ML20246J958
| ML20246J958 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/31/1988 |
| From: | Bohn T, Freeman, Serrano W EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20246J961 | List: |
| References | |
| CON-FIN-D-6034 EGG-PHY-8153, NUDOCS 8905170213 | |
| Download: ML20246J958 (33) | |
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l EGG-PHY-8153 July 1988 TECHNICAL EVALUATION REPORT x' '.
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i TECHNICAL EVALUATION REPORT FOR THE EVALUATION
);g OF ODCM UPDATED THP.00GH RE\\TSION 5 JAMES A. FITZPATRICK NUCLEAR POER PLANT
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EGG-PHY-8153 i
TECHNICAL EVALUATION REPORT for the i
EVALUATION OF ODCM UPDATED THROUGH REVISION 5 JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
NRC Docket No.150-333-NRC License No. DPR-59 l
i T. S. Bohn W. Serrano i
A. L. Freeman M. R. Winberg I
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Published July 1988 i
Idaho National Engineering Laboratory I
EG&G Idaho, Inc.
Prepared for the U. S. Nuclear Regulatory Commission 1
Washington, D. C. 20555 l
Under DOE Contract No. DE-AC07-76ID01570 i
FIN No. D6034 1
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l ABSTRACT The Offsite Dose Calculation Manual for the James A. FitzPatrick Nuclear Power Plant (JAFNPP) contains a description of the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program.
Revision 0, dated June 15, 1983, was submitted to the NRC with letter dated October 26, 1983 and was, in general, approved by the NRC on May 29, 1985.
Subsequent changes identified in Revisions 1 and 2 were reported to the NRC in the Semiannual Effluent Release Report for July-December 1985. The ODCM, updated threugh Revision 2 was, in general, approved on an interim basis by the NRC on October 2, 1986.
Changes identified in Revisions 3 and 4 were reported to the NRC in Semiannual Effluent Release Report for January-July 1986 and July-December 1986 respectively.
Also, Revision 5, dated March 1987, was submitted tn the NRC in letter dated April 2, 1987.
The NRC transmitted the ODCM updated through Revision 5 to the Idaho National Engineering Laboratory (INEL) for review.
The ODCM, complete through Revision 5, was reviewed in its entirety by EG&G Idaho at the INEL and the results of the review are presented in this report.
It was determined that the ODCM updated through Revision 5 uses methods that are, in general, in agreement with the guidelines of NUREG-0133. However, it is recommended that another revision to the ODCM be submitted to address the discrepancies identified in the review.
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s FOREWORD This report is submitted as partial fulfillment of the " Review of Radiological Issues" project being contracted by the Idaho National Engineering Laboratory for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation.
The U. S. Nuclear Regulatory Commission funded the work under FIN D6034 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 any third party's use, or the results of such use, of any information, 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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4 CONTENTS Page Abstract..............................
i Foreword..............................
11 1.
Introduction..........................
1 2.
Review Criteria..........................
2 3.
Evaluation...........................
3 4.
Conclusions...........................
18 5.
References............................
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l 1.
INTRODUCTION Puroose of Review This document reports the review and evaluation updated through Revision 5, of the Offsite Dose Calculation Manual (0DCM), submitted by the New York Power Authority, the Licensee for the James A. FitzPatrick Nuclear Power Plant (JAFNPP).
The ODCM is a supplementary document for i
implementing the Radiological Effluent Technical Specifications (RETS) in compliance with 10 CFR 50, Appendix I requirements. Ell i
Plant-Soecific Backoround The New York Power Authority submitted to the Nuclear Regulatory Commission (NRC) on October 26,1983,[2] ODCM Revision 0, dated June 15, 1983 for JAFNPP. The NRC reviewed the ODCM and found it to be, in general, acceptable as stated in letter dated May 29,1985.[3]
Subsequent changes identified in Revisions 1 and 2 were reported to the
. NRC in the Semiannual Effluent Release Report for July-December 1985.[43 The ODCM, updated through Revision 2 was, in general, approved on an interim basis by the NRC as stated in letter dated October 2, 1986.[5]
Changes identified in Revisions 3 and 4 were reported to the NRC in Semiannual Effluent Release Report.for January-June 1986,[6] and July-December 1986E73,respectively. Also, Revision 5, dated March 1987, was submitted to the NRC in letter dated April 2,1987.[8]
The NRC transmitted the ODCM updat'ed through Revision 5 to the Idaho National Engineering Laboratory (INEL) for review. The ODCM, updated through Revision 5, was reviewed in its entirety by EG&G Idaho at the INEL.
The results and conclusions of the review are presen; ; in this report.
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2.
REVIEW CRITERIA Review criteria for the ODCM were provided by the NRC in two documents:
NUREG-0473, RETS for BWRsl93 NUREG-0133, Preparation of RETS for Nuclear Power Plants.[10]
The following NRC guidelines were also used in the ODCM review:
" General Contents of the Offsite Dose Calculation Manual," Revision 1, Ell) and Regulatory Guide 1.109, Revision 1.[12]
As specified in NUREG-0473, the ODCM is to be developed by the Licensee to document the methodology and approaches used to calculate offsite doses and to 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 dose projections.
In addition, the ODCM should contain flow diagrams that define the treatment paths and the components of the radioactive liquid, gaseous, and solid waste management systems.
These flow diagrams should be consistent with the systems being used at the plant. A description and the location of samples in support of the environmental monitoring program are also needed in the ODCM.
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3.
EVALUATION JAFNPP is one of three operating BWR units located on the joint Nine Mile Point - James A. FitzPatrick site.
The other units are Nine Mile Point Units 1 (NMP-1) and 2 (NMP-2) and are located about 3000 ft to the west of JAFNPP.
The ODCM provides the methodology to calculate radiation doses to individuals in the vicinity of JAFNPP.
It also provides methodology for calculating effluent monitor setpoints and allowable release rates to ensure compliance with the Technical Specifications, Appendix B, of the New York Power Authority, JAFNPP Docket Number 50-333, and 10 CFR 20 release criteria.
Liauid Effluent Pathways JAFNPP is located on the southern shore of Lake Ontario in Oswego County, New York, approximately 36 miles northwest of Syracuse, New York.
JAFNPP has been in commercial operation since July of 1975.
Lake Ontario supplies water to the once-through circulating water system and the service water system which are used as the dilution water for liquid radwaste discharges. All radioactive liquid releases enter Lake Ontario where the circulating water discharge tunnel terminates on the lake bottom approximately 1400 ft from the shoreline.
The liquid radwaste system collects, monitors, and processes all potentially radioactive liquid wastes in a controlled manner.
Once the wastes are collected in waste tanks, they are processed by the liquid radwaste treatment system and are discharged from one of the following tanks:
Floor Drain Sample Tank, Waste Sample Tank, Laundry Drain Tank.
According to Figure F-5 of the ODCM (Figure 1 of this report) all 3
liquid radwaste effluent meant for release passes through the monitored liquid radwaste effluent line. According to the text in the ODCM, the radwaste effluents are then diluted with the circulating water.and service water before release to Lake Ontario. However, Figure F-5 of the ODCM does not show the point where the liquid radwaste effluent line joins the I
circulating water and the service water lines for dilution prior to discharge.
A simplified flow diagram of the solid radwaste treatment system is not included in the ODCM.
Liauid Effluent Monitor Setooints Sections 3.3.3 and 3.3.4 of the ODCM contain the methodology used to
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determine the setpoints for the radwaste liquid effluent and service water effluent monitors in compliance with Surveillance Technical Specification Appendix B, Section 2.1.a.
There is an automatic termination of release I
function for the liquid radwaste effluent line as indicated in Table 2.1-1 in Appendix B of the Technical Specification.
The setpoint for the liquid radwaste effluent line monitor is based on the radionuclides identified in each batch of liquid waste prior to i
release.
For continuous releases weekly composite samples are collected continuously in proportion to the rate of flow of the effluent stream and are analyzed.
Simultaneous releases are accounted for during both batch and continuous releases.
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In Section 3.3.3.b, the equation S - 0.5 x C/FL is given to determine the alarm / trip setpoint for the radwaste liquid effluent monitor.
However, when FL > 0.5, the value of the setpoint would be i
less than the e'x'pected count rate for the undiluted effluent at concentration C and the monitor would therefore alarm / trip continuously.
In Section 3.3.3.1.g a reference is made to Equation 3.3, however the equation identifier is missing for Equation 3.3 in Section 3.3.3.b.
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In Section 3.3.4 in Equation 3-3.c a summation over the nuclide index "i" is missing.
With the exception of the issues identified, the methodology for determining the alarm / trip setpoints for the liquid effluent radioactive monitors is within the guidelines of NUREG-0133 and is considered acceptable.
Gaseous Effluent Pathways According to Technical Specification 3.1.a, there are five gaseous environmental release points at JAFNPP:
Main Stack, Refuel Floor Vent, Reactor Building Vent, Turbine Building Vent, Radwaste Building Vent.
According to Appendix F, Table F-1, in the ODCM, the main stack has a height of 385-ft above grade and is considered an elevated release, whereas the other four are considered ground relear.es.
The. Technical Specifications and Appendix F, Table F-1 identify the five gaseous effluent release points listed above.
However, Figures F-1, F-2 and F-4 identify only four gaseous release points.
