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1 to Updated Final Safety Analysis Report, Chapter 11, Radioactive Waste Management
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Section 11.1 11.1.1 11.1.2 11.1.3 11.1.4 11.1.5 11.. 1.5.1 11.1.5.2 11.1.5.. 2.. 1 11.1.5.. 2.. 2 11.1.. 5.2.3 11.1.5.2.4 11.. 1.5.2.5 11.1.5.2.6 11.. 1.6 11.1.7 11.1.8 11.1.9 11.2 11.2.1 11.2.2 11.2.3 11.2.4 11.2.5 11.2.6 11.2.7 SGS-UFSAR SECTION 11 RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS SOURCE TERMS Determination of Activity in Reactor core Activities in the Fuel Rod Gap Fuel Handling Sources Reactor COolant Fission Product Activities Tritium Production General - overall sources Specific Individual Sources of Tritium Ternary Fissions -

Clad Diffusion Tritium Produced from Boron Reactions Tritium Produced from Lithium Reactions Control Rod Sources 11.1-8 Tritium Production from Deuterium Reactions Total Tritium Sources Volume Control Tank Activity Gas Decay Tank Activity Activity in Recirculated Sump Water References for Section 11.1 LIQUID WASTE SYSTEM Design Objectives

System Description

System Design Operating Procedures Performance Tests Estimated Releases Release Points 11-i 11.1-1 11.1-1 11.1-1 11.1-3 11.1-4 11.1-6 11.1-6 11.. 1-6 11.1-6 11.1-7 11.1-8 11.1-9 11.1-9 11.1-10 11.1-10 11.1-11 11.1-11 11.2-1 11.2-1 11.2-2 11.2-5 11.2-10 11.2-10a 11.2-11 11.2-12 Revision 16 January 31, 1998

Section 11.2.8 11.2.9 11.2.10 11.3 11.3.1 11.3.2 11.3.3 11.3.4 11.3.5 11.3.6 11.3.7 11.3.8 11.3.9 11.4 11.4.1 11.4.2 11.4.2.1 11.4.2.1.1 11.4.2.1.2 11.4.2.2 11.4.2.3 11.4.2.4 11.4.3 11.4.4 11.5 11.5.1 11.5.2 SGS-OFSAR TABLE OF CONTENTS (Cont)

Dilution Factors Estimated Doses (Potential)

Reference for Section 11.2 GASEOUS WASTE SYSTEM Design Objectives

System Description

System Design Operating Procedures Performance Tests Estimated Releases Release Points Dilution Factors Estimated Doses RADIOLOGICAL MONITORING Design Objectives Radiation Monitoring System Radiation Monitoring System Description Radiation Monitoring System - Unit 1 Radiation Monitoring System - Unit 2 Process Radiation Monitoring System Channel Description Process Filter Monitoring System Channel Description Area Monitoring System Channel Description Sampling Inservice Tests and Calibrations SOLID RADWASTE SYSTEM 11.5-1 Design Objectives

System Design

11-ii 11.2-12 11.2-13 11.2-17 11.3-1 11.3-1 11.3-2 11.3-6 11.3-13 11.3-15 11.3-15 11.3-16 11.3-16 11.3-16 11.4-1 11.4-1 11.4-1 11.4-3 11.4-3 11.4-5 11.4-7 11.4-17 11.4-18 11.4-21 11.4-23 11.5-1 11.5-2 Revision 23 October 17, 2007

Section 11.5.3 11.5.4 11.5.5 11.5.6 11.5.7 11.5.8 11.6 11.6.1 11.6.2 11.6.3 11.6.4 SGS-UFSAR TABLE OF CONTENTS (Cent)

Equipment Description Expected Volumes Packaging Storage Facilities Shipment Reference for Section 11.5 OFFSITE RADIOLOGICAL MONITORING PROGRAM Program Objective Preoperational and Operational Programs Expected Pathways Physical Characteristics of Samples 11-iii 11.5-2 11.5-3 11.5-3 11.5-3 11.5-3 11.5-4 11.6-1 11.6-1 11.6-2 11.6-2 11.6-2 Revision 6 February 15, 1987

Table 11.1-1 11.1-2 11.1-3 11.1-4 11.. 1-5 11.1-6 11.1-7 11.1-8 11.. 1-9 11.1-10 11.. 1-11 11.1-12 SGS-UFSAR LIST OF TABLES Core Activities Deleted Deleted Deleted Deleted Deleted Parameters Used in the Calculation of Reactor Coolant Fission Product Activities Reactor Coolant Equilibrium Fission and Corrosion Product Activities Tritium Production in the Reactor Coolant Tritium Sources from the Reactor Employing Ag-In-Cd Absorber Rods Volume Control Tank Activities Gas Decay Tank Activity 11-iv Revision 16 January 31, 1998

Table 11.1-13 11.1-14 11.1-15 11.2-1 11.2-2 11.2-3 11.2-4 11.2-5 11.2-6 11.2-7 11.2-8 11.2-9 SGS-UFSAR LIST OF TABLES (Cent)

Deleted Deleted Deleted Liquid Waste System Performance Data Estimated Annual Liquid Discharge to Waste Disposal (per unit)

Waste Disposal Component'S Code Requirements Component Data Summary Estimated Annual Liquid Discharge to Liquid Waste System Estimated Annual Liquid Release by Isotope {Two Unit Basis)

Concentration of Radionuclides in Fish Due to Plant Operation Concentration of Radionuclides in Blue Crabs Due to, Plant Operation Potential Radiation Exposure Pathways to Man {Liquid) (Based on 0.2 percent failed fuel defects) 11-v Revision 23 October 17, 2007 I

I I

Table 11.3-1 11.3-2 11.3-3 11.3-4 11.4-1 11.4-2 11.4-3 11.4-4 11.6-1 11.6-2 SGS-UFSAR LIST OF TABLES (Cent)

Estimated Annual Gaseous Release by Isotope from Gas Decay Tanks -

per Unit Estimated Total Radioactive Gaseous Releases -

Two Unit Basis Estimated Off-Site Radiation Exposures Releases -

Two Unit Basis Gaseous Radioactive Potential Radiation Exposure Pathways to Man {Gaseous)

Unit 1 Process Radiation Monitoring System Unit 2 Process Radiation Monitoring System Unit 1 Area Radiation Monitoring System Unit 2 Area Radiation Monitoring System Deleted Summary of Radionuclide Concentrations in Artificial Island Preoperational Radiological Environmental Monitoring Program Samples 11-vi Revision 15 June 12, 1996

Figure

11. 2-1A 11.2-18 11.3-1A 11.3-18 11.4-1 11.4-2 11.4-3 11.4-4 11.4-5 11.4-6 11.4-7 11.4-8 SGS-UFSAR LIST OF FIGURES Title Deleted:

Refer to Plant Drawing 205239 Deleted:

Refer to Plant Drawing 205339 Deleted:

Refer to Plant Drawing 205240 Deleted:

Refer to Plant Drawing 205340 Unit No. 2 Overall Radiation Monitoring System Makeup Unit No. 1 Air Particulate, Iodine and Gas Monitor (Typical)

Unit No. 2 Air Particulate, Iodine and Gas Monitor (Typical)

Unit No. 1 Liquid Monitor (Typical)

Unit No. 2 Liquid Monitor (Typical)

DELETED Unit No. 1 Area Monitor (Typical)

Unit No. 2 Area Monitor (Typical) 11-vii Revision 27 November 25, 2013

SECTION 11 RADIOACTIVE WASTE MANAGEMENT The purposes of this section performance evaluation of the are 1) to provide a complete description Radioactive Waste Treatment Systems and 2) and to demonstrate

that, under normal plant the anticipated operational will be in conformance regulations.

occurrences, with applicable operating conditions and during radioactive releases from the plant Nuclear Regulatory Commission (NRC) 11.1 SOURCE TERMS Source terms describe the quantity, chemical species, and timing of radioactive materials released from the core and/or primary coolant system for a specific accident.

The amounts of radioactive materials which are produced and stored in the reactor system are discussed in this section.

These sources have been calculated for the design basis accidents and for normal operation using the ORIGEN computer code for nuclide concentrations and activities from fuel depletion.

11.1.1 Determination of Activity in Reactor Core The total core activity was originally calculated for Salem with the ORIGEN computer code.

The alternative source term (AST, Ref.

1) analysis required additional radionuclide groups than originally calculated by ORIGEN.

The numerical values for core activity (60 isotopes) used in the AST LOCA analysis in UFSAR Section 15.4.1 are shown in Table 11.1-1 and were obtained from the default nuclide inventory file from the RADTRAD (3.02 and 3.03) computer program (Refs. 5, 6 & 7).

The LOCA dose analysis is based on these 60 isotope activities and a core thermal power of 3632 MWt

(=3459 x 1.05).

The aerosol inventory is multiplied by 1.10 in some analysis applications to make the RADTRAD default file conservative with respect to the SNGS plant-specific ORIGEN inventory file.

11.1-1 SGS-UFSAR Revision 28 May 22, 2015

I 11.1.2 Activities in The Fuel Rod Gap

'l'he activity contained in the space (gap) between the fuel pellets and the cladding is released when the cladding is breached.

Rod failure is typically caused by high fuel temperature and primary system depressurization.

'l'he fraction of core activity assumed to be in the gap can vary depending on the specific application.

Gap activity is the primary source term for the locked rotor, rod ejection and fuel handling accidents.

The gap activity basis is discussed as part of the assumptions described in the specific accident section of Chapter 15.

11.1-2 SGS-UF'SAR Revision 23 October 17, 2007

11.1.3 Fuel Handling Sources The inventory of fission products in a fuel assembly is dependent on the power rating of the assembly.

The parameters used for calculation of the highest rated assembly in the core to be discharged are provided in Section 15.4.6.

11.1-3 SGS-UFSAR Revision 20 May 6, 2003 I

11.1.4 Reactor Coolant Fission Product Activities The parameters used in the calculation of the reactor coolant fission product concentrations, including pertinent information concerning the expected coolant cleanup flow rate, demineralizer effectiveness, and volume control tank noble gas stripping behavior, are presented in Table 11.1-7. The results of calculations are presented in Table 11.1-8.

The table lists nuclides of fission and corrosioin products which are significant from a shielding standpoint as well as those nuclides which are listed in ANS standard ANSI/ANS-18.1-1984.

The values tabulated are the maxLmums that occur during the fuel cycle from startup through the equilibrium cycle.

In these calculations, small cladding defects in the equivalent of one percent of the fuel rods are assumed to be present at the initial core loading and uniformly distributed throughout the core.

Similar defects are assumed to be present in all reload regions. The fission product escape rate coefficients are, therefore, based upon an average fuel temperature.

The fission product activity in the reactor coolant during operation with defects in the cladding of the fuel rods is computed using the following differential equations:

For parent nuclides in the coolant:

SGS-UFSAR 11.1-4 Revision 16 January 31, 1998

For daughter nuclides in the coolant:

Where:

He NF t

R F

v Me l

OF QL

f

=

s

=

Concentration of nuclide in the reactor coolant (atoms/gram)

Inventory of nuclide in the fuel (atoms)

Operating t~e (seconds)

Nuclide release coefficient (1/sec) -F

v Fraction of fuel rods with defective cladding Fission product escape rate coefficient (1/sec)

Mass of reactor coolant (grams)

Nuclide decay constant (1/see)

Nuclide demineralizer decontamination factor Purification or letdown mass flow rate (grams/sec)

Nuclide volume control tank stripping fraction Fraction of parent nuclide decay events that result in the formation of the daughter nuclide D = Dilution coefficient for feed and bleed (1/sec) = :

p 1

BD-p. I DF B0 = Initial boron concentration (ppm) p = Boron concentration reduction rate (ppm/sec) and where:

SGS-UFSAR subscript i refers to the parent nuclide subscript j refers to the daughter nuclide 11.. 1-5 Revision 16 January 31, 1998

11.1.5 Tritium Production 11.1.5.1 General - overall sources Tritium is formed from several sources, the most abundant of which is the fissioning of uranium, which yields tritium as a ternary fission product.

Tritium atoms are generated in the fuel at a rate of approximately 8 x 10-5 atoms per fission, or 1.05 x 10-2 curies/mwt/day. Boron-bearing control rods can also be a potential source of tritium. These potential sources of tritium are only present in the reactor coolant to the extent that they diffuse through the fuel or control rod cladding.

A direct source of tritium is the reaction of neutrons with dissolved boron in the reactor coolant. Neutron reactions with lithium are also a direct source of tritium. Lithium is present in a pressurized water reactor (PWR) for pH control and as a product of boron reactions with neutrons. An extremely small amount of tritium is also produced by neutron reactions with naturally occurring deuterium in light water.

11.1.5.2 Specific Individual Sources of Tritium 11.1.5.2.1 Ternary Fissions - Clad Diffusion Because of the mode of operation of the PWR to minLmize any liquid or gaseous discharges from the plant, it has been possible to very accurately determine the buildup of tritium from various sources in the plant and to identify their origin.

A program was undertaken by Westinghouse to determine the source of tritium in the reactor coolant in operating plants with both stainless steel and zircaloy cladding.

This program clearly indicated that with the current generation of Westinghouse reactors with zircaloy-clad fuel, 1 percent or less of the tritium produced in the fuel will diffuse through the cladding into the coolant.

For those plants containing stainless steel cladding, operational data have shown that as high as 80 percent of the ternary tritium produced will diffuse through the cladding.

The tritium concentration in the reactor coolant in those plants having stainless steel and zircaloy fuel cladding has been substantially different.

Tritium concentrations at Yankee-Rowe (600 MWt), which has stainless steel cladding, has ranged from 11.1-6 SGS-UFSAR Revision 16 January 31, 1998

about 4.5 to 5 ~c/cc essentially throughout the core cycle.

A total discharge from the plant during the core cycle of -1500 curies of tritium was reported in the monthly operating reports.

In addition, with the stainless cores, there has been a continuing source of tritium to the reactor coolant during the power coastdown period when all the boric acid has been removed from the system.

This information, in particular, substantiates the high tritium diffusion through the stainless steel clad.

The experience at the R. E. Ginna plant has been substantially different. This plant operates at 1455 MWt and has zircaloy cladding. After approximately 8 months of operation at Ginna, the tritium concentrations were less than 0.3 ~c/cc in the reactor coolant and the monthly discharges averaged

-5 curies/month.

The experiences at Benzau and Zori ta were comparable.

An extensive program to follow the buildup of tri tiurn in the Ginna plant was initiated, and the results indicated a potential source from the core which is 1 percent or less of the ternary fissions generated in the fuel.

Based on this experience, the tritium sources during the operation of a PWR can be very accurately predicted.

In the past, Westinghouse has assumed that 30 percent of the tritium from ternary fissions would diffuse through the zircaloy fuel.

This value was used as a basis for systems and operational design and is clearly conservative.

11.1.5.2.2 Tritium Produced from Boron Reactions The neutron reactions with boron resulting in the production of tritium are:

10 B

(n, 2a) T 10 ( ). 7 B

n, a Ll

{n, na) T 11 9

B (n, T) Be SGS-UFSAR 11.1-7 Revision 6 February 15, 1987

(n, a) T Of the above reactions, only the first two contribute signi to the tritium production in a PWR.

The B11 (n, T) Be 9 reactions have a threshold of 14 Mev and a cross section of "'5 mb.

Since the number of neutrons produced at this energy is less than 10 9 n/cm2-sec the tritium produced from this reaction is negligible.

The B10 (n, d) reaction may be neglected since has been found to be unstable.

11.1.5.2.3 Tritium Produced from Lithium Reactions The neutron reactions with lithium in the production of tritium are:

Li 7 (n, na) T Li 6 (n, a) T In Westinghouse designed reactors, li thi urn is used for pH adjustment of the reactor coolant.

The reactor coolant is maintained at a maximum state level of 3. 5 +/- 0.15 ppm lithium by the addition of Li 70H and by a cation demineralizer included in the Chemical and Volume Control demineralizer will remove any excess of lithium such as could be the B10 (n, a) Li 7 reaction.

The Li 6 (n, a) T reaction is controlled by limiting the impurity in the This in OH used in the reactor coolant and by lithiating the demineralizers with 99.9 atom percent 11.1.5.2.4 Control Rod Sources In a fixed burnable poison rod, there are two primary sources of tritium generation:

the B10 {n, 2a) T and the P10 (n, a) Li 7 (n, na) T reactions.

Unlike the coolant where the level is controlled at a maximum steady state level of 3.5 +/- 0.15 ppm, there is a buildup of 11.1-8 SGS-UFSAR in the burnable Revision 24 May 11, 2009

poison rod.

The burnable poison rods are required during the first year of operation only. During this time the tritium production is 72 curies/pound a10

There are no tritium sources in Ag-In-Cd control rods.

11.1.5.2.5 Tritium Production from Deuterium Reactions Since the fraction of naturally occurring deuterium in water is less than 0.0015, the tritium produced from this reaction is negligible (less than 1 curie per year) 11.1.5.2.6 Total Tritium Sources Tritium sources released to the reactor coolant are listed in Table 11.1-9, based on 12 months of operation at full power (3558 MWt} and a 0.8 load factor.

Included in Table 11.1-9 is the amount of tritium produced in the reactor for all of the nuclear reactions described above.

Two columns of values of tritium released to the reactor coolant are given in Table 11.1-9, namely, a design value and an expected value.

The design values are based on a release of 30 percent of the tritium produced being diffused through the fuel cladding.

The present values are based on operating experience at existing PWR facilities where the data have indicated that the previous design values were unduly conservative. Based on this experience, the tritium released to the reactor coolant for a typical 3558 MWt reactor is reduced from -3815 to

-690 curiesjyear.

Basic parameters employed to calculate the tritium inventory are given in Table 11.1-10.

11.1-9 SGS-UFSAR Revision 6 February 15, 1987

11.1.6 Volume Control Tank Activity The radiation sources in the volume control tank (VCT) are basad on a nominal operating level in the tank of 200 cubic feet in the liquid phase and 200 cubic feet in the vapor phase, and on the stripping fractions given in Table 11.1-7, assuming no VCT purge.

Table 11.1-11 lists the activities for the vapor phase of the VCT with clad defects in 1 percent of the fuel rods.

11.1.7 Gas Decay Tank Activity The isotopic maximum inventories are determined in the RCS and VCT.

Since there is no continuous purge from the vol~e control tank, the activity values are obtained based on the following considerations:

At shutdown the radiogas inventory of the VCT is instantaneously transferred to the GOT.

The RCS (operating with one percent fuel defects) is stripped to the VCT at the maximum letdown with a stripping fraction of 1.0 over a 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months period at the end of which the VCT is again instantaneously transferred to the GOT.

The radioactive sources are calculated at the point where Xr-88 is a maximum in the GDT.

This provides a lLmiting gamma source.

The GDT activities for noble gas nuclides are presented in Table 11.1-12.

11.1-10 SGS-UFSAR Revision 16 January 31, 1998

11.1.8 Activity in Recirculated Sump Water The concentration of iodine isotopes in the recirculation loop at initiation of recirculation phase after the design basis loss-of-coolant accident (LOCA) was replaced with an al terna ti ve source term (AST) pursuant to Section 50. 67 of Title 10 of the Code Of Federal Regulations

( 10 CFR 50. 67), "Accident Source Term",

and the potential radiological consequences were re-evaluated.

The guidance provided in Regulatory Guide 1.183, Alternative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2002, was used in the re-evaluation.

The sump water volume is 328,600 gallons and 40% of the total core value of Iodine (in accordance with Reg. Guide 1.183, Table 2) is released to the containment sump.

Of this total, 4. 85%, which is elemental, is subject to becoming airborne in proportion to a flashing rate.

The remainder is assumed to remain waterborne since the sump water pH is maintained> 7.

The radioactivity in the containment would be an additional source of radiation to the auxiliary building following a LOCA.

The residual heat removal loop source and the containment source are used to calculate post-accident radiation doses in the Auxiliary Building.

The radioactivity leaking out of the recirculation flow path in the Auxiliary Building is Engineered Safety Features (ESF) Leakage and is assumed to be a total of 0.45 gpm in the Section 15.4.1 accident analysis.

11.1.9 References for Section 11.1

1.

Regulatory Guide 1.183, Alternative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2002

2.

CCC-217, ORIGEN Isotope Generator and Depletion Code, Matrix Exponential Method, April, 1975.

3.

Radioactive Source Term for Nominal Operation of Light Water Reactors, ANSI/ANS-18.1-1984, American Nuclear Society, December, 1984.

4.

Section 50. 67 of Title 10 of the Code Of Federal Regulations

( 10 CFR 50.67), "Accident Source Term"

5.

"RADTRAD:

A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", NUREG/CR-6604, USNRC, April 1998

6.

"RADTRAD:

A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", NUREG/CR-6604 Supplement 1, USNRC, June 8, 1999

7.

"RADTRAD:

A Simplified Model for RADionuclide Transport and Removal and Dose Estimation,"

W.C.

Arcieri (ITSC),

NUREG/CR-6604 Supplement 2

(October 2002) 11.1-11 SGS-UFSAR Revision 28 May 22, 2015

TABLE 11.1-1 CORE ACTIVITIES Isotope Core Inventory Isotope Core Inventory cur:i.es/MWt cu:ries/MWt C0-58 2.553E+02 TE-131M 3.707E+03 C0-60 1.953E+02 TE-132 3.690£+04 KR-85 3.056E+02 I-131 2.750E+04 KR-85M 7.222E+03 I-132 3.889E+04 KR-87 1.306E+04 I-133 5.556E+04 KR-88 l.B61E+04 I-134 6.111E+04 RB-86 1.496E+01 I-135 5.278E+04 SR-89 2.844E+04 XE-133 5.556E+04 SR-90 1.535E+03 XE-135 1.389E+04 SR-91 3.656E+04 CS-134 3.425E+03 SR-92 3.805E+04 CS-136 1.042E+03 Y-90 1.647E+03 CS-137 1.915E+03 Y-91 3.465E+04 BA-139 4.976E+04 Y-92 3.819E+04 BA-140 4.924E+04 Y-93 4.320E+04 LA-140 5.032E+04 ZR-95 4.377E+04 LA-141 4.615E+04 ZR-97 4.562E+04 LA-142 4.449E+04 NB-95 4.138E+04 CE-141 4.476E+04 M0-99 4.830E+04 CE-143 4.352E+04 TC-99M 4.169E+04 CE-144 2.697E+04 RU-103 3.598E+04 PR-143 4.273E+04 RU-105 2.340E+04 ND-147

1. 911E+04 RU-106 8.175E+03 NP-239 5.120E+05 RH-105 1.621E+04 PU-238 2.902E+Ol SB-127 2.208E+03 PU-239 6.545E+OO SB-129 7.820E+03 PU-240 8.254E+OO TE-127 2.132E+03 PU-241 1.390E+03 TE-l 27M 2.823E+02 AM-241 9.181E-01 TE-129 7.341E+03 CM-242 3.514E+02 TE-129M 1.935E+03 CM-244 2.056E+01 The LOCA dose analysis (Section 15.4.1) is based on these 60 isotope activities, which are from the RADTRAD default file.

A core thermal power of 3632 MWt

(=3459 x 1.05) was used.

Aerosol inventory is multiplied by 1.10 to make the RADTRAD default file conservative with respect to the SNGS plant-specific ORIGEN inventory file.

