ML19325C166

Technical Evaluation Rept for Evaluation of Odcm,Rev 7, Jm Farley Nuclear Plant,Units 1 & 2.

ML19325C166
Person / Time
Site:	Farley
Issue date:	08/31/1989
From:	Bohn T, Serrano W, Thomas Young EG&G IDAHO, INC.
To:	NRC
Shared Package
ML19325C167	List: ML19325C166
References
CON-FIN-D-6034 EGG-PHY-8674, NUDOCS 8909130029
Download: ML19325C166 (30)

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- August,'1989 4 '

TECHNICAL EVALUATION REPORT .

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/deho TECHNICAL EVALUATION REPORT FOR THE EVALUATION I

NeWons/ OF ODCM REVISION 7 JOSEPH M. FARLEY NUCLEAR ,

Eng/naer/np PLANT, UNITS 1 AND 2 Laboratory Managed T. E. Young l by the U.S. T. S. Bohn 1

Department M. Serrano ofEnergy l'

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• 1 TECHNICAL EVALUATION REPORT for the EVALUATION OF ODCM REVISION 7 -

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JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 NRC Docket No. 50-348 NRC LICENSE NO. NPF-2  :

NRC Docket No. 50-364 NRC LICENSE NO. NPF-8

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T. E. Young T. S. Bohn W. Serrano s

Published August 1989 Idaho National Engineering Laboratory ,

EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 i Prepared for the U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Under DOE Contract No. DE-AC07-76ID01570 ,

FIN No. D6034

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ABSTRACT i

The Offsite Dose Calculation Manual (ODCM) for the Joseph M. Farley Nuclear Plant Units 1 and 2 contains current methodology and parameters used in the calculation of offsite doses due to radioactive liquid and gaseous effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program. A copy of the most recent complete ODCM was transmitted to the NRC as Revision 7, with a letter dated April 13, 1989. The NRC transmitted Revision 7 to the Idaho National  :

Engineering Laboratory (INEL) for review. The revised ODCM was reviewed i by EG&G Idaho at the INEL and the results of the review are presented in this report. It was determined that ODCM Revision 7 uses methods that are, in general, in agreement with the guidelines of NUREG-0133. However, it is recommended that some changes and additions bc included in the Licensee's next revision to the ODCH.

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This report is submitted as partial fulfillment of the " Review of.

p Radiological Issues" project being conducted by the Idaho National Engineering Laboratory for the U. S. Nuclear Regulatory Commission, f Office of Nuclear Reactor Regulation. The U. S. Nuclear Regulatory Commission funded the work under FIN 06034 (Project 5) and NRC B&R Number 20 19 05 03.

!' This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warrant, expressed or implied, or assumes any legal liability or responsibility for .i any third party's use, or the results of such use, of any information, I apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately-owned rights.

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-Abstract .:. . . . . . . . . . . . . . . . . . . . . . . . . . . . . i i

e Foreword . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11  ;

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l. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2. Review Criteria. . . . . . . . . . . . . . . . . . . . . . . . . 2 1

3. Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

4. Conclusions. . . . . . . . . . . . . . . . . . . . . . . . . . . 19'

5. References'. . . . . . . . . . . . . . . . . . . . . . . . . . . 23 FIGURES

1. Radioactive liquid waste treatment system and effluent monitors. 4

2. Liquid effluent waste lines and liquid release point . . . . . . 5 l

3. Gaseous radwaste treatment system and effluent pathways. . . . . 8

4. Gaseous effluent release points. . . . . . . . . . . . . . . . . 9 .

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1. INTRODUCTION .

I 1.1 Purcose of Review .

This document reports the review and evaluation of the most recent version of the Offsite Dose Calculation Manual (ODCM) submitted by the Alabama Power Company (APC), the Licensee for the Jospeh M. Farley Nuclear Plant, Units 1 and 2 (FNP). The ODCM is a supplementary document for implementing the Radiological Effluent Technical Specifications (RETS) in compliance with 10 CFR 50, Appendix I requirementsIll, i

1.2 Plant-Soecific Backaround  ;

The APC transmitted a complete ODCM, designated Revision 7, effective April 3, 1989 to the Nuclear Regulatory Commission (NRC) with a letter dated April 13,1989[2),

The NRC transmitted Revision 7 of the ODCM to the Idaho National Engineering Laboratory (INEL) for review. The complete ODCM, updated '

through Revision 7, was reviewed by EG&G Idaho at the INEL. Results and conclusions of the review of Revision 7 are presented in this report.

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2. REVIEW CRITERIA

. Review criteria for the ODCM were provided by the NRC in two documents:

NUREG-0472, RETS for PWRsI33 -

NUREG-0133, Preparation of RETS for Nuclear Power PlantsI43 r The following NRC guidelines were also used.in the ODCM review: .