Figure F-4 also shows a simplified diagram of the gaseous radwaste treatment system.
Figure 2 in this report is a reproduction of Figure F-4.
Another diagram is needed however, showing a diagram of the components and pathways for the Standby Gas Treatment System (SBGTS).
j Technical Specification Appendix B, Section 3.6.a requires that the offgas treatment system be used to reduce the concentration of radioactive materials in gaseous effluents prior to release from the plant within 24 hrs after startup of the second turbine driven feedwater pump.
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Gaseous Radwaste Treatment Systems, Effluent Paths, and Controls for James'A. Fitzpatrick Nuclear Power Plant.
(Taken from Figure F-4 of the JAFNPP ODCM, Rev. 2 dated 1
June 1985.)
b According to the Bases statement, "this is due to the fact that excess air in-leakages in the main condenser as a result of operating only one turbine driven feedwater pump will exceed offgas treatment system limitations and consequently render the system inoperable.
Startup of the second turbine driven feedwater pump will decrease air in-leakage and assure offgas treatment system availability."
It is not clear why the feedwater pumps affect the in-leakage rate.
The Techniul Specifications identify noble gas monitors, iodine i
samplers and particulate samplers to monitor gaseous effluent releases.
Stack and vent effluents are continuously surveyed during release of noble gases.
Iodine and particulate samplers are routinely analyzed'in accordance with Technical Specification Appendix B, Table 3.2-1 Gaseous Effluent Monitor Setooints
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Section 4.3.2 of the ODCM contains the methodology used to determine the setpoints for the noble gas effluent monitors as required by Technical Specification Appendix B, Section 3.1.d.
Technical Specification Appendix B, Section 3.1.a states that radioactive gaseous wastes released to the environment from each of the five effluent release points shall be monitored and recorded. The monitors for releases from the refuel area exhaust and reactor building area exhaust have automatic isolation for the secondary containment prior to initiation of the SBGTS.
The total body dose rate limit is stated as being more restrictive than the skin dose rate and is used for determining the setpoints for the noble gas monitors.
The noble gas monitors maximum setpoints are based on the maximum concentration required to yield tne total body dose rate limit. The calculated setpoints account for simultaneous releases.
Technical Specification Appendix B, Section 3.2.a states that the dose rate limits apply to areas "at or beyond the site boundary."
Additionally, Figure 5.1-1 in the Technical Specifications shows the site 8
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boundary to include the JAFNPP as well as NMP-1 and NMP-2.
Therefore, it appears that the contribution from NMP-1 and NMP-2 must be considered when demonstrating compliance to the dose rate limit of Technical Specification Appendix B Section 3.2.a and not the dose rate from JAFNPP only.
In Section 4.3.2.b.(2) the value for the elevated release rate should be 2.1E+6 yCi/sec instead of 2.4E+6 yCi/sec.
With the exception of the uncertainty of including contributions for NMP-1 and NMP-2 when demonstrating compliance to the dose rate limit of Technical Specification Appendix B Section 3.2.a. the methods for determining the setpoints for the noble gas effluent monitors are, in general, in agreement with the guidelines of NUREG-0133 and are considered acceptable.
Concentrations in Liauid Effluents Section 3.2.2 contains the methodology for determining the radioactivity concentrations in liquid effluents as required by Surveillance Technical Specification Appendix B, Section 2.2.b.
In Section 3.2.2.a the word " percentage" should be removed from the definition for f.
Regardless, the methodology is within the guidelines 2
of NUREG-0133 and is considered acceptable.
Dose Rates in Gaseous Effluents Section 4.3.1 contains the methodology for determining dose rates from noble gases to the total body and skin as required by Surveillance Technical Specification Appendix B Section 3.2.a.
In Section 4.3.1.b.(2), the definition for DRTB for Equation 4-5 indicates that this quantity is the dose rate from nuclide i, whereas it in fact is the dose rate to the total body from all noble gas radionuclides.
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The definition of DRSKIN, for Equation 4-6, indicates that this quantity is the the dose rate from nuclide i, whereas it is in fact the dose rate to the skin from all noble gas radionuclides.
Also, the quantity DRSKINj is not defined for Equation 4-6, In Section 4.3.1.c.(7) in the equation for DRSKIN the release j
rate quantity, Q4 is missing.
With the exception of the discrepancies indicated above, the methodology for determining the dose rate due to the release of radioactive noble gases is in agreement with the guidelines of.NUREG-0133 and is considered acceptable.
j Section 4.3.3.b contains the methodology for determining the instantaneous dose rate due to the release of tritium, iodine-131, l
iodine-133 and all radionuclides in particulate form with half lives greater than 8 days for the inhalation pathway as required by Surveillance Technical Specification Appendix B, Section 3.2.b.
In Section 4.3.3 the text does not specifically indicate tritium as required in Technical Specification Appendix B, Section 3.2.b.
InSection4.3.3.b,inEquation4-9thequantityhjshouldbethe total release rate for nuclide i from all release points in order to determine a total dose rate. Also, in Equation 4-10, which determines the limiting iodine release rate, there should be a summation due to all radionuclides.
1 With the exception of the indicated discrepancies, the methodology for determining the dose rates due to the release of tritium, iodine-131, iodine-133 and all radionuclides in particulate form with half lives greater than 8 days is in agreement with the guidelines of NUREG-0133 and is considered acceptable.
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Dose Due to Liouid Effluents Section 3.4.1.b contains the methodology for determining the dose'due
' to radioactive material released in liquid effluents as required by Surveillance Technical Specification Appendix B, Section 2.3.a.
JAFNPP identifies fish and potable water pathways for the dose calculations, designating the adult as the maximum exposed individual and the total body and liver as the critical organs.
In Section 3.4.1.b the quantity Atj in Equation 3-4 should be
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the number of hours in the calendar year instead of the number of hours during liquid effluent releases for the calendar year.
Also, a summation of the dose due to all radionuclides is missing from Equation 3-4.
In Section 3.4.1.b the reference for the derivation of the quantity Air should be Appendix A, Table L-4.1 instead of Appendix A.
In Section 3.4.2.b.(1) the definition of Atj for Equation 3-5 should be the number of hours of the reporting period (e.g. monthly, quarterly, annually) and not the number of hours over which the release occurs. Also, the quantity (DF)) should be the total volume of dilution released during the reporting period. A summation of the dose due to all radionuclides should be included in Equation 3-5.
The methodology in Section 3.4.2.c should be made consistent with the previous comments in this review concerning the definition of time period Atj identified in Sections 3.4.1.b and 3.4.2.b.(1).
Appendix A, Table L-1 lists the maximum permissible concentrations in water in unrestricted areas.
The following discrepancies were detected by the reviewer:
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The first Zn-69 should be listed as Zn-65.
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The value for insoluble Y-93 should be 3E-4 pCi/ml instead of i
3E-5 pCi/ml.
Appendix A, Tables L-2 and L-3 contain liquid ddse conversion factors, Air for all age groups, pathways and radionuclides.
Using the methodology given in Appendix A, Table L-4.1 the reviewer (with the exception of the boating pathway) was unable to reproduce a large fraction of the dose conversion factors.
It is recommended that the Licensee ensure the validity of each and every dose conversion factor.
Appendix A, Table L-4.1 gives a derivation of the liquid dose factor Aj7 The following errors were detected by the reviewer:
In the definition for Z in Table L-4.1, the quantity M is not c
defined.
The units for the quantity Aj should be sec-1 instead of secs.
A reference should be provided supporting the quantities Dw, and Df in Equation L-4.1.
In the calculational example for swimming, the equation referred to should be Equation L-4.1 instead of Equation L-4.4.
Also, for the quantity DFi the table referred to should be Table L-4.4 instead of Table L-4-2.
The purpose for Table L-4.2 is unclear since all the activity values are unity and the table is not referred to in the ODCM.
The following tables are not referred to in the ODCM:
Tables L-4.3, L-5, L-5.1, and L-5.2.
The ODCM states that the dose convert, ion factors tubulated in Table L-4.4 for swimming and boating are generated using the 12
Impell computer code.
Either the methodology and parameters contained in the Impell code should be submitted for review or the Licensee should include the reference containing NRC approval for use of this code.
Primarily due to the incorrect definition for At1 and the values in the dose factor tables, the methodology for calculating the dose due to the release of radioactivity in liquid effluents is not in agreement with
.the guidelines of NUREG-0133 and is not considered acceptable.
Dose Due to Gaseous Effluents Sections 4.4.1.b.(1), 4.4.2.a and 4.4.2.b contain the methodologies for calculating the cumulative gamma and beta air doses due to the release of radioactive noble gases as required by Surveillance Technical Specification Appendix B, Section 3.3.a.
In Section 4.4.2.a.(2)(a), the reference for the detailed explanation and evaluation of'the quantity Meff should be Appendix E instead of Appendix C.
Also, the value for (X/Q) on Page 42 of the ODCM, for the 3
3 elevated release should be 3.8E-8 sec/m instead of 3.8E-7 sec/m,
In Section 4.4.2.b.(2) the second half of the first paragraph is evidently out of place since it discusses iodines and particulate in a section describing noble gas beta air dose methodology.
With the exception of these two discrepancies, the method for calculating -the air dose due to the release of radioactive noble gases is in agreement with the guidelines of NUREG-0133 and is considered acceptable.
