ML20207P113

Technical Evaluation Rept for Evaluation of Offsite Dose Calculation Manual,Updated Through Rev 20

ML20207P113
Person / Time
Site:	Catawba
Issue date:	08/31/1988
From:	Amaro C, Serrano W, Thomas Young EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20206F549	List: ML20206F546 ML20207P113
References
CON-FIN-D-6034 EGG-PHY-8214, NUDOCS 8810200126
Download: ML20207P113 (33)

Text

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.- . E!! CLOSURE

' EGG-PHY-8c;4 TECHNICAL EVALUATION REPORT for the EVALUATION OF 00CM (UPDATED THROUGH REVISION 20)

CATAWBA NUCLEAR STATION, UNITS 1 AND 2 NRC Docket No. 50-413 NRC License No. NPF-35 NRC Docket No. 50 414 NRC License No. NPF 52 i;

T. E. Young W. Serrano C. R. Amaro Published August 1988 1

Idaho National Engineering Laboratory EG1G Idaho, Inc.

Idaho Falls, Idaho 83415 I

Prepared for the

' U. S. Nuclear Regulatory Comission i-Washington. 0.C. 20555 Under DOE Contract no. DE.AC07 76ID01570 FIN Nc. 06034

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ABSTRACT The Offsite Dose Calculatio, Manual (ODCM) for the Catawba Nuclear Station (CNS) contains current methodology and parameters for calculation of offsite doses due to radioactive liquid and gaseous effluents, determination of gaseous and liquid effluent moni+. ors' alarm / trip setpoints, and for conduct of the environmental radiological monitoring program. The complete ODCM consists of a generic section that is applicable to all Duke Power Company (DPCo) nuclear stPtions and Appendix C, that applies only to CNS. The generic ODCM and Appendix C .

were reviewed for the NRC by Franklin Research Canter (FRC). The comments by FRC were transmitted to CNS by the NRC, and CNS responded by letter and subsequently submitted ODCM Revision 4 to the NRC. Revision 4 was approved, in general, by the NRC prior to commercial operation of CNS.

Changes to the approved version of the ODCM are required to be s';bmitted 1 to the NRC in the Se.niainual Radioactive Effluent Release Report for the l period in which the cha' ige (s) was made effective. The NRC transmitted the complete ODCM for CNS, consisting of the generic section and Appendix C with all revisions (through Revision 20) made since the initial approval, to the Idaho National Engineering Laboratory (INEL) for review. The ODCM, complete with all revisions, was reviewed by EG&G at the INEL and results i of the review are presented 'n this report. It was determined that the ODCM, updated through Revision 20, uses methods that are, in general, consistent with the guidelines of NUREG 0133. However, it is recommended

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that another revision by submitted by the Licensee to address  ;

the discrepancies identified in the review.

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FOREWORD This report is submitted as partial fulfillment of the "Review of Radiological Issues" project being conducted by the Idaho National Engineering Laboratory for the the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation. The U. S. Nuclear Regulatory Cummission 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 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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CONTENTS Page Abstract. ............................. i Foreword. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

1. Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . 1

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

3. Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . . 2

4. Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . 22

5. References. . . . . . . . . . . . . . . . . . . . . . . . . . . 28 FIGURES

1. Liquid radwaste effluent pathways for Catawba Nuclear Station, Units 1 and 2 . . . . . .. . . . . . . . . . . . . . . . . . . . 4

2. Gaseous radwaste system and effluent pathways for Catawba Nuclear Station, Units 1 and 2 . . . . . . . . . . . . . . . . . . . . . 8

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1. INTRODUCTION 1.1 Puroose of Review This document reports the review and evaluation of the most recent version of the Offsite Dose Calculation Manual (ODCM) submitted by the Duke Power Company (OPCo), the Licensee for Catawba Nuc.Nar Station (CNS) Units 1 and 2 (Westinghouse PWRs). The ODCM is a supplementary document for implementing the Radiological Effluent Technical Specifications (RETS) in compliance with 10 CFR 50, Appendix ! requirementsill.

