recommended updates to gale codes · (gale) codes from boiling-water reactors (bwr) and...

PNNL-18957

Recommended Updates to GALE Reactor Effluent Codes KJ Geelhood November 2009

PNNL-18957

Recommended Updates to GALE Reactor Effluent Codes

KJ Geelhood

November 2009

Prepared for

The U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Division of Risk Analysis

Washington, DC 20555-0001

under Contract DE-AC05-76RL01830

Pacific Northwest National Laboratory

Richland, Washington 99352

iii

Summary

This report describes work that Pacific Northwest National Laboratory (PNNL) is doing to assist U.S.

Nuclear Regulatory Commission (NRC) staff in their reviews of early site permit and combined license

applications for new nuclear power plants. This work focuses on assessing the potential consequences of

licensing actions for new reactor core designs. Potential updates are being considered for a suite of

models that are used to estimate the routine releases of radioactive materials to the environment from

pressurized water reactors and boiling water reactors. These models are the Gaseous and Liquid Effluents

(GALE) codes from Boiling-Water Reactors (BWR) and Pressurized-Water Reactors (PWR).

To direct the definition of what updates are needed to make the GALE codes applicable to facilities

using new reactor core designs, a sensitivity study was performed on the four GALE codes (Geelhood

et al. 2008)a. This study considered, based on selected representative nuclides, the sensitivity of hard-

wired parameters on the prediction of nuclide concentrations. In addition to the parameters identified in

the sensitivity study, the U.S. Nuclear Regulatory Commission also identified other parameters to be

examined and updated as necessary.

Recommendations are made for parameter and guidance updates to each of the GALE codes. The

parameters identified in the sensitivity study as having a significant impact on the output were

re-examined relative to modern reactor operations. Because facilities using new reactor core designs use

the same radwaste treatment processes, these operations at currently operating plants are applicable to the

radwaste operations at reactors using new core designs. A series of recommendations for making, or not

making, parameter changes in GALE codes are made. Each recommendation includes the reference to

the supporting data. When a change is recommended, the specifics of the change are defined. These

recommendations are the basis for new versions of the GALE codes that will be applicable to operations

at facilities using new reactor core designs.

a Geelhood KJ, MR Mitchell, and JG Droppo. 2008. Sensitivity Analysis of Hardwired Parameters in GALE Codes.

PNNL-18150, Pacific Northwest National Laboratory, Richland, Washington.

v

Acronyms, Initialisms, and Abbreviations

ANS American Nuclear Society

ANSI American National Standards Institute

BWR boiling water reactor

BWR-LE BWR GALE computer code for predicting liquid radioactive effluents

BWR-GE BWR GALE computer code for predicting gaseous radioactive effluents

Ci curie (unit of measurement of radioactivity)

DF decontamination factor

EPA U.S. Environmental Protection Agency

GALE computer code for estimating Gaseous and Liquid Effluents from Nuclear Electrical

Power Generation Reactors

GE gas effluent

HEPA high-efficiency particulate air (filter)

LE Liquid effluent

MW(t) megawatt thermal

NRC U.S. Nuclear Regulatory Commission

PNNL Pacific Northwest National Laboratory

PWR pressurized-water reactor

PWR-LE PWR GALE computer code for predicting liquid radioactive effluents

PWR-GE PWR GALE computer code for predicting gaseous radioactive effluents

vii

Contents

Summary ............................................................................................................................................... iii

Acronyms, Initialisms, and Abbreviations ............................................................................................ v

1.0 Introduction .................................................................................................................................. 1.1

2.0 GALE Parameter and Guidance Updates ..................................................................................... 2.1

2.1 Plant Capacity Factor ........................................................................................................... 2.2

2.1.1 Pressurized-Water Reactors ...................................................................................... 2.4

2.1.2 Boiling Water Reactors ............................................................................................. 2.4

2.2 Tritium Release .................................................................................................................... 2.4



2.3 Argon-41 Release ................................................................................................................. 2.6



2.4 Carbon-14 Release ............................................................................................................... 2.8



2.5 Adjustments to Liquid Release for Unexpected Release ..................................................... 2.10



2.6 Decontamination Factors for Demineralizers, Reverse Osmosis Units, Evaporators,

Centrifuges, and Filters ........................................................................................................ 2.14



2.7 Dynamic Adsorption Coefficients for Charcoal ................................................................... 2.17



2.8 Removal Efficiencies for HEPA Filters ............................................................................... 2.18


2.8.2 Boiling-Water Reactors ............................................................................................. 2.18

2.9 Iodine Removal Efficiencies for Charcoal Adsorbers .......................................................... 2.19



3.0 Conclusions .................................................................................................................................. 3.1

4.0 References .................................................................................................................................... 4.1

viii

Figures

Figure 2.1. Industry-Wide Capacity Factor for All U.S. Nuclear Power Plants .................................. 2.3

Figure 2.2. Median Plant Capacity Factor for U.S. PWRs and U.S. BWRs ........................................ 2.3

Figure 2.3. PWR Tritium Release Data, Mean, and Confidence Interval ............................................ 2.5

Figure 2.4. BWR Tritium Release Data, Mean, and Confidence Interval ........................................... 2.6

Figure 2.5. PWR Argon-41 Release Data, Mean, and Confidence Interval ........................................ 2.7

Figure 2.6. BWR Argon-41 Release Data from Plants with Measureable Argon-41 Release,

Mean, and Confidence Interval ..................................................................................................... 2.8

Tables

Table 2.1. Parameters Examined in this Report ................................................................................... 2.1

Table 2.2. Reported Carbon-14 Release from PWRs .......................................................................... 2.9

Table 2.3. Reported Carbon-14 Release from BWRs .......................................................................... 2.10

Table 2.4. Unplanned Liquid Release of Gamma Emitters from PWR Plants .................................... 2.12

Table 2.5. Unplanned Liquid Release of Gamma Emitters from BWR Plants .................................... 2.13

Table 2.6. GALE Recommended Decontamination Factors for Various Nuclides in Various

PWR Demineralizers .................................................................................................................... 2.14

Table 2.7. GALE Recommended Decontamination Factors for Various Nuclides in Various PWR

Evaporators, Liquid Radwaste Filters, and Reverse Osmosis Units ............................................. 2.14

Table 2.8. Measured DFs for Mixed Bed Demineralizers ................................................................... 2.15


BWR Demineralizers .................................................................................................................... 2.16


BWR Evaporators, Liquid Radwaste Filters, and Reverse Osmosis Units ................................... 2.16

Table 2.11. GALE Recommended Dynamic Adsorption Coefficients for Krypton and Xenon in

Charcoal Delay System ................................................................................................................. 2.17

Table 2.12. Iodine Removal Efficiencies Recommended in Regulatory Guide 1.140 Rev. 1 ............. 2.19

Table 2.13. Iodine Removal Efficiencies Recommended in Regulatory Guide 1.140 Rev. 2 ............. 2.20

1.1

1.0 Introduction

This report describes work that Pacific Northwest National Laboratory (PNNL) is doing to assist U.S.