Sections 4.4.1.b.(2) and 4.4.2.c contain the methodologies for calculating the cumulative dose due to the release of I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than eight days as required by Surveillance Technical Specification 13
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Appendix B, Section 3.4.a.
In Sections 4.4.1.b.(2) and 4.4.2.c, tritium has been omitted from the radionuclides included for determining cumulative dose but is required by l
Technical Specification Appendix B, Section 3.4.b.
I In Section 4.4.1.b.(2) the reference for the derivation of Rj7, should be to Appendix B, Table G-7.lB instead of Appendix B, Table G-8.
I In Section 4.4.2.c.(2) the subscript D should be removed from the quantity (D/Q) in Equations 4-25 and 4-26.
I Appendix B, Tables G-4, G-5, and G-6 contain gaseous dose conversion
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factors, Pj and Rj for all age groups, pathways, and radionuclides.
Using the methodology given in Appendix A, Table G-7 the reviewer was unable to reproduce a large fraction o'f the dose conversion factors.
It is recommended that the Licensee check the validity.of each and every dose conversion factor.
Appendix B, Tables G-7 gives a derivation of gaseous dose factors, Pg and R.
The following discrepancies were detected by the reviewer:
j In Table G-7.lA tritium has been omitted from the radionuclides included for determining the inhalation dose factor but is I
required by the Technical Specifications.
1 In Table G-7.1B methodology for determining the gaseous dose factors due to tritium in the cow-milk, cow-meat, goat-milk and fruit-vegetable pathways is missing.
In Table G-7.18,Section I the quantity (DFAi) should be written (DFA1)a to denote the age group.
j In Table G-7.18,Section III in the example calculation, the 1
quantity (DFli)g should be written (DFli)a-l l
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In Table G-7.1B the Rj factor values determined in the example
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calculations for Co-60 (child - total body) for all pathways i
l except inhalation are referenced as being listed in Appendix B, l
' Table G-6 in the adult portion instead of the child portion of the table.
The purpose of Table G-7.2 is unclear since all the activity values are unity and the table is not referred to in the ODCM.
-i The purpose for'some of the pages in Table G-7.3 is not clear.and also the table is not referred to in the ODCM.
The methodology for calculating the cumulative dose due to the release of I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than eight days is in agreement with the guidelines of -
l Regulatory Guide 1.109, Revision I and NUREG-0133 and is considered acceptable. However, since the reviewer was unable to verify the dose factors it is uncertain if the methodology can be used to demonstrate compliance to the Technical Specifications.
Dose Projections i
Section 3.5.2 of the ODCM describes the method used to project doses due to the expected releases of radioactive liquid effluents to determine when the liquid radwaste treatment system should be operated as required in Surveillance Technical Specification Appendix B, Section 2.4.a.
The methodology for determining the dose projection due to liquid radwaste effluents is in agreement with NUREG-0133 and is considered acceptable.
Section 4.5.2 of the ODCM describes the method used to project doses due to release of radioactive gases when the offgas treatment system is not in use as required in Surveillance Technical Specification Appendix B, Section 3.6.a.
It does not appear that there are ventilation exhaust treatment systems and the gaseous radwaste treatment system is in use whenever the second steam driven feedwater pump is in operation.
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Therefore, it is not clear why dose projections are required by the Technical Specifications since the purpose for the dose projections for a BWR is to determine required use of the ventilation exhaust treatment systems.
Section 4.5.2.a refers to the limits of Technical Specification Appendix B, Section 3.6.a whereas no limits are found in this Technical Specification.
Regardless, the methods for determining dose projections are within the guidelines of NUREG-0133 and are considered acceptable.
Total Dose Section 5.0 of the ODCM contains the methodology for calculating the total' dose contributions including direct radiation as require'd by Surveillance Technical Specification Appendix B, Section 5.1.a.
In Section 5.1 in the first sentence of the first paragraph the text should read: " Dose evaluation to demonstrate compliance with the 40 CFR 190 dose limits need only be performed if the quarterly s.t annual doses... exceed twice the dose limits of Technical Specifications Appendix B, Sections.2.3.a, 3.3.a.p.c 3.4.a respectively", in order to be consistent with the Technical Specifications.
With the exception of the indicated discrepancy, the methodology is in agreement with the requirements of the Technical Specifications and is considered acceptable.
Environmental Monitorina Procram
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Table H-1 in Appendix H of the ODCM contains specific parameters of distance and the direction sector from the site and additional information for each and every sample identified in Surveillance Technical Specification Section 6.1.
The environmental monitoring program is in compliance with Technical Specification Appendix B, Table 6.1-1 and is considered acceptable.
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l Summar.y In summary, the Licensee's ODCM uses documented and approved methods that are, in general, consistent with the methodology and guidance in NUREG-0133 and Regulatory Guide 1.109, Revision 1.
However, due primarily to the incorrect definition of Atj and the erroneous values for the liquid and gaseous dose factors it is recommended that the NRC request another revision.
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CONCLUSIONS The Licensee's ODCM updated through Revision 5, dated. March 1987 for JAFNPP was reviewed.
It was determined that the ODCM uses methods that are, in general, consistent with the guidelines of NUREG-0133. The methodology in most sections of the 00CM is acceptable for use in demonstrating compliance to the radiological effluent technical specifications.
However, it is recommended that another revision to the ODCM be submitted to address the discrepancies identified in the review.
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In Section 3.2.2.a the word " percentage" should be removed from the definition for f -
2 In Section 3.3.3.b, the equation S = 0.5 x C/FL is given to determine the alarm / trip setpoint for the radwaste liquid effluent monitor. However, when FL > 0.5, the value of the setpoint would be less than the expected count rate for the undiluted effluent at concentration C and the monitor would therefore alarm / trip continuously.
In Section 3.3.3.1.g a reference is made to Equation 3.3, however the equation identifier is missing for Equation 3.3 in Section 3.3.3.b.
In Section 3.3.4, a summation over the nuclide index "i" in Equation 3-3.c, is missing.
In Section 3.4.1.b the reference for the. derivation of the quantity Air should be more specifically Appendix A, Table L-4.1 instead of Appendix A.
In Section 3.4.1.b the quantity At1 in Equation 3-4 should be the number of hours in the calendar year instead of the number j
of hours during liquid effluent releases for the calendar year.
18 l
L.
Also, a summation of the dose due to all radionuclides is missing.
from Equation 3-4.
In Section 3.4.2.b.(1) the definition of Atj for Equation 3-5 should be the number of hours of the reporting period (e.g. monthly, quarterly, annually) and not the number of hours over which the release occurs. Also, the quantity (DF)1 should be the total volume of dilution released during the reporting period. A summation of the dose due to all radionuclides should be included in Equation 3-5.
The methodology in Section 3.4.2.c should be made to.'be consistent with the previous comments in this review concerning the definition of the time period Atj identified in Sections 3.4.1.b and 3.4.2.b.(1).
A simplified flow diagram of the solid radwaste treatment system is not included in the ODCM.
In Section 4.3.1.b.(2), the definition for DRTB for Equation 4-5 indicates that this quantity is the dose rate from nuclide i, whereas it in fact is the dose rate to the total body from all noble gas radionuclides.
The definition of DRSKIN, for Equation 4-6, indicates that this quantity is the the dose rate from nuclide 1, whereas it is in fact the dose rate to the skin from all noble gas radionuclides.
I f
In Section 4.3.1.b.(2), the quantity DRSKINj is not defined for Equation 4-6.
In Section 4.3.1.c.(7), in the equation for DRSKINj, the parameter Q is missing.
19
p L
l l
Technical Specification Appendix B, Section 3.2.a states that the dose rate limits apply to areas "at or beyond the site I
boundary." Additionally, Figure 5.1-1 in the Technical Specifications shows the site boundary to include the JAFNPP as l
well as NMP-1 and NMP-2. Therefore, it appears that the-contribution from NMP-1 and NMP-2 must be considered when demonstrating compliance to the dose rate limit of Technical Specification Appendix B Section 3.2.a and not the dose rate from JAFNPP only.
In Section 4.3.2.b.(2) the value for the elevated release rate should be 2.1E+6 sci /sec instead of 2.4E+6 sci /sec.
In Section 4.3.3 the text does not specifically indicate tritium as required in Technical Specification Appendix B, Section 3.2.a.b.
In Section 4.3.3.b, in Equation 4-9 the quantity 6 should be 4
the-t'otal release rate for nuclide i from all release points, in order to determine a total dose rate.
In Section 4.3.3.b in Equation 4-10, which determines the limiting iodine release rate, there should be a summation due to all radicauclides.
In Sections 4.4.1.b.(2) and 4.4.2.c, tritium has been omitted from the radionuclides included for determining cumulative dose but is required by Technical Specification Appendix B, Section 3.4.b.
In Section 4.4.1.b.(2) the reference for the derivation of Rj7, should be to Appendix B, Table G-7.18 instead of Appendix B, Table G-8.
20
4 In Section 4.4.2.a.(2)(a), the reference for the detailed explanation and evaluation of the quantity Meff should be Appendix E instead of Appendix C.
In Section 4.4.1.a.(2) the value for (X/Q) on Page 42 of the 3 instead ODCM, for the elevated release should be 3.8E-8 sec/m 3
of 3.8E-7 sec/m,
In Section 4.4.2.b.(2) the second half of the first paragraph seems to be displaced since it discusses iodines and particulate in a section describing noble gas beta air dose methodology.
