitted on March 7, 1985. Any revision to the program requires PORC approval.

The solid waste management system collects, monitors, processes, packages, and provides temporary storage facilities for radioactive spent bead and powdered resins and dry solid wastes for offsite shipment and permanent disposal. The solid waste management system does not have any safety-related functions. For the purpose of this section, the term "solid waste" is used for spent bead and powdered resins and dry solid waste produced from plant operation.

In addition, Class B and Class C Low-Level Radioactive Waste (LLRW) produced from plant operation may be shipped to Peach Bottom Atomic Power Station for storage.

Process and effluent radiological monitoring systems are discussed in Section 11.5.

11.4.1 Design Bases

a. The design objectives of the solid waste management system are:

1. Provide collection, processing, packaging, and storage of solid wastes resulting from normal plant operations without limiting the operation or availability of the plant

2. Provide a reliable means for handling solid wastes and allow system operation with ALARA radiation exposure to plant personnel

3. Package solid wastes in suitable containers for offsite shipment and burial

4. Prevent the release of significant quantities of radioactive materials to the environment so as to keep the overall exposure to the public well within 10CFR20 limits

b. Redundant and backup equipment, alternate routes, and interconnections are designed into the system to provide for operational occurrences such as refueling, abnormal leak rates, decontamination activities, equipment downtime, maintenance, and repair.

c. Equipment locations, compartments, drainage, ventilation, and components are designed to reduce maintenance, equipment downtime, leakage, and gaseous releases of radioactive materials to the structure atmosphere or to otherwise improve the system operations.

CHAPTER 11 11.4-1 REV. 18, SEPTEMBER 2016

LGS UFSAR

d. The solid waste management system is designed to package radioactive solid wastes for offsite shipment and burial in accordance with the requirements of applicable NRC and DOT regulations including 10CFR71 and 49CFR170 through 49CFR178. Final waste classification and waste form for offsite disposal will be consistent with the guidelines of 10CFR61. A shipping manifest consistent with 10CFR20.2006 will accompany the waste shipment. Waste stability for Class B and Class C wastes is provided by the use of high integrity containers which are designed in accordance with the "General Criteria for High Integrity Containers" established by the South Carolina Department of Health and Environmental Control on October 22, 1980. Concrete overpacks are used at the burial site to provide the compressive strength required for burial. This results in radiation exposures to individuals and the general population well within the limits of 10CFR20 and 10CFR50.

e. The solid waste management system is designed to seismic Category II requirements. The quality group classification and corresponding codes and standards that apply to the design of the solid waste management system are discussed in Section 3.2.

f. The expected and maximum radionuclide activity inventories of solid waste management system components containing significant amounts of radioactive liquids are shown in Tables 11.4-8 and 11.4-9.

1. Expected flow rates for streams shown in Figure 11.4-3 are given in Table 11.4-4.

2. Expected inputs and activities to the solid waste management system are shown in Tables 11.4-5 and 11.4-6.

g. A quality assurance program for the packaging and transportation of radioactive material shall be established which complies with the applicable requirements of 10CFR50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants", and 10CFR71, subpart H, "Packaging and Transportation of Radioactive Material, Quality Assurance".

11.4.2 SYSTEM DESCRIPTION The solid waste management system P&IDs are shown in drawings M-66 and M-67. The process flow diagram is shown in Figure 11.4-3. Major equipment design parameters are listed in Tables 11.4-1, 11.4-2, and 11.4-3.

11.4.2.1 System Operation The activities of the wastes entering the solid waste management system are dependent on the liquid activities in the various liquid systems such as the condensate, RWCU, fuel pool cleanup, equipment drain, and floor drain systems, whose activities are in turn a function of the reactor coolant activity. The activity of the reactor coolant is discussed in Section 11.1. Liquid waste management systems that supply input to the solid waste management system are discussed in Section 11.2.

CHAPTER 11 11.4-2 REV. 18, SEPTEMBER 2016

LGS UFSAR The quantities of solid wastes generated are dependent on the plant operating factor, extent of equipment leakage, plant maintenance and housekeeping, and decontamination requirements.

Input to the solid waste management system is predominantly captured solids from filter/demineralizers and bead resins from deep-bed demineralizers. Powdered and bead resins are dewatered and then packaged in HICs or other approved containers for offsite disposal (Section 11.4.2.4).

11.4.2.1.1 Wet Solid Waste Processing Wet solid wastes consist primarily of filtered solids from condensate filter/demineralizers and filtered solids, spent bead resins and powdered resins backwashed from RWCU, deep bed condensate demineralizers, floor drain, equipment drain, and fuel pool cleanup systems. Only spent resins from the RWCU system are expected to exceed LSA criteria. The rest of the solid wastes are expected to be low specific activity, as defined in 10CFR71.

The originally installed plant centrifuge system is no longer in service. Spent resins are transferred to the external processing station for dewatering through permanently installed plant systems/piping (Section 11.4.2.2.7).

The condensate filter/demineralizer are backwashed to the respective condensate backwash receiving tank and from there are pumped to one of two (per unit) phase separators. When the backwash material has settled in the phase separator, the clear water is decanted to the equipment drain tank for further processing. When a preset sludge level is reached in a phase separator, the other phase separator is used, while the sludge in the first phase separator is allowed to decay. After decay the sludge is pumped in slurry form to the external processing station for dewatering.

Backwash from the RWCU system is handled in a manner similar to the condensate filter/demineralizer system. There are two RWCU phase separators shared between two units.

Each RWCU phase separator is sized for a normal 60 day collection (from both units) and 60 day decay time.

Spent bead resin from the DBCD can be transferred to the floor and equipment drain demineralizers for re-use (Section 10.4.6.2). DBCD system spent resin can also be processed externally (vendor process) by pumping resin from the DBCD system spent resin tanks (SRT) to the external processing station for collection in liners or high integrity containers.

Spent powdered resins from the equipment drain and floor drain filters and fuel pool cleanup filter/demineralizer are backwashed to the waste sludge tank, transferred to a condensate phase separator for processing and then pumped to the external processing station for dewatering. The waste sludge tank contents can also be processed directly to the external processing station for dewatering, based on plant conditions. Equipment and floor drain demineralizer spent resins were designed to be sluiced to their respective intermediate spent resin tanks before being pumped to the waste sludge tank. However, for normal plant operation, the spent resin tanks are not used.

The exhausted bead resin from the liquid radwaste demineralizers will be sluiced directly to the waste sludge tank via a permanently installed bypass line.

The waste sludge tank holdup capacity does not approach the BTP ETSB 11-3 30 day criteria.

However, the tank contents can be routed to the condensate phase separator for holdup and decanting in lieu of immediate dewatering by external processing. The use of redundant CHAPTER 11 11.4-3 REV. 18, SEPTEMBER 2016

LGS UFSAR dewatering process trains is expected to preclude the need for significant upstream waste slurry storage from a process equipment availability standpoint.

Slurries from the phase separators or the waste sludge tank are pumped to the external processing station. The slurry is approximately 5% by weight solid content and is dewatered to approximately 40% to 60% by weight solid content.

The vendor's external dewatering system shall be operated in accordance with the system's Topical Report and applicable procedures.

During initial system testing (preoperational), process parameters were established that provided reasonable assurance that the dewatered sludge contained essentially zero free liquid.

Periodically during external waste processing, the free liquid parameters will be verified. Sludge samples may also be taken if an abnormal condition develops during waste processing. Records of process parameters will be maintained for individual waste batches. Dewatered waste shipments will comply with the applicable burial site criteria for maximum acceptable free liquid content, when shipped for burial.

11.4.2.1.2 Concentrated Liquid Waste Processing Installation of the radwaste evaporators was not completed for plant operation and the system has been abandoned (Section 11.2.2.1.3). Because the solid waste management system does not include an installed solidification subsystem, any future concentrates produced would be solidified prior to offsite shipment by an acceptable mobile solidification system connected to the external processing station.

11.4.2.1.3 Dry Solid Waste Inputs Dry wastes consist of air filters, miscellaneous paper, rags, etc, from contaminated areas; contaminated clothing, tools, and equipment parts that cannot be effectively decontaminated; and solid laboratory wastes. The activity of much of this waste is low enough to permit handling by contact. These wastes are collected in containers located in appropriate areas throughout the plant, as dictated by the volume of wastes generated during operation and maintenance.

Compressible wastes are packaged manually into strong, tight containers. Noncompressible wastes are packaged manually in similar strong, tight containers or in other suitable containers.

The filled containers are sealed and moved to a controlled access enclosed area for temporary storage. Because of its low activity, this waste can be stored until enough is accumulated to permit economical transportation to an offsite vendor for processing and/or to an offsite burial ground for final disposal.

11.4.2.1.4 Irradiated Reactor Internals Irradiated reactor internals being replaced are removed from the RPV underwater and stored for radioactive decay in the spent fuel storage pool. Section 9.1.4 describes reactor vessel and in-vessel servicing equipment used for handling reactor components.

Some irradiated reactor internals are processed and then stored in Storage Containers (SCs) in free-standing storage racks in the cask loading pit prior to shipping. The rack compartments hold LLW including Control Rod Blades (CRBs), fuel channels in sheaths, LPRM strings, velocity limiters, filters, stellite ball bearing and other small components. During LLW processing, processing equipment will be placed in the cask loading pit or on the Refuel Floor.

CHAPTER 11 11.4-4 REV. 18, SEPTEMBER 2016

LGS UFSAR An estimated average of seven of the control rod blades will be removed at each reactor outage (starting 10-15 years after operation) and stored on hangers on the fuel pool walls or in racks interspersed with the spent fuel racks. Offsite shipping is done in NRC licensed shipping casks.

An estimate of 30% of the power range monitor detectors will be replaced at each reactor outage.

Actual power range monitor detector replacement is as required based on detector life. Spent incore detectors and dry tubes are cut or bent as required to facilitate subsequent storage. The incore detectors and dry tubes are stored with adequate water shielding on hangers on the fuel pool walls or within suitable storage containers. A cutting tool allows remote cutting of the incore detectors and dry tubes in the fuel pool or cask pit. The cut incore monitors and dry tubes and other small sized reactor internals are shipped offsite in suitable containers and/or shielded casks that can be loaded underwater.

A trolley-mounted disposal cask with an internal cable drum is supplied with the NSSS for source and intermediate range neutron monitor detector cables and the TIP wires.

11.4.2.2 Process Equipment Description Major components of the solid waste management system include pumps, tanks, piping, centrifuges, discharge chute assemblies, capping machine, decontamination equipment, and handling equipment. The entire system is located in the radwaste enclosure with the exception of the condensate and reactor water cleanup backwash receiving tanks, which are located in the control enclosure and reactor enclosure, respectively. The designs of the radwaste, turbine, and reactor enclosures are discussed in Section 3.8.4. Equipment design parameters are listed in Tables 11.4-1, 11.4-2, and 11.4-3.

11.4.2.2.1 Pumps Process pumps are vertical in-line centrifugal pumps of ASME Section III, Class 3 construction.

All are made of carbon steel except the RWCU sludge discharge/mixing and the RWCU decant pumps, which are of stainless steel. All are provided with a mechanical-type seal.

11.4.2.2.2 Tanks System collection and phase separator tanks are sized for normal plant waste volumes with sufficient excess capacity to accommodate equipment downtime and expected maximum volumes that may occur. Cross-connections between tanks are provided as appropriate for greater operational flexibility. Air spargers or recirculation lines are provided in the tanks to create a homogeneous slurry for pumping. All tanks are provided with overflow lines to route any inadvertent overflow to liquid radwaste collection sumps. All tanks are vented to their enclosures' respective ventilation system. See Section 9.4.3 for a discussion of the radwaste enclosure ventilation system.

All tanks are constructed to API 650. See Table 11.4-1 for tank materials.

11.4.2.2.3 Piping The system piping material is carbon steel. The external processing station piping is stainless steel. Line sizing is based on maintaining adequate flow velocities to maintain slurries in CHAPTER 11 11.4-5 REV. 18, SEPTEMBER 2016

LGS UFSAR suspension. The piping is laid out to avoid low points and other features that could create local "hot spots." The lines are normally flushed with condensate after a pumping or draining operation.

Process piping is constructed to ASME Section III, Class 3 or ANSI B31.1 (See Section 3.2 for further details).

11.4.2.2.4 Centrifuges Two centrifuges were provided to dewater filter sludges and spent resins but are no longer in service. The water removed was returned to a condensate phase separator by gravity, and the dewatered sludges were directed to the HICs. The centrifuges are fabricated of stainless steel and are the continuous-feed, horizontal, solid bowl, sanitary-type.

11.4.2.2.5 Discharge Chute Assemblies Discharge chutes from the centrifuges are equipped with telescoping fill chutes to interface with the HIC openings. A drip pot and sluice pan mechanism is positioned below the chute during chute flush and while HIC filling is not in progress. Major components with surfaces contacting radioactive materials are fabricated of stainless steel. This equipment is no longer in service.

11.4.2.2.6 Capping Machines The capping machine automatically caps the HIC. Operation is controlled locally at the process cell viewed through the shielded glass window to verify proper closure. This equipment is no longer in service.

11.4.2.2.7 Solidification and Dewatering Equipment The solid waste management system does not include a permanently installed solidification capability and the originally installed plant centrifuge system, designed for dewatering, is no longer in service. An external processing station has been provided to accommodate the use of a mobile solidification or dewatering system. Waste processing via a vendor's mobile solidification or dewatering system will be accomplished consistent with the vendor's process control procedures.

Processed waste from the vendor's system will be packaged in suitable containers consistent with appropriate NRC and DOT regulations and receiving burial site requirements.

11.4.2.2.8 Decontamination Station Equipment HICs were washed down with spray nozzles and air blast dried within an enclosed ventilated space within the decontamination cell to minimize the spread of contamination. A swipe sample mechanism and a contact radiation monitor are provided to verify decontamination and to determine the radiation level for shipping considerations. This equipment is no longer in service.

11.4.2.2.9 Information in this section has been deleted.

11.4.2.2.10 Handling Equipment HIC handling is accomplished by an overhead crane and transfer carts.

CHAPTER 11 11.4-6 REV. 18, SEPTEMBER 2016

LGS UFSAR The overhead crane moves HICs to and from storage cells and to trucks for offsite shipping.

Operations can be viewed through a shielded glass window as well as on closed-circuit television monitors. Two area television cameras and one crane-mounted camera are provided.

HICs are moved in and out of process cells on railed, electric motor-driven transfer carts. If there is a motor failure, the carts can be manually placed in position by a push rod for HIC removal and for repair access. Cart operations are viewed and controlled from the shielded glass window of the process cell. Transfer carts are no longer in service.

11.4.2.3 Expected Volumes (Based on Original Plant Design Estimates)

It is estimated that approximately 12,720 cubic feet of dewatered waste will be generated per year. A breakdown of these wastes is given in Table 11.4-5.

Dry compressed waste is estimated at 28,500 cubic feet per year, representing about 3,886 fifty-five gallon drums per year. Noncompressible dry waste (filters, tools, etc) is estimated at 25,000 ft3/yr. A breakdown of the dry waste from air filters is given in Table 11.4-6.

Table 11.4-10 presents the expected annual offsite shipment of solid wastes and their curie content.

11.4.2.4 Packaging Low specific activity and other dewatered wastes, which meet the 10CFR61 classification as Class B and Class C waste, are packaged in large polyethylene HICs. The radioactivity contents of the shipping containers are listed in Table 11.4-7 for both expected and maximum conditions.

Class A dewatered wastes may be packaged in HICs, strong, tight containers or other suitable containers. All wet wastes are consistent with the receiving burial site maximum free liquid criteria.

Compressible dry waste is packaged in strong, tight containers. Noncompressible dry wastes are packaged in strong, tight containers or other suitable containers. All containers comply with applicable portions of 10CFR71, 49CFR170 through 49CFR178, and the receiving burial site container requirements, when shipped for direct burial.

11.4.2.4.1 HIC Service The HIC is designed to contain concentrated radioactive waste materials generated during the removal of corrosion and activation products from plant process systems. The specific activity of these materials will range from trace quantities to no greater than 350 Ci/cm3. These values of specific activity include only radionuclides whose half-lives are greater than 5 years. In this specific HIC application, the predominant radionuclides are Co-60, Cs-137, Ni-63, Zn-65, Cr-51, and Mn-54.

The containerized waste form has three main constituents other than the corrosion activation products. These are:

a. Fibrous filter aid material derived from wood pulp fibers (cellulose) - this material is stable with respect to gas evolution in both acidic and basic solutions. In highly acidic solutions, cellulose will liquefy. High levels of ionizing radiation cause a break in the long molecular chains (C6H10O5) with no significant loss of weight.

This material is nonprotein and highly resistant to micro-organism ingestion.

CHAPTER 11 11.4-7 REV. 18, SEPTEMBER 2016

LGS UFSAR

b. Ion Exchange Bead Resin - This material consists of cation and anion exchange resins in bead form (O.4 mm to O.5 mm diameter spheres). The individual resin beads are formed from sulfonated polystyrene. The material exhibits chemical stability in pH ranges of O to 14.

c. Ion Exchange Resin Powder - This is identical to the above material; however, it is crushed to powder prior to use.

11.4.2.4.2 HIC Material The container is to be molded polyethylene. The suitability of polyethylene for this particular application relies on its material characteristics: outstanding dielectric properties, excellent chemical resistance to solvents, acids, and alkalies, toughness, good barrier properties, high environmental stress cracking resistance, good cold impact strength, ultraviolet light stability, and radiation resistance.

11.4.2.4.3 HIC Integrity Polyethylene is an extremely corrosion-resistant material with a high degree of chemical resistance both to the contents of the HIC and the earthen environment in which it will be buried.

Testing will confirm the HIC's ability to withstand vibration, drop, compression, puncture, and pressure tests. Each container receives a variety of quality control checks to confirm individual HIC integrity.

The HIC is designed to maintain its physical integrity for 10 half-lives of the longest lived significant isotope. For routine resin wastes, this is Cs-137 which has a 30 year half-life.

Therefore, the lifetime of this HIC is 300 years.

Concrete overpacks are used at the burial site to provide the compressive strength required for burial.

11.4.2.5 Storage Facilities Storage is provided in storage bays for the HICs. Each HIC is located in its own shielded cubicle with a removable plug on top. There are 11 low specific activity and 12 high specific activity storage cubicles. At the expected waste generation rates presented in Table 11.4-5, the low specific activity storage bay provides a minimum storage capability of 30 days and the high specific activity storage bay provides a minimum storage capability of six months.

The storage compartments and process cell areas where the HICs are handled are equipped with floor drains for washdown of any spillage that may occur.

Compressible and other dry wastes are expected to be of low activity, and the strong, tight containers and other suitable containers will be stored in appropriately controlled unshielded areas throughout the plant before shipment. Storage area is available for at least one truckload of waste.

The general arrangements of the solid radwaste process cells, storage, and shipping areas are shown in Section 1.2.

CHAPTER 11 11.4-8 REV. 18, SEPTEMBER 2016

LGS UFSAR Sources of radioactivity not stored inside the plant structures are discussed in Section 12.2.1.7.

11.

4.3 REFERENCES

11.4-1 Letter from J.S. Kemper (PECo) to A. Schwencer (NRC), "Process Control Program,"

(November 30, 1983).

11.4-2 NRC letter dated May 31, 2011,

Subject:

Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments, Re: Storage of Low-Level Radioactive Waste Produced at Limerick Generating Station (TAC Nos. ME3092 and ME3093).

