ML22292A080
Text
U.S. NUCLEAR REGULATORY COMMISSION DOCKET 50-410 LICENSE NPF-69 NINE MILE POINT NUCLEAR STATION UNIT 2 UPDATED SAFETY ANALYSIS REPORT OCTOBER 2022 REVISION 25
NMP Unit 2 USAR Chapter 11 LIST OF EFFECTIVE FIGURES Revision Figure No.
Number Chapter 11 EF 11-1 Rev. 25, October 2022 11.1-1 A00 11.1-2 A00 11.1-3 A00 11.2-1b R16 11.2-1c R16 11.2-1d R19 11.2-1e R16 11.2-1f R25 11.2-1g R19 11.2-1h R25 11.2-1j R25 11.2-1k R19 11.2-1L R15 11.2-1m R19 11.3-1a R20 11.3-1b R23 11.3-1c R00 11.4-1a R10 11.4-1b R19 11.4-1c R09 11.4-1d R09 11.4-1e R20 11.4-1f R09 11.4-1g R20 11.4-1h R19
NMP Unit 2 USAR CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS Section Title Chapter 11 11-i Rev. 25, October 2022 11.1 SOURCE TERMS 11.1.1 General Electric Reactor Coolant and Main Steam Data 11.1.1.1 Fission Products 11.1.1.2 Activation Products 11.1.1.3 Tritium 11.1.2 Fuel Fission Product Inventory and Fuel Experience 11.1.2.1 Fuel Fission Product Inventory 11.1.2.2 Fuel Experience 11.1.3 Process Leakage Sources 11.1.4 Radioactive Sources in the Liquid Radwaste System 11.1.5 Radioactive Sources in the Offgas System 11.1.6 Radioactive Sources for Component Failures 11.1.7 References 11.2 LIQUID WASTE MANAGEMENT SYSTEMS 11.2.1 Design Bases 11.2.1.1 Power Generation Design Bases 11.2.1.2 System Design Basis 11.2.1.2.1 Applicable Codes and Standards 11.2.1.2.2 Structural Design 11.2.2
System Description
11.2.2.1 Waste Collector Subsystem 11.2.2.2 Floor Drain Collector Subsystem 11.2.2.3 Regenerant Waste Subsystem 11.2.2.4 Phase Separator Subsystem 11.2.2.5 System Operational Analysis 11.2.2.6 Instrumentation and Control 11.2.2.6.1 Tanks 11.2.2.6.2 Pumps 11.2.2.6.3 Radwaste Filters 11.2.2.6.4 Radwaste Demineralizers and Resin Traps 11.2.2.6.5 Floor Drain Filter (Retired, Removed) Associated Tanks and Pumps 11.2.2.6.6 Evaporators (Waste and Regenerant) 11.2.2.6.7 Diaphragm Seals 11.2.2.7 System Operation 11.2.2.8 Performance Tests 11.2.2.9 Summary 11.2.3 Radioactive Release and Doses 11.2.3.1 Release Points
NMP Unit 2 USAR CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS (contd.)
Section Title Chapter 11 11-ii Rev. 25, October 2022 11.2.3.2 Dilution Factors 11.2.3.3 Estimated Doses 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 11.3.1 Design Bases 11.3.2
System Description
11.3.2.1 Offgas System 11.3.2.2 Ventilation Systems 11.3.2.3 Steam and Power Conversion Systems 11.3.2.4 Miscellaneous Gaseous Releases 11.3.3 Radioactive Releases 11.3.3.1 Release Points 11.3.3.2 Design and Expected Releases 11.3.3.3 Dilution Factors 11.3.3.4 Estimated Doses 11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4.1 Design Basis 11.4.2 System Inputs 11.4.2.1 System Inputs Activity 11.4.3
System Description
11.4.3.1 Spent Resin/Filter Sludge Packaging 11.4.3.2 Evaporator Bottoms Packaging 11.4.3.3 Radwaste Backup System 11.4.3.3.1 Radwaste Dewatering System 11.4.3.4 Dry Waste Packaging 11.4.3.5 Incompressible Waste Packaging 11.4.3.6 Waste Packaging Controls 11.4.3.7 Waste Handling 11.4.4 Packaging 11.4.5 Storage Facilities 11.4.6 Shipment 11.4.7 Process Control Program 11.4.8 References 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5.1 Design Bases 11.5.1.1 Design Objectives 11.5.1.1.1 Radiation Monitors Required for Safety 11.5.1.1.2 Radiation Monitors Required for Plant Operation 11.5.1.2 Design Criteria 11.5.1.2.1 Monitors Required for Safety 11.5.1.2.2 Monitors Required for Plant Operation
NMP Unit 2 USAR CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS (contd.)
Section Title Chapter 11 11-iii Rev. 25, October 2022 11.5.2
System Description
11.5.2.1 Computer-Based Radiation Monitoring Systems 11.5.2.1.1 Computer-Based Monitor Descriptions 11.5.2.1.2 Monitors Required for Safety (DRMS) 11.5.2.1.3 Monitors Required for Plant Operations 11.5.2.2 Noncomputer-Based Process Radiation Monitoring System 11.5.2.3 Calibration, Maintenance, Inspection, and Tests 11.5.2.3.1 Inspection and Tests 11.5.2.3.2 Calibration 11.5.2.3.3 Maintenance 11.5.2.4 Sampling 11.5.3 Effluent Monitoring and Sampling APPENDIX 11A RADIOLOGICAL DOSES
NMP Unit 2 USAR CHAPTER 11 LIST OF TABLES Table Number Title Chapter 11 11-iv Rev. 25, October 2022 11.1-1 REACTOR COOLANT AND MAIN STEAM RADIONUCLIDE CONCENTRATIONS 11.1-2 PARAMETERS USED TO DETERMINE REACTOR COOLANT AND MAIN STEAM RADIONUCLIDE CONCENTRATIONS 11.1-3 GENERAL ELECTRIC DATA - NOBLE RADIOGAS SOURCE TERMS 11.1-4 GENERAL ELECTRIC DATA - OTHER FISSION PRODUCT RADIOISOTOPES IN REACTOR WATER 11.1-5 GENERAL ELECTRIC DATA - COOLANT ACTIVATION PRODUCTS IN REACTOR WATER AND STEAM 11.1-6 GENERAL ELECTRIC DATA - NONCOOLANT ACTIVATION PRODUCTS IN REACTOR WATER 11.1-7 GENERAL ELECTRIC DATA - POWER ISOLATION EVENT -
ANTICIPATED OCCURRENCE 11.1-8 GENERAL ELECTRIC DATA - REACTOR COOLANT FISSION PRODUCT RADIOHALOGENS 11.2-1 EQUIPMENT DESCRIPTIONS LIQUID WASTE MANAGEMENT SYSTEM 11.2-2 APPLICABLE CODES AND STANDARDS FOR LIQUID WASTE MANAGEMENT SYSTEMS 11.2-3 MATERIAL AND CONCENTRATION BALANCE FOR LIQUID WASTE MANAGEMENT SYSTEM 11.2-4 DECONTAMINATION FACTORS OF PROCESSING UNITS 11.2-5 EXPECTED ANNUAL LIQUID RELEASES 11.2-6 PRE-OPERATIONAL DESIGN ANNUAL LIQUID RELEASE ESTIMATES 11.3-1 EXPECTED RADIOACTIVE GASEOUS EFFLUENT FROM ALL SOURCES (CI/YR) 11.3-2 DATA USED IN CALCULATING ANNUAL RELEASES OF RADIOACTIVE GASEOUS EFFLUENTS 11.3-3 DESIGN ANNUAL AVERAGE GASEOUS RELEASES VS MPC
NMP Unit 2 USAR CHAPTER 11 LIST OF TABLES (contd.)
Table Number Title Chapter 11 11-v Rev. 25, October 2022 11.4-1 ANNUAL WET AND DRY SOLID WASTE QUANTITIES VOLUME 11.4-2 EXPECTED AND DESIGN WET SOLID WASTE ACTIVITIES 11.4-3 WET SOLID RADWASTE PRINCIPAL NUCLIDE INVENTORIES 11.4-4 SOLID WASTE MANAGEMENT SYSTEM MAJOR EQUIPMENT LIST 11.4-5 EXPECTED DRY SOLID RADWASTE ANNUAL NUCLIDE INVENTORIES 11.4-6 MAXIMUM DRY SOLID RADWASTE ANNUAL NUCLIDE INVENTORIES 11.4-7 PARAMETERS USED TO CALCULATE DRY SOLID RADWASTE INVENTORIES AND ANNUAL CURIE CONTENT 11.5-1 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS 11.5-2 GRAB SAMPLES FOR RADIOLOGICAL ANALYSIS
NMP Unit 2 USAR CHAPTER 11 LIST OF FIGURES Figure Number Title Chapter 11 11-vi Rev. 25, October 2022 11.1-1 NOBLE RADIOGAS DECAY CONSTANT EXPONENT FREQUENCY HISTOGRAM 11.1-2 RADIOHALOGEN DECAY CONSTANT EXPONENT FREQUENCY HISTOGRAM 11.1-3 NOBLE RADIOGAS LEAKAGE VERSUS 1-131 LEAKAGE 11.2-1 RADIOACTIVE LIQUID WASTE SYSTEM (SHEETS A THROUGH K) 11.2-1 RADWASTE SEAL WATER SYSTEM (SHEETS L AND M) 11.3-1 OFFGAS SYSTEM (SHEETS A THROUGH C) 11.4-1 RADIOACTIVE SOLID WASTE SYSTEM (SHEETS A THROUGH H) 11.5-1 OFF-LINE GAS AND PARTICULATE MONITOR 11.5-2 OFF-LINE GAS MONITOR 11.5-2a OFF-LINE GAS MONITOR 11.5-2b OFF-LINE GAS MONITOR 11.5-3 OFF-LINE LIQUID MONITOR (SHEETS A THROUGH D) 11.5-4 ON-LINE STEAM OR LIQUID MONITOR 11.5-5 BLOCK DIAGRAM TYPICAL GASEOUS EFFLUENT MONITORING SYSTEM 11.5-6 ACCEPTABLE CONTINUOUS AIRBORNE MONITOR 11.5-6a ACCEPTABLE CONTINUOUS AIRBORNE MONITOR 11.5-7 GASEOUS RADIATION MONITORING 11.5-8 LIQUID RADIATION MONITORING (SHEETS 1 AND 2)
NMP Unit 2 USAR Chapter 11 11.1-1 Rev. 25, October 2022 CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS Expected Activities The expected reactor coolant and main steam activities form the basis for estimating the average quantity of radioactive material released to the environment and reflect normal operating conditions, including operational occurrences. These data (Table 11.1-1) are based on methods that are consistent with ANSI/ANS-18.1-1999(12). Plant parameters used to determine the expected source terms are listed in Table 11.1-2.
Design Basis Activities The design basis radioactive material levels in the reactor coolant and main steam are also presented in Table 11.1-1.
These data conservatively represent the shielding, ventilation, and radwaste liquid release design basis fission product source terms. Design failed fuel conditions correspond to an offgas release rate of 100 uCi/sec/MWt, or 0.3536 Ci/sec, at 30-min delay and are developed by scaling up from expected NUREG-0016 Revision 1 source term data(1). In addition, the design basis source terms in Table 11.1-1 take into consideration the General Electric Company (GE) design basis data described in Section 11.1.1 and Tables 11.1-3 through 11.1-7. For the design basis concentrations, Table 11.1-1 presents the higher value for a given isotope from either the adjusted NUREG-0016 data or the GE data and represents a conservative data set.
11.1.1 General Electric Reactor Coolant and Main Steam Data GE has evaluated radioactive material sources (activation products and fission product release from fuel) in operating boiling water reactors (BWRs). These source terms are reviewed and periodically revised to incorporate up-to-date information.
The information provided in this section defines the design basis radioactive material levels in the reactor water, steam, and offgas. The various radioisotopes have been grouped as fission products, coolant activation products, and noncoolant activation products. The fission product activity levels are based on measurements of BWR water and offgas at several stations through mid-1971. Observations made at KRB and Dresden Unit 2 provided input to the design basis values. The design basis radioactive material levels do not necessarily include all the radioisotopes observed or theoretically predicted to be present. The radioisotopes included are considered significant to one or more of the following criteria:
NMP Unit 2 USAR Chapter 11 11.1-2 Rev. 25, October 2022
- 1.
Plant equipment design.
- 2.
Shielding design.
- 3.
Understanding of system operation and performance.
- 4.
Measurement practicability.
- 5.
Evaluation of radioactive material releases to the environment.
A combination of GE and NUREG-0016 data was used for the main steam radionuclide concentrations.
11.1.1.1 Fission Products Noble Radiogas Fission Products The noble radiogas fission product source terms observed in operating BWRs are generally complex mixtures whose sources vary from minuscule defects in cladding to minute quantities of "tramp" uranium on external cladding surfaces. The relative concentrations or amounts of noble radiogas isotopes relative to the other noble gas isotopes can be described as follows:
Equilibrium:
Rg K1y Recoil:
Rg K2y The terms in these and succeeding equations are defined in the nomenclature section. The constants K1 and K2 describe the fractions of the total fissions that are involved in each of the releases. The equilibrium and recoil mixtures are the two extremes of the mixture spectrum that are physically possible.
The equilibrium mixture results when a sufficient time delay exists between the fission event and the time of release of the radiogases from the fuel to the coolant for the radiogases to approach equilibrium levels in the fuel. When there is no delay or impedance between the fission event and the release of the radiogases, the recoil mixture is observed.
Prior to Vallecitos Boiling Water Reactor (VBWR) and Dresden Unit 1 experience, it had been assumed that noble radiogas leakage from the fuel would be the equilibrium mixture of the noble radiogases present in the fuel. VBWR and early Dresden Unit 1 experience indicated that the actual mixture most often observed approached a distribution that was intermediate in character to the two extremes(2). This intermediate decay mixture was termed the diffusion mixture. It must be emphasized that this diffusion mixture is merely one possible point on the mixture spectrum, ranging from the equilibrium to the recoil mixture, and does not have the absolute mathematical and
NMP Unit 2 USAR Chapter 11 11.1-3 Rev. 25, October 2022 mechanical basis for the calculational methods possible for equilibrium and recoil mixtures. This diffusion distribution pattern is described as follows(3):
Diffusion:
5.0 3
y K
Rg The constant K3 describes the fraction of total fissions that are involved in the release. The exponent of the decay constant,,
is midway between that of equilibrium, 0, and recoil, 1. The diffusion pattern value of 0.5 was originally derived from diffusion theory.
Although the previously described diffusion mixture has been used by GE as a basis for design since 1963, the design basis release magnitude used has varied from 0.5 Ci/sec to 0.1 Ci/sec as measured after 30-min decay (t=30 min). The noble radiogas source term rate after 30-min decay has been used as a conventional measure of the design basis fuel leakage rate, since it is conveniently measurable at existing BWR plants and is consistent with the nominal design basis 30-min offgas holdup system used in a number of plants. Since about 1967, the design basis release rate from the core was established at an annual average of 0.1 Ci/sec (t=30 min). The expected annual average is significantly below the design basis annual average. This design value was selected on the basis of operating experience rather than predictive assumptions. Several judgment factors, including the significance of environmental release, reactor water radioisotope concentrations, liquid waste handling, and effluent disposal criteria, building air contamination, shielding design, and turbine and other component contamination affecting maintenance, have been considered in establishing this level.
Although noble radiogas source terms from fuel above the design basis can be tolerated for reasonable periods of time, long-term operation at such levels would be undesirable. Continual assessment of this value is made on the basis of actual operating experience in BWRs(4).
While the GE noble radiogas source term magnitude was established at 0.1 Ci/sec (t=30 min) in 1967, it was recognized that there may be a more statistically applicable distribution for the noble radiogas mixture. Sufficient data were available from KRB operations from 1967 to mid-1971, along with Dresden Unit 2 data from operation in 1970 and several months in 1971, to more accurately characterize the noble radiogas mixture pattern from an operating BWR.
The basic equation for each radioisotope used to analyze the collected data is:
(
)(
)
t T
m g
g e
e y
K R
=
1 (11.1-1)
NMP Unit 2 USAR Chapter 11 11.1-4 Rev. 25, October 2022 With the exception of Kr-85 with a half-life of 10.74 yr, the noble radiogas fission products in the fuel are essentially at an equilibrium condition after an irradiation period of several months (rate of formation is equal to rate of decay).
Therefore, the term (
)
T e
1 approaches 1 and can be neglected when the reactor has been operating at steady state for long periods of time. The term (
)
t e
is used to adjust the releases from the fuel (t=0) to the decay time for which values are needed. Historically, t=30 min has been used. When discussing long steady state operation and leakage from the fuel (t=0), the following simplified form of Equation 11.1-1 can be used to describe the leakage of each noble radiogas isotope:
m g
g y
K R
=
(11.1-2)
The constant, Kg, describes the magnitude of total leakage. The rate of noble radiogas leakage with respect to each other (composition) is expressed in terms of the decay constant term,
, and the fission yield fraction, y.
Dividing both sides of Equation 11.1-2 by y and taking the logarithm of both sides results in the following equation:
(
)
( )
(
)
g g
K m
y R
log log
/
log
+
=
(11.1-3)
Equation 11.1-3 represents a straight line when log (Ry/y) is plotted versus log (); m is the slope of the line. By fitting actual data from KRB and Dresden Unit 2 (using least squares techniques), the value of m can be obtained. With radiogas leakage at KRB over the nearly 5-yr period varying from 0.001 to 0.056 Ci/sec (t=30 min), and with radiogas leakage at Dresden Unit 2 varying from 0.001 to 0.169 Ci/sec (t=30 min), the average value for m of 0.4 with a standard deviation of 0.07 was determined (Figure 11.1-1). As shown on this figure, variations in m were observed in the range m=0.1 to m=0.6.
After establishing the value of m=0.4, the value of Kg can be calculated by selecting a value for Rg. With g
R at 30 min =
0.1 Ci/sec, Kg is 2.6 x 107 and Equation 11.1-1 becomes:
(
)(
)
t T
g e
e y
x R
=
1 10 6.2 4.0 7
(11.1-4)
This updated noble radiogas source term mixture has been designated the 1971 Mixture to differentiate it from the diffusion mixture.
The noble gas source term for each radioisotope can be calculated from Equation 11.1-4. The resultant source terms are presented in Table 11.1-3 as leakage from fuel (t=0) and after 30-min decay. Since the NUREG-0016 values were higher than the GE data, the NUREG-0016 concentrations were used to calculate noble radiogas fission product releases.
Radiohalogen Fission Products
NMP Unit 2 USAR Chapter 11 11.1-5 Rev. 25, October 2022 Historically, the radiohalogen design basis source term was established by the same equation as that used for noble radiogases. In a manner similar to that used with gases, a simplified equation can be shown to describe the release of each halogen radioisotope:
n h
h y
K R
=
(11.1-5)
The constant, Kh, describes the magnitude of the total leakage from the fuel. The rate of halogen radioisotope releases with respect to each other (composition) is expressed in terms of the fission yield, y, and the decay constant. As was done with the noble radiogases, the average value was determined for n.
The average value for n is 0.5, with a standard deviation of 0.19 (Figure 11.1-2). As can be seen from this figure, variations in n were observed in the range from n=0.1 to n=0.9.
As mentioned, it appears that the use of the previous method of calculating radiohalogen leakage from fuel is overly conservative. Figure 11.1-3 relates KRB and Dresden Unit 2 noble radiogas releases versus I-131 leakage. While Dresden Unit 2 data during the period August 1970 to January 1971 show a relationship between noble radiogas and I-131 leakage under one fuel condition, there was no simple relationship for all fuel conditions experienced. Also, during this period, high radiogas leakages were not accompanied by high radioiodine leakage from the fuel. Except for one KRB datum point, all steady state I-131 leakages observed at KRB or Dresden Unit 2 were equal to or less than 505 uCi/sec. Even at Dresden Unit 1 in March 1965, when severe defects were experienced in stainless steel clad fuel, I-131 leakages greater than 500 uCi/sec were not experienced. Figure 11.1-3 shows that these higher radioiodine leakages from the fuel were related to noble radiogas source terms of less than the design basis value of 0.1 Ci/sec (t=30 min). This may be partially explained by inherent limitations due to internal plant operational problems that caused plant derating.
In general, one would not anticipate continued operation at full power for any significant time period with fuel cladding defects that would be indicated by I-131 leakage from the fuel in excess of 700 uCi/sec. When high radiohalogen leakages are observed, other fission products will be present in greater amounts.
By using these judgment factors and experience to date, the design basis radiohalogen source terms from fuel were established based on an I-131 leakage of 700 uCi/sec. This value, as seen on Figure 11.1-3, is consistent with the experiential data and the design basis noble radiogas source term of 0.1 Ci/sec (t=30 min). With the I-131 design basis source term established, Kh equals 2.4 x 107, and the halogen radioisotope release can be expressed by the following equation:
NMP Unit 2 USAR Chapter 11 11.1-6 Rev. 25, October 2022
(
)(
)
t T
h e
e y
x R
=
1 10 4.2 5.0 7
(11.1-6)
Concentrations of radiohalogens in reactor water are listed in Table 11.1-8 and can be calculated using the following equation:
)
(
+
+
= M Rh Ch (11.1-7)
Terms used in Equation 11.1-7 are defined in the nomenclature section.
Although carryover of most soluble radioisotopes from reactor water to steam is observed to be <0.1 percent (<0.001 fraction),
the observed carryover for radiohalogens has varied from 0.1 percent to about 2 percent on newer plants. The average of observed radiohalogen carryover measurements has been 1.2 percent with a standard deviation of 0.9. In the present source term definition, a radiohalogen carryover design basis of 2 percent (0.02 fraction) is used.
Other Fission Products The observations of other fission products (and transuranic nuclides, including Np-239) in operating BWRs are not adequately correlated by simple equations. For these radioisotopes, design basis concentrations in reactor water have been estimated conservatively from experiential data (Table 11.1-4). Carryover of these radioisotopes from the reactor water to the steam is estimated to be <0.1 percent (<0.001 fraction). In addition to carryover, however, decay of noble radiogases in the steam leaving the reactor results in production of noble gas daughter radioisotopes in the steam and condensate systems.
Some daughter radioisotopes (for example, yttrium and lanthanum) were not listed as being in reactor water. Their independent leakage to the coolant is negligible; however, these radioisotopes may be observed in some samples in equilibrium or approaching equilibrium with the parent radioisotope.
Except for Np-239, trace concentrations of transuranic isotopes have been observed on only a few samples where extensive and complex analyses were carried out. The predominant alpha emitter present in reactor water is Cm-242, at an estimated concentration of 10-6 uCi/g or less, which is below the maximum concentration allowed in drinking water for continuous use by the general public. The concentration of alpha-emitting plutonium radioisotopes is more than one order of magnitude lower than that of Cm-242. Plutonium-241 (a beta emitter) may also be present in concentrations comparable to the Cm-242 level.
Nomenclature
NMP Unit 2 USAR Chapter 11 11.1-7 Rev. 25, October 2022 The terms used in equations for source term calculations are defined as follows:
Rg =
Leakage rate of a noble radiogas isotope, uCi/sec Rh =
Leakage rate of a halogen radioisotope, uCi/sec y =
Fission yield of a radioisotope, atoms/fission
=
Decay constant of a radioisotope, sec-1 T =
Fuel irradiation time, sec t =
Decay time following leakage from fuel rod, sec m =
Noble radiogas decay constant exponent, dimensionless n =
Radiohalogen decay constant exponent, dimensionless Kg =
Constant establishing the level of noble radiogas leakage from fuel Kh =
Constant establishing the level of radiohalogen leakage from fuel Ch =
Concentration of a halogen radioisotope in reactor, uCi/g M =
Mass of water in the operating reactor, g
=
Cleanup system removal constant, sec-1
=
Cleanup system flowrate (g/sec)
M(g)
=
Halogen steam carryover removal constant, sec-1
=
Concentration of halogen radioisotope in steam (Ci/g Steam flow Ch(Ci/g) (g/sec)
M(g) 11.1.1.2 Activation Products Coolant Activation Products Coolant activation products are not adequately correlated by simple equations. Design basis concentrations in reactor water and steam have been estimated from experiential data and additional conservatism added by using the higher NUREG-0016 Revision 1 values. The resultant concentrations are listed in Table 11.1-5.
Noncoolant Activation Products
NMP Unit 2 USAR Chapter 11 11.1-8 Rev. 25, October 2022 The activation products formed by activation of impurities in the coolant or by corrosion of irradiated system materials are not adequately correlated by simple equations. The design basis source terms of noncoolant activation products have been estimated conservatively from experiential data(5) and, in some cases, more conservative values were based on NUREG-0016. The resultant concentrations are listed in Table 11.1-6. Carryover of these isotopes from the reactor water to the steam is estimated to be 0.1 percent (<0.001 fraction).
11.1.1.3 Tritium In a BWR, tritium, which is available for release in liquid and gaseous wastes, is produced by three principal methods:
- 1.
Activation of naturally-occurring deuterium in the primary coolant.
- 2.
Nuclear fission of UO2 fuel.
- 3.
Neutron capture reactions with boron used in reactivity control rods.
The tritium formed in control rods, which may be released from a BWR in liquid or gaseous effluents, is believed to be negligible. A prime source of tritium is that from activation of deuterium in the primary coolant. A secondary source of tritium may also be transferred from fuel to primary coolant. The following discussion is limited to the uncertainties associated with estimating the amounts of tritium generated in a BWR which are available for release.
Tritium produced by activation of deuterium in the primary coolant is available for release in liquid or gaseous effluents and can be determined by using the equation:
P x
V Ract 4
10 7.3
=
(11.1-8)
Where:
Ract = Tritium formation rate by deuterium activation, uCi/sec/MWt
= Macroscopic thermal neutron cross section, cm-1
= Thermal neutron flux, neutrons/(cm2), sec V
= Coolant volume in core, cm3
= Tritium radioactive decay constant, 1.78 x 10-9 sec-1 P = Reactor power level, MWt
NMP Unit 2 USAR Chapter 11 11.1-9 Rev. 25, October 2022 For recent BWR designs, Ract is calculated to be:
(
)
MWt uCi x
sec/
/
10 4.0 3.1 4
(11.1-9)
The uncertainty indicated is derived from the estimated errors in selecting values for the coolant volume in the core, coolant density in the core, abundance of deuterium in light water (some additional deuterium is present because of the H(n,) D reaction), thermal neutron flux, and microscopic cross section for deuterium.
The fraction of tritium produced by fission which may transfer from fuel to the coolant is much more difficult to estimate.
However, since zircaloy-clad fuel rods are used in BWRs, essentially all fission product tritium remains in the fue1 rods(6).