1 of 1 SGS-UFSAR Revision 23 October 17, 2007

TABLE 11.1-2 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 16 January 31, 1998

TABLE 11.1-3 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 16 January 31, 1998

TABLE 11.1-4 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 16 January 31, 1998

TABLE 11.1-5 THIS TABLE HAS BEEN DELETED SGS-UFSAR 1 of 1 Revision 16 January 31.1998

TABLE ll.l-6 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 16 January 31, 1998

TABLE 11.1-7 PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT FISSION PRODUCT ACTIVITIES( 1)

1.

2.

3.

4.

5.

6.

7.

Core thermal power, max. calculated, MWt Fraction of fuel containing clad defects 3

Reactor coolant liquid volume, ft Reactor coolant average temperature, °F Purification flow rate (normal), gpm Effective cation demineralizer flow, gpm Volume control tank volumes 3

a.

Vapor, ft b

.. d f 3 L1.qu1.,

t

8.

Fission product escape rate coefficients:

Noble

isotopes,

-1

a.

gas sec

b.

I, and Cs isotopes,

-1 Br, sec Te isotopes,

-1

c.

sec

d.

Mo isotopes,

-1 sec and

isotopes,

-1

e.

Sr Ba sec

f.

La, Pr isotopes,

-1 Y,

Ce, sec

9.

Mixed bed demineralizer decontamination factors:

10.

a.

Noble gases and Cs-134, 136, 137, Y-90, 91 and Mo-99

b.

All other isotopes Cation bed demineralizer decontamination factor for CS-134, 236, 237, Y-90, 91, and Mo-99 3600 0.01 10,892 (2) 568 77 7.5 200 200 6.5 X 10-8 1.3 X 10-8 1.0 X 10-9 2.0 X 10- 9 1.0 X 10-11 1.6 X 10-12 1.0 10.0 10.0

{1)

The RCS volume was increased by the introduction of new Unit 2 AREVA NP Model 61/19T steam generators.

However, the prior analysis remains conservative, therefore, the prior volumes and activities have not been adjusted.

{2)

Conservatively bounds 20% tube plugging in Series 51 steam generator and 10% tube plugging in Model-F steam generator.

1 of 2 SGS-UFSAR Revision 24 May 11, 2009

TABLE 11.1-7 (Cont.)

11.

Volume control tank noble gas stripping fraction (closed system):

Isotope Kr-85 Kr-Ssm Kr-87 Kr-88 Xe-133 xe-133m xe-135 Xe-135m xe-138 (No purge of VCT assumed in the calculation.

account only for radioactive decay.)

2 of 2 SGS-UFSAR Stripping Practi.on s.s x 10-s 5.4 X l0-1 8.0 X l0-1 6.5 X 10-1 2.9 X 10-2 6.7 X 10-2 2.9 X 10-1 9.4 X 10-1 9.4 X 10-1 Threfore, stri.pping fractions I Revision 16 January 31, 1998

Br-84 Rb-88 Rb-89 Sr-89 Sr-90 Sr-91 Sr-92 Y-90 Y-91 Y-92 Zr-95 Nb-95 Mo-99 I-131 I-132 I-133 I-134 I-135 Te-132 Te-134 Cs-134 TABLE 11.1-8 REACTOR COOLANT EQUILIBRIUM FISSION AND CORROSION PRODUCT ACTIVITIES( 1 )

(Based on Parameters Given in Table 11.1-7)

Activity gCi/gram 4.7 X 10-2 4.8 2.1 X 10-1 4.3 X 10-3 1.2 X 10-4 6.2 X 10-3 1.3 X 10-3 3.4 X 10-5

-4 5.7 X 10 1.2 X 10-3 6.5 X 10- 4 6.5 X 10-4 7.5 X 10-1 2.8 2.8 4.2 5.7 X 10-1 2.3 2.9 X 10-1 3.0 X 10-2 2.3 Isotope Cs-136 Cs-137 Cs-138 Ba-140 La-140 Ce-144 Pr-144 Kr-85 Kr-85m Kr-87 Kr-88 Xe-133 Xe-133m Xe-135 Xe-135m Xe-138 Mn-54 Mn-56 Co-58 Co-60 Fe-59 Activity gCi/gram 2.9 1.5 9.6 X 10-1 4.2 X 10-3 1.4 X 10-3 3.9 X 10-4 3.9 X 10-4 8.2 1.7 1.0 3.0 260 17.0 8.5 4.9 X 10-1 6.1 X 10-1 4.4 X 10-4 2.0 X 10-2 1.5 X 10-2 1.9 X 10-3 5.2 X 10-4 (1)

The RCS volume was increased by the introduction of new Unit 2 AREVA NP Model 61/19T steam generators.

However, the prior analysis remains conservative, therefore, the prior volumes and activities have not been adjusted.

SGS-UFSAR 1 of 1 Revision 24 May 11, 2009 I

TABLE 11.1-9 TRITIUM PRODUCTION IN THE REACTOR COOLANT Released to the Coolant Tritium Source Total Produced Ternary Fissions 10,926 Burnable Poison Rods (Initial Cycle) 973 Soluble Poison Boron (Initial Cycle) 397 (Equilibrium Cycle) 556 Li-7 Reaction 11 Li-6 Reaction 6

Deuterium Reaction 1

Totals Initial Cycle 12,314 Totals Equilibrium Cycle 11,500 (1) Weight of B2o3 = 221 (B10 = 13.58 )

Design Expected Value Value 3240 287 397 556 11 6

1 3942 3814 107 10 397 556 11 6

1 532 685 (2) Initial boron (hot, full power, equilibrium xenon) = 860 ppm (3) Initial boron (hot, full power, equilibrium xenon) = 1200 ppm 1 of 1 SGS-UFSAR Revision 6 February 15, 1987

TABLE 11.1-10 TRITIUM SOURCES FROM THE REACTOR EMPLOYING Ag-In-Cd ABSORBER RODS Basic Assumptions and Plant Parameters:

1.

2.

3.

4.

Core thermal power Plant load factor Core volume Core volume fractions

a.

uo 2

b.

Zr + SS

c.

H20

5.

Initial reactor coolant boron level

a.

Initial cycle

b.

Equilibrium cycle

6.

Reactor coolant volume

7.

Reactor coolant transport times

a.

In-core

b.

Out-of-core

8.

Reactor coolant peak steady state lithium level (99 pure Li7 )

9.

Core averaged neutron fluxes:

a.

E > 6 Mev

b.

E > 5 Mev

c.

3 Mev ~ E ~ 6 Mev

d.

1 Mev s E s 5 Mev

e.

E < 0.625 ev

10.

Neutron reaction cross-sections

a.

s10 (n, 2a) T:

cr(1 Mev s E s 5 Mev}

spectrum cr(E > 5 Mev) =

b.

Li7 (n, naV) T: cr(3 Mev s E s 6 Mev) cr(E > 6 Mev) 1 of 2 SGS-UFSAR 3558 MWt 0.8 1153 ft 3 0.3052 0.1000 0.5948 840 ppm 1100 ppm 12,560 ft 3 0.77 sec 10.87 sec 3.5 +/- 0.15 2

n/cm -sec 2.91 X 1012 7.90 X 1012 2.26 X 1013 5.31 X 1013 2.26 X 1013 31.6mb ppm (spectrum weighted)

> 5 rob 39.1 mb (spectrum weighted) 400 rob Revision 24 May 11, 2009

TABLE 11.1-10 (Cont)

11.

Fraction of ternary tritium diffusing through zirconium cladding Note:

a.

Design value

b.

Expected value 0.30 0.01 Although Unit 1 has Westinghouse Model-F and Unit 2 has AREVA NP Model 61/19T steam generators, the radioactivity values of Unit 1 and Unit 2 are bounded by the values in this Table.

The values contained in this Table were based on the original Westinghouse Series 51 steam generators.

2 of 2 SGS-UFSAR Revision 24 May 11, 2009

TABLE 11.1-11 VOLUME CONTROL TANK ACTIVITIES( 1)

Assumptions are given previously under reactor coolant (Table 11.1-7)

Total Activity Isotope (Curies)

Kr-83m 1.7 X 101 Kr-85 6.2 X 101 Kr-85m 1.0 X 102 Kr-87 2.7 X 101 Kr-88 1.4 X 102 Xe-131m 1.9 X 102 Xe-133 2.4 X 104 Xe-133m 1.5 X 103 Xe-135 6.6 X 102 Xe-135m 5.0 X 101 Xe-137 2.8 X 10-1 Xe-138 3.4 X 10° (1)

The RCS volume was increased by the introduction of new Unit 2 AREVA NP Model 61/19T steam generators.

However, the prior analysis remains conservative; therefore, the prior volumes and activities have not been adjusted.

1 of 1 SGS-UFSAR Revision 24 May 11, 2009

Assumptions:

SGS-UFSAR TABLE 11.1-12 GAS DECAY TANK ACTIVITY Volume of the tank immaterial to this calculation.

Clad Defects in one percent of fuel rods.

Operation at 3600 MWt for 497 days.

Tank contains entire gaseous activity stripped off from the Reactor Coolant System.

3 Reactor Coolant System Volume is 12,446 ft.

Isotope Kr-85 Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 Xe-133m Xe-135 1 of 1 Total Activity Curies 1.5 X 103 1.2 X 102 1.8 X 101 1.5 X 102 2.9 X 102 3.5 X 104 2.2 X 103 8.6 X 102 Revision 23 October 17, 2007 I

I

TABLE 11.1-13 This Table Has Been deleted 1 of 1 SGS-UFSAR Revision 23 October 17, 2007

TABLE 11.1-14 This Table Has Been Deleted 1 of 1 SGS-UFSAR Revision 23 October 17, 2007

TABLE 11.1-15 This Table Has Been deleted 1 of 1 SGS-UFSAR Revision 23 October 17, 2007

11.2 LIQUID WASTE SYSTEM The Liquid Waste System (LWS) provides controlled handling and disposal of liquid wastes generated during plant operation.

The system is designed to minimize exposure to plant personnel and the general public, in accord with Nuclear Regulatory Commission (NRC) regulations.

11.2.1 Design Objectives The design objectives of the LWS are the following:

1.

Maintain annual activity releases within the limits specified in 10CFR20

2.

Protect the public health and safety by maintaining radioactive releases as low as practicable

3.

Collect radioactive and potentially radioactive liquid wastes

4.

Provide processing of liquid wastes such that operation and availability of the stations are not limited

5.

Assure that exposures to the public are maintained below the design objectives set by Appendix I to 10CFR50 The design criteria for the LWS areas follows:

The facility design shall include those means necessary to maintain control over the plant radioactive liquid effluents.

Appropriate holdup capacity shall be provided for retention of

liquid, can be release particularly where unfavorable environmental conditions expected to require operational limitations upon the of radioactive effluents to the environment.

In all cases, the design for radioactivity control shall be justified 1) on the basis of 10CFR20 requirements, for both normal operations 11.2-1 SGS-UFSAR Revision 6 February 15, 1987

and for any transient situation that might reasonably be anticipated to occur and 2}

on the basis of 10CFR50. 67 dosage level limits for potential reactor accidents of exceedingly low probability of occurrence.

Liquid facilities are designed so that discharge of effluents and offsite shipments are in accordance with applicable governmental regulations.

Radioactive fluids entering the LWS are collected in tanks until determination of subsequent treatment can be made.

They are sampled and analyzed to determine the quantity of radioactivity, with an isotopic breakdown if necessary.

Liquid wastes are processed as required and then released under controlled conditions following isotopic analysis.

The system design and operation are directed toward minimizing releases to unrestricted areas.

Discharge streams are appropriately monitored and safety features are incorporated to preclude releases in excess of the limits of 10CFR20.

11.2.2

System Description

The bulk of the radioactive liquids discharged from the Reactor Coolant System

( RCS) is processed and retained inside the plant by the Chemical and Volume Control System

{CVCS) recycle train.

This minimizes liquid input to the LWS which processes relatively small quanti ties of generally low activity level wastes.

The processed water from waste disposal, from which most of the radioactive material has been removed, is discharged through a monitored line to the service water discharge header and then to the circulating water discharge.

During normal plant operation the LWS processes liquids from the following sources:

1.

Equipment drains and leaks SGS-UFSAR 11.2-2 Revision 23 October 17, 2007 *

2.

Radioactive chemical laboratory drains

3.

Hot shower drains

4.

Decontamination area drains

5.

eves demineralizer regenerant solutions and spent resins

6.

Sampling System In addition, piping has been installed to direct fluid valves in the following systems to the LWS: Residual Heat Removal (RHR),

ection (SIS),

Containment

eves, Sampling This minimizes the event of of accidents.

radioactive the continuous releases from steam generator blowdown from in the to below the of the steam blowdown radiation monitor are to the condenser or to the non-radioactive waste treatment system.

The LWS also collects and transfers liquids from the to the eves, to the waste tanks, or back to the tank (depending on fluid content) for

1.

Reactor coolant

2.

Pressurizer relief tank

3.

Reactor coolant pump seals

4.

Excess letdown (

5.

Accumulators

6.

Valve and reactor vessel leakoffs

7.

canal drains 11.2-3 SGS-UFSAR sources water storage Revision 25 October 26, 2010

These liquids flow to the reactor coolant drain tank and are-discharged either directly to the eves holdup tanks or to the waste holdup tanks by the reactor coolant drain pumps which are operated automatically by a level controller in the tank.

These pumps also return water from the refueling canal and cavity to the refueling water storage tank.

There is one reactor coolant drain tank with two reactor coolant drain tank pumps located inside containment.

Where possible, waste liquids drain to the waste holdup tanks by gravity flow.

Other waste liquids drain to the Auxiliary Building sump tank and are discharged to the waste holdup tanks by pumps operated automatically by a level controller for the Auxiliary Building sump tank.

With the exception of the shared pumps and tanks of the Laundry and Hot Shower Drains, the Chemical Drains, Portable Filter and the Portable Demineralizer, each unit has its own Liquid Waste Disposal System.

The Laundry and Hot Shower Drain Tanks and the Chemical Drain Tank are pumped to one of the Waste Hold-up Tanks or the Waste Monitor Hold-up Tank of either unit.

When a Waste Hold-up or Waste Monitor Hold-up Tank is filled, it is isolated and sampled while another tank is in service.

If analysis confirms that the activity level of the tankrs contents is suitable for discharge, the tank's contents may be pumped through a flow meter and a radiation monitor to the Service Water system.

Tanks requiring processing before release are routed on a batch basis through a portable filter and portable de.mineralizer.

The effluent of the portable system is returned either to the Waste Monitor Hold-up Tanks or the eves Monitor Tanks to be sampled,

analyzed, and either reprocessed or pumped through a flow meter and a radiation monitor to the Service Water System.

Although the Waste Monitor Hold-up or eves Monitor Tank analysis forms the basis for recording activity

releases, the radiation monitor provides surveillance over the operation by closing the discharge valve if the liquid activity exceeds a preset value.

11.2-4 SGS-UFSAR Revision 17 October 16, 1998

The system is capable of processing all liquid wastes generated during continuous operation of the primary system assuming that fission products escape to the reactor coolant by diffusion through defects in the cladding on one percent of the fuel rods.

At least two valves must be manually opened to permit discharge of liquid waste from the LWS.

The control valve will trip closed on a

high effluent radioactivity level signal.

The system is controlled from a central panel in the Auxiliary Building.

Malfunction of the system is alarmed in the Auxiliary Building, and annunciated in the Control Room.

All system equipment is located in the Auxiliary Building, except for the reactor coolant drain tank and drain tank pumps which are located in the reactor containment, and a 2-inch line from the drain tank pumps to the refueling water storage tank (RWST).

The LWS process flow diagram is shown on Plant Drawings 205239 and 205339.

Performance data for the LWS is given in Table 11. 2-1.

liquid discharged to the LWS is given in Table 11.2-2.

11.2.3

System Design

The LWS code requirements are given in Table 11. 2-3.

The estimated annual A summary of component system data is given in Table 11.2-4.

Note that Table 11.2-3 also contains code data for the Gaseous and Solid Radwaste Systems.

11.2-5 SGS-UFSAR Revision 27 November 25, 2013

Hot Shower Tanks Two stainless steel tanks collect liquid wastes originating from the hot shower and local sinks.

These tanks and their associated pumps are common to both Units' LWS.

The intention is that one tank will be available for filling, while the contents of the other tank are being pumped to a waste holdup tank to await processing.

A basket type strainer is provided downstream of this pump to prevent discharge of lint to other tanks..

The pump is started and stopped manually from a local control panel.

Chemical Drain Tank The chemical drain tank is stainless steel and collects drainage from the chemistry laboratory.

This potentially high activity waste is normally transferred to the waste holdup tanks to await processing. This tank and associated pump are common to both Units 1 and 2. The pump is started and stopped from a local control panel.

The suction lines from the chemical drain tank and laundry tanks are interconnected to allow the laundry pump and chemical drain pump to substitute when necessary.

Reactor Coolant Drain Tank The reactor coolant drain tank is a right circular cylinder with spherically dished heads.

The tank is constructed of stainless steel with welded seams.

The reactor coolant drain tank receives recyclable waste from the following sources:

1.

Reactor coolant pump seal and head tank leakoffs 11.2-6 SGS-UFSAR Revision 14 December 29, 1995

2.

Drains from each of the four primary coolant loops

3.

Reactor vessel flange leakage

4.

Accumulator drains

5.

Excess letdown

6.

Refueling canal drains During normal operation a nitrogen blanket 1 maintained in the reactor coolant drain tank.

at a pressure of 0. 5 psig, is

The tank is normally vented to the Gaseous Waste Disposal System vent header so that changes in liquid level will cause the tank to breathe to and from this header.

This eliminates the of hydrogen and radioactive gases to the containment. The contents of the reactor coolant drain tank can be transferred to one of the following

1.

eves holdup tanks

2.

Emergency Core System (ECCS) RWST

3.

LWS holdup tanks Normally all waste collected in the reactor coolant drain tank is transferred to the CVCS holdup tanks by the Nos. 11 1 21 and 12, 22 reactor coolant drain pumps.

Operation of these pumps is automatically controlled by tank level instrumentation.

Valves WL-12 and WL-13 are maintained in the in the reactor coolant drain tank pump discharge line shut following containment isolation until manually reset by the operator.

11.2-7 SGS-UFSAR Revision 26 May 21, 2012

Two waste holdup tanks are provided to accept liquid wastes from the eves, sump tank, chemical drain tank, reactor coolant drain tank, Steam Generator Blowdown System, floor and hot shower tanks.

The tanks are of welded stainless steel construction.

Individual air-operated valves in the common inlet manifold to these tanks are used to divert waste flow from one tank to the other.

Two tanks are with the intention that one tank will be available to accept waste, while the contents of the other tank are being held to await processing.

This an additional over and above that available from to allow shorter-lived radionuclides to An additional 25, waste monitor holdup tank is available to waste surges in the event of an emergency.

The containment sump accumulates all floor drains, washdowns from refueling decontamination operations, drains and condensate from the fan coil units and miscellaneous equipment drains of a potentially radioactive but non-reactor coolant nature.

The contents of this sump are pumped directly to the waste holdup tanks by two sump pumps that operate from sump level control instrumentation.

Both pumps can also be started manually.

All wetted parts of the pump are stainless steel.

The tank is all welded stainless steel.

The Unit 1 waste evaporator is abandoned in place.

No. 2 waste evaporator has been cancelled.

11.2-8 SGS-UFSAR Installation of the Unit Revision 26 May 21, 2012

Piping and valves which are in, contact with liquid wastes are constructed of stainless steel.

Piping con11ections are welded except where flanged connections are necessary to facilitate equipment maintenance.

Isolation valves are provided to isolate each piece of equipment for maintenance, to direct the flow of waste through the system, and to isolate storage tanks for radioactive Relief valves are for tanks containing radioactive wastes if the tanks might be over component malfunction.

The spent resin storage tank retains the zed by operation or resin discharged from the mixed bed, evaporator feed ion exchange, spent fuel

, deborating vessel and cation demineralizers.

A layer of water is maintained over the resin storage to prevent resin degradation due to heat from fission products.

The contents can be removed any time by flushing with nitrogen. Resin sample connections are supplied and downstream of the spent resin storage tank isolation valve.

The tank is all welded austenitic stainless steel.

11.2-9 SGS-UFSAR Revision 26 May 21, 2012

The Waste Monitor Tanks have been disabled and abandoned in place (West end of 84' elevation).

These tanks are of welded stainless steel construction.

The Waste Monitor Hold-up Tank, used as a steel tank and it serves a dual purpose.

WHUT, is also a welded stainless Its normal function is as a third waste holdup tank to receive abnormally large quantities of waste discharged to the system, but it can also serve as a waste monitor tank.

Portable Processing System Permanent provisions have been made to the 460 VAe supply, waste liquid piping, compressed air piping, and the demineralized water-restricted area to allow for the.installation and operation of a portable system to process liquid radwaste from either unit.

The Unit 2 liquid waste may be processed at a higher rate due to a difference in piping configuration.

The system is installed and operated in the 100' elevation of the truck bay of the The effluent of the is returned to either the Waste Monitor Holdup Tanks or the eves Monitor Tanks to be sampled, analyzed, and either reprocessed or disposed of.

Exhausted ion

media, or or both, is transferred to a burial site approved container after which it can be processed, classified, and shipped for The steam generator blowdown is described in section 10.4.8.

11.2.4 Procedures Verification is made to ensure that dilution flow sufficient to meet the of 10CFR20 is available whenever radioactive liquid wastes are released to the Plant Discharge System.

11.2-10 SGS-UFSAR Revision 26 May 21, 2012

Liquid waste releases are continuously monitored for gros*s activity during discharges to ensure that the activity limits specified in 10CFR20 for unrestricted areas are not exceeded.

The maximum allowable release at the plant is specified in the Technical Specifications.

Radioactive liquid batch wastes are sampled prior to releases to the Plant Discharge System and records of all releases are kept.

Continuous releases from the steam generator blowdown system are monitored and controlled in accordance with the Salem Offsite Dose Calculation Manual.

11.2.5 Performance Tests Samples are taken on each batch of liquid waste released.

Station records contain the quantity and concentration of radioactive isotopes, the volume of each batch and estimates of the water flow for dilution.

Each sample is analyzed for principal gamma emitters.

Composites are prepared from each batch released during a month and analyzed for the principal gamma emitting

nuclides, fission and activation products, gross alpha, and tritium.

A quarterly composite analysis is also performed for Sr-89, Sr-90, and Fe-55.

The sensitivities and frequencies of analyses comply with the requirements of Salem Technical Specifications.

Continuous releases from the steam generator blowdown system are sampled and analyzed to determine the quantity and concentration of isotopes present in the blowdown stream.

Composites are prepared from the samples.

The sensitivities and frequencies of analysis comply with the requirements of the Salem Technical Specifications.

11.2-lOa SGS-UFSAR Revision 18 April 26, 2000 I

THIS PAGE INTENTIONALLY LEFT BLANK 11.2-lOb SGS-UFSAR Revision 13 June 12, 1994

11.2.6 Estimated Releases Liquid wastes are primarily by plant maintenance and service operations, and, consequently, the quantities and activity concentrations of influents to the system.

Tables 11.2-5 and 11.2-6 are estimated values.

Therefore, considerable operational margin has been assigned between the design capability and the estimated system load as indicated in Table 11.2-5.

A conservative estimate of system ability activity released in the iiquid phase is to *limit dissolved and suspended summarized in Table 11.2-6:

This tabulation is based on the following assumptions.