" General Contents of the Offsite Dose Calculation Manual," Revision 1[5], l Regulatory Guide 1.109(6), and Regulatory Guide 1.111 I73 ,

s As specified in NUREG-0472, the ODCM is to be developed by the Licensee to document the methodology and approaches used to calculate offsite doses and maintain the operability of the radioactive effluent systems. As a minimum, the ODCM should provide equations and methodology for the following:

. Alarm and trip setpoints on effluent instrumentation

. Liquid effluent concentrations in unrestricted areas

. Gaseous effluent dose rates at or beyond the site boundary

. Liquid and gaseous effluent dose contributions

. Liquid and gaseous effluent dore projections In addition, the ODCM should contain flow diagrams, consistent with the systems being used at the station, defining the treatment paths and the components of the radioactive liquid, gaseous, and solid waste management systems. A description and the location of samples in support of the environmental monitoring program are also needed in the ODCH.

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3.- EVALUATION The Farley ODCM, Revision 7, reviewed below, is applicable to Units 1 and 2 of FNP. Revision 7 was submitted by the Licensee as a complete 00CM. It includes major changes in format and used more current data than previous revisions (e.g., dose factors from Regulatory Guide 1.109, Revision 1 instead of those from the original Regulatory Guide issued in 1976). - The methodology used in 00CM Revision 7 is based primarily on >

NUREG 0133 and Regulatory Guide 1.109, Revision 1. An appendix was added, giving the equations and references for calculations descrf bed in the body

, of the ODCM.

Revision 7 is generally better organized and more complete than previous revisions. However, some changes and additions appear to be needed to increase clarity and completeness and to correct errors.

3.1 Liauid Effluent Pathways The Joseph M. Farley Nuclear Plant consists of two pressurized water reactors (PWR) operating with closed cycle circulating water systems using cooling towers. Radioactive liquid waste is released in batch mode from Waste Monitor Tanks through Waste Lines No. 1 and No. 2, in continuous mode from the Steam Generator Blowdowns, and by either batch or continuous mode from the Turbine Building Sumps. Figure 1 (copied from the ODCM) and Figure 2 (copied from the Technical Specifications) show the Radioactive Liquid Waste Treatment System, the Steam Generator Blowdowns, the liquid effluent Waste Lines, and the release point. Table 2-1 in the ODCM, Tables 3.3-13 and 4.11-1 in the FNP Technical Specifications, and Tables 2A-1 and 2A-2 in the Licensee's Semiannual Radioactive Effluent Release Reports indicate possible release of radioactive material from the Turbine Building Sumps in addition to the Steam Generator Blowdowns and the Waste Monitor Tanks. The release pathways of radioactive liquids from ,

the various sources are not clearly described in the 00CM. The following 3

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(Reproduced from Figure 21 of the ODCM, Revision 7 for the Joseph M. Farley Nuclear Plant.)

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I Figure 2. Liquid effluent waste lines and liquid release point. (Reproduced from Figure 5.1-4 of RETS for the Joseph M. Farley Nuclear Plant, Unit No. 2, complete as of July 27,1984.)

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details'are not clear from Figures 1 and 2 and the discussions in the ODCM: [

(a) whether the systems shown are common to both reactor units or exist f

sepai'ately for each uniti (b) whether the Waste Lines shown in Figure 2 are e the same as the Liquid Radwaste Effluent Lines identified in Table 3.313 of f the Technical Specifications; and (c) whether the dilution flow rates f discussed are defined for the two units separately or for the site as a l

whole. The ODCM should clarify these points with a figure and/or a detailed l

verbal description of the pathways of radioactive liquid waste and dilution  !

water from each source to the discharge point in the Chattahoochee River.  !

The discussions of setpoints, dose rates, and doses, respectively, in Sections 3.2, 3.5, and 3.7 below are based on assumptions that the systems l

shown in Figure 1 exist separately for each reactor unit and that the  ;

" dilution water flow rate," F ,d refers to the flow rate of dilution water j from the site at the Waste Discharge Point shown in Figure 2.  !

3.2 Liouid Radwaste Monitor Setooints Section 2.2 of the ODCM contains methodology to determine setpoints for i the radioactive liquid waste discharge line monitors, as required by I Technical Specification 3.3.3.10. Technical Specification 3.3.3.10 requires l that setpoints for the liquid effluent monitors be established to ensure -

that the concentration of radioactive material released fr:m the site shall be limited to concentrations specified in 10 'FR 20, Appendix B Table !!,  ;

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For each reactor unit at FNP, Technical Specification Table 3.313 l

requires two radioactive liquid effluent monitors, one on the Liquid Radwaste Effluent Line (RE-18) and another on the Steam Generator Blowdown .