1.2 plant Soecific Backaround The ODCM for nuclear power plants operated by DPCo consists of a generic section applicable to all DPCo plants and appendices applicable only to the individual plants. Appendix C is applicable to CNS. Revisions to the ODCM are numbered sequentially, without regard to whether they change the generic section or one of the plant specific appendices. Following initial review of a proposed Appendix C, the DPCo submitted Appendix C (for CNS) as Revision 4 to the 00CM with a letter dated August 16,1984(2). The NRC reviewed Revision 4 and provided notification of approval, in general, of the ODCH ;o DPCo in a letter dated October 19,1984[3). Subsequent revisions were submitted to the NRC.

Revision SI43 applies to the generic section and Revisions 6(5), 8(5),

12(6),13 I73,19I83, and 20I93 apply to CNS. The NRC transmitted the Licensee's complete ODCM, updated through Revision 20, to an independent review team at the Idaho National Engineering Laboratory (INEL) for review. The complete 00CM was reviewed by EG&G Idaho at the INEL. The results and conclusions of the review are presented in this report.

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

NUREG0472,RETSforPWRs(10)

NUREG 0133 Preparation of RETS for Nuclear Power Plantsilll 1

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• 4 The following NRC guidelines were also used in the 00CM review:

"General Contents of the Offsite Oose Calculation Manual,"

Revision 1[12), and Regulatory Guide 1.109, Revision 1(13),

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As specified in NUREG-0472, the 00CM is to be developed by the Licensee to document the methodology and approaches used to calculate offsite doses and maintain the operability c~ the radioactive effluent  ;

systems. As a minimum, the 00CM 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. I In addition, the ODCM should contain flow dagrams, 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 00CM.

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3. EVALVATION The OOCH for the Cstawb4 Nuclear Station consists of a generic section l applicable to all DPCo nuclear stations (Oconee, McGuire, and Catawba) and l Appendix C, applicable only to CNS. The Technical Specifications for Unit I and Unit 2 are a single document.

i Several statements in Appendix C of the 00CM are written so as to l permit, but not require, calculations to be done in a certain way. Part of these should be written as requirements.

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s 3.1 Licuid Effluent Pathways The CNS is located adjacent to Lake Wylie in the north-central portion of South Carolina. It is on a peninsula bounded by Beaver Dam Creek to the north and Big Allisor. Creek to the south.

Main conden: ar cooling for the CNS is provided by water circulated through mechanical draft cooling towers. During normal operation, cooling water for the Nuclear Service Water and Low Pressure Service Water Systems is pumped from the Beaver Dam Creek arm of Lake Wylie and returned to Big Allison Creek. An onsite pond (Standby Nuclear Service Water Pond) is available for emergency use.

The liquid radwaste system at CNS is shared by the two units. A simplified diagram of the liquid radwaste effluent pathways is shown in Figure 1.

Technical Specification 3.3.3.10 requires the following radioactivity monitors providing alarm and automatic termination of release:

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1. Waste Liquid Discharge Monitor (EMF 49)

2. Turbine Building Sump Monitor (EMF 31)

3. Steam Generator Water Sample Monitor (EMF-34) (This does not i monitor releases directly offsite.)

Technical Specification 3.3.3.10 also requires a continuous composite sampler and sampler flow monitor on the Conventional Waste Water Treatment Line, which is the usual release pathway for liquids from the Turbine Building Sumps.

Revision 20 of the 00CM describes a release pathway through the Auxiliary Monitor Tanks, for which there are no monitoring requirements in the technical specifications dated May 1986. Revision 20 shows monitors and an isolation (shut off) valve on the effluent line from these tanks that are apparently identical to the monitors and shut off valve on the l effluent line from the Waste Monitor Tanks. Sections C2.1.1 and C3.1.1

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Letdown System, laundry and Containment Steam Reactor Coolant Hot Shower Ventilation Unit Generator l Drain Tank, Tank Condensate Tank Blowdown Waste Drain Tank, "

Steam Generator ' b EMF-44 Orain Tank, IH-EMr 34 Waste Evaporator Floor  :

feed Tank. Orain -

Tank Auxiliary Feed- Conden-water Sump Pumps sate and Floor Orain System i

V Sump Pumps -

Shared Recycle and Radwaste Systems: Boron Recycle and Liquid Waste Filters, Evaporators, O EMF 52 Demineralizers, and Tanks; and + -

Radwaste Tanks. Filters, and p y Demineralizer.