In Section 4.4.2.c.(2) the subscript D should be removed from the quantity (D/Q) in Equations 4-25 and 4-26.
Section 4.5.2.a refers to the limits of Technical Specification Appendix B, Section 3.6.a however, no limits are found in this Technical Specification.
In Section 5.1 in the first sentence of the first paragraph the text should read: " Dose evcluation to demonstrate compliance with the 40 CFR 190 dose limits need only be performed if the quarterly e annual doses... exceed twice the dose limits of Technical Specifications Appendix B, Sections 2.3.a, 3.3.a E 3.4.a respectively", in order to be consistent with the Technical Specifications.
i Appendix A, Table L-1 lists the maximum permissible concentrations in water in unrestricted areas.
The following discrepancies were detected by the reviewer:
The first Zn-69 should be listed as In-65.
i The value for insoluble Y-93 should be 3E-4 gCi/ml instead of 3E-5 #Ci/ml.
21 i
o
l Appendix A, Tables L-2 and L-3 contcin liquid dose conversion factors, Air for all age groups, pathways and radionuclides. Using the methodology given in Appendix A, Table L-4.1 the reviewer (with the exception of the boating pathway) was unable to reproduce a large fraction of the dose conversion factors.
It is recommended that the Licensee ensure the validity of each and every dose conversion factor.
Appendix A, Table L-4.1 gives a derivation of the liquid dose factor Aj7 The following errors were detected by the reviewer:
In the definition for Z in Table L-4.1, the quantity M is c
not defined.
The units for the quantity Aj should be sec-1 instead of secs.
A reference should be provided for the quantities Dw, and Df in Equation L-4.1.
In the calculational example for swimming, the equation referred to should be Equation L-4.1 instead of Equation L-4.4.
Also, for the quantity DFi the table referred to should be Table L-4.4 instead of Table L-4-2.
The purpose for Table L-4.2 is unclear since all the activity values are unity and the table is not referred to on the ODCM.
I The following tables are not referred to in the 00CM:
Tables L-4.3, L-5, L-5.1, and L-5.2.
j l
The ODCM states that the dose conversion factors tabulated in Table L-4.4 for swimming and boating are generated using 22
__x_
the Impe11 computer code.
Either the methodology and parameters contained in the Impell code should be submitted for review or the Licensee should include the reference containing NRC approval for use of this code.
Appendix B, Tables G-4, G-5, and G-6 contain gaseous dose conversion factors, Pj and Rj for all age groups, pathways, and radionuclides. Using the methodology given in Appendix A, Table G-7 the reviewer was unable to reproduce a large fraction of the dose conversion factors. It is reconaended that the Licensee check the validity of e,ach and every dose conversion factor.
Appendix B, Tables G-7 gives a derivation of gaseous dose factors, Pj and Rj. The following discrepancies were detected by the reviewer:
In Table G-7.lA tritium has been omitted from the radionuclides included for determining the inhalation dose factor but is required by the Technical Specifications.
In Table G-7.lB methodology for determining the gaseous dose factors due to tritium in the cow-milk, cow-meat, goat-milk and fruit-vegetable pathways is missing.
In Table G-7.1B,Section I the quantity (DFAi) should be written (DFA1)a to denote the age group.
In Table G-7.1B,Section III in the example calculation, the quantity (DFL1)j should be written (DFli)a*
In Table G-7.lB the Rj factor values determined in the example calculations for Co-60 (child - total body) for all pathways except inhalation are referenced as being listed in Appendix B, Table G-6 in the adult portion instead of the 23 L- _
______1-____
~
child portion of the table.
The purpose of Table G-7.2 is unclear since all the activity values are unity and the table is not referred to in the ODCM.
The purpose for some of the pages in Table G-7.3 is not clear and also the table is not referred to-in the ODCM.
The Technical Specifications and Appendix F, Table F-1 identify the five gaseous effluent release _ points listed above.
- However, Figures F-1, F-2 and F-4 identify only four gaseous r'elease points.
Figure F-4 also shows a simplified diagram of the gaseous radwaste treatment system. Another diagram is needed however, showing a diagram of the components and pathways for the Standby Gas Treatment System (SBGTS).
In Figure F-5 of the ODCM the diagram showing the liquid effluent release paths does not show the point where the liquid radwaste effluent line joins the circulating water and the service water lines for dilution prior to discharge.
l l
l l
24
5.
REFERENCES 1.
TJ:
10, Code of Federal Regulations, Part 50, Appendix I, " Numerical bu;oes for Design Objectives and Limiting Conditions for Operation to Meet the Criterion, 'As Low As Is Reasonably Achievable,' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents. "
2.
Letter from J. P. Bayne (New York Power Authority) to D. B. Vassallo (NRC),
Subject:
James A. FitzPatrick Nuclear Power Plant, Off-Site Dose Calculation Manual, October 26, 1983.
3.
Letter from H. I. Abelson (NRC) to J. P. Bayne (New York Power Authority),
Subject:
Approval of James A. FitzPatrick Nuclear Power P1 ant ODCM, May 29, 1985.
4.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. FitzPatrick Nuclear Power Plant, Semiannual Radioactive Effluent Release Reoort, February 28, 1986.
5.
Letter from H. I. Abelson (NRC) to J. C. Brons (New York Power Authority),
Subject:
Revision to the Offsite Dose Calculation Manual (0DCM), October 2,1986.
6.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. FitzPatrick Nuclear Power Plant, Semiannual Radioa,tive Effluent Release Reoort, August 29, 1986.
7.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. FitzPatrick Nuclear Power Plant, Semiannual Radioactive Effluent Release Report, February 27, 1987.
8.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. FitzPatrick Nuclear Power Plant, Docket No.
50-333, License No. DPR-59, Revision to the Off-Site Dose Calculation Manual (ODCM), April 2,1987.
25 b
p 9.
" Standard Radiological Effluent Technical Specifications for Boiling Water Reactors," Rev. 3, Draft 7", intended for contractor guidance in reviewing RETS proposals for operating reactors, NUREG-0473, September
'1982..
- 10. " Preparation of Radiological Effluent Technical Specifications for Nuclear, Power Plants, A Guidance Manual for Users of Standard Technical Specifications," NUREG-0133, October 1978.
I
- 11. " General Contents of the Offsite Dose Calculation Manual," Revision 1, Branch Technical Position, Radiological Assessment Branch, NRC,
. February 8, 1979.
- 12. " Calculation of Annual Doses of Evaluating Compliance with 10 CFR.50, Appendix I," Regulatory Guide 1.109, Rev. 1, October 1977.
4 b
26
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Fitzpatrick Nuclear Power Plant
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Nuclear Sciences EG&G Idaho, anc.
s' P. O. Box 1625 Idaho Falls, ID 83415
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Washington, DC 20555 13.uPPL u.NT an y me.T..
5 3. as.TR ACT,Jo. euru. ar 'sual The Offsite Dose Calculation Manual for the James A. FitzPatrick Nuclear Power Plant contains a description of the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints, and in the :onduct of the environmental radiological monitoring program. The NRC transmitted the ODCM updated through Revision 5 to the Idaho National Engineering Laboratory for review. The ODCM, complete through Revision 5, was reviewed in its entirety by EG&G~ Idaho at the INEL and the results of the review are resented in this report.
It was determined that the ODCM updated through Revision 5 uses methods that are, in general, in agreement with the guidelines of NUREG-0133. However, it is recommended that another revision to the ODCM be submitted to address the discrepancies identified in'the review.
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Docket No. 50-333 DISTRIBUTION Docket file NRC PDR PDI-1 Rdg Local PDR Mr. John C. Brons SYarga BBoger Executive Vice President, Nuclear Generation RCapra CVogan Power Authority of the State of New York DLaBarge OGC 123 Main Street EJordan BGrimes White Plains, New York 10601 WMeinke ACRS(10)
JJohnson, RI
Dear Mr. Brons:
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT, REVISION 5 TO THE OFFSITE DOSE CALCULATION MANUAL (TAC NO. 65216)
REFERENCES:
(a) October 2,1986 letter from H. I. Abelson to J. C. Brons (b) April 2, 1987 letter from R. J. Converse to T. E. Murley Reference (a) documented staff acceptance with certain caveats of Revision 2 to the Offsite Dose Calculation Manual (0DCM) for the James A. FitzPatrick Nuclear Power Plant.
In Reference (b) you submitted a complete Revision 5 to the ODCM and provided a response to questions raised in Refereme (a).
Pevision 5 to the ODCM has been reviewed for us in its entirety by our contractor, EG8G Idaho, Inc. Their discussion and comments are contained in the attachec' TER, EGG.PHY-8153. We agree with the contractor that Revision 5 to the ODCM uses methods that are, in general, consistent with staff guidelines and is an acceptable interim reference. However, there are a number of discrepancies and suggestions listed in the TER, some typographical in nature, others dealing with technical issues. Since these questions involve refinements to a basically sound document, there is no need for immediate action.
However, we ask that all of the points rt.ised in the Conclusions section of the TER be addressed when the next revision to the ODCM is issued or at least prior to startup for the 1990 Refueling Outage.
It is not necessary to stbmit the ODCM revision to us for review, but a record of the disposition of ecen item should be available for an audit.