CHAPTER 11 11.4-9 REV. 18, SEPTEMBER 2016

LGS UFSAR Table 11.4-1 SOLID WASTE MANAGEMENT SYSTEM TANK DESIGN PARAMETERS DESIGN DESIGN PRESSURE TEMPERATURE CAPACITY, EACH TANK QUANTITY (psig) (oF) TYPE MATERIAL (gal)

RWCU backwash 2 Atmos 212 Vert cyl Stainless 3,000 receiving Tank steel RWCU phase 2 Atmos 212 Vert cyl Stainless 7,500 separator steel Waste sludge tank 1 Atmos 212 Horiz cyl Carbon steel 16,000 Equipment drain spent 1 Atmos 212 Vert cyl Stainless 1,700 resin tank(2) steel Floor drain spent 1 Atmos 212 Vert cyl Stainless 1,700 resin tank(2) steel Condensate backwash 2 Atmos 212 Horiz cyl Carbon steel 20,000 receiving tank Condensate phase 4 Atmos 212 Vert cyl Carbon steel 15,000 separator (1) Deleted (2) The floor and equipment drain spent resin tanks are not used for normal plant operation (Section 11.4.2.1.1).

CHAPTER 11 11.4-10 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-2 SOLID WASTE MANAGEMENT SYSTEM PUMP DESIGN PARAMETERS DESIGN PRESSURE RATED FLOW RATED HEAD, TDH RATED POWER TEMPERATURE PUMP QUANTITY TYPE (gpm) (ft) (hp) (psig/F)

RWCU decant pump 1 Vert in-line 50 150 7.5 150/40-140 centrifugal RWCU sludge 1 Vert in-line 200 220 20 150/40-140 discharge mixing pump centrifugal Waste sludge 1 Vert in-line 200 219 25 150/40-140 discharge mixing pump centrifugal Equipment drain spent 1 Vert in-line 200 100 10 150/40-140 resin pump(2) centrifugal Floor drain spent 1 Vert in-line 200 100 10 150/40-140 resin pump(2) centrifugal Condensate backwash 4 Vert in-line 450 20 7.5 150/40-140 transfer pumps centrifugal Condensate decant 2 Vert in-line 450 100 25 150/40-140 pump centrifugal Condensate sludge 2 Vert in-line 400 224 40 150/40-140 discharge mixing pump centrifugal (1) Deleted (2) The floor and equipment drain spent resin pumps are not be used for normal plant operation (Section 11.4.2.1.1).

CHAPTER 11 11.4-11 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-3 SOLID WASTE MANAGEMENT SYSTEM PROCESS EQUIPMENT DESIGN PARAMETERS CENTRIFUGES Quantity 2 Type Solid bowl, horizontal, continuous solids discharge Rating Design feed rate 20 gpm Maximum feed rate 30 gpm Maximum solids capacity 750 lb/hr Centrifuge speed, bowl 3250 rpm conveyor 3224 rpm Developed G-forces 2050 Efficiency 98%

Drivers Type Ac motor, 460 V, 60 Hz, 3-phase Motor power rating 25 hp Nominal speed 1800 rpm SAMPLING EQUIPMENT Discharge Chute Resin Sampler Quantity 2 Type Pneumatic piston Sample Trolley Quantity 2 Driver Type Ac motor, 460 V, 60 Hz, 3-phase Motor power rating 1/2 hp Sample Station Quantity 1 Type Enclosed (with fume hood)

CHAPTER 11 11.4-12 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-3 (Cont'd)

DECONTAMINATION STATION Washdown Mechanism High pressure water Pump Type Single-stage centrifugal Rating Flow 30 gpm Head 1265' Temperature 40F-80F Driver Type Ac motor, 460 V, 60 Hz, 3-phase Motor power rating 40 hp Speed 3600 rpm Drying Mechanism Air blast nozzle system inside Flow 250 cfm Temperature 70F Radiation Monitoring Type GM tube Range 1-1x106 mRem/hr Swiping Mechanism Type Remote manipulator Container surface Coverage 50% minimum Detector type GM tube FILL STATION Discharge Chute Quantity 2 Material Stainless steel Telescopic Fill Chute Quantity 2 Material Stainless steel Positioning mechanism Pneumatic cylinder CHAPTER 11 11.4-13 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-3 (Cont'd)

Drip Pot/Sluice Pan Quantity 2 Material Stainless steel Positioning mechanism Pneumatic cylinder Radiation Monitor Quantity 2 Type GM tube Range Surface: to 100 R/hr 3': to 10 R/hr Swiping Mechanism Type Pole Capping Machine Quantity 2 Type of Cap Screw-on Cap Drive Type Air motor Positioning mechanism Pneumatic cylinder TRANSFER TABLES Large Transfer Tables Quantity 2 Capacity 8 tons Travel 14' Platform size 6.5'x6.5' Velocity 15 fpm CHAPTER 11 11.4-14 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-3 (Cont'd)

Vibrator Table Motors Quantity 4 Type Ac motor, 460 V, 60 Hz, 3-phase Motor power rating 6 hp Speed 1200 rpm HIC Platform Vacuum Pumps Quantity 2 Pump Type Rotary Driver Type Ac motor, 460 V, 60 Hz, 3-phase Motor power rating 11/2 hp OVERHEAD CRANE Quantity 1 Type Bridge Capacity 20 tons Maximum speed 50 fpm CHAPTER 11 11.4-15 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-4 SOLID WASTE MANAGEMENT SYSTEM FLOWS AVERAGE BATCH FREQUENCY FOR MAXIMUM BATCH NORMAL MAXIMUM NORMAL OPERATION FREQUENCY FOR VOLUME PER ACTIVITY ACTIVITY STREAM OF BOTH UNITS ONE UNIT(2) BATCH FLOW RATE CONCENTRATION CONCENTRATION NO.(1) (no. batches/no. days) (1/day) (gal) (gpm) (Ci/cc) (Ci/cc) 21 2/6.8 2/1 1100 By gravity 36.5 1110 22 4/6.8 2/1 1100 By gravity 35.8 1100 23 1/60 - 5600 20 20.6 91.8 24 7/10 4/1 9000 By gravity 1.1 26.6 25 7/10 4/1 9000 450 1.0 26.4 26 1/14.3 - 13000 20 0.7 5.1 27 1/125.8 - 1500 By gravity 0.26 4.61 28A(3)(5) 1/0.69 - 1925 By gravity 0.020 0.74 28B(3) 1/1.1 - 1925 By gravity 0.189 6.66 28C(3) 1/5 - 1925 By gravity 0.098 2.61 29 1/25.8 - 1500 By gravity 0.026 0.66 30(5)(6) 1/2.5 - 12800 20 0.062 2.27 31A(4) 1/60 - 51 ft3 By gravity 299 1340 31B(4) 1/14.3 - 235 ft3 By gravity 5.07 37.4 31C(4)(5)(6 1/2.5 - 27 ft3 By gravity 4.00 145 (1) See Figure 11.4-3 for the location of stream numbers.

(2) The maximum condition is assumed to happen 30 days per year per unit for the RWCU system and condensate filter/demineralizer system.

(3) 28A is the floor drain filter backwash.

28B is the equipment drain filter backwash.

28C is the fuel pool cleanup filter backwash.

(4) 31A is the RWCU sludge.

31B is the condensate sludge.

31C is the waste sludge.

(5) Batch frequencies, volumes, and activity concentrations are based on chemical waste processing via the floor drain subsystem (Section 11.2.2.1.3).

(6) Activity concentrations assume that exhausted bead resins from the liquid radwaste system are sluiced directly to the waste sludge tank (Section 11.4.2.1.1).

CHAPTER 11 11.4-16 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-5 WET WASTE INPUT TO THE SOLID WASTE MANAGEMENT SYSTEM(7)

NORMAL EXPECTED NORMAL BATCH EXPECTED MAXIMUM ANNUAL BATCH SIZE FREQUENCY ACTIVITY ACTIVITY VOLUME SOURCE WASTE TYPE (ft3) (days) (Ci/ft3) (Ci/ft3) (ft3)

RWCU phase Dewatered spent 51 60 8.5 037.9 424(2) separators powdered resin(1)

Condensate Dewatered spent 235 14 0.14 1.06 7670(2) phase separators powdered resin(1)/filtered solids Waste Dewatered spent resin 27 2.5 0.11 4.12 4622(4) sludge tank(5)(6) beads and powdered resin(1)(3)

(1) Density: 45 lb/ft3 .

(2) Assume 10% of operating time generating maximum input (Table 11.4-4) and 85% plant capacity factor.

(3) From radwaste filter/demineralizer and deep-bed demineralizer.

(4) Add 20% to account for abnormal input.

(5) Batch frequency, activities, and annual volume are based on chemical waste processing via the floor drain subsystem (Section 11.2.2.1.3).

(6) Activity concentrations assume that exhausted bead resins from the liquid radwaste system are sluiced directly to the waste sludge tank (Section 11.4.2.1.1).

(7) Values are historical data based on original plant design estimates.

CHAPTER 11 11.4-17 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-6 DRY WASTE FROM FILTERS TO THE SOLID WASTE MANAGEMENT SYSTEM EXPECTED(3) EXPECTED ANNUAL(3)

SIZE PER UNIT (in) NUMBER/WEIGHT VOLUME UNCOMPACTED SOURCE EQUIPMENT NOS WASTE TYPE OR WEIGHT PER UNIT OF FILTER PER YEAR (ft3)

Charcoal Offgas Treatment System(1)

HEPA (outlet) 10-F371 Glass fiber - steel 11x8 dia 1 1 20-F371 frame Control Structure -

Standby Gas Treatment System HEPA 0A-F170 Glass fiber - steel 24x24x12 12 48 0B-F170 frame Charcoal 0A-F169 Bulk activated 1200 lb 2400 lb 80(2) 0B-F169 charcoal HEPA 0A-F183 Glass fiber - steel 24x24x12 6 24 0B-F183 frame Control Structure -

Emergency Fresh Air Intake System Pre 0A-F159 Glass fiber 24x24x12 3 12 0B-F159 particle board frame HEPA 0A-F160 Glass fiber - steel 24x24x12 3 12 0B-F160 frame Charcoal 0A-F161 Bulk activated 770 lb 770 lb 26 0B-F161 charcoal HEPA 0A-F162 Glass fiber - steel 24x24x12 3 12 0B-F162 frame CHAPTER 11 11.4-18 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-6 (Cont'd)

EXPECTED(3) EXPECTED ANNUAL(3)

SIZE PER UNIT (in) NUMBER/WEIGHT VOLUME UNCOMPACTED SOURCE EQUIPMENT NOS WASTE TYPE OR WEIGHT PER UNIT OF FILTER PER YEAR (ft3)

Reactor Enclosure - Recirculation System Pre 1A-F259 Glass fiber, 24x24x12 80 320 1B-F259 particle board 2A-F259 frame 2B-F259 HEPA 1A-F260 Glass fiber - steel 24x24x12 80 320 1B-F260 frame 2A-F260 2B-F260 Charcoal 1A-F261 Bulk activated 12,900 lb 12,900 lb 430 1B-F261 charcoal 2A-F261 2B-F261 HEPA 1A-F262 Glass fiber - steel 24x24x12 40 160 1B-F262 frame 2A-F262 2B-F262 Turbine Enclosure -Equipment Compartment Exhaust Filters Pre 1A-F178 Glass fiber - steel 24x24x12 120 480 1B-F178 frame 2A-F178 2B-F178 HEPA 1A-F194 Glass fiber - steel 24x24x12 120 480 1B-F194 frame 2A-F194 2B-F194 Charcoal 1A-F179 Bulk activated 21,900 lb 43,800 lb 1460 1B-F179 charcoal 2A-F179 2B-F179 HEPA 1A-F157 Glass fiber - steel 24x24x12 60 240 1B-F157 frame 2A-F157 2B-F157 Turbine Enclosure - Condensate Backwash Receiving Tank Vent Pre 10-F166 Glass fiber - steel 24x24x12 (U2) 8 31 20-F166 frame 24x24x11.5 (U1)

HEPA 10-F189 Glass fiber - steel 24x24x12 (U2) 8 31 20-F189 frame 24x24x11.5 (U1)

CHAPTER 11 11.4-19 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-6 (Cont'd)

EXPECTED(3) EXPECTED ANNUAL(3)

SIZE PER UNIT (in) NUMBER/WEIGHT VOLUME UNCOMPACTED SOURCE EQUIPMENT NOS WASTE TYPE OR WEIGHT PER UNIT OF FILTER PER YEAR (ft3)

Control Structure -

Standby Gas Treatment Room Pre 0A-F196 Glass fiber - steel 24x24x12 4 16 0B-F196 frame HEPA 0A-F180 Glass fiber - steel 24x24x12 4 16 0B-F180 frame Charcoal 0A-F181 Tray type 24x24x4 4 8 HEPA 0B-F181 Glass fiber - steel 24x24x12 2 8 0A-F184 frame 0B-F184 Turbine Enclosure -

Condensate Backwash Area Floor Drain Sump Vent Pre 10-F172 Glass fiber - steel 8x8x6 12 3 20-F172 frame HEPA 10-F185 Glass fiber - steel 8x8x6 12 3 20-F185 frame Turbine Enclosure -

Condensate Backwash Area Equipment Drain Sump Vent Pre 10-F173 Glass fiber - steel 8x8x6 12 3 20-F173 frame HEPA 10-F186 Glass fiber - steel 8x8x6 12 3 20-F186 frame CHAPTER 11 11.4-20 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-6 (Cont'd)

EXPECTED(3) EXPECTED ANNUAL(3)

SIZE PER UNIT (in) NUMBER/WEIGHT VOLUME UNCOMPACTED SOURCE EQUIPMENT NOS WASTE TYPE OR WEIGHT PER UNIT OF FILTER PER YEAR (ft3)

Turbine Enclosure - Condensate Pump Area Floor Drain Sump Vent Pre 10-F174 Glass fiber - steel 8x8x6 12 6 20-F174 frame HEPA 10-F187 Glass fiber - steel 8x8x6 12 6 20-F187 frame Turbine Enclosure - Condensate Pump Area Equipment Drain Sump Vent Pre 10-F175 Glass fiber - steel 8x8x6 12 6 20-F175 frame HEPA 10-F188 Glass fiber - steel 8x8x6 12 6 20-F188 frame Reactor Enclosure - Equipment Compartment Exhaust Pre 1A-F254 Glass fiber - steel 24x24x12 140 560 1B-F254 frame 2A-F254 2B-F254 HEPA 1A-F255 Glass fiber - steel 24x24x12 140 560 1B-F255 frame 2A-F255 2B-F255 Charcoal 1A-F257 Bulk activated 21,900 lb 43,800 lb 1460 1B-F257 charcoal 2A-F257 2B-F257 HEPA 1A-F258 Glass fiber - steel 24x24x12 70 280 1B-F258 frame 2A-F258 2B-F258 Reactor Enclosure - RWCU Backwash Receiving Tank Vent Pre 10-F256 Glass fiber - steel 24x24x12 12 48 20-F256 frame HEPA 10-F263 Glass fiber - steel 24x24x12 12 48 20-F263 frame CHAPTER 11 11.4-21 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-6 (Cont'd)

EXPECTED(3) EXPECTED ANNUAL(3)

SIZE PER UNIT (in) NUMBER/WEIGHT VOLUME UNCOMPACTED SOURCE EQUIPMENT NOS WASTE TYPE OR WEIGHT PER UNIT OF FILTER PER YEAR (ft3)

Radwaste Enclosure -

Equipment Compartment Exhaust Pre 0A-F354 Glass fiber - steel 24x24x12 96 386 0B-F354 frame HEPA 0A-F355 Glass fiber - steel 24x24x12 96 386 0B-F355 frame Radwaste Enclosure -

Service and Control Area Exhaust Pre 0A-F356 Glass fiber - steel 24x24x12 12 48 0B-F356 frame HEPA 0A-F357 Glass fiber - steel 24x24x12 12 48 0B-F357 frame Radwaste Enclosure -

Centrifuge Room Exhaust Pre 0A-F360 Glass fiber - steel 24x24x12 12 48 0B-F360 frame HEPA 0A-F383 Glass fiber - steel 24x24x12 12 48 0B-F383 frame Radwaste Enclosure -

Laundry Room Exhaust Pre 00-F361 Glass fiber - steel 24x24x12 2 8 frame HEPA 00-F375 Glass fiber - steel 24x24x12 2 8 frame CHAPTER 11 11.4-22 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-6 (Cont'd)

EXPECTED(3) EXPECTED ANNUAL(3)

SIZE PER UNIT (in) NUMBER/WEIGHT VOLUME UNCOMPACTED SOURCE EQUIPMENT NOS WASTE TYPE OR WEIGHT PER UNIT OF FILTER PER YEAR (ft3)

Radwaste Enclosure -

Handling and Storage Area Exhaust Pre 00-F366 Glass fiber - steel 24x24x12 4 16 frame HEPA 00-F374 Glass fiber - steel 24x24x12 4 16 frame Radwaste Enclosure -

Common Tank Vent HEPA 00-F384 Glass fiber - steel 24x24x12 6 12 frame Radwaste Enclosure -

Centrifuge Vent Pre 0A-F367 Glass fiber - steel 12x12x6 12 6 0B-F367 frame HEPA 0A-F380 Glass fiber - steel 12x12x6 12 6 0B-F380 frame Radwaste Enclosure -

Floor and Equipment Drain Sump Vent HEPA 00-F368 Glass fiber - steel 2/8x8x6 12 3 00-F369 frame ______

Total 8345 (1) The charcoal guard beds are not expected to be replaced during the life of the plant.

(2) Charcoal density: 30 lb/ft3.

(3) Expected volumes are based on original plant design estimates.

CHAPTER 11 11.4-23 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-7 EXPECTED AND MAXIMUM SHIPPING CASK INVENTORIES(Ci)(1) (5)

EXPECTED(2) MAXIMUM Low High Low High Specific Specific Specific Specific Isotope Activities(4) Activities Activities(4) Activities Br-83 none none - -

Br-84 none none - -

Br-85 none none - -

I-131 1.33E+00 4.56E-04 1.19x101 4.11x10-3 I-132 3.81E-05 none 6.52x10 I-133 5.13E-06 none 8.35x10-5 -

I-134 none none - -

I-135 none none - -

Rb-89 none none - -

Cs-134 4.09E-02 4.98E+00 7.54x10-1 9.18x101 Cs-136 1.95E-03 3.94E-04 3.70x10-2 7.52x10-3 Cs-137 9.84E-02 1.35E+01 1.18 1.61x102 Cs-138 none none - -

Na-24 1.62E-08 none 3.72x10-9 -

P-32 2.90E-02 8.75E-03 2.87x10-3 8.67x10-4 Cr-51 2.05E+00 1.18E+01 2.02x10-1 1.17 Mn-54 7.92E-02 7.87E+00 5.24x10-2 5.19 Mn-56 none none - -

Fe-55 1.32E+00 1.72E+02 - -

Fe-59 1.98E-02 3.71E-01 5.17x10-2 9.76x10-1 Co-58 1.76E-01 6.80E+00 4.37 1.69x102 Co-60 5.84E-01 7.34E+01 7.24x10-1 9.07x101 Ni-63 1.49E-03 1.96E-01 - -

Ni-65 none none - -

Cu-64 none none - -

Zn-65 2.55E-01 2.35E+01 2.52x10-3 2.32x10-1 Zn-69 none none - -

Zn-69m none none - -

Sr-89 7.26E-02 1.72E-02 7.75 1.84x102 Sr-90 8.49E-03 1.16E+00 1.18 1.53x102 Sr-91 none none - -

Sr-92 none none - -

Y-91 5.20E-02 1.55E+00 1.24 3.85x101 Y-92 none none - -

Y-93 none none - -

Zr-95 5.85E-03 1.99E-01 1.16x10-3 3.92 Zr-97 none none - -

CHAPTER 11 11.4-24 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-7 (Cont'd)

EXPECTED(2) MAXIMUM Low High Low High Specific Specific Specific Specific Isotope Activities(4) Activities Activities(4) Activities Nb-95 8.46E-03 3.88E-01 1.70x10-1 7.71 Nb-98 none none - -

Mo-99 3.89E-03 none 1.48x10-1 -

Tc-99m 3.71E-03 none 1.42x10 Tc-101 none none - -

Tc-104 none none - -

Ru-103 1.18E-02 1.72E-01 3.89x10-2 5.66x10-1 Ru-105 none none - -

Ru-106 4.05E-03 4.19E-01 1.21x10-2 1.25 Ag-110m 1.29E-03 1.19E-01 7.62x10-2 7.08 Te-129m 2.04E-02 2.02E-01 7.02x10-2 6.98x10-1 Te-131m 1.36E-06 none - -

Te-132 3.71E-05 none 6.34x10-1 -

Ba-139 none none - -

Ba-140 4.65E-02 6.83E-03 3.62 5.31x10-1 Ba-141 none none - -

Ba-142 none none - -

La-142 none none - -

Ce-141 1.61E-02 1.49E-01 1.73x10-1 1.65 Ce-143 9.24E-08 none 3.83x10-6 -

Ce-144 3.94E-03 3.79E-01 1.58x10-1 1.53x101 Pr-143 5.61E-03 1.25E-03 1.91x10-2 4.22x10-3 Nd-147 2.53E-01 1.17E-05 4.07x10-2 1.89x10-3 W-187 3.99E-07 none 4.06x10-6 -

Np-239 6.61E-03 none 7.90x10-1 -

Other(3) 1.94E-01 1.51E+01 6.70 3.23x102 TOTAL 6.44E+00 3.36E+02 4.29x101 1.26x103 (1)

Container inventories are based on the specific activity levels of the source waste as processed by the external processing station and assume a filled container storage period prior to shipment of 15 days for low specific activity waste and 90 days for high specific activity waste.