Nuclear Regulatory Commission (NRC) staff in their reviews of early site permit and license applications

for new nuclear power plants. This work focuses on assessing the potential consequences of licensing

actions for new reactor core designs. Potential updates are being considered for a suite of models that are

used to estimate the routine releases of radioactive materials to the environment from pressurized water

reactors and boiling water reactors. These models are the Gaseous and Liquid Effluents (GALE) codes

from Boiling-Water Reactors (BWR) and Pressurized-Water Reactors (PWR).

New nuclear power generating facilities have been proposed for operation in the United States using

new reactor core designs. The objective of this effort is to define what updates are needed to the GALE

codes to make them applicable for use in the licensing process for these new facilities.

Making the GALE codes applicable to the proposed new nuclear power generating facilities requires

consideration of 1) updates to the GALE source codes and 2) updates to the guidance for using these

codes. The GALE updates for the new core designs are to be based a combination of a review of how the

new designs may change routine reactor radwaste operations and on more recent reactor operating

experience than was used for developing the original GALE codes.

Several aspects of the GALE source codes need to be considered. In terms of the formulations used

in the source codes, Geelhood et al. (2008) concluded the radwaste treatment technologies treated in the

GALE source codes will be appropriate for the proposed new reactors. However, it was also

recommended that new versions of the GALE with modern coding structure be created. The “spaghetti

structure” of the original GALE versions makes code verification and maintenance especially difficult.

The parameters used to characterize the liquid and gas effluents are a combination of input and hard-

wired parameters. Changes in input parameters can be documented as changes in the user guidance for

doing GALE simulations. The parameters hard wired in each of the GALE source codes may also need to

be updated. A sensitivity study was performed on the hard-wired parameters in four GALE codes

(Geelhood et al. 2008) to help direct the definition of which of these parameters may need to be updated.

The results of a review of parameter values for applications to facilities using new reactor core designs

are provided as follows.

2.1

2.0 GALE Parameter and Guidance Updates

A review of recent reactor operational experience and the design details for proposed new nuclear

power generating facilities is used as a basis for recommendations for updates to the GALE source codes

and their user guidance.

The sensitivity study performed for the four GALE codes (Geelhood et al. 2008) provides a basis for

evaluating which parameters may need to be updated. The sensitivity study considered, based on selected

representative nuclides, the sensitivity of hard-wired parameters on the prediction of nuclide

concentrations. In addition to the parameters identified in the sensitivity study, NRC also identified other

parameters to be examined and updated as necessary. Because facilities using new reactor core designs

use the same radwaste treatment processes, these operations at currently operating plants are applicable to

the radwaste operations at reactors using new core designs.

The first column in Table 2.1 lists the parameters in the GALE codes that were examined. The

second column has the section number where each parameter is discussed.

Table 2.1. Parameters Examined in this Report

Parameter

Section of this

Report

Plant capacity factor 2.1

Tritium release 2.2

Argon-41 release 2.3

Carbon-14 release 2.4

Adjustment to liquid release for unexpected release 2.5

Decontamination factors for demineralizers, reverse osmosis units

evaporators, centrifuges, and filters

2.6

Dynamic adsorption coefficients for charcoal 2.7

Removal efficiencies for HEPA filters 2.8

Iodine removal efficiencies for charcoal adsorbers 2.9

Several parameters examined in this report were updated on the basis of taking an average value of a

random sampling from reactor operations over the past 8 years. The approach used was to collect a

number of measurements and calculate the mean value using Equation 1 and the standard deviation using

Equation 2.

n

x

x

n

i

i

1 Equation 1

2.2

1

)( 2

1

n

xx

s

n

i

i

Equation 2

Where: x = mean of sample population

s = standard deviation of the population

xi = individual measurements

n = number of measurements

In order to determine if the number of measurements used to calculate the mean value was sufficient,

a 95% confidence interval was constructed around the mean value using a Student t-distribution. This

was done using Equation 3.

n

stxCI

)*)((

Equation 3

Where: CI = 95% confidence interval about the mean

x = mean of sample population

t* = value on t-distribution table for 95% confidence level and the number of samples taken

s = standard deviation of the population

n = number of measurements

If the confidence interval is small relative to the mean, then the calculated mean value can be used. If

the confidence interval remained large and more data were available, the process would be repeated with

more data. Once an appropriately small confidence interval was found or all the data was included, the

upper bound of the confidence level was used as the value recommended for GALE as a conservative

estimate of the mean.

2.1 Plant Capacity Factor

The plant capacity factor defines what fraction of full capacity of the facility occurs during long-term

operation of that facility. The current versions of the PWR GALE and BWR GALE codes use a

hardwired value of 80% for plant capacity factor.

Nuclear-industry-wide capacity factors have been compiled by the Energy Information

Administration of the U.S. Department of Energy and are tabulated in the Annual Energy Review (AER

2009). Figure 2.1 shows these capacity factors since 1973. The mean capacity factor between 2000 and

2008 was 89.8% with a standard deviation of 1.4%. These data may warrant increasing the plant capacity

factor used in the updated GALE codes.

2.3

Figure 2.1. Industry-Wide Capacity Factor for All U.S. Nuclear Power Plants

Figure 2.2 shows median plant capacity factor for three-year intervals compiled separately for BWRs

and PWRs by Blake (2009). Blake discusses using the median statistic rather than the mean to reduce the

impact of the outage of Browns Ferry-1. For the period from 2000 to 2008, the yearly mean value of the

median plant capacity is 89.6% for U.S. PWRs and is 90.6% for all U.S. BWRs. Therefore, it is

recommended that the plant capacity factor used in the updated PWR GALE and BWR GALE codes be

increased to 90%.

Figure 2.2. Median Plant Capacity Factor for U.S. PWRs and U.S. BWRs

2.4

2.1.1 Pressurized-Water Reactors

For the period from 2000 to 2008, Blake (2009) calculates the yearly mean value of the median plant

capacity to be 89.6% for U.S. PWRs. Using data reported by Blake (2009) for 69 individual U.S. PWRs,

the mean plant capacity factor is 89.8%, and the standard deviation is 5.3% during the three-year period

from 2006 through 2008. Therefore, it is recommended that the plant capacity factor used in the updated

PWR GALE codes be increased to 90%.