Effluent Line (RE-238). This table also identifies flow rate measurement [

devices for Waste Monitor Tank No. 1, Waste Monitor Tank No. 2, the Steam +

l Generator Blowdown Effluent Line, and the Discharge Canal Dilution Line.

l Liquid releases from the Turbine Building Sumps are not required to be monitored, but Technical Specification 4.11.1.1.3 requires sampling and i analysis of effluents from this source.

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1 Methodology should be included in the ODCMToNeount for simultaneous  !

releases from each of the six sources of radioactive liquid effluents f (2 Turbine Builcing Sumps, 2 Steam Generator 810wdowns, and Waste Monitor l

Tanks for each unit). Unless simultaneous releases are considered it would I be possible to exceed the concentration limits of 10 CFR 20, since  !

Section 2.2 addresses the setpoint for only one Liquid Radwaste Effluent l Line monitor (RE 18), and does not consider releases from the other sources. Equation 2 1 defines this setpoint in terms of the " anticipated dilution factor" (D,). However, 0, is not specifically defined. l although the last sentence in Section 2.2 implies that it is equal to the i

" actual dilution factor" (R ) defined by Equetion 2.5 in Section 2.3. i I

Methodo1rgy should be added to determine the required setpoints for the  !

Steam Generator Blowdown Effluent Line monitors (RE 23B) and to account for  !

simultaneous releases of radioactive liquid effluents from all sources on j the site. With these exceptions, the methodology to determine setpoints  :

for the liquid radioactivity monitors is considered to be within the guidelines of NUREG 0133.  !

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3.3 Gaseous Effluent Pathwayi j There are six release points for radioactive gaseous effluents at FNP, a Plant Stack Vent for each unit, a Turbine Building Vent for each unit, >

and an ILRT (integrated leak rate test) Vent for each unit. These vents ,

l are shown in Figure 3 (copied from the ODCM) and Figure 4 (copied from the i

! FNP Technical Specifications), t The Containment Ventilation (a continuous release which is also referred to as Containment Purge), the Auxiliary Building Vent, and the Waste Gas Decay Tanks exhaust through the Plant Stack Vents. The

• Containment Ventilation and Auxiliary Building Vent releases are continuous; the contents of the The Waste Gas Decay Tanks are released in i batch mode.

l The ILRT, which releases gases from the containment, is described as "a l

batch release performed infrequently." These batch releases are made at points separate from the Plant Stack Vents.

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(Reproduced from Figure 3-1 of 00CM, Revision 7 for the Joseph M.

Farley Nuclear Plaat.)

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I Releases from the SJAE (Steam Jet Air Ejector) are made in continuous mode through the Turbine Building Vents. ,

The Plant Vent Stacks release gaseous effluents to the atmosphere at l 145 ft 9 in (44.4 m) above grade. Other releases are at lower elevations, !

3.4 Gaseous Radwaste Monitor Setooints i

f Section 3.3 of the ODCM contains methodology to determine the  !

alarm / trip setpoints for the radioactive gaseous effluent monitors, as (

required by Technical Specification 3.3.3.11. Technical Specification  !

3.3.3.11 requires that the setpoints be established to ensure that the dose rates due to radioactive noble gases released offsite shall not  ;

exceed either 500 mrem /yr to the total body or 3000 mrem /yr to the skin. i Technical Specification 3.3.3.11 requires noble gas activity monitors at the Steam Jet Air Ejector (R-15, shown between the Turbine Building and [

the Turbine Building Vent in Figure 3), the Vent Header System (R 14 or  ;

R-22, shown on the lire to the Plant Vent Stack in Figure 3), and the l Gaseous Radwaste Treatment System (R-14, also sh,wn on line to the Plar.t j Vent Stack in Figure 3). Monitor R 14 provides automatic termination of release. Figure 3 shows two monitors not required by the technical  :

specifications, one on the line from the Waste Gas Decay Tanks to the ,

Plant Vent Stack (R 13) and one on the line leading from containment to I the Au::iliary Building Plenum (R 24). (In ODCM Figure 3-1, monitor R-13  ;

should apparently be shown on the effluent line from all the Waste Gas Decay Tanks instead of from only 3 of 8 as is shown now.)

Equation 3.17, used to implement the methodology for determining setpoints for the Vent Header System monitor (R-14 or R-22) and Gaseous Radwaste Treatment System monitor (R-14) uses the highest annual average atmospheric dispersion factor at the site boundary. The equation includes a factor, Fy , to account for simultaneous releases from the two monitored release points at each rractor unit and a factor, F, to account for the number of units (2) at FNP. The term " average" should be removed 10

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, from the definition of Qgy in Equation 3.11, since the dose rate limits l l are instantan2ous. The ODCM methodology to determine setpoints does not l address releases from the Unit I and Unit 2 ILRT Vents. The technical specifications do not require radioactivity monitors for these vents, but l the methodology should account for contributions to dose rates due to l releases from them on the basis of the sampling and analysis required by Technical Specification Table 4.11

• t The Licensee's methodology determines setpoints based on the assumption that all noble gas released is Kr 89. The metholology results  !

in conservative setpoints except of the failure to consider releases from I the ILRT Vents, as noted in the above paragraph, f i

The methodology given in the ODCM for determining the setpoints of  ;

gaseous effluent monitors is, in general, within the guidelines of  !