(2)

Low specific activity inventories are based on a prorated mixture of condensate sludge and waste sludge. High specific activity inventories are based on RWCU sludge.

Container inventories assume 90% fill.

CHAPTER 11 11.4-25 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-7 (Cont'd)

(3)

"Other" isotopes consist of daughter products resulting from radioactive decay of the influent isotopes during accumulation and storage periods.

(4)

The prorated portion of low specific activity due to waste sludge is based on direct sluicing of exhausted bead resins from the liquid radwaste system to the waste sludge tank.

(5)

Values are historical data based on original plant design estimates CHAPTER 11 11.4-26 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-8 EXPECTED RADIONUCLIDE INVENTORIES OF SOLID WASTE MANAGEMENT SYSTEM COMPONENTS (1)

CONDENSATE EQUIPMENT RWCU BACKWASH RWCU BACKWASH CONDENSATE FLOOR DRAIN(4) DRAIN(4) WASTE(2)(4)

NUCLIDE RECEIVING TANK PHASE SEPARATOR RECEIVING TANK PHASE SEPARATOR SPENT RESIN TANK SPENT RESIN TANK SLUDGE TANK 1Br-83 2.46E-01 8.97E-01 1.50E-01 9.56E-01 1.58E-04 2.50E-04 3.40E-03 Br-84 6.98E-02 1.81E-02 3.59E-02 1.62E-01 1.18E-05 1.90E-05 5.08E-04 Br-85 4.72E-04 1.86E-05 3.00E-03 2.06E-06 6.10E-08 9.81E-08 1.88E-05 I-131 1.91E+01 1.70E+02 1.59E+01 1.75E+02 1.59E-02 1.50E-01 3.55E-01 I-132 2.32E+00 8.43E+00 1.42E+00 8.98E+00 1.42E-03 2.32E-03 3.11E-02 I-133 1.69E+01 6.73E+01 1.08E+01 7.49E+01 2.82E-02 4.97E-02 5.09E-01 I-134 1.77E+00 5.44ED+00 9.93E-01 5.32E+00 4.51E-04 7.25E-04 1.46E-02 I-135 5.12E+00 1.98E+01 3.22E+00 2.17E+01 7.19E-03 1.04E-02 1.22E-01 Rb-89 2.62E-02 4.27E-02 5.36E-04 1.52E-03 7.54E-07 1.98E-05 1.09E-04 Cs-134 1.51E-01 5.21E+00 7.11E-03 1.41E-01 1.11E-03 4.29E-02 1.37E-03 Cs-136 8.54E-02 1.11E+00 3.72E-03 5.01E-02 3.99E-04 4.68E-03 8.36E-04 Cs-137 3.54E-01 1.24E+01 1.66E-02 3.33E-01 2.61E-03 1.06E-01 3.18E-03 Cs-138 1.42E-01 3.68E-01 3.65E-03 1.66E-02 6.74E-06 1.77E-04 5.75E-04 Na-24 5.69E+00 2.24E+01 1.81E-01 1.24E+00 9.83E-03 1.55E-02 1.70E-01 P-32 8.60E-01 1.15E+01 3.75E-02 5.13E-01 6.67E-04 1.07E- 51E-02 Cr-51 2.33E+01 4.63E+02 1.05E+00 1.69E+01 1.71E-02 5.05E-01 3.93E-01 Mn-54 3.01E-01 1.00E+01 1.41E-02 2.76E-01 2.11E-04 1.73E-02 4.89E-03 Mn-56 4.65E+00 1.69E+01 1.42E-01 9.10E-01 3.17E-03 4.99E-03 6.68E-02 Fe-55 5.06+00E 1.74E+02 2.39E-01 4.72E+00 none none none Fe-59 1.44E-01 3.52E+00 6.62E-03 1.15E-01 1.04E-04 4.54E-03 2.40E-03 Co-58 9.77E-01 2.69E+01 4.52E-02 8.29E-01 6.98E-04 3.83E-02 1.60E-02 Co-60 2.02E+00 7.07E+01 9.50E-02 1.89E+00 1.42E-03 1.31E-01 3.27E-02 Ni-63 5.06E-03 1.79E-01 2.38E-04 4.76E-03 3.54E-06 3.34E-04 8.18E-05 Ni-65 2.77E-02 1.01E-01 8.48E-04 5.42E-03 1.89E-05 2.96W-05 3.97E-04 Cu-64 1.64E+01 6.44E+01 5.28E-01 3.69E+00 none none none Zn-65 1.00E+00 3.27E+01 4.68E-02 9.13E+00 7.04E-04 5.59E-02 1.62E-02 Zn-69 7.13E-02 2.23E-01 2.01E-03 1.10E-02 3.09E-08 3.15E-05 5.88E-04 Sr-89 4.88E-01 1.24E+01 2.24E-02 3.99E-01 3.51E-04 1.63E-02 8.06E-03 Sr-90 3.04E-02 1.07E+00 1.43E-03 2.86E-02 2.12E-05 1.99E-03 4.92E-04 Sr-91 1.59E+00 6.22E+00 5.03E-02 3.44E-01 2.60E-03 3.79E-03 4.39E-02 Sr-92 9.86E-01 3.62E+00 3.02E-02 1.94E-01 7.05E-04 1.10E-03 1.46E-02 Y-91 3.08E-01 8.37E+00 1.44E-02 2.63E-01 2.09E-04 1.12E-02 4.91E-03 Y-92 1.83E+00 7.08E+00 5.74E-02 3.90E-01 2.23E-03 3.34E-03 3.97E-02 Y-93 1.70E+00 6.66E+00 5.39E-02 3.68E-01 2.83E-03 4.11E-03 4.77E-02 Zr-95 3.42E-02 9.25E-01 1.57E-03 2.87E-02 2.44E-05 1.29E-03 5.63E-04 Zr-97 3.61E-03 1.43E-02 1.14E-04 7.93E-04 6.18E-06 1.02E-05 1.09E-04 Nb-95 3.53E-02 1.14E+00 1.65E-03 3.24E-02 2.47E-05 1.77E-03 5.71E-04 Nb-98 1.02E-01 3.12E-01 2.86E-03 1.52E-02 2.56E-05 4.11E-05 8.37E-04 CHAPTER 11 11.4-27 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-8 (Cont'd)

CONDENSATE EQUIPMENT RWCU BACKWASH RWCU BACKWASH CONDENSATE FLOOR DRAIN(4) DRAIN(4) WASTE(2)(4)

NUCLIDE RECEIVING TANK PHASE SEPARATOR RECEIVING TANK PHASE SEPARATOR SPENT RESIN TANK SPENT RESIN TANK SLUDGE TANK Mo-99 4.80E+00 2.35E+01 1.71E-01 1.31E+00 5.27E-03 1.97E-02 1.10E-01 Tc-99m 8.94E+00 3.91E+01 3.00E-03 2.16E+00 1.08E-02 2.71E-02 2.05E-01 Tc-101 4.16E-01 6.18E-01 8.11E-03 2.10E-02 4.08E-05 6.57E-05 3.19E-03 Tc-104 5.34E-01 9.90E-01 1.16E-02 3.76E-02 6.05E-05 9.72E-05 3.89E-03 Ru-103 9.52E-02 2.20E+00 4.35E-03 7.47E-02 6.91E-05 2.67E-03 1.59E-03 Ru-105 3.41E-01 1.30E+00 1.06E-02 7.06E-02 3.68E-04 5.50E-04 6.71E-03 Ru-106 1.51E-02 5.07E-01 7.08E-04 1.39E-02 1.06E-05 8.90E-04 2.45E-04 Ag-110m 5.01E-03 1.64E-01 2.35E-04 4.58E-03 3.52E-06 2.82E-04 8.14E-05 Te-129m 1.89E-01 4.12E+00 8.61E-03 1.44E-01 1.38E-04 4.78E-03 3.16E-03 Te-131m 1.26E-01 5.17E-01 4.13E-03 2.88E-02 1.89E-04 4.05E-04 3.61E-03 Te-132 2.63E-02 1.37E-01 9.69E-04 7.64E-03 2.74E-05 1.16E-04 5.82E-04 Ba-139 4.55E-01 1.54E+00 1.34E-02 7.93E-02 1.75E-04 2.82E-02 4.58E-03 Ba-140 1.68E+00 2.10E+01 7.32E-02 9.65E-01 3.00E-03 1.93E-05 3.00E-02 Ba-141 6.69E-02 1.23E-01 1.45E-03 4.71E-03 8.33E-07 1.21E-06 4.86E-04 Ba-142 1.89E-02 2.14E-02 3.22E-04 6.40E-04 1.84E-07 2.69E-04 1.59E-04 La-142 2.93E-01 1.02E+00 8.74E-03 5.28E-02 1.39E-05 2.02E-03 3.14E-03 Ce-141 1.53E-01 3.29E+00 7.00E-03 1.16E-01 3.56E-04 3.64E-04 2.52E-03 Ce-143 4.17E-02 1.71E-01 1.36E-03 9.56E-03 2.24E-05 1.36E-04 1.15E-04 Ce-144 1.50E-02 4.99E-01 7.06E-04 1.38E-02 4.22E-05 8.61E-04 2.45E-04 Pr-143 1.80E-01 2.38E+00 7.89E-03 1.07E-01 3.34E-04 2.19E-03 3.13E-03 Nd-147 1.22E-02 1.39E-01 5.28E-04 6.62E-03 2.05E-05 1.24E-04 2.21E-04 W-187 3.06E-01 1.22E+00 9.81E-03 6.81E-02 1.42E-04 9.27E-04 9.04E-03 Np-239 1.50E+01 6.92E+01 5.24E-01 3.86E+00 1.05E-02 5.74E-02 3.61E-01 OTHERS(3) 3.26E+00 3.36E+01 1.16E-01 1.53E+00 1.59E-02 1.01E-01 3.04E-01 Total 1.51E+02 1.44+03 3.65E+01 3.33E+02 1.48E-01 1.45E+00 2.98E+00 (1)

Activity inventories are given in curies.

(2)

Activity inventory is based on chemical waste processing via the floor drain subsystem (Section 11.2.2.1.3).

(3)

Activity of daughter products resulting from radioactive decay of the influent isotopes during the accumulation period.

(4)

The floor and equipment drain spent resin tanks are not used for normal plant operation. Exhausted bead resin from the liquid radwaste sytems will be sluiced directly to the waste sludge tank (Section 11.4.2.1.1). The listed waste sludge tank curie inventory represents activity levels of the floor drain, equipment drain, and fuel pool filter/demineralizer backwashes. To determine the waste sludge tank inventory including an exhausted resin bed, the listed isotopic inventory from the appropriate spent resin tank should be added directly to the listed waste sludge tank inventory.

CHAPTER 11 11.4-28 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-9 MAXIMUM RADIONUCLIDE INVENTORIES OF SOLID WASTE MANAGEMENT SYSTEM COMPONENTS (1)

CONDENSATE EQUIPMENT RWCU BACKWASH RWCU BACKWASH CONDENSATE FLOOR DRAIN(4) DRAIN(4) WASTE(2)(4)

NUCLIDE RECEIVING TANK PHASE SEPARATOR RECEIVING TANK PHASE SEPARATOR SPENT RESIN TANK SPENT RESIN TANK SLUDGE TANK Br-83 5.34 1.94x10+1 3.25 2.07x10+1 3.82x10-4 5.40x10-3 7.35x10-2 Br-84 1.78 4.61 9.16x10-1 4.15 3.32x10- 5 4.82x10-4 1.30x10-2 Br-85 1.31x10-2 5.15x10-4 8.41x10-4 5.70x10-5 1.85x10-7 2.71x10-6 5.17x10-4 I-131 1.72x10+2 1.55x10+3 1.43x10+2 1.59x10+3 2.49x10-1 1.35 3.19 I-132 4.89x10+2 2.51x10+3 3.59x10+2 2.79x10+3 3.94x10-1 2.07 1.23x10+1 I-133 2.78x10+2 1.10x10+3 1.79x10+2 1.25x10+3 1.18x10-1 8.12x10-1 8.31 I-134 2.82x10+1 8.64x10+1 1.58x10+1 8.45x10+1 7.89x10-4 1.15x10-2 2.32x10-1 I-135 1.32x10+2 5.11x10+2 8.30x10+1 5.61x10+2 2.50x10-2 2.69x10-1 3.16 Cs-134 2.79 9.62x10+1 1.31x10-1 2.60 2.04x10-2 7.91x10-1 2.49x10-2 Cs-136 1.63 2.13x10+1 7.10x10-2 9.58x10-1 7.61x10-3 8.92x10-2 1.58x10-2 Cs-137 4.20 1.48x10+2 1.97x10-1 3.95 3.10x10-2 1.25 3.74x10-2 Cs-138 1.27x10+1 3.31x10+1 3.28x10-1 1.49 6.04x10-4 1.59x10-2 5.16x10-2 Na-24 1.32 5.17 4.16x10-2 2.87x10-1 4.68x10-4 3.58x10-3 3.92x10-2 P-32 8.52x10+1 1.15 3.73x10-3 5.09x10-2 1.60x10-4 1.06x10-3 1.49x10-3 Cr-51 2.30 4.58x10+1 1.04x10-1 1.68 5.28x10-3 4.99x10-2 3.85x10-2 Mn-54 1.99x10-2 6.60 9.30x10-3 1.82x10-1 5.66x10-4 1.14x10-2 3.20x10-3 Mn-56 5.45 1.99x10+1 1.66x10-1 1.07 4.15x10-4 5.84x10-3 7.81x10-2 Fe-59 3.80x10-1 9.27 1.75x10-2 3.05x10-1 9.55x10-4 1.17x10-2 6.26x10-3 Co-58 2.42x10+ 6.69x10+2 1.12 2.05x10+1 6.41x10-2 9.48x10-1 3.95x10-1 Co-60 2.50 8.73x10+1 1.17x10- 2.34 7.27x10-3 4.01x10-1 4.01x10-2 Ni-63 - - - - - - -

Ni-65 3.25x10-2 1.19x10-1 9.94x10-3 6.36x10-3 2.47x10-6 3.47x10-5 4.65x10-4 Cu-64 - - - - - - -

Zn-65 9.91x10-3 3.25x10-1 4.64x10-4 9.04x10-3 2.81x10-5 5.54x10-4 1.60x10-4 Zn-69m 1.83x10-2 7.14x10-2 5.76x10-4 3.96x10-3 6.14x10-6 4.84x10-5 5.36x10-4 Sr-89 5.22x10+1 1.33x10+3 2.40 4.28x10+1 1.35x10-1 1.75 8.56x10-1 Sr-90 4.03 1.42x10+2 1.89x10-1 3.79 1.17x10-2 2.64x10-1 6.45x10-2 Sr-91 1.04x10+2 4.00x10+2 3.23 2.21x10+1 2.66x10-2 2.44x10-1 2.81 Sr-92 4.42x10+1 1.62x10+2 1.35 8.71 3.54x10-3 4.94x10-2 4.99x10-1 Y-91 - - - - 6.83x10-3 1.88x10-1 6.38x10-2 Y-92 - - - - 1.01x10-2 4.94x10-2 4.99x10-1 Y-93 - - - - - - -

Zr-95 6.76x10-1 1.83x10+1 3.12x10-2 5.69x10-1 1.77x10-3 2.54x10-2 1.10x10-2 Zr-97 8.36x10-2 3.30x10-1 2.69x10-3 1.80x10-2 3.22x10-5 2.35x10-4 2.51x10-3 Nb-95 7.32x10-1 2.34x10+1 3.43x10-2 6.68x10-1 2.08x10-3 3.60x10-2 1.18x10-2 Nb-98 - - - - - - -

Mo-99 1.86x10+2 9.04x10+2 6.66 5.04x10+1 1.43x10-1 7.58x10-1 4.23 Tc-99m 4.15x10+2 1.78x10+3 1.38x10+1 9.85x10+2 1.78x10-1 1.18 9.50 CHAPTER 11 11.4-29 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-9 (Cont'd)

CONDENSATE EQUIPMENT RWCU BACKWASH RWCU BACKWASH CONDENSATE FLOOR DRAIN(4) DRAIN(4) WASTE(2)(4)

NUCLIDE RECEIVING TANK PHASE SEPARATOR RECEIVING TANK PHASE SEPARATOR SPENT RESIN TANK SPENT RESIN TANK SLUDGE TANK Tc-101 3.08 4.58 6.01x10-2 1.56x10-1 3.33x10-5 4.86x10-4 2.36x10-2 Tc-104 - - - - - - -

Ru-103 3.14x10-1 7.26 1.43x10-2 2.46x10-1 7.70x10-4 8.79x10-3 5.18x10-3 Ru-105 - - - - - - -

Ru-106 4.52x10-2 1.52 2.12x10-3 4.17x10-2 1.30x10-4 2.67x10-3 7.27x10-4 Ag-110m 2.97x10-1 9.78 1.39x10-2 2.72x10-1 8.44x10+4 1.67x10-2 4.79x10-2 Te-129m 6.54x10-1 1.43x10+1 2.98x10-2 5.00x10-1 1.57x10-3 1.65x10-2 1.09x10-1 Te-131m - - - - - - -

Te-132 4.53x10+2 2.36x10+3 1.67x10+1 1.32x10+1 3.81x10-1 1.99 9.95 Ba-139 3.15x10+1 1.07x10+2 9.25x10+1 5.48 1.34x10-3 1.95x10-2 3.17x10-2 Ba-140 1.32x10+2 1.65x10+3 5.72 7.54x10+1 2.37x10-1 1.50 2.32 Ba-141 5.37 9.95 1.17x10-1 3.79x10-1 6.69x10-5 9.76x10-4 3.89x10-2 Ba-142 2.57 2.92 4.37x10-2 8.69x10-2 2.49x10-5 3.65x10-4 2.15x10-2 La-142 - - - - 2.49x10-5 3.65x10-4 2.15x10-2 Ce-141 1.67 3.64x10+1 7.64x10-2 1.28 3.68x10-3 3.13x10-2 2.25x10-2 Ce-143 1.74x10-1 7.11x10-1 5.70x10-3 3.96x10-2 9.38x10-5 5.62x10-4 4.78x10-3 Ce-144 6.10x10-1 2.02x10+1 2.86x10-2 5.59x10-1 1.74x10-3 3.49x10-2 9.82x10-3 Pr-143 6.01x10-1 8.00 2.64x10-2 3.59x10-1 1.13x10-3 7.36x10-3 1.03x10-3 Nd-147 2.00 2.25x10+1 8.56x10-2 1.07 3.37x10-3 2.01x10-2 3.56x10-2 W-187 3.14 1.27x10+1 1.01x10-1 7.10x10-1 1.45x10-3 9.47x10-3 9.20x10-2 Np-239 1.81x10+3 8.32x10+3 6.34x10+1 4.64x10+2 1.27 6.90 4.32x10-2 OTHERS(3) 2.21x10+2 2.47x10+3 8.89 1.09x10+2 4.02x10-1 3.24 6.35

+3 +4 +2 +3 +1 Total 4.64x10 2.68x10 9.06x10 7.31x10 3.75 2.62x10 1.09x10+2 (1)

Activity inventories are given in curies.