The original version of PWR GALE uses a plant capacity factor to determine the yearly tritium

release per MW(t). However, the tritium release value in the new version of PWR GALE will be updated

based on actual plant measurements as discussed in Section 2.2 and will not require a calculation

involving plant capacity factor.

2.1.2 Boiling Water Reactors

For the period from 2000 to 2008, Blake (2009) calculates the yearly mean value of the median plant

capacity to be 90.6% for U.S. BWRs. Using data reported by Blake (2009) for 25 individual U.S. BWRs,

the mean plant capacity factor is 89.2%, and the standard deviation is 8.5% during the three-year period

from 2006 through 2008. As covered by Blake (2009), the extended (22 year) outage of Browns Ferry-1

did not end until almost half of this period had elapsed. Compensating for the influence of the extended

Browns Ferry-1 outage results in a mean plant capacity factor for this period of 90.6% with a standard

deviation of 3.8% for the remaining 34 U.S. BWRs. Therefore, it is recommended that the plant capacity

factor used in the updated BWR GALE codes be increased to 90%.

In the BWR GALE codes, iodine release rates and the carbon-14 release rates are a function of plant

capacity factor and are hard wired into the code. Iodine release rates from various buildings are given in

terms of Ci/yr/ Ci/g separately for normal operation and extended shutdown. Since the recommended

plant capacity factor is 90%, the revised iodine release rates for normal operation should be increased by

multiplying by 1.125 (90%/80%), and the revised iodine release rates for extended shutdown should be

decreased by multiplying by 0.5 (10%/20%). The recommendation that the carbon-14 release rate be

increased to use a 90% plant capacity factor in the formula that calculates carbon-14 production will be

discussed further in Section 2.4.

2.2 Tritium Release

Tritium release defines the expected rate of tritium release for long-term operation of a nuclear power

generating facility.


The current version of PWR GALE uses a value of 0.4 Ci/yr/MW(t) for tritium release. This tritium

is partitioned between liquid and gaseous effluents by assigning 1 Ci/ml of primary reactor coolant to

liquid release up to 90% of the total release. The remaining tritium is released as a gaseous effluent. The

value 1 Ci/ml is specified in ANS 18.1 (ANS 1984).

Surveys were performed on the tritium release from randomly selected reactors between the years

2000 and 2007. A total of 76 reactor-years were surveyed, and a mean tritium release value of

2.5

0.243±0.025 Ci/yr/MW(t) was calculated with a 95% confidence level. The data, mean, and confidence

interval are shown in Figure 2.3. Based on the upper bound average from this survey, it is recommended

that the tritium release from PWR be changed to 0.27 Ci/yr/MW(t). The partitioning between gaseous

and liquid effluents will remain the same as the most recent ANS 18.1 standard (ANS 1999), which still

specifies 1 Ci/ml of tritium in the primary coolant.

Figure 2.3. PWR Tritium Release Data, Mean, and Confidence Interval


The current version of BWR GALE uses a value of 0.03 Ci/yr/MW(t) for tritium release. This tritium

is partitioned between liquid and gaseous effluents by assigning 50% of the tritium to liquid release and

50% of the tritium to gaseous release.

Surveys were performed on the tritium release from randomly selected reactors between the years

2000 and 2008. A total of 71 reactor-years were surveyed, and a mean tritium release value of

0.0299±0.0068 Ci/yr/MW(t) was calculated with a 95% confidence level. The data, mean, and

confidence interval are shown in Figure 2.4. Based on this survey, it is recommended that the tritium

release from BWR remain the same at 0.03 Ci/yr/MW(t). The partitioning between gaseous and liquid

effluents was examined for those plants with both gaseous and liquid release, and an average value of

65% gas and 35% liquid was calculated. There was significant scatter in these ratios. As such, it is

recommended that the ratio of 50% liquid and 50% gaseous effluents be retained.

2.6

Figure 2.4. BWR Tritium Release Data, Mean, and Confidence Interval (previous value and sample

mean are identical)

2.3 Argon-41 Release

Argon-41 release defines the expected rate of argon-41 gas release for long-term operation of a

nuclear power generating facility.


The current version of PWR GALE uses a value of 34 Ci/yr for argon-41 release.

Surveys were performed on the argon-41 release from randomly selected reactors between the years

2000 and 2007. A total of 76 reactor-years were surveyed, and a mean argon-41 release value of

3.471±1.9904 Ci/yr was calculated with a 95% confidence level. The data, mean, and confidence interval

are shown in Figure 2.5. Based on this survey, it is recommended that the argon-41 release from PWR be

changed to the upper bound of the confidence level of the average value, or 6 Ci/yr.

2.7

Figure 2.5. PWR Argon-41 Release Data, Mean, and Confidence Interval


The current version of BWR GALE uses a combination of several sources for the argon-41 release.

The BWR GALE code assumes 15 Ci/yr is released from the drywell and includes this in the argon-41

released from the containment building. The BWR GALE code also assumes that 40 Ci/s (1262 Ci/yr)

is input to the main condenser offgas treatment downstream of the air ejectors. A holdup time for

argon-41 is calculated assuming dynamic adsorption coefficients for argon-41 in a charcoal delay system

of 6.4 cm³/g at ambient conditions and 16 cm³/g for chilled conditions. The argon-41 remaining after this

holdup is released from the air ejector.

The 15 Ci/yr of argon-41 from the drywell is based on a measured neutron flux in the drywell

interacting with natural argon-40 in the air and assuming 24 purges per year. It is not likely that the

neutron flux in the drywell has changed, and 24 purges per year approaches the value of an open drywell,

thus is probably conservative for this analysis

It was determined that the dynamic adsorption coefficients measured when the charcoal delay beds

were installed are still applicable for modern reactors. Section 2.7 provides more discussion on this topic.

Therefore, it is recommended that the assumed dynamic adsorption coefficients for argon-41 be retained

in GALE.

The 1262 Ci/yr input to the main condenser is based on air inleakage containing argon-40 at or

downstream of the main condenser. This argon-40 is activated in the reactor coolant to argon-41 and

2.8

released through the main condenser offgas treatment system. The value selected was an upper bound of

only several measurements made and represents a conservative upper bound. No additional

measurements have been made, but this value probably still represents a conservative upper bound.

Plants were surveyed for argon-41 release. Of the 71 reactor-years surveyed, 33 of those years had

measurable argon-41 release. The mean of these data was 11.1±7.0 Ci/yr with a 95% confidence level.

For the sample BWR case, GALE predicted an argon-41 release of 38 Ci/yr, which bounds this mean

value. The current data, mean, confidence interval, and previous GALE prediction for the sample case

are shown in Figure 2.6.