Turbine I

-* Building Sumps Recycle -m Waste -

Monitor Monitor IH-EMF-31 Tanks Tanks

, Conventional Waste

Water Treatment ,

8-Flow Rate Monitor System i IHEMF 49 ->

Sampling System:

IWL124 X4- Low Flow Interlock I

r Lake Wylie Y

Nuclear Service Water System Turbine Building Sumps,

'taam Generator Orain Tanks,.

, GasteEvaporatorFeedTank,j low Pressure Service Laundry and Hot Shower Tank.'

Water System (includes Flow and Floor Orain Tank  ;

Cooling Tower Blowdown) #- Rate j Monitor Auxiliary Tank Building C Radioactivity Monitor Process Area i 6 - Radioactivity Monitor ,

(with automatic y termination of release) < X-O-#- Auxiliary Monitor Tanks X - Shut off Valve IWLX28

1. - Flow Rate Monitor y Flow Rate Monitor Lake Wylie EMF 57 Fir r 1. Liquid radwaste effluent pathways for Catawba Nuclear Station, Units 1 and 2. (Figure drawn from information in Appendix C of the ODCM for Duke Nuclear Power Stations, updated through Revision 20.)

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s of the 00CM indicate that effluent lines frc.n both the Waste Monitor Tanks and the Auxiliary Monitor Tanks are components of the Waste Liquid Effluent Line. Both lines cumbine with the Low Pressure Service Water Line before discharge to the unrestricted area.

3.2 Liouid Effluent Monitor Setooints Sections 2.1 and C3.1 of the 00CM contain the methodology to determine the setpoints for the liquid radwaste effluent monitors, as required by Technical Specification 3.3.3.10. Technical Specification 3.3.3.10 requires that the setpoints for the liquid effluent monitors be set to ensure that radioactive materials released as effluents shall not result in concentrations in unrestricted areas in excess of the values specified in 10 CFR 20, Appendix B, Table II, Column 2(I43 The main pathway for the release of radioactive liquid effluents is the Waste Liquid Effluent Line, consisting of lines from the Waste Monitor Tanks and the Auxiliary Monitor Tanks, by which effluents are released to the Low Pressure Service Water Line for discharge offsite. The second sentence in Section C3.1.1 requires that the setpoints of the radioactivity monitors shall be set to terminate the release if the effluent activity should exceed that determined by laboratory analyses and used to calculate the release

, rate in Section C2.1.1. Therefore, the example in the next sentence of how the setpoints m_n be calculated is unnecessary. Additionally, the value of

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1.0E 07 gCi/ml for MPC in the example should be 3.0E 08 #Ci/ml for an unidentified mixture for compliance with 10 CFR 20.

Also, to prevent spurious alarms, it is reasonable to use a setpoint corresponding to slightly greater than the measured concentration and to reduce the allowed flow of undiluted radwaste calculated in Section C2.1.1 by a corresponding fraction.

Except for the use of 1.0E 07 pCi/mi instead of 3.0E 08 pCi/ml for the MFC of an unidentified mixture, the methodology in Section C3.1.1 to determine the setpoints of the waste liquid discharge monitors is, in general, within the guidelines of NUREG 0133. However, since the setpoint 5

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is to be set at the concentration measured, no calculational methodology is required. Thus, the licensee could either delete the calculational example provided or. modify the setpoint requirements.

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In addition to the monitor on the Waste Liquid Effluent Line, Technical Specification 3.3.3.10 requires alarm / trip radioactivity monitors on release lines from the Turbine Building Sumps and the Steam Generator Blowdown.

(Sections C3.1.2 and C3.1.3 of the 00CM discuss additional monitors on the Containment Ventilation' Unit Effluent Line and on the Auxiliary Feedwater Sump Pumps and floor Orain Sump Pump.) The Turbine Building Sumps are in the final release pathways for all of the monitored liquid streams except the Waste Liquid Effluent Line, unless the monitors indicate radioactivity.