(2)

Activity inventory is based on chemical waste processing via the floor drain subsystem (Section 11.2.2.1.3).

(3)

Activity of daughter products resulting from radioactive decay of the influent isotopes during the accumulation period.

(4)

The floor and equipment drain spent resin tanks are not used for normal plant operation. Exhausted bead resin from the liquid radwaste systems will be sluiced directly to the waste sludge tank (Section 11.4.2.1.1). The listed waste sludge tank curie inventory represents activity levels of the floor drain, equipment drain, and fuel pool filter/demineralizer backwashes. To determine the waste sludge tank inventory including an exhausted resin bed, the listed isotopic inventory from the appropriate spent resin tank should be added directly to the listed waste sludge tank inventory.

CHAPTER 11 11.4-30 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.4-10 EXPECTED ANNUAL SOLID WASTE MANAGEMENT SYSTEM CONTAINERS TO BE SHIPPED AND ACTIVITY CONTENTS ANNUAL EXPECTED QUANTITY OF ACTIVITY CONTAINERS CONTAINER TYPE CAPACITY CONTENTS Ci/CONTAINER FOR SHIPMENT High integrity 63 ft3 Dewatered (1) (3)

HSA sludge 337 8 High integrity 120 ft3 Dewatered (1) (3)

LSA sludge 6.4 118 Steel drum 55 gal Compacted (2) dry waste 2x10-2 3886 Wood box 100 ft3 Uncompacted (2) dry waste 6x10-3 250 (1)

Assumes fill to 90% of container capacity.

(2)

Estimates based on operational data from PBAPS.

(3)

Values are historical data based on original plant design estimates.

CHAPTER 11 11.4-31 REV. 13, SEPTEMBER 2006

LGS UFSAR 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS The process and effluent radiological monitoring and sampling systems, including the primary containment radiation monitoring system, are contained in the PRMS. The PRMS is provided to furnish information to operations personnel regarding radioactivity levels in principal plant process and effluent streams to assist in maintaining radiation levels as low as reasonably achievable.

The system is also provided to verify compliance with applicable governmental regulations for the containment, control, and release of radioactivity in liquid and gaseous effluents generated as a result of normal or abnormal operation of the plant. The design objectives and criteria are primarily determined by the system designation of either:

a. Safety-related systems

b. Monitoring systems for plant operation, and operator information.

c. Postaccident monitoring and sampling systems The PRMS is composed of the following process and effluent radiation monitors:

a. Safety-related systems

1. Main steam line radiation monitors

2. Reactor enclosure ventilation exhaust radiation monitors

3. Refueling area ventilation exhaust radiation monitors

4. Control room ventilation radiation monitors

5. Control room emergency fresh air radiation monitors

6. RHR service water radiation monitors

b. For plant operation and operator information

1. North stack effluent radiation monitors

2. South stack effluent radiation monitors

3. Radwaste equipment rooms ventilation exhaust radiation monitor

4. Charcoal treatment system process exhaust radiation monitors

5. Recombiner room and hydrogen analyzer compartments exhaust radiation monitors

6. Steam exhauster discharge and vacuum pump exhaust radiation monitors

7. Radwaste enclosure and Chemistry Laboratory Expansion ventilation exhaust radiation monitor CHAPTER 11 11.5-1 REV. 19 SEPTEMBER 2018

LGS UFSAR

8. Air ejector offgas effluent radiation monitors

9. Primary containment leak detector radiation monitors

10. Hot maintenance shop ventilation exhaust radiation monitor

11. Liquid radwaste discharge radiation monitor

12. Service water radiation monitors

13. Reactor enclosure cooling water radiation monitors.

c. Postaccident systems

1. Primary containment post-LOCA radiation monitors

2. North stack wide range accident monitoring system

3. Postaccident sampling system 11.5.1 DESIGN BASES AND SPECIFIC REQUIREMENTS In general, the radiation monitoring systems are designed to measure and record radioactivity levels, to alarm on high radioactivity levels, and to prevent the release of radioactive liquids, gases, and particulates. The PRMS aids in protecting the general public and plant personnel from exposure to radiation or radioactive materials in excess of those limits allowed by the applicable regulations.

The main objectives of the PRMS for normal operation are as follows:

a. To provide surveillance of radioactivity levels in process and effluent streams from minimum detectable levels to levels commensurate with Technical Specification limits by indicating and recording these levels, by alarming at abnormal activity levels, and by initiating or causing the initiation of corrective action when applicable.

b. To provide data for estimating total released activity.

For some anticipated operational occurrences the PRMS activates necessary isolation or diversion valves, thereby terminating releases if radioactivity levels exceed pre-established setpoints.

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 reactor enclosure refueling area, if predetermined radiation levels are exceeded.

All radiological effluent monitors that actuate an ESF system have been dynamically qualified to the requirements of IEEE 344 (1971). See Section 3.10. In addition, all safety-related radiological effluent monitors that are located in a harsh environment have been environmentally qualified to the requirements of NUREG-0588, Category II.

CHAPTER 11 11.5-2 REV. 19 SEPTEMBER 2018

LGS UFSAR Radiological effluent monitors that actuate an ESF system were procured in accordance with the requirements of 10CFR50, Appendix B, as interpreted at the time of their procurement. The design of these monitors and their system is consistent with the quality assurance criteria of Regulatory Guide 1.143, and the factors influencing instrumentation selection identified in ANSI 13.10 (1974) were considered.

Information regarding the process and effluent monitoring system's compliance with Regulatory Guide 1.97 (Rev 2) is provided in Section 7.5. LGS will meet the guidance of Regulatory Guide 4.15 (Rev 1) for roof vent and liquid effluent monitors.

11.5.1.1 General Design Criteria Specific design criteria for monitoring systems are discussed under the system description when applicable. General criteria taken into consideration in the design of the PRMS are as follows:

a. The PRMS is designed to monitor pathways for release of radioactive materials to the environment in conformance with GDC 60, GDC 63, and GDC 64 and 10CFR50, Appendix I, and Regulatory Guide 1.21.

b. The PRMS provides early warning of increasing radioactivity levels indicative of equipment failure, filter failure, system malfunction, or deteriorating system performance by using a high alarm setpoint.

c. The PRMS initiates prompt corrective action, either automatically or through operator response, on high radioactivity level by using a high-high alarm setpoint.

d. Monitors and detectors are selected with sensitivities and ranges in accordance with radiation levels anticipated at specific detector locations.

e. In general, monitors register full-scale if exposed to radiation levels exceeding full-scale indication. Monitors that do not have this characteristic are designed such that the maximum levels from the DBA do not exceed their upper range with the exception of the north stack effluent normal range monitors when radiation levels exceed the range of the north stack effluent normal range monitors the wide range accident monitor provides indication.

f. Radioactivity levels are continuously indicated and recorded in the control room with the exception of liquid radwaste effluent activity (recorded in the radwaste control room), the PASS system radiation monitors (indicated locally), Steam Packing Exhauster/Mechanical Vacuum Pump discharge, and the hot maintenance shop ventilation exhaust activity (indicated via the RMDS computer locally). The control room air supply, emergency fresh air, hot maintenance shop ventilation exhaust activity, and north and south stack effluent activity are indicated and trended on demand in the control room or indicated locally.

g. Audible control room alarms are initiated on high radioactivity levels and on signal, circuit, or power failures.

h. Monitor components requiring calibration, maintenance and inspection are accessible, and spare equipment is commercially available.

CHAPTER 11 11.5-3 REV. 19 SEPTEMBER 2018

LGS UFSAR

i. Insofar as practical, self-monitoring of components is provided to the extent that power failure or component malfunction causes annunciation and channel trip.

j. Off-line (sampling-type) radiation monitoring systems are designed to be unaffected by background radiation levels of up to 2.5 mRem/hour. Off-line monitors are located in areas where the maximum background does not increase the minimum detectable concentration above the required monitor lower range.

Accident dose rates have been considered in locating monitors that are required for safety or postaccident. In-line radiation monitoring systems will generally indicate changes in background radiation levels. Environmental qualification of safety-related components is discussed in Section 3.11.

k. Safety-related components of the PRMS required for safe shutdown are protected against the effects of extreme winds, floods, tornadoes, or missiles by locating them in a structure designed to withstand such conditions (Chapter 3).

l. Safety-related monitors are designed to seismic requirements consistent with the seismic design of the system being monitored.

m. Independence of redundant monitors that are safety-related is maintained by providing adequate separation of detectors, signal cabling, power supplies, and actuation circuits for isolation and diversion valves to meet IEEE 279 criteria.

11.5.1.2 Basis for Detector Location Selection Normal and potential paths for release of radioactive material are selected for monitoring as follows:

a. Process lines that may discharge radioactive fluids to the environs, in order to indicate the radioactivity level and to alarm in the control room when established limits for the release of radioactive materials are reached or exceeded.

b. Process lines that do not discharge directly to the environs, in order to indicate possible process system malfunctions by detecting increases in radioactivity levels.

Monitored processes and detector locations are listed in Table 11.5-1 and shown in drawing M-26.

11.5.1.3 Expected Radiation Levels The expected radioactivity concentrations in the process and effluent streams are such that radiation levels at the site boundary are a small fraction of 10CFR20 limits and will be ALARA.

11.5.1.4 Quantity to be Measured The principal radionuclides to be monitored are indicated in Table 11.5-1. All channels measure gross radioactivity.

11.5.1.5 Detector Type, Sensitivity, and Range CHAPTER 11 11.5-4 REV. 19 SEPTEMBER 2018

LGS UFSAR The detectors are Geiger-Muller tubes, ionization chambers, or scintillation crystals that detect either beta or gamma radiation, depending on the application. In general, ion chambers are used in high intensity or high temperature applications, GM tubes are used for gross measurements, beta scintillators are used for relatively precise measurements of noble gases, and gamma scintillators are used for relatively precise measurement of iodines and particulates. The sensitivity and range are selected so that the alarm setpoint is at least an order of magnitude higher than the detector threshold and so that the instrument reads on scale during normal operation. Detector type, sensitivity, and nominal ranges of each process and effluent monitor are indicated in Table 11.5-1.

11.5.1.6 Setpoints Setpoints for effluent monitors are established to meet Offsite Dose Calculation Manual Limits as required by the Technical Specifications which encompass 10CFR20 limits and ALARA guidelines. Setpoints for process monitors are established to provide a warning of increased system activity and to initiate corrective action where appropriate. In all cases, the setpoints are established to maintain offsite radiological effects within applicable regulation limits.

Changing of the setpoints is under the administrative control of the plant manager or his authorized delegate.

Two independently adjustable radiation setpoints are provided for most monitors. The high setpoint normally activates only an alarm, while the high-high setpoint activates an alarm and initiates corrective action where appropriate. Setpoints are at least twice the background level if practicable to reduce the number of spurious trips. Radiation monitoring system setpoints and associated functions are provided in Table 11.5-2.

11.5.1.7 Annunciators and Alarms All process and effluent radiation monitors indicate and audibly alarm in the control enclosure as shown in Table 11.5-2.

An operator can acknowledge the alarm and silence the audible alarm, but he cannot clear the annunciator window until the alarm condition has been cleared at the PRMS panels located in the control enclosure. Radiation alarms can be cleared only if the indication is less than the setpoint.

At the PRMS panels in the control enclosure, the channel that alarmed and the type of alarm are determined by the lights associated with three types of alarms. These alarm lights are as follows:

a. A high alarm light illuminates or a display indication is provided when the radioactivity exceeds preset limits that have been selected to provide an early warning.

b. A high-high alarm light illuminates or a display indication is provided when the radioactivity exceeds preset limits that have been selected to initiate appropriate corrective action. The high-high alarm actuates the trip auxiliaries in those applications where a control trip is automatic.

c. A low (failure) alarm light illuminates or a display indication is provided when the meter reaches a downscale trip point which indicates that there is a detector CHAPTER 11 11.5-5 REV. 19 SEPTEMBER 2018

LGS UFSAR signal, circuit, or power failure. In certain cases, as shown in Table 11.5-2, this downscale alarm also actuates the trip auxiliaries.

11.5.1.8 Calibration, Maintenance, Inspection, Decontamination, and Replacement All instruments are calibrated upon installation. Calibration sources are provided for periodic recalibration of the detectors.

Purge capability is provided to all off-line instrument racks. In event of severe contamination, the modules in question or the entire rack may be transferred to the hot maintenance shop for decontamination.

A schedule for periodic replacement of instrument components, such as photomultiplier tubes and detectors, is based on manufacturer's recommendations. Instrument rack installation facilitates access for inspection, testing, maintenance, and repairs.

Components of the radiological monitoring systems are designed for convenient replacement.

The assemblies are designed for a minimum life of 40 years when maintained in accordance with the manufacturers' recommendations including periodic parts replacement.

11.5.2 SYSTEM DESCRIPTION Specific information on the PRMS is tabulated in Tables 11.5-1 and 11.5-2 and arrangements are shown in drawing M-26.

11.5.2.1 Systems Required for Safety 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.

The system consists of four redundant instrument channels. Each channel consists of a local detector (gamma-sensitive ion chamber) and a control room radiation monitor with an auxiliary trip unit. Power for two channels (A and C) is supplied from RPS bus A and 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 2-pen recorder allows the output of any two selected channels to be recorded.

The detectors are physically located near the main steam lines just downstream of the outboard main steam line isolation valves in the space between the primary containment and secondary containment walls. 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.

Table 11.5-1 lists the sensitivity and range of the detectors.

Each radiation monitor has four alarm circuits: two upscale (high-high and high), one downscale (low), and one inoperative. Each alarm is visually displayed on the PRMS panels located in the control room. A high-high or inoperative signal results in a high-high/inoperative alarm, mechanical vacuum pump shutdown and mechanical vacuum pump suction valve closure. A high alarm actuates a main steam line high radiation annunciator common to all channels. A CHAPTER 11 11.5-6 REV. 19 SEPTEMBER 2018

LGS UFSAR downscale alarm actuates a main steam line downscale annunciator common to all channels.

Each radiation monitor visually displays the measured radiation level.

The MSL-RMS is safety-related and is discussed in detail in Section 7.6.

11.5.2.1.2 Reactor Enclosure Ventilation Exhaust Radiation Monitoring System This system monitors the radiation level of the air in the reactor enclosure ventilation system exhaust duct prior to its discharge from the structure.

The system consists of four redundant instrument channels. Each channel consists of a local detection assembly (a sensor and converter unit containing a GM tube and electronics) and a radiation monitor. Power for two channels (A and C) is supplied from RPS bus A and 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. Two 2-pen recorders allow the output of the channels to be recorded continuously. The detection assemblies are located outside the exhaust duct upstream of the ventilation isolation valves.

Each radiation monitor has three alarm circuits: high-high, high, and low. Two high-high trips in channels A and B initiate closure of the reactor enclosure ventilation isolation valves and the startup of the RERS and SGTS train A. The same conditions for channels C and D initiate closure of the corresponding tandem isolation valves and startup of the RERS and SGTS train B.

Alternate inboard and outboard valves are actuated independently by the two sets of trip channels. The same logic configuration of the four channels also initiates closure of the primary containment purge and vent valves.

An upscale trip is visually displayed on the affected radiation monitor and actuates a reactor enclosure ventilation high-high radiation annunciator common to all channels. A downscale trip is also visually displayed on the radiation monitor and actuates an annunciator common to all channels. An additional alarm for high radiation is provided by the recorder and actuates a common reactor enclosure ventilation high radiation annunciator. Each radiation monitor visually displays the measured radiation level.

The location and monitoring characteristics of the reactor enclosure ventilation radiation monitoring channels are adequate to provide detection capability for abnormal amounts of radioactivity in the reactor enclosure ventilation exhaust and to initiate isolation. The redundancy and arrangement of channels are sufficient to ensure that no single failure can prevent isolation when required. The upscale trips meet the design requirements of IEEE 279 (1971) as described in Section 7.3.2.2.

11.5.2.1.3 Refueling Area Ventilation Exhaust Radiation Monitoring System This system monitors the radiation level in the ventilation exhaust duct from the refueling area, including the area over the fuel pool. The monitoring system is identical to the REVE-RMS with the same channel trip logic and protective action initiation, with the exception of the initiation of the RERS.

During refueling operation (including criticality tests), the monitoring system acts as an emergency safety feature against the consequences of a refueling or control rod-drop accident. The response of the REVE-RMS to the refueling accident is discussed in Chapter 15.

CHAPTER 11 11.5-7 REV. 19 SEPTEMBER 2018

LGS UFSAR 11.5.2.1.4 Control Room Ventilation Radiation Monitoring System This system monitors the radiation level in the supply air to the control room. No measurable activity is expected to be present; however, in the event of a DBA, fission gases could escape from the plant structures and be drawn into the supply air intake. In the Control Room Radiation Isolation Mode HEPA/charcoal filter will remove radioactive particulates and iodines.

There are four independent monitors, separated in accordance with IEEE 279 (1971), that monitor air inside the control enclosure intake duct. These in-line monitors respond to the gross radioactivity in the vicinity of the detectors. Each monitor provides three alarm conditions:

low, high, and high-high. The low, high, and high-high alarms trip the control room annunciator.

The high-high alarm trips the control room fresh air isolation valves and starts the CREFAS, which provides for the filtration of the incoming air through HEPA/charcoal filters. The trip of any monitor A, B, C, or D shuts off the control room fresh air supply, and the trip of either monitors A and C, or B and D, starts the CREFAS. (See Section 6.4 for a more detailed discussion of control room isolation on detection of high radiation.)

11.5.2.1.5 Control Room Emergency Fresh Air Radiation Monitoring System Upon initiation of the CREFAS system, this monitor indicates radioactivity concentration levels downstream of control room ventilation HEPA/charcoal filters. Radioactive noble gas concentration is measured. These in-line monitors detect gross radiation only. Two monitors, separated in accordance with IEEE 279 (1971), monitor sample air from the control room emergency fresh air duct.

Each monitor provides three alarm conditions: low, high, and high-high. These alarms trip annunciators in the control room.

11.5.2.1.6 RHR Service Water Radiation Monitoring System This system is comprised of two monitors for sampling the combined RHRSW loop return flows to the spray pond or cooling tower. One monitor serves Loop A, the other serves Loop B.

Each monitor provides three alarm conditions: high-high/INOP/DNSCL, high radiation, and high/low sample pump flow. These signals trip annunciators in the control room, and the high-high/INOP/DNSCL trips the RHRSW pumps. Should a monitor failure occur, grab samples are used to measure radioactivity in the fluid.

These monitors are qualified as IEEE Class 1E.

The monitors for sampling the combined RHRSW loop return to the spray pond or cooling tower are safety-related.