This is confirmed by the study made at Dresden Unit 1 in 1968 by the U.S. Public Health Service (USPHS), which suggests that essentially all tritium released from the plant could be accounted for by the deuterium activation source(7). For purposes of estimating the leakage of tritium from defective fuel, it is assumed that the tritium leaks in a manner similar to the leakage of noble radiogases. Thus the empirical relationship described as the diffusion mixture is used for predicting the source term of individual noble gas radioisotopes as a function of total noble gas source term. The equation that describes this relationship is:
Ky Rdiff =
(11.1-10)
Where:
Rdiff = Leakage rate of the radioisotope, uCi/sec y
= Fission yield fraction
= Radioactive decay constant, sec-1 K
= Constant related to total leakage rate If the total noble radiogas source term is 105 uCi/sec after 30-min decay, leakage from the fuel is calculated to be about 0.24 uCi/sec of tritium. To place this value in perspective with the USPHS study, the observed leakage rate of Kr-85 (which has a half-life similar to that of tritium) was 0.06 to 0.4 times that calculated using the diffusion mixture relationship.
This would suggest that the actual tritium leakage rate might range from 0.015 to 0.10 uCi/sec. Since the annual average noble radiogas leakage from a BWR is expected to be less than 0.1 Ci/sec (t=30 min), the annual average tritium release rate from the fission source can be conservatively estimated at 0.12
NMP Unit 2 USAR Chapter 11 11.1-10 Rev. 25, October 2022 0.12 uCi/sec. Based on this approach, the estimated total tritium appearance rate in reactor coolant and the release rate in effluents can be estimated to be about 16 Ci/yr.
Tritium formed in the reactor is generally present as tritiated oxide (HTO) and, to a lesser degree, as tritiated gas (HT).
Tritium concentration in the steam formed in the reactor is the same as in the reactor water at any given time. This tritium concentration is also present in condensate and feedwater.
Since radioactive effluents generally originate from the reactor and power cycle equipment, radioactive effluents also have this tritium concentration. Condensate storage receives treated water from the radioactive waste system and rejects water to the condensate system. Thus, all plant process water has a common tritium concentration.
Offgases released from the plant contain tritium which is present as HT resulting from reactor water radiolysis. In addition, water vapor present in ventilation air due to process steam leaks or evaporation from sumps, tanks, and spills on floors also contains tritium. The remainder of tritium leaves the plant in liquid effluents.
Recombination of radiolysis gases in the air ejector offgas system forms water vapor that is condensed and returned to the main condenser. This reduces the amount of tritium leaving in gaseous effluents and results in a slightly higher tritium concentration in the plant process water. Reducing the amount of liquid effluent discharged also results in a higher process coolant equilibrium tritium concentration.
Essentially all tritium entering the primary coolant is eventually released to the environs, either as water vapor and gas to the atmosphere, or as liquid effluent to the plant discharge. Reduction due to radioactive decay is negligible due to the 12-yr half-life of tritium.
The USPHS study at Dresden Unit 1 estimated that approximately 90 percent of the tritium release was observed in liquid effluent, with the remaining 10 percent leaving as gaseous effluent(7). Efforts to reduce the volume of liquid effluent discharges may change this distribution; however, from a practical standpoint, the fraction of tritium leaving as liquid effluent may vary between 60 and 90 percent, with the remainder leaving in gaseous effluent. The tritium design basis for Unit 2 is taken from conservative values of NUREG-0016.
11.1.2 Fuel Fission Product Inventory and Fuel Experience 11.1.2.1 Fuel Fission Product Inventory Fuel rod and fuel plenum radioisotopic inventory, along with escape rate coefficients and release fractions, are not used in
NMP Unit 2 USAR Chapter 11 11.1-11 Rev. 25, October 2022 establishing BWR design basis source term coolant activities.
Fuel fission product inventory information is used in establishing fission product source terms for accident analysis and is therefore discussed in Chapter 15.
11.1.2.2 Fuel Experience A discussion of fuel experience gained for BWR fuel, including failure experience, burnup experience, and thermal conditions under which the experience was gained, is available in NEDO-10505, NEDO-20922, and NEDO-10173(3,8,9).
11.1.3 Process Leakage Sources Process leakage results in potential release paths for noble gases and other volatile fission products via ventilation systems. All liquid from process leaks is collected and routed to the liquid-solid radwaste system. Radionuclide releases via ventilation paths are at extremely low levels and have been insignificant compared to process offgas from operating BWR plants. However, because the implementation of improved process offgas treatment systems makes the ventilation release comparatively significant, measurements have been conducted to identify and qualify these low-level release paths.
Leakage of fluids from the process system results in the release of radionuclides into plant buildings. In general, the noble radiogases remain airborne and are released to the atmosphere with little delay via the building ventilation exhaust ducts.
The radionuclides are partitioned between air and water, and airborne radioiodines may plateout on metal surfaces, concrete, and paint. A significant amount of radioiodine remains in the air or is desorbed from surfaces. Radioiodines are found in ventilation air as methyl and inorganic iodines, which are defined here as particulate, elemental, and hypoiodous acid forms of iodine. Particulates are also present in the ventilation exhaust air.
The airborne radiological releases from BWR building heating, ventilating, and air conditioning (HVAC) and the main condenser mechanical vacuum pump have been compiled and evaluated in NEDO-21159(10). This report is periodically updated to incorporate the most recent data on airborne emissions. The results of these evaluations (Section 12.2) are based on data obtained by owner and utility personnel and special in-plant studies of operating BWR plants by independent organizations and GE.
11.1.4 Radioactive Sources in the Liquid Radwaste System The radioactive sources for the liquid radwaste system are described in Section 11.2.
NMP Unit 2 USAR Chapter 11 11.1-12 Rev. 25, October 2022 11.1.5 Radioactive Sources in the Offgas System The radioactive sources for the offgas system are described in Section 11.3.
11.1.6 Radioactive Sources for Component Failures The radioactive sources considered for evaluating the radiological consequences of component failures are described in Section 15.7.
11.1.7 References
- 1.
NUREG-0016, Revision 1. Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors (BWRs), 1979.
- 2.
Brutschy, F. J. A Comparison of Fission Product Release Studies in Loops and VBWR. Presented at Tripartite Conference on Transport of Materials in Water Systems, Chalk River, Canada, February 1961.
- 3.
Williamson, H. E. and Ditmore, D. C. Experience with BWR Fuel Through September 1971, NEDO-10505, May 1972 (Update).
- 4.
Skarpelos, J. M. and Gilbert, R. S. Technical Derivation of BWR 1971 Design Basis Radioactive Material Source Terms, NEDO-10871, March 1973.
- 5.
Elkins, R. B. Experience with BWR Fuel Through December 1976, NEDO-21660, July 1977.
- 6.
Ray, J. W. Tritium in Power Reactors. Reactor and Fuel-Processing Technology, 12 (1), p 19-26, Winter 1968-1969.
- 7.
Kahn, B., et al. Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor, BRH/DER 70-1, March 1970.
- 8.
Elkins, R. B. Experience with BWR Fuel Through September 1974, NEDO-20922, June 1975.
- 9.
Williamson, H. E. and Ditmore, D. C. Current State of Knowledge of High Performance BWR Zircaloy Clad UO2 Fuel, NEDO-10173, May 1970.
- 10. Marrero, T. R. Airborne Releases From BWRs for Environmental Impact Evaluations, NEDO-21159, March 1976.
- 11. GE Document GE-NE-187-08-1090, Revision 1, July 1992.
NMP Unit 2 USAR Chapter 11 11.1-13 Rev. 25, October 2022
- 12. ANSI/ANS-18.1-1999, Radioactive Source Term for Normal Operation of Light Water Reactors, American Nuclear Society, September 1999.
NMP Unit 2 USAR Chapter 11 11.1-14 Rev. 25, October 2022 TABLE 11.1-1 REACTOR COOLANT AND MAIN STEAM RADIONUCLIDE CONCENTRATIONS(1)
Reactor Coolant Main Steam Design Design Basis Expected(4)
Basis Expected(4)
Isotope (uCi/g)
(uCi/g)
(uCi/g)
(uCi/g)
Noble Gases(2)
Kr-83m 6.4-3(3) 5.9E-04 Kr-85m 1.1-2 1.0E-03 Kr-85 3.5-5 4.0E-06 Kr-87 3.9-2 3.3E-03 Kr-88 3.9-2 3.3E-03 Kr-89 2.4-1 2.1E-02 Kr-90 5.3-1 Kr-91 6.4-1 Kr-92 6.4-1 Kr-93 1.7-1 Kr-94 4.2-2 Kr-95 3.9-3 Kr-97 2.5-5 Xe-131m 2.8-5 3.3E-06 Xe-133m 5.3-4 4.9E-05 Xe-133 1.5-2 1.4E-03 Xe-135m 5.0-2 4.4E-03 Xe-135 4.2-2 3.8E-03 Xe-137 2.8-1 2.6E-02 Xe-138 1.6-1 1.5E-02 Xe-139 5.3-1 Xe-140 5.7-1 Xe-141 4.6-1 Xe-142 1.3-1 Xe-143 2.3-2 Xe-144 1.1-3 Halogens Br-83 2.3-2 3.5-4 Br-84 3.0-2 5.6-4*
Br-85 1.7-2*
3.4-4*
I-131 1.3-2*
3.3E-03 2.6-4*
7.1E-05 I-132 2.2-1 2.7E-02 3.3-3 5.9E-04 I-133 1.6-1 2.2E-02 2.6-3 4.7E-04 I-134 4.0-1 4.4E-02 7.8-3 9.9E-04 I-135 1.7-1 3.0E-02 2.8-3 6.6E-04
NMP Unit 2 USAR Chapter 11 11.1-15 Rev. 25, October 2022 TABLE 11.1-1 (Cont'd.)
Reactor Coolant Main Steam Design Design Basis Expected(4)
Basis Expected(4)
Isotope (uCi/g)
(uCi/g)
(uCi/g)
(uCi/g)
Cesium and Rubidium Rb-89 2.2-2 3.6E-03 2.2-5 1.3E-05 Cs-134 8.5-5 1.7E-05 8.5-8 5.9E-08 Cs-136 5.5-5 1.1E-05 5.5-8 3.9E-08 Cs-137 2.2-4 4.5E-05 2.2-7 1.6E-07 Cs-138 1.6-1*
7.2E-03 1.6-4*
2.5E-05 Water Activation Products N-13 5.8-2*
7.0-3 N-16 1.0+2 6.0+1 1.0+2 5.0+1 N-17 1.8-2*
3.6-2*
O-19 1.1+0*
5.9-1*
F-18 4.8-2*
4.0-3 Tritium H-3 1.0-2 1.0-2 1.0-2 1.0-2 Other Nuclides Na-24 4.1-3 1.3E-03 4.1-6 4.4E-06 P-32 7.9-5 2.5E-05 7.9-8 8.7E-08 Cr-51 2.4-3 1.9E-03 2.4-6 6.5E-06 Mn-54 4.4-5*
2.2E-05 4.4-8*
7.6E-08 Mn-56 5.5-2*
1.7E-02 5.5-5*
5.9E-05 Fe-55 3.9-4 6.2E-04 3.9-7 2.2E-06 Fe-59 8.8-5*
1.9E-05 8.8-8*
6.5E-08 Co-58 5.5-3*
6.2E-05 5.5-6*
2.2E-07 Co-60 5.5-4*
1.2E-04 5.5-7*
4.4E-07 Ni-63 3.9-7 6.2E-07 3.9-10 2.2E-09 Ni-65 3.3-4*
3.3-7*
Cu-64 1.3-2 1.9E-03 1.3-5 6.7E-06 Zn-65 7.9-5 6.2E-05 7.9-8 2.2E-07 Zn-69m 1.1-3 1.1-6 Sr-89 1.2-3*
6.2E-05 1.2-6*
2.2E-07 Sr-90 8.8-5*
4.4E-06 8.8-8*
1.5E-08 Sr-91 3.2-2*
2.6E-03 3.2-5*
9.0E-06 Sr-92 6.6-2*
6.7E-03 6.6-5 2.3E-05 Y-91 1.1-4 2.5E-05 1.1-7 8.7E-08 Y-92 2.0-2 4.0E-03 2.0-5 1.4E-05
NMP Unit 2 USAR Chapter 11 11.1-16 Rev. 25, October 2022 TABLE 11.1-1 (Cont'd.)
Reactor Coolant Main Steam Design Design Basis Expected(4)
Basis Expected(4)
Isotope (uCi/g)
(uCi/g)
(uCi/g)
(uCi/g)
Y-93 1.2-2 2.6E-03 1.2-5 9.0E-06 Zr-95 2.3-5 5.0E-06 2.3-8 1.7E-08 Zr-97 1.8-5 1.8-8 Nb-95 2.3-5 5.0E-06 2.3-8 1.7E-08 Nb-98 1.6-2 1.6-5 Mo-99 8.9-3*
1.3E-03 8.9-6*
4.4E-06 Tc-99m 6.2-2 1.3E-03 6.2-5 4.4E-06 Tc-101 3.9-1 3.9-4 Tc-104 3.4-1 3.4-4 Ru-103 5.6-5 1.2E-05 5.6-8 4.4E-08 Ru-105 6.4-3 6.4-6 Ru-106 8.5-6 1.9E-06 8.5-9 6.5E-09 Ag-110m 6.6-5*
6.2E-07 6.6-8*
2.2E-09 Te-129m 1.3-4*
2.5E-05 1.3-7*
8.7E-08 Te-131m 2.8-4 6.3E-05 2.8-7 2.2E-07 Te-132 5.6-3*
6.3E-06 5.6-6*
2.2E-08 Ba-139 1.1-1*
1.1-4*
Ba-140 3.4-3*
2.5E-04 3.4-6*
8.7E-07 Ba-141 1.6-1*
1.6-4*
Ba-142 1.6-1*
1.6-4*
La-142 1.8-2 1.8-5 Ce-141 8.5-5 1.9E-05 8.5-8 6.5E-08 Ce-143 8.5-5 8.5-8 Ce-144 1.3-5*
1.9E-06 1.3-8*
6.5E-09 Pr-143 1.1-4 1.1-7 Nd-147 8.5-6 8.5-9 W-187 3.3-3*
1.9E-04 3.3-6*
6.6E-07 Np-239 9.5-2*
5.0E-03 9.5-5*
1.8E-05 (1)
Design basis values marked with an asterisk are based on GE document GE-NE-187-08-1090, Revision 1, July 1992; all other design basis values are based on NUREG-0016, Revision 1.
(2)
The design and expected concentrations for noble gases in reactor coolant are negligible.
(3) 6.4-3 = 6.4x10-3.
(4)
Only isotopes that correspond to the design basis are listed.
NMP Unit 2 USAR Chapter 11 11.1-17 Rev. 25, October 2022 TABLE 11.1-2 PARAMETERS USED TO DETERMINE REACTOR COOLANT AND MAIN STEAM RADIONUCLIDE CONCENTRATIONS Variable Unit*
Rated thermal power 3,988 MWt Rated steam flow rate 17,636,000 lb/hr Weight of reactor coolant in RPV 600,000 lb at rated power Reactor coolant cleanup system 270,000 lb/hr flow rate Ratio of condensate demineralizer 0.69 flow rate to steam flow rate Assumed moisture carryover 0.35%
NMP Unit 2 USAR Chapter 11 11.1-18 Rev. 25, October 2022 TABLE 11.1-3 GENERAL ELECTRIC DATA NOBLE RADIOGAS SOURCE TERMS Source Term Source Term at t=0 at t=30 min Isotope Half-Life (uCi/sec)
(uCi/sec)
Kr-83m 1.86 hr 3.4 x 103 2.8 x 103 Kr-85m 4.48 hr 6.1 x 103 5.7 x 103 Kr-85 10.72 yr 2.0E + 01 2.0E + 01 Kr-87 76.0 min 2.0 x 104 1.5 x 104 Kr-88 2.84 hr 2.0 x 104 1.8 x 104 Kr-89 3.16 min 1.3 x 105 1.8 x 102 Kr-90 32.3 sec 2.8 x 105 Kr-91 8.6 sec 3.3 x 105 Kr-92 1.84 sec 3.3 x 105 Kr-93 1.29 sec 9.3 x 104 Kr-94 0.21 sec 2.3 x 104 Kr-95 0.78 sec 2.1 x 103 Kr-97 0.1 sec 1.4 x 101 Xe-131m 11.92 day 1.5 x 101 1.5 x 101 Xe-133m 2.19 day 2.9 x 102 2.9 x 102 Xe-133 5.25 day 8.2 x 103 8.2 x 103 Xe-135m 15.3 min 2.6 x 104 6.7 x 103 Xe-135 9.10 hr 2.2 x 104 2.1 x 104 Xe-137 3.84 min 1.5 x 105 6.7 x 102 Xe-138 14.1 min 8.9 x 104 2.1 x 104 Xe-139 40.0 sec 2.8 x 105 Xe-140 13.6 sec 3.0 x 105 Xe-141 1.72 sec 2.4 x 105 Xe-142 1.20 sec 7.3 x 104 Xe-143 0.30 sec 1.2 x 104 Xe-144 1.2 sec 5.6 x 102 Total 2.4 x 106 1.0 x 105 NOTE: 3.4x103 = 3.4E+03
NMP Unit 2 USAR Chapter 11 11.1-19 Rev. 25, October 2022 TABLE 11.1-4 GENERAL ELECTRIC DATA OTHER FISSION PRODUCT RADIOISOTOPES IN REACTOR WATER Concentration*
Isotope Half-Life (uCi/g)
Sr-89 50.52 day 1.2 x 10-3 Sr-90 29 yr 8.8 x 10-5 Sr-91 9.5 hr 3.2 x 10-2 Sr-92 2.71 hr 6.6 x 10-2 Zr-95 64.03 day 1.5 x 10-5 Zr-97 16.8 hr 1.4 x 10-5 Nb-95 34.98 day 1.6 x 10-5 Mo-99 65.94 hr 8.9 x 10-3 Tc-99m 6.01 hr 4.0 x 10-2 Tc-101 14.2 min 1.4 x 10-1 Ru-103 39.24 day 7.5 x 10-6 Ru-106 372.6 day 1.0 x 10-6 Te-129m 33.4 day 1.3 x 10-4 Te-132 78.2 hr 5.6 x 10-3 Cs-134 2.065 yr 6.1 x 10-5 Cs-136 13.1 day 4.1 x 10-5 Cs-137 30.17 yr 9.3 x 10-5 Cs-138 32.2 min 1.6 x 10-1 Ba-139 83.1 min 1.1 x 10-1 Ba-140 12.76 day 3.4 x 10-3 Ba-141 18.3 min 1.6 x 10-1 Ba-142 10.7 min 1.6 x 10-1 Ce-141 32.5 day 1.5 x 10-5 Ce-143 33.0 hr 1.4 x 10-5 Ce-144 284.4 day 1.3 x 10-5 Pr-143 13.58 day 1.5 x 10-5 Nd-147 10.99 day 5.4 x 10-6 Np-239 2.35 day 9.5 x 10-2 Based on noble gas release rate of 0.1 Ci/sec after 30 min.
NMP Unit 2 USAR Chapter 11 11.1-20 Rev. 25, October 2022 TABLE 11.1-5 GENERAL ELECTRIC DATA COOLANT ACTIVATION PRODUCTS IN REACTOR WATER AND STEAM Reactor Steam Water Concentration Concentration Isotope Half-Life (uCi/g)
(uCi/g)
N-13 9.97 min 1.5 x 10-3 5.8 x 10-2 N-16 7.13 sec 5.0 x 101 4.1 x 101 N-17 4.17 sec 3.6 x 10-2 1.8 x 10-2 O-19 26.9 sec 5.9 x 10-1 1.1 x 10-0 F-18 109.8 min 4.4 x 10-4 4.8 x 10-2
NMP Unit 2 USAR Chapter 11 11.1-21 Rev. 25, October 2022 TABLE 11.1-6 GENERAL ELECTRIC DATA NONCOOLANT ACTIVATION PRODUCTS IN REACTOR WATER Concentration Isotope Half-Life (uCi/g)
Na-24 14.97 hr 2.2 x 10-3 P-32 14.28 day 2.2 x 10-5 Cr-51 27.7 day 5.5 x 10-4 Mn-54 312.2 day 4.4 x 10-5 Mn-56 2.579 hr 5.5 x 10-2 Co-58 70.91 day 5.5 x 10-3 Co-60 5.272 yr 5.5 x 10-4 Fe-59 44.51 day 8.8 x 10-5 Ni-65 2.520 hr 3.3 x 10-4 Zn-65 243.8 day 2.2 x 10-6 Zn-69m 13.8 hr 3.3 x 10-5 Ag-110m 249.8 day 6.6 x 10-5 W-187 23.9 hr 3.3 x 10-3
NMP Unit 2 USAR Chapter 11 11.1-22 Rev. 25, October 2022 TABLE 11.1-7 GENERAL ELECTRIC DATA POWER ISOLATION EVENT - ANTICIPATED OCCURRENCE Isotopic Spiking Activity Isotope (Ci/bundle)
I-131 2.1 I-132 3.2 I-133 5.0 I-134 5.4 I-135 4.8 Kr-83m 0.9 Kr-85m 2.2 Kr-85 0.5 Kr-87 4.3 Kr-88 6.1 Kr-89 8.0 Xe-131m 0.1 Xe-133m 0.3 Xe-133 11.6 Xe-135m 1.8 Xe-135 11.0 Xe-137 10.5 Xe-138 10.6
NMP Unit 2 USAR Chapter 11 11.1-23 Rev. 25, October 2022 TABLE 11.1-8 GENERAL ELECTRIC DATA REACTOR COOLANT FISSION PRODUCT RADIOHALOGENS Concentration*
Isotope Half-Life (uCi/g)
Br-83 2.39 hr 1.4 x 10-2 Br-84 31.8 min 2.8 x 10-2 Br-85 2.87 min 1.7 x 10-2 I-131 8.040 day 1.3 x 10-2 I-132 2.28 hr 1.2 x 10-1 I-133 20.8 hr 8.4 x 10-2 I-134 52.5 min 2.4 x 10-1 I-135 6.585 hr 1.2 x 10-1 Based on noble gas release rate of 0.1 Ci/sec after 30 min.
NMP Unit 2 USAR Chapter 11 11.2-1 Rev. 25, October 2022 11.2 LIQUID WASTE MANAGEMENT SYSTEMS 11.2.1 Design Bases 11.2.1.1 Power Generation Design Bases The objective of the liquid waste management system (LWS) is to collect, monitor, and process for reuse or disposal all potentially-radioactive liquid wastes in a controlled manner that meets the objectives of 10CFR20 and 10CFR50 Appendix I, and at the same time does not limit operation of the plant.
The system provides for recycling of water for reuse in the plant. Sufficient treatment and diversity of types of equipment are available to process normal plant wastes to condensate quality. Discharge of processed liquid waste should be necessary only when the plant has a complete water inventory.
The system design also includes all items of reasonably demonstrated technology which, for a favorable cost-benefit ratio, effect reductions in the dose to the population reasonably expected to be within 50 mi of the reactor. The cost-benefit analysis is included in the report entitled Nine Mile Point Nuclear Station Unit 2 Compliance with 10CFR50 Appendix I (Appendix 11A).
The LWS has the capacity and capability of processing the anticipated quantities and activities of liquid wastes resulting from normal operation, maintenance, and anticipated operational occurrences. Cross-connections between the liquid waste subsystems provide additional flexibility for processing of wastes by alternate methods. These cross-connections are explained in Section 11.2.2 and illustrated on Figure 11.2-1.
The equipment and instrumentation selected for the LWS ensure that the radioactive concentrations resulting from liquid discharges from the plant are within the limits set forth in 10CFR20.
11.2.1.2 System Design Basis The LWS is designed to minimize operational radiation exposure to plant personnel. Equipment has been selected, arranged, and shielded to permit operation, inspection, and maintenance while keeping personnel radiation exposure within the limits of 10CFR20 and as low as reasonably achievable (ALARA) in accordance with Regulatory Guide (RG) 8.8.
Where possible, equipment has been selected that requires minimum maintenance and has been located in accessible areas.
Tanks, processing equipment, and associated piping that may contain significant quantities of radioactivity are shielded from personnel access areas and from controls or equipment requiring regular maintenance or operation. There are no
NMP Unit 2 USAR Chapter 11 11.2-2 Rev. 25, October 2022 outdoor tanks containing potentially-radioactive liquid.
Technical Requirements Manual (TRM) Section 3.7.7 provides requirements for unprotected outdoor tanks, should any be installed.
In some cases, several storage tanks are located in a common shielded area, since no regularly-scheduled maintenance is anticipated within the shielded tank compartments. All regularly-maintained equipment (pumps, valves, etc.) associated with these tanks is located outside the tank compartment but in other shielded and more accessible areas. In all cases, excess tank capacity is provided so that if an operational problem occurs within a compartment, the tank can be isolated without loss of unit operating capability.
To minimize maintenance, operational problems, and personnel exposure, the following items have been incorporated into the system design:
- 1.
Air and condensate connections are provided at various points in all subsystems to facilitate flushing of piping and components.
- 2.
Piping subsystems that can expect a high solids content (i.e., phase separator and evaporator bottoms piping) have been given special consideration in piping arrangement details as follows:
- a.
Where possible, laterals and five-diameter bends are used in place of tees and elbows.
- b.
Butt-welded fittings are used in place of socket-welded fittings.
- c.
All butt welds are made using consumable inserts.
- d.
Piping and components have been designed to minimize pockets and crevices that could create crud traps.
- e.
Low-point drains have not been used in slurry piping runs.
- f.
Pumping capacity and pipe sizes have been selected to ensure adequate flow velocity to prevent sedimentation.
- g.
Instrumentation connections have been located in a manner to prevent crud traps. Indicators are in accessible areas.
- 3.
Radiation and corrosion-resistant materials are used to reduce the need for component replacement over the 40-yr plant life.
NMP Unit 2 USAR Chapter 11 11.2-3 Rev. 25, October 2022
- 4.
Plug valves are used extensively to minimize valve stem leakage.
- 5.
Equipment is employed which has proven reliability in radwaste service as well as other process industries.
- 6.
Redundancy in components as well as processing flexibility between subsystems is provided to permit continued operation during periods of equipment repair. In addition, this redundancy permits appropriate scheduling of maintenance activities to minimize personnel exposure.
- 7.
Pump seal leakage is minimized by the use of mechanical seals for most of the radwaste pumps.
Double mechanical seals are used in pumps that handle slurry. Seal life is extended by the continuous supply of cool, clean seal water provided by the radwaste seal water system, thereby minimizing pump maintenance. Seal water pressure also is kept higher than the stuffing box pressure so that process fluid cannot leak out from the pumps.
The power supply to the system components is provided from non-Class 1E power sources.
Table 11.2-1 lists the equipment design data of each component in the LWS. Surge capacity is provided to cover contingencies such as processing equipment outages, back-to-back refueling outages, or abnormal evolutions resulting in the production of excessive waste volumes. Tank volume provided is in excess of that required by NUREG-0016 Revision 1, Table 1-4. See Section 11.2.2 for a description of holdup volumes and processing capacities.
NUREG-0016 Revision 1 provides guidance in calculational methods associated with liquid radwaste systems. The liquid radioactive waste tank holdup capacity is in excess of that required based on the daily expected average input flows from Table 1-4 of this NUREG (Section 11.2.2.9).
Decontamination factors used in analysis are consistent with Table 1-5 of this NUREG (Section 11.2.2.5).