1.

Fission product concentration in the reactor coolant is based on 1

defective fuel.

2.

of reactor grade coolant which is and discharged from the is assumed to be a

total of 2,102 gallons per day.

3.

Non-reactor non-radioactive which is and discharged from the is assumed to be a

total of 4,308 gallons per day.

4.

It is assumed that the liquid wastes are accumulated in the waste SGS-UFSAR holdup tank and then cooling water.

for 11.2-11 overboard with the raw Revision 26 May 21, 2012

5.

A dilution factor is used in determining the discharge concentrations of liquids released from the waste rnoni tor holdup and waste holdup tanks.

This is discussed in Section 11.2.8.

6.

The tri tiurn that is formed in the fuel (the predominant source) diffuses through the zircaloy clad and eventually becomes available for dispersal to the environment.

The expected release of ternary produced tritium is about 1 percent; the total annual tritium release expected is indicated in Table 11.2-6.

All of the sources of tritium accumulating in the reactor coolant, shown in Section 11.1, are included in the annual release.

The release estimates given in Table 11.2-6 are based on continuous operation with 1 percent fuel defects in both units.

Based on experience with operating pressurized water reactors to date, 0.2 percent is a more realistic estimate of fuel defects averaged over a year of operation. Hence, in order to evaluate expected releases, all release values, except for H-3 and corrosion activation products given in Table 11.3-6, could be reduced by a factor of 5.

The release estimates for continuous releases from the steam generator blowdown system are based on assumptions consistent with NUREG-001 7.

Releases are monitored by the steam generator radiation monitors which provide a signal to close the isolation valves to maintain releases within the requirements of the ODCM.

At higher primary system activity levels, steam generator blowdown would be isolated at lower primary to secondary leakage rates.

11.2.7 Release Points Release points are shown on the system flow diagram, Plant Drawings 205239 and 205339.

11.2.8 Dilution Factors Monitored waste released from the Liquid Waste Disposal System is normally pumped into the Circulating Water System discharge lines via the Service Water System.

The maximum release rates are calculated based on how many circulators are available in the release path.

The release rate is controlled by throttling the discharge valve to maintain the maximum release rate or less.

Stearn generator blowdown is directed to the non radioactive waste basin either from the blowdown flashtank or from the condensate polishing system and then to the circulating water system.

11.2-12 SGS-UFSAR Revision 27 November 25, 2013

11.2.9 Estimated Doses (Potential)

There are generally two main pathways by which the general population could receive radiation exposure from liquid releases.

One would be by drinking the water from the river in the vicinity of the station, the other would be by the consumption of marine life (such as fish and shellfish) that inhabit the general river and bay area (marine life has a tendency to reconcentrate certain radioactive isotopes above background water levels).

Less important pathways would be from swimming in the river, boating or fishing, standing on the shoreline, etc.

Section 2 contains detailed information on public water supplies and known water wells in the Salem site vicinity.

Because of the brackish condition of the river, no potable water supplies are drawn from it in the area.

Ground water is the primary source of water supply in the vicinity with the exception of the city of Salem, New Jersey, which obtains about two-thirds of its water supply from Quinton, on Alloways Creek. This water supply is a dammed fresh water stream some 9 miles upstream from the Delaware River - Alloways Creek confluence.

Hence, no radioactive releases would reach this public water supply from the Salem station as the flow is from Alloways Creek into the Delaware River.

Consideration was given to the possibility of radioactivity reaching public and private water supplies that are present in the Salem area. The Delaware River is not recharging aquifers that are in use. The upper sand layers in the region are saline aquifers.

The artesian aquifers, located at much greater depths than the saline aquifers, are separated from the upper sand layers by an impervious clay strata. Hence, no Hydraulic communication exists. It is concluded that no liquid radioactive releases could reach ground water drinking supplies in the Salem area (additional information on ground water hydrology is given in Section 2).

11.2-13 SGS-UFSAR Revision 14 December 29, 1995

Consideration was given to the radiation exposure that might be received through the fish food chain.

The fundamental approach in evaluating this pathway is dependent on the radionuclide concentration provided by the fish in question.

The concentration of the stable element {and the radionuclide) by the organism is related to the natural biological demand which the organism has for the element in question and the ratio of the concentration of the element to the elemental concentration in its water environment.

Tables 11.2-7 and 11.2-8 relate the concentration factor (by the marine life), the radioactivity in the marine life and the resulting concentration to which the individual would be exposed upon consumption of fish and blue crabs, respectively.

The following assumptions and data were used to develop the concentration factors, ingestion factors and fraction of MPC in Tables 11.2-7 and 11.2-8:

1.

Concentration factors are based on a literature review of the stable element chemistry and the radionuclide concentration contained in "Concentration Factors of Chemical Elements in Edible Aquatic organisms,~ w. H. Chapman et al, UCRL-50564, December 30, 1968.

2.

100 percent of the effluent radionuclides are assumed to be ingested or absorbed by the marine life in question (the organisms are considered to remain in the vicinity of the discharge pipe 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day, 52 weeks per year).

3.

Individual consumption will include approximately 4-5 meals per week (100 grams, or 140 grams per day, 52 weeks per year).

The marine life in question are comprised of bullhead catfish, carp, weakfish, striped bass, American eel and blue crab. These organisms represent the major cross section of edible marine life caught in the area.

11.2-14 SGS-UFSAR Revision 6 February 15, 1987

4.

The ingestion factor is calculated based on the following assumptions:

a.

Equivalent density for "fish flesh" and water

b.

Individual daily consumption of "fish flesh" is approximately 140 grams

c.

MPC values given in 10CFR20 are based on a total daily intake of 2.2 liters of water I

t.

F t

140 -- 6.36 X 10-2 nges 10n ac or =

2200 This approach to an analysis of radiation exposure through the food chain is extremely conservative.

The assumption that the marine life in question will remain adjacent to the discharge pipe 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day, 52 weeks per year, is highly conservative.

The reduction in radioactivity due to radioactive decay and biological turnover was not assumed.

Based on Table 11.2-7 and 1 percent failed fuel, the potential exposure from eating 50,000 grams (110 lbs) per year of fish can

-5 be calculated as 6.1 x 10 x 500 mrem = 0. 030 mrem per year.

Similarly, from Table 11.2-8, the potential exposure from eating 50,000 grams per year of crabmeat (net weight) can be calculated

-4 as 2.4 x 10 x 500 mrem = 0.12 mrem per year.

Based on 0.2 percent fuel defects, the exposures from ingestion of fish and crabmeat would be 0.01 mrem/yr and 0.03 mrem/yr, respectively.

Consideration was also given to individuals swimming, fishing, and boating on the river.

Calculation of these doses is given below.

11.2-15 SGS-UFSAR Revision 6 February 15, 1987

The dose to an individual swimming near the discharge canal was calculated using the following basic equation for the dose in an "infinite" homogeneous source (1):

R = 51 CE where!

R = dose, rads/day C = concentration of radioactive material, uCi gm E = decay energy, Mev/dis.

The release concentrations given in Table 11.2-6 (based on 1 percent fuel defects) would be adjusted to 2.0 x 10-ll uCi/cc, excluding H-3, and 4.5 x 10-7 uCi/cc for H-3 (based on 2 percent fuel defects).

It is assumed that an individual swims near the discharge canal for 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months per year.

It is also assumed that the average decay energy is I Mev/dis for all isotopes except H-3.

An average decay energy of 0.0057 Mev/dis is used for H-3.

The calculations to a maximum individual follow:

Dose due to all isotopes except H-3 D = 51(2.0xl0-ll !£)

cc D

-9 ~

= 8.3xl0 yr 11.2-16 SGS-UFSAR (l Mev) dis (200 hrs) yr Revision 6 February 15, 1987

Dose due to H-3 D

uCi cc Mev hrs 1 day s1 (4.5x 10-' --)--ri -> (o.oos7 --> {2oo --) c

)

cc gm dis yr 24 hrs rads D

1. 1x 1 o-6 --

yr The total dose is thus 1.1x10-6 rads per year.

The dose to an individual fishing along the shore or on a boat on the river can be estimated in a manner similar to the one above.

In this case, it is more appropriate to think in terms of a

"semi-infinite" medium, since essentially no radioactivity from releases to the river is present in the air above the river (in the swimmdng case, it was assumed that the individual was "submerged" in the water; hence, the individual would be exposed to radiation from all directions}.

The doses to an individual fishing 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months per year on shore or on a boat would thus be one-half of the doses previously calculated, or 4. 3x10-9 rads per year due to all isotopes except H-3 and 5.5xl0-1 rads per year due to H-3, for a total dose of S.Sxl0- 7 rads per year.

The dose pathways considered, and the resulting doses to the maximally exposed individual based on realistic liquid releases at 0.2 percent failed fuel defects, are summarized in Table 11.2-9.

Doses and releases, including steam generator blowdown continuous releases are controlled in accordance with the Salem Offsite Dose Calculation Manual.

11.2.10

1.

SGS-UFSAR Reference for Section 11.2 Evans, R. D. "The Atomic Nucleus," McGraw-Hill, 1955, p. 742.

11.2-17 Revision 18 April 26, 2000

TABLE 11.2-1 LIQUID WASTE SYSTEM PERFORMANCE DATA Evaporator Design Life Normal process capacity, liquids Evaporator load factor Average During peak week Annual liquid discharge(!)

Reactor grade water Nonreactor grade water Total Activity other than tritium Tritium Portable Demineralizer normal Process capacity Portable system normal process capacity Portable system load factor Average During Peak Week NOTES:

{1)

Estimate based on Table 11.2-2.

1 of 1 SGS-UFSAR 39 years 16.5 gpm (Unit 1 only)

Inactive Inactive 766,950 gal 1,572,500 gal 2,339,450 gal 0.145 curies 690 curies 28 gpm (Unit 1}

38 gpm Unit 2) 25-40 gpm (at 28 gpm) 6 percent 42 percent Revision 19 November 19, 2001

TABLE 11.2-2 ESTIMATED ANNUAL LIQUID DISCHARGE TO WASTE DISPOSAL (per unit}

SGS-UFSAR source BAE distallate Hot showers Laboratory Equipment drains, leaks Decontamination Total Annual Discharge. gal 1,411,000 119,800 91,250 702,400 15.000 Total waste disposal system 2,339,450 1 of 1 t

Revision 14 December 29, 1995

TABLE 11.2-3 WASTE DISPOSAL COMPONENTS CODE REQUIREMENTS Component Chemical Drain Tank Reactor Coolant Drain Tank Sump Tank Waste Holdup Tank Waste Monitor - Holdup Tank Waste Monitor Tank Laundry and Hot Shower Tank Waste Evaporator Forced Circulation Concentrator Waste Filter Piping and Valves (11}

Spent Resin Storage Tank Pumps and Compressors (7) (11)

Evaporator Bottoms Holdup Tank Reactor Coolant Drain Tank Pumps Portable Liquid Radwaste Processing System SGS-UFSAR Code ASME VIII (4} (not code stamped)

ASME III, ( 1) Class c ASME III, {1) Class c ASME III,

( 1) Class c ASME III,

( 1) Class c { 9)

ASME III,

( 1) Class c (10)

ASME VIII ( 4}

(not code stamped)

ASME VIII (5)

ASME III, (1) Class C ANSI B31.7

{ 2) Section 1 ANSI B31.1 ( 3)

ASME III,

( 1) Class c ASME Draft Code for Pumps and Valves for Nuclear November, 1968 ASME VIII {5}

NNS, Class D+ (6)

(8) 1 of 2

Power, Revision 19 November 19, 2001 I

TABLE 11.2-3 (Cont.)

WASTE DISPOSAL COMPONENTS CODE REQUIREMENTS Component Code NOTES:

(1)

ASME III -

American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Nuclear Vessels, 1968 Edition.

(2)

For piping not supplied by the NSSS

supplier, material inspection, fabrication, and quality control conform to ANSI 831.7.

Where not possible to comply with ANSI B31. 7, the requirements of ASME III-1971, which incorporated ANSI B31.7, were adhered to.

(3)

ANSI 831.1 -Used for design and material selection.

(4)

ASME VIII - American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section VIII, Pressure Vessels, 1968 Edition.

(5)

ASME VIII -

American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section VIII, Pressure Vessels, 1971 Edition.

(6)

Quality Group D is augmented by the addition of Quality Assurance Requirements necessary to ensure an acceptable level of confidence that the pumps perform their intended functions.

(7)

Does not include the Reactor Coolant Drain Tank Pumps (listed separately).

( 8)

Design criteria for the portable liquid radwaste processing system meets the intent of section 3 of the OFSAR and was specified to be compatible with Reg Guide 1.143, except for the seismic criteria which are addressed by the basic reactor facility, even though Salem does not commit to the Reg Guide.

(9)

12 Waste Holdup Tank {1WLE11) abandoned in place.

(10)

Waste Monitor Tank (1&2) WLE {3&4) abandoned in place.

(11)

Some Waste Gas (WG) System components can be reclassified to nonsafety-related, non-nuclear, and Seismic Class II in accordance with safety evaluation S-C-NSlO-MSE-0319, "Safety Classification of the Gas Waste Disposal Systernn.

However, all piping and valves providing isolation for the Waste Gas Decay Tanks shall be classified safety-related, Nuclear Class III, and Seismic Class I.

2 of 2 SGS-UFSAR Revision 19 November 19, 2001 I

Tanks Quantity Reactor Coolant Drain 1

(per unit)

Laundry and Hot Shower 2

Chemical Drain 1

Auxiliary Building sump Tank 1

Waste Holdup 2

Waste Monitor Holdup 1

Waste Monitor {4) 2 Spent Resin 1

Evaporator Bottoms 1

Holdup Tank SGS-UFSAR TABLE 11.2-4 COMPONENT DATA

SUMMARY

~

Volume Horiz 565 gal Vert 600 gal Vert 600 gal Horiz 1000 gal Horiz 25,000 gal Horiz 25,000 gal Vert 950 gal Vert 300 ft 3 Vert 1500 gal 1 of 2 Design Pressure 25 psig Atm Atm 15 psig 15 psig 15 psig 15 psig 100 psig 0.5 psig Design Temp.

OF 267 180 180 110 180 180 180 180 250 Material(!)

ss ss ss ss ss ss ss ss Inconel 625 Revision 14 December 29, 1995

Tanks Quantity Reactor Coolant 2

Drain Chemical Drain 1{3)

Laundry 1{3)

Auxiliary Building Sump Tank 2

Waste Evaporator 1

Feed Waste Monitor ( 4) 2 Waste Monitor 1

Holdup NOTES:

( 1) Material contacting fluid.

( 2) Mechanical seal provided.

( 3) Shared by Units 1 and 2.

( 4) Waste Monitor Tanks and Pumps SGS-UFSAR TABLE 11.2-4 Flow Head

~

9..2ffi

!.L_

Horiz 50 175 cent canned Horiz 20 100 cent(2)

Horiz 20 100 cent(2)

Horiz 40 60 cent Horiz 16.5 150 cent Horiz 40 100 cent Horiz 60 110 cent abandoned in place.

2 of 2 Design Design Pressure Temp QSig OF 150 200 150 200 150 200 150 180 150 200 150 180 150 180 Material(!)

ss ss ss ss ss ss ss Revision 14 December 29, 1995

TABLE 11.2-5 ESTIMATED ANNUAL LIQUID DISCHARGE TO LIQUID WASTE SYSTEM Source of Liquid Waste Tank Drains Filter Strainer Drains Heat Exchanger Drains Demineralizers Valve Leakoffs Pump Leakages Sample Sink Drains Total Reactor Grade Component Cooling Heat Exchanger Drains Component Cooling Valve Leakoffs Hot Showers Decontamination and Floor Scrubbing BAE Distillate Total Nonreactor Grade Grand Total SGS-UFSAR Type Reactor Grade Reactor Grade Reactor Grade Reactor Grade Reactor Grade Reactor Grade Reactor Grade Nonreactor Grade Nonreactor Grade Nonreactor Grade Nonreactor Grade Nonreactor Grade 1 of 1 Amount <gal/yr) 239,000 51,000 63,700 49,000 78,000 195,000 91,250 766,950 20,500 6,200 119,800 15,000 1.411.000 1,572,500 2,339,450 Revision 14 December 29, 1995

Isoto~e H-3 Cr-51 Mn-54 Mn-56 Co-58 Co-60 Fe-59 Sr-89 Sr-90 Sr-91 Sr-92 Y-90 Y-91 Y-92 Zr-95 Mo-99 Nb-95 1-131 1-132 1-133 1-134 I-135 Te-132 Te-134 SGS-UFSAR TABLE 11. 2-6 ESTIMATED ANNUAL LIQUID RELEASE BY ISOTOPE (TWO UNIT BASIS)

Discharge Annual Canal Release Concentration (IJCi)

(IJCi/cc) 1.37 X 109 4.54 X 10 -7 5.56 X 10 1 1.88 X 10-14 7.24 X 10 1 2.45. X 10 -14 9.84 X 10° 3.32 X 10-15 1.98 X 103 6.69 X 10-13 7.28 X 101 2.46 X 10-14 7.34 X 101 2.48 X 10-14 2.74 X 102 9.25 X 10-14 1.34 X 10 1 4.53 X 10-15 2.64 X 10 0 8.92 X 10-!6 2.84 X 10 -1 9.60 X 10-17 1.39 X 10 0 4.70 X 10-16 4.64 X 10 2 1.57 X 10-13 3.54 X 10 -1 1.20 X 10-16 5.40 X 10 -1 1.82 X 10-14

4. 78 X 10 4 1.61 X 10-11 4.40 X 10 1 1.49 X 10-14 6.18 X 104 2.09 X 10-11 2.86 X 102 9.66 X 10-14 1.13 X 104 3.82 X 10-12 7.12 X 10 1 2.41 X 10-14

1. 94 X.103 6.55 X 10-13 2.68 X 10 3 9.05 X 10-13 3.00 X 10° 1.01 X 10-15 1 of 2 Fraction of (MPC)w (MPC)w (JJCi/cc) 3 X 10-3 1.51 X 10 -4 2 X 10-3 9.40 X 10-12 1 X 10-4 2.45 X 10-10 1 X 10-4 3.32 X 10-11 1 X 10-4 6.69 X 10-9 5 X 10-5 4.92 X 10-10 6 X 10-5 4.13 X 10-10 3 X 10-6 3.08 X 10-8 3 X 10-7 1.51 X 10-8 7 X 10-5 1.27 X 10-11 7 X 10-5 1.37 X 10-12 2 X 10-5 2.35 X 10-11 3 X 10-5 5.23 X 10 -9 6 X 10-5 2.00 X 10-12 6 X 10-S 3.03 X 10-10 2 X 10-4 8.05 X 10 -8 1 X 10-4 1.49 X 10-10 3 X 10-7 6.97 X 10 -5 8 X 10-6 1.21 X 10 -8 1 X 10-6 3.82 X 10 -6 2 X 10-5 1.20 X 10 -9 4 X 10-6 1.64 X 10 -7 3 X 10-5 3.02 X 10 -8

-~~~~~~~~---

~._.,..., ___

Revision 6 February 15, 1987

Annual Isoto2e Release (IJCi)

Cs-134 2.52 X 104 Cs-136 6.18 X 103 Cs-137 1.29 X 105 Cs-138 7.30 X 10 1 Ba-140 1.58 X 102 La-140 7.30 X 10° Ce-144 2.96 X 101 Pr-144 1.20 X 10-2 Totals 2.90 X 105 (1)

NOTE:

(1) Excluding H-3 SGS-UFSAR TABLE 11.2-6 (Cont)

Discharge Canal Concentration (1JCi/cc) 8.51 X 10-12 2.09 X 10-}2 4.36 X 10-11 2.46 X 10-14 5.34 X 10-14 2.46 X 10-15 1.00 X 10-14 4.05 X 10-18 9.80 X 10-11 (1) 2 of 2 Fraction of (MPC)w (MPC)w (IJCi/cc) 9 X 10-6 9.46 X 10 -7 9 X 10-5 2.32 X 10 -8 2 X 10-5 2.18 X 10 -6

--~-... ~

8 X 10-5 1.78 X 10-9

2 X 10-5 1.23 X 10-10 1 X 10-5 1.00 X 10-9

~

-4 2.27 X 10 Revision 6 February 15, 1987

TABLE 11.2-7 CONCENTRATION OF RADIONUCLIDES IN FISH DUE TO PLANT OPERATION Discharge Activity Activity in Fraction Canal Concentration in Fish x of Isoto:12e Concentration Factor Fish (MPC)w Ingestion Factor (MPC)w (JJCi/cc)

(J.1Ci/cc)

(iJCi/cc)

(JJCi/cc)

H-3 4.54 X 10-7 9.3 X 10 -1 4.2 X 10-7 3 X 10-3

2. 7 X 10-s 9.0 X 10-6 Cr-51 1.88 X 10-14 4 X 102 7.5 X 10-12 2 X 10-3 4.8 X 10-13 2.4 X 10-10 Mn-54 2.45 X 10-14 3 X 102 7.4 X 10-12 1 X 10-4 4.7 X 10-13

4. 7 X 10-9 Mn-56 3.32 X 10-15 3 X 102 1.0 X 10-12 1 X 10-4.