If no radioactivity is detected, effluents from the Turbine Building Sumps Discharge lines are normally released to Lake Wylie via the Conventional Waste Water Treatment System Effluent Line. The monitors' setpoints are initially set at 1.0E 06 pCi/ml (the monitors' minimum practical setpoints) plus background. If a monitor setpoint is exceeded, the release through the monitored line is terminated or routed to another system, after which the routing or treatment is administratively determined. The rerouting and tr a tment choices include continued release through the Conventional Wasce Water Treatment System Effluent Line, discharge through the Waste liquid Effluent Line, and treatment by the liquid radwaste treatment system followed by release through the Waste Liquid Effluent Line.

Section C3.1.5 contains an example of how a setpoint mm be calcalated for a turbine building sump monitor to ensure that the concentration released is not greater than 1.0E 07 pCi/ml for an unidentified nixture. For compliance with 10 CFR 20, the limits on the cencentr, tion of in unidentified mixture released should be 3.0E 08 #Ci/ml. In thit section it is assumed that 10?. of the minimum flow in the Low Pressure Service Water (RL) Line will be used to dilute the discharge from the Turoine Building Sumps via the Conventional Waste Water Treatment System Effluent (WC) Line. The 00CM states, "This flow rate will allow the WC system to discharge 10?. of the stations MPC and dose limits." This methodology is considered acceptable if the Conventional Waste Water Treatment System Effluent (WC) Line and the Low Pressure Service Water Line 6

combine before discharr,ing to the unrestricted area, but does not ensure compliance with the concentration limits of 10 CFR 20 if the two lines are-released to the unre'.tricted area at different locations, as appears to be indicated by both being identified as release points'in 00CM Lection C2.1 Except for the 1eed to limit the concentration of an unidentified mixture to 3.0E-08 pCi/ml, the methodology in Section C3.1.5 to determine the setpoints of the Turbine Building Sump Discharge Line monitors is, in general, within the guidelines of NUREG-0133 if the Conventional Waste Water Treatment System Effluent Line and the low Pressure Service Water Line discharge to the unrestricted area at the same point, 3.3 Gaseous Effluent Pathways Technical Specification 3.0.3.11 requires noble gas radioactivity monitors on the following gaseous effluent pathways:

Waste Gas Holdup System, EMF 50 (1 per station)

Condenser Evacuation System, EMF-33 (each unit)

Vent System, EMF 36 (each unit)

Containment Purge System EMF 39 (each unit)

Containment Air Release and Addition System, EMF 39 (each unit)

Figure C1.0.2 of the OOCH shows the gaseous radwaste systems, radioactivity monitors, and release pathways. This figure is reproduced in this report as Figure 2. Most significant gaseous effluents are released through the unit vents and monitored by the unit vent system monitors.

Revision 20 of the OOCH describes a release pathway 'or which monitoring is not required by the technical specifications dated May 1986. This pathway is the exhaust from the Auxiliary Monitor Tank Building (See Figure 2).

Turbine Building ventilation air is discharged from unmonitored roof vents.

3.4 Gaseous Effluent Monitor Setcoints Section 2.2 and C3.2 of the 00CM contains the e a dology used to determine the alarm setpoints for the noble gas radiation monitors, as 7

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• a a t.C yb a t t s.L 't a C C=4acca,assCaste Figure 2. Gaseous radwaste system and effluent pathways for Catawba Nuclear Station Units 1 and 2. (Figure taken from Appendix C of M fsite Dose e

Calculation Manual for Duke Nuclear Power Stations, updated through Revision 20.)

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e AUXILI ARY MONITOR TANK BUILDING PARTICULATE at NOOLE CAS

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UB HOODS Figure 2. Caseous radwaste system and effluent pathways for Catawba Nuclear Station, Units 1 and 2. (Figure taken from Appendix C of Offsite Dose Calculation Manual for Duke Nuclear Power Stations, updated through Revision 20.) (Cont.)

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al Specification 3.3.3.11. The methodology depends partially on definitions and discussions in Section C2.2. Technical Specification 3.3.3.11 requires that the setpoints for the gaseous effluent monitors be established to ensure that the limits of Specification 3.11.2.1 are not exceeded Specification 3.11.2.1 limits the dose rates due to noble i' gas effluents to less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.