11.5.2.2 Systems Required for Plant Operation 11.5.2.2.1 North Stack Effluent Radiation Monitoring System The NSE-RMS is comprised of two subsystems:

a. North stack effluent monitors for normal plant operation

b. Wide range accident monitoring subsystem CHAPTER 11 11.5-8 REV. 19 SEPTEMBER 2018

LGS UFSAR The objectives of the normal plant operation subsystem are to indicate whether the limits of actual release of radioactive material to the environs are reached or exceeded, and to measure the quantity of release of radioactive material during normal plant operation, in compliance with 10CFR50 and Regulatory Guide 1.21.

The NSE-RMS, including the isokinetic sampling system and the wide range accident monitoring subsystem, is designed to carry out the following functions:

a. To provide continuous isokinetic and representative samples of the stack flow in compliance with the requirements of GDC 64, Regulatory Guide 1.21, and ANSI 13.1 (1971).

b. To continuously record releases of radioactive particulates, iodines and noble gases to the environs so that the total quantity of radioactive material released can be evaluated.

c. To alarm, in event that specified rates of release of radioactive material are exceeded.

d. To provide continuous real-time indications of radioactive releases during the accident and postaccident modes of operation.

e. Provide an isolation signal to the containment purge valves in the event of high radiation in the north stack effluent.

The north stack exhausts from the following systems:

a. Unit 1 turbine enclosure exhaust

b. Unit 1 turbine enclosure equipment compartment exhaust (including mechanical vacuum pump exhaust)

c. Unit 2 turbine enclosure exhaust

d. Unit 2 turbine enclosure equipment compartment exhaust (including mechanical vacuum pump exhaust)

e. Radwaste enclosure equipment compartment exhaust

f. Chemistry Laboratory fume hood exhaust

g. Chemistry Laboratory air exhaust

h. Control structure battery compartment exhaust

i. Unit 1 steam packing condenser and effluents from the recombination system

j. Unit 2 steam packing condenser and effluents from the recombination system

k. SGTS enclosure exhaust CHAPTER 11 11.5-9 REV. 19 SEPTEMBER 2018

LGS UFSAR

l. Unit 1 battery compartment exhaust

m. Unit 2 battery compartment exhaust

n. Control structure toilet room exhaust

o. SGTS filter exhaust

p. Drywell purge system exhaust

q. Offgas treatment system exhaust

r. Radwaste enclosure air exhaust Sampling stubs are provided on the exhaust ducts of most of the systems listed above for the purpose, of extracting grab samples as needed (see drawing M-26 for specific locations).

11.5.2.2.1.1 Isokinetic Sampling and Radiation Monitoring Systems Units 1 and 2 share the north stack and consequently the same RMS. Under normal plant operation, the stack flow rate varies from about 350,000 cfm for Unit 1 only in operation to about 668,450 cfm for both units in operation. The expected composition and concentrations of the effluent under normal plant operation are given in Table 11.5-4. Following an accident flow may be reduced.

Under this condition, the flow rate will be below the range capability of the isokinetic sampling system of the north stack normal range monitors, but the wide range accident subsystem will continue to provide representative data. The north stack is provided with three equally spaced honeycomb grids that serve the purpose of stabilizing, equalizing, and collimating the stack flow in order that the exhaust flow rate can be measured accurately and representative air sampling can be achieved. A flow rate sensing array is provided, consisting of 128 uniformly spaced total pressure sensors and 32 uniformly spaced static pressure sensors for providing an instantaneous traverse across the stack. A differential pressure sensor connected to this sensing array provides the stack flow rate signal. To allow maintenance without losing monitoring capability, a redundant stack flow rate signal is provided by nine thermal-type velocity sensors connected to an averaging network. A local switch allows selection of one of the two stack flow signals to be input to the microprocessor-based north stack effluent and wide range accident monitors for isokinetic control of the sample flow rates. The stack flow rate and sampling flow rates are indicated on demand in the control room via the wide range accident monitoring subsystem readout module or the RMDS display consoles (Section 11.5.6).

Two independent sampling arrays, each consisting of a set of 64 uniformly spaced isokinetic nozzles are provided for extracting representative samples at the stack cross-section. One array provides a sample for the normal plant operation radiation monitoring subsystem and the other for the wide range accident monitoring subsystem. Sample bypasses with associated pumps and flow controls are provided to allow the sampling array flow to follow the stack flow rate isokinetically without exceeding the maximum sample flow range of the monitors. The system has the capability to maintain isokinetic conditions with variations in stack flow rate of +/-25%. The sample is split into parallel paths. Each half is passed through a particulate filter provided with a radiation detector indicating the corresponding integrated measurement of the particulate effluent, CHAPTER 11 11.5-10 REV. 19 SEPTEMBER 2018

LGS UFSAR an iodine filter provided with an in-place detector, and a noble gas monitoring chamber. Thus, each of the two redundant monitoring racks provide the following outputs:

a. Sampling flow rates

b. Particulate radioactivity, integrated

c. Iodine radioactivity, integrated

d. Noble gas radioactive concentration From these data and the stack flow measurement, the total radioactive effluent may readily be evaluated. Readouts from the detectors are fed into microprocessors, which in turn provide outputs to the RMDS display consoles in the control room. The microprocessors are provided with memory retention capability to preclude the loss of data in event of a power failure.

There are one downscale and two upscale common alarms which annunciate abnormal conditions on either monitor in the control room. The upscale alarms indicate high and high-high radiation in any of the three detector channels, and the downscale alarm indicates instrument malfunction.

For the normal plant operation mode, the characteristics of the isokinetic sampling system and radiation monitoring subsystem provide plant operations personnel with complete and accurate data of radioactive materials released to the environs from the north stack. The system thus enables personnel to control activity release rates. Sufficient redundancy is provided to allow maintenance and checking of one channel without losing monitoring capability.

11.5.2.2.1.2 Wide Range Accident Monitor The wide range accident monitoring subsystem is independent of the normal plant operation monitoring subsystem and operates continuously. Effluent samples are drawn via two sample flow paths. During normal plant operation, one sample, drawn from the second 64 nozzle array described above, is passed through a particulate filter, iodine filter, and low range noble gas detector assembly. This provides redundancy to the normal plant operation monitoring subsystem. If the low range detector approaches its upscale limit, the system automatically starts pumping effluent sample from a separate comb-type probe located downstream of the isokinetic nozzle arrays. This sample is passed through shielded particulate and iodine filters and a midrange and high range noble gas detector assembly. When the midrange detector reaches a preselected point, the low range detector is automatically purged and flow through it is stopped.

Purging ensures that the low range detector can resume monitoring when activity again decreases. A similar automatic purge/shutdown cycle is performed on the midrange and high range detectors when activity decreases.

The low range detector assembly consists of a shielded chamber and a beta-sensitive plastic scintillation detector. The midrange and high range detectors use cadmium telluride solid-state sensors housed in a shielded chamber. Detector outputs are fed into the microprocessor that applies conversion factors, determines if alarm setpoints have been exceeded and retains data for each detector channel in history files. In addition, the microprocessor automatically calculates effluent release rate per unit time based on detector measurement and stack flow rate.

Background subtraction is provided based on manually entered background values. Outputs of the microprocessor are transmitted to a readout module, the RMDS display consoles, and CHAPTER 11 11.5-11 REV. 19 SEPTEMBER 2018

LGS UFSAR recorders in the control room. One 3-pen recorder provides indication of radioactivity concentration in the low range, midrange, and high range channels, while effluent release rate is recorded in a single-pen recorder. The RMDS display consoles can also provide trend printouts on demand of the microprocessor history files. The microprocessors have memory retention in event of loss of power. One downscale and two upscale alarms annunciate abnormal monitor conditions to the control room. The wide range accident monitoring subsystem also provides a nonsafety-related isolation signal to the containment purge valves on high-high radiation and loss of control power to the system.

The monitor is initially calibrated by the manufacturer using gaseous and solid sources including Sr-90. Transfer sources are then used to transfer this calibration data base to the plant by realigning each detector to the conditions established during primary calibration. Calibration is performed periodically in the plant at designated intervals according to the plant Technical Specifications.

Dissemination of information from this monitor via the RMMS data links to the control room, TSC, and EOF, as well as consideration of radionuclide distribution as a function of time after shutdown in dose assessment is discussed in Section 11.5.6.

The particulate and iodine filters of the wide range accident monitoring subsystem are used as grab sample modules to provide the capability of collecting representative samples of iodines and particulates for onsite analysis during and following an accident. The sample lines are heat traced to preclude entrained moisture in the effluent stream that could degrade the filters. Three removable filter modules are provided in both sample flow paths to allow continuous collection.

This also allows the control room operator to select a clean set of filters in order to prevent appreciable concentrations of noble gases produced by iodine decay in a loaded filter, which could be falsely interpreted by the noble gas detectors as high activity in the effluent stream.

Controls are provided in the control room to select filters, to grab samples locally and remotely, and to initiate automatically timed grab sampling. Filters on the high activity sample flow path are housed in shielded enclosures designed for ease of removal and replacement of filter media.

Filter removal is provided by means of quick disconnect couplings. After removal, the filter is placed in a shielded cask for transport to the onsite analysis facility. The filter enclosure and transport cask have been designed and the access routes have been selected in accordance with the requirements of NUREG-0737 to keep personnel exposure in sample handling and transport below the GDC 19 limits of 5 rem whole body exposure and 75 rem to the extremities during the duration of the accident. The sample filter, constructed of silver zeolite, has a collection efficiency of 99% for iodine and for 0.3 micron particles.

11.5.2.2.2 South Stack Effluent Radiation Monitoring System The objectives and functions of the SSE-RMS are the same as those of the north stack normal plant operation monitoring subsystem. A system for postaccident monitoring is not provided because any HVAC exhaust to this stack containing accident effluents is automatically isolated.

The south stack encloses two independent exhaust ducts servicing the reactor enclosures for Unit 1 and Unit 2, respectively. The stack exhausts ventilation air from the following systems:

a. Unit 1 Duct

1. Unit 1 reactor enclosure exhaust CHAPTER 11 11.5-12 REV. 19 SEPTEMBER 2018

LGS UFSAR

2. Unit 1 reactor enclosure equipment compartment exhaust

3. Refueling floor Unit 1 side exhaust

b. Unit 2 Duct

1. Unit 2 reactor enclosure exhaust

2. Unit 2 reactor enclosure equipment compartment exhaust

3. Refueling floor Unit 2 side exhaust Each of these two ducts is monitored by means of two redundant monitors. Consequently, four independent sets of data are obtained of stack flow rates and corresponding sampling rates.

Flow rates in each of the two ducts vary from about 43,000 cfm to about 223,000 cfm. The expected composition and concentrations of the effluent under normal plant operation are given in Table 11.5-4. The stack flow is collimated to provide a uniform velocity distribution over the entire cross-section to assure representative sampling. Flow rate in each duct is measured by a manifold containing 64 uniformly spaced total pressure sensors and 16 static pressure sensors to provide an instantaneous ongoing velocity traverse. A differential pressure sensor connected to the manifold provides the flow rate signal to both monitors associated with each duct for isokinetic control of the sample flow rate. Sampling is done by an array of 32 uniformly spaced isokinetic nozzles.

Radiation detection in each monitor is done by means of a particulate filter, iodine filter and noble gas chamber in series. Each of these items is shielded and provided with a dedicated detector.

The gas chamber has a beta scintillation detector consisting of a beta-sensitive crystal optically connected to a photomultiplier tube.

The other two detectors are similar in construction except that the iodine detector is gamma-sensitive. The output from the preamplifier is fed to a microprocessor, which in turn provides output to the RMDS display consoles in the control room. Digitized outputs are also available locally. Stack flow rates, sampling flow rates, and concentration of particulates, iodines, and noble gases are indicated on demand.

There are one downscale and two upscale common alarms that are annunciated in the control room for abnormal conditions in any of the four monitors. The downscale alarm indicates instrument malfunction and the upscale alarms indicate high and high-high radiation in any of the three detector channels.

11.5.2.2.3 Radwaste Equipment Rooms Ventilation Exhaust Radiation Monitoring System The common duct that collects exhaust air from the charcoal offgas treatment equipment compartments ventilation ducts is continually monitored for airborne radioactivity. High-high, high, and low radiation are annunciated in the control room. Although particulate/iodine concentration is not anticipated, the monitors are provided with particulate/iodine filters as backups. In the event of a high radiation alarm, the ventilation exhaust from each compartment can be monitored separately.

11.5.2.2.4 Charcoal Treatment System Process Exhaust Radiation Monitoring System CHAPTER 11 11.5-13 REV. 19 SEPTEMBER 2018

LGS UFSAR Two charcoal offgas trains (Unit 1 and Unit 2) exhaust the processed gases via HEPA filters to the north stack. Each of the exhaust pipes is monitored to detect malfunction in the corresponding offgas train. Low, high, and high-high radiation are annunciated in the control room.

11.5.2.2.5 Recombiner Room and Hydrogen Analyzers Exhaust Radiation Monitoring System RRHAC-RMS monitor detects airborne radioactivity in the ventilation ducts of the recombiner compartments, drain sump rooms, and hydrogen analyzer compartments in the control structure.

Exhaust air in the common duct from the recombiner compartments and drain sump rooms and the common duct from the hydrogen analyzer compartments is continuously monitored prior to return to the turbine enclosure equipment compartment ventilation filters for eventual release through the north stack. Low, high, and high-high radiation are annunciated in the control room.

The source of the radiation can be identified by means of a hand selector switch/solenoid valve arrangement which allows switching the sample line to the individual return duct of potentially contaminated compartments.

11.5.2.2.6 Steam Exhauster Discharge and Vacuum Pump Exhaust Radiation Monitoring System The SEDVP-RMS monitor provides an indication of radioactivity in the discharge of the steam packing exhauster to the north vent stack. The sampling arrangement allows the capability to also monitor the mechanical vacuum pump exhaust at the water separator discharge when this equipment is operated during startup.

11.5.2.2.7 Radwaste Enclosure and Chemistry Laboratory Expansion Ventilation Exhaust Radiation Monitoring System Monitoring of the main exhaust duct of the radwaste enclosure and Chemistry Laboratory Expansion provides for general surveillance of gaseous effluents from equipment compartments and access areas of these structures prior to discharge to the north stack. The REV-RMS instrumentation also provides backup for the RERV-RMS monitor. Low, high, and high-high radiation alarms annunciate in the control room.

11.5.2.2.8 Air Ejector Offgas Effluent Radiation Monitoring System The AEOE-RMS monitors radioactivity in the main condenser offgas after it has passed through the SJAE condenser, recombiner and aftercondenser. This is representative of gaseous radioactivity released from the reactor and therefore indicates the condition of the fuel cladding.

A continuous sample is extracted from the offgas pipe via a stainless steel sample line and passed through a sample chamber and a sample panel before being returned to the suction side of the SJAE. The sample chamber is a steel pipe that is internally polished to minimize plateout.

It can be purged with room air to check detector response to background radiation. The sample panel measures and indicates sample line flow. Heat tracing is provided in the sample line between the sample chamber and panel to prevent moisture condensation in the sample panel.

The sample chamber is monitored by three channels. Each channel has a gamma-sensitive ionization chamber mounted outside the sample chamber. Two channels have logarithmic radiation monitors with low, high, and high-high alarm outputs that annunciate in the control room.

CHAPTER 11 11.5-14 REV. 19 SEPTEMBER 2018

LGS UFSAR The third monitoring channel has linear radiation sensing capability designed to detect radioactive fragments resulting from fuel cladding deterioration. Small changes in the offgas gross fission product concentration can be detected by the continuous use of the linear radiation monitor. The linear radiation monitor is not a process monitor such as the channels described above, but is utilized as an expanded scale device for aiding in measurement of small changes in the offgas radiation level. The detector is a gamma-sensitive ionization chamber that monitors the same sample as the logarithmic monitors. The system uses a linear readout with a range switch instead of a logarithmic readout. The output from the monitor is recorded on a 1-pen recorder.

Grab samples can be obtained for determining isotopic composition by using the semiautomatic vial sampler panel. To draw a sample, a serum bottle is inserted into a sample chamber, the sample lines are evacuated, and a 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.

11.5.2.2.9 Primary Containment Leak Detector Radiation Monitoring System The design objective of the PCLD-RMS is to detect leakage in the primary containment. The leak detection monitor is designed to monitor and alarm for gaseous radioactivity (Section 5.2.5.2.1.4).

The monitoring system is provided with four main control room annunciated alarms (high-high, high, downscale, and abnormal sample flow). Rate meter output is recorded on a recorder in the main control room.

11.5.2.2.10 Hot Maintenance Shop Ventilation Exhaust Radiation Monitoring System Equipment serviced in the hot maintenance shop is expected to be contaminated with residual particulate radioactivity. A small quantity of radioactive iodine might also be present. No radioactive gases are anticipated.

Continuous isokinetic sampling of the maintenance shop exhaust duct, downstream of the HEPA filters, is conducted in accordance with ANSI N13.1. The flow controller in the monitor skid maintains isokinetic sample flow proportional to the constant exhaust duct flow. This sample is passed through fixed particulate and iodine filters. Separate detectors measure the gross radioactivity accumulated on these filters. The microprocessor-based monitor calculates radioactivity concentration based on the value of sample flow in memory and measured accumulated radioactivity over a time period. A recording of this radioactivity concentration is made locally, and an annunciator alarms in the event of discharges in excess of limits specified in the radioactive Effluent Controls Program defined in the Technical Specifications. All instrumentation is mounted locally.

11.5.2.2.11 Liquid Radwaste Discharge Radiation Monitoring System Waste liquid may be discharged on a batch basis from sample tanks in the liquid waste management system as discussed in Section 11.2.3. Prior to discharge, the liquid in the tank is recirculated, sampled, and analyzed for radioactivity, and the release and dilution rates are determined. The expected composition and concentrations of the effluent from each sample tank under normal plant operation are given in Table 11.5-5. The liquid radwaste effluent discharges into the cooling tower blowdown line. The LRD-RMS monitor measures activity in the discharge CHAPTER 11 11.5-15 REV. 19 SEPTEMBER 2018

LGS UFSAR line to prevent concentration in the cooling tower blowdown line from exceeding ten-times the 10CFR20, Appendix B, Table 2 activity limits (post-1994).

Monitoring is performed in an off-line sample rack in preference to in-line monitoring. This arrangement affords improved sensitivity and precludes the necessity of shutting down the radwaste discharge line in order to purge accumulated radioactive sludge from the monitoring system.

The monitoring channel consists of a gamma scintillation detector/ preamplifier, a rate meter in the auxiliary equipment area, and a 1-pen recorder in the radwaste control room. A low alarm trip will initiate discharge valve closure because the trip circuit has been designed to "fail-safe" in the event of a loss of power. The high trip alarms in the radwaste control room and the high-high trip initiates discharge valve closure. Since the radwaste release is based upon batch analysis, the basis for the alarm setpoint on the monitor is that an alarm be given on a gross release in the range 10-6 Ci/cc to 10-2 Ci/cc as a gross check against significant operator error. The alarm setpoint may vary from batch to batch depending upon the activity concentration of the batch and the available cooling tower discharge flow that is used to dilute the fluid effluent prior to leaving the site boundary.

11.5.2.2.12 Service Water Radiation Monitoring System Plant service water return flow discharges to the cooling tower. The return flow is monitored by the SW-RMS monitor. No activity attributable to reactor operation is expected to be present in this line. For radioactivity to be present, leakage would have to occur simultaneously in equipment cooled by the RECW system and in the RECW heat exchanger. Thus, this monitor provides a backup for the RECW-RMS monitor.

Off-line monitoring is selected to facilitate decontamination without shutdown of the service water system. The channel consists of a detector, preamplifier, a rate meter in the auxiliary equipment room, and a 1-pen recorder in the control room. The rate meter provides a high and high-high alarm in the control room. Annunciation due to low sample flow rate also is provided. The alarm setpoint is based upon detecting leakage into the service water with the setpoint set sufficiently above background to preclude spurious alarms.