6.4 X 10-14 6.4 X 10-10 Co-58 6.69 X 10-13 5 X 102 3.3 X 10-10 1 X 10-4 2.1 X 10-11 2.1 X 10-7 Co-60 2.46 X 10-14 5 X 102 1.23 X 10-}1 5 X 10-5 7.9 X 10-13 1.6 X 10-s Fe-59 2.48 X 10-14 3 X 103 7.4 X 10-11 6 X 10-5 4.7 X 10-12 7.8 X 10-8 Sr-89 9.25 X 10-14 5 X 10-1 4.6 X 10-14 3 X 10-6 2.9 X 10-15 9.7 X 10-10 Sr-90 4.53 X 10-15 5 X 10-l 2.3 X 10-15 3 X 10-7 1.5 X 10-16 5.0 X 10-10 Sr-91 8.92 X 10-16 5 X 10-1 4.5 X 10-16 7 X 10-5 2.9 X 10-17 4.1 X 10-13 Sr-92 9.60 X 10-17 5 X 10-1 4.8 X 10-17 7 X 10-5 3.1 X 10-18 4.4 X 10-14 Y-90 4.70 X 10-16 1 X 102

4. 7 X 10-14 2 X 10-5 3.0 X 10-15 1.5 X 10-10 Y-91 1.57 X 10-13 1 X 102 1.6 X 10-11 3 X 10-5 1.0 X 10-12 3.3 X 10-s Y-92 1.20 X 10-16 1 X 102 1.2 X 10-14 6 X 10-s 7.7 X 10-16 1.3 X 10-11 Zr-95 1.82 X 10-14 1 X 102 1.8 X 10-12 6 X 105 1.2 X 10-13 2.0 X 10-9 Mo-99 1.61 X 10-11 1 X 101 1.6 X 10-10 2 X 10-4 1.0 X 10-11 5.0 X 10-8 Nb-95 1.49 X 10-14 3 X 104 4.5 X 10-10 1 X 10-4 2.9 X 10-11 2.9 X 10-7 I-131 2.09 X 10-11 1 X 101 2.1 X 10-10 3 X 10-7 1.3 X 10-11 4.3 X 10-5 1 of 2 SGS-UFSAR Revision 6 February 15, 1987

e TABLE 11.2-7 (Cont)

Discharge Activity Activity in Fraction Canal Concentration in Fish x of Isoto~e Concentration Factor Fish (MPC)w Ingestion Factor (MPC)w (JJCi/cc)

(JJCi/cc)

(JJCi/cc) 1-132 9.66 X 10-14 1 X 101 9.7 X 10-6 8 X 10-6 6.2 X 10-14 7.8 X 10-9 1-133 3.82 X 10-12 1 X 101 3.8 X 10-11 1 X 10-6 2.4 X 10-12 2.4 X 10-6 1-134 2.41 X 10-14 1 X 101 2.4 X 10-13 2 X 10-5 1.5 X 10-14 7.5 X 10-10 1-135 6.55 X 10-13 1 X 101 6.6 X 10-12 4 X 10-6 4.2 X 10-13 1.0 X 10-7 Te-132 9.05 X 10-13 1 X 101 (1) 9.0 X 10-12 3 X 10-5 5.8 X 10-13 1.9 X 10-8 Te-134 1.01 X 10-15 1 X 101 (1) 1.0 X 10-14 6.4 X 10-16 Cs-134 8.51 X 10-12 3 X 101 2.6 X 10-10 9 X 10-.6 1.7 X 10-11 1.9 X 10-6 Cs-136 2.09 X 10-12 3 X 101 6.3 X 10-11 9 X 10-5 4.0 X 10-12 4.4 X 10-8 Cs-137 4.36 X 10-11 3 X 101 1.3 X 10-9 2 X 10-5 8.3 X 10-11 4.2 X 10-6 Cs-138 2.46 X 10-14 3 X 101 7.4 X 10-13

4. 7 X 10-14 Ba-140 5.34 X 10-14 1 X 101 5.3 X 10-13 3 X 10-5 3.4 X 10-14 1.1 X 10-9 La-140 2.46 X 10-15 1 X 102 2.5 X 10-13 2 X 10-5 1.6 X 10-14 8.. 0 X 10-10 Ce-144 1.00 X 10-14 1 X 102 1.0 X 10-13 1 X 10-5 6.4 X 10-14 6.4 X 10 -9 Pr-144 4.05 X 10-18 1 X 102 4.0 X 10-16 2.6 X 10-17 TOTAL 9.80 X 10-11 (2)

-9 2.6 X 10 (2)

6. 1 X 10 -5 NOTES:

(1) Abstracted from reference "A Model for the Approximate Calculation of Safe Rates of Discharge of Radioactive Waste into Marine Environs," A. M. Freke, Health Physics, Vol.. 13, pp. 743-758, 167.

(2) Excluding H-3.

SGS-UFSAR 2 of 2 Revision 6 February 15, 1987

Discharge Canal Isoto:2e Concentration

(~Ci/cc)

H -3 4.54 X 10-7 Cr - 51 1.88 X 10-14 Mn - 54 2.45 X 10-14 Mn-56 3.32 X 10-15 Co - 58 6.69 X 10-13 Co - 60 2.46 X 10-14 Fe - 59 2.48 X 10-14 Sr - 89 9.25 X 10-14 Sr - 90 4.53 X 10-15 Sr - 91 8.92 X 10-16 Sr - 92 9.60 X 10-17 y -

90 4.70 X 10-16 y -

91 1.57 X 10-13 y -

92 1.20 X 10-16 Zr - 95 1.82 X 10-14 Mo - 99 1.61 X 10-11 SGS-UFSAR TABLE 11.2-8 CONCENTRATION OF RADIONUCLIDES IN BLUE CRABS DUE TO PLANT OPERATION Activity Activity in Concentration in Fish x Factor Crab (MPC)w Insestion Factor

(~Ci/cc)

(JJCi/cc)

(tJCi/cc) 9.3 X 10 1 4.2 X 10-7 3 X 10-3 2.7 X 10-s 2 X 103 3.8 X 10-11 2 X 10-3 2.4 X 10-12 5 X 103 1.2 X 10-10 1 X 10-4 7.7 X 10-12 5 X 103 1.7 X 10-11 1 X 10-4 1.1 X 10-12 1 X 103

6. 7 X 10-10 1 X 10-4 4.3 X 10-11 1 X 103 2.5 X 10-11 5 X 10-s 1.6 X 10-12 2 X 104 5.0 X 10-10 6 X 10-s 3.2 X 10-11 6.3 X 10° 5.8 X 10-13 3 X 10-6 3.7 X 10-14 6.3 X 10° 2.9 X 10-14 3 X 10-7 1.9 X 10-15 6.3 X 10° 5.6 X 10-15 7 X 10-5 3.6 X 10-16 6.3 X 10 0 6.0 X 10-16 7 X 10-5 3.8 X 10-17 1 X 103

4. 7 X 10-13 2 X 10-s 3.0 X 10-14 1 X 103 1.6 X 10-10 3 X 10-5 1.0 X 10-11 1 X 103 1.2 X 10-13 6 X 10-5 7.7 X 10-15 1 X 103 1.8 X 12-11 6 X 10-5 1.2 X 10-12 1 X 101 1.6 X 10-10 2 X 10-4 1.0 X 10-11 1 of 3 Fraction of (MPC)w 9.0 X 10-6 1.2 X 10-9 7.7 X 10-8 1.1 X 10-s 4.3 X 10-7 3.2 X 10-s 5.3 X 10-7 1.2 X 10-8 6.3 X 10-9 5.1 X 10-12 5.4 X 10-13 1.5 X 10-9 3.3 X 10-7 1.3 X 10-10 2.0 X 10-8 5.0 X 10-s Revision 6 February 15, 1987

Discharge Canal Concentration Isoto;ee Concentration Factor

(}JCi/cc)

Nb - 95 1.49 X 10-14 1 X 102 I -

131 2.09 X 10-11 5 X 101 I -

132 9.66 X 10-14 5 X 101 I -

133 3.82 X 10-12 5 X 101 I -

134 2.41 X 10-14 5 X 101 I -

135 6.55 X 10-13 5 X 101 Te - 132 9.. 05 X 10-13 1 X 101 (1)

Te - 134 1.01 X 10-15 1 X 101 (1)

Cs - 134 8.51 X 10-12 2 X 101 Cs - 136 2.09 X 10-12 2 X 101 Cs - 137 4.36 X 10-11 2 X 101 Cs - 138

. 2.46 X 10-14 2 X 101 Ba - 140 5.34 X 10-14 2 X 102 SGS-UFSAR TABLE 11.2-8 (Cont)

Activity in Crab (MPC)w (JJCi/cc)

(JJCi/cc) 1.5 X 10-12 1 X 10-4 1.0 X 10-9 3 X 10-7 4.8 X 10-12 8 X 10-6 1.. 9 X 10-10 1 X 10-6 1.2 X 10-12 2 X 10-s 3.3 X 10-11 4 X 10-6 9.0 X 10-12 3 X 10-5 1.0 X 10-14 1.7 X 10-10 9 X 10-6 4.2 X 10-11 9 X 10-5 8.7 X 10-10 2 X 10-5 4.9 X 10-13 1.1 X 10-11 3 X 10-5 2 of 3 Activity in Fish x Ingestion Factor

(}JCi/cc) 9.. 6 X 10-14 6.. 4 X 10-11 3.1 X 10-13 1.2 X 10-11 7.7 X 10-14 2.1 X 10-12 5.8 X 10-13 6.4 X 10-16 1.1 X 10-11 2.7 X 10-12 5.6 X 10-11 3.1 X 10-14 7.0 X 10-13 Fraction of (MPC)w 9.. 6 X 10-10 2.. 1 X 10-4 3.. 9 X 10 -8 1.2 X 10-5 3.8 X 10-9 5.2 X 10-7 1.9 X 10-8 1.2 X 10-6 3.0 X 10 -8 2.8 X 10-6 2.3 X 10-8 Revision 6 February 15, 1987

Isotope La - 140 Ce - 144 Pr - 144 TOTALS NOTES:

D.ischarge Canal Concentration (J,JCi/cc) 2.46 X 10-15 1.00 X 10-14 4.05 X 10-18 9.80 X 10-11 (2)

Concentration Factor 1 X 103 1 X 103 1 X 103 TABLE 11.2-8 (Cont)

Activity in Crab (f.JCi/cc) 2.5 X 10-12 1.0 X 10-11 4.0 X 10-15 4.0 X 10-9 (2)

(MPC)w (JJCi/cc) 2 X 10-5 1 X 10-5 Activity in Fish x Ingestion Factor (JJCi/cc) 1.6 X 10-13 6.4 X 10-13 2 6 10-16 X

Fraction of (MPC)w 8.0 X 10-9 6 4 10-8 X

-4 2.4 X 10 (1) Abstracted from reference "A Model for the Approximate Calculation of Safe Rates of Discharge of Radioactive Waste into Marine Environs," A.M. Freke, Health Physics, Vol. 13, pp. 743-758, 167.

(2) Excluding H-3.

SGS-UFSAR 3 of 3 Revision 6 Feb~uary 15, 1987

TABLE 11.2-9 POTENTIAL RADIATION EXPOSURE PATHWAYS TO MAN (LIQUID)

(Based on 0.2 percent failed fuel defects)

Pathway Drinking River Water Radioactivity in Ground Water Ingestion of Fish Ingestion of Crabmeat Swimming in River Boating on River or Fishing on Shoreline SGS-UFSAR Individual Assumed Exposed None-water not potable None - See Section 11.1.3.1 of the FSAR Individual eats 50,000 gm/yr of fish Individual eats 50,000 gm/yr of crabmeat Individual swims 200 hr/yr in discharge canal Individual boats or fishes 200 hr/yr 1 of 1 Calculated Exposure 0

0 0.01 mrem/yr 0.03 mrem/yr 0.0011 mrem/yr 0.00055 mrem/yr Revision 6 February 15, 1987

Figure F11.2-1A Sheets 1 through 5 of 5 intentionally deleted.

SGS-UFSAR Refer to plant drawing 205239 in DCRMS Revision 27 November 25, 2013

Figure F11.2-1B Sheets 1, 2 & 3 of 3 deleted.

intentionally Refer to plant drawing 205339 in DCRMS SGS-UFSAR Revision 27 November 25, 2013

11.3 GASEOUS WASTE SYSTEM The Gaseous Waste System (GWS) provides controlled handling and disposal of gaseous wastes generated during plant operation.

The system also supplies hydrogen and nitrogen to primary systems' components as required during normal operation.

The system is designed to minimize exposure to plant personnel and the general public, in accordance with Nuclear Regulatory Commission (NRC) regulations.

In this section, the system is described and evaluated.

11.3.1 Design Objectives Design objectives for the GWS are the following:

1.

To provide sufficient capacity and storage to process and store the volume of gaseous effluent expected for a period of 45 days

2.

To provide cover gas for the liquid holdup tanks

3.

To assure that releases of radioactive gaseous wastes are kept as low as practicable.

4.

To maintain releases below the limits set by 10CFR20

5.

To assure that exposures to the public are maintained below the design objective of 10CFRSO Appendix I The design criteria for the GWS are as follows:

The facility design shall include those means necessary to maintain control over the plant radioactive gaseous effluents.

Appropriate holdup capacity shall be provided for retention of gaseous effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment.

In all 11.3-1 SGS-UFSAR Revision 6 February 15, 1987

cases, the design for radioactivity control shall be justified 1) on the basis of 10CFR20 requirements, for both normal operations and for any transient situation that might reasonably be anticipated to occur and 2) on the basis of I 10CFR50, 67 dosage level limits for potential reactor accidents of exceedingly low probability of occurrence.

Gaseous waste facilities are designed so that discharge of effluents are in accordance with applicable governmental regulations.

Radioactive gases entering the GWS are collected in tanks to allow for decay and isotopic analysis.

The system design and operation is directed toward minimizing releases to unrestricted areas. Discharge streams are appropriately monitored and safety features are incorporated to preclude releases in excess of the limits of 10CFR20.

Radioactive gases are pumped by compressors through a manifold to one of the gas decay tanks where they are held for a suitable period of time to allow for decay.

Cover gases in the Nitrogen Blanketing System can be reused to minimize gaseous wastes.

intermittently at During normal operation, decayed gases are discharged a controlled rate from these tanks through the moni tared plant vent.

The system is provided with discharge controls.

11.3.2

System Description

During plant operations, gaseous wastes will originate from the following:

1.

Degassing reactor coolant discharge to the Chemical and Volume Control System (CVCS)

2.

Displacement of cover gases as liquids accumulate in various tanks 11.3-2 SGS-UFSAR Revision 23 October 17, 2007

3.

Miscellaneous equipment vents and relief valves

4.

Sampling operations and automatic gas analysis for hydrogen and oxygen in cover gases.

The GWS consists of two waste gas compressors and four waste decay tanks. During normal operation, the GWS supplies nitrogen to plant components.

Two liquid nitrogen storage tanks, each with a self contained (ambient) vaporizer are supplied.

One storage tank and its vaporizer is used at a tLme to supply the operating headers for both units. The pressure regulator in the operating header of each unit is set for 100 psig discharge.

Each operating header is backed up by a nitrogen (gaseous) cylinder manifold with a pressure regulator set at 90 psig.

When the operating header is below 100 psig, an alarm will alert the operator.

The backup header will come into service automatically at 90 psig to assure a continuous supply of gas. After the operating header has been switched over to the standby liquid nitrogen storage tank, and the operating header pressure restored to 100 psig, the flow from the backup header will drop to zero.

In addition to use as a backup nitrogen supply to the Waste Disposal System (WDS), the nitrogen (gaseous) cylinder manifold also supplies high pressure nitrogen gas for recharging accumulators.

A hydrogen cylinder manifold is included in the Gaseous Waste Disposal System. It serves as a backup supply for hydrogen feed to the volume control tank. Normal feed is from the bulk hydrogen Control System.

Most of the gas received by the WDS during normal operation is cover gas displaced from the eves holdup tanks as they fill with liquid.

Since this gas must be replaced when the tanks are emptied during processing, facilities are provided to return gas from the decay tanks to the holdup tanks. A backup supply from the nitrogen header is provided for makeup if return flow from the gas decay tanks is not available.

To avoid the possibility of hydrogen combustion in the vent header system while gas is being displaced from holdup tanks to the vent header, components discharging to the vent header system are restricted to those 11.3-3 SGS-UFSAR Revision 6 February 15, 1987

containing no air or aerated liquids and the vent header itself is designed to operate at a slight positive pressure (0.5 psig minimum to 4.0 psig maximum) to prevent in leakage. out leakage from the system is minimized with Saunders patent diaphragm valves, bellows seals, self-contained pressure regulators and soft-seated packless valves throughout the radioactive portions of the system.

Gases vented to the vent header flow to the waste gas compressor suction header.

One of the two compressors is in continuous operation with the second unit instrumented to act as backup for peak load conditions or failure of the first unit.

From the compressors, gas flows to one of four gas decay tanks.

The control arrangement on the gas decay tank inlet header allows the operator to place one tank in service and to select one tank for backup.

When the tank in service becomes pressurized to 92 psig, a pressure transmitter automatically closes the inlet valve to that tank, opens the inlet valve to the backup tank and sounds an alarm to alert the operator so he may select a new backup tank.

Pressure indicators are provided to aid the operator in selecting the backup tank.

Gas held in the decay tanks can either be returned to the eves holdup tank or discharged to the atmosphere if it has decayed sufficiently for release.

Generally, the last tank to receive gas will be the first tank emptied back to the holdup tanks which permits the maximum decay time before releasing gas to the environment.

However, the header arrangement at the tank inlet gives the operator the option to fill, reuse or discharge gas to the environment simultaneously without restriction by operation of the other tanks.

During degassing of the reactor coolant prior to a cold shutdown, for example, it may be desirable to pump the gas purged from the volume control tank into a particular gas decay tank and isolate that tank for decay rather than reuse the gas in it. This is done merely by aligning the control to open the inlet valve to the 11.3-4 SGS-UFSAR Revision 15 June 12, 1996

desired tank and closing the outlet valve to the reuse header. Simultaneously, one of the other tanks can be opened to the reuse header if desired, while another is discharged to atmosphere.

Before a tank is discharged to the environment, it is sampled and analyzed to determine and record the activity to be released, and then discharged to the plant vent at a controlled rate, and monitored for gross activity.

During operation, gas samples are drawn automatically from the gas decay tanks and automatically analyzed to determine their hydrogen and oxygen content.

There should be no significant oxygen content in any of the tanks, and an alarm will warn the operator if any sample shows 2 percent or higher by volume of oxygen.

This allows time to take required action before the combustible limits of hydrogen-oxygen mixtures are reached. Another tank is placed in service while the operator locates and eliminates the source of oxygen.

The system is controlled from a central panel in the Auxiliary Buildings.

Malfunction of the system is alarmed in the Auxiliary Building, and annunciated in the Control Room.

Building.

All system equipment is located in the Auxiliary The Unit 1

& Unit 2 auxiliary feedwater storage tanks are provided with a nitrogen purge/blanket system in order to control the dissolved oxygen concentration in the water.

Each nitrogen purge/blanket system is provided with a dedicated nitrogen source.

The GWS process flow diagrams are shown on Plant Drawings 205240 and 205340.

11.3-5 SGS-UFSAR Revision 27 November 25, 2013

11.3.3 system Design Gas Decay Tanks Four welded carbon steel tanks per unit are provided to contain waste gases (hydrogen, nitrogen, and fission gases).

Each tank conforms to ASHE Boiler and Pressure Vessel Code Section III, Class c.

Design data are as follows:

3 volume, each (ft }

Design pressure (psig)

Design temperature (°F)

Operating pressure {psig)

Operating temperature (°F)

Type Waste Gas Compressors 525 150 180 0 -

92 50 -

150 Vertical cylinder There are two waste gas compressors per system to provide continuous removal of gases discharged to the vent header.

Only one unit is normally in operation.

The second unit is provided for backup during peak load conditions, such as when degassing the reactor coolant or for service when the first unit is down for maintenance.

The compressors are water sealed, rotary, positive displacement units in which the water is used to displace and compress the gas being moved.

Each compressor has a capacity of 40 cfm at 105 psig.

The seal water is cooled, in a heat exchanger, by the component cooling water.

Makeup water for the seal is supplied to the compressor suction from the Component Cooling System.

Each compressor contains a mechanical seal to minimize leakage of seal water.

The compressor discharges a mixture of waste gas and water into the separator.

In the separator, the water is centrifuged out of the mixture and is accumulated in the bottom of the separator.

11.3-6 SGS-UFSAR Revision 15 June 12, 1996

The discharge from the separator is saturated at the discharge pressure and temperature of the gas.

At 40 cfm, 105°F cooling water and 105 psig discharge, water vapor carryover is based on the following inlet conditions:

Saturated N2 Saturated N2 65% by Volume N2 and 35%

H2 65% by Volume N2 and 35%

H2 Inlet Temperature Vapor Carryover 0.87 l.b/min 0.352 1b/min 0.019 lb/min 0.355 lb/min In order to assure that there will be sufficient pressure to circulate seal water at startup, the compressor discharge control valve on the separator is set to open at 50 psig.

Proper water level in the separator is maintained by means of a liquid high level transmitter and a low level alarm and makeup switch.

The liquid level transmitter actuates the high level drain valve. If the water level falls below the low level cutoff point, the low level switch opens the water makeup valve and water is introduced to the compressor through the inlet. Design data for the compressor are as follows:

11.3-7 SGS-UFSAR Revision 13 June 12, 1994

Compressor Number per unit Type Design flow rate, N2 (at 140°F, 2 psig) cfm Design pressure (psig)

Design temperature (°F)

Normal operating pressure (psig)

Suction Discharge Normal operating temperature (°F)

Compressor Motor H.P..

RPM volts Phase Cycle Rise Ambient temperature Dripproof Enclosure Class B Powerhouse insulation Nitrogen Manifold 2

Liquid piston rotary type 40 150 180 o.s - 4.0 0 -

92 70 - 130 25 3500 460 3

60 90°C 40°C A supply header from the Liquid Nitrogen System supplies nitrogen gas to purge the vapor spaces of various components, to reduce hydrogen concentrations or replace fluid in emptying tanks. Pressure controllers ( I-PIA-1066) which. switch from the normal bulk supply to a backup gaseous cylinder header, assure a continuous flow of gas.

Pressure regulator 1-PCV-1043 in the backup header is set at 90 psig which is lower than the 100 psig 11.3-8 SGS-UFSAR Revision 15 June 12, 1996

in the operating header.

When the operating header supply from one bulk tank is exhausted, the discharge pressure of this header will fall below the setpoint pressure of the backup header. ~hich will come into service automatically.

This system has the additional function of supplying N2 at 800 psi to the accumulator in the Safety Injection System.

If the need ever arises, this pressurized gas will inject borated liquid from the accumulators into the reactor coolant loops.

Design data for the manifold are as follows:

Type Automatic switching dual header Number per unit 1

Number of separate headers per package 2 Number of cylinders per header 18 Design flow rate, scfm 40 Design delivery pressure, psig 100 Station Bulk Lpw Pressure Nitrogen Supply A station bulk low pressure {LP) nitrogen supply package has been added to the above system to provide additional capability.

Two liquid nitrogen storage

tanks, each with a

self-contained vaporizer are supplied.

One storage tank and its vaporizer are used at a time to supply the operating headers for both units.

Design data are as follows:

11.3*9 SGS-UFSAR Revision 13 June 12. 1994

Type Argon Operating pressure (Max)

Design pressure (Max)

Design temperature Empty weight Hydrogen Manifold Vertical cylindrical, Double walled 55,866 scf 69,030 scf 67,470 scf 245 psig 249 psig 4,400 lbs A dual manifold serves as a backup to the Bulk Hydrogen System to supply hydrogen to the volume control tank and to maintain the hydrogen partial pressure as hydrogen dissolves in the reactor coolant.

A pressure controller (1-PIA-1065) which automatically switches from the normal system to the backup system, assures a continuous supply of gas.

The operation of the backup header is essentially the same as for the Nitrogen Manifold System.

Design data are as follows:

Type Automatic switching Number per unit 1

Number of separate headers per package 2

Number of cylinders per header 6

11.3-10 scs.. uFsAR dual header Revision 6 February 15, 1987

Design flow rate, scfm 30 Design delivery pressure, psig 100 Gas Analyzer Redundant gas analyzers, one in each Salem Unit and both cross-connected, are provided in accordance with the recommendations of NUREG-0472 to automatically monitor the concentrations of oxygen and hydrogen in the system, in order to indicate when the accumulation of these gases approaches an explosive mixture.

Upon indication by alarm that the oxygen level is approaching a hazardous level, provisions must be made to either isolate the component or purge with nitrogen to the GWS.

The gas analyzer has sui table connections for sampling when necessary from the following components:

Waste gas to plant vent Reactor coolant drain tank Spent resin storage tank Gas decay tanks (2 points) eves holdup tanks Boric acid evaporator and gas stripper Volume control tank Pressure relief tank Gas decay tank samples are analyzed continuously to ensure that the oxygen concentration remains less than or equal to 2 percent.

Separate feed lines with calibration gases are provided for analyzer calibration purposes.