NUREG 0133. However, contributions to the offsite dose rates due to i releases of noble gases via the ILRT Vents should be specifically j addressed 6nd the use of the modifier " average" with the release rate  !

should be removed or justified, i 3.5 Concentrations in Liouid Effluents i

Section 2.3 of the ODCM contains methodology to assure that the  ;

concentration of radioactive materials in liquid effluents is mtintained l L within the limits of Technical Specification 3.11.1.1 as required by 2 l

Specifications 3.11.1.1.1, 4.11.1.1.2, and 4.11.1.1.3. Technical j Specification 3.11.1.1 requires that the concentration of radioactive l 1

material released from the site in liquid effluents shall be limited to i

the concentrations specified in 10 CFR 20, Appendix B, Table 11, Column 2 l for radionuclides other than dissolved or entrained noble gases, and to 2 X 10'4 uC1/mL total activity for noble gases. Technical l Specifications 4.11.1.1.1, 4.11.1.1.2, and 4.11.1.1.3 require the results of pre-release t.nd post-release analyses to be used to assure that the concentration limits at the point of release are/were not exceeded.

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e i Equation 2.4 of Section 2.3 defines a conservative " required dilution factor," D r by including a " safety factor" of 2. However, releases  :

from only one liquid radwaste tank (Waste Monitor Tank) are specifically accounted for in the calculations. Possible simultaneous releases from l the Steam Generatnr Blowdowns, Turbine Building Sumps, and more than one  :

Waste Monitor Tank should be addressed, as discussed in Section 3.2 l

above. The source of dilution water flow, Fd , for each waste effluent

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stream should be identified.

With the exceptions noted in the previous paragraph, the methodology to determine that the concentrations of radioactive material in liquid ,

effluents are/were within the limits of 10 CFR 20 is considered to be I within the guidelines of NUREG 0133.  ;

3.6 Dose Rates Due to Gaseous Effluents  ;

Section 3.4 of the ODCM contains methodology to determine that the dose rates in unrestricted areas due to the release of radioactive materials in gaseous effluents are within the limits of Technical l Specification 3.11.2.1, as required by Technical Specification 4.11.2.1.1 l and 4.11.2.1.2. Technical Specification 3.11.2.1 requires that the dose  ;

rate due to radioactive. materials release 6 in gaseous effluents from the site does not exceed 500 mrem /yr to the total body or 3000 mrem /yr to the ,

skin for noble gases, and that the dose rate to any organ does not exceed 1500 mrem /yr due to all radioiodines and all radioactive materials in particulate form and radionuclides (other than noble gases) with half lives greater than 8 days. The methodology of Section 3.4 limits the doses to one half of these values for each of the two reactor units on the site.

The definition of Qjy in Equations 3.2.3, 3.2.4, and 3.2.5 states,

" Noble gases may be averaged over a period of I hour, and any other nuclides may be averaged over a period of I week." These time periods should be justified, since Section 2.2 of NUREG recommends the "second" as the practical unit for establishing release rates to show compliance with 12

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the requirements of 10 CFR 20. (Since these calculations are intended to .

verify compliance with instantaneous dose rate limits, the averaging period  ;

should be as short as is feasible with the installed instrumentation.) l Dose rates to the critical organ of an infant (for each radionuclide) due to gaseous radioactive effluents other than noble gases are calculated at the nearest receptor in each sector and summed over applicable pathways for comparison with the organ dose limit of Technical Specification 3.11.2.1. This methodology is expected to calculate a conservative (high) dose rate.

The statement in Section 3.4(b) concerning radionuclides that are to be -

considered in determining compliance with the limiting dose rate specified i in Technical Specification 3 ll 2.l(b) contains two errors. First,

" radionuclides" should read "radioiodines;" and second, the statement in the 00CM does not include tritium, which is included in the methodology.

r All assumptions used in the calculations of the Rgg's should be furnished in the ODCM, (e.g., it could be stated that the default values of Regulatory Guide 1.109 are used for consumption, transit times, etc.) The definition of Wmv should specify the use of (x/Q)jpy for carbon 14 for all pathways (See Equation C-8 in Regulatory Guide 1.109. .