This closes TAC No. f5216.
Sincerely, l
ORIGINAL SIGNED BY David E. LaBarge, Project Director l
Project Directorate I 1 Division of Reactor Pro,1ects I/II
Enclosure:
Technical Evaluation Report cc: See next page
[ REVISION 5 TAC 65216]
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EGG-PHY-8153 July 1988 TECHNICAL EVALUATION REPORT
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TECHNICAL EVALUATION REPORT FOR THE EVALUATION OF ODCM UPDATED THROUGH REVISION 5 JAMES A. FITZPATRICK NUCLEAR POER PLANT
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2 T. S. Eohn W. Serrano A. L. Freeman M. R. Winberg 4
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Prepared for the
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U.S. NUCLEAR REGULATORY COMMISSION s
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EGG-PHY-8153 TECHNICAL EVALUATION REPORT for the EVALUATION OF ODCM UPDATED THROUGH REVISION 5 JAMES A. FITZPATRICK NUCLEAR POWER PLANT NRC Docket No. 50-333 NRC License No. DPR-59 T. S. Bohn W. Serrano A. L. Freeman M. R. Winberg Published July 1988 Idaho National Engineering Laboratory EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the i
U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. D6034
_ _ _ _ - _ _- _A
4 1
ABSTRACT The Offsite Dose Calculation Manual for the James A. FitzPatrick Nuclear Power Plant (JAFNPP) contains a description of the methodology and l
parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program.
Revision 0, dated June 15, 1983, was submitted to the NRC with letter dated October 26, 1983 and j
was, in general, approved by the NRC on May 29, 1985.
Subsequent changes identified in Revisions I and 2 were reported to the NRC in the Semiannual Effluent Release Report for July-December 1985.
The ODCM, updated through Revision 2 was, in general, approved on an interim basis by the NRC on October 2, 1986.
Changes identified in Revisions 3 and 4 were reported to the NRC in Semiannual Effluent Release Report for January-July 1986 and July-December 1986 respectively. Also, Revision 5,datedMarch198Nwas submitted to the NRC in letter dated April 2, 1987.
The NRC transmitted the ODCM updated through Revision 5 to the Idaho National Engineering Laboratory (INEL) for review.
The ODCM, complete through Revision 5, was reviewed in its entirety by EG&G Idaho at the INEL and the results of the review are presented in this report.
It was determined that the ODCM updated through Revision 5 uses methods that are, in general, in agreement with the guidelines of NUREG-0133.
However, it is recommended that another revision to the ODCM be submitted to address the discrepancies identified in the review.
i
l
^
l FOREWORD l
This report is submitted as partial fulfillment of the " Review of Radiological Issues" project being contracted by the Idaho National I
Engineering Laboratory for the U. S. Nuclear Regulatory Commission,' 0ffice i
of Nuclear Reactor Regulation.
The U. S. Nuclear Regulatory Commission y
funded the work under FIN D6034 and NRC B&R Numhe.' 20 19 05 03.
l 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 any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately-owned rights.
ii
CONTENTS Page r
Abstract..............................
i Foreword................................
11 l.- Introduction..........................
1
- 2. Review Criteria...........................
-2
- 3. Evaluation.....-......................
3-
- 4. Conclusions............................
18 i
5.
References...........,..................
26 I
1 l
iii j
l
+-
.)
1.
INTRODUCTION
)
Puroose of Review This document reports the review and evaluation updated through Revision 5, of the Offsite Dose Calculation Manual (0DCM), submitted by the New York Power Authority, the Licensee for the James A. FitzPatrick Nuclear Power Plant (JAFNPP).
The ODCM is a supplementary document for
~
implementing the Radiological Effluent Technical Specifications (RETS) in ccepliance with 10 CFR 50, Appendix I requirements. Ell Plant-Soecific Backaround The New York Power Authority submitted to the Nuclear Regulatory Commission (NRC) on October 26,1983,[2] ODCM Revision 0, dated June 15, 1983 for JAFNPP.
The NRC reviewed the ODCM and found it to be, in general, acceptable as stated in letter dated May 29,1985.[3]
~
Subsequent changes identified in Revisions 1 and 2 were reported to the NRC in the Semiannual Effluent Release Report for July-December 1985.[4]
The ODCM, updated through Revision 2 was, in general, approved on an interim basis by the NRC as stated in letter dated Oci:ober 2, 1986.[5] Changes iden'tified in Revisions 3 and 4 were reported to the-NRC in Semiannual Effluent Release Report for January-June 1986,[6] and July-December 1986E73,respectively.
Also, Revision 5, dated March 1987, was submitted to the NRC in letter dated April 2,1987.[8] The NRC transmitted the ODCM updat'ed through Revision 5 to the Idaho National Engi n ering Laboratory (INEL) for review. The ODCM, updated through Revision 5, was reviewed in its entirety by EG&G Idaho at the-INEL.
The results and conclusions of the review are presented in this report.
I
2 2.
REVIEW CRITERIA Review criteria for the ODCM were provided by the NRC in two documents:
NUREG-0473, RETS for BWRsl93 NUREG-0133, Preparation of-RETS for Nuclear Power Plants.[10]
The following NRC guidelines were also used in the ODCM review:
" General Contents of the.0ffsite Dose Calculation Manual," Revision 1,[113 and Regulatory Guide 1.109, Revision 1.[12]
As specified in NUREG-0473, the ODCM is to be developed by the Licensee to document the methodology and approaches used to calculate offsite doses and to 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 dose projections.
In addition, the ODCM should contain flow diagrams that define the treatment paths and the components of the radioactive liquid, gaseous, and solid waste management systems. These flow diagrams should be consistent-with the systems being used at the plant.
A description and the location of samples in support of the environmental monitoring program are also needed in the 00CM.
)
2
)
a
3.
EVALUATION JAFNPP is one of three operating BWR units located on the joint Nine Mile Point - James A. FitzPatrick site.
The other units are Nine Mile Point Units 1 (NMP-1) and 2 (NMP-2) and are located about 3000 ft to the west of JAFNPP.
The ODCM provides the methodology to calculate radiation doses to individuals in the vicinity of JAFNPP.
It also provides methodology for calculating effluent monitor setpoints and allowable release rates to ensure compliance with the Technical Specifications, Appendix B, of the New York Power Authority, JAFNPP Occket Number 50-333, and 10 CFR 20 release criteria.
Liouid Effluent Pathways JAFNPP is located on the southern shore of Lake Ontario in Oswego County, New York, approximately 36 miles northwest of Syracuse, New York.
JAFNPP has been in commercial operation since July of 1975.
Lake Ontario supplies water to the once-through circulating water system and t'he service water system which are used as the dilution water for liquid radwaste discharges.
All radioactive liquid releases enter Lake Ontario where the circulating water discharge tunnel terminates on the lake bottom approximately 1400 ft from the shoreline.
The liquid radwaste system collects, monitors, and processes all potentially radioactive liquid wastes in a controlled manner. Once the wastes are collected in waste tanks, they are processed by the liquid radwaste treatment system and are discharged from one of the following tanks:
Floor Drain Sample Tank, Waste Sample Tank, Laundry Drain Tank.
According to Figure F-5 of the ODCM (Figure 1 of this report) all i
3 4
__._____....____.______._m.
l liquid radwaste effluent meant for release passes through the monitored i
liquid radwaste effluent line. According to the text in the ODCM, the radwaste effluents are then diluted with the circulating water and service
{
water before release to Lake Ontario.
However, Figure F-5 of the ODCM
]
does not show the point where the liquid radwaste effluent line joins the circulating water and the service water lines for dilution prior to discharge.
A simplified flav diagram of the solid radwaste treatment system is not included in t i. UDCM.
Liouid Effluent Monitor Setooints Sections 3.3.3 and 3.3.4 of the ODCM contain the methodology used to determine the setpoints for the radwaste liquid effluent and service water effluent monitors in compliance with Surveillance Technical Specification Appendix B, Section 2.1.a.
There is an automatic termination of release function for the liquid radwaste effluent line as indicated in Table 2.1-1 in Appendix B of the technical Specification.
The setpoint for the liquid radwaste effluent line monitor is based on the radionuclides identified in each batch of liquid waste prior to release.
For continuous releases weekly composite samples are collected continuously in proportion to the rate of flow of the effluent stream and are analyzed.
Simultaneous releases are accounted for during both batch and continuous releases.
In Section 3.3.3.b, the equation S = 0.5 x C/FL is given to determine the alarm / trip setpoint for the radwaste liquid effluent monitor.
However, when FL > 0.5, the value of the setpoint would be
~
less than the e'x'pected count rate for the undiluted effluent at concentration C and the monitor would therefore alarm / trip continuously.
In Section'3.3.3.1.g a reference is made to Equation 3.3, however the
]
equation identifier is missing for Equation 3.3 in Section 3.3.3.b.
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In Section 3.3.4 in Equation 3-3.c a summation over the nuclide index "1" is missing.
With.the exception of the-issues identified, the methodology for determining the alarm / trip setpoints for the liquid effluent radioact) e monitors is within the guidelines of NUREG-0133 and is considered acceptable.
Gaseous Effluent Pathways According to Technical Specification 3.1.a, there are five gaseous environmental release points at JAFNPP:
Main Stack, Refuel Floor Vent, Reactor Building Vent, a
Turbine Building Vent, Radwaste Building Vent.