Given a further lack of specific data on argon-41 or relevant system upgrades, it is recommended that

the current methodology for calculating argon-41 release from BWRs be retained in GALE.

Figure 2.6. BWR Argon-41 Release Data from Plants with Measureable Argon-41 Release, Mean, and

Confidence Interval (GALE Calculation for the Sample Case also is shown)

2.4 Carbon-14 Release

Carbon-14 release defines the expected rate of carbon-14 release for long-term operation of a nuclear

power generating facility.

2.9


The current version of PWR GALE uses a value of 7.3 Ci/yr for carbon-14 release based on

measurements from seven reactors between 1975 and 1978.

NRC does not require licensees to report carbon-14 in their annual radioactive release reports.

Because of this, it is not possible to compile a large data set of carbon-14 release values as with the

tritium (hydrogen-3) and argon-41. A search of literature published since the time of the last survey was

performed to determine if more recent data would indicate a different carbon-14 release.

An EPRI report (Vance et al. 1995) presented data from two U.S. and six German PWRs in 1984 and

1985 that had an average carbon-14 release of 5.9 Ci/yr. Another source (Yim et al. 2006) referenced an

EPA study in 1981 that found the release to be about 5 Ci/yr. These data are shown in Table 2.2.

Table 2.2. Reported Carbon-14 Release from PWRs

Reactor Carbon-14 release, Ci/yr

E(a)

9.264

F(a)

5.684

H(b)

3.01

I(b)

8.19

J(b)

1.242

K(b)

4.3692

L(b)

5.985

M(b)

9.49

Average 5.9

EPA study 5

(a) U.S. reactor.

(b) German reactor.

It is recommended that the value in the PWR GALE be decreased from 7.3 Ci/yr to 5.9 Ci/yr based on

average values from the most recent surveys. This value also compares well with value from the

aforementioned EPA study.


The current version of BWR GALE uses a value of 9.5 Ci/yr for carbon-14 release based on the

assumption that all the carbon-14 produced in the reactor coolant will be released and the formula for

carbon-14 production provided in Equation 4.

Q = Noσoφmtps Equation 4

Where: Q = carbon-14 release rate, Ci/yr

No = 1.3x1022

atoms O-17/kg in water

2.10

σo = 2.4x10-25

cm-2

thermal cross section for O-17

φ = 3x1013

n/cm²s average thermal flux

m = 3.9x104 kg water in reactor core

t = 3.15x107 s/yr

p = 0.8, plant capacity factor

s = 1.03x10-22

Ci/atom carbon-14

Based on these values, a release rate for carbon-14 of 9.5 Ci/yr can be calculated.

As discussed in Section 2.1.2, the plant capacity factor should be increased to 0.9. If this is done, a

release rate for carbon-14 of 10.7 Ci/yr is calculated.

In order to confirm that this calculated value is reasonable, a literature search of recent publications

was performed. An EPRI report (Vance et al. 1995) presented data from one U.S. and three German

BWRs in 1984 and 1985 that had an average carbon-14 release of 10.7 Ci/yr. Another source (Yim et al.

2006) referred to an EPA study from 1981 that found the release to be about 9 Ci/yr. This source also

stated that in a BWR, all the produced carbon-14 is released from the condenser steam jet air ejector.

These data are shown in Table 2.3. These references confirm the calculated value of 10.7 Ci/yr, as well

as the assumption that all the carbon-14 produced will be released.

Table 2.3. Reported Carbon-14 Release from BWRs

Reactor Carbon-14 release, Ci/yr

3 10.18

7 10.05

8 12.397

9 10.34

Average 10.7

EPA study 9

Calculated value 10.7

Based on the updated calculation, it is recommended that the value in the BWR GALE be increased

from 9.5 Ci/yr to 10.7 Ci/yr. Recent surveys confirm that this value is reasonable.

2.5 Adjustments to Liquid Release for Unexpected Release

Adjustment factors are used in GALE code to account for unexpected releases during normal

operations.

2.11


The current version of PWR GALE uses a value of 0.16 Ci/yr to adjust the liquid release of all

nuclides except for tritium. This value of 0.16 Ci/yr is partitioned to the nuclides based on their relative

distribution calculated from other sources.

Surveys were performed on the unexpected release from all the U.S. operating PWRs for up to five

years between the years 2000 and 2008. Palo Verde units 1, 2, and 3 were excluded from this survey

because they have no liquid release. In addition, Point Beach units 1 and 2 were excluded as no mention

was made of unexpected release in the annual release reports for these plants. A total of 320 reactor-years

worth of data were examined; the average unexpected liquid release value for all nuclides other than

tritium was 1.6x10-4

Ci/yr. The data are shown in Table 2.4

Based on this survey, it is recommended that release from unexpected liquid releases be changed

from 0.16 Ci/yr to 1.6x10-4

Ci/yr.

2.12

Table 2.4. Unplanned Liquid Release of Gamma Emitters from PWR Plants

Reactor Date Release, Ci Reactor Date Release, Ci Reactor Date Release, Ci Reactor Date Release, Ci

Watts Bar 2007 0 Vogtle-2 2008 0 Calvert Cliffs 1 2007 0.00000084 Waterford 3 2008 0

2006 0.00000492 2007 0 2005 0 2007 0

2005 0 2006 0 2004 0 2006 0

2004 0 2005 0 2003 0 2005 0

2003 0 2004 0 2002 0 2004 0

2002 0 Comanche Peak 1 2008 0 Calvert Cliffs 2 2007 0.00000084 Braidwood-1 2008 0