Section C3.2, which is the ;ection of the 00CM included specifically to describe the method for determining monitor setpoints, should include instructions for determining the setpoint for the gaseous effluent monitor en the Auxiliary Monitor Tank Building. The section requires that the sum of the dose rates permitted by the monitors at the final release points be

less than or equal to 500 mrem /yr to the total body by reference to the j definition of Qj in Section C2.2.2. It would be prudent to require the j setpoints to be below the release limits, so alarm is given prior to

) exceeding the technical specification limits.

Except for omission of instructions for determining the setpoint of the j effluent monitor on the Auxiliary Monitor Tank Building, the methodology of

! Section C3.2 for determining the radiation monitor setpoints for gaseous l effluents is, in general, within the guidelines of NUREG 0133, and is

! considered acceptable, i

3.5 (pncentrations in Liauid Effluents Sections 1.1 and C2.1 of the 00CM contain the methodology for determining that the radionuclide concentrations in the released liquid effluents are maintained within the limits of Technical Specification 3.11.1.1, as required by Technical Specification 4.11.1.1.2. Technical Specification 3.11.1.1 requires compliance to the concentration limits of 10 CFR 20, Appendix B, Table !!, Column 2 for radionuclides other than dissolved and entrained noble gases. For dissolved and entrained noble gases, the concentration is required to be limited to 2 X 10*4 pCi/ml total activity.

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Since releases may be made simultaneously from the Waste Monitor Tanks and the Auxiliary Monitor Tanks, as indicated in Section C3.1.1, the equation in Section C2.1.1 that accounts for simultaneous releases from the WL and WC lines should be exp3nded to account for simultaneous releases from these two sets of monitor tanks into the WL line. With this exception, the methodology of Section C2.1.1 to ensure that the concentration of radionuclides released to an unrestricted area due to releases by way of the Liquid Waste Effluent Discharge Line is, in general, within the guidelines of NUREG 0133 and is considered acceptable.

Section C2.1.2 contains the methodology to ensure that the radionuclide concentration limits of Technical Specification 3.11.1.1 will not be exceeded due to radioactive materials released via the Conventional Waste Water Treatment System Effluent (WC) l.ine. A discussion of the handling of station effluents that are normally discharged by way of the WC line requires use of information in Section C3.1 (on monitor setpoints) as well as Section C2.1.2. The following four systems (lines) that normally feed directly or indirectly to the WC line have monitors initially set at 1.0E 06 4Cl/ml (the monitors' minimum practical setpoint) plus background:

1. Containment Ventilation Unit Condensate Effluent Line,

2. Auxiliary Feedwater Sump Pumps and Fleer Drain Sump Pump,

3. Steam Generator Blowdown Line, and 4 Turbine Building Sump Discharge Line.

Each of these systems (lines) is discharged automatically, but the discharge is terminated and other procedures followed for releasing the liquids if a monitor setpoint is exceeded. These other procedures include primarily routing to other systems, but in the case of the Turbine Building Sumps Discharge Line the liquids can also be "routed directly to the liquid waste effluent discharge line, or allowed to continue being discharged via the circuit with proper administrative controls to assure that release limits are not exceeded." An example of these controls restricting the radionuclide concentration in the released liquid to 1.0E 07 uCi/ml is 11

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given in Section C3.1.5 of the 00CM. This limit should be 3.0E-08 uC1/ml. A procedure to be used for ensuring that the  :

radionuclide concentration in liquids discharged to unrestricted areas does not exceed the technical specification limits when releases from the Turbine

. Building Sumps Discharge Line are released directly to the Liquid Waste Effluent Discharge Line should be added to the ODCM.  ;

L The acceptability of the methodology in Section C2.1.2 (as expanded by information in Section C3.1) depends on the discharge point of the [

j Conventional Waste Water Treatment System Effluent Line. Section C2.1 states that the Liquid Waste Effluent Discharge Line (WL) and the Conventional Waste l Water Treatment System Effluent Line (WC) are two potential release points. {

j The WL release is via the Low Pressure Service Water System (LPSW) Effluent  ;

Line. Ten percent of the LPSW flow should not be used to calculate diluted 1 concentrations for releases via the WC line if the WC and LPSW lines  ;

f j discharge to the unrestricted area of Lake Wylie at different points, because j the liquids are not combined before release to the unrestricted area. If the

! releases are at different points in the unrestricted area, the monitor set at 1.0E 06 pCl/ml on the Turbine Building Sumps effluents would also appear j to allow radionuclide concentrations in excess of the 10 CFR 20 limits to be l released to unrestricted areas, because of the low flow in the WC line.