According to. Appendix F, Table F-1, in the ODCM, the main stack has a height of 385-ft above grade r 1 is considered an elevated release, whereas the other four are considered ground releases.
The. Technical Specifications and Appendix F, Table F-1 identify the five gaseous effluent release points listed above. However, Figures F-1, F-2 and F-4 identify only four gaseous release points.
Figure F-4 also shows a simplified diagram of the gaseous radwaste treatment system.
Figure 2 in this report is a reproduction of Figure F-4.
Another diagram is needed however, showing a diagram of the componer.ts and pathways for the Standby Gas Treatment System (SBGTS).
Technical Specification Appendix B, Section 3.6.a requires that the offgas treatment system be used to reduce the concentration of radioactive materials in gaseous effluents prior to release from the plant within 24 hrs after startup of the second turbine driven feedwater pump.
6
=
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.ees,e,,
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,,e,,eee l
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Figure 2.
Gaseous Radwaste Treatment Systems, Effluent Paths, and Controls for James A. Fitzpatrick Nuclear Power Plant.
i (Taken from Figure F-4 of the JAFNPP ODCM, Rev. 2 dated June 1985.)
According to the Bases statement, "this is due to the fact that excess air in-leakages in the main condenser as a result of operating only one turbine driven feedwater pump will exceed offgas treatment system limitations and consequently render the system inoperable.
Startup of the second turbine driven feedwater pump will decrease air in-leakage and assure offgas treatment system avai1 ability."
It is not clear why the feedwater pumps affect the in-leakage rate.
I The Technical Specifications identify noble gas monitors, iodine samplers and particulate samplers to monitor gaseous effluent releases.
Stack and vent effluents are continuously surveyed during release of noble gases.
Iodine and particulate samplers are routinely analyzed'in accordance with Technical Specification Appendix B, Table 3.2-1 Gaseous Effluent Monitor Setooints Section 4.3.2 of the ODCH contains the methodology used to determine the setpoints for the noble gas effluent monitors as required by technical Specification Appendix B, Section 3.1.d.
Technical Specification Appendix B, Section 3.1.a states that radioactive gaseous wastes released to the environment from each of the five effluent release points shall be monitored and recorded. The monitors for releases from the refuel area exhaust and reactor building area exhaust have automatic isolation for the secondary containment prior to initiation of the SBGTS.
The total body dose rate limit is stated as being more restrictive than the skin dose rate and is used for determining the setpoints for the noble gas monitors.
The noble gas monitors maximum setpoints are based on the maximum concentration required to yield the total body dose rate limit.
The calculated setpoints account for simultaneous releases.
Technical Specification Appendix B, Section 3.2.a states that the dose rate limits apply to areas "at or beyond the site boundary."
Additionally, Figure 5.1-1 in the Technical Specifications shows the site 8
j i
l 1
boundary to include the JAFNPP as well as NMP-1 and NMP-2. Therefore, it i
appears that the contribution from NMP-1 and NMP-2 must be consid red when 1
demonstrating compliance to the dose rate limit of Technical Specification
)
Appendix B Section 3.2.a and not the dose rate from JAFNPP only.
(
I In Section 4.3.2.b.(2) the value for the elevated release rate should be 2.1E+6 pCi/sec instead of 2.4E+6 yCi/sec.
With the exception of the uncertainty of including contributions for NMP-1 and NMP-2 when demonstrating compliance to the dose rate limit of Technical Specification Appendix B Section 3.2.a. the methods for determining the setpoints for the noble gas effluent monitors are, in I
general, in agreement with the guidelines of NUREG-0133 and are considered acceptable.
Concentrations in Liouid Effluents Section 3.2.2 contains the methodology for determining the radioactivity concentrations in liquid effluents as required by Surveillance Technical Specification Appendix 8, Section 2.2.b.
In Section 3.2.2.a the word " percentage" should be removed from the definition for f.
Regardless, the methodology is within the guidelines 2
of NUREG-0133 and is considered acceptable.
Dose Rates in Gaseous Effluents Section 4.3.1 contains the methodology for determining dost rates from noble gases to the total body and skin as required by Surveillance
)
Technical Specification Appendix B Section 3.2.a.
1 In Section 4.3.1.b.(2), the definition for DRTB for Equation 4-5 indicates that this quantity is the dose rate from nuclide i, whereas it in fact is the dose rate to the total body from a].1, noble gas radionuclides.
9
d The definition of DRSKIN, for Equation 4-6, indicates that this quantity is the the dose rate from nuclide i, whereas it is in fact the dose rate to the skin from all noble gas radionuclides. Also, the quantity ORSKINj is not defined for Equation 4-6.
In Section 4.3.1.c.(7) in the equation for OR the release SKINj rate quantity, Qj is missing.
With the exception of the discrepancies indicated above, the methodology for determining the dose rate due to the release of radioactive noble gases is in agreement with the guidelines of.NUREG-0133 and is considered acceptable.
Section 4.3.3.b contains the methodology for determining the instantaneous dose rate due to the release of tritium, iodine-131,
)
iodine-133 and all radionuclides in particulate form with half lives greater than 8 days for the inhalation pathway as required by Surveillance Technical Specification Appendix B, Section 3.2.b.
l In Section 4.3.3 the text does not specifically indicate tritium as required in Technical Specification Appendix B, Section 3.2.b.
InSection4.3.3.b,inEquation4-9thequantityhjshouldbethe total release rate for nuclide i from all release points in order to determine a total dose rate. Also, in Equation 4-10, which determines the limiting iodine release rate, there should be a summation due to all radionuclides.
]
i With the exception of the indicated discrepancies, the methodology for determining the dose rates due to the release of tritium, iodine-131, iodine-133 and all radionuclides in particulate form with half lives greater than 8 days is in agreement with the guidelines of NUREG-0133 and is considered acceptable.
10
y I
Dose Due to Liouid Effluents I
l Section 3.4.1.b contains the methodology for determining the dose due to radioactive material released in liquid effluents as required by l
l Surveillance Technical Specification Appendix B, Section 2.3.a.
i' JAFNPP identifies fish and potable water pathways for the dose calculations, designating the adult as the maximum exposed individual and the total body and liver as the critical organs.
i In Section 3.4.1.b the quantity Atj in Equation 3-4 should be the number of hours in the calendar year instead of the number'of hours during liquid effluent releases for the calendar year. Also, a summation of the dose due to all radionuclides is missing from Equation 3-4.
In Section 3.4.1.b the reference for the derivation of the quantity Aj7 should be Appendix A, Table L-4.1 instead of Appendix A.
In Section 3.4.2.b.(1) the definition of Atj for Equation 3-5 should be the number of hours of the reporting period (e.g. monthly, quarterly, annually) and not the number of hours over which the release Also, the quantity (DF)j should be the total volume of dilution occurs.
released during the reporting period. A summation of the dose due to all radionuclides should be included in Equation 3-5.
The methodology in Section 3.4.2.c should be made consistent with the previous comments in this review concerning the definition of time period Atj identified in Sections 3.4.1.b and 3.4.2.b.(1).
Appendix A, Table L-1 lists the maximum permissible concentrations in water in unrestricted areas. The following discrepancies were detected by the reviewer:
j The first Zn-69 should be listed as Zn-65.
l 11
The value for insoluble Y-93 should be 3E-4 gCi/ml instead of 3E-5 gCi/ml.
Appendix A, Tables L-2 and L-3 contain liquid dose conversion factors, A
for all age groups, pathways and radionuclides. Using the j7 methodology given in Appendix A, Table L-4.1 the reviewer (with the exception of the boating pathway) was unable to reproduce a large fraction of the dose conversion factors.
It is recommended that the Licensee ensure the validity of each and every dose conversion factor.
Appendix A, Table L-4.1 gives a derivation of the liquid dose factor Aj7 The following errors were detected by the reviewer:
In the definition for Z in Table L-4.1, the quantity M is not c
defined.
The units for the quantity Aj should be sec-1 instead of secs.
A reference should be provided supporting the quantities Dw, and Df in Equation L-4.1.
In the calculational example for swimming, the equation referred to should be Equation L-4.1 instead of Equation L-4.4.
Also, for the quantity DFi the table referred to should be Table L-4.4 instead of Table L-4-2.
The purpose for Table L-4.2 is unclear since all the activity values are unity and the table is not referred to in the ODCM.
The following tables are not referred to in the ODCM:
Tables L-4.3, L-5, L-5.1, and L-5.2.
The ODCM states that the dose conversion factors tabulated in Table L-4.4 for swimming and boating are generated using the 12
Impell computer code.- Either the methodology and parameters contained in the Impe11 code should be submitted for review or the Licensee ~should include the reference containing NRC approval for use of this code.
Primarily due to the incorrect definition for At1 and the values in the dose factor tables, the methodology for calculating the dose due to the release of radioactivity in liquid effluents is not in agreement with the guidelines of NUREG-013J and is not considered acceptable.
Dose Due to Caseous Effluents Sections 4.4.1.b.(1), 4.4.2.a and 4.4.2.b contain the methodologies for calculating the cumulative gamma and beta air doses due to the release of radioactive noble gases as required by Surveillance Technical
' Specification Appendix B,.Section 3.3.a.