(Note, however, that C 14 calculations could be omitted since it is not i included in recent NRC guidance (3).)

l The following corrections should be made in Tabic 3-4, Infant Dose Factors, which is used with Equation 3.25 for calculating the organ dose rate: (a) The ground plane dose factor for C 14 should be 0.0; (b) The l Cow-Hilk and Goat Milk dose factors for C-14 should be 3.23E+06 mrem /yr per #Ci/m3 instead of 2.34E+09 m 2(mrem /yr) per pCi/s, based on the dose methodology for C-14 in Section C.3.c and Sections 3.a and-3.b of l Appendix C of Regulatory Guide 1.109 ; and (c) Goat Milk dose factors for many radionuclides should be non-zero, since Section 1, Appendix E, of Regulatory Guide 1.109 specifies use of F, for the Cow Hilk pathway from Table E-1 when +here is no Fm given for the Goat-Hilk pathway in ,

Table E 2. The 3 in Equation 3.6 of Section 3.2 should be (3)l\2, l

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The methodology given in the ODCM to determine that the dose rates ,f offsite due to releases of radioactive materials in gaseous effluents are .

.not exceeded is, in general, within the guidelines of NUREG 0133 but the correctionf. identified in the previous paragraph should be made and the use of I hour as the averaging period for doses due to noble gas effluents should be justified. ,

3.7 Dose Due to Liouid Effluents Section 2.4 of the ODCM contains the methodology to determine that doses due to radioactive materials in liquid effluents are within the limits of '

Technical Specification 3.11.1.2, as required by Technical Specification 4.11.1.2. Technical Specification 3.11.1.2 requires that the dose or dose commitment to an individual due to liquid effluents from each reactor unit  !

be limited to 1.5 mrem to the total body and to 5 mrem to any organ during any calendar quarter, and to 3 mrem to the total body and to 10 mrem to any ,

organ during any calendar year.

Section 2.4 considers only doses due to adult fish consumption. The  ;

subject of doses due to shoreline sediment and to potable water and  !

invertebrate consumption should be addressed (e.g., either by adding [

methodology to calculate the doses or stating that the pathways do not exist, as appropriate). A subscript should be added to the F in the definition of Atg .  !

Table 2-3, Liquid Dose Conversion Factors, should include Mo 99, since it is one of the principal gamma emitters for which analysis is required by [

Notation e of Technical Specification Table 4.11-1, The Licensee may wish to consider the following' changes. Change the bioaccumulation factor for P in Table 2-2 from 1.00E+05 to the currently recommended value of 3.00E+03(9,10). Use the 1000 cfs per reactor dilution flow permitted by Section 4.3 of NUREG 0133 instead of "the average flow from the site discharge structure to unrestricted receiving waters times 5" in the definition of F , (since 3

sources other than the ODCM indicate that the product now used is less than 1000 cfs per reactor).

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.- j Calculate doses for the reporting periods, calendar quarter or calendar  !

year, using the average dilution flow for the reporting period instead of l the actual dilution flows during releases. (This is acceptable to the NRC  !

Staff, and permits independent determination of the calculated dose commitments for the reporting periods.)

The subjact of doses due to shoreline sediment and to potable water and  !

invertebrate consumption should be addressed in the ODCM. With this f

exception, the methodology to calculate dose commitments due to liquid effluents is considered to be within the guidelines of NUREG 0133.

3.8 Dose Due to Gaseous Effluents  !

Noble Gases. Section 3.6.1 of the ODCM contains methodology to determine offsite doses due to the release of noble gases from the site, as i required by Technical Specification 4.11.2.2. Technical Specification 4.11.2.2 requires that cumulative dose contributions be determined at least

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once per 31 days. Specification 3.11.2.2 requires that the air dose due to l noble gases released in gaseous effluents, from each reactor unit, from the j site be limited to less than or equal to 5 mrad per quarter and 10 mrad per i year for gamma radiation, and 10 mrad per quarter and 20 mrad per year for  !

beta radiation.  ;

The ODCM contains two methods for calculating the gamma and beta air i doses. " Method A" uses real time meteorological input and " Method B" uses l

annual average values of x/Q. The Mg should be defined in this  !

section. All release points, including the ILRT Vents, should be ,

specifically included in the calculations, possibly by identifying the vents included and summing over the subscript v.  !

t If all release points are specifically included in the calculations, the _

Licensee's methods to determine site boundary noble gas air doses are, in general, within the guidelines of NUREG 0133.