11.2.1.2.1 Applicable Codes and Standards The LWS is a nonnuclear safety, non-Category I system. Table 11.2-2 lists the applicable codes and standards for system equipment. The safety class and material selections for equipment in the LWS satisfy the criteria of RG 1.143 (Section 1.8).
The atmospheric storage tanks are filament-wound fiberglass reinforced plastic tanks. They are designed to meet or exceed
NMP Unit 2 USAR Chapter 11 11.2-4 Rev. 25, October 2022 National Bureau of Standards Voluntary Product Standard PS-15-69 and the American Society of Testing and Materials Specification No. ASTM D3299-74.
11.2.1.2.2 Structural Design The radwaste equipment is housed in the radwaste building (Figures 1.2-13 and 1.2-14). Section 3.8 describes the building structural design criteria.
11.2.2 System Description The LWS is divided into the following four subsystems: the waste collector subsystem, the floor drain collector subsystem, the regenerant waste subsystem, and the phase separator subsystem. These subsystems permit wastes from various sources to be combined according to similarity of conductivity and isotopic concentrations for appropriate processing. Flow paths and equipment are shown on Figure 11.2-1. There is no provision for laundry waste processing at Unit 2. Laundry facilities at Nine Mile Point Nuclear Station - Unit 1 (Unit 1) are shared by Nine Mile Point Nuclear Station - Unit 2 (Unit 2) for the decontamination of radiation protection apparel and breathing apparatus. Laundry waste effluent will be treated by the radwaste treatment system at Unit 1. Laundry services may be provided by outside vendors.
11.2.2.1 Waste Collector Subsystem The waste collector subsystem collects, monitors, and processes for reuse or disposal relatively low-conductivity waste (less than 50 umho/cm) of variable radioactivity. This subsystem removes radioactivity from the liquid via filtration and ion exchange.
The subsystem normally receives input from the following sources:
- 1.
Draindown from the spent fuel pool cooling and cleanup (SFC) system.
- 2.
Draindown from the reactor water cleanup (RWCU) system.
- 3.
Distillate from the waste evaporator.
- 4.
Distillate from the regenerant evaporator.
- 5.
Radwaste demineralizer effluent.
- 6.
Reactor building equipment drains.
- 7.
Decant from the phase separator tanks.
NMP Unit 2 USAR Chapter 11 11.2-5 Rev. 25, October 2022
- 8.
Turbine building equipment drains.
- 9.
Draindown from the residual heat removal (RHR) system.
- 10. Waste collector surge contents.
- 11. Condensate demineralizer (CND) backwash flush water.
In addition, the subsystems can receive input from the following sources:
- 1.
Floor drain filter effluent.
- 2.
Floor drain collector surge tank.
- 3.
Recovery sample tank recycle.
The waste collector subsystem consists of three waste collector tanks which receive the above inputs, and three waste collector pumps. In addition, a waste collector surge tank and two associated waste collector surge pumps are provided. The surge tank provides a surge capacity in excess of that of the waste collector tanks.
The processing equipment in this subsystem consists of two etched disc-type radwaste filters for removal of insoluble particles, and two mixed bed-type demineralizers for the removal of soluble and colloidal ionic material.
The waste collector subsystem has two recovery sample tanks that receive the effluent from the waste collector process. They are used for holding and testing the processed water before reuse or disposal via the recovery sample pumps.
The waste collector subsystem influents previously listed are collected and directed primarily to the waste collector tanks via a common header. In the event that additional storage capacity is required, influents can be routed to the waste collector surge tank. However, the surge tanks are normally reserved to receive overflows from the main tanks, rather than direct inputs from other systems. Inputs to the waste collector surge and floor drain collector surge tanks are run separately.
All tanks are individually isolated from the collection header by air-operated valves (AOV). These valves permit tank isolation for processing and/or influent quality segregation.
Overflows from the three waste collector tanks are tied together and are gravity fed to the waste collector surge tank. This provides for direct collection of tank overflows and minimizes the potential for spillage. Overflow from the waste collector surge tank is directed to the floor of its compartment. The potential for surge tank overflow is extremely remote and would result from failure of high-level alarms in both the normal on-line collector tanks and the surge tanks. Alarms are
NMP Unit 2 USAR Chapter 11 11.2-6 Rev. 25, October 2022 provided in the radwaste control room where the Operators can identify and isolate inputs to the tanks as necessary. The tank level detection is not interlocked to shut the waste inflows into the surge tanks. However, an overflow from the surge tank would indicate an excessive influx of liquid into the system which could possibly exceed the collection capacity of the radwaste building floor and equipment drains. Therefore, overflow from the surge tank is directed to the floor of the compartment where it is totally contained. This arrangement prevents the uncontrolled release of radioactive liquid to other sections of the radwaste building and, therefore, meets the intent of RG 1.143. The floor is sloped to a local small sump.
A portable, submersible pump and hose would be used to transfer any overflow to the floor drain system and from there back to the liquid waste system. The steel liner prohibits any leakage to the environs.
The contents of the waste collector surge tank can be transferred back to the waste collector tanks, to the radwaste filter, or to the floor drain filter via the waste collector surge pumps. The recirculation lines from the waste collector surge pumps have connections for the addition of acid or caustic to neutralize the contents of the tank as necessary.
The waste collector pumps can take suction from any waste collector tank. The pumps, including the surge pumps, can provide recirculation to each respective tank through their mixing spargers. The spargers mix the tank contents to maintain a uniform volume prior to sampling and processing.
The waste collector pumps can transfer the contents of any of the waste collector tanks to the waste collector surge tank, the etched disc radwaste filters, or the Thermex system (Section 11.2.2.2). In no case can the contents of the waste collector tanks bypass a filtration process.
The etched disc filters are 100-percent capacity units. One is in operation while the other is in standby. The filters can be set to backwash either on high differential pressure or by manual initiation. When backwashing is required the standby filter is put into service and the operating filter is isolated.
This isolated filter is then backwashed by high-pressure air from the air receiver tank. The filter backwash is sent to the radwaste filter backwash tank. After several discharges, when the filter backwash tank is full, its contents are transferred, via the radwaste filter backwash pumps, to the spent resin tank (Section 11.2.2.4), the solid waste sludge tank, or the regenerant waste tank. The radwaste filter backwash pump can also provide recirculation to the tank to mix its contents. The effluent from the radwaste filters can be sampled and directed to the waste discharge sample tanks, for discharge to the service water system (SWP) discharge bay, the floor drain collector tank or the waste collector surge tank for
NMP Unit 2 USAR Chapter 11 11.2-7 Rev. 25, October 2022 reprocessing, or the radwaste demineralizers for further processing, depending on its characteristics.
The radwaste demineralizers are 100-percent capacity, mixed, deep bed types. Each can be operated singly, with the other in standby or in parallel, depending on pressure drop through the system. The demineralizers are charged with resin from the condensate regeneration subsystem (Section 10.4.6). The resins are of identical type, volumes and anion/cation ratio as those used in the CNDs. Therefore, the condensate regeneration subsystem is designed to handle, clean and regenerate the condensate as well as the radwaste demineralizer system. When the resin is depleted or the demineralizers have a high differential pressure, the resin is transferred back to the condensate regeneration subsystem for regeneration, or to the spent resin tank for further processing and disposal.
The radwaste demineralizer effluent can be sampled and then sent through a backwashable strainer. The strainers prevent resin from being introduced to other parts of the system in the event of a failure of the demineralizer resin retention system.
Demineralizer effluent can then be sent to one of three places:
to the waste discharge sample tank for sampling and disposal, to the waste collector tanks for reprocessing, or to the recovery sample tanks for recycling.
In addition to receiving input from the radwaste demineralizers, the recovery sample tanks can receive input from the waste sample tanks of Unit 1 (intertie piping is cut and capped to isolate units until inter-unit service is required). The inlets of the recovery sample tanks can be individually isolated so that one tank can receive input while the other transfers its contents. Overflow from the recovery sample tank is directed to the waste collector surge tank. The contents of a recovery sample tank are mixed, via the recovery sample pump recirculation line and tank sparger, to provide a uniform tank mixture, and then sampled to determine water quality. Depending on water quality and plant water inventory, the contents of the recovery sample tanks can be sent to one of the following places: the waste collector tanks for reprocessing, the condensate storage tanks (CST) for reuse, the waste sample tanks of Unit 1 (under strict administrative control) for use in that plant's water inventory (intertie piping is cut and capped to isolate units until inter-unit service is required), or through the radwaste discharge radiation monitors for release to the SWP discharge bay.
11.2.2.2 Floor Drain Collector Subsystem The floor drain collector subsystem collects, monitors, and processes for reuse or disposal potentially-high conductivity (greater than 50 umho/cm) waste of variable radioactivity. This subsystem normally receives input from the following sources:
NMP Unit 2 USAR Chapter 11 11.2-8 Rev. 25, October 2022
- 1.
Radwaste building floor drains.
- 2.
Reactor building floor drains (including the drywell floor drain tank).
- 3.
Auxiliary boiler area floor drains.
- 4.
CST area floor drains.
- 5.
Main stack area floor drains.
- 6.
Effluent from the Thermex system.
- 7.
Floor drain collector surge tank.
- 8.
Decant from the spent resin tank.
- 9.
Turbine building floor drains.
The system can receive inputs from the following sources:
- 1.
Distillate from the waste evaporator.
- 2.
Distillate from the regenerant evaporator.
- 3.
Effluent from the radwaste filters.
- 4.
Water from the waste discharge sample tanks.
- 5.
Waste collector surge tank.
- 6.
Floor drain collector tank at Unit 1 (intertie piping is cut and capped to isolate units until inter-unit service is required).
- 7.
CND backwash flush water.
The floor drain collector subsystem includes two floor drain collector tanks which receive the above inputs, and two floor drain collector pumps. In addition, a floor drain collector surge tank and two associated floor drain collector surge pumps are provided. This tank provides a surge capacity in excess of that of the floor drain collector tanks.
The processing equipment in the subsystem consists of the Thermex system and a forced circulation-type evaporator for the concentration of soluble and insoluble waste.
Finally, the floor drain collector subsystem has two waste discharge sample tanks that receive the effluent from the floor drain collector processes. They are used for holding and testing the processed water before reuse or disposal via the waste discharge sample pumps.
NMP Unit 2 USAR Chapter 11 11.2-9 Rev. 25, October 2022 The floor drain collector subsystem inputs are collected and directed to the floor drain collector and floor drain collector surge tanks via a common header. The tanks are individually isolated from the collection header by AOVs. These valves permit tank isolation for processing, influent quality segregation, or maintenance.
Overflows from the floor drain collector tanks are tied together and are gravity fed to the floor drain collector surge tank.
Overflow from the floor drain collector surge tank is directed to the floor of its compartment. The floor is sloped to permit the use of a portable, submersible pump to remove any overflow.
The arrangement is the same as that for the waste collector surge tank (see Section 11.2.2.1). The steel liner prohibits any leakage to the environs.
The contents of the floor drain collector surge tank can be transferred back to the floor drain collector tanks, the waste or regenerant evaporators, the radwaste filters, or via the floor drain collector surge pumps. The floor drain collection surge pump recirculation lines have connections for the addition of acid or caustic to neutralize the contents of the tank as necessary.
The floor drain collector pumps can take suction from either floor drain collection tank. The pumps (including the surge pumps) provide recirculation to their tanks through their mixing spargers. The spargers mix the tank contents to ensure a uniform volume prior to sampling and processing.
The floor drain collector pumps can transfer the contents of either of the floor drain collector tanks to the following, depending on waste quality, radioactivity level, and equipment availability: the waste or regenerant evaporators, the radwaste filters, or the floor drain surge tank. In no case can the contents of the floor drain collection tanks bypass a filtration or evaporation process.
The Thermex system is a modular-designed wastewater treatment system utilizing reverse osmosis, ion exchange electrodionization and ultraviolet photo degradation methodologies. The modular design allows for simple setup and maintenance in addition to a high degree of operational flexibility. The Thermex system separates the dissolved and suspended solids from recycled wastewater. The dissolved and suspended solids waste products are concentrated, treated, and disposed of by an offsite vendor, or processed through 2LWS-TK7.
The purified water from the Thermex effluent is returned to the plant systems via the floor drain filter effluent tank, and is transferred via the floor drain filter effluent pump to one of the following tanks, depending on the effluent quality: the waste discharge sample tanks for disposal, the floor drain collector tanks, the waste collector tanks, or the regenerant waste tanks for further processing.
NMP Unit 2 USAR Chapter 11 11.2-10 Rev. 25, October 2022 The waste evaporator is a forced circulation design with a reboiler providing process heat and an overhead entrainment separator and distillation column that minimizes liquid droplets, particulates, and volatiles in the vapor. The evaporator recirculation line has connections for the addition of acid or caustic to neutralize the contents of the evaporator as necessary. Concentrated evaporator bottoms are sent to the evaporator bottoms tank via the evaporator bottoms pumps. When the evaporator bottoms tank is full, its contents are transferred to the radioactive solid waste system for processing and disposal (Section 11.4). During operation, the evaporator overhead is condensed and subcooled. If the quality of the distillate is satisfactory, it is directed to the waste collector tanks to await demineralization. If the distillate quality is unacceptable, it can be returned to the floor drain collector tanks for reprocessing. If the distillate is to be discharged to the environment, it is directed to the waste discharge sample tanks.
The waste discharge sample tank can receive input from both the waste evaporator and the regenerant evaporator as well as the radwaste filters, the radwaste demineralizers, or the floor drain filter. The inlets to the waste discharge sample tank can be individually isolated so that one tank can receive input while the other transfers its contents. Overflow from these tanks is directed to the floor drain collector surge tank. The contents of the waste discharge sample tanks are mixed via the waste discharge sample pump recirculation line and the tank sparger to provide a uniform tank volume prior to sampling. If the results of sample analysis indicate that further processing or reprocessing is required, the contents of the waste discharge sample tank can be directed to the waste or regenerant evaporators or the floor drain collection tanks. If the results of the sample analysis indicate that no further processing is required, the tank contents can be discharged to the discharge bay. The final discharge is again checked by a radiation monitor. If predetermined activity levels are exceeded, the flow is automatically halted. The radiation monitor is located sufficiently upstream of the discharge valve to prevent any unacceptable release.
11.2.2.3 Regenerant Waste Subsystem The main purpose of the regenerant waste subsystem is to collect, monitor, and process chemical solutions resulting from the acid/caustic regeneration of CND resins (conductivities greater than 1,000 umho/cm). This subsystem can also collect input from those radwaste building floor drains with a high anticipated chemical content, effluent from the Thermex system, and backwash from the radwaste filters for further processing.
The regenerant waste evaporator is identical to the waste evaporator in all respects.
NMP Unit 2 USAR Chapter 11 11.2-11 Rev. 25, October 2022 The regenerant waste subsystem includes two regenerant waste tanks, which collect the inputs described above, and two regenerant waste pumps.
The processing equipment of this subsystem consists of a forced circulation-type evaporator for the concentration of radioactive sodium sulfate. This evaporator is identical to the waste evaporator described in Section 11.2.2.2. This includes its sources of influent and the possible discharge paths of its effluent.
This subsystem also shares the evaporator bottoms tank and pumps and the waste discharge sample tanks and pumps of the floor drain collector subsystem. The regenerant waste subsystem inputs are collected and directed to the regenerant waste tanks via a common header. The tanks are individually isolated from the collection header by AOVs. These valves permit tank isolation for processing influent quality segregation or maintenance.
The regenerant waste pumps can take suction from either regenerant waste tank. The pumps also provide recirculation to either tank through their mixing spargers. The spargers mix the tanks' contents to maintain a uniform volume prior to sampling and processing. The pump recirculation lines have acid and caustic addition connections to permit neutralization of the tank contents as necessary.
The regenerant waste pumps can transfer the contents of the regenerant waste tanks to any of the following components, depending on waste quality and equipment availability: the floor drain filter, the waste evaporator, or the regenerant evaporator. The description of the process from this point on through reuse, reprocessing, or disposal is identical to that of the floor drain collector subsystem (Section 11.2.2.2).
11.2.2.4 Phase Separator Subsystem The phase separator subsystem collects, decants, and holds for radioactive decay, the backwash from the RWCU filter/demineralizers, the SFC filter/demineralizers, the radwaste filters, and spent resins from the condensate and radwaste demineralizers.
The phase separator subsystem consists of three phase separator tanks, three associated phase separator pumps, one spent resin tank, and one associated spent resin pump.
Two of the phase separator tanks receive input from the RWCU filter/demineralizers. The third phase separator tank receives input from the SFC filter/demineralizer and can also receive the contents of the RWCU phase separator tank via the RWCU phase separator pumps. Each tank is capable of being isolated by an
NMP Unit 2 USAR Chapter 11 11.2-12 Rev. 25, October 2022 AOV. These valves permit tank isolation for holdup or maintenance.
The overflows from the RWCU phase separator tanks are tied together and are gravity fed to the reactor building floor drains. The overflow from the SFC phase separator is also gravity fed to the reactor building floor drains.
The RWCU phase separator pumps can take suction from either RWCU phase separator tank and can also provide recirculation to either tank through their mixing eductors. The eductors mix the tanks' contents to maintain a homogeneous mixture prior to transferring their contents. The SFC phase separator and the spent resin tank, although separately piped, have the same features.
After the phase separators have received a backwash, the relatively clean water is drawn off through decant screens by the associated pump and transferred to the waste collector tanks. When the phase separator tanks are filled to their solids capacity, the contents are transferred, via their respective pumps, to the spent resin tank for holdup prior to disposal in the solid waste system.
In addition to receiving input from the phase separator tanks, the spent resin tank receives input from the radwaste filters and demineralizers, decant from the solid waste system sludge tank, and CNDs.
The spent resin tank can decant its contents to the floor drain collector tanks. This decant is gravity drained or pumped.
Overflow from this tank is directed to the radwaste building floor and equipment drains. Spent resin tank contents are transferred, via the spent resin pump, to the radioactive solid waste system for further processing and disposal.
11.2.2.5 System Operational Analysis The flow rate and the activity concentrations (fractions of primary coolant concentration) (Table 11.2-3) were developed using a material balance calculation and data from NUREG-0016 Revision 1.
Decontamination factors of the processing units are shown in Table 11.2-4.
11.2.2.6 Instrumentation and Control The radwaste process computer receives input of system conditions throughout the LWS and provides indication of process operation, equipment performance, system status, and central control of process equipment via computer terminals and video displays. The radwaste control room is located in the turbine
NMP Unit 2 USAR Chapter 11 11.2-13 Rev. 25, October 2022 building, adjacent to the south side of the radwaste building at el 279 ft.
Fundamentally, the components of the waste collector, floor drain collector, regenerant waste, and phase separator subsystems employ similar basic components such as tanks, pumps, filters, demineralizers, and evaporators (in two of the subsystems). Like components have similar instrumentation.
Instrumentation for the different components is described in the following sections.
11.2.2.6.1 Tanks In multiple tank subsystems (e.g., waste collector tanks 1A, 1B, and 1C), the inlet valves to each tank are interlocked in such a way that one is always open so that one tank is available to accept influents.
Liquid radwaste tanks have level transmitters which provide a trouble alarm when a high-level setpoint is reached.
11.2.2.6.2 Pumps The liquid radwaste process pumps (e.g., waste collector pumps 1A, 1B, and 1C) rely on tank level transmitters and the radwaste computer to shut down the pump on low tank levels.
The discharge side of the pumps has the following instrumentation: pressure, temperature, and conductivity. The suction side of the pumps has pH instrumentation.
Pumps with double mechanical seals in the radwaste building are supplied with recirculating, filtered, and cooled seal water.
Radwaste pumps in the reactor building are supplied with filtered once-through seal water. Instrumentation has been provided to regulate the seal water flow, to monitor the condition of the seals via pressure retention, and to trip pumps if the seal water flow is interrupted. Flow regulation and monitoring instruments are located in a general access area to minimize radiation exposure to personnel. The seal water system is shown on Figure 11.2-1.
11.2.2.6.3 Radwaste Filters The radwaste filters have differential pressure transmitters that monitor the pressure drop across the filter and automatically initiate the backwash cycle when a high differential pressure setpoint is reached.
The radwaste filters utilize time delays for the backwash cycle sequence of operation. This permits the filters to be put back into service.
11.2.2.6.4 Radwaste Demineralizers and Resin Traps
NMP Unit 2 USAR Chapter 11 11.2-14 Rev. 25, October 2022 Each radwaste demineralizer has a differential pressure transmitter to monitor the pressure drop across the demineralizer, and a conductivity element in the demineralizer outlet line to monitor effluent conductivity. The demineralizers can be backwashed, regenerated, or ultrasonically cleaned due to either high differential pressure or high effluent conductivity.
These operations will be performed utilizing the equipment in the CND system.
Influent to the demineralizers is maintained by a flow element and transmitter that controls a flow control valve (FCV).
The resin traps on the discharge side of the demineralizers also have differential pressure instruments. Radwaste Operators monitor demineralizer strainer differential pressure on a weekly basis for evidence of demineralizer retention screen resin leakage.
11.2.2.6.5 Floor Drain Filter (Retired, Removed) Associated Tanks and Pumps The tanks associated with the floor drain filter (retired, removed) (body feed, precoat, and effluent) have level switches to shut down their associated pumps on low level and alarm on high level. The pumps have discharge pressure indication.
11.2.2.6.6 Evaporators (Waste and Regenerant)
The evaporators have level transmitters to control the inlet feed flow and to provide trouble alarms.
Pressure, temperature, and liquid level are measured at various points throughout the evaporator system to monitor performance and initiate the various modes of operation.
The conductivity of the effluent distillate is monitored to determine the required flow path (i.e., further processing, disposal, or reuse).
11.2.2.6.7 Diaphragm Seals Instruments that measure parameters (pressure or level) of liquids with high radioactivity and solids content have diaphragm seals. These seals physically separate the radioactive liquid from a signal transmitting fluid located in a shop-sealed capillary tube including the sensor. This reduces personnel exposure to radiation because it keeps the radioactive liquid within a shielded equipment cubicle while the instrument racks, located in general access areas, contain only nonradioactive transmitting fluid. These diaphragm seals also
NMP Unit 2 USAR Chapter 11 11.2-15 Rev. 25, October 2022 reduce the possibility of instrument lines plugging with radioactive crud.
11.2.2.7 System Operation The operating procedures used for all liquid radwaste equipment are based on batch processing through radwaste systems. This type of operation allows time to sample and check the feed and effluent streams before and after each process step to prevent the inadvertent discharge of waste having a radioactivity level above the predetermined limit.
It also allows for maximum reuse of water that has been processed through the radwaste system.
As previously noted, the influents to the LWS can vary widely in activity, conductivity, and volume, depending on the operating conditions of the plant. A description of the conditions that may be used for guidance for the disposition of batches is as follows:
- 1.
High conductivity: evaporate, demineralize, and recover.
- 2.
Low gross activity and high conductivity: filter, evaporate, demineralize, and recover.
- 3.
High gross activity and low conductivity: filter, demineralize, and recover.
- 4.
High activity and high conductivity: filter, evaporate, demineralize, and recover.
11.2.2.8 Performance Tests Performance tests that monitor the removal of insolubles and organics by filters, reduction of conductivity by demineralizers, and the removal of solubles by the evaporators are conducted on a periodic basis. The radioactivity of the input and output streams of each piece of equipment is checked periodically. Performance tests verify the decontamination factors and other aspects including:
- 1.
Filters:
- a.
Insoluble material removal efficiency.
- b.
Measurement of influent and effluent activity.
- 2.
Demineralizers:
- a.
Conductivity reduction factor (i.e.,
demineralizer efficiency).
NMP Unit 2 USAR Chapter 11 11.2-16 Rev. 25, October 2022
- b.
Measurement of influent and effluent activity.
- 3.
Evaporators:
- a.
Insoluble material removal efficiency.
- b.
Measurement of influent and effluent activity.
- c.
Distillate conductivity.
Overall system tests for the waste collector subsystem, the floor drain collector subsystem, and the regenerant waste subsystem establish the overall decontamination factor for the entire subsystem.
11.2.2.9 Summary Over 225,000 gal of bulk storage capacity is available to collect and store the average liquid radwaste influent volumes of approximately 41,000 gpd. Storage facilities are also available to receive treated liquid. The treated liquid bulk storage capacity is 100,000 gal which allows sufficient time for sampling and disposal of approximately 2,000 gpd to the environs and recycling approximately 39,000 gpd tank.
11.2.3 Radioactive Release and Doses Table 11.2-5 is a tabulation of the expected annual liquid releases. The release estimates corresponding to EPU conditions from the liquid effluent stream in Ci/yr per nuclide are given in Table 11.2-6, and correspond to operation with design-failed fuel conditions as discussed in Section 11.1. The release is based on effluent discharge to the environs only through the recovery sample tanks and/or waste discharge sample tanks.
Tritium release from liquid pathways is anticipated at 55 Ci/yr.
11.2.3.1 Release Points All liquid releases from Unit 2 are fed into the SWP discharge bay that is directed to Lake Ontario. Figure 11.5-8 shows all systems that feed the discharge bay. Figure 11.2-1 shows the release point of each system to the discharge line. Figure 1.2-29 shows the physical location of the discharge into the bottom of the lake.
11.2.3.2 Dilution Factors The dilution factor used in evaluating the release of radioactive liquid effluents is that provided by the cooling tower blowdown and service water discharge.
NMP Unit 2 USAR Chapter 11 11.2-17 Rev. 25, October 2022 Treated radioactive effluents are diluted in the discharge bay with a combined plant discharge (cooling tower blowdown and service water discharge) flow of 30,428 gpm (6.05 x 1013 cc/yr).
Prior to discharging any LWS effluents to the lake, the waste discharge sample tank and/or the recovery sample tank contents are recirculated (to provide a homogeneous mixture) and sampled.
The analysis of this sample is used to determine the discharge rate and amount of dilution needed for a release within the allowable limits.
To prevent any accidental discharge of effluents greater than the allowable limits, an isolation valve in the LWS discharge will close upon receipt of a high radiation signal. Once the effluent stream has been isolated and sampled, it may be recirculated through the LWS to reduce the activity concentration.
11.2.3.3 Estimated Doses A summary of the estimated annual radiation doses is presented in Appendix 11A, which shows that the estimated annual doses from liquid releases are below the dose criteria set forth in 10CFR50 Appendix I.
The design liquid release isotopic radioactive concentration estimates (corresponding to EPU conditions) in the unrestricted area are shown in Table 11.2-6, along with the indication that these concentrations are fractions of Maximum Permissible Concentration (MPC) limits given in 10CFR20 Appendix B, Table II, Column 2.
NMP Unit 2 USAR Chapter 11 11.2-18 Rev. 25, October 2022 TABLE 11.2-1 EQUIPMENT DESCRIPTIONS LIQUID WASTE MANAGEMENT SYSTEM A. Tanks Item No.