In Section 4.4.2.a.(2)(a), the reference for the detailed explanation and evaluation of 'the quantity M,ff should be Appendix E instead of Appendix C.
Also, the value for (X/Q) on Page 42 of the ODCM, for the 3
3 elevated release should be 3.8E-8 sec/m instead of 3.8E-7 sec/m.
In Section 4.4.2.b.(2) the second half of the first paragraph is evidently out of place since it discusses iodines and particulate in a section describing noble gas beta air dose methodology.
With the exception of these two discrepancies, the method for calculating.the air dose due to the release of radioactive noble gases is in agreement with the guidelines of NUREG-0133 and is considered acceptable.
i Sections 4.4.1.b.(2) and 4.4.2.c contain the methodologies for calculating the cumulative dose due to the release of I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than eight days as required by Surveillance Technical Specification 13 A
Appendix B, Section 3.4.a.
In Sections 4.4.1.b.(2)'and 4.4.2.c, tritiurn has been omitted from the radionuclides included-for determining cumulative dose but. is required by Technical Specification Appendix B, Section 3.4.b.
In Section 4.4.1.b.(2) the reference for the derivation of Rj7, should be to Appendix B, Table G-7.18 instead of Appendix B. Table G-8.
In Section 4.4.2.c.(2) the subscript D should be removed from the quantity (D/Q) in Equations 4-25 and 4-26.
Appendix B, Tables G'4, G-5, and G-6 contain gaseous dose conversion factors, Pj and Rj for all age groups, pathways, and radionuclides.
Using the methodology given in Appendix A, Table G-7 the reviewer was unable to reproduce a large fraction o'f the dose conversion l
factors.
It is recommended that the Licensee check the validity of each and every dose conversion factor.
Appendix B, Tables G-7 gives a derivation of gaseous dose factors, Pj and Rj. The following discrepancies were detected by the reviewer:
1 l
In Table G-7.1A tritium has been omitted from the radionuclides included for determining the inhalation dose factor but is l
required by the Technical Specifications.
)
In Table G-7.18 methodology for determining the gaseous dose I
factors due to tritium in the cow-milk, cow-meat, goat-milk and fruit-vegetable pathways is missing.
)
In Table G-7.18,Section I the quantity (DFAi) should be written (DFAi), to denote the age group.
In Table G-7.1B,Section III in the example calculation, the quantity (DFli)j should be written (DFli)a' 14
~
i In Table G-7.1B the Rj factor values determined in the example
{
calculations for Co-60 (child - total body) for all pathways
)
except inhalation are referenced as being listed in Appendix B, Table G-6 in the adult portion instead of the child portion of the table.
The purpose of Table G-7.2 is unclear since all the activity values are unity and the table is not referred to in the ODCM.
The purpose for some of the pages in Table G-7.3 is not c1t id also the table is not referred to in the ODCM.
The methodology for calculating the cumulative dose due to the release of I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than eight days is in agreement with the guidelines of Regulatory Guide 1.109, Revision I and NUREG-0133 and is considered acceptable.
However, since the reviewer was unable to verify the dose factors it is uncertain if the methodology can be used to demonstrate compliance to the Technical Specifications.
Dose Pro.iections l
l Section 3.5.2 of the ODCM describes the method used to project doses due to the expected releases of radioactive liquid effluents to determine when the liquid radwaste treatment system should be operated as required in Surveillance Technical Specification Appendix B, Section 2.4.a.
The methodology for determining the dose projection due to liquid radwaste effluents is in agreement with NUREG-0133 and is considered acceptable.
l Section 4.5.2 of the ODCM describes the method used to project doses due to release of radioactive gases when the offgas treatmenY system is not in use as required in Surveillance Technical Specification Appendix B, Section 3.6.a.
It does not appear that there are ventilation exhaust treatment systems and the gaseous radwaste treatment system is in use whenever the second steam driven feedwater pump is in operation.
i 15 l
Therefore, it is not clear why dose projections are required by the Technical Specifications since the purpose for the dose projections for a BWR is to determine required use of the ventilation exhaust treatment systems.
Section 4.5.2.a refers to the limits of Technical Specification Appendix B, Section 3.6.a whereas no limits are found in this Technical Specification.
Regardless, the methods for determining dose projections are within the guidelines of NUREG-0133 and are considered acceptable.
Total Dose Section 5.0 of the ODCM contains the methodology for calculating the total dose contributions including direct radiation as require'd by Surveillance Technical Specification Appendix B, Section 5.1.a.
In Section 5.1 in the first sentence of the first paragraph the text should read: " Dose evaluation to demonstrate compliance with the 40 CFR 190 dose limits need only be performed if the quarterly 2r annual doses... exceed twice the dose limits of Technical Specifications Appendix B, Sections 2.3.a, 3.3.a ju: 3.4.a respectively", in order to be consistent with the Technical Specifications.
With the exception of the indicated discrepancy, the methodology is in agreement with the requirements of the Technical Specifications and is I
considered acceptable.
Environmental Monitorina procram 1
Table H-1 in Appendix H of the ODCM contains specific parameters of distance and the direction sector from the site and additional information for each and every sample identified in Surveillance Technical Specification Section 6.1.
The environmental monitoring program is in compliance with Technical Specification Appendix B, Table 6.1-1 and is considered acceptable.
l 16
Sumary In summary, the Licensee's ODCM uses documented and approved methods that are, in general, consistent with the methodology and guidance in NUREG-0133 and Regulatory Guide 1.109, Revision 1.
However, due primarily to the incorrect definition of Atj and the erroneous values for the liquid and gaseous dose factors it is recommended that the NRC request another revision.
17
--__________-___-__-____-____-___________,_______--_-_-____-_____-_--__--__________________-___-_-__-_---A
4.
CONCLUSIONS The Licensee's ODCM updated through Revision 5, dated March 1987 for JAFNPP was reviewed.
It was determined that the ODCM uses methods that are, in general, consistent with the guidelines of NUREG-0133.
The methodology in most sections of the ODCM is acceptable for use in demonstrating compliance to the radiological effluent technical specifications. However, it is recommended that another revision to the ODCM be submitted to address the discrepancies identified in the review.
In Section 3.2.2.a the word " percentage" should be removed from the definition for f -2 In Section 3.3.3.b, the equation S - 0.5 x C/FL is given to determine the alarm / trip setpoint for the radwaste liquid effluent monitor.
However, when FL > 0.5, the value of the setpoint would be less than the expected count rate for the undiluted effluent at concentration C and the monitor would therefore alarm / trip continuously.
In Section 3.3.3.1.g a reference is made to Equation 3.3, however the equation identifier is missing for Equation 3.3 in Section 3.3.3.b.
In Section 3.3.4, a summation over the nuclide index "i" in Equation 3-3.c, is missing.
In Section 3.4.1.b the reference for the. derivation of the quantity Aj7 should be more specifically Appendix A, Table L-4.1 instead of Appendix A.
In Section 3.4.1.b the quantity Atj in Equation 3-4 should be the number of hours in the calendar year instead of the number of hours during liquid effluent releases for the calendar year.
18
r, 4
Also, a sumation of the dose due to all radionuclides is missing from Equation 3-4.
In Section 3.4.2.b.(1) the definition'of Atj for Equation 3-5 should be the number of hours of the reporting period (e.g. monthly, quarterly, annually) and not the number of hours over which the release occurs. Also, the quantity (DF))
should be'the total volume of dilution released during the reporting period. A summation of the dose due to all radionuclides should be included in Equation 3-5.
The methodology in Section 3.4.2.c should be made tos 'be consistent with the previous comments in this review concerning the definition of the time period At1 identified in Sections 3.4.1.b and 3.4.2.b.(1).
A simplified flow diagram of the solid radwaste treatment system is not included in the ODCM.
In Section 4.3.1.b.(2)
M e definition for DRTB for Equation 4-5 indicates knat this quantity is the dose rate from nuclide 1, whereas it in fact is the dose rate to the total body l
from 1].]. noble gas radionuclides.
The definition of DRSKIN, for Equation 4-6, indicates that this quantity is the the dose rate from nuclide i, whereas it is in fact the dose rate to the skin from 31]. noble gas radionuclides.
In Section 4.3.1.b.(2), the quantity DRSKIN is not defined j
for Equation 4-6.
In Section 4.3.1.c.(7), in the equation for DRSKIN, the j
parameterdismissing.
19
y Technical Specification Appendix B, Section 3.2.a states that the l
dose rate limits apply to areas "at or beyond the site.
boundary." Additionally, Figure 5.1-l'in the Technical i
Specifications shows the site boundary to include the JAFNPP as well as NMP-1 and NMP-2. Therefore,-it appears that the.
.j contribution from NMP-1 and NMP-2 must be considered when demonstrating compliance to the dose' rate limit of Technical.
Specification Appendix B Section 3.2.a and not the dose rate from JAFNPP only.
1 In Section 4,3.2.b.(2) the value for the elevated release rate should be 2.lC+6 pCi/sec instead of 2.4E+6 pCi/sec.
In Section'4.3.3 the text does not specifically indicate tritium as required in Technical Specification Appendix B, Section 3.2.a.b.
InSection4.3.3.b,inEquation4-9thequantity6jshouldbe l
the total release rate for nuclide i from all release points, in order to determine a total dose rate.