2001 0 2007 0 2005 0 2007 0

2000 0 2006 0 2004 0 2006 0.0002685

Indian Point 2 2004 0 2005 0 2003 0 2005 0

2005 0.0005025 2004 0 2002 0 2004 0

2006 0.0002831 Comanche Peak 2 2008 0 Gina 2008 0 Braidwood-2 2008 0

2007 0.000035 2007 0 2007 0 2007 0

Indian Point 3 2004 0 2006 0 2005 0 2006 0.0002685

2005 0.0005025 2005 0 2004 0 2005 0

2006 0.0002831 2004 0 2003 0 2004 0

2007 0.000035 Sequoyah-1 2005 0 Millstone-2 2007 0 Byron-1 2008 0

Kewaunee 2007 0 2004 0 2006 0 2007 0

2006 0 2003 0 2005 0 2006 0

2005 0 2002 0 2004 0 2005 0

2004 0 2001 0 2003 0.000296 2004 0

2003 0 Sequoyah-2 2005 0 Millstone-3 2007 0 Byron-2 2008 0

2002 0 2004 0 2006 0 2007 0

2001 0 2003 0 2005 0 2006 0

2000 0 2002 0 2004 0 2005 0

Crystal River 3 2007 0 2001 0 2003 0 2004 0

2006 0 South Texas 1 2008 0 North Anna-1 2006 0 Beaver Valley-1 2008 0

2005 0 2007 0 2005 0 2007 0

2004 0 2006 0 2004 0 2006 0

2003 0 2005 0 2003 0.0178 2005 0.000163

2002 1.60E-06 2004 0 2002 0 2004 0.0003645

2001 4.40E-06 South Texas 2 2008 0 North Anna-2 2006 0 Beaver Valley-2 2008 0

2000 0 2007 0 2005 0 2007 0

San Onofre 2 2007 0 2006 0 2004 0 2006 0

2006 0 2005 0 2003 0.0178 2005 0.000163

2005 0 2004 0 2002 0 2004 0.0003645

2004 0 Virgil Summer 2008 0 Surry-1 2008 0 Davis Besse 2007 0

2003 0 2007 0 2007 0 2006 0

San Onofre 3 2007 0 2006 0 2006 0 2004 0

2006 0 Shearon Harris 2008 0 2005 0 2003 0

2005 0 2007 0 2004 0 2002 0

2004 0 2006 0 Surry-2 2008 0 St. Lucie-1 2008 0

2003 0 2005 0 2007 0 2007 0

Wolf Creek 2007 0 2004 1.55E-05 2006 0 2006 0

2006 0 Robinson-2 2008 0 2005 0 2005 0

2005 0 2007 0 2004 0 2004 8.77E-03

2004 0 2006 0 McGuire-1 2008 0 St. Lucie-2 2008 0

2003 0 2005 0 2007 0 2007 0

Catawba 1 2007 0 2004 0 2006 0 2006 0

2006 0 Diablo Canyon 1 2008 0 2005 0 2005 0

2005 0 2007 0 2004 0 2004 0

2003 0 2006 0 McGuire-2 2008 0 Turkey Point-3 2008 0

2002 0 2005 0 2007 0 2007 0

Catawba 2 2007 0 2004 0 2006 0 2006 0

2006 0 Diablo Canyon 2 2008 0 2005 0 2005 0

2005 0 2007 0 2004 0 2004 0

2003 0 2006 0 Oconee-1 2008 0 Turkey Point-4 2008 0

2003 0 2005 0 2007 0 2007 0

Salem-1 2008 0 2004 0 2006 0 2006 0

2007 0 Fort Calhoun 2005 0 2005 0 2005 0

2006 0 2004 0 2004 0 2004 0

2005 0 Prairie Island 1 2008 0 Oconee-2 2008 0 Seabrook 2007 0

2004 0 2007 0 2007 0 2005 0

Salem-2 2008 0 2006 0 2006 0 2001 0

2007 0 2004 0 2005 0 2000 0

2006 0 2003 0 2004 0 Donald Cook -1 2008 0.000000056

2005 0 Prairie Island 2 2008 0 Oconee-3 2008 0 2007 0

2004 0 2007 0 2007 0 2006 0

Farley-1 2008 0 2006 0 2006 0 2005 0

2007 0 2004 0 2005 0 2004 0

2006 0 2003 0 2004 0 Donald Cook -2 2008 0.000000056

2005 0 Callaway 2007 0 ANO-1 2006 0 2007 0

2004 0 2006 0 2005 0 2006 0

Farley-2 2008 0 2005 9.28E-04 2004 0 2005 0

2007 0 2004 1.82E-03 2002 0 2004 0

2006 0 TMI-1 2008 0 2001 0 Palisades 2007 0

2005 8.81E-08 2007 0 ANO-2 2006 0 2006 0

2004 0 2006 0 2005 0 2005 0

Vogtle-1 2008 0 2004 0 2004 0 2004 0

2007 0 2001 0 2002 0 2003 0

2006 0 2001 0

2005 0

2004 0

2.13


The current version of BWR GALE uses a value of 0.1 Ci/yr to adjust the liquid release of all

nuclides except for tritium. This value of 0.1 Ci/yr is partitioned to the nuclides based on their relative

distribution calculated from other sources.

Surveys were performed on the unexpected release from all the U.S. operating BWRs for up to five

years between the years 2000 and 2008. Columbia, Duane Arnold, Clinton, and Fermi-2 were excluded

from this survey because they had no liquid release. In addition, Peach Bottom units 2 and 3 were

excluded as the release from unexpected release was not separately quantified from the normal release in

the annual release reports for these plants. A total of 147 reactor-years were examined; the average

unexpected liquid release value for all nuclides other than tritium was 0.014 Ci/yr. The data are shown in

Table 2.5.

Table 2.5. Unplanned Liquid Release of Gamma Emitters from BWR Plants

Reactor Date Release, Ci Reactor Date Release, Ci Reactor Date Release, Ci Reactor Date Release, Ci

Grand Gulf 2007 0 Browns Ferry 1 2007 0.0617 Nine Mile Point 2 2008 0 Limerick-1 2008 0.0000418

2006 0.955 2005 0.0375 2007 0 2007 0

2005 0 2004 0 2006 0 2006 0.00092325

2004 0 2003 0.837 2005 0 2005 0

2003 0 2002 0 2004 0 2004 0

2002 0 Browns Ferry 2 2007 0 FitzPatrick 2008 0 Limerick-2 2008 0.0000418

2001 0 2005 0 2006 0 2007 0

2000 0 2004 0 2005 0 2006 0.00092325

Dresden 2 2008 0 2003 0 2004 0 2005 0

2007 0 2002 0 2003 0 2004 0

2006 0 Browns Ferry 3 2007 0 Pilgrim 2008 0 Perry-1 2008 0

2005 0.13 2005 0 2007 0 2007 0

2004 0.00000566 2004 0 2006 0 2006 0

Dresden 3 2008 0 2003 0 2004 0 2005 0

2007 0 2002 0 Vermont Yankee 2008 0 2004 0

2006 0 Susquehanna 1 2008 0 2007 0 Cooper 2006 0

2005 0 2007 0 2006 0 2005 0

2004 0 2006 0 2005 0 2004 0

Quad Cities 1 2007 0 2004 0 2004 0 2003 0

2006 0.0001047 2003 0 River Bend 2008 0 2001 0

2005 0.00205 Susquehanna 2 2008 0 2007 0 Montecello 2008 0

2004 0.00444 2007 0 2006 0 2007 0.0000237

2003 0.0562 2006 0 2004 0 2006 0

Quad Cities 2 2007 0 2004 0 2002 0 2005 0

2006 0 2003 0 LaSalle-1 2008 0 2004 0.00000742

2005 0 Oyster Creek 2008 0 2007 0 Hope Creek 2008 0

2004 0 2007 0 2006 0 2007 0

2003 0 2006 0.000141 2005 0 2006 0

Brunswick 1 2007 0 2005 0 2004 0 2005 0

2006 0 2004 0 LaSalle-2 2008 0 2004 0

2005 0 Nine Mile Point 1 2008 0 2007 0 Hatch-1 2008 0

2004 0 2007 0 2006 0 2007 0

2003 0 2006 0 2005 0 2006 0

Brunswick 2 2007 0 2005 0 2004 0 2005 0

2006 0 2004 0 2004 0.0000065

2005 0 Hatch-2 2008 0

2004 0 2007 0

2003 0 2006 0

2005 0

2004 0

Based on this survey, it is recommended that release from unexpected liquid releases be changed

from 0.1 Ci/yr to 0.014 Ci/yr.