I i Summarizing the discussion of releases via the Conventional Waste Water

{ Treatment System Line, if this line discharges to the unrestricted area  ;

l separately from the Low Pressure Service Water Line the methodology does nut ,

ensure that the concentration limits of 10 CFR 20 are not exceeded. If these f two lines combine prior to release to the unrestricted area or at the point l

of release, and the limit on the concentration of an unidentified mixture is l ,

! changed to 3.0E 08 4C1/ml, the methodology of Sections C2.1.2 and C3.1.5  ;

is, in general, within the guidelines of NUREG 0133 and is considered i

acceptable. I f

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! 3.6 Dese Rates due to Gaseous Effluents t

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I i f Sections 1.0, 1.2, 2.2. C2.2, and C2.2.1 of the 00CM contain the methodology for determining that the dose rates due to the release of  !

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radioactive noble gases are within the limits of Technical Specification 3.11.2.1.a as required by Specification 4.11.E.1.1. Technical Specification 3.11.2.1.a requires that the dose rates due to noble gases at or beyond the site boundary be limited to less than or equal to 500 mrem /yr to the whole body and less than or equal to 3000 mrem /yr to the skin.

Section C2.2 should specifically require that gaseous effluents from the Auxiliary Monitor Tank Building be considered when determining offsite dose rates due to gaseous effluents. As stated previously in Section 3.4, it is recommended that setpoints be set to give an alarm prior to release rates being exceeded. However, the methodology to determine that the dose rates due to the release of radioactive noble gases are within the limits of Technical Specification 3.ll.2.1.a is, in general, within the guidelines of NUREG 0133 and is considered acceptable.

Sections 1.0, 1.2. C2.2, and C2.2.2 of the ODCM contain the methodology l

for determining that the dose rate due to release of radioactive gaseous effluents (other than noble gases) are within the limits of Technical l

Specification 3.11.2.1.b. as required by Specification 4.11.2.1.2.

Technical Specification 3.11.2.1.b requires that the dose rate due to I-131,1-133, H 3, and all radionuclides in particulate form with half lives greater than 8 days be limited to less than or equal to 1500 mrem /yr to any organ. Releases from the Auxiliary Monitor Tank Building should be included in the calculations of the offsite dose rate due to releases from CNS.

The identification of radionuclides released in gaseous effluents to be considered when determining that the organ dose rate limit is not exceeded differ slightly in Sections 1.2 and C.2.2. However, if the identifications are interpreted using the equations in Sections 1.2.2 and C2.2.2 and the data in Table 1.2.2, both radionuclide identifications are considerably more restrictive than required by Technical Specification 3.ll.2.1.b. Also, the Licensee may wish to change the requirements to match the recomendation in the bases statement for Technical Specification 3 ll 2.1.b for CNS and in NUREG 0472; i.e., that the organ dose limit may be applied to the thyroid of a child via the inhalation pathway.

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e The source of the data and the equations used to calculate the values in Table 1.2.2 should be specified in the 00CM. The dose parameters, I Licensee's(!), in Table 1.2.2 were checked using data from Regulatory Guide l.109, Rev. 1. The reviewers' values for the food and ground plane dose  ;

parameters ranged from the same as the Licensee's values to 507, higher. [

However, even if the values are in error or calculated by a different  ;

method, the dose rate calculations of the 00CM are still acceptable, since the inclusion of the food and ground plane pathways makes the Licensee's calculations more conservative than required. It is not clear how the l values for Cd ll5m, Sn 123, Sn 126, Sb 124, and Sb-125 were obtained.  !