Radionuclides in particulate Form and Tritium. Section 3.6.2 of the ODCM contains methodology to determine "the dose to an individual from 15

t i

~

. j radiciodines and radioactive materials in particulate form, with half lives  ;

greater than 8 days, in gaseous effluents released from each reactor at the site to unrestricted areas." The methodology in tt4 section conforms '

closely to the requirements of Technical Specification 4.11.2.3, but the introductory statement in quotes above should be relocated so it applies to {

both Method A and Method B, and reworded to read. exactly like the technical l specification requirement. As it now reads it does not require calculation I of doses due to tritium (although the methodology does include tritium).  :

Technical Specification 4.11.2.3 requires that cumulative dose contributions be determined at least once per 31 days. Technical Specification 3.11.2.3 I requires that "the dose to an individual from radioiodines and radioactive  ;

materials in particulate form, and radionuclides (other than noble gases) {

with half lives greater than 8 days in gaseous effluents released, from each  ;

reactor unit, from the site," be limited to less than or equal to 7.5 mrem  ;

to any organ per quarter and to less than or equal to 15 mrem to any organ j per year.

The ODCM contains two methods for calculating doses due to radioiodines l and particulates. " Method A" uses real time meteorological input and f

" Method B" uses annual average values of x/Q and D/Q. Both of these methods should contain methodology to specifically include releases of

(

{

radioactive material from the !LRT Vents as well as releases from the Plant l Vent Stacks and the Turbine Building Vents (e.g., by indicating 6 summation i over y and specifying the vents included.) The basic equation for Method B should require summation over the applicable pathways, as does Equation 3.30 i for Method A.  !

i The definitions of R ipa' We jv, and W should specify the use l of(X/Q) jay for carbon 14 for all pathways (See Equation C 8 in Regulatory Guide 1.109). Alternatively, the calculation of doses due to  :

C-14 could be eliminated since recent NRC guidance does not require it(3). The definition of Rjpy, in Section 3.6.2 should reference Table 3 5 instead of "an appendix," Dose factors for Mo 99 should be added to Table 3-5, since it is in the list of principal gamma emitters in Notation e of Technical Specification Table 4.11 1. All assumptions used in the calculations of the Rgj's should be specified in the ODCM, 16

i J

(e.g., it ceuld be stated that the default values of Regulatory Guide 1.109 l are used for consumption, transit times, etc). i 4

i The determination of the " maximum exposed individual" by summing the l doses to all organs for each age group implies that~the maximum organ doses j for comparison with the dose limits of Technical Specification 3.11.2.3 are j calculated by this method. This is a very conservative method, since the  ;

dose limits apply to the dose to a single organ (with dose commitments due to all pathways and all radionuclides considered). The Licensee may wish to l revise the methodology of the ODCM to use the highest dose to a single organ i for comparison with the technical specification limits. I Both Method A and Method B for calculating doses due to iodines and i particulates are, in general, within the guidelines of NUREG 0133. However,  :

the deficiencies noted above should be addressed, e l

3.9 Dose Proitetions  !;

Sections 2.6 and 3.8 of the ODCM require projection of doses based on expected operating conditions as required by Technical Specifications 4.11.1.3.1 and 4.11.2.4.1 respectively. '

I 3.10 Djjarams of Effluent Pathways  !

l Figure 2 1 in the ODCM is a diagram of the liquid *adwaste treatment i system at FNP, and Figure 3 1 is a diagram of the gaseous release sources and release points. Figure 2 1 should be clarified or another figure added to indicate the pathways from sources to the release point, as discussed in ,

Section 3.1 above. No diagram of the solid waste management system is included in the ODCM. ,

3.11 Total Dose Section 4 of the ODCM contains a discussion of the Special Report l required by Technical Specification 4.11.4 if any of the dose limits of ,

Technical Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3 are exceeded.

Assessment techniques used to determine dose contributions from direct 17

l i

J radiation from the plant and its components will be documented in the  !

Special Report. The Licensee states that there are no other uranium fuel l cycle sources within a 50 mile radius. (The Licensee may desire to amend  !

Technical Specification 3.11.4 so it corresponds to Specification 3.11.4 of i NUREG 0472 Revision 3, Draft 7," which requires a special report only if the limits of 40 CFR 190 have been exceeded during a calendar year.) j 3.12 Environmental Monitorina Procram i

Section 5 of the ODCM contains a table and figures giving locations for )

obtaining each and every radiological environmental monitoring sample, as  !

required by Technical Specification 4.12.1, except for the locations of  !

indicator locations for a milk animals. The locations of milk animals as reported in the annual land use censuses changes frequently, so it is not certain if a milk animals existed within 5 miles of the plant at the time 3 the latest revision of the ODCH was prepared. However, if no milk animals l

exist within 5 miles of the plant this fact should be included in the ODCM. i For completeness, a future revision of the ODCM should modify the " sample f

location" column in the table of monitoring locations to unambiguously define the units of entries that appear to be distance in miles. The  !

descriptions of locations for obtaining radiological environmental l monitoring samples are, however, in general, within the guidelines of l l l NUREG 0133.

l j 3.13 Summarv l

The Licensee's ODCM uses documented and approved method: that are, in

, general, consistent with the methodology and guidance in NUREG-0133, Regulatory Guide 1.109, and Regulatory Guide 1.111. However, it is recommended that deficiencies noted in this review be addressed in a future .