The 11.3-11 SGS-UFSAR Revision 20 May 6, 2003 I

I

high-span calibration gas is nominally 4% oxygen, and low-span calibration gas is nominally 1% oxygen. The balance of the calibration mixtures consists of nitrogen, except for small amounts of hydrogen (between 1% and 2.5%). The gas mixture allows calibration of the analyzer to the profile expected in the sample stream at alarm conditions. Design data for the analyzers are as follows:

Oxygen By partial pressure measurement 0-5% O2 Range Hydrogen By partial pressure measurement 0-25% H2 Range Recorder printout (chart)

Waste Gas Decay Tank: every 3 minutes Sequential sampling (cover gas): each point All major equipment in the Gaseous Radwaste Disposal System is located outside of the Reactor Containment Building in the Auxiliary Building, Elevation 64 feet and 122 feet.

11.3-12 SGS-UFSAR Revision 31 December 5, 2019

Piping Gas piping is mainly carbon steel with stainless steel piping in some sections installed as part of modifications. Piping connections are welded except where flanged connections are necessary to facilitate equipment maintenance.

Valves exposed to gases are either carbon steel or stainless steel. Isolation valves are provided to isolate each piece of equipment for maintenance, to direct the flow of waste through the system, and to isolate storage tanks for radioactive decay.

Relief valves are provided for tanks containing radioactive wastes if the tanks might be over-pressurized by improper operation or component malfunction.

Codes and Standards Additional information is presented in Table 11.2-3 for system piping, valves and compressors.

11.3.4 Operating Procedures The gaseous wastes processed by this system consist primarily of hydrogen stripped from reactor coolant during boron recycle and degassing operations and nitrogen from the various tank cover gases and from the degassing operation.

These gases are discharged to the vent header which feeds the suction of the waste gas compressors.

One of the two waste gas compressors will be operating with the other compressor being on standby. The operating compressor maintains a vent header pressure of 0.5 to 4.0 psig. If the vent header pressure rises to 4 psig, the standby compressor automatically energizes. The compressors can be used to: 1) pump gas to the waste decay tanks; 2) transfer gas between tanks; and 3) pump gas directly to the CVCS holdup tanks.

11.3-13 SGS-UFSAR Revision 31 December 5, 2019

'l'o pump gas to the gas decay tanks 1 the operator selects two tanks at the auxiliary control panel No. 104: one to receive gas, and one for standby. When the tank in-service is pressurized to 92 psig, flow is automatically switched to the standby tank and an alarm alerts the operator to select a new standby tank. The decay tank being filled is sampled automatically by the gas analyzer and an alarm will alert the operator to a high oxygen content.

The tank must then be isolated and the operator is required to direct flow to the standby tank and select a new standby tank.

As the liquid in the CVCS holdup tanks is processed by the boric acid evaporator, gas must be provided as cover gas to replace the processed liquid.

The cover gas may be provided from any of the gas decay tanks or from the nitrogen supply.

The gas decay tank supplying the returning cover gas is selected manually at the auxiliary control panel No.

104 by opening the appropriate valve in the return line header.

To maximize total residence time for gas decay in the system, the last tank filled should be the first tank returned as cover gas.

A backup supply of gas to the holdup tanks is provided from the bulk nitrogen header for makeup when return flow is not available from the decay tanks.

Before a gas decay tank is discharged to the plant vent for release to atmosphere, a sample must be taken to determine activity concentration of the gas and total activity inventory in the tank.

Total tank activity inventory is determined from the activity concentration and pressure in the tank.

To release the gas, the appropriate local manual stop valve is opened to the plant vent and the gas discharge modulating valve is opened at the auxiliary control panel.

If the Plant Vent Radiation Monitor detects high activity during release, the modulating valve automatically trips closed. To reopen the valve, the switch must first be reset by returning it to the closed posit ion. The valve can then be repositioned.

The equipment which connects with the vent header system is limited in number.

Under normal operating conditions no air is permitted to enter the vent header.

During maintenance operations air could enter the boric acid evaporator vent condenser or the waste evaporator vent condenser.

During maintenance operations on either of these pieces of equipment, the valve on the equipment discharge line to the vent header is closed.

When maintenance operatj.ons are completed, and prior to opening the valves, the equipment is filled with nitrogen to purge the air.

During discharge, the nitrogen purge is continued.

No fluids can get into the vent header.

11.3-14 SGS-UFSAR Revision 23 October 17, 2007

The maximum allowable release rate of gaseous radioactivity is specified in the Offsite Dose Calculation Manual.

A record of all releases is kept.

11.3.5 Performance Tests Periodic inspection of waste gas compressors shall be done in accordance with the manufacturer's technical manual.

11.3.6 Estimated Releases HISTORICAL NOTE:

The radiological release values originally contained in this section were calculated in support of initial licensing and have been deleted.

Off-site releases during normal plant operations are controlled by the Radioactive Effluent Control Program.

Several sources of potential release of gaseous radioactivity to the environment have been identified.

Each is discussed separately below.

Gas Decay Tanks Gaseous wastes consist primarily of hydrogen stripped from coolant discharged to the eves holdup tanks during boron dilution, nitrogen and hydrogen gases purged from the eves volume control when degassing the reactor coolant, and nitrogen from the closed gas blanketing system. The gas decay tank capacity will permit adequate time to allow for decay of waste gas activity release based on 1-percent defective fuel clad, and 34 23 MWt power with daily load reductions to 50-percent power for several hours.

11.3-15 SGS-UFSAR Revision 29 January 30, 2017

Containment Purging Purging of the containment will take place infrequently, on the order of two to three times a year per unit, to keep concentrations of radioactive gases in the containment within specified limits to allow plant personnel to enter the containment periodically for maintenance and inspection.

Section 9.4 describes the methods employed to minimize release to the environment from containment purging.

Diaphragm valves are used in the GWS vent header to eliminate steam leakage.

All pipe connections in the vent headers are welded.

Therefore, there will be no effect on the annual release of gaseous radioisotopes.

11.3.7 Release Points Release points are shown on the system flow diagrams (Plant Drawings 205240 and 205340).

11.3.8 Dilution Factors See Section 11.3.9.

11.3.9 Estimated Doses The radiological release doses originally contained in this section were calculated in support of initial licensing and have been deleted.

Off-site doses during normal plant operations are controlled by the Radioactive Effluent Control Program and the ODCM.

11.3-16 SGS-UFSAR Revision 27 November 25, 2013

Isotope Kr 85 Kr 85m, 87, 88 Xe 133 TABLE 11.3-1 (H~storical Informat~on)

ESTIMATED ANNUAL GASEOUS RELEASE BY ISOTOPE(!)

FROM GAS DECAY TANKS

{Per Unit)

Activity Release to Environment Curies/yr Xe 133m, 135, 135m 138 5450 Negligible 2000 Negligible Total 7450 NOTE:

(1)

Based on 1 percent defective fuel, 3423 MWt core, load follow and 45 days holdup.

1 of 1 SGS-UFSAR Revision 19 November 19, 2001 I

Isotope H-3 Kr-85 Kr-85m Kr-87 Kr-88 Xe-133 Xe-133m Xe-135 Xe-135m I-131 SGS-UFSAR TABLE 11.3-2 (Histori.ca:t Information}

ESTIMATED TOTAL RADIOACTIVE GASEOUS RELEASES

{TWO UNIT BASIS}

Estimated Annual Release (Curies) 248 11450 195 113 309 26850 240 633 10 0.23 1 of 1 Revision 19 November 19, 2001 I

Annual Isotope Release (Curies)

H-3 248 Kr-85 11450 Kr-85m 195 Kr-87 113 Kr-88 309 Xe-133 26850 Xe-133m 240 Xe-135 633 Xe-135m 10 I-131 0.23 Totals 401000 NOTES:

(1) gamma energy (2) beta energy SGS-UFSAR TABLE 11.3-3 (Historical Information)

ESTIMATED OFF-SITE RADIATION EXPOSURES GASEOUS RADIOACTIVE RELEASES (Two Unit Basis}

Radiation ExEosure (mrem) at 1270 Meters, North Sector Finite Cloud*

Semi-Infinite Cloud(1)

Infinite Cloud(2)

Whole BodY: Dose Whole BodY: Dose Skin Dose

{rnrem)

(mrern)

(mrem) 0 0

0.0007 0.0033 0.012 1.12 0.0039 0.008 0.023 0.0059 0.046 0.073 0.058 0.15 0.053 0.227 0.55

1. 94 0.0062 0.0016 0.023 0.017 0.042 0.087 0.0003 0.0011 0.0004 negl negl negl 0.269 0.81 3.32 1 of 1 Revision 19 November 19, 2001 I

TABLE 11.3-4 (Historica1 Information)

POTENTIAL RADIATION EXPOSURE PATHWAYS TO MAN {GASEOUS}

Pathway External Exposure Ingestion of Milk SGS-UFSAR Individual Assumed Exposed Individual stands at nearest site boundary 100 percent of time Young child drinks entire intake of milk from nearby dairy farms.

Cows assumed to graze 9 months per year with grass making up to 100 percent of diet during that time.

1 of 1 Calculated Exposure 0.054 rnrem/yr

{whole body-finite cloud) 0.16 rnrem/yr (whole body-semi - infinite cloud}

0.66 rnrem/yr (skin - infinite cloud, beta energy only}

0.26 rnrem/yr (farm 4.1 miles NW of site)

Revision 19 November 19, 2001 I

Figure F11.3-1A Sheets 1, 2 & 3 of 3 deleted.

intentionally Refer to plant drawing 205240 in DCRMS SGS-UFSAR Revision 27 November 25, 2013

Figure F11.3-1B Sheets 1, 2 & 3 of 3 deleted.

intentionally Refer to plant drawing 205340 in DCRMS SGS-UFSAR Revision 27 November 25, 2013

11.4 RADIOLOGICAL MONITORING 11.4.1 Design Objectives Design objectives for the Radiation Monitoring System (RMS) are as follows:

1.

Warn of any radiation hazard which might develop

2.

Give early warning of a plant malfunction which might lead to a

radiation hazard or plant damage

3.

Provide assurance that personnel exposure does not exceed 10CFR20 limits

4.

Provide assurance that atmospheric releases will not exceed the design objectives of 10CFR50

5.

Record the activity present at various plant locations

6.

Provide data for radiological analyses and reports

7.

Monitor process system filters for radiation buildups 11.4.2 Radiation Monitoring System The RMS provides instrument channels, located at selected points in and around the plant to detect, compute, and record the radiation levels.

In the event the radiation level should rise above a desired setpoint, an alarm will be initiated in the Control Room.

The RMS operates in conjunction with regular and special radiation surveys and with chemical analyses performed by the plant staff to meet the radiological monitoring design objectives presented in Section 11.4.1.

The RMS signal processing equipment is centralized in six cabinets for Unit 1 and in three cabinets for Unit 2. High reliability and 11.4-1 SGS-UFSAR Revision 27 November 25, 2013

ease of maintenance are emphasized in the design of this system.

Cabinet equipment is equipped with sliding channel drawers for rapid replacement of units, assemblies and entire channels.

It is possible to completely remove the various chasses from the cabinet, after disconnecting the cables from the rear of these units.

The components of the RMS are designed to meet or exceed the requirements of normal and DBA conditions for temperature, humidity, pressure and radiation, as stated in the Salem Generating Station Environmental Design Criteria.

The RMS is divided into the following subsystems:

1.

The Process Radiation Monitoring System monitors various gaseous and liquid streams for indication of increasing radiation levels, and all identified effluent paths to establish the quantity of radioactivity being discharged to the environment.

2.

SGS-UFSAR The Process Filter Monitoring System monitors the buildup of radioactivity on various process filters to warn of unexpected radiation and to indicate the need for changing or cleaning the filter.

11.4-2 Revision 19 November 19, 2001

3.

The Area Radiation Monitoring System monitors radiation levels in various locations of the plant to warn personnel of a deteriorating radiological condition.

It is also useful in assessing the spread of radioactivity in a given area.

11.4.2.1 Radiation Monitoring System Description 11.4.2.1.1 Radiation Monitoring System-Unit 1 Except for 1R18, 1R2, 1R3, 1R4, 1R5,

1R6A, 1R 7,

1R9, 1R1 OA,

1R11A, 1R12A,

1R128, 1R13A,

1R138, 1R15,

1R17A, 1R178,

1R18, 1R19A,

1R198, 1R19C,

1R19D, 1R41A, 1R418, 1R41C,

1R41D, 1R4 6A, 1R4 68, 1R4 6C, 1R4 6D, 1R53A, 1R538, 1R53C, and 1R53D, the Unit 1 RMS consists of analog channels which monitor radiation levels in various plant locations and operating systems.

Monitors 1R18, 1R41A, 1R418, 1R41C, 1R4 6A, 1R4 68, 1R4 6C, 1R4 6D, 1R53A, 1R538, 1R53C and 1R53D have microprocessor-based electronics.

A digital control and display module is located in the Control Equipment Room for monitors 1R18,

1R41A, 1R418 and 1R41C, 1R4 6A, 1R4 68, 1R4 6C, 1R4 6D, and in the Relay Room for monitors 1R53A,

1R538, 1R53C and 1R53D.

The output from each detector is transmitted via cables to the RMS cabinets in the Control Room area where the radiation level is indicated on a meter and pre-selected channels are recorded on a multipoint recorder.

For area monitors, the radiation level is also indicated locally at the detector.

High radiation level alarms are annunciated on the Control Room overhead annunciator and further identified at the RMS cabinets.

For area

monitors, a high radiation level is also alarmed at the detector location, except for area monitors located in the Control Room.

Each channel contains a completely integrated modular assembly, which includes the following.

1.

Level Amplifier/Discriminator Discriminates and amplifies the detector output to provide a discriminated and shaped pulse output to the log level amplifier.

2.

Log Level Amplifier -Accepts the shaped pulse of the level amplifier output,* performs a log integration (converts total pulse rate to a logarithmic analog signal),

and amplifies the resulting output for suitable indication and recording.

Note, monitors 1R18, 1R41A/8/C/D and 1R53A/8/C/D have microprocessor based electronics that provide a direct digital conversion of detector output to CPM.

1R18 contains two channel inputs and monitors both Control Room area inlet ducts as illustrated on Figure 11.4-9.

11.4-3 SGS-UFSAR Revision 25 October 26, 2010

3.

Power Supplies - Individual power supplies are contained in each drawer for furnishing the positive and negative voltages for the transistor circuits, relays and alarm lights and for providing the high voltage for the detector.

4.

Test-Calibration Circuitry-These circuits provide a

pre-calibrated pulsed and/or analog signal to perform a channel test, and a solenoid operated radiation check source to verify the channel's operation.

A light on the Control Room overhead annunciator indicates when any channel is in the test-calibrate mode, except 1R1B.

5.

Radiation Level Meter - This meter, mounted on the assembly drawer, has 101 10 6 a scale callbrated logarl thmlcally from to for process monitor channels and in mR/hr for area channels.

For monitors

1R1B, 1R41A/B/C/D, and counts per minute and filter monitor 1R53A/B/C/D, the displays are digital.

Pre-selected signals are also recorded and displayed in the Control Room area.

6.

Indicating Lights -

These lights indicate high radiation levels and circuit failures. A light on the Control Room overhead annunciator is actuated on a high radiation signal and a yellow light on the RMS recorder panel indicates which channel.

provides discriminate 1R1B channel alarms.

applicable for 1R1B.

The Control Room alarm CRT The yellow light is not

7.

Bistable Circuits-Two bistable circuits are provided, one to alarm on high radiation (actuation point may be set at any level over the range of the instruments) and one to alarm on loss of signal (circuit failure).

8.

Check Source -A remotely-operated long half-life radiation check source is furnished in each channel.

The energy emissions are similar to the radiation energies being monitored.

The source strength is sufficient to cause a visible increase in the meter indication.

During checksource operation on R1B indication is frozen for both channels. If insufficient count rate is achieved (check source count rate compared against a setpoint), a norm failure alarm is provided.

11.4-4 SGS-UFSAR Revision 19 November 19, 2001

RMS channels 1R1 7A, 1R1 78, 1R2, 1R3, 1R4,

1R18, 1R19A,

1R5, 1R19B,

1R6A, 1R7, 1R9, 1R10A, 1R13A, 1R13B, 1R15,

1R19C, 1R19D, and 1R4 0 are microprocessor based digital instrumentation systems. This equipment is designed to power, operate, and monitor various types of radiation detectors.

The system consists of a detector (GM tube, scintillation, or ion chamber), a local monitor, and a remote monitor.

Local monitors perform pulse discrimination and shaping.

Also, all calibration constants are stored in the local monitor.

Remote monitors communicate with the local monitors via serial communication ports and provide analog outputs to the plant computer and indicators and contact outputs to alarm and interlocks.

RMS channels 1R11A, 1R12A, and 1R12B are microprocessor based digital instrumentation systems. A sample skid with a local microprocessor monitors containment atmosphere for particulate iodine and noble gas. Detector signals are processed locally. Alarm contacts are located at the microprocessor.

Calibration and database constants are stored locally. A remote display unit communicates with the local microprocessor via a serial communication port. The remote display provides analog outputs to plant computers, indicators and contact outputs for annunciation.

11.4.2.1.2 Radiation Monitoring System-Unit 2 The Unit 2 RMS is primarily a microprocessor-based digital monitoring system.

The system is basically a two-tiered structure with local field units and remote units in the Control Equipment Room.

A simplified diagram of the general system structure is shown on Figure 11. 4-1.

2R1B has microprocessor-based electronics and a digital control and display assembly in the Control Equipment room as illustrated on Figure 11.4-9.

Local Field Units The local digital field units are located at selected points in the plant to detect airborne radioactivity, filter buildup radioactivity and process system radioactivity.

Each field unit consists of a

detector and a

microprocessor I electronics cabinet.

The units detect, compute, and indicate radiation data at their respective location.

A digital display is provided for radiation level indication.

Indicating lights are provided for alarm, warning, failure, and check source operation information.

Auxiliary relay contacts are available for control system functions to indicate alarm, warning, and failure conditions.

Keylock controls are used for testing, calibration, and entering data such as setpoints, conversion factors, and confidence levels.

11.4-5 SGS-UFSAR Revision 27 November 25, 2013

This page intentionally left blank 11.4-6 SGS-UFSAR Revision 27 November 25, 2013

The system was designed to provide for the safe operation of the plant, to assure that personnel exposure does not exceed 10CFR20 limits, and to assure that environmental releases do not exceed Technical Specification limits.

The Unit 2 system was designed to meet the same requirements as the Unit 1 system.

The two Radiation Monitoring Systems perform essentially the same functions.

There are,

however, some differences in sensitivities, detector types, and monitoring channels.

11.4.2.2 Process Radiation Monitoring System Channel Description The process monitors are utilized for monitoring process systems for potential radiation leakage and effluent discharge paths for normal releases and those following potential accidents.

liquid or gas sampling system.

The monitors typically incorporate an offline Some of the units monitor the process stream directly.

Typical functional block diagrams of the process monitors are shown on Figures 11.4-2 through 11.4-7.

The Process Radiation Monitoring System is summarized in Tables 11. 4-1 and 11.4-2 and consists of the following radiation 11.4-7 SGS-UFSAR Revision 27 November 25, 2013

monitoring channels. The prefix numbers indicate monitors associated with Unit 1 or Unit 2.

Control Room Area Intake Duct Monitors (1-R1B and 2-R1B)

The Control Room Intake Air Radiation Monitoring System is a shared system. The R1B monitors provide redundant functions to monitor air drawn into the Control Room through the Unit 1 and Unit 2 air intakes to the Control Room. The dual channel processors with beta scintillation detectors in each of the Unit 1 and Unit 2 Control Room intake ducts provide the redundant initiation signals to place the ventilation system into its accident-pressurized mode of operation.

In addition, the monitoring system provides continuous indication and recording of the radiation levels and annunciates in the Control Room the failures of the radiation monitoring equipment and the development of warning and alarm level radiation conditions.

This is a safety-related channel.

Containment - Air Particulate Monitors (1-R11A and 2-R11A)

These monitors are provided to measure air particulate beta radioactivity in the containment and to ensure that the release rate through the plant vent during purging is maintained below specified limits.

For Unit 1, channel R-11A takes a continuous air sample from the containment atmosphere.

For Unit 2, channel 2-R11A takes a sample from the containment. The sample is drawn from the containment through a

closed, sealed system and monitored by a

scintillation counter-moving filter paper detector assembly. The filter paper collects 99 percent of all particulate matter greater than 1 micron in size on its constantly moving surface and is viewed by a photomultiplier-scintillation crystal combination.

11.4-8 SGS-UFSAR Revision 25 October 26, 2010

The sample is returned to the containment or vent, depending on which source is being monitored.

The detector assembly is in a completely enclosed housing.

The detector is a hermetically sealed photomultiplier tube scintillator combination.

The filter paper has a 25-day minimum supply at normal speed.

Lead shielding is provided to reduce the background level to where it does not interfere with the detector's sensi ti vi ty.

The filter paper mechanism, an electro-mechanical assembly which controls the filter paper

movement, is provided as an integral part of the detector unit.

Channel 1R11A (particulate) monitors containment for leak detection and effluent releases.

Digital indication for 1R11A is located at the local and remote monitors. The monitors are set to indicate radiation from 10 1 to 10 6 cpm.

The local monitor provides Normal, Warn and Alarm indications.

The remote monitor in the Control Equipment Room, in addition to display and status indications on the monitor panel, provides analog outputs to the Safety Parameter Display System (SPDS), Plant Computer P250, indicators on Panel 1RP1, indication of high radiation on Panel 1RP1, and High/Trouble alarm on the Overhead Annunciator.

Containment/Plant Vent Radioactive Gas Monitors (1-R12A, 2-R12A, 1-R41C and 2-R41D)

These monitors are provided to measure gaseous radioactivity in the containment, and to ensure that the release rate through the plant vent during purging is maintained below specified limits.

High radiation level initiates closure of the containment purge supply and exhaust duct valves and pressure relief line valves.

discharge valve.

For Unit 2, high radiation level also closes the waste gas For Unit 1, channel 1-R12A takes a continuous air sample from the containment atmosphere.

Channel 1-41D samples only the plant vent.

For Unit 2, channel 2-R12A takes a sample from the containment and channel 2-R41D from the plant vent.

All samples reach the gaseous detector after passing through the air particulate monitor or an air particulate sampler (1-R41D only).

The sample is constantly mixed in the fixed, shielded volume, where it is viewed by beta scintillator.

The sample is then returned to the source being monitored.

The detector assembly is in a completely enclosed housing containing a beta-gamma sensitive detector mounted in a constant gas volume container.

Lead shielding is provided to reduce the background level to a point where it does not interfere with the detector's sensitivity.

11.4-9 SGS-UFSAR Revision 25 October 26, 2010

Channel 1R12A (noble gas) monitors containment for effluent releases.

Digital indication for 1R12A is located at the local and remote monitors. The monitors are set to indicate radiation from 10 1 to 10 6 cpm.

The local monitor provides

Normal, Warn and Alarm indications and provides an alarm relay contact for initiating closure of containment ventilation closure/isolation valves 1VC1, 4, 5, and 6 for Modes 1, 2, 3, 4 & 5. The remote monitor in the Control Equipment

Room, in addition to display and status indications on the monitor panel, provides analog outputs to the Safety Parameter Display System ( SPDS), Plant Computer P250, indicators on Panel 1RP1, indication of high radiation on Panel 1RP1, and High/Trouble alarm on the Overhead Annunciator.

Containment Fixed Filter Iodine Monitor (1-R12B, 2-R12B)

Iodine is one of the more prominent isotopes requiring special surveillance.

The containment monitoring system has been designed so that the sample flows first through the filter paper assembly and then through a charcoal cartridge.