2.14

2.6 Decontamination Factors for Demineralizers, Reverse Osmosis Units, Evaporators, Centrifuges, and Filters

The GALE code uses decontamination factors to define the expected performance of radwaste

systems during long-term operation of a nuclear power generating facility.


For all the liquid radwaste systems in PWR GALE, except the primary coolant demineralizer and the

condensate demineralizer, the user is required to input the expected decontamination factor (DF). For the

primary coolant demineralizer, GALE assumes DFs of 2 for cesium and rubidium and 10 for all other

nuclides. For the condensate demineralizer, GALE assumes DFs of 2 for cesium and rubidium, 100 for

anions, and 50 for other nuclides.

The GALE input document provides recommended values to use for all other liquid radwaste

systems. These values are shown in Tables 2.6 and 2.7.


PWR Demineralizers

Anion Cesium, Rubidium Other Nuclides

Mixed bed purification systems (LiBO3) 100(a)

2(a)

50(a)

10(b)

Boron recycle system 10 2 10

Evaporator condensate (H+OH

-) 5 1 2

(b) 10

Radwaste (H+OH

-) 10²(10) 2(10) 10²(10)

Steam Generator Blowdown 10²(10) 10(10) 10²(10)

Cation bed (H+) (any system) 1(1) 10(10) 10(10)

Anion bed (OH-) (any system) 10²(10) 1(1) 1(1)

Powdex (any system) 10(c)

(10) 2(c)

(10) 10(c)

(10)

(a) Hard wired for Condensate Demineralizer.

(b) Differences between GALE and ANS 55.6-1993 are shown in bold italic.

(c) Hard wired for Primary Coolant Demineralizer.

Table 2.7. GALE Recommended Decontamination Factors for Various Nuclides in Various PWR

Evaporators, Liquid Radwaste Filters, and Reverse Osmosis Units

All Nuclides Except Iodine Iodine

Miscellaneous radwaste

evaporators

10³ 10²

Boric acid evaporators 10³ 10²

Separate evaporator for detergent

wastes

10² 10²

Liquid radwaste filters 1 1

Reverse osmosis units 10(30 for laundry) 10(30 for laundry)

2.15

The latest ANS standard on liquid radioactive waste processing (ANS-55.6-1993 R 2007) contains a

table of DFs for various evaporators, demineralizers, and reverse osmosis units. The values in this

standard were compared with the values in Tables 2.6 and 2.7. In all cases except for two, the

recommendations in GALE were identical to those in ANS-55.6-1993 R (2007). The values that are

different are shown in bold italic font (b) in Table 2.6 after the recommended value.

ANS-55.6-1993 R (2007) does not provide recommended DFs for liquid radwaste filters. One

reference, Mandler et al. (1981), provided 216 measurements at three PWR plants that gave an average

DF of 1.56 for liquid radwaste filters. The primary purpose of a filter is not to reduce the activity of the

wastewater, but to reduce solids input to other systems (ANS-55.6-1993 R 2007). Based on this, it is

appropriate and slightly conservative to retain the current GALE recommended DF of 1.0 for liquid

radwaste filters.

In order to confirm the DFs that are hard wired in GALE as noted in Table 2.6, a brief literature

search was performed for these types of demineralizers. For the primary coolant demineralizer, the DFs

from the Powdex row are used. A recent evaluation of the Powdex process (Salem and LaTerra 1999)

stated that Powdex processes could remove more than 90% of all metals from the condensate (DF=10).

This confirms the DF assumed for all nuclides except cesium and rubidium. This evaluation (Salem and

LaTerra 1999) also stated that the Powdex process could remove 50% (DF=2) to 90% (DF=10) of various

iron oxides. Given that cesium and rubidium are highly reactive with water and will probably exist in the

water as oxides, a DF of 2 for these elements seems reasonable.

For the condensate demineralizer, the DFs from the mixed bed row are used. One reference (Mandler

et al. 1981) provides measured DFs from mixed beds for various nuclides taken from four PWRs. These

measurements are shown in Table 2.8. The average values found in these measurements are 1191 for

anions, 6.9 for cesium, and 96.4 for all others. This comparison demonstrates the DFs that GALE and

ANS-55.6-1993 R (2007) use are conservative; however, it is not clear that this would justify increasing

the DFs as individual values at or below the GALE recommended values were measured.

Table 2.8. Measured DFs for Mixed Bed Demineralizers (Mandler et al. 1981)

Measurement Anion Cesium Other

1 2600 1.3 69

2 3600 1.3 95

3 100 1.2 13

4 18 1.2 20

5 380 1.0 40

6 450 1.0 40

7 36 15

8 35 6.7

9 0.95 680

10 0.95 170

11 1.2 6.7

12 1.2 1.3

Average: 1191 6.9 96.4

GALE Value 100 2 50

ANS-55.6-1993 Value 100 2 10

2.16


For all the liquid radwaste systems in BWR GALE, except the condensate demineralizer, the user is

required to input the expected DF. For the condensate demineralizer, GALE assumes DFs of 2 for cesium

and rubidium and 10 for other nuclides.

The GALE input document provides recommended values to use for all other liquid radwaste

systems. These values are shown in Tables 2.9 and 2.10.


BWR Demineralizers

Anion Cesium, Rubidium Other Nuclides

Mixed bed (H+OH

-)

Reactor coolant

10

2

10

Condensate 10(a)

2(a)

10(a)

Clean waste 10²(10) 10(10) 10²(10)

Dirty waste (floor drains) 10²(10) 2(10) 10²(10)

Cation bed (H+)

Dirty waste

1(1)

10(10)

10²(10) 10(10)(b)

Powdex (any system) 10(10) 2(10) 10(10)

(a) Hard wired for Condensate Demineralizer.

(b) Differences between GALE and ANS 55.6-1993 are shown in bold italic.