2LWS-TK-Name Capacity Each (gal)
Design Material Quantity Temperature
(°F)
Pressure (psig) 1A,B,C Waste collector 25,000 150 Atmosphere Fiberglass 3
2A,B Floor drain collector 25,000 150 Atmosphere Fiberglass 2
3A,B Regenerant waste 25,000 150 Atmosphere Fiberglass 2
4A,B Recovery sample 25,000 150 Atmosphere Fiberglass 2
5A,B Waste discharge sample 25,000 150 Atmosphere Fiberglass 2
6A,B RWCU phase separator 8,200 150 15 Stainless steel 2
7 Spent resin 7,000 150 15 Stainless steel 1
8 Waste evaporator distillate 470 350 100 Stainless steel 1
9 Regenerant evaporator distillate 470 350 100 Stainless steel 1
10 CFS phase separator/Evaporator bottoms 8,000 350 25 Incoloy 1
12 Floor drain filter body feed 150 104 Atmosphere Carbon steel plastic lined 1
14 Floor drain filter effluent 200 104 Atmosphere Carbon steel plastic lined 1
15 Air receiver 60 250 350 Stainless steel 1
16 Radwaste filter backwash 1,100 150 150 Stainless steel 1
17 Floor drain collector surge 29,800 150 Atmosphere Fiberglass 1
18 Waste collector surge 29,800 150 Atmosphere Fiberglass 1
20 Floor drain filter precoat 1,000 104 Atmosphere Carbon steel plastic lined 1
30 SFC phase separator 8,200 150 15 Stainless steel 1
NMP Unit 2 USAR Chapter 11 11.2-19 Rev. 25, October 2022 TABLE 11.2-1 (Cont'd.)
B. Pumps Item No.
2LWS-P-Name Capacity (gpm)
TDH (ft)
Design Pressure (psig)
Quantity 1A,1B,1C Waste collector 220 391 275 3
2A,2B Floor drain collector 110 312 275 2
3A,3B Regenerant waste 110 330 275 2
4A,4B Recovery sample 110 211 275 2
5A,5B Waste discharge sample 165 171 275 2
6A,6B RWCU phase separator 110 238 275 2
7 Spent resin 66 49 275 1
8 Waste evaporator distillate transfer 30 22 255 1
9 Regenerant evaporator distillate transfer 30 22 225 1
10A,10B Evaporator bottoms 60 105 225 2
13 Waste evaporator recirculation 4,000 98 80 1
11 Regenerant evaporator recirculation 4,000 98 80 1
15 Floor drain filter effluent 95 140 125 1
16A,16B Radwaste filter backwash 110 53 275 2
17A,17B Floor drain collector surge 110 331 275 2
18A,18B Waste collector surge 220 431 275 2
20 Caustic addition 1
110 1
21 Acid addition 1
110 1
27 Floor drain filter precoat 200 55 125 1
28 Floor drain filter body feed 61.8 508 165 1
30 SFC phase separator 110 295 275 1
NMP Unit 2 USAR Chapter 11 11.2-20 Rev. 25, October 2022 TABLE 11.2-1 (Cont'd.)
C. Miscellaneous Heat Exchangers Item No.
2LWS-E-Name Duty Btu/Hr x 106 Design Temperature
(°F)
Shell/Tube Design Pressure (psig)
Shell/Tube Material Shell/Tube Quantity 1
Waste evaporator condenser 12.99 300/150 Full vacuum 100/200 Stainless steel/
stainless steel 1
2 Waste evaporator distillate cooler 1.76 150/300 150/Full vacuum and 100 Stainless steel/
carbon steel 1
4 Waste evaporator reboiler 15.65 350/350 Full vacuum and 100 (both)
Incoloy/
carbon steel 1
5 Regenerant evaporator condenser 12.99 300/150 Full vacuum and 100/200 Stainless steel/
stainless steel 1
6 Regenerant evaporator distillate cooler 1.76 150/300 200/Full vacuum and 100 Stainless steel/
carbon steel 1
7 Regenerant evaporator reboiler 15.65 350/350 Full vacuum and 100 (both)
Incoloy/
carbon steel 1
D. Demineralizers Item No.
2LWS-DEMN-Capacity (gpm)
Volume (ft3)
Design Type Material Quantity Temperature
(°F)
Pressure (psig) 4A,B 100 110/110 140 150 Mixed bed Rubber lined carbon steel 2
NMP Unit 2 USAR Chapter 11 11.2-21 Rev. 25, October 2022 TABLE 11.2-1 (Cont'd.)
E. Evaporators Item No.
2LWS-EV-Capacity (gpm)
Design Operating Batch Volume (gal)
Material Quantity Temperature
(°F)
Pressure (psig)
Temperature
(°F)
Pressure (psig) 1,2 30 350 100/full vacuum 250 15 3,450***
Stainless steel and Incoloy 2
F. Filters Item No.
2LWS-FLT-Name Design Capacity (gpm)
Flow Flux (gpm/ft2)
Material Quantity Temperature
(°F)
Pressure (psig) 1A,B Radwaste 150 350 200 6.7 Stainless steel 2
G. Strainers Item No.
2LWS-STR-Capacity (gpm)
Design Operating Material Quantity Temperature
(°F)
Pressure (psig)
Temperature
(°F)
Pressure (psig) 5A,B 100 140 150 120 105 Carbon steel 2
- Progressive cavity-type 80 psig maximum discharge pressure.
- Not applicable (positive displacement-type pump).
- Includes liquid in reboiler and connecting piping.
NMP Unit 2 USAR Chapter 11 11.2-22 Rev. 25, October 2022 TABLE 11.2-2 APPLICABLE CODES AND STANDARDS FOR LIQUID WASTE MANAGEMENT SYSTEMS Description Safety Class Code Code Class Earthquake Criteria Tornado Criteria Quality Assurance Category 1 Tanks (steel or alloy) 4 ASME VIII*
No No No Tanks (fiberglass) 4 NBS PS15-69 No No No Etched disc filter 4
ASME VIII No No No Demineralizers 4
ASME VIII No No No Heat exchangers and reboilers 4
ASME VIII No No No Evaporators 4
ASME VIII No No No Pumps 4
HIS No No No All except the RWCU phase separator tanks (2LWS-TK6A and 6B), which were not rehydrotested after field modification and, therefore, had their code stamps removed.
NMP Unit 2 USAR Chapter 11 11.2-23 Rev. 25, October 2022 TABLE 11.2-3 MATERIAL AND CONCENTRATION BALANCE FOR LIQUID WASTE MANAGEMENT SYSTEM Fraction of Average Primary Flow Coolant (gpd)
Activity (PCA)
Input Floor drains Reactor building 4,320 0.001 Radwaste building 1,000 0.001 Turbine building 2,500 0.001 Lab drains 500 0.02 Chemical lab waste 100 0.02 Floor drain subtotal 8,420 Equipment drains Drywell 3,400 1.00 Secondary containment and auxiliary bays 3,700 0.10 Radwaste building 1,100 0.10 Turbine building 3,000 0.001 Equipment drain subtotal 11,200 Other sources Ultrasonic resin cleaner 15,000 0.05 Resin rinse 3,150 0.002 Phase separator decant 1,050 0.002 Regenerants 1,725 Total 40,545
NMP Unit 2 USAR Chapter 11 11.2-24 Rev. 25, October 2022 TABLE 11.2-4 DECONTAMINATION FACTORS OF PROCESSING UNITS Equipment DF Remarks Radwaste and floor 1
For corrosion/activation drain collector products (insolubles) filters Demineralizers High-purity wastes 10 For Cs and Rb 100 For anions and other nuclides Low-purity wastes 2
For Cs and Rb 100 For anions and other nuclides Waste and regenerant 1,000 For anions evaporators 10,000 For all nuclides except anions
NMP Unit 2 USAR Chapter 11 11.2-25 Rev. 25, October 2022 TABLE 11.2-5 EXPECTED ANNUAL LIQUID RELEASES*
Activity Released Isotope (uCi/ml)
(Ci/yr)
Na-24 1.6E-18 9.5E-11 P-32 1.1E-12 6.3E-05 Cr-51 2.5E-10 1.5E-02 Mn-54 9.4E-12 5.5E-04 Mn-56 0.0 0.0 Fe-55 3.0E-10 1.9E-02 Fe-59 4.3E-12 2.5E-04 Co-58 1.8E-11 1.1E-03 Co-60 5.8E-11 3.5E-03 Ni-63 3.0E-13 1.9E-05 Ni-65 0.0 0.0 Cu-64 0.0 0.0 Zn-65 2.6E-11 1.6E-03 Zn-69 0.0 0.0 Sr-89 1.6E-11 9.5E-04 Sr-90 2.2E-12 1.3E-04 Sr-91 0.0 0.0 Sr-92 0.0 0.0 Y-91 8.6E-12 5.2E-04 Y-92 0.0 0.0 Y-93 0.0 0.0 Zr-95 1.4E-12 8.4E-05 Zr-97 0.0 0.0 Nb-95 8.8E-13 5.3E-05 Nb-98 0.0 0.0 Mo-99 2.1E-13 1.2E-05 Tc-99m 0.0 0.0 Tc-101 0.0 0.0 Tc-104 0.0 0.0 Ru-103 2.4E-12 1.4E-04 Ru-105 0.0 0.0 Ru-106 8.6E-13 5.2E-05 Ag-110m 2.7E-13 1.6E-05 Te-129m 4.2E-12 2.7E-04 Te-131m 1.5E-16 9.0E-09 Te-132 1.7E-15 1.0E-07 Ba-139 0.0 0.0 Ba-140 8.0E-12 4.8E-04
NMP Unit 2 USAR Chapter 11 11.2-26 Rev. 25, October 2022 TABLE 11.2-5 (Cont'd.)
Activity Released Isotope (uCi/ml)
(Ci/yr)
Ba-141 0.0 0.0 Ba-142 0.0 0.0 La-142 0.0 0.0 Ce-141 3.2E-12 1.9E-04 Ce-143 0.0 0.0 Ce-144 8.1E-13 4.9E-05 Pr-143 0.0 0.0 Nd-147 0.0 0.0 W-187 6.3E-17 3.8E-09 Np-239 4.2E-13 2.5E-05 Br-83 0.0 0.0 Br-84 0.0 0.0 Br-85 0.0 0.0 I-131 7.2E-11 4.3E-03 I-132 1.0E-15 5.8E-08 I-133 1.8E-14 1.1E-06 I-134 0.0 0.0 I-135 0.0 0.0 Rb-89 0.0 0.0 Cs-134 3.5E-10 2.3E-02 Cs-136 1.7E-11 1.0E-03 Cs-137 1.1E-09 6.5E-02 Cs-138 0.0 0.0 H-3 9.1E-07 5.5E+01 Anticipated operational occurrences 1.00E-01 Ci/yr added to release. Dilution release rate (gm/yr) 6.05E+13.
Total release (excluding tritium) is 1.4E-01. Total release (excluding tritium) is 2.3E-09 uCi/gm.
NOTE: 4.8E-18 = 4.8x10-18
NMP Unit 2 USAR Chapter 11 11.2-27 Rev. 25, October 2022 TABLE 11.2-6 PRE-OPERATIONAL DESIGN ANNUAL LIQUID RELEASE ESTIMATES Design Activity Released MPC(1)
Fraction Isotope uCi/ml Ci/yr (uCi/ml) of MPC(2)
Na-24 5.0E-18 3.0E-10 2.0E-4 2.5E-14 P-32 3.4E-12 2.0E-4 2.0E-5 1.7E-7 Cr-51 3.1E-10 1.9E-2 2.0E-3 1.6E-7 Mn-54 1.9E-11 1.1E-3 1.0E-4 1.9E-7 Mn-56 0.0 0.0 1.0E-4 0.0 Fe-55 1.9E-10 1.2E-2 8.0E-4 2.4E-7 Fe-59 2.0E-11 1.2E-3 6.0E-5 3.3E-7 Co-58 1.6E-9 9.7E-2 1.0E-4 1.6E-5 Co-60 2.6E-10 1.6E-2 5.0E-5 5.3E-6 Ni-63 1.9E-13 1.2E-5 3.0E-5 6.3E-9 Ni-65 0.0 0.0 1.0E-4 0.0 Cu-64 0.0 0.0 3.0E-4 0.0 Zn-65 3.3E-11 2.0E-3 1.0E-4 3.3E-7 Sr-89 3.0E-10 1.8E-2 3.0E-6 1.0E-4 Sr-90 4.4E-11 2.5E-3 3.0E-7 1.5E-4 Sr-91 0.0 0.0 7.0E-5 0.0 Sr-92 0.0 0.0 7.0E-5 0.0 Y-91 3.8E-11 2.3E-3 3.0E-5 1.3E-6 Y-92 0.0 0.0 6.0E-5 0.0 Y-93 0.0 0.0 3.0E-5 0.0 Zr-95 6.3E-12 3.9E-4 6.0E-5 1.1E-7 Zr-97 0.0 0.0 2.0E-5 0.0 Nb-95 4.0E-12 2.4E-4 1.0E-4 4.0E-8 Mo-99 1.4E-12 8.3E-5 2.0E-4 7.2E-9 Tc-99m 0.0 0.0 6.0E-3 0.0 Ru-103 1.1E-11 6.7E-4 8.0E-5 1.4E-7 Ru-105 0.0 0.0 1.0E-4 0.0 Ru-106 3.8E-12 2.3E-4 1.0E-5 3.8E-7 Ag-110m 2.9E-11 1.7E-3 3.0E-5 9.6E-7 Te-129m 2.2E-11 1.4E-3 3.0E-5 7.3E-7 Te-131m 6.5E-16 4.0E-8 6.0E-5 1.1E-11 Te-132 1.5E-12 9.1E-5 3.0E-5 5.1E-8 Ba-140 1.1E-10 6.6E-3 3.0E-5 3.6E-6 Ce-141 1.4E-11 8.5E-4 9.0E-5 1.6E-7 Ce-143 0.0 0.0 4.0E-5 0.0 Ce-144 5.5E-12 3.4E-4 1.0E-5 5.5E-7 Pr-143 0.0 0.0 5.0E-5 0.0 Nd-147 0.0 0.0 6.0E-5 0.0
NMP Unit 2 USAR Chapter 11 11.2-28 Rev. 25, October 2022 TABLE 11.2-6 (Cont'd.)
Design Activity Released MPC(1)
Fraction Isotope uCi/ml Ci/yr (uCi/ml) of MPC(2)
W-187 1.1E-15 6.6E-8 7.0E-5 1.6E-11 Np-239 8.0E-12 4.8E-4 1.0E-4 8.0E-8 Br-83 0.0 0.0 3.0E-6 0.0 I-131 2.8E-10 1.7E-2 3.0E-7 9.4E-4 I-132 8.5E-15 4.8E-7 8.0E-6 1.1E-9 I-133 1.3E-13 8.3E-6 1.0E-6 1.3E-7 I-134 0.0 0.0 2.0E-5 0.0 I-135 0.0 0.0 4.0E-6 0.0 Cs-134 1.8E-9 1.1E-1 9.0E-6 2.0E-4 Cs-136 8.5E-11 5.1E-3 9.0E-5 9.4E-7 Cs-137 5.3E-9 3.2E-1 2.0E-5 2.7E-4 H-3(3) 9.1E-7 5.5E+1 3.0E-3 3.0E-4 Total(4) 1.1E-8 6.4E-1 1.7E-3 Dilution Release Rate = 6.05 x 1013 gm/yr (1)
MPC values are from 10CFR20 Appendix B, Table 11, Column
- 2.
(2)
Fraction of MPC = design activity released (uCi/ml)/MPC (uCi/ml).
(3)
Tritium release is in accordance with NUREG-0016 Rev. 1, 1/79, pp 1-8, Section 1.5.1.10.
(4)
All totals are excluding tritium.
NOTE: 4.8E-18 = 4.8x10-18
NMP Unit 2 USAR Chapter 11 11.3-1 Rev. 25, October 2022 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS This section describes the capabilities of Unit 2 to collect, control, process, store, and dispose of gaseous radioactive waste generated from normal operation and anticipated operational occurrences. The gaseous waste management systems include the offgas (OFG) system, standby gas treatment system (SGTS), and various building ventilation systems. The SGTS and building ventilation systems are discussed in Sections 6.5.1 and 9.4, respectively. Section 3.1 discusses compliance with the General Design Criteria (GDC).
11.3.1 Design Bases The design of the gaseous waste management systems meets the following criteria:
- 1.
The systems have the capability to meet the requirements of 10CFR20 and the dose design bases specified in Appendix I to 10CFR50, including provisions to treat gaseous radioactive waste in such a way that:
- a.
The calculated total annual quantity of all radioactive material released from Unit 2 to the atmosphere does not result in an estimated annual external dose from gaseous effluents to any individual in unrestricted areas beyond the site boundary in excess of 5 mRem to the whole body or 15 mRem to the skin.
- b.
The calculated total annual quantity of all radioactive iodine and radioactive material in particulate form released from Unit 2 in the effluents to the atmosphere does not result in an estimated annual dose for any individual in an unrestricted area beyond the site boundary in excess of 15 mRem to any organ.
- c.
The concentrations of radioactive materials in gaseous effluents released to an unrestricted area do not exceed the limits in 10CFR20 Appendix B, Table II, Column 1.
- 2.
The OFG system is designed to meet the anticipated processing requirements of the plant. Adequate capacity is provided to process gaseous wastes during periods when major processing equipment may be down for maintenance and during periods of greater than normal gaseous waste generation.
- 3.
In accordance with RG 1.29, the radioactive gaseous waste systems are classified as nonseismic. An analysis of a postulated waste gas tank failure was
NMP Unit 2 USAR Chapter 11 11.3-2 Rev. 25, October 2022 performed according to the guidance provided in RG 1.98. A description of this analysis and the results are given in Section 15.7.1. In accordance with RG 1.26, the OFG system is classified as nonnuclear safety, Quality Group D, and is designed to withstand the effects of a hydrogen detonation where the potential for an explosive mixture exists.
Instrumentation with automatic alarm is provided to monitor the concentrations of the hydrogen gas in those portions of the system having the potential for an explosive gas mixture.
- a.
Each hydrogen analyzer annunciates both on the local panel and in the control room. The control room annunciation is a trouble alarm.
- b.
The hydrogen analyzers are of the nonsparking thermo-conductivity type. The analysis is off line. Gas is extracted from the main stream through instrument piping, analyzed, and returned to the main stream. Because the analyzers are off line, a hydrogen explosion will not affect the analyzers, although they are not explosion proof.
- c.
See Figures 11.3-1a, 11.3-1b, and 11.3-1c.
- d.
See Section 11.3.2.1 for a discussion of protective features.
(The description of hydrogen detonation design criteria is vendor proprietary and was submitted under separate Proprietary cover in response to Nuclear Regulatory Commission (NRC)
Question F460.12.)
The design pressure due to hydrogen detonation is as follows:
psig All tanks and vessels 350 Piping:
Air ejector to preheater 575 Preheater to recombiner 260 Recombiner to condenser 260 Condenser to discharge manifold 575 Vacuum pump 200
- 4.
The system design contains provisions to control leakage and to facilitate operation and maintenance in accordance with the guidelines of RG 1.143.
NMP Unit 2 USAR Chapter 11 11.3-3 Rev. 25, October 2022 Table 11.3-1 gives the expected annual quantities of radioactive gaseous effluent from all sources (by radionuclide) in Ci/yr.
Table 11.3-2 lists the data used in calculating the annual releases of radioactive gaseous effluents. Table 11.3-3 shows the design annual radioactive gaseous releases versus the maximum permissible concentration (MPC). Figure 11.3-1 shows the piping and instrumentation diagram (P&ID) for the OFG system.
Appendix 11A, in compliance with 10CFR50 Appendix I, shows that the Unit 2 systems contain all items of reasonably demonstrated technology that will effect a reduction in dose to the population reasonably expected to be within a 50-mi radius of the plant. The OFG system is located in the turbine building.
This portion of the turbine building is designed to meet Category I requirements. Section 15.7.1 describes the analysis of a radioactive gaseous waste system failure.
The following design features are incorporated to minimize maintenance, equipment downtime, leakage, and radioactive gaseous releases. These features facilitate offgas operation and assist in maintaining Operator exposures ALARA.
- 1.
Major components are duplicated to provide two parallel offgas trains.
- 2.
Components requiring servicing are placed in individual shielded cubicles to minimize personnel exposure during maintenance.
- 3.
The hydrogen concentration in the offgas is kept below the flammable limits by maintaining the required steam flow for dilution at all times.
- 4.
The operating pressure of the OFG system is subatmospheric. Because of the inherent pressure differential between the system and the atmosphere, leakage will be negligible.
- 5.
Leakage of radioactive gas is further reduced by using welded construction wherever possible.
- 6.
The OFG system can be operated from a single remote control panel which is located in a non-radiation area. The offgas chillers may also be operated locally via switches on each chiller skid.
- 7.
The system is supplied from non-Class 1E power sources.
Hydrogen concentration will be maintained below the 4 volume percent flammability limit by maintaining the required steam flow for dilution at all power levels. In addition, hydrogen analyzers will be used to confirm these concentrations.
NMP Unit 2 USAR Chapter 11 11.3-4 Rev. 25, October 2022 Operability requirements regarding offgas system explosive gas monitoring instrumentation are described in TRM Section 3.3.11.
11.3.2 System Description 11.3.2.1 Offgas System Noncondensable radioactive process offgas is continuously removed from the main condensers by the steam jet air ejectors (SJAE). The offgas from the main condensers will be the major source of radioactive gas, and is greater than all other sources combined.
The condenser offgas will normally contain activation gases, principally N-16, O-19, and N-13. It will also contain the radioactive noble gas parents of Sr-89, Sr-90, Ba-140, and Cs-137 (Table 11.3-1).
In addition to removing process offgas from the main condensers, the air ejectors also provide adequate mode of force for the gases to flow to the OFG system. The design feed to each offgas train is:
Dry air 40 scfm @ 60°F H2 136 scfm O2 68 scfm H2O 3,740 scfm as dilution steam Noble gases Negligible The offgas mixture enters the system through a preheater. This preheater uses steam to raise the temperature of the gaseous mixture from 260°F to 290°F. The increase in temperature will increase the reaction rate in the recombiner. Hydrogen and oxygen react in the recombiner to form water, thus reducing the potential for hydrogen flammability. Once the system is in normal operation, the steam supply to the preheaters may be secured.
The newly-formed water vapor and dilution steam are removed downstream of the recombiner by a condenser. This also serves to remove the heat of reaction developed in the recombiner. The condensate returns to the main condenser hotwell. Downstream of the OFG system condenser is a 75-sec delay pipe, where the short-lived radioisotopes, such as N-16, N-17, and O-19, decay.
The gas is then passed through dryers which lower the moisture content. This is desirable since moisture reduces the efficiency of the charcoal beds downstream. The charcoal beds are arranged in two parallel trains of four each. Each train is
NMP Unit 2 USAR Chapter 11 11.3-5 Rev. 25, October 2022 designed to handle 50 percent of the gas flow. The trains can be operated in parallel or in series.
Xenon and krypton isotopes will be adsorbed on the charcoal and delayed from the bulk carrier gas (essentially air) permitting them to decay, thereby significantly reducing the offsite dose.
Design basis holdup times of 20 days for xenon and 26.6 hr for krypton have been selected. Following the charcoal beds is a high-efficiency particulate air (HEPA) filter to collect solid decay products and charcoal fines.
Vacuum pumps are utilized to provide the motive force in drawing the offgas through the system. These pumps discharge the waste gas through the main stack, where it is monitored for radioactivity prior to release to the atmosphere.
The hydrogen water chemistry (HWC) system includes an oxygen injection system to offgas, upstream of the offgas recombiner, to maintain a stoichiometric mixture of hydrogen and oxygen in the recombiner. The system is provided due to an excess ratio of hydrogen to oxygen at the entrance to the offgas system because of hydrogen injection through the condensate system.
The HWC system includes an additional offgas sample system for monitoring of the offgas oxygen concentration from the recombiners to assure that the oxygen addition flows are properly balanced. The sample supply for the oxygen analyzers is taken from the same general location as the existing hydrogen samples.
Control Features of Offgas System Control Panel The control panel, located in a non-radiation area, provides all remote monitoring and control functions of the OFG system. It contains visual display of various system parameters, annunciator systems (lights, audible alarms, and controls), system status lights and control switches, and a system mimic display mounted on the face of the control panel.
The panel will alarm and control all necessary components so the Operator will have complete knowledge of system status. The offgas chillers may also be operated locally via switches on each chiller skid. The offgas chiller trouble alarm may be silenced via a maintenance switch located on each chiller skid.
The system has been designed to operate completely unattended during normal operation. The control panel has alarm provisions for the following system parameters:
- 1.
System inlet temperature.
- 2.
System inlet pressure.
- 3.
Recombiner inlet temperature (low).
- 4.
Recombiner inlet temperature (high).
NMP Unit 2 USAR Chapter 11 11.3-6 Rev. 25, October 2022
- 5.
Catalytic recombiner outlet temperature.
- 6.
Condenser outlet temperature.
- 7.
Condenser outlet hydrogen concentration.
- 8.
Condenser high level.
- 9.
Condenser low level.
- 10. Condenser combined outlet hydrogen concentration.
- 11. Dryer temperature.
- 12. HEPA filter differential pressure.
- 13. Vacuum pump inlet pressure.
- 14. Train shutdown.
- 15. Vacuum pump motor overload.
- 16. Chiller trouble.
- 17. Pretreatment high radiation.
- 18. Dryer drain tank water level (high).
- 19. Dryer drain tank water level (low).
- 20. Dryer differential pressure (high).
- 21. Discharge valve to stack closed.
The offgas control panel is located in the offgas area in the turbine building. The control room has a trouble alarm which alarms on all the items noted in Section 11.3. In addition, the specified reason for an alarm is printed out on a computer log located in the control room.
Moisture Transmitter The moisture transmitter is located upstream of the charcoal adsorbers and is designed to provide a dew point readout for the process steam. The alarm for high process dew point is obtained from the dryer intermediate stage temperature sensing loop, which is essentially the output temperature of the dryers due to the freeze-out coils being eliminated from the dryer loops. This temperature is an appropriate measure of dew point as the gas in the dryer intermediate stage is at saturation, thus producing a 100-percent humidity level. The temperature alarm serves to guard the charcoal adsorbers from moisture contamination resulting from freeze-out dryer or condenser damage.
NMP Unit 2 USAR Chapter 11 11.3-7 Rev. 25, October 2022 Hydrogen Analyzers Three hydrogen analyzers are installed in the process flow line to monitor hydrogen concentration. One hydrogen analyzer is installed downstream of each condenser and analyzes process flow hydrogen concentration and transmits a signal to an annunciator on the control room panel and offgas control panel.
The third analyzer is installed in the common process piping to the dryers and is designed to monitor hydrogen concentration.
This analyzer also transmits a signal to an alarm annunciator on the control panel.
Oxygen Analyzers Two oxygen analyzers are installed in the process flow lines to monitor oxygen concentration in the offgas flow. The analyzers can be configured to monitor the offgas flow at any two of three sample points where the hydrogen analyzers are installed (A train, B train, or common header).
The oxygen analyzers will be used to support the operation of the hydrogen water chemistry (HWC) system. A low percent oxygen in the offgas will provide a signal to the HWC system to isolate the hydrogen and oxygen injection lines.
Protective Features All critical system parameters are automatically monitored during automatic operation. Certain abnormal conditions will automatically shut down the process train in which the abnormal condition exists. These conditions are as follows:
- 1.
Condenser high outlet temperature.