It is a scintillation type detector.

For Unit 1, the sample is drawn from the containment for channel 1-R12B.

Channel 1-R41C samples only the plant vent.

For Unit 2, channel 2-R12B takes a sample from the containment.

High radiation level initiates closure of the containment purge supply and exhaust duct valves and pressure line relief valves.

The abnormal conditions are alarmed in the Control Room and Control Equipment Room. A solenoid-operated check source is provided to give an instant checkout of the system functional status.

Channel 1R12B (iodine) monitors containment for effluent releases.

Digital indication for 1R12B is located at the local and remote monitors. The monitors are set to indicate radiation from 10 1 to 10 6 cpm. The local monitor provides

Normal, Warn and Alarm indications and provides an alarm relay contact for initiating closure of containment ventilation closure/isolation valves 1VC1, 4, 5, and 6.

The remote monitor in the Control Equipment Room, in addition to display and status indications on the monitor panel, provides analog outputs to the Safety Parameter Display System (SPDS), Plant Computer P250, indicators on Panel 1RP1, indication of high radiation on Panel 1RP1, and High/Trouble alarm on the Overhead Annunciator.

Note The containment radiation monitors (channels R11, R12A, and R12B) have elements common to all three channels of the particulate/noble gas/ iodine monitoring assembly.

These are described as follows:

11.4-10 SGS-UFSAR Revision 24 May 11, 2009

1.

The flow control assembly includes a pump unit and selector valves that provide a representative sample (or a "clean" sample) to the detectors.

2.

The pump unit consists of:

a.

A pump to obtain the air sample

b.

A flowmeter to indicate the flow rate

c.

A flow control valve to provide flow adjustment

d.

A flow alarm assembly to provide low and high flow alarm signals

3.

Selector valves are used to direct the desired sample to the detector for monitoring and to block flow from the sampling area when the channel is in the maintenance or "purging" condition.

4.

A temperature sensor is used to protect the system from high temperature. This unit automatically closes the inlet motor operated valve upon a high temperature condition.

5.

Purging is accomplished with a valve control arrangement whereby the normal sample flow is blocked and the detector purged with a "clean" sample. This facilitates detector calibration by establishing the background level and aids in verifying sample activity level.

6.

For Unit 1, the flow control panel in the Control Room radiation monitoring racks permits remote operation of the flow control assembly.

By operating a sample selector switch on the control panel, either the containment or a local "clean" sample may be monitored. For Unit 2, these controls are located on 2RP1 in the control room.

7.

Deleted.

8.

Containment isolation valves are provided for the containment sample piping. In addition, the containment isolation valves (for regular and backup flow paths) limit switch auxiliary relays are wired to the Auxiliary Annunciator System to provide a 1R11/12 loss of flow path alarm whenever both backup and regular flow paths are lost.

9.

For Unit 1 and Unit 2, the containment particulate and gaseous monitors (1-R11A, 1-R12A, 2-R11A and 2-R12A) are also used as part of the Reactor Coolant Leak Detection System.

11.4-11 SGS-UFSAR Revision 30 May 11, 2018

Alarm lights are actuated by the following:

1.

Flow alarm assembly (low or high flow)

2.

The pressure sensor assembly (high pressure)

3.

The filter paper sensor (paper drive malfunction)

4.

The pump power control switch (pump motor on)

Containment Fan Cooler Radiation Monitors (1-R13A and B and 2-R13A and B)

Service water is used as the cooling medium for the containment fan coolers and could be contaminated if the cooling coil leaks.

Since the Service Water System discharges to the river, the fan cooler units will be monitored for radioactivity.

This is done through the use of two monitors for the five fan coolers. Each monitor employs an in-line detector.

in the Control Room are also provided.

Condenser Air Removal Gas Monitors (1-R15 and 2-R15)

Remote alarms and readout This channel monitors the discharge from the condenser air removal exhaust header for gaseous radiation which is indicative of a primary to secondary system leak.

The gas discharge is routed to the plant vent.

In Unit 1 a gamma sensitive scintillation detector is used to monitor the radiation level.

The detector is inserted in an inline fixed volume container which includes adequate shielding to reduce the background radiation to where it does not interfere with the detector's sensi ti vi ty.

coupled to a photomultiplier tube is used.

11.4-12 SGS-UFSAR In Unit 2, a gamma scintillator Revision 27 November 25, 2013

Component Cooling Liquid Monitors (1-R17A, Band 2-R17A,B)

These channels continuously monitor the component cooling water for radiation.

Leakage from the Reactor Coolant System and other systems' components to the component cooling water is detected by a scintillation counter located in an inline well.

A high radiation level alarm signal initiates closure of the gas valve located in the component cooling surge tank vent line to prevent gaseous radiation release.

Waste Disposal System Liquid Effluent Monitors (1-R18 and 2-R18)

This channel continuously rnoni tors all Waste Disposal System liquid releases from the plant.

Automatic valve closure action is initiated by monitor after a high radiation level is indicated and alarmed in the Control Room.

A scintillation counter in a fixed volume assembly monitors liquid effluent as it is discharged.

Remote indication and annunciation are provided on the Waste Disposal System control board.

Stearn Generator Blowdown Liquid Monitors (1-R19A, B, C, D, and 2-R19A, B, C, D)

Each of these channels (four channels per unit) monitors the liquid phase of the steam generators for radioactivity, which would indicate a primary-to-secondary system leak.

The four steam generator blowdown sample lines each have a radiation monitor.

In SGS-UFSAR 11.4-13 Revision 21 December 6, 2004

Unit 1, the monitors are located in the Sampling Room where the blowdown sample has been cooled.

In Unit 2, an offline sampling system is used.

A high radiation alarm signal will close the No.

12

( 22) steam generator blowdown tank inlet valves and the steam generator blowdown isolation valves on the affected steam generator.

Letdown Line Monitors (1-R31A and 2-R31)

The Letdown Monitoring System for each unit consists of a single channel for monitoring total gross activity of the letdown line concentration.

These monitors are also called failed fuel monitors.

This purpose is to detect the failure of the cladding of one or more fuel elements by the gamma emission of fission products released into the reactor coolant.

1.

Unit 1 System The system continuously measures gamma radiation intensity in a

continuously flowing sample stream of reactor coolant water using a gamma scintillation detector.

Besides continuous indication of the reactor coolant activity, high radiation conditions are alarmed in the Control Room.

6 The detector is capable of measuring up to 10 cpm at which position it will reach the saturation condition.

Provision is made for desensitizing the system by two or more decades to compensate for permanent activity buildup resulting from long-term normal operation.

This 11.4-14 SGS-UFSAR Revision 15 June 12, 1996

is accomplished by insertion of a lead spacer between the sensitive end of the detector and the letdown line.

2.

Unit 2 System A gamma scintillator is used to monitor the total gross gamma letdown line activity concentration.

Signal processing is performed by the digital RMS to provide data on a significant increase of gross gamma activity.

A significant increase of gross gamma activity would be indicative of a fuel cladding failure.

The detector is capable of measuring up to 1 X 10 9 cprn.

Provision is made for desensitizing the system by two or more decades to compensate for permanent activity buildup resulting from long-term normal operation.

This is accomplished by insertion of a lead spacer between the sensitive end of the detector and the letdown line.

Evaporator and Feed Preheaters Condensate (1-R36)

Heating steam is supplied to the boric acid and waste evaporators and feed heater.

Condensate from the evaporators and feed heater is returned to the condensate receivers from whence it is pumped back to the heating boiler. Stearn is used in the tubes of the evaporators and in the heater for process heating.

Since the evaporators and heater can contain radioactive fluids, a tube rupture could result in a contamination of the Condensate System, Heating Boiler System, and Heating Stearn System.

11.4-15 SGS-UFSAR Revision 27 November 25, 2013

This channel continuously monitors the activity in the common condensate piping from each unit's evaporators.

This channel employs an offline sampler. A high radiation level alarm will automatically close the condensate line valve for each unit's evaporator packages.

Alarm and indication are provided in the Control Room.

A manually valved drain is provided for disposal of any contaminated condensate to the Waste Disposal System.

Nonradwaste Basin Discharge (2-R37)

The nonradwaste basin provides a potential release path due to the fact that steam generator blowdown is offline sampler is provided nonradwaste basin.

directed to the basin during plant startup.

An to continuously monitor the discharge from the Plant Vent High Range Monitors (1-R41B-D and 2-R41B-D, 1-R45 and 2-R45)

The Plant Vent High Range Monitors consist of High Range Noble Gas Monitors (1-R41B-D and 2-R41B-D) and High Range Particulate and Iodine Sampling Skids (1-R45 and 2-R45).

The High Range Noble Gas Monitors comply with NUREG-0737, Item II.F.1, and the intent of Regulatory Guide 1. 97, type C

& E Category 2 requirements.

The system provides a sampling capability of 10 5

!-LCi/cc for noble gases.

The monitors are safety grade and qualified for the post-accident environment.

However, the monitors do not perform any safety related function.

The system is comprised of four noble gas channels as follows: low range noble gas (R41A),

intermediate range noble gas (R41B), high range noble gas (R41C) and composite noble gas (R41D).

Microprocessor software logic is used for the composite channel to display the effluent release rate based on low-, intermediate-or high-range noble gas concentrations (R41A, B or C, respectively) and the plant vent flow rate.

The High Range Particulate and Iodine Sampling Skids (1-R45 and 2-R45) comply with NUREG-0737, Item II.F.1, and the intent of Regulatory Guide 1.97, type E Category 3 requirements.

The 2

system provldes a sampllng capablll ty of 10

!-LCi/cc for iodines and particulates.

Nonsafety-related samplings conditions.

SGS-UFSAR lines heat tracing during plant is provided to preclude outages coincident with 11.4-16 freeze-up adverse of the weather Revision 28 May 22, 2015

Main Steam High Range Monitors (1-R46A-D and 2-R46A-D)

The Main Steam High Range Monitoring System complies with NUREG-0737, Item II.F.1, and the intent of Regulatory Guide 1.97. The system provides a detection capability of 103 ci/cc. The monitors are safety grade and qualified for the post-accident environment. Channels R46A-D each monitor one of the main steam lines.

Main Steam Line N16 Radiation Monitors (1R53A/B/C/D and 2R53A/B/C/D)

The Main Steam Line N16 Radiation Monitoring System is one of the means used to detect and trend primary-to-secondary leakage in the main steam generators. It consists of four channels; each channel continuously monitors the N16 gamma radiation from one of the main steam lines. The detectors are high temperature NaI (T1) gamma scintillators with an integral Am241 check source. They are located upstream of the mixing bottle and as close to the main steam lines as practical. Detector output is processed by a multi-channel-analyzer, and the count rates in two energy windows are monitored for each channel. The high energy window is sensitive only to N16. Because N16 is present only during power operation, this monitor is used only during Mode 1.

The monitor is not included in the Reg. Guide 1.97 monitoring systems, has no interlocks with other monitors or systems, and is non-safety grade.

Technical Support Center (1R51)

The Technical Support Center monitors the radiation inside of the outside air intake duct.

11.4.2.3 Process Filter Monitoring System Channel Description Area-type radiation monitors are provided on these liquid (process) filters to determine when they should be replaced by indicating the level of activity given off by the filter. A high radiation level alarm is initiated in the Control Room. A radiation indicator and alarm light are located at the filter.

The filters which are monitored include the following:

1.

Reactor coolant filters (1-R26 and 2-R26)

2.

Condensate filters (1-R40 and 2-R40) 11.4-17 SGS-UFSAR Revision 30 May 11, 2018

-1 4

All Unit 1 process filter monitors are GM tubes and have a range of 10

-10 mR/hr.

All Unit 2 process filter monitors are ion chambers or Geiger Mueller detectors and have a range of 10-1 -10 6 mR/hr.

They perform no control function.

The following liquid (process) filters are monitored for differential pressure and radioactivity to determine when the filters should be replaced:

1. Seal water injection filters

2. Seal water filters

3. Liquid waste filters

4. Spent fuel pool filters

5. Spent fuel pool skimmer filters

6. Refueling water purification filters

7. Ion exchange filters 11.4.2.4 Area Monitoring System Channel Description Each channel consists of either a gamma sensitive GM tube detector or an ion chamber detector.

Functional block diagrams of the area monitors are shown on Figures 11.4-7 and 11.4-8.

For the Unit 1 analog channels, a meter is mounted on the front of each indicating-control module and is calibrated logarithmically from 0.1 mR/hr to 10 R/hr.

In addition, one of the containment area monitors in Unit 1 is calibrated from 10 mR/hr to 1000 R/hr.

A remote meter, calibrated logarithmically from 0.1 mR/hr to 10 R/hr (or from 10 mR/hr to 1000 R/hr) is mounted at the detector assembly.

Radiation Monitoring System cabinet alarms consist of indicator lights for high radiation and detector or circuit failure.

The remote meter and alarm assembly at the detector contains a

red indicator light and an audible alarm which are actuated on high radiation.

Both units have containment monitors capable of indicating area radiation from 1R/hr to 107R/hr.

Unit 1 digital channels 1R2, 1R3, 1R4, 1R5,

1R6A, 1R7, 1R9, 1R10A, and 1R40 provide digital indication at the local and remote monitors.

The monitors are set to indicate radiation from 0.1 mR/h to 10 R/h.

The local monitors provide Normal, Fail, and Alert indications and initiates buzzer and beacon light on high alarm.

The remote monitors in the Control Equipment Room, in addition to display and status indication on the monitor panel, provide analog outputs to the plant computer

( P250) and indicators on Panel 1RP1, indication of high radiation on Panel 1RP1, High/Trouble alarm on the Overhead Annunciator, and interlock contacts when required.

Tables 11.4-3 and 11.4-4 give a summary of System.

A listing of the area monitors is additional or special information added.

11.4-18 SGS-UFSAR the Area Radiation Monitoring also included below with any Revision 26 May 21, 2012

1.

Control Room Area (Channel 2-R1A, 1-R1A) -

This channel continuously monitors the Control Room area.

This area monitor does not have its own integral flashing beacon and horn since it is located in the Control Room and an alarmed condition is indicated by the annunciator and audible alarm (Unit 2 is provided with LED alarm indication and an adjustable volume horn) vital power supply.

This is a non-safety-related unit with a

2.

Containment Area (Low Range)

(1-R2, 2-R2)

3.

Radiochemistry Laboratory (R3)

4.

Charging Pump Room (1-R4, 2-R4)

5.

Fuel Handling Building (Channels 1-R5 and 2-R5) -

These channels continuously monitor the fuel storage areas.

A high radiation alarm from either unit will initiate charcoal filtration of the Fuel Handling Building atmosphere.

The Fuel Handling Accident in the Fuel Handling Building was analyzed without credit for filtration by the Fuel Handling Building Ventilation System.

For Unit 2

the high radiation alarm will automatically start the exhaust fans.

In addition to the integral alarm horn and flashing beacon, these units actuate an emergency evacuation horn in the building and radiation alert lights outside of the building.

vital power supply.

Each unit is on a separate

6.

Sampling Room (R6A)

7.

In-core Seal Table (1-R7, 2-R7)

8.

Fuel Storage Area (1-R9, 2-R9) -

These channels continuously monitor SGS-UFSAR the fuel storage areas.

A high radiation alarm from either unit will automatically start the exhaust fans (Unit 2

only) and initiate charcoal 11.4-19 Revision 22 May 5, 2006

filtration of the Fuel Handling Building atmosphere. The Fuel Handling Accident in the Fuel Handling Building was analyzed without credit for charcoal filtration by the Fuel Handling Building Ventilation System.

In addition to the integral alarm horn and flashing beacon, these units actuate an emergency evacuation horn in the building and radiation alert lights outside of the building. Each unit is on a separate vital power supply.

9.

Containment Personnel and Equipment Hatches (1-R10A, B and 2-R10A, B)

10. Counting Room (R20B)

11. Containment Area (High Range) (1-R44 A and B and 2-R44 A and B) -These channels continuously monitor the containment area and are provided with a special ion chamber detector for extended range capability in a post-accident environment. This is a safety-related unit with a vital power supply.

12. Public Service Control Point (R23)

13. Fuel Handling and Cask Handling Cranes (1-R32 A and B, 2-R32 A and B) -

These channels are not connected to the central Radiation Monitoring System and are not provided with integral horns and flashing beacons. A flashing beacon and alarm bell on the cranes are initiated.

14. Mechanical Penetration Area (1-R34 and 2-R34)

15. Condensate Filter Area (1-R40 and 2-R40)

16. (Deleted)

17. (Deleted) 11.4-20 SGS-UFSAR Revision 31 December 5, 2019

11.4.3 Sampling Samples are taken as required by the plant Technical Specifications. The plant vent will be continuously monitored for gross radioactivity. Additionally, a fixed paper particulate filter followed by a charcoal cartridge (Cesco or equivalent) is installed, both of which will be changed weekly. The charcoal cartridge and particulate filter will be analyzed by gamma spectroscopy within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from change out. A sample will be taken manually at a frequency not to exceed monthly, and an isotopic analysis performed to determine the identity and quantity of noble gases.

The sample will be taken at a time when there are no gas decay tank releases or containment purges in progress, since these releases are not representative of a continuous release. Therefore, gas decay tank releases and containment purges will be analyzed isotopically on a batch basis.

The method employed will be a grab sample, utilizing a gas collection device, taken from either of two sampling points in the plant vent. These samplers are the same as those used in conjunction with the continuous plant vent gas monitors outlined in Section 11.4.2.2.

In order to ensure that the inline filter and cartridge sample is representative of the plant vent exhaust gas, a weekly isotopic analysis of the particulate filter and cartridge is performed. This isotopic inventory is used to determine the isotopic composition of the plant effluent. In the event that this equilibrium is upset by a refueling, process change, or a deviation of greater than 20 percent in the isotopic ratio established from the previous isotopic analysis (this does not apply while releasing gas decay tanks or during containment purging), a new isotopic analysis will be performed.

11.4-21 SGS-UFSAR Revision 30 May 11, 2018

In order to ensure sampling during a radiological emergency that might render the normal plant vent sampling station uninhabitable, a supplemental Plant Vent Sampling System is provided.

The supplemental Plant Vent Sampling System is located in a region of the plant that is expected to be habitable during all accident conditions (west side of the Fuel Handling Building).

The supplemental system is designed such that it can be used during normal or accident conditions.

The sample lines are heat traced to help ensure that a representative sample is being delivered to the supplemental sampling station.

The controlled facilities ventilation duct is equipped with offline particulate and iodine samples that operate in parallel with its system operation.

The cartridges are removed and analyzed whenever necessary.

Samples are taken on each batch of liquid waste released.

Station records contain the quantity and concentration of radioactive isotopes, the volume of each batch and estimates of the water flow for dilution.

Each sample is analyzed for principal gamma emitters.

Composites are prepared from each batch released during a month and analyzed for the principal gamma emitting nuclides, fission and activation products, gross alpha, and tritium.

A quarterly composite analysis is also performed for Sr-89, Sr-90 and Fe-55.

The sensitivities and frequencies of analyses comply with the requirements of Salem Technical Specifications.

Information on sampling of the containment atmosphere and the Reactor Coolant System is included in Section 9.3.

11.4-22 SGS-UFSAR Revision 18 April 26, 2000

11.4.4 Inservice Tests and Calibrations Radiation monitors of the Radiation Monitoring System are initially calibrated to standards traceable to the National Bureau of Standards, most often referred to as "primary calibration". The area monitors undergo a range calibration by exposing the detectors to at least three radiation intensities from a Co-60 or Cs-137 source.

The exception is the Containment High Range Area Monitors, which are calibrated with at least one intensity.

The liquid and gas process monitors undergo "wet" isotopic calibrations with isotopes of an average energy comparable to those of the isotopes expected to be monitored.

At least three concentrations of isotopes are used in calibrating the instruments.

Beyond the initial primary calibration all radiation detectors undergo point source calibrations in a fixture with repeatable geometry, most often referred to as a "secondary calibration" or "transfer calibration".

This fixture and the button sources are utilized for periodic field calibrations. It is typical practice to use three secondary sources of the same isotope to validate detector linearity and stability; however, single point calibrations using secondary sources where detectors are inherently linear are acceptable.

In addition, the detectors are provided with check sources which can be used to indicate functional operability.

Tests, functional checks and calibrations are performed periodically in accordance with Technical Specification requirements and operating procedures.

11.4-23 SGS-UFSAR Revision 28 May 22, 2015

Channel No.

1-R1B(1) 1-R11A{1) 1-R12A(1) 1-R12B(1) 1R13A 1R13B 1-R15 1-R17A SGS-UFSAR Type of Detector Beta Scintillation Scintillation GMTube Scintillation Scintillation Scintillation Scintillation (In-Line)

Scintillation Channel Description Control Room Intake Duct Containment or Vent Air Containment or Vent Gas Containment or Vent Iodine Nos.11,12,

& 13ContFanCoiiUnit Cooling Water Nos. 13, 14 & 15 Cont Fan Coil Unit Cooling Water Condenser Air Ejector Component Cooling-Liquid TABLE 11.4-1 UNIT 1 PROCESS RADIATION MONITORING SYSTEM Minimum Detectable 10-6 J.tCi/cc 133xe 10-9 J.tCilcc 137 Cs Particulate (moving filter) 10-6 flCilcc 133xe Effluent 10-11 J.tCilcc 1311(2)

Effluent (fixed filter) 10-7 J.tCilcc 137 Cs 10-7 J.tCilcc 137 Cs 10-6 J.tCilcc 133xe 1 of4 Control Functionllnter1ocks Isolates the Control Room Envelope from outside air and places the ventilation system in accident pressurized mode Containment Ventilation Isolation

{Mode6)

Containment Ventilation Isolation Containment Ventilation Isolation Surge Tank Vent Valve Closure Revision 25 October 26, 2010 I

I

Type of Chamel ctameiNo.

Detector Description 1--R17B Scintillation Component Coolirq.Uquid 1--R18(1)

Scintillation liquid Waste Disposal Closure 1-R19A Scintillation No. 11 Steam Generator BlcNvdown 1-R19B Scint8lation No. 12 Steam Genemtor BIDwdcMn 1-R19C ScintiRafion No. 13 Stecm Genelaor BkMdown

-R19D Scinb"Dation No. 14 Stecm Generator BkJM:iown 1-R31A santilalion letdown line (Gross)

SGS-UFSAR TABLE 11.4-1 (Cont.)

Mininum Detedabie level 10 ~ pCilcc 137 Cs 10-5 p(llcc 137 Cs 10 ~ pCilcc 137 Cs 10-5 pCilcc 137 Cs 10 ~ pCi/cc 137 Cs 10 ~ pCilcc 137 Cs 10-4 pCilcc 60 Co 2of4 Control FunctionllnteJb::ks Sulge Tank Vent V;fie Closure Uqukt Waste Discharge V*e High No.12 BlaMbvn Tank Inlet Vfirlr!s High: SG BlowdcNm Isolation vaves High No.12 BlcJIMiMn Tri: Inlet Vftles Hi,jl: SG BlaMiown Isolation Valves High No.12 BII:MI:brm Tri: Inlet Valves High: SG Blowdown Isolation Valves High No.12 Bk7MJown Tank Inlet Valves High: SG BkMdown lsofalion V*e.s RevisiJn23 October 17, 2007 I

I

Channel No.

1-R36 1-R41A 1-R41B 1-R41C 1-R41D SGS-UFSAR TABLE 11.4-1 (Cont.)