revision of the ODCM.

l l 18 l

1

i

4. CONCLUSIONS The' Licensee's most recent complete 00CM, Revision 7, effective April 3, 1989 and transmitted to the NRC with a letter dated April 13, '

1989, was reviewed. It was determined that the ODCM uses methods that are, in general, consistent with the guidelines of MUREG 0133, Regulatory Guide 1.109, and Regulatory Guide 1.111. However, it is recommended that -

deficiencies noted in the review be addressed in a future revision of the ODCM.  ;

The following summarizes the identified deficiencies:

. In Section 2.1, another figure and/or a detailed description l should be added to clearly define the pathways of radioactive liquid waste and dilution water from each source to the the discharge point in the Chattahoochee River, ,

. Figure 2-1 should be expanded to show the solid waste management 3 system or a figure showing this system should be added to the  :

ODCM.

. Section 2.2 should include methodology to determine setpoints for the monitors on the Steam Generator Blowdowns, i

. The methodology to determine setpoints of liquid effluent monitors -

in Section 2.2 and compliance with the 10 CFR 20 limits in .

Section 2.3 should specifically address the simultaneous release ,

of radioactive materials from several sources, including the Turbine Building Sumps. Also, the source of dilution, water flow for each radioactive effluent stream should be identified.

. Liquid pathways other than fish consumption should be addressed in  ;

Section 2.4, either by including the methodology for calculatirg doses or stating that the pathways do not exist, as appropriate. <

19

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. Table 2-3, Liquid Dose Conversion Factors, should include Mo-99,  :

since it is one of the principal gamma emitters for which analysis is required by Notation e of Technical Specification Table 4.11-1. l

. In Section 3.3, the term " average" should be removed from the definition of Qjy, since the dose rate limits are instantaneous.

. The methodology in Section 3.3 to determine setpoints of the gaseous monitors should account for releases from the ILRT Vents.

. The averaging of noble gas releases in Section 3.4.over a period I of I hour and other releases over 1 week should be changed or justified, since NUREG-0133 recommends "I second" for the averaging period.  ;

. The following corrections should be made in Table 3-4, Infant Dose l

Factors, which is used with Equation 3.25 for calculating the organ dose rate: (a) The ground plane dose factor for C-14 should be 0.0; (b) The Cow-Milk and Goat-Milk dose factors for C-14 t

/ yr per yCi/m3 should be 3.23E+06 mrent instead of 2.34E+09 m 2 (mrem /yr) per pCi/s, based on the dose -

methodology for C-14 in Appendix C of Regulatory Guide 1.109 ; and (c) Goat Milk dose factors should be non zero for all nuclides ,

with a transfer factor in either Taole E-1 or E-2 in Regulatory Guide 1.109

. In Section 3.4(b), all assumptions used in calculations of the i Raj's should be furnished in the ODCM, (e.g., it could be stated that the default values of Regulatory Guide 1.109 are used for consumption, transit times, etc.)

. The definitions of Wmv in Section 3.4(b) and of Ripa' Wp jy, and W in Section 3.6.2 should specify the use of (X/Q)jev for carbon 14 for all pathways (See Equation C-8 in Regulatory Guide 1.109.) Alternatively, the calculation of doses due to C-14 could be eliminated since it is not required by recent NRC guidance (3).

20

• '4 l

• In Section 3.6.1 and 3.6.2, all release pointr,, including the ILRT

! Vents, should be specifically included in the calculations (e.g., by l identifying the vents included and summing over the subscript v.)

. The basic equation for Method B in Section 3.6.2 should require the summation of doses over the applicable pathways.

I .

The first sentence of Section 3.6.2 should be relocated so it applies to both Method A and Method B, and reworded so it requires i calculation of doses due to tritium releases.

u

. Indicator locations for milk sampling should be added to Table 51, ,

or an explanation given for their omission, t i

The following mainly editorial and typographical corrections should be made in a future revision of the ODCM:  !

. A subscript should be added to F in the fefinition of At s in Section 2.4

. The 3 in Equation 3.6 of Section 3.2 should be (3)1\2, A definition of Mj should be included in Section 3.6.1.  !

. In ODCM Figure 31, monitor R-13 should apparently be shown on the effluent line from all the Waste Gas Decay Tanks.

The word " radionuclides" in the first sentence of Section 3.4(b) should read "radioiodines." This sentence should also be reworded so doses due to tritium are required to be calculated.

l i The definition of Rjp7, in Section 3.6.2 should reference Table 3-5 instead of "an appendix."

L

. The units of distance should be specified in Table 5-1, 21 l

l

F'

• I i

i

0. , j l .

i

, The following are possible changes the Licensee may wish to consider.