Table 2.10. GALE Recommended Decontamination Factors for Various Nuclides in Various BWR

Evaporators, Liquid Radwaste Filters, and Reverse Osmosis Units

All Nuclides Except Anions Anions

Miscellaneous radwaste

evaporator

104 10

3(a) 10

3 10

2(a)

Separate evaporator for detergent

wastes

102

102

Liquid radwaste filters 1 1

Reverse osmosis units 10(30 for laundry) 10(30 for laundry)

(a) Differences between GALE and ANS 55.6-1993 are shown in bold italic.

The latest ANS standard on liquid radioactive waste processing (ANS-55.6-1993 R 2007) contains a

table of DFs for various evaporators, demineralizers, and reverse osmosis units. The values in this

standard were compared with the values in Tables 2.9 and 2.10. In all cases except for two, the

recommendations in GALE were identical to those in ANS-55.6-1993 R (2007). The values that are

different are shown in bold italic font in Tables 2.9 and 2.10, following their recommended value.

ANS-55.6-1993 R (2007) does not provide recommended DFs for liquid radwaste filters. Based on

the recent data with PWR liquid radwaste filters (previously discussed) and the fact that the primary

purpose of a filter is not to reduce the activity of the wastewater but to reduce solids input to other

2.17

systems (ANS-55.6-1993 R 2007), it is appropriate and slightly conservative to retain the current GALE

recommended DF of 1.0 for liquid radwaste filters.

It is recommended that he following changes be made to the GALE recommended DFs:

PWR mixed bed purification systems (LiBO3) DF for other nuclides should change from 50 to 10.

PWR evaporator condensate (H+OH

-) DF for cesium, rubidium should change from 1 to 2.

BWR cation bed (H+) dirty waste DF for other nuclides should change from 100(10) to 10(10).

BWR miscellaneous radwaste evaporator DFs should change from 103 for anions and 10

4 for other

nuclides to 102 for anions and 10

3 for other nuclides.

In addition, since the DFs for PWR mixed bed purification systems are hard wired in GALE for the

condensate demineralizer, the PWR GALE code should be changed to use a DF for other nuclides of 10

rather than 50.

2.7 Dynamic Adsorption Coefficients for Charcoal

The GALE codes use dynamic adsorption coefficients to define the expected performance charcoal

filter systems during long-term operation of a nuclear power generating facility.


In the PWR GALE code, the user must specify the holdup time for fission gases stripped from the

primary coolant. There are several different holdup systems, described below, with the input required for

PWR GALE:

1. Pressurized storage tanks—calculate holdup times and fill time using the equations provided in the

PWR GALE manual.

2. Charcoal delay systems—calculate holdup times using the equations provided in the PWR GALE

manual and the table of dynamic absorption coefficients provided in the PWR GALE manual.

3. Cover gas recycle system—assumes a 90-day holdup time

The Table 2.11 shows the dynamic adsorption coefficients that are recommended for a charcoal delay

system.

Table 2.11. GALE Recommended Dynamic Adsorption Coefficients for Krypton (Kr) and Xenon (Xe)

in Charcoal Delay System

Operating 77°F

Dew Point 45°F

Operating 77°F

Dew Point 0°F

Operating 77°F

Dew Point -40°F

Operating 0°F

Dew Point -20°F

Krypton 18.5 25 70 105

Xenon 330.0 440 1160 2410

In an e-mail conversation with Charles Myers, president of Columbus, Ohio-based NCS Corp., a

longtime industry leader in the testing and servicing of safety-related HVAC systems, he shared that little

2.18

work has been done in the field of activated charcoal that would improve on these dynamic adsorption

coefficients since the original systems were installed. Several more recent measurements (Sun et al.

1994) of dynamic adsorption coefficients show reasonable agreement with the data used in the GALE

derivation. Based on this, it is recommended that the table of recommended dynamic adsorption

coefficients shown in Table 2.11 be retained in the user input guidelines.


In the BWR GALE code, the user must specify the xenon and krypton dynamic adsorption

coefficients for the charcoal delay system in the condenser air ejector offgas treatment system.

The Table 2.11 shows the dynamic adsorption coefficients that are recommended for a charcoal delay

system.

The charcoal delay systems in a BWR are the same as those in a PWR. Therefore, similar to the

PWR recommendation, it is recommended that the table of recommended dynamic adsorption coefficients

shown in Table 2.11 be retained in the user input guidelines.

2.8 Removal Efficiencies for HEPA Filters

The GALE code uses removal efficiencies to define the expected performance of HEPA filter systems

during long-term operation of a nuclear power generating facility.


In the PWR GALE code, the user is required to enter the removal efficiency for any HEPA filters in

use in the waste gas system and fuel handling and auxiliary buildings. In addition, removal efficiencies

for HEPA filters in use in the containment internal cleanup system, high volume purge, and low volume

purges must be specified.

The PWR GALE manual suggests a removal efficiency of 99% on Regulatory Guide 1.140 rev. 1

(NRC 1979). This regulatory guide was updated in 2001 (NRC 2001), but the updated version still

recommends a removal efficiency of 99%.

Based on this, the recommendation for user input if HEPA filters are used should stay at 99%. If a

manufacturer is able to provide information that its HEPA filters will remove more than this, the utility

can input a greater removal rate and reference the manufacturer’s data.

2.8.2 Boiling-Water Reactors

In the BWR GALE code, the user is required to enter the removal efficiency for any HEPA filters in

use in the containment, turbine, auxiliary, and radwaste buildings.

The BWR GALE manual suggests a removal efficiency of 99% on Regulatory Guide 1.140 rev. 1

(NRC 1979). This regulatory guide was updated in 2001 (NRC 2001), but the updated version still

recommends a removal efficiency of 99%.

2.19

Based on this, the recommendation for user input if HEPA filters are used should stay at 99%. If a

manufacturer is able to provide information that its HEPA filters will remove more than this, the utility

can input a greater removal rate and reference the manufacturer’s data.

2.9 Iodine Removal Efficiencies for Charcoal Adsorbers

The GALE code uses removal efficiencies to define the expected performance of iodine removal

systems during long-term operation of a nuclear power generating facility.


In the PWR GALE code, the user is required to enter the iodine removal efficiency for any charcoal

adsorbers in use in the fuel handling and auxiliary buildings. In addition, removal efficiencies for

charcoal adsorbers in use in the containment internal cleanup system, high volume purge, and low volume

purges must be specified.

The PWR GALE manual provides a table of removal efficiencies based on Regulatory Guide 1.140

rev. 1 (NRC 1979). The 1979 table is repeated here as Table 2.12. This regulatory guide and its

corresponding table have been updated, and the user should use the updated table from Regulatory

Guide 1.140 rev. 2 (NRC 2001). The 2001 version of the table is repeated as Table 2.13.