Type of Channel Minimum Detectable Detector Description Level Control Function/Interlocks Gamma Evaporator and Feed Heater 1 0-5 1-1Ci/cc 137 Cs Condensate Line Valve Scintillator Condensate Beta Scintillator Plant Vent Noble Gas (Low)

-7

. 133 3x1 0 1-1Ci/cc Xe

Beta-Gamma Scintillator Plant Vent Noble Gas (Inter.)

-4

. 133 7x1 0 1-1Ci/cc Xe Beta-Gamma Scintillator Plant Vent Noble Gas (High) 10-1 1-1Ci/cc 13\\e N/A Plant Vent Noble N/A Containment Ventilation Isolation Gas (composite)

Closes Waste Gas Discharge Valve

This MDL reflects the design range of the detector. The actual detection level may be higher than the MDL due to the masking effect of background radiation at the installed location.

3 of4 Revision 28 May 22, 2015

TABLE 11.4-1 (Cont.)

Type of Channel Minimum Detectable Channel No.

Detector Description Level Control Function/Interlocks 1-R46A Ion Chamber Main Steam Line No. 11 0.1 mr/hr to 10,000 mr/hr(4) 1-R46B Ion Chamber Main Steam Line No. 12 0.1 mr/hr to 10,000 mr/hr(4) 1-R46C Ion Chamber Main Steam Line No. 14 0.1 mr/hr to 10,000 mr/hr(4) 1-R46D Ion Chamber Main Steam Line No. 13 0.1 mr/hr to 10,000 mr/hr(4) 1R51 Beta Scintillator Technical Support Center 10 cpm (nominal) (5) 1-R53A Nal(T1) Gamma Scint.

Main Steam Line No. 11 10 cpm (nominal) (5) 1-R53B Nal(T1) Gamma Scint.

Main Steam Line No. 12 10 cpm (nominal) (5) 1-R53C Nal(T1) Gamma Scint.

Main Steam Line No. 14 10 cpm (nominal) (5) 1-R53D Nal(T1) Gamma Scint.

Main Steam Line No. 13 10 cpm (nominal) (5)

NOTES:

(1) Also performs a safety function (2) Assumes 1-week collection time (3) The upper range corresponds to at least 105 Ci/cc (Xe-133)

(4) The upper range corresponds to at least 103 Ci/cc (Xe-133)

(5) > Background (electrical noise) 4 of 4 SGS-UFSAR Revision 30 May 11, 2018

Type of Channel Channel No.

Detector Description 2-R18 Beta Scintillator Control Room Vent Intake Duct (2 channels) 2-R11A Beta Scintillator Containment Particulate 2-R12A Beta Scintillator Containment Noble Gas 2-R128 Gamma Scintillator Containment Iodine 2-R13A Gamma Scintillator 21,22,23 Fan Cooler Service Water Discharge 2-R138 Gamma Scintillator 23,24,25 Fan Cooler Service Water Discharge 2-R15 Gamma Scintillator Condenser Air Ejector 2-R17A Gamma Scintillator 21 Component Cooling Loop 2-R178 Gamma Scintillator 22 Component Cooling Loop 2-R18(1)

Gamma Scintillator Liquid Waste Discharge SGS-UFSAR TABLE 11.4-2 UNIT 2 PROCESS RADIATION MONITORING SYSTEM 1 of4 Minimum Detectable Level 10-6 flCilcc 133xe 10-11 J.tCVcc 137 Cs 10-6 J.tCilcc 133xe 10-11 JJ.Cilcc 1311(2) 10-7 J.tCi/cc 137 Cs 10-7 JJ.Cilcc 137 Cs 10-6 J.tCilcc 133xe 10-7 J.tCilcc 137 Cs 10-7 JJ.Cilcc 137 Cs 10-7 JJ.Cilcc 137 Cs Control Functionnnterlocks Isolates the Control Room Envelope from outside air and places the ventilation system in accident pressurized mode Containment Ventilation Isolation (Mode6)

Containment Ventilation Isolation Containment Ventilation Isolation Closes Surge Tank Vent Valve Closes Surge Tank Vent Valve Closes Liquid Waste Discharge Valve Revision 25 October 26, 2010

TABLE 11.4-2 (Cont.)

UNIT 2 PROCESS RADIATION MONITORING SYSTEM Type of Channel Minimum Detectable Channel No.

Detector Description Level Control Function/Interlocks 2-R19A Gamma Scintillator 21 Steam Generator 3.91x10-8 Ci/cc137Cs Warn: Closed Blowdown Tank Inlet Valves Blowdown High: Isolate 21 Steam Generator Blowdown 2-R19B Gamma Scintillator 22 Steam Generator 3.91x10-8 Ci/cc137Cs Warn: Closed Blowdown Tank Inlet Valves Blowdown High: Isolate 22 Steam Generator Blowdown 2-R19C Gamma Scintillator 23 Steam Generator 3.91x10-8 Ci/cc137Cs Warn: Closed Blowdown Tank Inlet Valves Blowdown High: Isolate 23 Steam Generator Blowdown 2-R19D Gamma Scintillator 24 Steam Generator 3.91x10-8 Ci/cc137Cs Warn: Closed Blowdown Tank Inlet Valves Blowdown High: Isolate 24 Steam Generator Blowdown 2-R31(1)

Gamma Scintillator Letdown Line 10-6 Ci/cc137Cs 2-R37 Gamma Scintillator Non Radwaste Basin 10-8 Ci/cc137Cs None 2-R41A Beta Scintillator Plant Vent Noble Gas 3x10-7 Ci/cc133Xe(2)

None (low) 2-R41B Beta-Gamma Plant Vent Noble Gas 7x10-4 Ci/cc133Xe(2)

None Scintillator (inter.)

2-R41C Beta-Gamma Plant Vent Noble Gas 10-1 Ci/cc133Xe None Scintillator (high) 2-R41D N/A Plant Vent Noble Gas N/A Containment Ventilation Isolation; (composite)

Closes Waste Gas Discharge Valve

This MDL reflects the design range of the detector. The actual detection level may be higher than the MDL due to the masking effect of background radiation at the installed location.

2 of 4 SGS-UFSAR Revision 30 May 11, 2018

Channel No.

2-R46A 2-R46B 2-R46C 2-R46D SGS-UFSAR Type of Detector ion Chamber ion Chamber lon Chamber ion Chamber Channel Description Main Steam Line No. 21 Main Steam Line No. 22 Main Steam Line No. 24 Main Steam Line No. 23 TABLE 11.4-2 (Cont.)

UNIT 2 PROCESS RADIATION MONITORING SYSTEM 3 of4 Minimum Detectable Level 0.1 mr/hr to 10,000 mr/hr(4) 0.1 mr/hr to 10,000 mr/hr(4) 0.1 mr/hr to 10,000 mr/hr(4) 0.1 mr/hr to 10,000 mr/hr(4)

Control Function/Interlocks Revision 28 May 22, 2015

Type of Channel No.

Detector 2-R53A Nai(T1) Gamma Scint.

2-R53B Nai(T1) Gamma Scint.

2-R53C Nai(T1} Gamma Scint 2-R53D Nai(T1) Gamma Scint NOTES:

( 1 )

Also perfonns a safety function (2) Assumes 1 week collection time (3) The upper range corresponds to 105 f.!Cii~X {Xe-133)

(4) The upper range coo-esponds to at least 103 llCilcc (Xe-133)

(5) > Background (electrical noise)

SGS-UFSAR TABLE 11.4--2 (Cont.)

UNIT 2 PROCESS RADIATION MONITORING SYSTEM Channel Description Main Steam Line 21 Main Steam Line 22 Main Steam Une 23 Main Steam Une 24 4of4 Minimum Detectable Level 10 cpm (nominal}

(5) 10 cpm (nominal)

(5}

10 cpm (nominal}

(5) 10 cpm (nominal}

(5)

Control F unctionllntel1ocks Revision 19 November 19, 2001

1-R1A 1-R2 R3 1-R4 1-R5(1)

R6A 1-R7 1-R9(1) 1-R10A 1-R10B R20B SGS-UFSAR TABLE 11.4-3 UNIT 1 AREA RADIATION MONITORING SYSTEM Channel Control Room Containment Radio-Chem Laboratory Charging Pump Room Fuel Handling Sampling Room In-core Seal Table GM Tube GM Tube GM Tube GM Tube GM Tube GM Tube GM Tube Fuel Storage Area GM Tube Personnel Hatch to Containment (el 100 ft)

Personnel Hatch to Containment (el 100 ft)

Counting Room GM Tube GM Tube GM Tube 1

mR/hr 10-1 -10 4 mR/hr 10-1 -10 4 mR/hr 10-1 -10 4 mR/hr 10-1 -10 4 mR/hr

-10 4 mR/hr 10-1 -10 4 mR/hr

-10 4 mR/hr 1

-10 4 mR/hr 1 of 2

1)

Fuel Handling Area Hi-Rad Evacuation Alarm

2)

Fuel Handling Area Ventilation 1

Exhaust Filter Units

1)

Fuel Handling Area Hi-Rad Evacuation Alarm 2}

Fuel Area Ventilation Exhaust Filter Units Containment Area High Radiation Alarm Containment Area High Radiation Alarm Actuates High Radiation Signal at Hatch Revision 25 October 26, 2010 I

TABLE 11.4-3 (Cont.)

Channel Type of Channel No.

Description Detector Range Control Function/Interlocks R23 Monitoring Room GM Tube 10-1 -104 mR/hr 1-R32A (2)

Fuel Handling GM Tube 10-1 -104 mR/hr Crane Monitor 1-R34 Mechanical GM Tube 10-1 -104 mR/hr Penetration Area 1-R44A Containment 100 - 107 R/hr (High Range) 1-R44B Containment 100 - 107 R/hr (High Range)

NOTES:

(1) Also performs a safety function (2) Local monitor only. Not indicated, recorded, or alarmed in the control room.

2 of 2 SGS-UFSAR Revision 31 December 5, 2019

(

Chann~l rfo.

2-R1A 2-R2 2-R4 2-RS(l) 2-R7 2-R9(1) 2-RlOA 2-RlOB 2-R32A(2)

SGS-UFSAR Channel Description Control Room Containment Charging Pump Room l'uel Handling Building Incore Seal Table Fuel Handling Building Containment Personnel Hatch (el 100 ft) containment Personnel Hatch (el 130 ft)

Fuel Handling Crane

(

TABLE 11.4-4 UNIT 2 AREA RADIATION MONITORING SYSTEM Type of Detector GM Tube GM Tube GM Tube GM Tube GM Tube GM Tube GM Tube GM Tube GM Tube Range 10-l - 104 mR/hr 10 104 mR/hr 10 104 mR/hr 10 104 mR/hr 10 104 mR/hr 10 104 mR/hr 1 of 2 Control Functions/Interlocks Actuates High Radiation Sign at Hatches (el 100 ft and el 130 ft)

Initiates Charcoal Filtration and Evacuation Horns in Fuel Handling Building and automatically starts exhaust fans.

Actuates High Radiation Sign at Hatch Actuates High Radiation Sign at Hatch Revision 16 January 31, 1998

TABLE 11.4-4 (Cont.)

Channel Type of Channel No. Description Detector Range Control Functions/Interlocks 2-R34 Mechanical GM Tube 10-1 -106 mR/hr Penetration Area 2-R44A Containment Ion Chamber 100 -107 R/hr (High Range) 2-R44B Containment Ion Chamber 100 -107 R/hr (High Range)

NOTES:

(1) Also performs a safety function.

(2) Local only - Not connected to RMS monitor in the Control Equipment Room.

2 of 2 SGS-UFSAR Revision 31 December 5, 2019

FIELD

-CONTROL FUNCTIONS CONTROL ROOM AREA PLANT COMPUTER 1---------1 L DC AL 1----+------1 REM 0 T E DETECTOR MONITOR MONITOR TYPICAL RADIATION MONITORING STRUCTURE ANNUNCIATOR PSEG NUCLEAR LLC SALEM GENERATING STAT ION UNIT NOa 2 OVERALL RADIATION MONITORING SYSTEM MAKEUP Updated FSAR Sheet 1 of 1 REVISION 27, NOVEMBER 25,2013 F 1 g a 11 a 4 -1

AIR PARTICULATE 8c GAS MONITOR FROM CONTAINMENT I

ALARM ANNUNCIATOR ISOLATION/ALARM SIGNALS -------------.

~~

MOVING FILTER PARTICULATE ASSEMBLY SAMPLE SKID CONTROLS FIXED FIL TERt--------t IODINE AND NOBLE GAS DETECTOR ASSEMBLY TO CONTAINMENT LOCAL MICROPROCESSOR AND DISPLAY PSEG Nuclear.

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CONTROL ROOM ALARM LAMP Revision 19, Nov.1q, 2001 Solem Nuclear Generotin~ Station UNIT NO.1 AIR PARTICULA E IODINE AND GAS MONITOR (TYPICAL)

Air

{

Sample Containment

~---~---

or Plant Vent Particulate Filter Mech.

Iodine Filter Sampler Gas Sampler I

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I I

I t

l I

J l

J 1

I *--

l I /

-0 Alarm

/

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/

' -lwl Fail

/

-~ Checksource

/

I '

X.XXXE:tyy 2-R11Af 12A, B Control Room Display Control Functions (Tables 11.2-7)

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Unit No. 2 Air Particulate, Iodine and Gas Monitor

{Typical)

Updated FSAR Revision 16 Figure 11.4-3, Sheet 1 of 1 January 31, 1998

1 I

I I

I EFFLUENT IN I

I I

I DETECTOR I

I I

SAMPLER I

I I

EFFLUENT OUT LIQUID PROCESS MONITOR

CONTROL ROOM ANNUNCIATOR CHECK SOURCE HIGH RADIATION VISUAL ALARM RELAYS CONTROL EQUIP. ROOM READ OUT MODULE RECORDER

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<METER>

WARNING ALARM CHANNEL FAILURE ALARM I

I I

~:s~on-2~ ~~-:,. -2~ 2"010" PSEG Nuclear, LLC Solem Nuclear Generating Station UNIT NO.1 LIQUID MONITOR <TYPICAL>

SALEM NUCLEAR GENERATING STATION ted fSAR SHEET 1 Of 4 figure 11.4-4 2000 PSEG Nttle<r. LtC. All Ri4Jts Reserved.

HIGH _RAD AUDIBLE ALARM DIGITAL LIQUID -PROCESS MONITOR (TYPICAL>

HIGH RAD VISUAL ALARM 0ALERT HIGH RAD VISUAL

ALARM.

IN CONT.ROOM COMPUTER CONTROL ROOM ANNUNCIATOR CONTROL INTERLOCKS ROOM (IF SHOWN INDICATION IN THE

<METER)

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0000 HIGH ALERT NORMAL FAIL

~------~~

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Updated FSAR.

SHEET 2 Figure 11.4-4

DIGITAL LIQUID PROCESS MONITOR

<TYPICAL)

HIGH RAD AUDIBLE ALARM HIGH RAD VISUAL ALARM

~ALERT I'DISPLAYI ~FAIL

+

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0 *~ 0 ~

HIGH ALERT NORMAL FAIL

~------~~

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Rev1s1on 21, Dec. 6, 2~~4 Solem Nuclear Generating Station UNfT NO.1 COND. AIR EJECT RAD MONITOR 1R15 (TY Updated FSAR SHEET 3 Figure 11.4-4 Reserved.

DJGiT~. LlQUlD PROCESS MONlTOR HlOH RAO AUDIBLE AL~RM HIGH RAD VISUAL ALARM

~ALERT

[ DlSPC§yJ

~FAlL

~NORMAl.

l.OCAL MONit: OR HlOH RAD AUDIBLE At..AAM 1R13A

HIGH RAD VISUAL ALARM

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ROOM ANNUNCIATOR REMOTE MONlTOR CONTROL EQUt?MEtJT ROOM PSEG NUCLEAR LLC SALEM GENERATING STAT ION UNIT NOa 1 LIQUID MONITOR 1R13 A & B Updated FSAR Sheet 4 REVISION 27, NOVEMBER 25,2013 F 19a lla4 -4

DIGITAL LIQUID -PROCESS MONITOR tTYPICALl HIGH.. RAD HIGH. RAD AUDIBLE **

VISUAL ALARM ALARM

~-ALERT COMPUTER*

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Updated FSAR

,Sheet 1 of 2 FIQ ure 11.4*5

HIGH RAD AUDIBLE ALAR~t HIGH RAD VISUAL ALARM DJGIT AL LIQUID PROCESS MONiTOR 2R13A & B mALEAT 16isPCa~l WFAIL mNORMAL LOCAL MONITOR HIGH RAO AUDIBLt:

ALARM 1cW PIPE 12 HIGH RAO VISUAL ALARM

~ALERT

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~NORMAL LOCAL MONITOR 1CW PIPE 11 COMPUTER CONTROL ROOM ANNUNCIATOR mw:m:~

...__---=i HIGH ALERT NORMAL FAlL 1 oiseceYJ REMOTE ~tONJTOR CONTROL EQUIPMENT ROO~

PSEG NUCLEAR LLC SALEM GENERATING STATION UNIT NOa 2 LIQUID MONITOR 2R13A & B Updated FSAR REVISION 27, NOVEMBER 25,2013 Sheet 2 F l9a lla4-5

THIS FIGURE HAS BEEN DELETED Revtslon 21,.Dec. 6, 2004 PSEG Nuclear, LLC.

Salem Nuclear. Generating

. Station UNIT 2 IN-LINE GAS MODEL <TYPICAL>

SALEM NUCLEAR GENERATING STATION doted FSAR Figure 11.4-6

CHECK SOURCE HIGH RADIATION AUDIBLE ALARM DETECTOR LOCAL READOUT HIGH RADIATION VISUAL I ALARM PSEG Nuclear, LLC AREA MONITOR

CONTROL ROOM ANNUNCIATOR HIGH RADIATION VISUAL ALARM CONTROL EQUIP. ROOM READ OUT MODULE RECORDER (CONTROL EQUIP.

RM.)

INTERLOCKS CIF SHOWN IN THE TABLE)

CONTROL ROOM READ OUT

<METER>

WARNING ALARM Revtston }q, Nov.lq, 2001 Solem Nuclear Generating Station UNJT NO.1 AREA MONITOR (TYPICAL)

SALEM NUCLEAR GENERATING STATION Updated FSAR SHEET 1 OF 2 Figure 11.4-7

~II Rights Reservoo.

DIGITAL AREA MONITOR CTYPICAL)

HIGH RAD AUDIBLE ALARM HIGH RAD VISUAL ALARM w:ALERT HIGH RAD VISUAL ALARM IN CONT.ROOM COMPUTER CONTROL ROOM ANNUNCIATOR CONTROL INTERLOCKS ROOM CIF SHOWN INDICATION IN THE CMETER>

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w:w:mw:

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w:FAIL

~------+-~

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1R2,1R3,1R4,1R5,1R6A,1R7,1R9,1R10A PSEG Nuclear, LLC SALEM NUCLEAR GENERATING STATION I DISPLAYI REMOTE MONITOR CONTROL EQUIPMENT ROOM Rev1s1on 19, Nov.l9, 2001 Salem Nuclear Generating Station UNIT NO.1 AREA MONITOR (TYPICAL>

Updated FSAR SHEET 2 OF 2 Figure 11.4-7 CD 2000 PSE G Nuc l ecr. ll C. A II Rights Reserved.

c e

t e c

t c

r

[

TO REMOTE UNIT

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(AS SHOWN IN TABLE 11-2

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PSEG NUCLEAR LLC SALEM GENERATING STAT ION UNIT NOa 2 AREA MONITOR (TYPICAL)

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UNIT 1 ClRPl>

CONTROL ROOM AREA RATE METER UNITt (2>

CONTROL EQUIPMENT ROOM CPM CPM CHANNEL 1

/

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/ '

/

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/

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/ '

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l.CHANNEL 1 INTERLOCKS WITH RESPECTIVE UNITS CONTROL CIRCUITS.CHANNEL 2 INTERLOCKS WITH OPPOSITE UNITS CONTROL CIRCUITS.

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SUPPLY 1 MICROPROCESSOR ASSEMBLY CONTROL FUNCTIONS I INTERLOCKS

[AS SHOWN IN TABLE 11.4.1(2)]

CONTROL ROOM AREA SEE NOTE 1 VENTILATION ENTRY CONTROL ROOM AREA VENTILATIO~ ENTRY DESIGNATED UNIT CHANNEL 1 DETECTOR 1 RlB SHOWN 2 RlB TYPICAL PSEG Nuclear, LLC OPPOSITE UNIT CHANNEL 2 DETECTOR Rev1s1on 21, Dec. 6, 2004 Salem -Nuclear Generating Station CONTROL ROO~ AREA* INT f.J<E DUCT RADIATION MONITOR SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 11.4-9

11.5 SOLID RADWASTE SYSTEM The Solid Radwaste System collects, processes,

, and provides temporary storage for radioactive solid wastes due for offsi te shipment and permanent 11.5.1 Design Objectives

1.

To a means of demineralizer resins during plant

2.

To provide a means of spent resins and expended filters in containers suitable for transfer from the plant site.

3.

To provide a means of packaging low level contaminated solid wastes such as glassware and clothing.

4.

The To provide a means of shipment.

spent resins and packaging for offsite criteria for the Solid Radwaste Treatment System are as follows:

The control over design shall include the plant radioactive those means necessary to maintain solid waste.

Appropriate storage shall be provided for retention of solid wastes.

In all *cases, the design for radioactivity control shall be justified (a) on the basis of 10CFR20 requirements, for both normal operations and for any transient situation that might be anticipated to occur and (b) on the basis of 10CFR50.67 level limits for potential reactor accidents of exceedingly low of occurrence.

11.5-1 SGS-UFSAR Revision 26 May 21, 2012

The solid waste facility is designed so that offsite are in accordance with applicable governmental The spent resins from the demineralizers and filter cartridges are packaged and stored onsite until shipment offsite for disposal.

11.5.2

System Design

The solid processing portion of the Waste Disposal System is designed to package all solid wastes for removal to volume reduction or burial facilities.

All packaging shall meet DOT/NRC as applicable upon contents.

for burial will conform to burial site criteria.

The resins are transferred to appropriate shipping containers for processing as necessary prior to Active Wastes (DAW) are shipped to an off-site volume ~eduction facility.

11.5.3 Equipment The Solid Radwaste System consists of components and subsystems described below.

11.5-2 SGS-UFSAR Revision 26 May 21, 2012

The resin processing equipment is operated remotely from a control station that a fill head on the appropriate shipping container.

The processing is done within the Auxiliary Building to control the release of air and liquid to the environment.

Activity levels of the contents are monitored to limit the doses during Off Site Volume Reduction DAW is collected in Sea-vans and then shipped off-site for volume reduction processing by a licensed contractor.

The volume-reduced DAW is placed into packages that will meet DOT/NRC approval and then shipped for disposal at a licensed burial site.

11.5.4 Expected Volumes See Section 11.5.7 11.5.5 Packaging Packaging is done in DOT/NRC approved packages, as appropriate, depending upon contents.

11.5.6 Storage Facilities The storage areas are shielded to protect in accessible portions of the solid radwaste area. The is designed.to meet the of 10CFR20.

Low-level radwaste can be stored in the Low-Level Radwaste (LLRSF) for up to 5 years if to a Low-Level Radwaste is denied.