11.5.2.2.13 Reactor Enclosure Cooling Water Radiation Monitoring System The RECW system cools components that may contain radioactive liquids, but does not normally carry any radioactive materials unless the cooled components leak. A radiation monitor is provided to measure radioactivity in the system.

An off-line monitor is employed, to facilitate decontamination without shutdown of the reactor enclosure cooling water system. The channel consists of a gamma scintillation detector, preamplifier, rate meter, and 1-pen recorder. The channel is provided with a high and high-high alarm that annunciate in the control room. The high alarm setpoint is set sufficiently above background to preclude spurious alarms. A low flow annunciator also is provided.

11.5.2.2.14 Process Sampling System A process sampling system is provided to allow grab sampling for evaluation of water quality and radioactivity levels in liquid process waste streams. The sample analysis results will provide operators with information for taking necessary corrective actions. This system is designed to CHAPTER 11 11.5-16 REV. 19 SEPTEMBER 2018

LGS UFSAR provide representative samples from process streams at central sample stations for use in minimizing leakage, spillage, and potential radiation exposure to operational personnel. Where applicable, means are provided for sample water cooling and for maintaining a fixed or measured sample flow rate. This system is described in Section 9.3.2.

No provisions have been made for sampling and analysis of laboratory and sample waste systems because all laboratory and sample station drains are routed to either the liquid radwaste, normal waste, or condensate systems, depending on their origin and quality. Processing takes place in their respective systems.

There are no liquid radwaste tanks outside buildings. The condensate and refueling water storage tanks that are located outside of buildings are sampled downstream of their respective transfer pumps via connections to the process sampling system.

Although the process sampling system is designed to provide liquid samples from many plant process streams, radionuclide sampling will be periodically performed on the following process systems:

a. FPCC

b. RECW

c. Liquid radwaste - equipment drain processing

d. Liquid radwaste - floor drain processing

e. Liquid radwaste - chemical and laundry processing 11.5.2.3 Postaccident Systems 11.5.2.3.1 Primary Containment Post-LOCA Radiation Monitoring System The PCPL-RMS is comprised of four ion chamber sensors for the primary containment in the event of a LOCA. After such a postulated accident, the monitoring system measures the gross radioactivity present in the containment atmosphere. This information is transmitted to control room personnel to provide them with a basis for making safety-related decisions. The PCPL-RMS provides a trip signal to the containment sump pumps on an upscale alarm indication. A downscale annunciator is provided to indicate instrument malfunction.

The sensors are located in separate areas of containment to provide independent measurements and to view large fractions of the containment volume (drawings N-119 and N-134).

Consideration to accessibility for maintenance and calibration was given in the selection of the sensor locations. The sensors are located in relatively open areas to prevent shielding that could impair their detection function.

The monitoring system provides energy response from 60 KeV to 3 MeV, with uniform response within +/-20% from 80 KeV to 3 MeV. Onsite calibration of the monitors will be performed with a calibrated 100 millicurie Cs-137 gamma source that will provide an effective dose rate of approximately 10 R/hr. A built-in current source is provided in the monitors to allow calibration checks through electronic signal substitution for the decades above 10 R/hr.

CHAPTER 11 11.5-17 REV. 19 SEPTEMBER 2018

LGS UFSAR Under high drywell temperature conditions, Insulation Resistance (IR) leakage current will cause a system error. Because the instrument signal at low radiation levels is very weak, high temperature IR leakage current significantly affects the accuracy of the indicated readings up to a maximum of 112.5 Rad/hr at the maximum design drywell temperature of 340°F. As a result, the indicated readings below 112.5 Rad/hr may not be within the factor of two accuracy recommendation of Regulatory Guide 1.97 Rev. 2. The induced error decreases exponentially with drywell temperature and becomes insignificant below 230°F. This induced error is significant only under low radiation conditions coincident with high drywell temperatures, whereas the system will operate to perform its principal function under normal and varying temperature conditions during and following an accident.

11.5.2.3.2 North Stack Wide Range Accident Monitoring System The north stack wide range accident monitoring system is discussed in Section 11.5.2.2.1.

11.5.2.3.3 Postaccident Sampling System The PASS is discussed in Section 11.5.5.

11.5.3 EFFLUENT MONITORING AND SAMPLING The requirements of GDC 64 are implemented with respect to effluent discharge paths by means of the following monitoring stations:

a. Gaseous Effluents:

1. Reactor enclosure ventilation exhaust (Section 11.5.2.1.2)

2. Refueling area ventilation exhaust (Section 11.5.2.1.3)

3. Standby gas treatment system (Section 11.5.2.2.1)

4. North stack ventilation exhaust (Section 11.5.2.2.1)

5. South stack ventilation exhaust (Section 11.5.2.2.2)

6. Charcoal offgas system compartments ventilation exhaust (Section 11.5.2.2.3)

7. Charcoal offgas system effluent (Section 11.5.2.2.4)

8. Recombiner, hydrogen/oxygen analyzers, and equipment drain sump (Section 11.5.2.2.5)

9. Steam seal effluent (Section 11.5.2.2.6)

10. Radwaste enclosure and Chemistry Laboratory Expansion ventilation exhaust (Section 11.5.2.2.7)

11. Hot shop ventilation exhaust (Section 11.5.2.2.10).

Sampling stubs are provided on all major exhaust ducts for the purpose of extracting grab samples as required.

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LGS UFSAR

b. Liquid Effluents:

1. Residual heat removal service water (Section 11.5.2.1.6)

2. Liquid radwaste discharge (Section 11.5.2.2.11)

3. Plant service water (Section 11.5.2.2.12).

11.5.4 PROCESS MONITORING AND SAMPLING The requirements of GDC 60 are implemented with respect to the automatic termination of gaseous and liquid effluent discharges by means of the following monitoring systems:

a. Main steam line (Section 11.5.2.1.1)

b. Reactor enclosure ventilation exhaust (Section 11.5.2.1.2)

c. Refueling area ventilation exhaust (Section 11.5.2.1.3)

d. Residual heat removal service water (Section 11.5.2.1.6)

e. Liquid radwaste discharge (Section 11.5.2.2.11).

The requirements of GDC 63 are implemented with respect to the monitoring of radiation levels in radioactive fuel and waste storage systems by means of the following monitoring systems:

a. Area radiation monitor channels 31 and 32 (Section 12.3.4.1)

b. Refueling area ventilation exhaust (Section 11.5.2.1.3)

c. Radwaste enclosure and Chemistry Laboratory Expansion ventilation exhaust (Section 11.5.2.2.7)

d. Liquid radwaste discharge (Section 11.5.2.2.11).

The following liquid process systems are provided with grab sample stations for laboratory measurement of radioactive concentrations for satisfying the requirements of GDC 63 and GDC 64:

a. Liquid radwaste systems (Sections 11.5.2.2.14.c, 11.5.2.2.14.d, and 11.5.2.2.14.e)

b. Reactor enclosure cooling water (Section 11.5.2.2.14.b)

c. Spent fuel pool treatment system (Section 11.5.2.2.14.a)

d. Residual heat removal service water (Section 11.5.2.1.6) 11.5.5 POSTACCIDENT SAMPLING SYSTEMS The PASS is designed to obtain representative liquid and gas grab samples from the primary coolant system and from within the primary and secondary containments for radiological and chemical analysis under accident conditions. The grab samples are subsequently transported to CHAPTER 11 11.5-19 REV. 19 SEPTEMBER 2018

LGS UFSAR the Chemistry Laboratory Expansion and counting facility for chemical and radioisotopic analyses, or shipped offsite for analysis.

The PASS was designed to satisfy certain requirements of NUREG-0737 (Item II.B.3).

Limerick license amendment numbers 166/129 approved the elimination of the requirement to have and maintain the Post Accident Sampling System. The following items were committed to as part of the license amendment numbers 166/129.

1. Limerick has developed contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, suppression pool, and containment atmosphere. The contingency plans are contained in the Limerick chemistry procedures. Establishment of contingency plans is considered a regulatory commitment.

2. The capability for classifying fuel damage events at the Alert level threshold has been established at a level of core damage associated with radioactivity levels of 300 micro-curies/gm dose equivalent iodine in the primary coolant system. This capability is described in Limericks emergency plans and emergency plan implementing procedures. The capability for classifying fuel damage is considered a regulatory commitment.

3. Limerick has established the capability to monitor radioactive iodines that have been released offsite to the environs. This capability is described in the emergency plans and emergency plan implementing procedures. The capability to monitor radioactive iodines is considered a regulatory commitment.

The following information contained in the UFSAR regarding the regulatory requirements for post accident sampling is retained for historical purposes.

The system design minimizes operating complexities and "in-line" instrumentation, is modular for maintenance and contamination control purposes, and is compact in size to reduce the amount of shielding required. The system can be used to provide samples under all plant conditions, ranging from normal shutdown and power operation to postaccident conditions.

The PASS P&ID is shown in drawing M-30. The equipment includes isolation and control valves, piping racks, shielded sample stations (gas and liquid), liquid chillers, and control panels. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the PASS are discussed in Section 3.2. A separate PASS is provided for each unit with common demineralized water and nitrogen support systems.

11.5.5.1 System Description 11.5.5.1.1 Sample Points

a. Wetwell and Drywell Atmospheres Sample lines are installed to obtain atmosphere samples from two separate areas in both the drywell and wetwell. Drywell samples are taken at el 291' and el 242'.

Wetwell samples are taken at el 222' on opposite sides of the containment. The sample lines tap into the CAC system sample lines outside the primary containment and outboard of the second containment isolation valve. Containment CHAPTER 11 11.5-20 REV. 19 SEPTEMBER 2018

LGS UFSAR gas samples will be representative of conditions throughout the primary containment because the containment is not compartmentalized and the atmosphere is fully mixed.

b. Secondary Containment Atmosphere A sample line is provided to allow sampling of secondary containment atmosphere noble gases to aid in determining postaccident accessibility of the reactor enclosure. Samples are taken in the vicinity of access doors 191 (Unit 1) and 287 (Unit 2) on el 217'.

The reactor enclosure is not a vital area requiring post accident access as defined in Section 1.13. For design basis events the reactor enclosure is largely inaccessible due to airborne noble gas activity. With use of self-contained breathing apparatus (SCBA), noble gases are the limiting factor throughout the accident. Once accessibility is determined using PASS, reactor enclosure airborne iodine activity can be determined using portable equipment available to HP personnel, in order to assess the need for SCBA in long term accident recovery operations.

c. Reactor Coolant and Suppression Pool When the reactor is pressurized, reactor coolant samples are obtained from a tap off the jet pump pressure instrument system. The sample point is on a noncalibrated jet pump instrument line outside the primary containment and downstream of the excess flow check valve. This sample point location is preferred over the normal reactor sample points on the RWCU system inlet line and recirculation line because the RWCU is expected to remain isolated under accident conditions, and it is possible that the recirculation line containing the sample line may be isolated. The jet pump pressure tap is in a location protected from damage and debris. This sample point provides representative samples of reactor coolant under various reactor conditions:

1. Normal operation/small pipe break: Reactor water level can be maintained at or near normal water level. With a nearly normal water level, or at least water in the upper plenum, natural circulation will occur with a large loop from the downcomer to the shroud region via the jet pumps. With thermal conditions pumping water up through the core and back down past the tap from which the PASS sample is taken, a representative relationship will exist which will allow the results of the sample to be related to the condition of the core.

2. Large Pipe Break: A large pipe break, such as a recirculation pump suction line break, may occur wherein the water level may be controlled only by the height of the jet pumps and the ability to add water to the vessel. The sample taps are located sufficiently low to permit sampling at a reactor water level even below the lower core support plate. As reactor pressure decays, LPCI is initiated into the core region. This water volume supplies more coolant than is boiled off by the decay heat. This excess water will flow down past the core, up through the jet pumps, and out CHAPTER 11 11.5-21 REV. 19 SEPTEMBER 2018

LGS UFSAR through the postulated break, assuring a representative sample at the sample point.

To ensure a representative liquid sample from the jet pumps at low (<1%) power conditions for small break or nonbreak events, the reactor water level will be raised to the level of the moisture separator when this action is not inconsistent with station EOPs. This will fully flood the separators and will provide a thermally induced recirculation flow path for mixing. Alternatively, operating at least one reactor recirculation pump 1(2)A-P201 or 1(2)B-P201 at minimum speed or higher will provide sufficient mixing.

Samples will be taken from the reactor via the jet pump pressure instrument lines as long as possible. This allows a more direct and therefore faster response to core conditions. Upon decay or loss of reactor pressure, the jet pump sample point is lost, and the RHR loops sample points must be employed for sampling.

Reactor coolant and/or suppression pool samples may be taken from the RHR sample lines, depending on the mode of RHR operation. These sample lines tap off downstream of the second system isolation valve in the RHR system sample lines at the discharge of each RHR heat exchanger.

1. LPCI: Suppression pool water is injected into the core, flows up through the jet pumps, and back to the suppression pool via the postulated break.

The system will be operated for an estimated 30 minutes minimum prior to sampling of the suppression pool water to ensure that a representative sample is obtained at the sample taps.

2. Shutdown Cooling: The RHR system, aligned in the shutdown cooling mode, provides cooling and circulation of reactor coolant through the core, resulting in a representative sample at the RHR sample taps.

3. Suppression Pool Cooling: The RHR system, aligned in the suppression pool cooling mode, provides cooling and circulation of the suppression pool water. The system will be operated for an estimated 30 minutes minimum prior to sampling of the suppression pool water to ensure that a representative sample is obtained at the RHR sample taps.

11.5.5.1.2 Isolation Valves and Sample Lines Containment isolation for the drywell and wetwell gas sample lines is provided by the existing CAC system sample line isolation valves. Jet pump instrument line containment isolation is provided by the existing restricting orifice and excess flow check valve upstream of the sample tap.

Containment isolation for the RHR sample lines is provided by the existing RHR sample line automatic isolation valves. All automatic isolation valves can be overridden from the control room.

PASS remote-operated sample line valves are controlled from local panels located adjacent to the sample stations.

All safety-related solenoid isolation and control valves that are part of the PASS are environmentally qualified for the conditions in which they must operate. Table 3.2-1 contains a description of the safety-related portions of the PASS, and Section 3.11 contains a description of the equipment qualification program.

CHAPTER 11 11.5-22 REV. 19 SEPTEMBER 2018

LGS UFSAR Nonsafety-related valves that are part of the PASS are not included in the LGS equipment qualification program. However, those valves that are not accessible for repair after an accident do not contain materials that, if degraded, would prevent the PASS from performing its sampling function.

Sample line routings are as direct and short as practical. Recirculation flow rates in the liquid sample lines are maintained in the turbulent flow regime. The gas sample lines are equipped with heat trace which was originally intended to prevent plate-out of iodine vapor. The PASS is not used for iodine sampling. Therefore, the heat trace is not required for proper gas sampler operation.

11.5.5.1.3 Piping Rack The piping rack, which is installed within the reactor enclosure, includes sample coolers and control valves that determine the liquid sample flow path to the sample station. The rack provides a flow path to recirculate liquid samples, bypassing the sample stations, until a representative sample condition is obtained. The cooling water is supplied by the RECW system.

11.5.5.1.4 Sample Station and Control Panels The sample station consists of a floor stand, frame, and sample enclosures, and is mounted flush against the outside of the secondary containment wall. Included within the sample station are equipment trays that contain modularized liquid and gas samplers. The liquid sample portion of the sample station is shielded with 6 inches of lead brick, whereas the gas sampler has 2 inches of lead shielding. The various sample and return lines enter the sample station enclosure through the back by way of a penetration through the reactor enclosure wall. Control instrumentation is installed in two control panels mounted side-by-side. One of these panels contains the conductivity and radiation level readouts. The other control panel contains the flow, pressure, and temperature indicators, and various valve controls and switches. A graphic display is provided directly below the main control panel which shows the status of the pumps and valves. A radiation monitor provides an indication of dose rates in the immediate vicinity of the control panel and is used to limit operator exposure.

11.5.5.1.4.1 Gas Sampler The gas sample system is designed to operate at pressures ranging from subatmospheric to the design pressures of the primary containment one hour after a LOCA. The sample flow is chilled to remove moisture, and a 15 cc nominal grab sample can be taken for determination of gaseous activity and gas composition by gas chromatography. The gas is collected in an evacuated vial using hypodermic needles. When purging the drywell and wetwell gas sample lines to obtain a representative sample, the flow is returned to the wetwell. During purging of the secondary containment line and when flushing the sample panel lines with nitrogen, flow is returned to secondary containment. The sample station design allows for sample gas or nitrogen flushing of the entire sample panel line downstream of the four-position selector valve. This capability will minimize cross-contamination between the various samples.

11.5.5.1.4.2 Liquid Sampler The liquid sample system is designed to operate at pressures from 0-1150 psi. The design flow rate of 0.2 to 1 gpm is sufficient to maintain turbulent flow in the sample line and serves to minimize cross-contamination between samples. The flow is returned to the suppression pool.

CHAPTER 11 11.5-23 REV. 19 SEPTEMBER 2018

LGS UFSAR The liquid sampling system is designed to allow demineralized water flushing of the system lines from a point in the piping station through the sampling needles. A radiation monitor in the sample enclosure monitors the sample return to provide assessment of sample activity level and the effectiveness of the demineralized water flush after sampling.

a. Diluted Liquid Sample All liquid samples are taken into nominal 15 milliliter septum bottles mounted on sampling needles. In the sampling lineup, the sample flows through a conductivity cell (0.1-1000 mho/cm) and through a ball valve bored to nominal 0.10 milliliter volume. After flow through the sample panel is established, the ball valve is rotated 90, and a syringe is used to flush the sample and a measured volume of diluent (generally 10 milliliters) through the valve and into the sample bottle. This provides an initial dilution of up to 100:1 nominal. The sample bottle is contained in a shielded cask and remotely positioned on the sample needles through an opening in the bottom of the sample enclosure.

b. Nondiluted Liquid The sample station can provide depressurized samples of the primary coolant.

Ten milliliter aliquots nominal of degassed liquid may be taken for offsite (or onsite depending on activity level) analyses, which requires a relatively large undiluted sample. This sample is obtained remotely using the large volume cask and cask positioner through a needle on the underside of the sample station enclosure .

The PASS system was originally designed to obtain dissolved gas samples from the primary coolant as a means of assessing core damage per NUREG-0737. However, core damage assessment is accomplished by use of the CAC system as described in UFSAR Section 6.2.5. Therefore, this PASS feature is not used.

11.5.5.1.4.3 Sample Station Ventilation The sample station enclosure is vented to a Zone V room in the secondary containment.

Ventilation is facilitated by differential pressure between the control structure and reactor enclosure. The ventilation rate required for heat removal and proper sweep velocity during operation is about 40 scfm. A pressure gauge is attached to the sample station enclosure to monitor the pressure differential between the enclosure and the general sampling area in the control structure. The pressure differential will assure the operator that airborne activity in the sample enclosure will be swept into secondary containment.

11.5.5.1.4.4 Sample Station Sump The sample station is provided with a bottom sump to collect liquid leakage. This sump can be isolated, pressurized, and discharged into the sample station liquid return line to the suppression pool.

11.5.5.1.4.5 Sample Handling Tools and Transport Containers Appropriate sample handling tools and transporting casks are used. Gas vials are installed and removed by use of a vial positioner through the front of the gas sampler. The vial is manually CHAPTER 11 11.5-24 REV. 19 SEPTEMBER 2018

LGS UFSAR lowered into a shielded cask directly from the positioning tool. This allows the operator to maintain a distance of about three feet from the unshielded vial. The cask provides about 1-1/8 inches of lead shielding. A 1/8 inch diameter hole is drilled in the cask so that an aliquot can be withdrawn from the vial with a gas syringe without exposing the analyst to the unshielded vial.