- 2.
Recombiner high outlet temperature.
- 3.
Preheater low-low outlet temperature.
- 5.
Manual or automatic vacuum pump shutdown.
- 6.
Pretreatment high radiation level.
Radiation Monitoring Two pretreatment gross gamma radiation monitors mounted in parallel are located downstream of the dryers just upstream of the charcoal adsorber vessels. Their function is to monitor the activity level of the process stream and to alarm when an abnormal condition exists. When this occurs, the system outlet valve automatically closes, thereby terminating discharge to the main stack. Process and effluent radiological monitoring systems are further described in Section 11.5.
11.3.2.2 Ventilation Systems The following ventilation systems potentially contain radioactive gas.
NMP Unit 2 USAR Chapter 11 11.3-8 Rev. 25, October 2022
- 1.
Reactor building HVAC/containment purge (Section 9.4.2).
- 2.
Turbine building HVAC system (Section 9.4.4).
- 3.
Radwaste building HVAC system (Section 9.4.3).
The expected concentrations of airborne radioactivity at various in-plant areas are calculated based on data given in NUREG-0016 and EPRI-495. The concentrations in the reactor building, both for normal operation and anticipated operational occurrences, e.g., refueling, are presented in Table 12.2-15. Airborne radioactivity concentrations in various turbine and radwaste building areas for full-power operation are also provided in Table 12.2-15.
11.3.2.3 Steam and Power Conversion Systems During startup of the unit, air is removed from the main condenser by a mechanical vacuum pump. In this mode of operation, when little or no radioactive gas is present in the condenser, vacuum pump discharge is directed to the main stack.
During normal operation, the SJAEs provide sufficient force for condenser air removal. This system is further discussed in Section 10.4.2. The turbine gland sealing system utilizes steam to pressurize the turbine seals. This prevents radioactive steam from escaping through the seals to atmosphere. This system is discussed further in Section 10.4.3.
11.3.2.4 Miscellaneous Gaseous Releases Exposure of the drywell air to neutron leakage fluxes around the reactor vessel results in some activation products. Activity may also be introduced into the drywell atmosphere by steam leaks and by venting of the primary system safety valves into the suppression chamber. Except for leakage of 2 scfm, the drywell essentially forms a closed system. When access is required, the drywell may be purged with normal reactor building air. The drywell can also be vented during startup to accommodate the expansion of air as its temperature increases.
The discharge of this air is to the main stack by way of the SGTS (Section 6.5.1).
11.3.3 Radioactive Releases 11.3.3.1 Release Points The plant airborne radioactive releases to the environs are from two monitored points, the main stack and the combined radwaste/reactor building vent. Figure 1.2-2 shows the release points for each effluent source.
The main stack is 430 ft above grade, which is 2.5 times the height of the reactor building, the tallest structure in the
NMP Unit 2 USAR Chapter 11 11.3-9 Rev. 25, October 2022 power block. The main stack releases exhaust air from the following plant areas and systems:
- 1.
Turbine building.
- 2.
Containment purge.
- 3.
Turbine generator gland seal and exhaust steam system.
- 4.
OFG system.
- 5.
Mechanical vacuum pump discharge (intermittent release).
- 6.
SGTS.
- 7.
The combined radwaste/reactor building vent is located 186 ft above ground level and releases exhaust air from the radwaste building equipment and area exhaust, the auxiliary boiler building area exhaust, and the reactor building ventilation exhaust, above and below the refueling floor. Figure 11.3-1 shows the P&ID of the radioactive gaseous waste system.
11.3.3.2 Design and Expected Releases The parameters used to calculate the design and expected releases are presented in Table 11.3-2. Expected ventilation exhaust, offgas, mechanical vacuum pump, and containment purge releases are presented in Table 11.3-1. The design annual average release concentrations are shown in Table 11.3-3 as activity in uCi/cc, and as a fraction of maximum permissible concentration.
11.3.3.3 Dilution Factors The atmospheric dilution factor associated with normal plant releases is based on the average annual meteorological conditions applicable to the site as well as the effective release height of the effluent discharge pathway. The site meteorological conditions are given in Section 2.3.
11.3.3.4 Estimated Doses A summary of the estimated annual radiation doses is presented in Appendix 11A and shows that the estimated annual doses from gaseous effluents are below the dose criteria set forth in 10CFR50 Appendix I and are well below the dose criteria specified in 40CFR190 and 10CFR20.
The maximum hypothetical gamma and beta air doses from noble gas releases occur at the exclusion area boundary (EAB), 1,603 m east of the site. The doses at this location are estimated to
NMP Unit 2 USAR Chapter 11 11.3-10 Rev. 25, October 2022 be 7.9E-02 mrad/yr gamma and 5.3E-02 mrad/yr beta, as compared with the 10CFR50 Appendix I design objective for gamma and beta air doses of 10.0 mrad/yr and 20.0 mrad/yr, respectively.
NMP Unit 2 USAR Chapter 11 11.3-11 Rev. 25, October 2022 TABLE 11.3-1 EXPECTED RADIOACTIVE GASEOUS EFFLUENT FROM ALL SOURCES (CI/YR)
Mech.
Reactor Turbine Radwaste Vacuum Offgas Isotope Bldg Bldg Bldg Pump System I-131 4.4E-02 3.0E-01 2.0E-02 1.1E-01 I-132 3.6E-01 2.5E+00 1.6E-01 9.4E-01 I-133 2.9E-01 2.0E+00 1.3E-01 7.5E-01 I-134 5.8E-01 4.2E+00 2.6E-01 1.6E+00 I-135 4.0E-01 2.8E+00 1.8E-01 1.0E-00 Kr-83m 2.7E+00 Kr-85m 3.0+00 2.5+01 1.3E+03 Kr-85 2.7E+02 Kr-87 2.1+00 6.1+01 2.2E-01 Kr-88 3.8+00 9.1+01 4.6E+02 Kr-89 2.4+00 5.8+02 2.9+01 Xe-131m 7.5E+01 Xe-133m 9.0E+00 Xe-133 1.1+02 1.5+02 2.2+02 1.3+03 8.3E+03 Xe-135m 6.1+01 4.0+02 5.3+02 Xe-135 1.3+02 3.3+02 2.8+02 5.0+02 4.2E-10 Xe-137 1.9+02 1.0+03 8.3+01 Xe-138 7.3+00 1.0+03 2.0+00 Cr-51 9.7-04 9.0-04 7.0-06 Mn-54 1.3-03 6.0-04 4.0-05 Fe-59 3.5-04 1.0-04 3.0-06 Co-58 2.9-04 1.0-03 2.0-06 Co-60 5.1-03 1.0-03 7.0-05 Zn-65 4.1-03 6.0-03 3.0-06 Sr-89 1.7-04 6.0-03 Sr-90 7.9-06 2.0-05 Zr-95 6.2-04 4.0-05 8.0-06 Nb-95 1.1-02 6.0-06 4.0-08 Mo-99 6.3-02 2.0-03 3.0-08 Ru-103 5.9-04 5.0-05 1.0-08 Ag-110m 2.6-06 Sb-124 2.1-05 1.0-04 7.0-07 Cs-134 4.3-03 2.0-04 2.4-05 Cs-136 4.9-04 1.0-04 Cs-137 5.7-03 1.0-03 4.0-05 Ba-140 1.4-02 1.0-02 4.0-08 Ce-141 7.9-04 1.0-02 7.0-08 Ar-41 1.5+01 1.06E+02 H-3 3.1E+01 3.1E+01 C-14 7.8E+01 NOTE:
3.0+00 = 3.0x100 = 3.0E00
NMP Unit 2 USAR Chapter 11 11.3-12 Rev. 25, October 2022 TABLE 11.3-2 DATA USED IN CALCULATING ANNUAL RELEASES OF RADIOACTIVE GASEOUS EFFLUENTS Parameter Data Rated thermal power 3,988 (MWt)
Rated steam flow rate 17,633,000 lb/hr Offgas charcoal bed holdup*
times per NUREG-0016 (Kr) 25.4 hr (Xe) 18.67 days Plant capacity factor 95%
Design releases source term 0.3536 Ci/sec failed fuel basis (at 30 min)
Offgas system charcoal mass/train 48,000 lb Dynamic adsorption coefficients (Kr) 25 cm3/gm (Xe) 440 cm3/gm Charcoal delay system normal 70°F operating temperature Charcoal delay system dew 50°F point temperature Ventilation systems See Section 9.4 Decontamination factors Normal operations Radwaste building 99% efficient HEPA filter All other buildings/systems Unfiltered Shutdown Radwaste building 99% efficient HEPA filter
NMP Unit 2 USAR Chapter 11 11.3-13 Rev. 25, October 2022 TABLE 11.3-2 (Cont'd.)
Parameter Data Reactor building 10% of effluents -
unfiltered 90% of effluents -
90% efficient iodine filter 99% efficient HEPA filter All other buildings/systems Unfiltered The offgas system holdup times presented are based on NUREG-0016 calculation methods. The holdup time values in this table are used only to calculate annual radioactive gaseous effluent releases from Unit 2. The offgas system holdup times discussed in Section 11.3.2.1 are based on design system parameters.
NMP Unit 2 USAR Chapter 11 11.3-14 Rev. 25, October 2022 TABLE 11.3-3 DESIGN ANNUAL AVERAGE GASEOUS RELEASES VS. MPC Isotope Main Stack Releases Radwaste/Reactor Building Vent Releases Total Release (fraction of MPC)
Continuous Annual Average Intermittent (MVP)
Annual Average MPC*
(uCi/cc)
Activity at EAB (uCi/cc)
Fraction of MPC Activity at EAB (uCi/cc)
Fraction of MPC Activity at EAB (uCi/cc)
Fraction of MPC I-131 I-132 I-133 I-134 I-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Cr-51 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Mo-99 Ru-103 Ag-110m Sb-124 Cs-134 Cs-136 Cs-137 Ba-140 Ce-141 Ar-41 H-3 C-14 Total 3.1E-16 4.1E-15 3.2E-15 9.7E-15 3.5E-15 1.4E-15 1.7E-12 4.9E-13 1.3E-13 6.2E-13 1.2E-12 1.1E-13 8.4E-15 1.4E-11 8.3E-13 6.7E-13 2.1E-12 2.0E-12 2.6E-19 2.7E-19 2.1E-19 2.0E-17 1.0E-18 1.7E-18 5.4E-17 1.8E-19 8.6E-20 1.6E-20 6.5E-18 1.1E-19 3.1E-22 2.0E-19 4.2E-19 2.1E-19 2.1E-18 6.2E-17 2.1E-17 2.5E-14 7.9E-15 1.9E-14 3.1E-06 1.4E-06 8.0E-06 1.6E-06 3.5E-06 4.6E-08 1.7E-05 1.6E-06 6.3E-06 3.1E-05 4.0E-05 2.8E-07 2.8E-08 4.5E-05 2.8E-05 6.7E-06 7.0E-05 6.8E-05 3.3E-12 2.7E-10 1.0E-10 1.0E-08 3.4E-09 8.7E-10 1.8E-07 6.1E-09 8.6E-11 5.5E-12 9.3E-10 3.5E-11 1.0E-12 2.9E-10 1.0E-09 3.5E-11 4.2E-09 6.2E-08 4.1E-09 6.3E-07 3.9E-08 1.9E-07 3.3E-04 7.2-16 9.1-15 7.2-15 2.1-14 7.7-15 1.6-11 6.0-12 7.2E-06 3.0E-06 1.8E-05 3.5E-06 7.7E-06 5.3E-05 6.0E-05 1.5E-04 2.0E-15 2.1E-14 1.6E-14 3.9E-14 1.7E-14 1.0E-13 7.6E-14 1.4E-13 1.1E-12 1.2E-11 2.1E-11 1.5E-11 9.8E-12 3.3E-13 5.0E-18 1.1E-17 1.3E-17 1.0E-16 9.0E-17 2.1E-17 2.7E-17 1.2E-18 2.3E-17 4.0E-16 3.6E-15 2.1E-17 2.2E-18 7.9E-19 1.6E-16 1.7E-17 2.0E-16 1.5E-15 2.8E-17 1.4E-13 2.0E-05 7.0E-06 3.9E-05 6.4E-06 1.7E-05 1.0E-06 3.8E-06 6.8E-06 3.7E-05 4.0E-05 7.1E-04 1.5E-04 3.3E-04 1.1E-05 6.2E-11 1.1E-08 6.6E-09 5.1E-08 3.0E-07 1.0E-08 8.9E-08 4.2E-08 2.3E-08 1.3E-07 5.1E-07 7.1E-09 7.4E-09 1.1E-09 3.9E-07 2.9E-09 4.1E-07 1.5E-06 5.6E-09 6.8E-07 1.4E-03 1.0E-10 3.0E-09 4.0E-10 6.0E-09 1.0E-09 3.0E-08 1.0E-07 3.0E-07 2.0E-08 2.0E-08 3.0E-08 4.0E-07 3.0E-07 3.0E-07 3.0E-08 1.0E-07 3.0E-08 3.0E-08 8.0E-08 1.0E-09 2.0E-09 2.0E-09 3.0E-10 2.0E-09 3.0E-10 3.0E-11 1.0E-09 3.0E-09 7.0E-09 3.0E-09 3.0E-10 7.0E-10 4.0E-10 6.0E-09 5.0E-10 1.0E-09 5.0E-09 4.0E-08 2.0E-07 1.0E-07 1.9E-05 8.2E-06 4.7E-05 8.0E-06 2.0E-05 4.6E-08 1.8E-05 1.6E-06 1.0E-05 3.8E-05 7.7E-05 2.8E-07 2.8E-08 8.5E-05 7.4E-04 1.6E-04 4.0E-04 7.9E-05 6.6E-11 1.1E-08 6.7E-09 6.1E-08 3.1E-07 1.1E-08 2.7E-07 4.8E-08 2.3E-08 1.3E-07 5.1E-07 7.1E-09 7.4E-09 1.4E-09 3.9E-07 2.9E-09 4.1E-07 1.6E-06 9.8E-09 6.3E-07 7.2E-07 1.9E-07 1.7E-03 In accordance with 10CFR20 Appendix B.
NMP Unit 2 USAR Chapter 11 11.4-1 Rev. 25, October 2022 11.4 SOLID WASTE MANAGEMENT SYSTEM Power plant operation results in various solid radioactive wastes that require disposal. The radioactive solid waste system is designed to collect, hold, monitor, process, package, and provide temporary storage facilities for radioactive materials prior to shipment offsite and ultimate disposal. The solid waste management system is shown on Figure 11.4-1.
11.4.1 Design Basis The radioactive solid waste system is designed to the following criteria:
- 1.
The system provides for the dewatering/solidification and packaging of wet solid wastes into shipping containers prior to shipment for offsite disposal.
- 2.
All solid waste containers, shipping casks, and methods of packaging meet applicable state and federal regulations. Wastes will be shipped to a licensed burial site in accordance with applicable NRC and Department of Transportation (DOT) regulations (i.e.,
10CFR71 and 49CFR171-178).
- 3.
The filling of containers, dewatering/solidification, and storage of radioactive solid waste conforms to 10CFR20 and 10CFR50 requirements and RG 8.8 guidelines in terms of ALARA doses to plant personnel and the general public.
- 4.
Remote automatic and/or manual operation is provided by the radwaste dewatering system control panel and waste handling control panel.
- 5.
System reliability is emphasized through proven design of system components, compartmentalization of equipment layout, shielding, containment of possible spills, accurate process monitoring, and interlocking of process controls.
- 6.
The seismic criteria and analytical procedures for structures housing the solid radwaste system are given in Section 3.2.1 and 3.7. The quality group classification for the system components and piping is given in Table 3.2-1.
11.4.2 System Inputs Radioactive solid wastes that result from plant operation consist of concentrated liquid wastes from the evaporators in the radioactive LWS, spent resins from all plant demineralizers handling radioactive liquids, filter sludges from LWS filters, phase separators, and miscellaneous solid materials that become
NMP Unit 2 USAR Chapter 11 11.4-2 Rev. 25, October 2022 contaminated during plant operation and maintenance. Table 11.4-1 is a conservatively high estimate of the expected volumes of both the wet and dry radioactive solid waste generated by the unit.
11.4.2.1 System Inputs Activity Expected and design wet solid waste activities are given in Table 11.4-2. The prediction of solid radwaste principal nuclide curie inventories was obtained from a mathematical model of expected and design values of reactor coolant and main steam radionuclide concentrations, as discussed in Section 11.1 and found in Table 11.4-3.
Expected and maximum dry solid radwaste inventories and annual curie content for both compactible and noncompactible wastes are provided in Tables 11.4-5 and 11.4-6. The values in these tables are based on data on radionuclides present in dry solid radwaste from operating plants(1), and are calculated for Unit 2 using the parameters in Table 11.4-7.
11.4.3 System Description The original plant radioactive waste solidification system (WSS) consisted of an asphaltic-based solidification process described in Topical Reports WPC-VRS-001(2) and WPC-VRS-002(3). This system primarily consisted of a waste fill station, a monitor and capping station, an extruder/evaporator, control console, and piping, metering pumps, and process equipment required for transfer and solidification of wastes. The asphaltic-based solidification process is no longer used and most of the associated process equipment has permanently been abandoned in place. Table 11.4-4 lists the major equipment of the originally supplied system and identifies the equipment that has been abandoned in place.
In place of the asphaltic-based solidification process, a radwaste dewatering system described in Section 11.4.3.3.1 provides the primary method of waste volume reduction.
The spent resin and filter sludge handling system consists of a waste sludge tank with agitator, redundant transfer pumps, and a decant pump. The system provides for holdup, recirculation, and sampling of waste, and decanting of excess water prior to transfer to the solidification portion of the system.
The evaporator bottoms handling system includes a concentrated waste transfer pump. In conjunction with the LWS evaporator bottoms tank, this system provides for holdup, recirculation, and sampling of evaporator concentrates prior to transfer to dewatering/solidification and phase separation of condensate filtration system (CFS) filter backwash water. Originally, all lines and components in the waste concentrate system were heat traced (to prevent crystallization of the waste salts in the
NMP Unit 2 USAR Chapter 11 11.4-3 Rev. 25, October 2022 system). Operating experience and current waste handling practices determined that waste salts are no longer a concern.
Thus, some heat trace circuits have been removed or abandoned in place. In addition, waste concentrate lines are equipped with relief valves, where required, that are discharged to the building floor drains.
The various plant systems that interface with the waste dewatering/solidification system are shown on Figures 11.4-1a to 11.4-1h and are as follows:
- 1.
LWS provides the waste feed streams.
- 2.
Radwaste auxiliary steam (ASR) provides hot flush water.
- 3.
Steam released from relief valves is discharged to the auxiliary boiler steam (ABM) relief header.
- 4.
Makeup water system (MWS) supplies boiler feedwater to the WSS boiler.
- 5.
Condensate makeup and drawoff water (CNS) is used for equipment flushing.
- 6.
Instrument air system (IAS) supplies air to all instrument and air-operated devices.
- 7.
Service air system (SAS) provides air for clearing lines via hose connections.
- 8.
All equipment is provided with drains to the radwaste building floor and equipment drains system (DFW). All liquid effluent from the WSS is returned to the liquid waste system.
- 9.
Exhaust air from the equipment is vented to the radwaste building ventilation system (HVW).
- 10. Turbine building closed loop cooling water (CCS) provides component cooling.
- 11. Radwaste sampling system (WSS) provides process sampling.
- 12. Radwaste seal water system (SWR) provides seal water for pumps with double mechanical seals. Details are given in Section 11.2.
A compactor is provided to compress dry wastes such as paper, rags, and plastic for packaging in metal boxes. Incompressible solid wastes are packaged in metal boxes, 55-gal drums, or encapsulated in liners ranging in size from 50 to 200 cu ft.
NMP Unit 2 USAR Chapter 11 11.4-4 Rev. 25, October 2022 11.4.3.1 Spent Resin/Filter Sludge Packaging The waste sludge tank receives spent resin/filter sludge from the LWS spent resin tank and radwaste filter backwash tank.
Excess water may be removed from the sludge tank by the decant pump through retention screens to prevent resin carryover.
Contents of the waste sludge tank are fed by one of two redundant waste sludge pumps to the radwaste truck bay for treatment by the radwaste dewatering system or cement solidification by an approved vendor.
11.4.3.2 Evaporator Bottoms Packaging Evaporator bottoms are processed from the LWS evaporator bottoms tank. Contents are transferred to an approved container in the radwaste truck bay. The waste is then processed to final form in the truck bay or shipped offsite for vendor processing to final form.
11.4.3.3 Radwaste Backup System The radwaste backup system provides an alternate method for the solidification of spent resin/filter sludge and evaporator bottoms by a mobile, skid-mounted solidification system which would be temporarily located in the radwaste building truck bay.
In the event that solidification is required, three pipelines will bypass the extruder and go directly to a hose station in the truck bay area. These three pipelines consist of:
- 1.
A 1 1/2-in diameter, electrically-traced line that carries evaporator bottoms waste directly from the evaporator bottoms transfer pump.
- 2.
A 1 1/2-in diameter line that carries spent bead resins and filter sludge directly from the waste sludge tank pumps.
- 3.
A 2-in diameter line that carries spent powdered resin from the influent header feed line of the spent resin tank.
Flexible hoses will be used to transfer the spent waste from the hose station to a mobile solidification unit.
The truck bay area is classified as Radiation Zone I/IV, restricted during the radwaste loading operation.
This backup facility consists entirely of transfer piping and valves. It is part of and is designed and constructed to the same criteria as the LWS and WSS systems with which it interfaces. All transfer operations are manually controlled from a control panel located in the truck dock area.
NMP Unit 2 USAR Chapter 11 11.4-5 Rev. 25, October 2022 11.4.3.3.1 Radwaste Dewatering System The radwaste dewatering system provides an acceptable method of waste volume reduction by a self-contained, freestanding, portable dewatering system located in the radwaste building truck bay, el 261'. This unit consists of a dewatering skid, a plant connection stand, a control module, a container fillhead, and a waste container and associated interconnecting hoses and cables. Plant services required to properly operate the dewatering system include service air, service water, electrical power, and waste as shown on Figure 11.4-1. The design, operation, and safety evaluation of this dewatering system is described in Chem Nuclear Systems, Inc., Topical Report RDS-25506-01-P/NP(4), which has been reviewed and accepted by Unit 2 management and the NRC as applicable.
11.4.3.4 Dry Waste Packaging Dry activated waste (DAW) is collected at various locations in the plant and transported to the radwaste building. The DAW, such as paper and rags, is packaged in approved shipping containers prior to shipment to a waste disposal facility, or offsite waste processor. The packaging and shipment is done in accordance with approved Station procedures.
11.4.3.5 Incompressible Waste Packaging Components of low activity, such as contaminated tools, can be packaged in available or specially-designed shipping containers and stored, if necessary, in the radwaste building prior to shipment.
Spent core components with very high activity levels are handled underwater within the reactor refueling cavity and fuel transfer canal, and stored in the fuel pool until adequate packaging is provided for offsite shipment. Refer to Section 9.1 for additional information on spent fuel handling.
11.4.3.6 Waste Packaging Controls Complete solidification of the processed waste is ensured by preoperational testing and the implementation of a process control program. Waste sludge and evaporator bottoms tanks are provided with means for obtaining representative samples via a shielded sampling station located in the radwaste building.
Waste packaging controls and monitoring are provided from the solidification system control panel. The system is designed to prevent external contamination of the containers by instrumentation interlocks that prevent overfilling.
Specific requirements regarding radioactive waste solidification are described in TRM Section 3.11.1.
11.4.3.7 Waste Handling
NMP Unit 2 USAR Chapter 11 11.4-6 Rev. 25, October 2022 Waste handling is provided by a 30-ton, overhead, traveling bridge crane with an 8 1/2-ton auxiliary hoist. Remote handling of processed waste is done by closed-circuit TV monitoring from the waste handling control panel.
11.4.4 Packaging Filling of containers and storage of radioactive solid wastes conforms with 10CFR20 and 10CFR50 requirements. Packages meet shipping regulations of 49CFR171-178 and 10CFR71, as applicable.
The waste packaging procedure is described in Section 11.4.3.
Containers used are 50- to 200-cu ft liners, 55-gal drums, and metal boxes with supplementary lead or steel shielding as required for shipment.
11.4.5 Storage Facilities The radwaste building was designed to process and store Unit 2 low-level radioactive waste (LLRW). El 245 ft of the radwaste building has been modified with the specific intent of providing interim onsite storage of LLRW (primarily compacted DAW and incinerator ash), including storage from both Unit 1 and Unit 2.
The need to store LLRW onsite is the result of the federal Low-Level Radioactive Waste Policy Act as amended in 1985, which initiated the process by which the three existing LLRW disposal sites (Barnwell, SC; Beatty, NV; and Hanford, WA) would no longer be required to receive LLRW. El 245 ft of the radwaste building is capable of providing at least 5 yr of interim storage of low activity LLRW produced at both Unit 1 and Unit 2.
The storage of Unit 1 LLRW at Unit 2 is considered acceptable based on the following:
- 1.
The isotopic library to be considered is essentially the same for both units;
- 2.
The isotopic distributions for the two units are similar and release pathways will not be affected.
The higher Co-60 weighting factor for Unit 1 waste will have no appreciable effect on shielding, release analyses or handling;
- 3.
The material to be stored at Unit 2 is of very low activity level; and
- 4.
The transfer of by-product material between Unit 1 and Unit 2 will be conducted in accordance with approved radiation protection implementing procedures.
The Unit 1 radwaste solidification and storage building (RSSB) provides interim storage of higher activity waste generated at Unit 1 and Unit 2.
NMP Unit 2 USAR Chapter 11 11.4-7 Rev. 25, October 2022 Dewatered/solidified waste is stored in a shielded area within the radwaste building in liners ranging in size from 50 to 200 cu ft. The storage capacity of processed waste is approximately a 4-month output of packaged waste at expected generation rates.
11.4.6 Shipment Shipment of radioactive solid wastes conforms with 10CFR50, 10CFR61, 10CFR71, and 49CFR171 through 49CFR178 requirements.
Higher-activity wastes are shipped in shielded casks, as applicable.
Solid waste is transported by a licensed disposal contractor or a common carrier to a licensed burial site.
Tables 11.4-1 and 11.4-2 summarize the annual number of packaged containers, expected number of shipments to be made, and expected and design activities of the wastes.
11.4.7 Process Control Program The Process Control Program (PCP) was approved by the NRC before implementation. It contains the current formula sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of radioactive wastes, based on demonstrated processing of actual or simulated wet or liquid wastes, will be accomplished in such a way as to assure compliance with 10CFR20, 10CFR61, 10CFR71, and Federal and State regulations and other requirements governing the transport and disposal of radioactive waste.
Licensee-initiated changes to the PCP shall become effective upon review and acceptance by the PORC and shall be submitted to the NRC in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:
- 1.
Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
- 2.
A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- 3.
Documentation of the fact that the change has been reviewed and found acceptable by the PORC.
11.4.8 References
- 1.
Phillips, J., Feizollahi, F., Martineit, R., and Bell, W.
Waste Report for Reactor and Fuel-Fabrication Facility Wastes, ONWI-20/NUS-3314, NUS Corporation, March 1979.