! r l Section C2.2 or C2.f. 2 of the 00CM should contain commitments to perform f the dose rate calculations consistent with the frequency of analyses i i required by Technical Specification Table 4.11-2, as required by Technical

Specification 4.11.2.1.2. With this exception, the methodology in Sections l j 1.0,1.2, and C2.2 of the 00CM for determining the dose rate for comparison  !

with the limits of Technical Specification 3.11.2.1.b is, in general, within l '

i the guidelines of NUREG 0133 and is considered acceptable.

1  !

4 3.7 Dose Oue to Licuid Effluents  :

i l Sections 3.1.1 and C4.1 of the 00CM contain the methodology to determine j j the dose or dose commitment to the maximum exposed member of the public due (

to radioactive materials in liquid effluents to demonstrate compliance with I the dose limits of Technical Specification 3.11.1.2, as required by Technical Specification 4.11.1.2. However, the introduction to the generic l

! section of the 00CM states that normally the LADTAP code will be used to f I

calculate offsite doses, with the methods given in the OOCH used when t.ADTAP <

I I is not available. Therefore, for compliance with the recommendations of j NUREG 0133, all of the input parameters of LADTAP should be included in the  :

i 00CH.  !

i i l In Section C4.3.1, the location of the potable water intake should be l l specified, and the value 37.7 for Dy should be referenced or justified.  !

! [

i

! t 14

- ._ - _ - . - _ i

Section C4.1 of the 00CM states that the 31 day, quarterly, semiannual, and annual dose contributions shall be calculated using the methodology in the generic information section (Section 3.1.1). The purpose of Section C4.2.1 is not clear. Section C4.c.1 states that "...the maximum exposed individual would be an adult who consumed fish caught in the discharge canal and who drank water from the nearest ' downstream' potable j water intake.' As written, Section C4.2.1 appears to contain no requirements, but information only. If it is intended that only the adult

age group be considered for the calculations required by Section C4.1, this fact should be stated explicitly in Section C4.2.1, otherwise doses to all

, four age groups should be calculated as required by 00CM Section 3.1.1. If all age groups are used, then the tables of site related ingestion dose comcitment factors (Aair) should be added for teenager, child, and j infant age groups. If the information in Section C4.2.1 applies to only the dose estimate calculations described in Section C4.3.1, there is no conflict l

between the gener'c and plant specific sections of the 00CM.

Since the limits on offsite doses are on a per unit basis, the problem of assigning releases of radioactive materials in liquid effluents to the two units should be addressed, especially for cases of disproportionate releases from the two units.

Althuugh the use of the average dilution flow (F) during the liquid effluent release is acceptable for the calculations of doses due to liquid effluents in Section 3.1.1, the Licensee may wish to take advantage of the 1000 cfs per reactor permitted by Section 4.3 of NUREG 0133 for reactors with closed cycle cooling systems. Using this value would require increasing the value of the recirculation factor (a) and probably reducing the value of Og for drinking water, but should result in a lower and more realistic calculated dose than the present calculational method.

As another alternative, the NRC Staff allows the dose for a reporting period (calendar quarter or calendar year) to be calculated using the average dilution flow for the reporting period. If this method is chosen for the dose calculations, the total dilution flow for the reporting period should be included in the semiannual report.

15

The Licensee should correct the bioaccumulation factor for Na in fish in Table 3.1-1 to 1.0E+02 and add the bioaccumulation factor for P (for which the "best values" are now 3000 for fish and 600 for invertebrates). The values for P-32 should be added to Tables 3.1-2 through 3.1-5.

The ingestion dose factors in Tables 3.1-2 through 3.1-5 should be corrected to agree with the values from Regulatory Guide 1.109 shown below:

Ace Group (Tablel Nuclide Kidney GI-LLI Adult (3.1-2) Te-127m 2.75E 05 Teenager (3.1 3) Tc 101 8.75E 17 Teenager (3.1-3) Ni 65 5.19E 06 Infant (3.1-5) 1 132 3.76E 06 The Licensca's methodology to determine the dose or dose commitment to the maximum exposed member of the public due to radioactive materials released in liquid effluents is, in general, within the guidelines of NUREG 0133. However, any revision of the 00CM should include the above corrections to the input tables and justify the value used for Oy.