The small volume (diluted) liquid sample cask is a cylinder with a lead wall thickness of about 2 inches. The cask weighs approximately 50 pounds and has a handle which allows it to be carried by one person.

The nominal 10 milliliter undiluted sample is taken in a 700 pound lead shielded cask which is transported and positioned by a four-wheel dolly. The sample is shielded by about 51/2 inches of lead. A licensed shipping cask for transport of the undiluted samples to the offsite analysis facility (Section 11.5.5.2.2) has been procured in conjunction with a group of other utilities. This cask will be placed in a centrally located, continuously attended warehouse facility.

11.5.5.1.4.6 Sample Station Power Supply The PASS isolation and control valves, sample station control panels, and auxiliary equipment are connected to an instrument ac distribution panel which is powered from an engineered safeguard system bus. Following a LOOP, the engineered safeguard system bus is powered from the onsite diesel generators. The RECW system, which is needed for the sample coolers, is also powered from the emergency diesel generators following a LOOP.

All electrically operated components associated with the PASS are capable of being supplied with power and operated within 30 minutes of an accident in which there is core degradation, assuming a LOOP.

11.5.5.2 Description of Sample Preparation/Chemistry and Nuclear Counting Facilities After the samples are obtained from the sample station, they will be transported to the Chemistry Laboratory Expansion where they will be diluted as necessary and appropriate aliquots taken for chemical and radioisotopic analysis. The radioisotopic analysis will be done in an area where background radiation can be kept to a minimum. The primary facility for performance of these analyses is the chemistry laboratories and counting room in the Chemistry Laboratory Expansion.

In addition to these onsite facilities, which are intended to handle the gas samples and the diluted liquid samples, prior arrangements will be made with an offsite laboratory for supplemental and confirmatory analysis of samples as required.

11.5.5.2.1 Onsite Facilities The chemistry lab is equipped to handle the gas samples and the nominal 0.1 ml diluted liquid samples. The maximum activities of these samples will be 0.50 Ci and 0.32 Ci, respectively, using one hour decay and the fractional releases of core inventory as discussed in Section 11.5.5.5.

The laboratory will maintain a dedicated inventory of items such as lead bricks for shielding, gas syringes, gloves, reagents for analysis, etc., which will be needed in case of an accident. The laboratory will be equipped with a gas chromatograph, pH meter, conductivity meter, turbidimeter and other instrumentation needed to perform the required analyses. This equipment, however, may not be dedicated exclusively to postaccident analysis. Supplied air or self-contained breathing CHAPTER 11 11.5-25 REV. 19 SEPTEMBER 2018

LGS UFSAR masks will be available in the event of high activity levels in the ventilation supply or accidental spills in the laboratory.

The primary counting facility located adjacent to the chemistry laboratories in the Chemistry Laboratory Expansion is equipped to handle the gamma spectra analyses required for postaccident samples. The detectors in the counting room are equipped with lead shields connected to computer based analyzer systems which have automatic peak search and isotope identification capabilities. The capability to purge the volume within the shield with compressed gas is available for at least one detector assembly. This will help prevent atmospheric noble gas activity released during an accident from swamping the detector.

It is expected that the first set of postaccident samples will be analyzed in the chemistry lab/counting room facilities located in the Chemistry Laboratory Expansion approximately two hours after the start of an accident. At this time, the chemistry lab will be a Zone III area and therefore accessible for performing the required chemical analyses. The lab becomes a Zone II area within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following an accident. The counting room is a Zone II area within two hours, and a Zone I area within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following an accident. Shielding for the PASS is further discussed in Section 1.13.2 (Item II.B.2).

The most direct route from the sampler location to the chemistry lab is through the control structure and Unit 1 turbine enclosure to the radwaste enclosure on el 217' and then through the doorway (ramp) of the west wall of the radwaste building to the Chemistry Laboratory Expansion.

However, during the first 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following an accident, high radiation in portions of this access path require that an alternate route from sampler to labs be taken. Operators must exit through the north of the turbine enclosure, travel around the west end of the radwaste building, and enter the lab/counting room area in the Chemistry Laboratory Expansion through the doorway of the west wall radwaste building.

11.5.5.2.2 Arrangements for Offsite Analyses A part of the LGS approach to postaccident sampling is the establishment of prior arrangements with an offsite laboratory for confirmatory and supplemental analyses.

11.5.5.3 Sample Collection and Transport Procedures It is anticipated that the first set of samples will be taken within one hour following a LOCA, with samples taken approximately every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the remainder of the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Following day one, it is expected that three samples per day may be taken for the remainder of the first week, with samples once per day following the first week for the duration of the accident.

The following is a conservative time sequence for sampling, transport, and analysis to demonstrate that samples can be obtained and analyzed within the specified 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> period:

a. Recirculate sample, install sample vial - 15 min

b. Operate sample station - 15 min

c. Transport sample to lab - 20 min

d. Analyze sample - 60 min CHAPTER 11 11.5-26 REV. 19 SEPTEMBER 2018

LGS UFSAR Atmosphere sample lines from the CAC system and liquid samples from the RHR system are automatically isolated following a containment isolation signal. The sample station operator(s) must confirm with the control room personnel that the necessary isolation valves are opened. A telephone extension to the control room is provided near the PASS control panel.

Controls for all other PASS sample line valves are located on or adjacent to the PASS control panel. All controls for valves that are a part of the sampler or piping rack are located on the PASS control panel.

Following a series of presampling checks and procedures, including adjustment of the enclosure damper to ensure adequate ventilation, checks of demineralized water and nitrogen supplies, flushing of system with demineralized water, draining the trap and sump, etc., the system is ready for obtaining the samples.

11.5.5.4 Chemical/Radiochemical Procedures The PASS provides a means of obtaining primary coolant, suppression pool, and primary and secondary containment air samples for radiochemical and chemical analysis following a major reactor accident. Because of the extremely high radioactivity levels associated with extensive fuel damage, the PASS was developed to provide the capability of obtaining the necessary samples and performing analyses, as required, for immediate plant needs, or as defined by regulatory requirements. Procedures have been established for shipping samples to appropriate facilities to perform detailed analyses on multicurie level samples.

11.5.5.4.1 Sample Preparation All sample bottles, etc., will be identified prior to sampling to eliminate unnecessary exposure resulting from handling high level samples. A centralized logging system will be developed to track sample aliquot identification, dilution factors, sample disposition, etc.

Liquid samples will be taken at the sample station in septum-type bottles and transported to the analysis facility in lead containers.

Sample aliquots are taken from the septum bottles for analysis or further dilution. Aliquoting and transfer will be performed using shielded containers, or behind a lead brick pile. Calibrated hypodermic syringes will be used for aliquoting the higher activity samples. Tongs or other holding/clasping devices will be available for holding the sample bottle during the transfer and dilutions to reduce hand and body exposure. Unless prohibited by the intended analysis, dilutions will be done using very dilute (about 0.01 N) nitric acid as the diluent to minimize sample plateout problems.

Primary coolant samples obtained from the sampling station are diluted by a factor of 100 (0.1 ml coolant diluted to 10 ml). Under severe accident conditions, a calibrated syringe would be used to obtain an aliquot for this sample for further dilutions. At the maximum expected primary coolant activity level (3 Ci/cc), a dilution factor of 1x105 would be required for gamma spectroscopy.

Direct counting of the initial nominal 100:1 dilution sample would allow analysis at coolant activity levels down to 1 Ci/cc. In addition, the degassed, undiluted nominal 10 ml sample available from the sample station could be used for analysis of samples in the 10-2 Ci/cc to 10-3 Ci/cc range.

Thus, useful samples may be obtained from the postaccident sampling station for coolant activity CHAPTER 11 11.5-27 REV. 19 SEPTEMBER 2018

LGS UFSAR levels ranging from DBA source terms to well below the maximum level that can be tolerated at the normal reactor sample station.

Gas samples are taken at the sample station in a nominal 15 milliliter septum bottle. A lead carrier is furnished with a small hole at the septum end so that a gas sample can be withdrawn from the carrier using a hypodermic syringe without having to handle the bottle.

Samples taken from the gas sample bottle will either be injected into a gas chromatograph for analysis or diluted for gamma spectroscopy. The dilutions will be performed in a manner analogous to the liquid samples. Fractional milliliter samples can be transferred to new 15 ml gas bottles without concern for sample leakage due to pressurization. For larger volume aliquots, a gas syringe will be used to draw a partial vacuum in the bottle prior to sample transfer.

11.5.5.4.2 Chemical Analysis Approved chemistry procedures are chosen on the basis of simplicity, stability, minimum of radiation exposure and least likelihood to cause major contamination problems. They have been tested for radiation sensitivity and are suitable for use at the PASS design basis source term of 3.13 Ci/gm, and where applicable, with the design basis 0.1 ml to 10 ml nominal dilution at the sample station.

a. Gross activity, gamma spectra analyses will be accurate within at least a factor of two over a coolant activity range of 10 Ci/cc to 10 Ci/cc (Additional information is provided in Section 11.5.5.4.1).

b. Boron concentrations will be determined by an approved Chemistry procedure.

Concentrations between 50 ppm and 1100 ppm are of interest for BWR reactivity control in the event sufficient control rods are not inserted to shutdown the reactor.

The approved method with the 100:1 diluted sample will have an accuracy of

+/-10% over the described range.

Tests have been performed to verify that none of the expected post accident chemical constituents (I, Cs+, Ba+2, La+3, Ce+4, Cl-, B, Li+, NO-3, NH+4, K+) will interfere with this analysis method.

c. The chloride analysis performed onsite will be accomplished using an approved Chemistry procedure. The use of the method with the nominal 100:1 diluted sample will result in a accuracy of at least +/-10% for coolant concentration over the range of 500 ppb to 20 ppm.

Tests have been performed to verify that none of the expected post accident chemical constituents listed above will interfere with the results of this analysis for the procedures and equipment which will be employed.

Offsite provisions for chloride analysis will be accurate +/-10% over the range 0.5 ppm to 20 ppm and +/-0.05 ppm below concentrations of 0.5 ppm.

d. A combination electrode will be used to measure the pH of coolant samples.

Testing performed by GE has verified that expected levels of irradiation result in a shift of less than 0.3 pH units.

CHAPTER 11 11.5-28 REV. 19 SEPTEMBER 2018

LGS UFSAR

e. The postaccident sample station is equipped with a 0.1 cm-1 conductivity cell. The conductivity meter has a linear scale with a six-position range selector switch to give conductivity ranges of 0-3 mho/cm, 0-10 mho/cm, 0-30 mho/cm, 0-100 mho/cm, 0-300 mho/cm, and 0-1000 mho/cm when using the 0.1 cm-1 cell.

This conductivity measurement system will be used to determine the primary coolant or suppression pool conductivity. During normal operation the LGS Technical Specifications require maintaining the primary coolant below 1.0 mho/cm (at 25C), and conductivity measurements are the primary method of coolant chemical control.

Conductivity measurements are, of course, nonspecific, but they serve the important function of indicating changes in chemical concentrations and conditions.

Perhaps even more important, in the case of the BWR primary coolant, the conductivity measurements can establish upper limits of possible chemical concentrations and can eliminate the need for additional analyses.

The conductivity measurement can also be used to bound the possible range of pH values.

11.5.5.4.3 Radiochemical Analysis/Gamma Ray Spectroscopy After the samples have been brought to the chemistry laboratory and appropriately diluted, they can be carried without shielding to the counting room which is adjacent to the chemistry laboratory. The appropriate dilution factors will be somewhat dependent on the detector and shelf arrangements available. A prior determination of the maximum desirable dose rates for the various shelf configuration will be made to minimize this problem. The present Gamma Spectrometer in the onsite counting room will handle the analysis of these samples within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the time a decision is made to sample.

The gas samples will be counted in the PASS gas sample vials, and the liquid samples will be counted in the standard sample bottles used during normal operation because calibration curves for these geometries will be available and regularly updated. In general, the counting of the postaccident sample will follow the normal counting room procedures. The peak search and identification library will contain the principal gamma rays of the following isotopes in addition to the standard activated corrosion products:

a. Noble gases: Kr-85, Kr-85m, Kr-87, Kr-88, Xe-131m, Xe-133, Xe-133m, Xe-135

b. Iodines: I-131, I-132, I-133, I-135

c. Cesiums: Cs-134, Cs-137

d. Others: Ba/La-140, Ce-141, Ce-144, Ru-106, Te-129, Te-129m, Te-131, Te-131m, Np-239 If the levels of noble gases in the ambient atmosphere surrounding the detector are high enough to cause significant interference or to overload the detector, a compressed air or nitrogen purge of the detector shield volume will be maintained.

CHAPTER 11 11.5-29 REV. 19 SEPTEMBER 2018

LGS UFSAR The onsite radiological and chemical laboratory facilities are equipped with gamma spectral analysis equipment to quantify the radionuclides present in gas and liquid samples. Shielding is provided for the radiation detectors to minimize the effect of background radiation. Initial dilutions are performed in the process of taking liquid samples at the sample stations. Any additional dilutions required will be performed in the laboratory fume hood behind a lead brick pile.

11.5.5.4.4 Gas Analysis A gas chromatograph will be used to measure hydrogen and oxygen concentrations in containment atmosphere. The accuracy of the containment gas analysis will be at least +/-5%

between 0.1% and 30% (volume) of the constituent.

11.5.5.4.5 Determination of Extent of Core Damage A generic procedure to assess the extent of core damage based on radionuclide concentrations and other parameters has been prepared by the BWROG and transmitted to the NRC by letter from T.J. Dente (BWROG) to D. Eisenhut (NRC) dated June 17, 1983. A licensee procedure based on this methodology has been prepared and transmitted to the NRC.

11.5.5.4.6 Storage and Disposal of Sample Short-term sample storage areas will be provided in the chemistry laboratory and counting room facilities. The area to be used for long-term storage of the samples is the source storage room, number 244, in the radwaste enclosure. Low level wastes generated by the chemistry procedures will be flushed to radwaste. Ultimate procedures for disposal of the samples will be determined later; however, after a sufficiently long decay period, the activity levels will be significantly reduced.

This will ease exposure problems during disposal.

11.5.5.4.7 System Testing and Operator Training Equipment used for post-accident sampling and analyses will be calibrated and tested per an approved preventive maintenance program. The testing program is designed so that sampling operations are conducted approximately every twenty-four months.

11.5.5.5 Dose Rate Analysis The post-LOCA core inventory of fission products was calculated assuming a three year irradiation, 100% availability, and reactor operation at 102% of rated power. Fractional releases of fission products from the fuel to the reactor water, suppression pool, and containment atmosphere were based on Regulatory Guide 1.3 and Regulatory Guide 1.7 assumptions. The resulting source terms were used in the design of PASS shielding and in determining doses to operators.

The sampling and analysis provisions at LGS have been designed such that it will be possible to obtain and analyze a sample following an accident without exceeding the criteria of GDC 19. Time sequences and calculated dose rates to verify compliance for a sample taken one hour post-LOCA are given in Table 11.5-3.

11.5.6 RADIATION AND METEOROLOGICAL MONITORING SYSTEM 11.5.6.1 General Description CHAPTER 11 11.5-30 REV. 19 SEPTEMBER 2018

LGS UFSAR The RMMS is a system consisting of data acquisition computers with report generating capabilities designed to perform the following functions:

a. Collect radioactivity level information from selected microprocessor-based process radiation monitors throughout the plant

b. Remotely operate radiation monitors

c. Display radiation levels

d. Generate radioactive effluent reports for compliance with Regulatory Guide 1.21

e. Calculate offsite dose consequences from normal plant operation releases

f. Calculate offsite dose consequences from accident conditions in accordance with NUREG-0654.

g. Meteorological data is transmitted from the met towers to Unit I PMS for display, averaging and transmission to EPDS.

11.5.6.2 RMMS Hardware Description Deleted 11.5.6.2.1 Radiation Monitoring and Display System The RMDS is a dual-computer system that monitors, stores, and displays information obtained from selected process radiation monitors described in Section 11.5.2.

The RMDS also transmits radioactivity concentration and stack flow rate information obtained from the process radiation monitors to the EPDSto be used in calculating offsite doses.

The RMDS dual-computer configuration is designed with one computer being the primary computer. The computer collects data from the process radiation monitors and displays the data on RMDS workstations. The second computer is redundant and will take over the functions of the primary computer, if the primary computer fails. The dual hardware configuration gives increased assurance of the availability of radiation monitoring data to plant personnel through the RMDS workstations and to the EPDS through a data link.

In the RMDS dual-computer configuration, each computer is capable of storing dynamically created data. RMDS workstations located in the control room and TSC are interfaced to the RMDS computers to allow the access of information from the computers and process radiation monitors and to display process radiation monitors status.

The RMDS has data links that communicate with process radiation monitors located throughout the plant. The RMDS also has a data link that communicates with the EPDS. The EOF has access to RMDS data through the EPDS. The two RMDS computers communicate with each other and with the workstations over a local area network.

CHAPTER 11 11.5-31 REV. 19 SEPTEMBER 2018

LGS UFSAR The RMDS provides process radiation monitors status and RMDS status through annunciation of alarm conditions at the RMDS display consoles to supplement the process radiation monitors alarms at the plant annunciator panel. The operator is alerted of abnormal conditions through visual alarms located at each RMDS display console. Trend displays of the process radiation monitors data are available as demand prints are available for display or hard copy at the RMDS display consoles.

11.5.6.3 RMMS Software Description 11.5.6.3.1 RMDS Software Description The RMDS software is designed to help the plant operator display and retrieve information from the RMDS and the process radiation monitors memory, and to initiate process radiation monitors control functions. The RMDS software also handles the communications interface between the process radiation monitors and the RMDS and between the RMDS and the EPDS.

The RMDS software consists of the following programs:

a. Operator interface programs to allow operator communications via the RMDS workstations

b. Alarm log and display programs

c. System fail-over software

d. Communications software and control programs.

e. Collect, quality check (evaluate), and report radiological effluent release data from onsite monitors 11.5.6.3.2 Effluent Release Software Description The Effluent Dose Calculation Computer Program is designed to help the plant operators and health physicists meet NRC regulatory requirements for measuring, evaluating, and reporting releases of radioactive materials in gaseous and liquid effluents and for classifying and reporting the categories and the curie content of radioactive solid wastes. The software functions are grouped into the following two categories:

a. Generate a data base and store data in a retrievable form

b. Compute short-time or accumulated offsite radiation doses from various effluent discharge pathways caused by the release of gaseous and/or liquid radioactive materials.

In performing the above functions, the procedures used adhere to specific guidelines.

Calculations of liquid dispersion use either site specific factors or equations that conform to the standards of Regulatory Guide 1.113; and calculations of radiation doses conform to Regulatory Guide 1.109 (Rev 1). Effluent release summaries and the environmental pathway dose summary are prepared in accordance with Regulatory Guide 1.21 (Rev 1). The above regulatory guides and 10CFR50, Appendix I were used to develop the Effluent Dose Calculation software.

CHAPTER 11 11.5-32 REV. 19 SEPTEMBER 2018

LGS UFSAR The functions that the Effluent Dose Calculation software performs are related to routine operations. The operator is provided with the capability to inspect the contents of the data base and to perform dose calculations. The operator can perform both gaseous and liquid effluent calculations. The method used to perform gaseous calculations is based on hourly averages of meteorological and effluent data. Liquid calculations are based on batch release data.

To satisfy NRC reporting requirements for routine operation, the Effluent Dose Calculation Software has the capability to generate two reports:

a. An effluent release summary, describing releases of gaseous and liquid effluents and shipments of solid waste

b. An environmental pathway dose summary, showing total body and body organ doses by age group for environmental pathways for gaseous and liquid effluents.