NMP Unit 2 USAR Chapter 11 11.4-8 Rev. 25, October 2022
- 2.
Radwaste Volume Reduction and Solidification System-Topical Report, Report Number WPC-VRS-001, Revision 1, Werner and Pfleiderer Corporation, May 1978.
- 3.
Topical Report, 10CFR61, Waste Form Conformance Program for Solidified Process Waste Products by a Wastechem Corporation Volume Reduction and Solidification (VRS)
System, Report No. VRS-002, Revision 1, August 1987.
- 4.
Topical Report, RDS-1000 Radioactive Waste Dewatering System, Report No. RDS-25506-01-P/NP, Revision 1, Chem Nuclear Systems, Inc.
NMP Unit 2 USAR Chapter 11 11.4-9 Rev. 25, October 2022 TABLE 11.4-1 ANNUAL WET AND DRY SOLID WASTE QUANTITIES VOLUME Source of Wet Wastes Expected Design Unsolidified Waste Volume (ft3) 50 ft3 Containers Packaged Volume(3)
(ft3)
Shipments Unsolidified Waste Volume (ft3) 50 ft3 Containers Packaged Volume(3)
(ft3)
Shipments Spent resins (radwaste demineralizer, condensate demineralizer) 1,210 31 1,628 19 4,400 113 5,933 68 Filter sludges 4,840 124 6,510 74 9,064 232 12,180 140 Evaporator bottoms
- 1. Radwaste
- 2. Regenerative Subtotal 2,487 5,653 14,190 8
39 202 420 2,048 10,606 35(1) 128 4,504 10,236 28,204 14 71 430 735 3,728 22,576 64(1) 272 Source of Dry Wastes Metal Boxes(2) 55-Gal Drums(4)
Packaged Volume (ft3)
Shipments Metal Boxes(2) 55-Gal Drums(4)
Packaged Volume (ft3)
Shipments Compacted 1,228 9,210 18 2,052 15,390 29 Miscellaneous 8
1,024 1
13 1,664 2
Subtotal 8
1,228 10,234 19 13 2,052 17,054 31 Total 20,840 147 39,630 303 (1)
Combined shipments of radwaste and regenerative evaporator bottoms.
(2)
Metal boxes contain 128 ft3 volume.
(3)
Packaged volume based on 52.5 ft3 of external volume.
(4)
Packaged volume based on 7.5 ft3.
NMP Unit 2 USAR Chapter 11 11.4-10 Rev. 25, October 2022 TABLE 11.4-2 EXPECTED AND DESIGN WET SOLID WASTE ACTIVITIES Source of Wet Wastes Expected Total Activity (Ci/yr)
Design Total Activity (ci/yr) uCi/cc Ci/ft3 uCi/cc Ci/ft3 Spent resin (radwaste demineralizer, condensate demineralizer) 58.62 1.66 2.01x103 497.96 14.10 6.20x104 Filter sludges Radwaste filter 1.58 0.05 43.42 2.60 0.07 133.47 Evaporator bottoms (radwaste and regenerative) 35.50 1.00 8.18x103 293.02 8.30 1.22x105
NMP Unit 2 USAR Chapter 11 11.4-11 Rev. 25, October 2022 TABLE 11.4-3 WET SOLID RADWASTE PRINCIPAL NUCLIDE INVENTORIES Activity Isotope (uCi/cc)
Radwaste Filter Sludge Na-24 0.202+00 P-32 0.725-02 Cr-51 0.217+00 Mn-54 0.334-02 Mn-56 0.411-03 Fe-55 0.374-01 Fe-59 0.759-02 Co-58 0.477+00 Co-60 0.479-01 Ni-65 0.266-02 Cu-64 0.536+00 Zn-65 0.750-02 Zn-69m 0.382-01 Ag-110m 0.576-02 Ag-110 0.115-03 W-187 0.188+00 Nb-98 0.398-02 Waste and Regenerant Evaporator Bottoms Co-60 0.655+00 Ni-65 0.459-05 Cu-64 0.365-01 Zn-65 0.951-01 Zn-69m 0.293-02 Ag-110m 0.733-01 Ag-110 0.147-02 W-187 0.362-01 Np-239 0.123+02 Nb-98 0.316-05 Tc-104 0.409-06 Ba-139 0.237-03 Ba-140 0.341+01 Ba-141 0.234-06 Ba-142 0.542-08 La-140 0.362+01 La-141 0.105-02 La-142 0.382-04 Ce-141 0.112+00 Ce-143 0.183-02 Ce-144 0.430-01
NMP Unit 2 USAR Chapter 11 11.4-12 Rev. 25, October 2022 TABLE 11.4-3 (Cont'd.)
Activity Isotope (uCi/cc)
Pr-143 0.478-01 Pr-144 0.430-01 Nd-147 0.461-02 Pm-147 0.157-03 Na-24 0.182-01 P-32 0.329-01 Cr-51 0.157+01 Mn-54 0.494-01 Mn-56 0.807-03 Fe-55 0.504+00 Fe-59 0.692-01 Co-58 0.502+01 Nb-97m 0.175-03 Nb-97 0.197-03 Mo-99 0.141+01 Tc-99m 0.143+01 Tc-101 0.102-06 Ru-103 0.444-01 Ru-105 0.721-03 Ru-106 0.105-01 Rh-103m 0.444-01 Rh-105m 0.723-03 Rh-105 0.631-04 Rh-106 0.105-01 Te-129m 0.649-01 Te-129 0.651-01 Te-131m 0.514-02 Te-131 0.104-02 Te-132 0.385+01 Cs-134 0.206+00 Cs-136 0.447-01 Cs-137 0.317+00 Cs-138 0.528-05 Ba-137m 0.292+00 Br-83 0.415-02 Br-84 0.141-04 I-129 0.452-12 I-131 0.620+02 I-132 0.400+01 I-133 0.236+02 I-134 0.366-01 I-135 0.125+01 Rb-89 0.933-08
NMP Unit 2 USAR Chapter 11 11.4-13 Rev. 25, October 2022 TABLE 11.4-3 (Cont'd.)
Activity Isotope (uCi/cc)
Sr-89 0.281+01 Sr-90 0.306+00 Sr-91 0.980-01 Sr-92 0.206-02 Y-90 0.290+00 Y-91m 0.644-01 Y-91 0.556+00 Y-92 0.124-01 Y-93 0.195-01 Zr-95 0.392-01 Zr-97 0.182-03 Nb-95m 0.753-03 Nb-95 0.503-01 Reactor Water Cleanup Sludge*
Co-58 0.503+02 Co-60 0.750+01 Ni-65 0.786-03 Cu-64 0.352+00 Zn-65 0.105+01 Zn-69m 0.420-01 Ag-110m 0.809+00 Ag-110 0.162-01 W-187 0.196+00 Np-239 0.560+02 Nb-98 0.472-02 Tc-104 0.123-01 Ba-137m 0.335+01 Ba-139 0.133+00 Ba-140 0.237+02 Ba-141 0.653-02 Ba-142 0.214-02 La-140 0.258+02 La-141 0.773-01 La-142 0.174-01 Ce-141 0.984+00 Ce-143 0.866-02 Ce-144 0.479+00 Pr-143 0.337+00 Pr-144 0.479+00 Nd-147 0.309-01 Pm-147 0.205-02 F-18 0.741-01
NMP Unit 2 USAR Chapter 11 11.4-14 Rev. 25, October 2022 TABLE 11.4-3 (Cont'd.)
Activity Isotope (uCi/cc)
Na-24 0.146+00 P-32 0.235+00 Cr-51 0.133+02 Mn-54 0.553+00 Mn-56 0.134+00 Fe-55 0.576+01 Fe-59 0.644+00 Nb-95 0.290+00 Nb-97m 0.126-02 Nb-97 0.139-02 Mo-99 0.660-01 Tc-99m 0.300+01 Tc-101 0.863-02 Ru-103 0.405+00 Ru-105 0.410-01 Ru-106 0.118+00 Rh-103m 0.405+00 Rh-105m 0.411-01 Rh-105 0.232-03 Rh-106 0.118+00 Te-129m 0.741+00 Te-129 0.742+00 Te-131m 0.252-01 Te-131 0.510-02 Te-132 0.187+02 Cs-134 0.132+01 Cs-136 0.613-02 Cs-137 0.364+01 Cs-138 0.567-05 Br-83 0.541-01 Br-84 0.347-02 Br-85 0.126-04 I-129 0.156-10 I-131 0.191+02 I-132 0.196+02 I-133 0.868+01 I-134 0.127+00 I-135 0.207+01 Rb-89 0.578-03 Sr-89 0.267+02 Sr-90 0.349+01 Sr-91 0.139+01 Sr-92 0.317+00
NMP Unit 2 USAR Chapter 11 11.4-15 Rev. 25, October 2022 TABLE 11.4-3 (Cont'd.)
Activity Isotope (uCi/cc)
Y-90 0.343+01 Y-91m 0.898+00 Y-91 0.348+02 Y-92 0.601+00 Y-93 0.213-02 Y-94 0.530-10 Zr-97 0.131-02 Floor Drain Filter Sludge Na-24 0.677+00 P-32 0.263-01 Cr-51 0.786+00 Mn-54 0.139-01 Mn-56 0.930-03 Fe-55 0.136+00 Fe-59 0.275-01 Co-58 0.173+01 Co-60 0.174+00 N-165 0.338-02 Cu-64 0.177+01 Zn-65 0.271-01 Zn-69m 0.127+00 Ag-110m 0.209-01 Ag-110 0.418-03 W-187 0.648+00 Nb-98 0.783-02 Buildup activity in the spent resin tank corresponding to RWCU F/D backwashes is higher than that corresponding to the spent resin from the condensate demineralizer. Also, this activity is either equal to or less than the SFC F/D sludge.
NMP Unit 2 USAR Chapter 11 11.4-16 Rev. 25, October 2022 TABLE 11.4-4 SOLID WASTE MANAGEMENT SYSTEM MAJOR EQUIPMENT LIST Component Parameter Waste sludge tank 2WSS-TK8 Number 1
Capacity, gal 1,355 Material of construction Type 316L stainless steel
- Asphalt storage tank 2WSS-TK2 Number 1
Capacity, gal 10,800 Material of construction Carbon steel
- Extruder/evaporator 2WSS-EV25 Number 1
Capacity, gpm Varies (~1.0)
Material of construction Mfg standard
- Asphalt metering pump 2WSS-P5A&B Number 2
Capacity, gpm 0.3 Material of construction Cast iron
- Asphalt recirc pump 2WSS-P3A&B Number 2
Capacity, gpm 20 Material of construction Cast iron Waste sludge transfer pump 2WSS-P50A&B Number 2
Capacity, gpm 50 Material of construction High chrome iron
- Waste sludge metering pump 2WSS-P12A&B Number 2
Capacity, gpm 0.2-0.8 Material of construction Type 316 stainless steel Decant pump 2WSS-P10 Number 1
Capacity, gpm 40 Material of construction Type 316 stainless steel Waste concentrate transfer pump 2WSS-P6 Number 1
Capacity, gpm 50 Material of construction Alloy 20
NMP Unit 2 USAR Chapter 11 11.4-17 Rev. 25, October 2022 TABLE 11.4-4 (Cont'd.)
Component Parameter
- Waste concentrate metering pump 2WSS-P7A&B Number 2
Capacity, gpm 0.4-0.8 Material of construction 316TI stainless steel
- Steam dome boilout tank 2WSS-TK46 Number 1
Capacity, gal 3
Material of construction Type 304 stainless steel Overhead crane 2MHN-CRN1 Number 1
Capacity - main hoist 30 tons auxiliary hoist 8-1/2 tons This equipment is no longer utilized and has permanently been abandoned in place.
NMP Unit 2 USAR Chapter 11 11.4-18 Rev. 25, October 2022 TABLE 11.4-5 EXPECTED DRY SOLID RADWASTE ANNUAL NUCLIDE INVENTORIES Isotope Compactible (Ci)
Noncompactible (Ci)
Cr-51 5.3-01*
4.0+01 Mn-54 4.3-01 3.3+01 Fe-59 5.1-02 3.9+00 Co-58 1.4-01 1.1+01 Co-60 1.9+00 1.4+02 Zn-65 7.5-01 5.7+01 Zr-95 1.3-02 9.6-01 Nb-95 4.5-02 3.4+00 Cs-134 6.5-01 5.0+01 Cs-137 1.2+00 8.9+01 Total 5.6+00 4.3+02 5.3-01 is 5.3 x 10-1
NMP Unit 2 USAR Chapter 11 11.4-19 Rev. 25, October 2022 TABLE 11.4-6 MAXIMUM DRY SOLID RADWASTE ANNUAL NUCLIDE INVENTORIES Isotope Compactible (Ci)
Noncompactible (Ci)
Cr-51 8.9-01*
6.7+01 Mn-54 7.2-01 5.5+01 Fe-59 8.6-02 6.6+00 Co-58 2.4-01 1.8+01 Co-60 3.1+00 2.4+02 Zn-65 1.3+00 9.5+01 Zr-95 2.1-02 1.6+00 Nb-95 7.5-02 5.7+00 Cs-134 1.1+00 8.3+01 Cs-137 1.9+00 1.5+02 Total 9.4+00 7.2+02 8.9-01 is 8.9 x 10-1
NMP Unit 2 USAR Chapter 11 11.4-20 Rev. 25, October 2022 TABLE 11.4-7 (Historical)
PARAMETERS USED TO CALCULATE DRY SOLID RADWASTE INVENTORIES AND ANNUAL CURIE CONTENT Average plant waste activity for BWRs Compactible 5.20-03 Ci/MW(e)-yr (for trash)(1)
Noncompactible 3.97-01 Ci/MW(e)-yr Net (expected) electrical output for Unit 2 1,080 MWe Volume of dry solid radwaste (design) for Unit 2 17,054 ft3 Volume of dry solid radwaste (expected) for Unit 2 10,234 ft3 (1)
Phillips, J., Feizollahi, F., Martineit, R., and Bell, W.
Waste Report for Reactor and Fuel-Fabrication Facility Wastes, ONWI-20/NUS-3314, NUS Corporation, March 1979 (Table 4.2-49 and pp 4-88).
NMP Unit 2 USAR Chapter 11 11.5-1 Rev. 25, October 2022 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5.1 Design Bases The process and effluent radiological monitoring and sampling systems are provided to allow determination of the content of radioactive material in various gaseous and liquid process and effluent streams. The design objective and criteria are primarily determined by the system function of either:
- 1.
Monitors/sampling required for safety, or
- 2.
Monitors/sampling required for plant operation.
11.5.1.1 Design Objectives 11.5.1.1.1 Radiation Monitors Required for Safety The main objective of radiation monitoring systems (RMS) required for safety is to initiate appropriate manual or automatic protective action to limit the potential release of radioactive materials from the reactor vessel, primary and secondary containment, and fuel storage areas if predetermined radiation levels are exceeded in major process/effluent streams, and to provide main control room personnel with radiation level indication throughout the course of an accident. Additional objectives are to have these systems available under all operating conditions including accidents, and to provide main control room personnel with an indication of the radiation levels in the major process/effluent streams, including alarm annunciation if high radiation levels are detected.
Radiation monitors provided to meet these objectives are:
- 1.
- 2.
Reactor building ventilation exhaust (above refueling floor).
- 3.
Reactor building ventilation exhaust (below refueling floor).
- 4.
Main control room air intakes.
- 5.
RHR heat exchanger service water discharge.
11.5.1.1.2 Radiation Monitors Required for Plant Operation The main objective of RMSs required for plant operation is to provide operating personnel with a measurement of the level of radioactivity in potentially-radioactive effluents and process streams. This allows demonstration of compliance with plant normal operational Technical Specifications and the Offsite Dose
NMP Unit 2 USAR Chapter 11 11.5-2 Rev. 25, October 2022 Calculation Manual by providing gross radiation level monitoring and off-line isotopic analysis of gaseous effluents including halogens and particulates. Additional objectives are to initiate valve isolation on the offgas system, containment purge, and LWS if predetermined release rates are exceeded.
Grab samples can be taken at all process and liquid effluent monitor and both gaseous effluent sample tap locations to allow laboratory determination of isotopic content. In addition to grab samples, provisions have been included for the offgas pretreatment process monitors to allow alternate sampling with local analysis capabilities (Figure 11.3-1b). Connection capability is provided at selected locations for continuous airborne monitors (CAM).
The radiation monitors provided to meet these objectives are:
- 1.
For gaseous effluent streams:
- a.
Combined radwaste/reactor building ventilation exhaust.
- b.
Plant main stack exhaust.
- 2.
For liquid effluent streams:
- a.
Liquid radwaste effluent.
- b.
Cooling tower blowdown line (circulating water system [CWS]).
- c.
Service water discharge.
- 3.
For gaseous process streams:
- a.
Offgas pretreatment.
- b.
Standby gas treatment discharge (monitors containment purge system exhaust).
- c.
Turbine building ventilation exhaust (CAM only).
- d.
Radwaste building area ventilation exhaust (CAM only).
- e.
Radwaste building equipment exhaust (CAM only).
- f.
Radwaste tank vent exhaust (CAM only).
- g.
Radwaste equipment service area (CAM only).
- h.
Drywell atmosphere monitoring (CAM).
- 4.
For liquid process streams:
NMP Unit 2 USAR Chapter 11 11.5-3 Rev. 25, October 2022
- a.
Spent fuel pool cooling and cleanup (SFC).
- b.
Turbine building closed loop cooling water (TBCLCW).
- c.
Reactor building closed loop cooling water (RBCLCW).
11.5.1.2 Design Criteria 11.5.1.2.1 Monitors Required for Safety The radiation monitors are designed to the following criteria:
- 1.
The monitors are designed to Category I criteria to withstand the effects of natural phenomena (e.g.,
earthquakes) without loss of capability to perform their functions.
- 2.
The monitors perform their intended safety function in the environment resulting from normal, abnormal, and postulated accident conditions.
- 3.
The monitors meet the reliability, testability, independence, and failure mode requirements of engineered safety features (ESF).
- 4.
The monitors provide continuous outputs in the main control room panel.
- 5.
The monitors permit checking of the operational availability of each channel during reactor operation with provision for calibration function and source checks.
- 6.
The monitors assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
- 7.
The monitors initiate prompt protective action before exceeding plant Technical Specification and the Offsite Dose Calculation Manual limits.
- 8.
The monitors provide warning of increasing radiation levels indicative of abnormal conditions by alarm annunciation.
- 9.
The monitors provide annunciation to indicate power failure or component malfunction.
- 10. The monitors register full-scale output if radiation detection exceeds full scale.
NMP Unit 2 USAR Chapter 11 11.5-4 Rev. 25, October 2022
- 11. The monitors have sensitivities and ranges compatible with anticipated radiation levels.
- 12. The monitors have safety-related power supplies.
11.5.1.2.2 Monitors Required for Plant Operation The operations design criteria are that the monitors:
- 1.
Provide continuous indication of selected radiation levels in the main control room.
- 2.
Provide warning by alarm annunciation of increasing radiation levels indicative of abnormal conditions.
- 3.
Provide annunciation to indicate power failure or component malfunction.
- 4.
Monitor a sample representative of the bulk stream or volume.
- 5.
Have provisions for calibration, function, and source checks.
- 6.
Have sensitivities and ranges compatible with anticipated radiation levels, and Offsite Dose Calculation Manual limits.
- 7.
Register full-scale output if radiation detection exceeds full scale.
The monitors that detect radioactivity in the discharges from the containment purge system (via SGTS), the liquid radwaste treatment system, and the offgas system pretreatment, have provisions to alarm and to initiate automatic closure of the discharge valve on the affected system prior to exceeding the limits specified in the Offsite Dose Calculation Manual.
11.5.2 System Description The process and effluent RMS consists of a computer-based digital radiation monitoring system (DRMS), a microprocessor-based gaseous effluent monitoring system (GEMS), and nondigital monitors supplied as part of the reactor protection system (RPS).
11.5.2.1 Computer-Based Radiation Monitoring Systems Digital Radiation Monitoring System The function of the DRMS is to measure, evaluate, and report radioactivity in process streams, in liquid effluents, and in selected plant areas (Section 12.3.4), and to annunciate and/or initiate an automatic control function for abnormal system or
NMP Unit 2 USAR Chapter 11 11.5-5 Rev. 25, October 2022 plant operating conditions. The DRMS process and liquid effluent monitors are shown in Table 11.5-1, and the DRMS area monitors are shown in Table 12.3-1. Each monitoring channel has a microprocessor located near the detector or sample panel. The DRMS computer system continuously polls the local microprocessors collecting and storing radiation levels, alarms, and status information for these monitoring channels. This information, stored in the computer, is available on demand for analysis of plant conditions, trending of radiation levels, and maintenance purposes.
All monitoring channels have two alarm states: alert radiation (alarm) and high radiation (trip). Alarms are annunciated locally at the detector and in the main control room.
Readouts for all DRMS monitors are provided locally at the monitor electronics panel and at consoles located in the radiation protection (health physics) room, the main control room, the emergency operations facility (EOF), the technical support center (TSC), and the computer room located on el 288'-
6" of the control building. An additional control panel for the safety-related monitors also is located in the main control room.
Gaseous Effluent Monitoring System The function of the GEMS is to measure, evaluate, and report radioactivity in gaseous effluents and to annunciate if release levels approach limits specified in the Offsite Dose Calculation Manual. The GEMS provides real time noble gas activity monitoring and continuous iodine and particulate sample collection for main stack and radwaste/reactor building vent releases. The information from the GEMS is manually combined with information from the meteorological computer (VAX 11/780) to generate the gaseous release calculations for RG 1.21 report generation.
GEMS readouts are provided in the main control room, the EOF, and the TSC.
11.5.2.1.1 Computer-Based Monitor Descriptions Six basic types of computer-based monitoring or sampling systems are provided, as indicated in Table 11.5-1, for the process and effluent monitoring systems: off-line gas and particulate, off-line gas, off-line liquid, CAM, on-line liquid, and off-line gaseous effluent. Diagrams of these types of monitors are shown on Figures 11.5-1 through 11.5-6. Figures 11.5-7 and 11.5-8 show the gaseous and liquid RMS. Safety and operationally-required monitors are shown on these composite diagrams.
Off-line Gas and Particulate Monitor
NMP Unit 2 USAR Chapter 11 11.5-6 Rev. 25, October 2022 The typical off-line gas and particulate monitor consists of a moving particulate filter with detector, iodine filter cartridge, gas sample chamber with detector, and associated pump and valving. Connections are available for taking grab samples of the process stream and taking samples for tritium analysis downstream of the filter units. Check sources that are remotely operated are provided with each detector to check the function of each channel periodically. Remote purging capability is provided for the gas sample chamber and sample tubing.
Detectors are shielded and designed to obtain the sensitivities and ranges indicated in Table 11.5-1. Process streams in which plateout due to condensation could be a problem have been heat traced so that particulate sampling is representative of the process stream. Plateout is also minimized by using stainless steel for sample tubing and maintaining minimum bend radii in the sample lines in accordance with ANSI N13.1-1969.
Off-line Gas Monitor The typical off-line gas monitor consists of an isokinetic sampling system, fixed particulate and charcoal filters, a gas sample chamber with detector, and associated pump with valving.
Connections are available for taking grab samples of the process streams. All filters are removable for laboratory analysis.
Check sources and purging capabilities are provided as described for the off-line gas and particulate monitor. Detector type, ranges, and alarm setpoints are given in Table 11.5-1.
Isokinetic sampling systems are used to draw samples for the off-line gas monitor. The isokinetic sampling systems for these monitors are designed in accordance with ANSI N13.1-1969.
On-line Liquid Monitor The on-line liquid monitor consists of a detector, shielding, and remotely-operated check source. The monitor is mounted to view a solid radwaste system line and shielded to obtain the sensitivities indicated in Table 11.5-1.
Off-line Liquid Monitor The typical off-line liquid monitor consists of a sample chamber with detector and associated pump, piping, and valving. The detector has a remotely-operated check source and is shielded to obtain the sensitivities indicated in Table 11.5-1. Connections for taking a grab sample from the process stream and purging the liquid sample chamber and sample tubing are provided. Heat exchangers are provided on sampling systems that contain high temperature fluids.
Off-line Gaseous Effluent Monitor Each off-line GEM consists of a noble gas detector and iodine and particulate filters with associated valving, electronics,
NMP Unit 2 USAR Chapter 11 11.5-7 Rev. 25, October 2022 and microprocessor. The system automatically determines gaseous effluent noble gas activity, provides the activity level to the PPC, stores the information, and retrieves the data on command.
Iodine and particulate filters are manually removed and analyzed in the laboratory. The system automatically purges itself and also automatically performs daily maintenance routines such as background count and diagnostics. Isokinetic sample probes are used to draw samples as previously described.
Continuous Airborne Monitor CAMs are used to monitor process ventilation systems and local area ventilation activity conditions. The monitors contain noble gas and particulate detectors. They are also equipped to accommodate removable charcoal filter cartridges which can be installed, as needed, to obtain grab samples. All CAMs are equipped with a check source and purge capability, and mounted on movable carts to allow greater flexibility in the monitored areas depending on current plant conditions or maintenance schedules.
11.5.2.1.2 Monitors Required for Safety (DRMS)
Reactor Building Ventilation Redundant off-line gas monitors are provided on the reactor building ventilation exhaust air ductwork above and below the refueling floor. These monitors function to indicate airborne levels of activity in the reactor building (Section 12.3.4). On a high-radiation alarm signal, the reactor building ventilation exhaust air is discontinued and reactor building air is recirculated, with a small fraction being diverted through the SGTS. A CAM connection is provided on the recirculation duct.
These monitors are also designed to perform their required functions under all environmental conditions (Section 3.11).
Main Control Room Air Intakes Redundant off-line gas monitors are provided at the main control room air intake. The main control room ventilation intake monitors divert the intake and recirculated air through HEPA/charcoal filter trains on high radiation. The main control room ventilation intake monitors indicate airborne radiation levels present in the main control room (Section 12.3.4). These monitors are also designed to perform their required functions under all environmental conditions (Section 3.11.1).
RHR Heat Exchanger Service Water Discharge An off-line liquid monitor is located in the service water effluent on each of the two RHR heat exchangers. These monitors function to detect and alarm on contamination of the service water effluent due to leaks in the heat exchangers following a loss-of-coolant accident (LOCA) or under normal operating
NMP Unit 2 USAR Chapter 11 11.5-8 Rev. 25, October 2022 conditions. These monitors are also designed to perform their required function under all environmental conditions (Section 3.11.1).
11.5.2.1.3 Monitors Required for Plant Operations Liquid Effluent Monitors (DRMS)
- 1.
Cooling tower blowdown line (CWS) monitor - The cooling tower blowdown monitor detects and alarms on high levels of radioactivity in the plant CWS effluent to the discharge bay.
- 2.
Liquid radwaste effluent monitor - The liquid radwaste effluent monitor terminates a LWS release if discharge concentrations approach the Offsite Dose Calculation Manual limits.
- 3.
Service water discharge effluent monitors - Two service water discharge monitors are installed in the SWP effluent lines upstream of the discharge bay. These monitors will alert Operators if discharge concentrations approach Offsite Dose Calculation Manual limits.