In Section 4.3.3.b in Equation 4-10, which determines the limiting iodine release rate, there should be a summation due to all radionuclides.
l In Sections 4.4.1.b.(2) and 4.4.2.c, tritium has been omitted
')
from the radionuclides included for determining cumulative dose.
but is required by Technical Specification Appendix B, Section 3.4.b.
I In Section 4.4.1.b.(2) the reference for the derivation of Rj7, should be to Appendix B, Table G-7.18 instead of Appendix 8. Table G-8.
20 LE.
9 In Section 4.4.2.a.(2)(a), the reference for the detailed explanation and evaluation of the quantity M,ff should be Appendix E instead of Appendix C.
In Section 4.4.1.a.(2) the value for (X/Q) on Page 42 of the 3
ODCM, for the elevated release should be 3.8E-8 sec/m instead 3
of 3.8E-7 sec/m.
In Section 4.4.2.b.(2) the second half of the first paragraph seems to be displaced since it discusses iodines and particulate in a section describing noble gas beta air dose methodology.
In Section 4.4.2.c.(2) the subscript D should be removed from the quantity (D/Q) in Equations 4-25 and 4-26.
Section 4.5.2.a refers to the limits of Technical Specification Appendix B, Section 3.6.a however, no limits are found in this Technical Specification.
In Section 5.1 in the first sentence of the first paragraph the text should read: " Dose evaluation to demonstrate compliance with the 40 CFR 190 dose limits need only be performed if the quarterly at annual doses... exceed twice the dose limits of Technical Specifications Appendix B, Sections 2.3.a, 3.3.a 2r 3.4.a respectively", in order to be consistent with the Technical Specifications.
Appendix A, Table L-1 lists the maximum permissible concentrations in water in unrestricted areas.
The following discrepancies were detected by the reviewer:
The first Zn-69 should be listed as In-65.
The value for insoluble Y-93 should be 3E-4 yCi/ml instead of 3E-5 pCi/ml.
21
.mma-__-
o Appendix A, Tables L-2 and L-3 contain liquid dose conversion factors, Aj7 for all age groups, pathways and radionuclides. Using the methodology given in Appendix A, Table L-4.1 the reviewer (with the exception of the boating pathway) was unable to reproduce a large fraction of the dose conversion factors.
It is recommended that the Licensee ensure the validity of each and every dose conversion factor.
Appendix A, Table L-4.1 gives a derivation of the liquid dose factor Ajf.
The following errors were detected by the reviewer:
In the definition for Z in Table L-4.1, the quantity M is c
not defined.
The units for the quantity Aj should be sec-1 instead of secs.
A reference should be provided for the quantities Dw, and Of in Equation L-4.1.
In the calculational example for swimming, the equation referred to should be Equation L-4.1 instead of Equation L-4.4.
Also, for the quantity DFi the table referred to should be Table L-4.4 instead of Table L-4-2.
The purpose for Table L-4.2 is unclear since all the activity values are unity and the table is not referred to on the ODCM.
The following tables are not referred to in the ODCM:
Tables L-4.3, L-5, L-5.1, and L-5.2.
The ODCM states that the dose conversion factors tabulated l
in Table L-4.4 for swimming and boating are generated using 22
4 l
the Impe11 computer code.
Either the methodology and parameters contained in the Impell code should be submitted for review or the Licensee should include the reference containing NRC approval for use of this code.
Appendix B, Tables G-4, G-5, ar.d G-6 contain gaseous dose conversion factors, Pj and Rj for all age groups, pathways, and radionuclides. Using the methodology given in Appendix A, Table G-7 the reviewer was unable to reproduce a large fraction of the dose conversion factors. it is recommended that the Licensee check the validity of e,ach and every dose conversion factor.
Appendix B, Tables G-7 gives a derivation of gaseous dose factors, Pj and R. The following discrepancies were j
detected by the reviewer:
In Table G-7.1A tritium has been omitted from the radionuclides included for determining the inhalation dose factor but is required by the Technical Specifications.
In Table G-7.1B methodology for determining the gaseous dose factors due to tritium in the cow-milk, cow-meat, goat-milk and fruit-vegetable pathways is missing.
In Table G-7.18,Section I the quantity (DFAi) should be written (DFA1)a to denote the age group.
In Table G-7.1B,Section III in the example calculation, the quantity (DFL1)j should be written (DFli)a-I In Table G-7.1B the R factor values determined in the example calculations for Co-60 (child - total body) for all pathways except inhalation are referenced as being listed in Appendix B, Table G-6 in the adult portion instead of the 23
)
4 child portion of the table.
The purpose of Table G-7.2 is unclear since all the activity values are unity and the table is not referred to in the ODCM.
The purpose for some of the pages in Table G-7.3-is not clear and also the table is not referred to in the ODCM.
The Technical Specifications and Appendix F, Table F-1 identify the five gaseous effluent release points listed above.
- However, Figures F-1, F-2 and F-4 identify only four gaseous r'elease points.
Figure F-4 also shows a simplified diagram of the gaseous radwaste treatment system. Another diagram is needed however, showing a diagram of the components and pathways for the Standby Gas Treatment System (SBGTS).
In Figure F-5 of the ODCM the diagram showing the liquid effluent release paths does not show the point where the liquid radwaste effluent line joins the circulating water and the service water lines for dilution prior to discharge.
24
5.
REFERENCES 1.
Title 10, Code of Federal Regulations, Part 50, Appendix I, " Numerical
)
Guides for Design Objectives and Limiting Conditions for Operation to j
Meet the Criterion, 'As Low As Is Reasonably Achievable,' for I
Radioactive Material in Light-Water-Cooled Nuclear Power Reactor l
Effl uents. "
2.
Letter from J. P. Bayne (New York Power Authority) to D. B. Vassallo (NRC),
Subject:
James A. FitzPatrick Nuclear Power Plant, Off-Site Dose Calculation Manual, October 26, 1983.
I 3.
Letter from H. I. Abelson (NRC) to J. P. Eayne (New York Power
)'
Authority),
Subject:
Approval of James A. FitzPatrick Nuclear Power P1 ant ODCM, May 29, 1985.
4.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. FitzPatrick Nuclear Power Plant, Semiannual Radioactive Effluent Release Reoort, February 28, 1986.
5.
Letter from H. I. Abelson (NRC) to J. C. Brons (New York Power Authority),
Subject:
Revision to the Offsite Dose calculation Manual (ODCM), October 2, 1986.
6.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. FitzPatrick Nuclear Power Plant, Semiannual Radioactive Effluent Release Reoort, August 29, 1936.
7.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. Fiittrick Nuclear Power Plant, Semiannual Radioactive Effluent Relech2qtt, February 27, 1987.
8.
Letter from R. J. Converse (New York Power Authority) to T. E. Murley (NRC),
Subject:
J. A. FitzPatrick Nuclear Power Plant, Docket No.
50-333, License No. DPR-59, Revision to the Off-Site Dose Calculation l
Manual (00CM), April 2,1987.
25
M..
4-
'9.
" Standard Radiological Effluent Technical Specifications _for Boiling-Water Reactors," Rev. 3, Draft'7", intended for contractor guidance in reviewing RETS proposals for operating reactors, NUREG-0473, September
'1982.
- 10. " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, A Guidance Manual for Users of Standard Technical Specifications," _NUREG-0133, October 1978.
- 11. " General Contents of the Offsite Dose Calculation Manual," Revision 1, Branch Technical Position, Radiological Assessment Branch, NRC, February 8,1979.
- 12. " Calculation of Annual Doses of Evaluating Compliance with 10 CFR 50, Appendix I," Regulatory Guide 1.109, Rev.1, October 1977.
i 26
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eleuoGRAPHic DATA SHEET EGG-PHY-8073
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- "Ye'cIElai Evaluation Report for the Evaluation of 0DCM Updated Through Revision 5, James A Fitzpatrick Nuclear Power Plant
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.1988
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. oar....o.,,uw.o T. S. Bohn, W. Serrano, A. L. Freeman, M. R. Winberg i
July 1988 7 p.asonesa=0 caGamilar,om maas. Asso maste=0 ADom. uaeme to Co.es
. ##oa.CTir4.E'wro#4 wm,r aiwas..m Nuc1 ear Sciences EG&G Idaho, Inc.
P. O. Box-1625 Idaho Falls, ID 83415 06934 80.Poss.caimo omeamigariosi esaw. Amo maissNQ acom.m s,arsve te casw i t.. rWP. os m.Pom, U. S. Nuclear Regulatory Commission NRR/DRPEP/RPB
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Washington, DC 20555 u.s.
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- 3. a tmeCT (JW seen er ' anal The Offsite Dose Calculation Manual for the James A. FitzPatrick Nuclear Power Plant contains a description of the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints, and in the :onduct of the environmental radiological monitoring program.
The NRC transmitted the ODCM updated through Revision 5 to the Idaho National Engineering Laboratory for review.
The ODCM, complete through Revision 5, was reviewed in its entirety by EG&G Idaho at the INEL and the results of the review are resented in this report.
It was determined that the ODCM updated through Revision 5 uses methods that are, in general, in agreement with the guidelines of NUREG-0133. However, it is recommended that another revision to the ODCM be submitted to address the discrepancies identified in the review.
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,. neu.,rv euni.. care e tae Unc assified 2o. =ti.....,o t...
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