The dose assessment model performs calculations of atmospheric dispersion and dose for accidental release of gaseous effluent. To perform calculations of atmospheric dispersion (X/Q values), the dose assessment model prompts the operator to enter meteorological data.

Meteorological data is used to compute the dispersion conditions for a given release point. Dose calculations can be done using radiological release data obtained automatically from the EPDS or entered manually by the operator.

The dose assessment model is described in detail in the Emergency Plan.

The results of the accidental dose calculations, in addition to being available at the control room, TSC, and EOF, are available to local, State, and Federal emergency officials.

CHAPTER 11 11.5-33 REV. 19 SEPTEMBER 2018

LGS UFSAR Table 11.5-1 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS TOTAL NO.

OF PRINCIPAL MONITORED CHANNELS RADIONUCLIDES PROCESS (BOTH UNITS) DETECTOR TYPE DETECTOR LOCATION CHANNEL RANGE SENSITIVITY MEASURED 6

Main steam 8 Gamma-ion Immediately 1 to 10 1 mrem/hr Coolant line chamber downstream of mrem/hr activation main steam gases isolation valve Reactor enclosure 8 Gamma-GM Reactor enclosure 0.01 to 102 0.01 mrem/hr Kr-85 exhaust exhaust duct mrem/hr Refueling area 8 Gamma-GM Refueling area 0.01 to 102 0.01 mrem/hr Kr-85 exhaust chamber exhaust duct mrem/hr Control room 4 Beta-scint Main intake 10-4 to 101 1x10-6 Ci/cc Kr-85 supply ventilation duct Ci/cc (Xe-133)

Control room 2 Beta-scint Emergency fresh air 10-4 to 101 1x10-3 Ci/cc Kr-85 emergency fresh air supply duct Ci/cc (Xe-133)

RHR service 2 Gamma-scint Common RHRSW/ESW 10 to 106 1.29x10-7 Ci/cc Cs-137 water Discharge Lines counts/sec (Cs-137) Co-60 (1)

North stack Normal range 6 a. Beta-scint(Part.) Cross-section of a. 10-12 to 1x10-7 Ci/cc Kr-85, I-131 subsystem b. Gamma-scint(Iodine) north stack at 10-6 Ci/cc (Xe-133) Sr-90, Y-90

c. Beta-scint(Noble Gas) el 402' b. 10-11 to 10-5 Ci/cc

c. 10-7 to 10-1 Ci/cc Wide range 3(3) Beta-scint Cross-section of 10-7 to 10-1, 1 x 10-7 Ci/cc subsystem north stack at 10-4 to 102 and (Xe-133) el 402' 10-1 to 105 Ci/cc( 3 )

CHAPTER 11 11.5-34 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.5-1 (Cont'd)

TOTAL NO.

OF PRINCIPAL MONITORED CHANNELS RADIONUCLIDES PROCESS (BOTH UNITS) DETECTOR TYPE DETECTOR LOCATION CHANNEL RANGE SENSITIVITY MEASURED South stack 12(2) Beta-scint, Cross-section of (2) 1x10-7 Ci/cc Kr-85, I-131 gamma-scint, south stack at (Xe-133) Sr-90, Y-90 particle/iodine el 402' filter Charcoal offgas 1 Beta-GM, Charcoal offgas 10 to 106 1.46x10-6 Ci/cc Kr-85 compartment particle filter compartment counts/min (Xe-133) exhaust exhaust ducts Charcoal offgas 2 Beta-GM, Charcoal offgas 10 to 106 1.46x10-6 Ci/cc Kr-85 effluent particle filter exhaust pipe counts/min (Xe-133)

Recombiner 2 Beta-GM, Recombiner compt 10 to 106 1.46x10-6 Ci/cc Kr-85 compartment particle filter and H2/O2 units counts/min (Xe-133) exhaust exhaust ducts Containment 8 Gamma-ion Inside primary 1 to 106 102 to 108 post-LOCA chamber containment counts/min rads/hr Steam 2 Gamma-GM, Downstream of 10 to 106 1.46x10-6 Ci/cc Kr-85 seal particle filter junction of counts/min (Xe-133) mechanical vacuum pump exhaust and steam seal exhaust Radwaste 1 Gamma-GM, Main exhaust duct 10 to 106 1.46x10-6 Ci/cc Kr-85 enclosure particle filter of radwaste counts/min (Xe-133) exhaust enclosure Air ejector 6 Gamma-ion SJAE discharge 10 to 106 mr/hr 1 mrem/hr Noble gas effluent chamber pipes to recombiner fission products Primary 2 Beta-scint Primary 10 to 107 1x10-6 Ci/cc Fission gas containment leak (1 per particle containment counts/min (Xe-133) daughter and detection unit) filter at el 292' 1x10-9 Ci/cc corrosion (active) (Sr-90) activation products CHAPTER 11 11.5-35 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.5-1 (Cont'd)

TOTAL NO.

OF PRINCIPAL MONITORED CHANNELS RADIONUCLIDES PROCESS (BOTH UNITS) DETECTOR TYPE DETECTOR LOCATION CHANNEL RANGE SENSITIVITY MEASURED PRINCIPAL Hot maintenance 1 Beta-scint filter Ventilation exhaust a. 10-12 to 1x10-11 Ci/cc I-131, Sr-90 exhaust shop a. particle duct from hot 10-6 Ci/cc Y-90, Co-60

b. Iodine maintenance shop b. 10-11 to 10-5 Ci/cc Radwaste 1 Gamma-scint Radwaste drain line 10 to 106 1.29x10-7 Ci/cc Cs-137 discharge to cooling tower counts/sec (Cs-137) Co-60 blowdown line Service water 2 Gamma-scint Main pipe to 10 to 106 1.29x10-7 Ci/cc Cs-137 cooling tower counts/sec (Cs-137) Co-60 Reactor enclosure 2 Gamma-scint Reactor enclosure 10 to 106 1.29x10-7 Ci/cc Cs-137 cooling water cooling water counts/sec (Cs-137) Co-60 upstream of pumps Dose rate at PASS 2 Gamma-ion PASS control panel 1 to 104 1 mrem/hr (4) control panel chamber mrem/hr PASS liquid sample 2 Gamma-ion PASS liquid 0.01 to 102 0.01 mrem/hr(4) chamber sample return mrem/hr PASS gaseous 2 Gamma-ion PASS gas sample 1 to 104 1 mrem/hr (4) sample chamber particulate/iodine mrem/hr filter (1)

Three detectors in each of two redundant systems: a) Particulate, b) Iodine, and c) Noble gas.

(2)

Four noble gas detectors (10-7 to 5x10-1 Ci/cc), four particulate detectors (10-12 to 10-6 Ci/cc), and four iodine detectors (10-11 to 10-5 Ci/cc).

(3)

Three noble gas detectors with overlapping ranges.

(4)

Not applicable.

CHAPTER 11 11.5-36 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.5-2 RADIATION MONITORING SYSTEM SETPOINTS AND FUNCTIONS LOW HIGH HIGH-HIGH MONITORED PROCESS SETPOINT(1) FUNCTION SETPOINT(1) FUNCTION SETPOINT(1) FUNCTION (2)

Main steam line Below Annunciate 1.5 x Full Annunciate 3 x Full Power Annunciate Background(6) Power Trip mechanical vacuum pump

& suction valve Reactor enclosure Below Annunciate Above Annunciate Technical Annunciate ventilation exhaust Background(6) Background(6) Specifications Trip HVAC valves Start RERS Start SGTS Trip containment purge & vent valves Refueling floor Below Annunciate Above Annunciate Technical Annunciate ventilation exhaust Background(6) Background(6) Specifications Trip HVAC valves Start SGTS Trip containment purge & vent valves (7)

Control room Annunciate Above Annunciate Technical Annunciate air supply(3) Background(6) Specifications Trip fresh air isolation valves Start emergency fresh air system (7) (11) (11)

Control room Annunciate Annunciate Annunciate emergency fresh air(3)

RHR service Below (Same as Above Annunciate Technical Annunciate water(4)(2) Background(6) high-high) Background(6) Specifications Trip RHRSW pumps North stack(3) (7)

Annunciate a.Calculated Annunciate a.Calculated Annunciate normal range per ODCM per ODCM (Noble (Noble Gas) Gas) b.3.8x10- b.7.66x10-7 Ci/cc Ci/cc (Iodine) (Iodine)

CHAPTER 11 11.5-37 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.5-2 (Cont'd)

LOW HIGH HIGH-HIGH (1) (1)

MONITORED PROCESS SETPOINT FUNCTION SETPOINT FUNCTION SETPOINT(1) FUNCTION North stack(3) (7)

Annunciate Calculated Calculated Annunciate wide range er ODCM per ODCM Trip containment (Noble Gas) purge & vent valves South stack(3) (7)

Annunciate a. Calculated Annunciate a. Calculated Annunciate per ODCM per ODCM (Noble (Noble Gas) Gas)

b. 1.1x10-8 b. 2.27x10-6 Ci/cc Ci/cc Iodine) (Iodine)

(9) (9) (9)

Radwaste equipment Annunciate Annunciate Annunciate rooms ventilation exhaust(3)

(9) (9) (9)

Charcoal treatment Annunciate Annunciate Annunciate system process exhaust(3)

(9) (9) (9)

Recombiner Annunciate Annunciate Annunciate compartment & H2 analyzers exhaust(3)

(7)(8)

Primary containment Annunciate 50 R/hr --- 1x102 R/hr Trip reactor enclosure post-LOCA sump pumps (9) (9)

Steam exhauster Annunciate 1.5xBackground Annunciate Annunciate discharge &

vacuum pump exhaust (9) (9) (9)

Radwaste enclosure Annunciate Annunciate Annunciate exhaust(3)

Air ejector/after Below (Same as 1.5xNormal Full Annunciate Technical Annunciate condenser Background(6) high-high) Power Background Specifications effluent(2)(3) (12)

CHAPTER 11 11.5-38 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.5-2 (Cont'd)

LOW HIGH HIGH-HIGH MONITORED PROCESS SETPOINT(1) FUNCTION SETPOINT(1) FUNCTION SETPOINT(1) FUNCTION (10) (10)

Primary containment Below Annunciate Annunciate Annunciate leak detection(3) Background (10)

(7) (11) (11)

Hot maintenance shop Annunciate Annunciate Annunciate exhaust(5) Locally (9)

Radwaste (Same as 7.6x10-6 Annunciate Calculated Annunciate discharge(4) high-high) Ci/cc per ODCM Trip radwaste discharge valve Service water(4) (9)

Annunciate Above Annunciate Calculated Annunciate Background per ODCM (9)

Reactor enclosure Annunciate Above(6) Annunciate Technical Annunciate cooling water(4) Background Specifications Dose rate at PASS N/A N/A N/A control panel PASS liquid sample N/A N/A N/A PASS gaseous sample N/A N/A N/A (1)

NA designation in this column means that an associated alarm is not provided.

(2)

Monitoring system inoperative alarms are also provided for this process. The alarm performs the same function as the high-high alarm.

(3)

In addition to the alarms shown, these processes also are provided with high/low flow alarms that annunciate in the control room.

(4)

In addition, low sample pump flow alarms annunciate in the control room.

(5)

This system is used for particulate/iodine monitoring only.

(6)

Setpoints are determined after normal background values are established during startup testing. The high setpoint is set at twice the background value if practicable.

(7)

Setpoints are factory set based on parameters of the detector.

(8)

A bias source is provided in this monitor to maintain minimum indication when the ambient radiation is below channel range.

(9)

Low setpoint is set below expected radioactivity concentration for normal operation. High setpoint is set at twice this value if practicable.

(10)

Setpoints are set significantly above background to prevent nuisance alarms (Section 5.2.5.2.1.5).

(11)

This monitor is used for information only. Setpoints may be set for operator information above normal radioactivity concentration values established during startup testing.

(12)

Setpoint not to exceed 2100 mR/Hr.

CHAPTER 11 11.5-39 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.5-3 POSTACCIDENT SAMPLING DOSE ASSESSMENT(1)(2)

Background Sample Integrated Dose Dose Dose Time Rate Rate Rate Liquid Sample (min) (mR/hr) (mR/hr) (Rem)

Recirculate Sample 5 722 27 0.06 Operate Sampler 5 722 112 0.07 Transport Sample 10 3766 8 0.63 Cask Handle Sample 10 sec 9 182 0.01 Analyze Sample 20 9 91 0.03 Total 0.80 Gas Sample Recirculate Sample 5 722 75 0.07 Operate Sampler 5 722 289 0.08 Handle Bottle 1 722 332 0.02 Transport Sample 10 3766 43 0.63 Cask Analyze Sample 20 9 37 0.01 Total 0.82 Total (All samples) 1.62 (1)

Doses listed are 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-LOCA, except for those associated with MSIV Leakage Alternate Drain Pathway. Worst case doses resulting from this pathway peak are at 4 days and are:

Liquid Gas Iodine/Particulate Sample Sample Sample Total WHOLE BODY(REM) 0.036 0.037 0.036 0.11 BETA SKIN(REM 0.0007 0.0008 0.0007 0.0032 THYROID (REM) 0.20 0.20 0.20 0.60 (2)

Some additional dose to extremities will result from the limited handling of samples in the laboratory. Because of the use of sample dilutions, small volume samples, shielded casks, lead brick piles, and laboratory extension devices (i.e., tongs), doses to the extremities are estimated to be 100 mR to 200 mR for each sample.

CHAPTER 11 11.5-40 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 11.5-4 GASEOUS EFFLUENTS COMPOSITION AND CONCENTRATIONS Concentrations(1)

(Ci/cc)

South Stack North Stack Noble Ar-41 7.18x10-9 (4)

Gases Kr-83m (3) (4)

Kr-85m 1.72x10-9 1.50x10-8 Kr-85 (3) 5.67x10-8 Kr-87 1.72x10-9 2.63x10-8 Kr-88 1.72x10-9 4.65x10-8 Kr-89 (3) (4)

Xe-131m (3) 1.42x10-9 Xe-133m (3) (4)

Xe-133 3.73x10-8 5.29x10-9 Xe-135m 2.64x10-8 1.32x10-7 Xe-135 1.95x10-8 2.07x10-7 Xe-137 (3) (4)

Xe-138 4.02x10-9 2.83x10-7 Halogens(2) I-131 2.87x10-11 1.62x10-11 I-133 1.15x10-10 4.05x10-11 Airborne Particulate C- 14 (3) 1.92x10-9 Cr-51 1.72x10-15 4.45x10-14 Mn-54 1.72x10-14 6.20x10-14 Co-58 3.45x10-15 1.03x10-14 Fe-59 2.30x10-15 3.14x10-14 Co-60 5.74x10-14 1.86x10-13 Zn-65 1.15x10-14 3.44x10-15 Sr-89 5.17x10-16 1.30x10-14 Sr-90 2.87x10-17 6.48x10-16 Zr-95 2.30x10-15 3.04x10-16 Sb-124 1.15x10-15 7.09x10-16 Cs-134 2.30x10-14 1.03x10-14 Cs-136 1.72x10-15 1.42x10-15 Cs-137 3.16x10-14 2.14x10-14 Ba-140 2.30x10-15 2.47x10-15 Ce-141 5.74x10-16 6.48x10-15 (1) Expected concentrations at maximum stack flow rates may increase in proportion to power.

(2) Includes both gaseous and particulate releases.

(3) Less than 1.44x10-10 Ci/cc.

(4) Less than 5.06x10-10 Ci/cc.

CHAPTER 11 11.5-41 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.5-5 LIQUID EFFLUENTS COMPOSITION AND CONCENTRATION Concentration(1)

(Ci/cc)

Floor Drain Equipment Drain Laundry Drain Isotope Sample Tank(2) Sample Tank(2) Sample Tank Br-83 1.72x10-8 6.12x10-8 -

Br-84 - 2.22x10-9 -

Br-85 - - -

I-131 6.54x10-7 8.55x10-7 9.65x10-7 I-132 1.55x10-7 5.52x10-7 -

I-133 1.38x10-6 2.52x10-6 -

I-134 4.84x10-8 1.60x10-7 -

I-135 4.91x10-7 1.38x10-6 -

Rb-89 5.50x10-9 2.45x10-9 -

Cs-134 1.07x10-7 2.67x10-7 2.09x10-5 Cs-136 6.75x10-7 1.74x10-7 -

Cs-137 2.49x10-6 6.22x10-7 3.86x10-5 Cs-138 3.88x10-7 2.33x10-7 -

Na-24 5.15x10-7 1.04x10-6 -

P-32 2.71x10-8 3.49x10-8 -

Cr-51 6.94x10-7 8.81x10-7 -

Mn-54 8.49x10-9 1.06x10-8 1.61x10-6 Mn-56 3.35x10-7 1.19x10-6 -

Fe-55 1.42x10-7 1.66x10-6 -

Fe-59 4.20x10-9 5.30x10-9 -

Co-58 2.81x10-8 3.53x10-8 6.43x10-6 Co-60 5.68x10-8 7.11x10-8 1.45x10-5 Ni-63 1.42x10-10 1.78x10-1 -

Ni-65 1.99x10-9 7.09x10-9 -

Cu-64 1.50x10-6 3.38x10-6 -

Zn-65 2.82x10-8 3.53x10-8 -

Zn-69 8.84x10-9 7.39x10-9 -

Sr-89 1.41x10-8 1.78x10-8 -

Sr-90 8.49x10-10 1.06x10-9 -

Sr-91 1.55x10-7 3.73x10-7 -

Sr-92 7.31x10-8 2.58x10-7 -

Y-91 8.14x10-9 8.90x10-9 -

Y-92 1.78x10-7 5.77x10-7 -

Y-93 1.65x10-7 3.88x10-7 -

Zr-95 9.81x10-10 1.24x10-9 2.25x10-6 Zr-97 3.14x10-10 6.08x10-1 -

CHAPTER 11 11.5-42 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 11.5-5 (Cont'd)

Concentration(1)

(Ci/cc)

Floor Drain Equipment Drain Laundry Drain Isotope Sample Tank(2) Sample Tank(2) Sample Tank Nb-95 9.95x10-10 1.24x10-9 3.22x10-6 Nb-98 3.89x10-9 8.90x10-9 -

Mo-99 2.26x10-7 3.21x10-7 -

Tc-99m 6.24x10-7 1.50x10-6 -

Tc-101 2.44x10-10 4.44x10-1 -

Tc-104 9.12x10-10 2.20x10-9 -

Ru-103 2.79x10-9 3.51x10-9 2.25x10-7 Ru-105 3.02x10-8 9.78x10-8 -

Ru-106 4.25x10-10 5.32x10-1 3.86x10-6 Ag110M 1.42x10-10 1.78x10-1 7.08x10-7 Te129M 5.58x10-9 7.06x10-9 -

Te131M 8.70x10-9 1.43x10-8 -

Te132 1.17x10-9 1.63x10-9 -

Ba-139 2.14x10-8 7.50x10-8 -

Ba-140 5.40x10-8 6.96x10-8 -

Ba-141 1.14x10-10 2.74x10-1 -

Ba-142 2.18x10-12 3.88x10-1 -

La-142 1.53x10-8 5.41x10-8 -

Ce-141 4.52x10-9 5.62x10-9 -

Ce-143 2.73x10-9 4.37x10-9 -

Ce-144 4.25x1-10 5.32x10-1 -

Pr-143 5.55x10-9 7.04x10-9 -

Nd-147 4.02x10-10 5.19x10-10 -

W-187 2.33x10-8 4.05x10-8 -

Np-239 7.59x10-7 1.10x10-6 -

(1) Concentrations are prior to dilution in the cooling tower blowdown, and may increase in proportion to power. Sample tank pump capacities are presented in Table 11.2-10.

(2) Equipment drain sample tank discharges are routed to the cooling tower blowdown line via the floor drain system.

(3) Deleted CHAPTER 11 11.5-43 REV. 16, SEPTEMBER 2012