Gaseous Effluent Monitors
- 1.
Radwaste/Reactor Building Vent Ventilation exhaust effluents from the reactor and radwaste buildings are combined and released from a single vent between the reactor building and turbine building. This release path is monitored by an off-line noble gas GEM which also continuously samples for iodine and particulate activity. The primary function of this monitor is to collect data for RG 1.21 report generation. Major process streams exhausted through the radwaste/reactor building vent include reactor building ventilation exhaust air, radwaste building area ventilation exhaust air, and radwaste building equipment exhaust air. The radwaste/reactor building vent monitor has extended range to cover post-accident monitoring requirements.
- 2.
Main Plant Stack Exhaust Effluent from the main stack (consisting of turbine building ventilation system exhaust air, containment purge exhaust, SGTS exhaust, condenser mechanical vacuum pump discharge, gland seal condenser gaseous discharge, and offgas system exhaust) is monitored by an off-line noble gas GEM which also continuously samples for iodine and particulate activity. The primary function of this monitor is to collect data for RG 1.21 report generation. The main stack monitor has the necessary range to cover possible activities released during and following an accident.
NMP Unit 2 USAR Chapter 11 11.5-9 Rev. 25, October 2022 Process Gaseous Monitors (DRMS)
- 1.
Standby Gas Treatment Discharge Monitor An off-line gaseous monitor is installed on the discharge of the SGTS which isolates the normal containment purge system on a high radiation alarm.
- 2.
Drywell Atmosphere Monitors Redundant off-line gas and particulate monitors are provided to monitor drywell airborne activity levels and detect reactor coolant pressure boundary (RCPB) leakage in accordance with RG 1.45 requirements. The two drywell monitors pull samples through sampling trees located in the drywell and return the samples to the drywell. The sampling trees provide representative samples of drywell air by extracting samples from various elevations.
- 3.
Offgas Pretreatment Monitors The offgas process flow upstream of the charcoal adsorbers is monitored by off-line gaseous monitors equipped with iodine and particulate sampling capabilities. The offgas pretreatment monitors isolate the offgas effluent upon receipt of a high radiation signal.
- 4.
Turbine Building Ventilation A connection tap is provided in the system exhaust ductwork for a CAM.
- 5.
Radwaste Building Area Exhaust and Tank Vent Exhaust Monitors A connection tap for a CAM is provided in the exhaust ductwork of each of the above ventilation subsystems upstream of the filtration units.
Process Liquid Monitors (DRMS)
The following process streams are monitored by off-line liquid monitors for detection of radiation levels:
- 1.
SFC system pumps discharge.
- 2.
- 3.
The SFC system monitor provides indication of radioactivity levels in the spent fuel pool water and will warn Operators of failures in spent fuel elements stored in the pool. The turbine
NMP Unit 2 USAR Chapter 11 11.5-10 Rev. 25, October 2022 and RBCLCW monitors detect and alarm following increases in the radioactivity level of the water in these systems.
A process radiation monitor was not installed on the secondary side of the alternate decay heat (ADH) system because the possibility of secondary side contamination has been analyzed and found not to be a credible event. The basis for this conclusion is that the secondary side operates at a higher pressure than the primary side, a switch shuts down the primary side pumps on low secondary to primary side differential pressure, and plate heat exchanger design results in primary side leakage being to the reactor building rather than into the secondary side. As an added precaution and because the secondary side of ADH interfaces with a radioactive system, the ADH secondary side is routinely sampled for radioactivity.
11.5.2.2 Noncomputer-Based Process Radiation Monitoring System Main Steam Line Radiation Monitoring System This system monitors the gamma radiation level exterior to the MSLs. The normal radiation level is produced primarily by coolant activation gases plus smaller quantities of fission gases being transported with the steam. In the event of a gross release of fission products from the core, this monitoring system provides annunciation in the control room.
The system consists of four redundant instrument channels. Each channel consists of a local on-line steam detector (gamma-sensitive ion chamber) and a main control room radiation monitor with an auxiliary trip unit. A diagram of this type of monitor is shown on Figure 11.5-4.
The detectors are physically located near the MSLs just downstream of the outboard main steam isolation valves (MSIV).
The detectors are geometrically arranged so that this system is capable of detecting significant increases in radiation level with any number of MSLs in operation. Table 11.5-1 lists the range of the detectors.
Each radiation monitor has four trip circuits: two upscale (high-high and high), one downscale (low), and one inoperative.
Each trip is visually displayed locally on the affected radiation monitor panel. A high-high or inoperative trip in the radiation monitor results in a one-out-of-two twice channel trip. This initiates mechanical vacuum pump shutdown, and discharge valve closure. A high-high trip or inoperative trip actuates a MSL high-high radiation main control room annunciator. A high trip actuates a MSL high radiation main control room annunciator. A downscale trip actuates a MSL downscale main control room annunciator common to all channels.
High and low trips do not result in a channel trip. Each radiation level is visually displayed locally.
NMP Unit 2 USAR Chapter 11 11.5-11 Rev. 25, October 2022 11.5.2.3 Calibration, Maintenance, Inspection, and Tests Calibration, inspection, and tests of monitors required for safety (except for the RHR heat exchanger service water discharge monitors whose testing is described in Section 7.5.3),
effluent monitors, the offgas pretreatment monitor, the SGTS discharge monitor, and the drywell atmosphere monitors are performed in accordance with the plant Technical Specifications and the Offsite Dose Calculation Manual. Calibration, inspection, and tests of monitors not covered by the Technical Specifications' surveillance program or the Offsite Dose Calculation Manual's surveillance program are conducted in accordance with the manufacturer's recommendations. Maintenance of all process and effluent monitors is performed in accordance with appropriate plant procedures.
The design of all process and effluent radiation monitors meets the applicable criteria cited in RG 4.15. Each monitor is equipped with purge capability for analysis of background radiation and check or keep-alive source(s) for determination of changes in detector counting rates, efficiencies, or energy resolution. Effluent monitoring systems are designed using the guidelines established in ANSI 13.10. Calibration standards and frequency are described in Section 11.5.2.3.2.
The quality assurance (QA) criteria is consistent with the function of the monitor. Safety-related monitors are procured and designed to 10CFR50 Appendix B criteria. Nonsafety-related monitors are designed and procured under standards that meet the criteria established in RG 1.143.
11.5.2.3.1 Inspection and Tests Inspection and testing of the monitors specified in Section 11.5.2.3 is described in Station procedures and either the Technical Specifications or the Offsite Dose Calculation Manual.
11.5.2.3.2 Calibration The radiation monitor's calibration is traceable to the National Institute of Standards and Technology (NIST) and is accurate to at least 15 percent. The source-detector geometry during initial calibration is identical to the sample-detector geometry in actual use. Secondary standards that were counted in a reproducible geometry during the initial calibration are used for calibration after installation. Where applicable, each monitor (except for the RHR heat exchanger service water discharge monitors whose calibration is described in Section 7.5.3) is calibrated in accordance with the frequencies provided in the plant Technical Specifications or the Offsite Dose Calculation Manual during plant operation or during the refueling outage if the detector is not readily accessible.
11.5.2.3.3 Maintenance
NMP Unit 2 USAR Chapter 11 11.5-12 Rev. 25, October 2022 The channel detector, electronics, and recorder are serviced and maintained in accordance with manufacturers' recommendations to ensure reliable operations. Such maintenance includes cleaning, lubrication, and assurance of free movement of the recorder in addition to the replacement or adjustment of any components required after performing a test or calibration check. If any work is performed that would affect the calibration, a recalibration is performed at the completion of the work.
11.5.2.4 Sampling Section 9.3.2 discusses various process and effluent samples periodically taken for chemical and radiochemical analysis.
Liquid process and effluent samples are periodically taken and monitored for radioactivity. Those provisions for sampling not covered in Section 9.3.2 are described in the individual system design sections. Sampling of these fluid systems is via local sampling connections. The Technical Specifications and the Offsite Dose Calculation Manual describe various liquid samples and the analysis required, including the sampling frequencies.
Additionally, the process and effluent radiological monitoring system can provide grab samples that are used to locate manually a specific source of high radioactivity when a continuous radiological monitor signals a high radioactivity alarm in the main control room. Tritium in the plant areas is determined on the basis of representative grab samples collected from the effluent points or ventilation exhaust ducts. Grab samples are obtained from locations indicated in Table 11.5-2 and the samples are analyzed. In addition to grab samples, provisions have been included for the offgas pretreatment process monitors to allow alternate sampling with local analysis capabilities (Figure 11.3-1b).
11.5.3 Effluent Monitoring and Sampling All potentially-radioactive gaseous and liquid effluent discharge paths are either continuously monitored or routinely sampled for radiation level during discharge (Section 11.5.2).
Solid waste shipping containers are monitored with gamma-sensitive portable survey instruments. The following gaseous effluent paths are sampled and monitored:
- 1.
Plant main stack exhaust.
- 2. Combined radwaste/reactor building ventilation exhaust.
The following liquid effluent paths are sampled and monitored:
- 1.
LWS effluent.
NMP Unit 2 USAR Chapter 11 11.5-13 Rev. 25, October 2022
- 2.
CWS cooling tower blowdown line.
- 3.
SWP discharge.
All monitor ranges are listed in Table 11.5-1.
An isotopic analysis is performed periodically on samples obtained from each liquid effluent release path to verify the adequacy of effluent processing to meet the discharge limits to unrestricted areas.
This effluent monitoring and sampling program is comprehensive and provides the information for the effluent measuring and reporting programs required by 10CFR50 Section 36a, Appendix A, GDC 64, and Appendix I and RG 1.21 in semiannual reports to the NRC. The frequency of the periodic sampling and analysis described in the Technical Specifications, the Offsite Dose Calculation Manual, and in TRM Section 3.4.8 is a minimum and is increased if effluent levels approach Technical Specification or Offsite Dose Calculation Manual limits. Activity level of noble gas effluents is continuously monitored by off-line monitors.
Iodine and particulate in the gaseous effluents are continuously sampled with isotopic content determined by laboratory analysis.
All potentially-significant radioactive discharge paths are equipped with a control system to isolate the discharge automatically on indication of a high radiation level. These include:
- 1.
Offgas pretreatment.
- 2.
SGTS discharge (isolates containment purge system).
- 3.
Reactor building ventilation exhaust.
- 4.
Liquid radwaste effluent.
The effluent isolation functions for each monitor are given in Table 11.5-1 and in Section 11.5.2.
Radiation levels in radioactive and potentially-radioactive process streams are monitored by the process and effluent monitors given in Table 11.5-1.
Airborne radioactivity in the fuel-handling area and the radwaste building is detected by CAM and area radiation monitors. Airborne radioactivity in the drywell is detected by the drywell atmosphere monitors and the SGTS discharge monitor which isolates the containment purge on high radioactivity.
These monitors are also described in Section 12.3.4 since they are used to monitor in-plant airborne radioactivity to protect plant personnel. The area RMS is also described in Section 12.3.4. A system level/qualitative-type failure modes and effects analysis (FMEA) of the MSL radiation monitoring is provided in Appendix 15A. Originally, the FMEA for other
NMP Unit 2 USAR Chapter 11 11.5-14 Rev. 25, October 2022 safety-related radiation monitors was contained in the Unit 2 FMEA document, which is historical. FMEAs for plant systems are now performed and controlled by the design process.
NMP Unit 2 USAR Chapter 11 11.5-15 Rev. 25, October 2022 TABLE 11.5-1 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS Monitor Location Monitor Type Range(4)
Isotope Trip/High Setpoint Function Monitors Required for Safety Reactor building ventilation above and below refueling floor (2HVR*CAB14A,B; 2HVR*CAB32A,B)
Main control room intake (2HVC*CAB18A,B,C,D)
RHR heat exchanger service water (2SWP*CAB23A,B)
(2MSS*RT46A,B,C,D)
Offline gaseous Offline gaseous Offline liquid Online steam 10-7 to 10-1 uCi/cc 10-7 to 10-1 uCi/cc 10-7 to 10-1 uCi/cc 1-106 mr/hr Xe-133, Kr-85 Xe-133, Kr-85 Cs-137 N-16 Allowable Value in Tech. Spec.
Allowable Value in Tech. Spec.
3.3 x 10-4 uCi/cc Allowable Value in Tech. Spec.
Monitors radiation levels in the reactor building ventilation system.
Isolates reactor building(5)
Monitors incoming control room air; activates Category I HEPA/charcoal filters(5)
Monitors service water effluent from heat exchangers for contamination(6)
Monitors main steam lines for fuel damage and carryover to turbine building, trips mechanical vacuum pumps and associated isolation valve.
Monitors Required for Plant Operation Drywell and containment atmosphere(1)
(2CMS*CAB10A,B)
Service water system discharge monitors (2SWP*CAB146A,B)
Radwaste/reactor building vent(2)
(2RMS-SKD180A&B, 2RMS-PNL180A&B)
Offline gaseous Particulate Offline liquid Offline gaseous 10-7 to 10-1 uCi/cc 10-11 to 10-5 uCi/cc 10-7 to 10-1 uCi/cc 10-6 to 105 uCi/cc Xe-133, Kr-85 I-131 Cs-137 Xe-133, Kr-85(7)
N2-RTP-129 N2-RTP-129 ODCM ODCM Monitors drywell for airborne radiation - RCPB leak detection(5)
Monitors service water system discharge(5)
Monitors reactor and radwaste building ventilation effluent releases for RG 1.21 report generation(5)
NMP Unit 2 USAR Chapter 11 11.5-16 Rev. 25, October 2022 TABLE 11.5-1 (Cont'd.)
Monitor Location Monitor Type Range(4)
Isotope Trip/High Setpoint Function Main stack exhaust(2)
(2RMS-SKD170A&B, 2RMS-PNL170A&B)
Fuel pool cooling pumps discharge (2SFC-CAB142)
Turbine plant closed loop cooling water (2CCS-CAB152)
Liquid radwaste effluent (2LWS-CAB206)
Cooling tower blowdown line (2CWS-CAB157)
Standby gas treatment discharge (2GTS-CAB105)
Reactor plant closed loop cooling water
- a.
SFC heat exchanger cooling water discharge (2CCP-CAB115)
- b.
RWCU heat exchanger cooling water discharge (2CCP-CAB131)
Radwaste building equipment, tank vents and area ventilation exhausts
- a. Radwaste equipment exhaust (2HVW-CAB195)
Offline gaseous Offline liquid Offline liquid Offline liquid Offline liquid Offline gaseous Offline liquid Offline liquid CAM 10-6 to 105 uCi/cc 10-7 to 10-1 uCi/cc 10-7 to 10-1 uCi/cc 10-7 to 10-1 uCi/cc 10-7 to 10-1 uCi/cc 10-6 to 10-1 uCi/cc 10-7 to 10-1 uCi/cc 10-7 to 10-1 uCi/cc gas: 10-7 to 10-1 uCi/cc part: 10-11 to 10-5 uCi/cc Xe-133, Kr-85(7)
Cs-137 Cs-137 Cs-137 Cs-137 Xe-133, Kr-85 Cs-137 Cs-137 Kr-85, Xe-133 I-131 ODCM 2.2 x 10-2 uCi/cc
<3.0 x 10-3 uCi/cc ODCM ODCM Allowable Value in Tech. Spec.
1.6 x 10-2 uCi/cc
<9.1 x 10-3 uCi/cc 2.3 x 10-2 uCi/cc NA Monitors content of effluent releases for RG 1.21 report generation(5)
Monitors fuel pool cooling water for fuel damage(6)
Monitors CCS for detection of radioactive in-leakage(6)
Monitors radwaste effluent discharge to environment isolates system at trip level(5)
Monitors circulating water system discharge to environment(5)
Monitors containment purge exhaust; isolates primary containment discharge lines(5)
Monitors each CCP loop for detection of radioactive in-leakage(6)
Monitors radiation levels in radwaste building equipment, tank vents, and area ventilation exhaust prior to filtration and exhaust to the radwaste/reactor building vent(6)
NMP Unit 2 USAR Chapter 11 11.5-17 Rev. 25, October 2022 TABLE 11.5-1 (Cont'd.)
Monitor Location Monitor Type Range(4)
Isotope Trip/High Setpoint Function
- b.
Radwaste tank vents (2HVW-CAB196)
- c.
Radwaste building ventilation (2HVW-CAB197)
- d.
Radwaste equipment service area (2HVW-CAB199)
Turbine building ventilation exhaust (2HVT-CAB206)
Offgas pretreatment (2OFG-CAB13A,B)
CAM CAM CAM CAM Offline gaseous gas: 10-7 to 10-1 uCi/cc part: 10-11 to 10-5 uCi/cc gas: 10-7 to 10-1 uCi/cc part: 10-11 to 10-5 uCi/cc gas: 10-7 to 10-1 uCi/cc part: 10-11 to 10-5 uCi/cc gas: 10-7 to 10-1 uCi/cc part: 10-11 to 10-5 uCi/cc 10-3 to 102 uCi/cc Kr-85, Xe-133 I-131 Kr-85, Xe-133 I-131 Kr-85, Xe-133 I-131 Kr-85, Xe-133 I-131 Xe-133, Kr-85 2.3 x 10-2 uCi/cc NA 2.3 x 10-2 uCi/cc NA 2.3 x 10-2 uCi/cc NA 4.0 x 10-2 uCi/cc NA ODCM Monitors radiation levels in turbine building ventilation prior to exhaust to the main stack(6)
Monitors process before charcoal adsorbers; isolates offgas discharge(5)
(1)
RCPB leak detection in accordance with RG 1.45.
(2)
PAM - Post-accident monitor.
(3)
Monitors are not part of computer-based radiation monitoring systems, supplied separately as part of the process system.
(4)
Design range is stated for DRMS and main steam monitors. Due to ambient conditions, the low end of the range may not be available or required. For GEMS monitors, the low end of the ranges required by NUREG-0473 and the high end of the ranges required by RG 1.97 are given.
(5)
On actuation of high trip, action is required in accordance with the Technical Specifications.
(6)
On actuation of high trip, new trip setpoint and/or system corrective actions will be specified.
(7)
Kr-85 used to establish calibration curve and subsequently removed from analysis during normal operation to avoid erroneous identification.
NMP Unit 2 USAR Chapter 11 11.5-18 Rev. 25, October 2022 TABLE 11.5-2 GRAB SAMPLES FOR RADIOLOGICAL ANALYSIS Sample Point Location (No.)
Grab Sample at Sample Station Local Grab Sample Grab Sample at Radiation Monitor Reactor Steam Supply System Reactor recirculation system pump discharge Main steam line X
X Reactor Water Cleanup System Common filter/demineralizer influent Individual filter/demineralizer effluents(4)
X X
Fuel Pool Cooling and Cleanup System Pump discharge Common filter influent Individual filter effluents(2)
Individual heat exchanger effluents(2)
X X
X X
Reactor Building Closed Loop Cooling Water Cooling water sample (outlet of RWCU and SFC heat exchangers)
X Turbine Building Closed Loop Cooling Water Cooling water sample (common outlet of radwaste system exchangers)
X Residual Heat Removal System Individual heat exchanger outlet (service water)(2)
Individual heat exchanger outlet (RHR)(2)
X X
Control Rod Drive System Common CRD filter effluent X
High-Pressure Core Spray System Test return line to condensate storage tank X
NMP Unit 2 USAR Chapter 11 11.5-19 Rev. 25, October 2022 TABLE 11.5-2 (Cont'd.)
Sample Point Location (No.)
Grab Sample at Sample Station Local Grab Sample Grab Sample at Radiation Monitor Radwaste System Individual waste collector tank pump effluents(3)
Individual demineralizer effluents(2)
Filtrate pump effluent Individual filter effluents(2)
Demineralizer (acid influent)
Demineralizer (caustic influent)
Individual recovery sample pump effluents(2)
Individual floor drain collector pump effluents(2)
Floor drain filter effluent pump discharge Liquid radwaste final discharge Regenerant waste pump effluents(2)
Regenerant recirculation pump suction and discharge(2)
Phase separator tank pump discharge Waste evaporator recirculation pump suction and discharge(2)
Waste evaporator distillate Individual waste sample pumps, discharge(2)
Regenerant evaporator distillate Common discharge floor drain collector surge pumps Common discharge waste collector surge pumps Individual radwaste auxiliary steam cooler effluents(2)
X X
X X
X X
X X
X X
X X
X X
X X
X X
X X
X X
Water Treating System Dilute acid effluent Dilute caustic effluent Waste water effluent X
X X
Condensate Demineralizer System Common demineralizer influent Common demineralizer effluent Resin hold tank effluent Individual demineralizers effluent(9)
Resin mix tank effluent Cation regeneration tank effluent Anion regeneration tank effluent Recovered acid tank effluent Regeneration system effluent Low conductivity waste tank effluent Demineralizer waste neutralizing tank effluent Dilute acid effluent Recovered caustic tank effluent Dilute caustic effluent Recovered water sump effluent X
X X
X X
X X
X X
X X
X X
X X
X X
NMP Unit 2 USAR Chapter 11 11.5-20 Rev. 25, October 2022 TABLE 11.5-2 (Cont'd.)
Sample Point Location (No.)
Grab Sample at Sample Station Local Grab Sample Grab Sample at Radiation Monitor Condensate Makeup and Drawoff System Condensate transfer line X
Makeup Water System Demineralizer water transfer line X
Condensate System Condensate pump discharge Condenser hotwells(6)
LP heater drains(3)
Common effluent fourth-point heaters LP heater string common effluent X
X X
X X
Reactor Feedwater System Feedwater (after last heater)
X Circulating Water System Effluent (blowdown line)
X X
Auxiliary Steam System Auxiliary boiler (steam outlet)
Feedwater (pump discharge)
Auxiliary boiler (blowdown)
Auxiliary boiler recirc pump seal heat exchanger outlet and sample cooler discharge (Service Water)(4)
X X
X X
Sealing Steam System Individual clean steam reboiler outlets X
Reactor Building Ventilation System Reactor/radwaste building ventilation exhaust Main plant stack exhaust Containment atmosphere Containment purge X
X X
X Standby Gas Treatment System SGTS effluent X
Control Building Ventilation System Main control room intakes X
X
NMP Unit 2 USAR Chapter 11 11.5-21 Rev. 25, October 2022 TABLE 11.5-2 (Cont'd.)
Sample Point Location (No.)
Grab Sample at Sample Station Local Grab Sample Grab Sample at Radiation Monitor Radwaste Building Ventilation System Ventilation exhaust Radwaste tank vents X
X X
Turbine Building Ventilation System Ventilation Mechanical vacuum pump discharge Offgas pretreatment(2)
Turbine gland seal discharge X
X X
X Service Water System Final discharge X
X Storm, Underdrain Water, and Site Sewage Systems Final discharge X
Alternate Decay Heat Removal System Discharge to storm drain X
NOTE: See Section 9.3.2 for details regarding the reactor, turbine, and radwaste sample systems.
NMP Unit 2 USAR Chapter 11 Rev. 25, October 2022 APPENDIX 11A RADIOLOGICAL DOSES
NMP Unit 2 USAR APPENDIX 11A RADIOLOGICAL DOSES TABLE OF CONTENTS Section Title Chapter 11 11A-i Rev. 25, October 2022 11A.1
SUMMARY
OF ANNUAL RADIATION DOSES 11A.2 COST-BENEFIT ANALYSIS LIST OF TABLES Table Number Title 11A.1-1 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM LIQUID EFFLUENTS 11A.1-2 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM LIQUID EFFLUENTS 11A.1-3 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM LIQUID EFFLUENTS 11A.1-4 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM LIQUID EFFLUENTS 11A.1-5 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS At Maximum Residence Location 11A.1-6 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS At Maximum Residence Location 11A.1-7 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS At Maximum Residence Location 11A.1-8 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS At Maximum Residence Location 11A.1-9 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS At Maximum Cow Location
NMP Unit 2 USAR APPENDIX 11A RADIOLOGICAL DOSES LIST OF TABLES (contd.)
Table Number Title Chapter 11 11A-ii Rev. 25, October 2022 APPENDIX 11A LIST OF TABLES (Cont'd.)
Table Number Title 11A.1-10 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS At Maximum Cow Location 11A.1-11 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS At Maximum Cow Location 11A.1-12 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS At Maximum Cow Location 11A.1-13 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS At Maximum Beef Animal Location 11A.1-14 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS At Maximum Beef Animal Location 11A.1-15 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS At Maximum Beef Animal Location 11A.1-16 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS At Maximum Beef Animal Location 11A.1-17 COMPARISON OF MAXIMUM CALCULATED DOSES FROM UNIT 2 WITH APPENDIX I DESIGN OBJECTIVES 11A.1-18 CALCULATED ANNUAL DOSES FOR POPULATION WITHIN 80-KM (50-MI) RADIUS 11A.1-19 CALCULATED POPULATION DOSE COMMITMENT (Contiguous U.S. Population Dose)
NMP Unit 2 USAR APPENDIX 11A RADIOLOGICAL DOSES LIST OF TABLES (contd.)
Table Number Title Chapter 11 11A-iii Rev. 25, October 2022 11A.2-1 BASE CASE ANNUAL POPULATION DOSES DUE TO LIQUID EFFLUENTS 11A.2-2 BASE CASE ANNUAL POPULATION DOSES DUE TO GASEOUS EFFLUENTS
NMP Unit 2 USAR Chapter 11 11A.1-1 Rev. 25, October 2022 APPENDIX 11A RADIOLOGICAL DOSES 11A.1
SUMMARY
OF ANNUAL RADIATION DOSES The calculated annual radiation doses to maximum individuals from normal operation of Unit 2 are presented in Tables 11A.1-1 through 11A.1-16. Table 11A.1-17 demonstrates that the calculated annual radiation doses are below the design objectives of 10CFR50 Appendix I. In providing guidance for the implementation of Appendix I, the NRC has made use of the maximum exposed individual approach. Maximum individuals are characterized as maximum with regard to food consumption, occupancy, and other usage of the region in the vicinity of the plant site and, as such, represent individuals with reasonable deviations from the average individual considered representative of the population in general.
For gaseous radioactive releases the analyzed pathways included:
standing on contaminated ground, ingestion of vegetation, and inhalation of and submersion in gaseous effluents. These pathways were considered for all of the resident locations.
Resident locations having cows, goats, and meat animals were also analyzed for ingestion of cow milk, goat milk, and beef meat, respectively. Additionally, the doses associated with ingestion of deer were conservatively added to all resident locations analyzed.
The calculated organ dose due to radioiodines and particulates is 2.8E+00 mRem. This represents the dose to the thyroid of an infant living at the residence location 2,350 m east-southeast of the site. The majority of this dose was due to consumption of cow milk. The highest calculated external annual doses to the total body and skin from immersion in noble gases at an occupied location were 3.9E-02 and 8.3E-02 mRem, respectively.
These occurred at the residence location 1,693 m east of the site.
The highest calculated beta and gamma air doses at an unoccupied location from noble gas releases were 5.3E-02 and 7.9E-02 mrad, respectively. These occurred at the EAB 1,603 m east of the site.
For liquid releases, the maximum individual consumed fish whose principal habitat was assumed to be the edge of the initial mixing zone. This location was also conservatively used in calculating doses from swimming and boating.