It is recommended that the PWR GALE user input be updated to reference the removal efficiencies in

Table 2.13.

Table 2.12. Iodine Removal Efficiencies Recommended in Regulatory Guide 1.140 Rev. 1 (1979)

Activated Carbon Bed Depth Removal Efficiencies for Radioiodine

2 inches. Air filtration systems designed to operate

inside reactor containment

90%


outside reactor containment and relative humidity is

controlled at 70%

70%



controlled at 70%

90%



controlled at 70%

99%

2.20


In the BWR GALE code, the user is required to enter the removal efficiency for any charcoal

adsorbers that are in use in the containment, turbine, auxiliary, and radwaste buildings.

The BWR GALE manual provides a table of removal efficiencies based on Regulatory Guide 1.140

rev. 1 (NRC 1979) . The 1979 table is repeated here as Table 2.12. This regulatory guide and its

corresponding table have been updated, and the user should use the updated table from Regulatory

Guide 1.140 rev. 2 (NRC 2001). This table is repeated as Table 2.13.

It is recommended that the BWR GALE user input instructions be updated to reference the removal

efficiencies in Table 2.13.

Table 2.13. Iodine Removal Efficiencies Recommended in Regulatory Guide 1.140 Rev. 2 (2001)

Activated Carbon Total

Bed Depth

Maximum Assigned Activated

Carbon Decontamination

Efficiencies

Methyl Iodine Penetration Acceptance

Criterion for Representative Sample

2 inches Elemental iodine

Organic iodine

95%

95%

Penetration ≤ 5% when tested in

accordance with ASTM D-3803-1989

4 inches or greater Elemental iodine

Organic iodine

99%

99%

Penetration ≤ 1% when tested in

accordance with ASTM D-3803-1989

3.1

3.0 Conclusions

Code and guidance updates needed in each of the GALE codes are identified. The supporting data for

each of the recommended parameter changes listed below are given in the main text. These changes are

to be the basis for new versions of the GALE codes that will be applicable to operations at facilities using

new reactor core designs.

The following changes to the GALE codes are proposed in this document:

Plant capacity factor should be increased from 0.8 to 0.9 (80% to 90%)

BWR iodine release rates from various buildings during normal operations should be increased by

multiplying by 1.125

BWR iodine release rates from various buildings during extended shutdown should be decreased by

multiplying by 0.5

PWR tritium release rate should be decreased from 0.4 Ci/yr/MW(t) to 0.27 Ci/yr/MW(t)

PWR argon-41 release rate should be decreased from 34 Ci/yr to 6 Ci/yr

PWR carbon-14 release rate should be decreased from 7.3 Ci/yr to 5.9 Ci/yr

BWR carbon-14 release rate should be increased from 9.5 Ci/yr to 10.7 Ci/yr

PWR unexpected release rate should be decreased from 0.16 Ci/yr to 1.6x10-4

Ci/yr

BWR unexpected release rate should be decreased from 0.1 Ci/yr to 0.014 Ci/yr

PWR condensate demineralizer DF for other nuclides should be changed from 50 to 10

In addition, the following changes to the GALE user input instructions are proposed in this document:

For both PWR and BWR GALE, change the recommended values for iodine removal efficiencies

from those in Table 2.12 (1979) to those in Table 2.13 (2001)

PWR mixed bed purification systems (LiBO3) DF for other nuclides should change from 50 to 10

PWR evaporator condensate (H+OH

-) DF for cesium, rubidium should change from 1 to 2

BWR cation bed (H+) dirty waste DF for other nuclides should change from 100(10) to 10(10)

BWR miscellaneous radwaste evaporator DFs should change from 103 for anions and 10

4 for other

nuclides to 102 for anions and 10

3 for other nuclides

4.1

4.0 References

ANS 18.1-1984. December 1984. American Nuclear Society Standard Radioactive Source Term for

Normal Operation of Light Water Reactors. American Nuclear Society, Illinois.

ANS 18.1-1999. September 1999. American Nuclear Society Standard Radioactive Source Term for

Normal Operation of Light Water Reactors. American Nuclear Society, Illinois.

ANS-55.6-1993. 2007R. American Nuclear Society Standard for Liquid Radioactive Waste Processing

System for Light Water Reactor Plants. American Nuclear Society, Illinois.

Blake EM. May 2009. “U.S. capacity factors: Can older reactors keep up the pace?” Nuclear News, 29-

33.

EIA - Energy Information Administration. 2009. Annual Energy Review 2008. DOE/EIA-0384(2008).

Energy Information Administration, U.S. Department of Energy, Washington, D.C.

Geelhood KJ, MR Mitchell, and JG Droppo. 2008. Sensitivity Analysis of Hardwired Parameters in

GALE Codes. PNNL-18150, Pacific Northwest National Laboratory, Richland, Washington.

Mandler JW, AC Stalker, ST Croney, CV McIsacc, GA Soli, JK Hartwell, LS Loret, BG Motes, TE Cox,

DW Akers, NK Bihl, SW Duce, JW Tkachyk, CA Pelletier, and PG Voileque. 1981. In-Plant Source

Term Measurements at Four PWR’s. NUREG/CR-1992, Nuclear Regulatory Commission,

Washington, D.C.

NRC - U.S. Nuclear Regulatory Commission. 2001. Design, Inspection, and Testing Criteria for Air

Filtration and Adsorption Units of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear

Power Plants. Regulatory Guide 1.140 Rev. 2, U.S. Nuclear Regulatory Commission, Washington D.C.

NRC - U.S. Nuclear Regulatory Commission. 1979. Design, Testing, and Maintenance Criteria for

Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear

Power Plants. Regulatory Guide 1.140 Rev. 1, U.S. Nuclear Regulatory Commission, Washington, D.C.

Salem E and T LaTerra. 1999. Half a Century of Condensate Polishing. Graver Water Systems, Inc.,

Cranford, New Jersey.

Sun CL, JC Chen, YW Yu, HC Ha, CS Lu, and TY Lee. 1994. “Dynamic Adsorption Properties of Kr

and Xe Isotopes in Charcoal.” Journal of Radioanalytical and Nuclear Chemistry, 181(2):291-299.

Vance JN, JE Cline, and DE Robertson. 1995. Characterization of Carbon-14 Generated by the Nuclear

Power Industry. EPRI-TR-105715, Electric Power Research Institute, Palo Alto, California.

Yim MS and F Caron. 2006. “Life cycle and management of carbon-14 from nuclear power generation.”

Progress in Nuclear Energy, 48:2-36.

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