The LLRSF has been in accordance with guidelines provided in Generic Letter 11.5-3 SGS-UFSAR Revision 26 May 21, 2012 I

11.5.7 Shipment The average annual volumes of solid wastes (on a two unit basis) shipped from the Salem Generating Station are as follows:

Spent resins, filter sludges Dry compressible waste, contaminated equipment 25 cubic meters 100 cubic meters The Process Control Program (PCP) has been approved by the Nuclear Regulatory Commission (NRC) which outlines the in-plant measures and controls to assure the suitability of radioactive waste products for and/or at a licensed burial site.

11.5.8 Reference for Section 11.5

1. Public Service Electric and Gas Letter, Liden to Varga, November 28, 1983 11.5-4 SGS-UFSAR Revision 26 May 21, 2012

11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM Since 1968 an Offsite Radiological Environmental Monitoring Program (REMP) has been conducted on the Artificial Island Site.

Starting in December 1972, more extensive radiological monitoring programs were initiated with the purpose of identifying and quantifying the concentration of various radioactive elements in the different environmental media surrounding the Salem Generating Stations.

11.6.1 Program Objective The objectives of the operational radiological environmental program are:

1.

To fulfill the obligations of the radiological Environmental Monitoring sections of the Technical Specifications for the Salem Generating Station

2.

To determine whether any significant increase occurs in the concentration of radionuclides in critical pathways

3.

To determine if the station has caused an increase in the radioactive inventory of long lived radionuclides

4.

To detect any change in ambient gamma radiation levels

5.

To verify that the Salem Generating Station operations have no detrimental effects on the health and safety of the public or on the environment The basis for the selection of sample locations was determined from site meteorological, hydrological, and geographical constraints enclosed by local demography and land use.

Sample locations chosen are consistent with the U.S.

Nuclear Regulatory Commission Branch Technical Position on the REMP.

Sample locations were divided into two basic classes, indicator and control stations.

Indicator stations were situated in areas most likely to be affected by a station release. All control stations were presumed to be located at a distance from the station to be unaffected by normal and accident plant operation. In this manner, fluctuations in the level of radionuclides and direct radiation at an indicator station could be compared to those, if any, at a control location to determine if the station was the cause of the fluctuations.

11.6-1 SGS-UFSAR Revision 16 January 31, 1998

11.6.2 Preoperational and Operational Programs The operational REMP for the Artificial Island Site was initiated in December 1976 after Unit 1 achieved criticality.

The current operational program is described in the Offsite Dose Calculation Manual (ODCM).

Table 11.6-2 provides the results obtained during the preoperational phase of the program.

11.6.3 Expected Pathways The scope of the REKP is to determine whether any significant increases have occurred in the concentration of radioactive materials in the critical pathways as well as to determine if there is an increase in the radioactive inventory of long-lived radionuclei. During the months when cows are in pasture, the critical pathway has been determined to be the air, grass, milk, to man pathway.

During those months when cows are not on pasture the air, broadleaf vegetation, to man pathways predominates. The selection of sample type was based upon these established critical pathways of the transfer of radionuclei, from the environment to man as well as experience of the preoperational phase of the program.

11.6.4 Physical Characteristics of Samples The physical characteristics of the samples are summarized in the Offsite Dose Calculation Manual (ODCM).

The sensitivity of the analytical procedures used are consistent with the Branch Technical Position on the procedures for Environmental surveillance programs and are provided in the Radiological Monitoring sections of the Technical Specifications. The sensitivity of the analytical measurements was selected to verify that the measurable concentrations of radioactive materials are not higher than expected based on effluent monitoring and modeling.

The sensitivity requirements listed in the Radiological Monitoring sections of the Technical Specifications are sufficiently low to discern dose commitments which are a fraction of 10CFRSO, Appendix I for there predicted critical pathways.

It should be noted that deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or the malfunction of automatic sampling equipnent. All deviations from the sampling schedule are documented in the annual report.

Descriptions of the analytical methods utilized are provided in the Annual Radiological Reports.

11.6-2 SGS-UFSAR Revision 16 January 31, 1998

TABLE 11.6-1 This Table Deleted 1 of 1 SGS-UFSAR Revision 15 June 12, 1996

Med_;_um and Anc.lysis Performed AQUATIC ENVIRONMENT Surface Water H-3 Alpha (soluble)

Alpha (insoluble}

Alpha (total)

Beta (soluble)

Beta (insoluble}

Beta (total)

K-40 Sr-89 Sr:-90 Gamma K-40 ZrNb-95 cs-137 Ra-226 Solids Chlorides Edible Fish Flesh H-3 {aqueous)

H-3 (organic)

H-3 /organic)

Sr-89 Sr-90 SGS-UFSAR TABLE 11.6-2

SUMMARY

OF RADIONUCLIDE CONCENTRATIONS IN ARTIFICIAL ISLAND PREOPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLES Number of Number Average samEles Analyzed Above MDL Minimum Maximum

+/- 2 Si~

259 222

<80 600+100 213+165 136 6

<1.5 33+32 136 5

<1.5 3.3+1. 0 123 22

<1.5 27+20 136 136 3.8+/-2.5 120+/-16 42+/-52 136 14

<3.0 4.8+/-3.2 123 123 3.3+/-2.7 110+/-11 32+/-45 136 136 0.20+/-0.02 120+/-12 41+/-63 65 5

<0.37 1.5+/-0.8 77 20

<0.28 1.6+/-0.4 259 200

<6 200+/-30 48+/-64 2

<0.4 1.9+/-0.6 2

<0.5 0.86+/-0.56 9

<0.9

4. 0+/-1. 4 136 136 160+/-16 14000+/-1400 5748+/-7137 136 136 32+/-3 17000+/-1700 3345+/-5567 28 15

<80 460+/-78 165+/-206 19 17

<130 480+/-69 285+/-205 8

4

<80 390+/-80 158+/-230 9

0

<4.1

<100 12 1

<4.1 67+/-11 1 of 9 Units pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/l pCi/1 pCi/1 pCi/1 mg/1 mg/1 pCi/1 pCi/1 pCi/kg(dry) pCi/kg(wet}

pCi/kg{wet)

Revision 6 February 15, 1987

Medium and Number of An""lysis Performed SamEles Analyzed Edlble Fish Flesh (:ontl Gam'lla 31 K-40 cs-137 Ra-226 Edible Fish Bone Sr-89 18 Sr-90 18 Blue Crab -

Edible Hard Shell (Flesh}

H-3 (Aqueous) 8 H-3 (Organic) 2 H-3 (Organic) 4 H-3 (Total Tr-itium) 7 Sr-89 10 Sr-90 10 Gamma

[(-40 16 ZrNb-95 Ra-226 Blue Crab -

Edible Soft Shell (Total}

H-3 (Aqueous) 4 H-3 (Organic) 2 H-3 (Organic) 2 H-3 (Total Tritium) 4 Sr-89 6

sr-90 8

Gamma K-40 7

?.a-226 TABLE 11. 6-2 Number Above MDL Minimum 31 1000+/-100 5

8.7+/-6.6 1

<4 3

100+/-60 14

<24 7

<80 2

95+/-78 1

<80 5

<80 0

<6.0 7

<5.1 16 960+/-384 1

<5 3

<10 3

<80 0

<140 1

<80 3

<80 1

<7.5 5

<7.8 7

770+/-462 1

<10 2 of 9 (Contl Maximum 13000+/-1000 11+/-6 130+/-80 100+/-70 940+/-100 330+/-71 420+/-120 90+/-80 420+/-60

<51 150+/-26 12000+/-1000 120+/-12 33+/-19 320+/-110

<220 130+/-80 500+/-68

<26 39+/-6 3000+/-300 37+/-15 Average

+ "

4-Sigma 2914+/-4351 335+/-614 180+/-168 259+/-455 219+/-259 40+/-103 2835+/-5048 190+/-197 105+/-71 230+/-373 21+/-24 2040+/-1359 Units pCi/kg(wet) pCi/kg(wet) pCi/kg(wet) pCi/kg(wet dry) pCi/kg(wet dry) pCi/1 pCi/1 pCi/kg{dry}

pCi/1 pCi/kg(wetJ pCi/kg(wet) pCi/kg(wet dryl pCi/kg (wet) pCi/kg(wet}

pCi/1 pCi/1 pCi/kg(dry) pCi/1 pCi/kg(wet) pCi/kg{wet}

pCi/kg{wet) pCi/kg(wetJ Revision 6 February 15, 1987

TABLE 11.6-2 Medium and Number of Nmnber Analysis Performed SamEles Analvzed Above MDL Minimum Blue Crab - Nonedible Hard Shell (Shell)

Sr-89 B

3

<57 SR-90 8

8 330+/-30 Prey Fish Sr-89 16 2

<5.8 Sr-90 16 8

<4.8 Gamma 29 K-40 29 970+/-100 ZrNb-95 2

<3 Cs-137 1

<2 Ra-226 1

<5 Benthos Sr-89 12 1

<44 Sr-90 12 5

<120 Ganuna K-40 4

2

<1 Mn-54 1

<0.03 Nb-95 2

<0.09 Ru-106 1

<0.3 Cs-137 1

<0.07 Ra-226 4

0.26+/-0.08 Th-232 4

0.52+/-0.13 Zooplankton Sr-89 8

2

<0.21 Sr-90 8

5

<0.51 3 of 9 SGS-UFSAR (Cont}

Maximum 210+/-170 990+/-87 320+/-22 66+/-11 13000+/-1000 17+/-5 42+/-38 13+/-9

<36000

<30000

6. 9+/-1. 8 0.13+/-0.09 0.11+/-0. 08 0.91+/-0.65

<0.1 0.48+/-0.16 1.2+/-0.4 4.6+/-4.4

<4.9 Average

+/- 2 Si9!!.!a 614+/-511 28+/-40 4604+/-6667 3.4+/-5.0 0.11+/-0.13 0.38+/-0.18 0.80+/-0.61

1. 3+/-2. 9 Units pCi/kg(dry) pCi/kg(dry}

pCi/kg{wet dry) pCi/kg(wet dry) pCi/kg(wet dry}

pCi/kg{wet) pCi/kg(wet) pCi/kgCwet) pCi/kg(dry) pCi/kg(dry) pCi/g(dry}

pCi/g{dry) pCi/g(dry) pCi/g(dry) pCi/g{dry) pCi/g(dry) pCi/g (dry) pCi/g (dry) pCi/g{dryl Revision 6 February 15, 1987

Medium and Number of Analysis Performed SamEles Analyzed Zooplankton (Cent)

Gamma 20 Be-7 K-40 ZrNb-95 Ra-226 Sediment Sr-89 12 Sr-90 16 Gamma 41 Be-7 K-40 Nb-95 Zr-95 Ru-103 Ru-106 Sb-125

.:s-137

e-1<';4 Ra-2~ 6 Th-232 ATMOSPHERIC ENVIRONMENT Air Particulates Alpha 1045 Beta 1088 Sr-89 60 Sr-90 60 SGS-UFSAR TABLE 11. 6-2 Nwn.ber Above MDL Minimum 1

3

<3 2

<0.09 1

<0.2 0

<0.03 4

<0.03 3

<0.1 41 3.4+/-0.4 8

<0.01 1

<0.02 1

<0.01 2

0.03+/-0.02 8

0.05+/-0.04 35

<0.01 2

<0.1 41 0.28+/-0.04 41 0.21+/-0.11 788

<0.16 1585 5.0 18

<0.15 46

<0.20 4 of 9 (Cent)

Maximum 3.0+/-2.8 llO+/-BO

<10

<30

<1.0 0.32+/-0.05 2.3+/-0.3 21+/-2 2.6+/-0.6 0.70+/-0.30 0.31+/-0.13 0.30+/-0.11 0.27+/-0.12 0.40+/-0.04 0.48+/-0.14 1.2+/-0.1 1.3+/-0.1 7.9+/-3.1 920+/-24 4.7+/-0.5 3.0+/-0.3 Average

+/- 2 Sigma 15+/-23 0.15+/-0.22 0.76+/-0.43 0.84+/-0.54 1.1+/-2. 8 74+/-280

0. 9+/-1. 6 Units pCi/gtdryl pCi/g (dry) pCi/G {dry) pCi/g(dry) pCi/g(dry) pCi/g(dryl pCi/g (dryl pCi/g (dryl pCi/g (dryl pCi/g{dryl pCi/g(dryl pCi/g(dry) pCi/g(dryl pCi/g(dryl pCi/g(dryl pCi/g(dryl pCi/g(dryl 1D-3pCi/rrr 10-3pCi/rrr 10-3pci/rrf 10-3pci/ml Revision 6 February 15, 1987

Medium and Number of Analysis Performed SamEles Analyzed Air Particulates {Cant)

Gamma 127 Be-7 Mn-54 Zr-95 Nb-95 Ru-103 Ru-106 Sb-125 Cs-237 BaLa-140 Ce-141 Ce-144 Ru-226 Th-232 Air Iodine I-131 519 Precipi t atior, R-3 63 Alpha 63 Beta 63 K-40 35 Sr-89 16 Sr-90 20 Gamma 22 Be-7 K-40 Kr/Nb-95 Ru-103 Cs-137 BaLa-140 ce-144 SGS-UFSAR TABLE 11.6-2 Number Above MDL Minimum 127 12+/-7 31

<0.1 80

<0.3 74

<0.4 90

<0.1 60

<0.9 55

<0.4 122

<0.2 3

<0.4 38

<0.2 99

<1 26

<0.6 24

<0.5 20

<0. 72 49

<80 6

<0.45 57

<3.0 32

<0.01 7

<0.43 12

<0.48 16 0.75+/-0.22 3

<5 9

<0.4 2

0.48+/-0.29 1

<0.5 1

<0.4 1

<3 5 of 9 (Cont)

Maximum 330+/-33 2.6+/-0.4 44+/-5 27+/-4 84+/-8 46+/-4 6.2+/-0.8 11+/-1 27+/-8 46+/-5 120+/-15 16+/-7 3.1+/-0.18 42+/-4 610+/-90

4. 7+/-1.8 71+/-8 0.52+/-0.05

5. 6+/-1.2

3. 8+/-1.1 79+/-74 18+/-5 9.5+/-1.0 3.4+/-0.5 1.2+/-0.4 2.2+/-0.9 6.2+/-2.2 Average

+/- 2 Sigma 109+/-126 5+/-15 4+/-13 13+/-39 2.2+/-4.7 19+/-54 216+/-290 19+/-36 0.13+/-0.21 1.5+/-2.2 29+/-45 Units 10-" pCi/ni' 10-" pCi/Jn'l 10-3 pCi/Jn'l 10-3 pCi/m>

10-" pCi/m>

10-" pCi/Jn'l 10-" pCi/Jn'l 10-3 pCi/m>

10-3 pCi/m>

10-' pCi/Jn'l 10-3 pCi/Jn'l 10-' pCi/m>

10-' pCi/Jn'l 10-3 pCi/Jn'l pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 Revision 6 February 15, 1987

Medium and Number of Anal:rsis Performed Sarnt::les Analyzed TERRESTRIAL ENVIRONMENT We_l Water l-l-3 144 Alpha 144 Beta 144 K-40 134 sr-es 24 Sr-90 24 Gamma 39 Be-7 K-40 ZrNb-95 Ra-226 Potable Water (Raw and treated>

H-_-i 94 Alpi"'.a 94 Beta 94 K-40 84 Sr-89 24 Sr-90 28 Gamma 32 Ra-226 Milk Sr-89 177 Sr-90 193 I-131 167 SGS-UFSAR TABLE 11.6-2 Number Above MDL Minimum 23

<40 4

<1.0 126

<2.1 134 1.1+/-0.1 1

<0.49 3

<0.36 3

32+/-7 10

<0.5 2

<0.4 1

<0.8 73

<80 16

<0.53 63

<2.6 84 0.50+/-0.10 3

<0.33 10

<0.26 1

<0.8 35

<0.55 169

<0.23 44

<0.03 6 of 9

{Cont)

Maximum 380+/-77 9.6+/-2.5 38::!:6 19+/-2

<2.1 0.87+/-0.52 45+/-9 30+/-14 2.0+/-0.7 2.0+/-1.2 350+/-80

2. 7+/-1. 3 9.0+/-3.4 10+/-1 1.1+/-0.9 2.1+/-1.2 1.4+/-1.0 14+/-2 12+/-3 65::!:6 Average

+/- 2 Sigma 9+/-10 7.8+/-7.6 179+/-173 4.2+/-3.0

l. 7+/-2.2 3.5+/-7.5 Units pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/l pCi/1 pClll pC_;_/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 pCi/1 Revision 6 February 15, 1987

TABLE 11.6-2 Medium and NUI!lber of NUI!lber Anal~sis Performed Sam£les Analyzed Above MDL Minimum Milk (Contl Gamma 270 K-40 270 800+/-100 I-13l 5

<0.4

s-L37 147

<0.4

l.a-2::6 9

<0.9 Food ?roducts sr-89 29 3

<2.5 sr-90 45 24

<1.5 Gamma 52 K-40 52 70+/-50 Cs-137 7

<1 Fodder crops (dry)

Gamma 41 Be-7 21 0.80+/-30 K-40 41 2.9+/-0.5 Co-58 3

<0.02 ZrNb-95 18

<0.02 Zr-95 3

4.3+/-0.4 Nb-135 3

3.0+/-0.3 Ru-103 4

0.09+/-0.02 RuRj-206 14

<0.01 I-131 3

<0.01 I-133 3

<0.02 cs-137 6

<0.01 BaLa-140 3

<0.02

Additional egg plant sampled 10-13-76 reported in units of pCi/kg(dry}.

Results were: Sr-89, <55 pCi/kg(dry); Sr-90, 32+/-24 pCi/kg(dry).

{These data not included in ranges, or average).

7 of 9 SGS-UFSAR

{Cont)

Maximum 2000+/-200 63+/-14 14+/-11

<30 10+/-6*

24+/-6*

4800+/-480 59+/-33 9.3+/-0.9 80+/-8 0.07+/-0.05 9.5+/-4.8 6.3+/-0.6 4.2+/-0.4

1. 3+/-0.1 2.5+/-0.2 0.53+/-0.06 0.46+/-0.11 0.11+/-0.02 3.1+/-0.3 Average

+/- 2 Sigma 1437+/-440 3.0+/-5.1 7+/-11 2141+/-1628 22+/-33 Units pCi/1 pCi/1 pCi/1 pCi/1 PCi/kg(wet)

PCi/kg(wet)

PCi/kg(wet) pCi/gldry) pCi/g(dry) pCi/g(dry) pCi/g(dry) pCl/g (dry) pCi/g (dry) pCi/g{dry) pCi/g (dry}

pCi/g(dry) pCi/g (dry) pCi/g(dry}

pCi/g(dry)

Revision 6 February 15, 1987

Medium and Number of Analysis Performed Samples Analyzed Fodder Crops (dry) (Cent}

Garruua Ce-141 4

Ce-144 5

Ra-226 1

Th-232 3

Fodder Crops (wet)

Gamma 5

Be-7 4

K-40 5

ZrNb-95 1

Zr-95 4

Nb-95 4

Mo-99 4

Ru-103 4

I-131 4

I-132 4

Te-132 4

I-133 4

Ba-140 4

La-140 4

Ce-141 4

Ce-144 1

Np-239 3

Game Sr-89 11 1

Sr-90 11 11 Gamma 16 K-40 16 RuRh-106 1

cs-137 1

Ra-226 2

Th-232 2

SGS-UFSAR TABLE 11.6-2 Number Above MDL Minimum 0.28+/-0.03 5.1+/-0.5

<0.1

1. 4+/-0.3

<0.04 0.10+/-0.04

<0.07 0.39+/-0.08 0.47+/-0.17 4.7+/-0.5 2.9+/-0.5 16+/-2 0.03+/-0.02 0.28+/-0.04

1. 9+/-0.2 0.04+/-0.02 0.36+/-0.04

<0.01 2.2+/-0.2 0.04+/-0.02 0.59+/-0.05

<0.01 2.4+/-0.2 0.23+/-0.06 1.4+/-0.1

<0.01

1. 8+/-0.2

<0.01 0.3+/-0.05 0.50+/-0.22 2.8+/-2.0 0.13+/-0.03 3.0+/-0.3 0.06+/-0.03 1.1+/-0.1

<0.08 0.45+/-0.25 0.72+/-0.20 3.2+/-0.3

<90

<800 32+/-16 1800+/-600 1600+/-200 27000+/-800

<10

<66+/-55

<3 620+/-50

<6 1000+/-100 20+/-10 140+/-40 8 of 9 (Cent)

Maximum 2.0+/-3.7 7+/-11 1.4+/-1.5 0.27+/-0.30 1.1+/-1. 9 0.21+/-0.46 1.1+/-2. 0 0.79+/-0.96

0. 7+/-1.5 0.19+/-0.31 1.8+/-2.0

1. 8+/-2. 4 0.77+/-0.96 2.0+/-2.5 249+/-540 4444+/-12124 Average

+/- 2 Si~

pCi/g(dry) pCi/g{dry) pCi/g{dry) pCi/g(dry) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g(wet) pCi/g{wet) pCi/g(wet) pCi/g(wet) pCi/g{wet) pCi/kg(dry}

pCi/kg(dry) pCi/kg(wet dry) pCi/kg(wet) pCi/kg(wet) pCi/kg(wet) pCi/kg(wet)

Units Revision 6 February 15, 1987

Medium and Number of Analysis Performed Samples Analyzed Fodder Crops (wet) (Cent.)

Thyroid Garruna 7

K-40 1

I-131 1

Soil Sr-90 21 18 Gam.rna 30 Be-7 4

K-40 30 Mn-54 2

ZrNb-95 1

zr-95 2

Nb-95 3

Sb-125 4

Cs-137 29 Ce-144 2

Ra-226 30 Th-232 30 DIRECT RADIATON Garruna Dose 3872M 3872 1256Q 1256 260SA 260 SGS-UFSAR TABLE 11. 6-2 Number Above MDL Minimum

<400 1800+/-600

<30 89+/-1

<0.03 1.1+/-0.1 0.18:+/-0.13 21+/-10 3.4+/-0.4 24+/-2

<0.01 0.01+/-0.01

<0.02 0.02+/-0.02 0.03+/-0.02 0.06+/-0.04 0.07+/-0.02 0.09+/-0.02 0.05+/-0.04 0.27+/-0.07

<0.1 2.8+/-0.3

<0.09 0.51+/-0.12 0.27+/-0.05

1. 5+/-0.2 0.14+/-0.06 1.4+/-0.1 2.07+/-0.14 7.28+/-0.36

1. 07+/-0. 24 5.60+/-0.54 2.83+/-0.10 5.21+/-0.22 9 of 9 (Cont}

Maximum 0.26+/-0.50 10+/-8

o. 8+/-1.5 0.87+/-0.65 0.74+/-0.58 4.26+/-1.53 4.62+/-1.18

4. 30+/-1.15 Average

+2 Sigma PCi/kg(wet}

PCi/kg(wet) pCi/g{dry) pCi/g{dry) pCi/g{dry) pCi/g(dry) pCi/g(dry) pCi/g(dry) pCi/g(dry) pCi/g{dry) pCi/g(dry) pCi/g(dry) pCi/g{dry) pCi/g{dry) rnrad/std.mon.

mrad/std.mon.

mrad/std.rnon.

Units Revision 6 February 15, 1987

ML19360A108

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ML19360A108

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