The calculated annual doses to the population residing within an 80-km radius of the site are presented in Table 11A.1-18.
Population doses were calculated for a projected population of 1.2 million residing within 80 km of the site in the year 2010.
NMP Unit 2 USAR Chapter 11 11A.1-2 Rev. 25, October 2022 Gaseous pathways considered in the population dose analysis included: inhalation, exposure to ground deposits, ingestion of vegetation, cow milk, and beef meat, and submersion in noble gases. The calculated annual dose to the population was 0.6 man-Rem/yr total body(1) and 3.3 man-Rem/yr thyroid(2).
Liquid pathways considered in the population dose analysis included: ingestion of potable water and fish, shoreline recreation, and swimming and boating. The calculated annual dose to the population was 4.3 man-Rem/yr total body and 1.2E-01 man-Rem/yr thyroid.
In addition to the 80-km (50-mi) radius population doses, population doses associated with the export of food crops produced within the 80-km (50-mi) region and the atmospheric and hydrospheric transport of the more mobile effluent species, such as noble gases, tritium, and carbon-14, were calculated.
These calculated annual gaseous and liquid doses to the contiguous U.S. population are presented in Table 11A.1-19. For liquid effluents, the calculated doses to the contiguous U.S.
population were 4.3 man-Rem/yr total body and 1.2E-01 man-Rem/yr thyroid. For gaseous effluents, the calculated doses to the contiguous U.S. population were 35.57 man-Rem/yr total body(3) and 43.04 man-Rem/yr thyroid(4).
(1),(2)
Refer to the power uprate footnotes (3 and 4) and the revised dose values shown in Table 11A.1-18.
(3),(4)
Refer to the power uprate footnote (3) shown in Table 11A.1-19.
NMP Unit 2 USAR Chapter 11 11A.1-3 Rev. 25, October 2022 TABLE 11A.1-1 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM LIQUID EFFLUENTS (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Potable water 3.5E-04 0.0 1.7E-04 4.1E-04 2.9E-04 2.6E-04 2.1E-04 2.0E-04 Fish consumption 5.8E-01 0.0 7.0E-01 8.2E-01 3.6E-03 2.7E-01 8.9E-02 4.3E-02 Shoreline recreation 2.3E-05 2.6E-05 2.3E-05 2.3E-05 2.3E-05 2.3E-05 2.3E-05 2.3E-05 Fresh vegetation 5.9E-05 0.0 4.4E-05 7.5E-05 2.6E-05 3.6E-05 2.2E-05 2.0E-05 Stored vegetation 4.1E-04 0.0 3.0E-04 5.2E-04 1.1E-04 2.5E-04 1.6E-04 1.4E-04 Duck consumption 1.3E-04 0.0 2.1E-03 2.2E-04 1.4E-06 4.7E-05 6.4E-06 2.9E-04 Swimming exposure 3.3E-05 4.2E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 Boating exposure 3.3E-05 4.2E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 Total dose 5.8E-01 1.1E-04 7.0E-01 8.2E-01 4.1E-03 2.7E-01 8.9E-02 4.4E-02 NOTE:
3.5E-04 = 3.5x10-4 (1)
All annual doses shown in Table 11A.1-1 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 5.35.
NMP Unit 2 USAR Chapter 11 11A.1-4 Rev. 25, October 2022 TABLE 11A.1-2 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM LIQUID EFFLUENTS (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Potable water 2.2E-04 0.0 1.6E-04 3.5E-04 2.2E-04 3.2E-04 1.6E-04 1.4E-04 Fish consumption 3.3E-01 0.0 7.6E-01 8.4E-01 3.3E-03 2.8E-01 1.1E-01 3.3E-02 Shoreline recreation 1.3E-04 1.5E-04 1.3E-04 1.3E-04 1.3E-04 1.3E-04 1.3E-04 1.3E-04 Fresh vegetation 3.2E-05 0.0 3.9E-05 6.3E-05 1.8E-05 5.1E-05 1.7E-05 1.3E-05 Stored vegetation 4.4E-04 0.0 5.3E-04 8.7E-04 1.5E-04 5.6E-04 2.4E-04 1.8E-04 Duck consumption 1.0E-04 0.0 1.8E-03 1.8E-04 9.9E-07 3.0E-04 6.0E-06 1.8E-04 Swimming exposure 3.3E-05 4.2E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 Boating exposure 3.3E-05 4.2E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 3.3E-05 Total dose 3.3E-01 2.3E-04 7.6E-01 8.4E-01 3.9E-03 2.8E-01 1.1E-01 3.4E-02 NOTE:
2.2-04 = 2.2x10-4 (1)
All annual doses shown in Table 11A.1-2 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 5.34.
NMP Unit 2 USAR Chapter 11 11A.1-5 Rev. 25, October 2022 TABLE 11A.1-3 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM LIQUID EFFLUENTS (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Potable water 3.3E-04 0.0 4.6E-04 6.9E-04 4.7E-04 3.9E-04 3.0E-04 2.5E-04 Fish consumption 1.4E-01 0.0 9.5E-01 7.5E-01 3.3E-03 2.4E-01 8.5E-02 1.4E-02 Shoreline recreation 2.6E-05 3.1E-05 2.6E-05 2.6E-05 2.6E-05 2.6E-05 2.6E-05 2.6E-05 Fresh vegetation 2.5E-05 0.0 6.8E-05 7.9E-05 2.4E-05 3.4E-05 2.0E-05 1.4E-05 Stored vegetation 4.7E-04 0.0 1.3E-03 1.5E-03 2.4E-04 6.3E-04 3.8E-04 2.6E-04 Duck consumption 1.6E-04 0.0 3.4E-03 2.5E-04 1.5E-06 4.2E-05 7.1E-06 1.1E-04 Swimming exposure 1.8E-05 2.4E-05 1.8E-05 1.8E-05 1.8E-05 1.8E-05 1.8E-05 1.8E-05 Boating exposure 1.9E-05 2.4E-05 1.9E-05 1.9E-05 1.9E-05 1.9E-05 1.9E-05 1.9E-05 Total dose 1.4E-01 7.9E-05 9.6E-01 7.5E-01 4.1E-03 2.4E-01 8.6E-02 1.5E-02 NOTE:
3.3E-04 = 3.3x10-4 (1)
All annual doses shown in Table 11A.1-3 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 5.34.
NMP Unit 2 USAR Chapter 11 11A.1-6 Rev. 25, October 2022 TABLE 11A.1-4 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM LIQUID EFFLUENTS (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Potable water 2.9E-04 0.0 4.7E-04 7.9E-04 6.0E-04 3.9E-04 3.0E-04 2.5E-04 Total dose 2.9E-04 0.0 4.7E-04 7.9E-04 6.0E-04 3.9E-04 3.0E-04 2.5E-04 NOTE:
2.9E-04 = 2.9x10-4 (1)
All annual doses shown in Table 11A.1-4 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 5.32.
NMP Unit 2 USAR Chapter 11 11A.1-7 Rev. 25, October 2022 TABLE 11A.1-5 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS*
At Maximum Residence Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 7.9-03 9.2-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 Inhalation 1.6-04 0.0 1.6-04 2.4-04 1.3-02 3.0-04 2.8-04 2.1-04 Fresh vegetation 7.5-04 0.0 1.7-03 1.1-03 1.3-01 1.2-03 2.1-04 8.0-04 Stored vegetation 2.7-03 0.0 6.3-03 3.2-03 4.5-03 1.8-03 1.1-03 1.8-03 Deer 1,603 m east 1.4-04 0.0 1.7-04 1.9-04 3.2-04 9.2-05 3.1-05 3.3-04 Total dose 1.2-02 9.2-03 1.6-02 1.3-02 1.6-01 1.1-02 9.5-03 1.1-02 Analysis performed at maximum residence location is 4,106 m (13,471 ft) east.
NOTE:
7.9-03 = 7.9x10-3 (1)
All annual doses shown in Table 11A.1-5 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated value represents the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-8 Rev. 25, October 2022 TABLE 11A.1-6 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS*
At Maximum Residence Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 7.9-03 9.2-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 Inhalation 1.8-04 0.0 2.2-04 2.9-04 1.7-02 3.8-04 3.8-04 2.3-04 Fresh vegetation 5.4-04 0.0 1.6-03 1.0-03 1.1-01 1.1-02 1.9-04 5.8-04 Stored vegetation 3.5-03 0.0 1.1-02 5.7-03 7.3-03 6.0-02 2.0-03 2.7-03 Deer 1,603 m east 7.8-05 0.0 1.4-04 1.5-04 2.4-04 7.6-03 2.7-05 1.9-04 Total dose 1.2-02 9.2-03 2.1-02 1.5-02 1.4-01 8.7-02 1.0-02 1.2-02 Analysis performed at maximum residence location is 4,106 m (13,471 ft) east.
NOTE:
7.9-03 = 7.9x10-3 (1)
All annual doses shown in Table 11A.1-6 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated value represents the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-9 Rev. 25, October 2022 TABLE 11A.1-7 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS*
At Maximum Residence Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 7.9-03 9.2-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 Inhalation 1.8-04 0.0 3.0-04 2.8-04 2.1-02 3.6-04 3.3-04 1.9-04 Fresh vegetation 7.1-04 0.0 2.9-03 1.4-03 1.6-01 1.4-03 3.2-04 5.3-04 Stored vegetation 5.7-03 0.0 2.8-02 1.1-02 1.5-02 6.2-03 4.5-03 4.6-03 Deer 1,603 m east 8.6-05 0.0 2.5-04 2.0-04 3.6-04 9.5-05 4.0-05 1.2-04 Total dose 1.5-02 9.2-03 3.9-02 2.1-02 2.0-01 1.6-02 1.3-02 1.3-02 Analysis performed at maximum residence location is 4,106 m (13,471 ft) east.
NOTE:
7.9-03 = 7.9x10-3 (1)
All annual doses shown in Table 11A.1-7 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-10 Rev. 25, October 2022 TABLE 11A.1-8 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS*
At Maximum Residence Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 7.9-03 9.2-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 7.9-03 Inhalation 1.2-04 0.0 2.2-04 2.2-04 1.9-02 2.3-04 2.4-04 1.1-04 Total dose 8.0-03 9.2-03 8.1-03 8.1-03 2.7-02 8.1-03 8.1-03 8.0-03 Analysis performed at maximum residence location is 4,106 m (13,471 ft) east.
NOTE:
7.9-03 = 7.9x10-3 (1)
All annual doses shown in Table 11A.1-8 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-11 Rev. 25, October 2022 TABLE 11A.1-9 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS*
At Maximum Cow Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 1.3-02 1.6-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 Inhalation 1.4-04 0.0 1.1-04 2.0-04 9.7-03 2.5-04 2.7-04 1.9-04 Fresh vegetation 7.0-04 0.0 1.3-03 1.1-03 1.3-01 1.1-03 1.3-04 7.3-04 Stored vegetation 2.4-03 0.0 4.3-03 3.0-03 4.1-03 1.4-03 6.9-04 1.4-03 Cow milk 2.4-03 0.0 3.0-03 3.7-03 2.2-01 2.7-03 4.5-04 1.1-03 Deer 1,603 m east 1.4-04 0.0 1.7-04 1.9-04 3.2-04 9.2-05 3.1-05 3.3-04 Total dose 1.9-02 1.6-02 2.2-02 2.1-02 3.8-01 1.9-02 1.5-02 1.7-02 Analysis performed at maximum cow location is 2,350 m (7,710 ft) east-southeast.
NOTE:
1.3-02 = 1.3x10-2 (1)
All annual doses shown in Table 11A.1-9 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-12 Rev. 25, October 2022 TABLE 11A.1-10 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS*
At Maximum Cow Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 1.3-02 1.6-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 Inhalation 1.5-04 0.0 1.5-04 2.4-04 1.3-02 3.1-04 3.6-04 2.1-04 Fresh vegetation 4.8-04 0.0 1.2-03 9.5-04 1.1-01 1.2-02 1.2-04 5.1-04 Stored vegetation 2.8-03 0.0 7.7-03 5.2-03 6.5-03 6.5-02 1.3-03 1.9-03 Cow milk 3.0-03 0.0 5.3-03 6.4-03 3.5-01 4.7-03 8.5-04 1.5-03 Deer 1,603 m east 7.8-05 0.0 1.4-04 1.5-04 2.4-04 7.6-03 2.7-05 1.9-04 Total dose 2.0-02 1.6-02 2.7-02 2.6-02 4.9-01 1.0-01 1.6-02 1.7-02 Analysis performed at maximum cow location is 2,350 m (7,710 ft) east-southeast.
NOTE:
1.3-02 = 1.3x10-2 (1)
All annual doses shown in Table 11A.1-10 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-13 Rev. 25, October 2022 TABLE 11A.1-11 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS*
At Maximum Cow Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 1.3-02 1.6-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 Inhalation 1.5-04 0.0 2.1-04 2.3-04 1.6-02 2.9-04 3.1-04 1.7-04 Fresh vegetation 5.8-04 0.0 2.2-03 1.3-03 1.6-01 1.3-03 1.8-04 3.9-04 Stored vegetation 3.8-03 0.0 1.9-02 9.3-03 1.3-02 4.4-03 2.6-03 2.6-03 Cow milk 4.3-03 0.0 1.3-02 1.1-02 7.1-01 7.9-03 1.6-03 1.8-03 Deer 1,603 m east 8.6-05 0.0 2.5-04 2.0-04 3.6-04 9.5-05 4.0-05 1.2-04 Total dose 2.2-02 1.6-02 4.8-02 3.5-02 9.1-01 2.7-02 1.8-02 1.8-02 Analysis performed at maximum cow location is 2,350 m (7,710 ft) east-southeast.
NOTE:
1.3-02 = 1.3x10-2 (1)
All annual doses shown in Table 11A.1-11 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-14 Rev. 25, October 2022 TABLE 11A.1-12 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS*
At Maximum Cow Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 1.3-02 1.6-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 1.3-02 Inhalation 9.6-05 0.0 1.5-04 1.8-04 1.5-02 1.8-04 2.3-04 9.5-05 Cow milk 6.6-03 0.0 2.3-02 2.2-02 1.7+00 1.3-02 3.2-03 4.7-03 Total dose 2.0-02 1.6-02 3.6-02 3.5-02 1.7+00 2.6-02 1.6-02 1.8-02 Analysis performed at maximum cow location is 2,350 m (7,710 ft) east-southeast.
NOTE:
1.3-02 = 1.3x10-2 (1)
All annual doses shown in Table 11A.1-12 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-15 Rev. 25, October 2022 TABLE 11A.1-13 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE ADULT GROUP FROM GASEOUS EFFLUENTS*
At Maximum Beef Animal Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 2.4-02 2.8-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 Inhalation 2.3-04 0.0 1.3-04 3.1-04 1.2-02 3.6-04 4.9-04 3.2-04 Fresh vegetation 1.6-03 0.0 2.7-03 2.4-03 2.6-01 2.3-03 2.4-04 1.7-03 Stored vegetation 5.7-03 0.0 8.4-03 7.4-03 7.9-03 3.1-03 1.3-03 3.0-03 Beef 9.7-04 0.0 1.7-03 1.4-03 1.5-02 8.1-04 3.2-04 4.4-03 Deer 1,603 m east 1.4-04 0.0 1.7-04 1.9-04 3.2-04 9.2-05 3.1-05 3.3-04 Total dose 3.3-02 2.8-02 3.7-02 3.6-02 3.2-01 3.1-02 2.6-02 3.4-02 Analysis performed at maximum beef animal location is 1,693 m (5,555 ft) east.
NOTE:
2.4-02 = 2.4x10-2 (1)
All annual doses shown in Table 11A.1-13 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-16 Rev. 25, October 2022 TABLE 11A.1-14 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE TEEN GROUP FROM GASEOUS EFFLUENTS*
At Maximum Beef Animal Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 2.4-02 2.8-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 Inhalation 2.4-04 0.0 1.8-04 3.6-04 1.6-02 4.3-04 6.6-04 3.4-04 Fresh vegetation 1.1-03 0.0 2.5-03 2.2-03 2.1-01 3.2-02 2.3-04 1.2-03 Stored vegetation 6.3-03 0.0 1.5-02 1.3-02 1.3-02 1.7-01 2.4-03 4.0-03 Beef 6.1-04 0.0 1.4-03 1.1-03 1.1-02 9.2-02 2.6-04 2.5-03 Deer 1,603 m east 7.8-05 0.0 1.4-04 1.5-04 2.4-04 7.6-03 2.7-05 1.9-04 Total dose 3.2-02 2.8-02 4.3-02 4.1-02 2.7-01 3.3-01 2.8-02 3.2-02 Analysis performed at maximum beef animal location is 1,693 m (5,555 ft) east.
NOTE:
2.4-02 = 2.4x10-2 (1)
All annual doses shown in Table 11A.1-14 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-17 Rev. 25, October 2022 TABLE 11A.1-15 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE CHILD GROUP FROM GASEOUS EFFLUENTS*
At Maximum Beef Animal Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 2.4-02 2.8-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 Inhalation 2.3-04 0.0 2.4-04 3.4-04 2.0-02 4.0-04 5.6-04 2.6-04 Fresh vegetation 1.2-03 0.0 4.5-03 2.8-03 3.3-01 2.6-03 3.1-04 7.8-04 Stored vegetation 7.6-03 0.0 3.6-02 2.2-02 2.6-02 9.2-03 4.4-03 4.4-03 Beef 7.9-04 0.0 2.5-03 1.4-03 1.7-02 8.8-04 4.4-04 1.6-03 Deer 1,603 m east 8.6-05 0.0 2.5-04 2.0-04 3.6-04 9.5-05 4.0-05 1.2-04 Total dose 3.4-02 2.8-02 6.7-02 5.1-02 4.2-01 3.7-02 3.0-02 3.1-02 Analysis performed at maximum beef animal location is 1,693 m (5,555 ft) east.
NOTE:
2.4-02 = 2.4x10-2 (1)
All annual doses shown in Table 11A.1-15 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of he values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-18 Rev. 25, October 2022 TABLE 11A.1-16 ANNUAL DOSES TO MAXIMUM INDIVIDUAL IN THE INFANT GROUP FROM GASEOUS EFFLUENTS*
At Maximum Beef Animal Location (Annual Dose in mRem/yr)(1)
Pathway Total Body Skin Bone Liver Thyroid Kidney Lung GI Tract Contaminated ground 2.4-02 2.8-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 2.4-02 Inhalation 1.4-04 0.0 1.8-04 2.5-04 1.8-02 2.5-04 4.1-04 1.4-04 Total dose 2.4-02 2.8-02 2.4-02 2.4-02 4.2-02 2.4-02 2.4-02 2.4-02 Analysis performed at maximum beef animal location is 1,693 m (5,555 ft) east.
NOTE:
2.4-02 = 2.4x10-2 (1)
All annual doses shown in Table 11A.1-16 reflect a power level of 3,323 MWt. Due to an operation at EPU/MELLLA+ at 3,988 MWt, each of the values shown must be multiplied by a factor of 1.79; the uprated values represent the maximum possible increase.
NMP Unit 2 USAR Chapter 11 11A.1-19 Rev. 25, October 2022 TABLE 11A.1-17 COMPARISON OF MAXIMUM CALCULATED DOSES FROM UNIT 2 WITH APPENDIX I DESIGN OBJECTIVES Appendix I Unit 2 Design Calculated Criterion Objective(1)
Dose Gaseous effluents Gamma air dose(2),
mRad/yr 10 7.9E-02 Beta air dose(2),
mRad/yr 20 5.3E-02 Noble gas - total body(3), mRem/yr 5
3.9E-02 Noble gas - skin(3),
mRem/yr 15 8.3E-02 Iodines and particulates(4)
Any organ (thyroid),
mRem/yr 15 2.8E+00 Liquid effluents Total body, mRem/yr 3
1.8E-00 Any organ(5), mRem/yr 10 5.1E+00 NOTE:
6.7E-02 = 6.7x10-2 (1)
Per reactor.
(2)
Calculated at exclusion area boundary 1,603 m (5,259 ft) east.
(3)
Calculated at 1,693 m (5,554 ft) east.
(4)
Infant thyroid dose from cow milk 2,350 m (7,710 ft) east-southeast.
(5)
Child bone dose is calculated to be the highest organ dose.
Nine Mile Point Unit 2 USAR Chapter 11 11A.1-20 Rev. 25, October 2022 TABLE 11A.1-18 CALCULATED ANNUAL DOSES FOR POPULATION WITHIN 80-KM (50-MI) RADIUS Total Body Thyroid (man-Rem)
(man-Rem)
Liquid Effluents Ingestion of potable water 1.2E-01 6.7E-02 Ingestion of fish 4.1E+00 9.0E-03 Shoreline recreation 6.8E-02 4.3E-02 Swimming 1.4E-04 8.8E-05 Boating 6.8E-05 4.3E-05 Total 4.3E+00 1.2E-01 Gaseous Effluents(3)
Submersion 3.3-01 3.3-01 Inhalation 1.4-02 9.9-01 Standing on contaminated ground 7.3-02 7.3-02 Ingestion of fruits, grains, and vegetation 1.7-01 1.2+00 Ingestion of cow milk 4.8-02 6.8-01 Ingestion of meat 3.8-03 7.1-03 Total(4) 6.4-01 3.3+00 NOTES:
- 1. Based upon a projected 80-km (50-mi) population of 1.2+06 for the year 2010.
- 2. 4.0E-02 = 4.0x10-2.
- 3. The values shown for "Gaseous Effluents" are based on a power level of 3,323 MWt. Due to an increase in the power level to 3,988 MWt, the "Submersion" values are to be multiplied by a factor of 1.37 and the rest by a factor of 1.66.
- 4. The "Total" values shown are based on a 3,323 MWt power level.
NMP Unit 2 USAR Chapter 11 11A.1-21 Rev. 25, October 2022 TABLE 11A.1-19 CALCULATED POPULATION DOSE COMMITMENT (Contiguous U.S. Population Dose)
Annual Dose Per Site Total Body Thyroid (man-Rem)
(man-Rem)
Liquid effluents 4.3E+00 1.2E-01 Noble gas effluents(1) 1.14+00 1.37+00 Radioiodines and particulates*(2) 1.9+01 2.3+01 Total(3) 2.1+01 2.5+01 NOTE:
1.5E+00 = 1.5x100 (1)
The annual dose values shown in Table 11A.1-19 are based on a power level of 3,323 MWt. Due to a power increase to 3,988 MWt, the values must be multiplied by 1.37. The uprated values represent the maximum possible increase.
(2)
The annual dose values shown in Table 11A.1-19 are based on a power level of 3,323 MWt. Calculation disposition UE-005-01A and calculation change notice ECP-12-000448-CN-008 UE-005-01.00 provide the required information to adjust for an EPU level of 3,988 MWt.
(3)
The "Total" values represent doses at 3,323 MWt. To adjust for 3,988 MWt (except for "Liquid effluents" which has already been incorporated), adjust the values as noted in notes (1) and (2) above.
Carbon-14 and tritium have been added to this category, applicable for the 3,323 MWt power level values.
NMP Unit 2 USAR Chapter 11 11A.2-1 Rev. 25, October 2022 11A.2 COST-BENEFIT ANALYSIS This section presents the results of cost-benefit analyses performed in accordance with Section II.D of 10CFR50 Appendix I.
Augments to the liquid and gaseous effluent systems and respective potential reductions to the annual population exposure are taken from RG 1.110. The beneficial savings of each augment were calculated by multiplying the calculated dose reduction by $1,000/man-Rem or $1,000/man-Rem/thyroid.
Augments to the Liquid Effluent Treatment System Table 11A.2-1 presents the calculated base case (updated for EPU) annual total body dose (man-Rem) and thyroid dose (man-Rem/thyroid) associated with the operation of the plant LWS system for the population expected to live within an 80-km radius of the plant for the year 2010. Assuming that each augment is capable of reducing the population doses to zero (an extremely conservative assumption), the maximum benefit to be derived from any augment would be $1,400 for reducing man-Rem exposures to zero and $61 for reducing man-Rem/thyroid exposures to zero.
In an analysis of the annualized procurement, installation, operation, and maintenance costs, the least expensive liquid radwaste augment was found to be $20,000/yr for a plant located in the northeastern United States. Since the benefit from this augment would be less than the corresponding total annualized cost, the cost-benefit ratio is greater than 1. The operation of additional equipment for the purpose of reducing the annual population dose would not be cost effective. Therefore, the most cost-beneficial system has been included in the current plant design.
Augments to the Gaseous Effluent Treatment System Table 11A.2-2 presents the calculated base case (updated for EPU) annual total body dose (man-Rem) and thyroid dose man-Rem/thyroid associated with the operation of the gaseous radwaste system for the 80-km radius population.
Assuming that each augment is capable of reducing the population doses to zero, the maximum benefit to be derived from any augment would be $640 for reducing man-Rem exposures to zero and
$3,300 for reducing man-Rem/thyroid exposures to zero.
In an analysis of the annualized procurement, installation, operation, and maintenance costs, the least expensive gaseous radwaste augment was found to be $10,580/yr for a plant located in the northeastern United States. Since the benefit from this augment would be less than the corresponding total annualized cost, the cost-benefit ratio is greater than 1. The operation
NMP Unit 2 USAR Chapter 11 11A.2-2 Rev. 25, October 2022 of additional equipment for the purpose of reducing the annual population dose would not be cost effective. Therefore, the most cost beneficial system has been included in the current plant design.
NMP Unit 2 USAR Chapter 11 11A.2-3 Rev. 25, October 2022 TABLE 11A.2-1 BASE CASE ANNUAL POPULATION DOSES DUE TO LIQUID EFFLUENTS Total Body Dose Thyroid Dose Pathway (man-Rem)
(man-thyroid-Rem)
Ingestion of fish 4.1E+00 9.0E-03 Ingestion of potable water 1.2E-01 6.7E-02 Shoreline recreation 6.8E-02 4.3E-02 Swimming 1.4E-04 8.8E-05 Boating 6.8E-05 4.3E-05 Total 4.3E+00 1.2E-01 NOTE:
1.4E+00 = 1.4x100
NMP Unit 2 USAR Chapter 11 11A.2-4 Rev. 25, October 2022 TABLE 11A.2-2 BASE CASE ANNUAL POPULATION DOSES DUE TO GASEOUS EFFLUENTS Total Body Dose Thyroid Dose Pathway (man-Rem)(1)
(man-thyroid-Rem)(1)
Submersion 3.3-01 3.3-01 Inhalation 1.4-02 9.90-01 Standing on contaminated ground 7.30-02 7.30-02 Ingestion of fruits, grains, and vegetation 1.70-01 1.20+00 Ingestion of cow milk 4.8-02 6.80-01 Ingestion of meat 3.8-03 7.10-03 Total(2) 6.40-01 3.30+00 NOTE:
3.3-01 = 3.3x10-1 (1)
The values for dose shown in Table 11A.2-2 are based on a power level of 3,323 MWt. Due to an increase in the power level to 3,988 MWt, the "Submersion" values are to be multiplied by a factor of 1.37 and the remaining values by a factor of 1.66.
(2)
The "Total" values shown are based on a 3,323 MWt power level.