. In Section 3.4, revise the technical specifications and/or the ODCM requirements so as to use the calculated dose to the thyroid of a  ;

child via the inhalation pathway at or beyond the site boundary as  !

the criteria for satisfying the organ dose limit for gaseous l effluents, as recommended by the basis statement for Specification  !

3/4.11.2.1 in NUREG 0472. ,

. Change the bioaccumulation factor for P in Table 2-2 from 1.00E+05 ,

to the currently recommended value of 3.00E+03(9,10),  ;

. In Section 2.4, use the 1000 cfs per reactor dilution flow permitted by Sectiori 4.3 of NUREG 0133 instead of "the average flow from the site discharge structure to unrestricted receiving waters times 5" in the definition of F , (since s

sources other than the ODCM indicate that the product now used is less than 1000 cfs per reactor).

. Calculate doses due to liquid effluents, in Section 2.4, for the reporting periods, calendar quarter or calendar year, using the ,

average dilution flow for the reportir,g period instead of the actual dilution flows during releases. (This is act.eptable to the NRC .;

Staff, and permit.c independent determination of the calculated dose  !

cormitments for the reporting periods.)

t  :

. Revise the methodology of Section 3.6.2 so the maximum dose to an I

I single organ is calculated for comparison with the organ dose limits of Technical Specification 3.11.2.3.

1

. Amend Technical Specification 3.11.4 and Section 4 of the ODCM so they correspond to Specification 3.11.4 of NUREG 0472, Revision 3, Draft 7," which requires a special report only if the limits of 40 CFR 190 have been exceeded during a calondar year.

22

[ ,,

5. REFERENCES

1. Title 10, gode of Federal Reculations, Part 50, Appendix 1, S= ical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion, 'As low As !s Reasonably Achievable,' for Radioactive Material in Light Water Cooled Nuclear Power Reactor

2. Letter from R. P. Mcdonald (APC) to U. S. Nuclear Regulatory Ccamission, [ transmitting the Joseph M. Farley Offsite Dose Calculation Manual (FNP-0-M-Oll, Rev 4)). April 13,1988.

3. " Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors," Rev. 3, Draft 7", intended for contractor guidance in reviewing RETS proprosals for operating reactors, NUREG 0472, September 1982.

4. " Preparation of Radiological Effluent Technical Specifications for >

Nuclear Power Plants, A Guidance Manual for Users of Standard Technical i Specifications," NUREG-0133, October 1978.

d

5. ' General Contents of the Offsite Oose Calculation Manual," Revision 1 Branch Technical Position, Radiological Assessment Branch, NRC, February 8, 1979. 7 i

6. " Calculation of Annual Doses to Man from Routine Releases of Reactor l Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, [

Appendix I," Regulatory Guide 1.109, Rev. 1, October 1977. l

7. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors," i Regulatory Guide 1.111, Revision 1 July 1977. l t

8. Title 10, Code of Federal Reaulations, Part 20, Appendix B,

" Concentrations in Air and Water above Natural Background."

4:

23 l

F 4 0

9. E. F. Branagan, C. R. Nichols, and C. A. Willis, "importance of P 32 in Nuclear Reactor Liquid Effluents," Health Physics Society Annual Meeting, June 1982.

10. B. Kahn and K. S. Turgeon, "The Bioaccumulation for Phosphorus-32 in Edible Fish Tissue," NUREG/CR-1336, March 1980, 24

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TE5iNdE'EVALUATIONREPORTFORTHEEVALUATION 0F ODCli REVISION 7 JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2

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.....a.. August 1989 T. E. Young ......n.....

T. S. Bohn **'- "a l

W. Serrano - Aucust 10A0

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Nuclear Sciences EG&G Idaho. Inc., *

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U. S. Nuclear Regulatory Commission

)ffice of Nuclear Reactor Regulation F na1

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Washington, D. C. 20555

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The Offsite Dose Calculation Manual (ODCri) for the Joseph M. Farley Nuclear Plant Units 1 and 2 contains current methodology and parameters used in the calculation of offsite doses due to radioactive liquid and gaseous

• effluents, ,

in the calculation of gasecus and liquid effluent monitoring alarm / trip set- i points, and in the conduct of the environmental radiological monitoring program, i A copy of the most recent complete ODCM was transmitted to the '4RC as Revision

• 7, with a letter dated April 13. 1939. The NPC transmitted Revision 7 to the Idaho National Engineering Laboratory (INEL) for review. The revised ODCM was reviewed by EG&G Idaho at the INEL and the results of the review are presented '

in this report. It was detemined that ODCM Revition 7 uses methods that are. '

in general. in agreement with the guidelines of NUREG-0133. However. it is recommended that some changes and additions be included in the Licensee's next revision to the ODCM, ,

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