3.8 Dose due to Gaseous Effluent.1 Sections 3.1.2.1, C4.1, and C4.2.2.1 of the 00CM contain the methodology for calculating the gamma air dose and beta air dose to demonstrate ccepliance with the air dose limits of Technical Specification 3.!!.2.2, as required by Technical Specification 4.11.2.2. Technical Specification 3.11.2.2 requires that the air doses from each reactor unit be limited to less than or equal to 5 mrad for gamma radiation and 10 mrad for beta radiation during any calendar quarter, and less than or equal to 10 mrad for garna radiation and 20 mrad for beta radiation during any calendar year.

)

Since the introduction to the generic section states that GASPAR will l

normally be used for calculating the organ doses due to gaseous effluents,

! all the required site specific input data for GASFAR should be included in I

the 00CM.

16 I

The terminology in Sections 3.1.2 and C.4.2.2.1, for nsable gases, should be changed to indicate that the doses to be calculated are air doses and not doses to an individual. The air doses calculated are correct however, so the methodology in Sections 3.1.2.1 and C.4.2.2.1 is within the guidelines of NUREG 0133 and is considered acceptable.

Sections 3.1.2.2, C4.1, and C.4.2.2 of the ODCM contain the methodology for calculating the maximum dose to any organ of a me-ber of the public due to releases of "all radioiodines, radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than 8 days" in gaseous effluents. The equations in Section 3.1.2.2 result in more conservative calculations than are required by Technical Specification 4.11.2.3, since all radioiodines are included. The consideration of both H 3 and C 14 is implied in the classification of radionuclides with half lives greater than 8 days, but C 14 is not specifically addressed in the calculations nor is it required by the technical specifications. A value of 2 days is used for t f, the transport time from pasture to receptor for the grass cow meat pathway. The Licensee may wish to increase this to the 20 days recommended in Regulatory Guide 1.109.

The statement is made in Section C.4.1 that dose calculations are made using the methodology in the generic section, which considers all appropriate age groups. Section C4.2.2 states that it is assumed that the maximum exposed individual is a child or infant. Therefore, the same question arises here as in 'he discussion of liquid @sts. Does Section C4.2 apply to the general dose calculations or only to Section C.47 The Licensee's R; tables (Tables 3.1 13 through 3.1 30) were checked against results obtained by the reviewers using equations from NUREG 0133.

Agreement was obtair,td with Rj values for the inhalation pathway (except for the adult lung for which the factor for Cr 51 should be 1.44E+04) and for the vegetable pathway. The Licensee's values for the meat and milk pathways (Tables 3.1 16 through 3.126) could not be reproduced exactly, except for H 3. The Licensee's values are generally 15*. to 30?. lower than 17

the values calculated by the reviewers, with the range of values depending mainly on the radionuclide. An exception to the Licensee's values being  !

below the reviewer's values exists in the case of the cerium' isotopes, for whicn the Licensee's milk pathway Rj values are 5 times the reviewer's values. The reasons for the differences between the Licensee's values and the reviewer's values were not determined.

The Licensee's methodology for calculating the organ dose due to gaseous effluents generally follows the procedures of NUREG-0133 almost exactly.

However, the following inconsistencies and errors should be corrected:

• Regulatory Guide 1.109 permits the assumption that only one-half of the radioiodines are in elemental form, it is acceptable to include this assumption in the calculation of doses via the food chain pathways, but the various equations of the ODCM should be consistent. The factor E, accounting for only one half of the iodines being in elemental form, is included only in the equation for. the grass cow milk pathway factor, but is apparently used in calculations of the Rj values for the grass cow meat pathway and the vegetable pathway. The factor should either be added to the equations for these pathways or removed from the calculated values.

. The E factor for radiolodines permitted in the R value equations j for the food chain pathways, is not allowed in the calculation of i R values for the ground plane pathway, but was apparently used in the calculations. The R values for the ground plane should be recalculated without this factor.

. The inhalation dose factors in 00CM Tables 3.1 6 and 3.1 7 shculd be corrected to agree with the values from Regulatory Guide 1.109 shown below:

Ace Groue (Table) Nuclide Luno Adult (3.1 6) Zr 95 2.21E 04 Teenager (3.1 7) Ba la0 2.54E 04 18

. The value of the stable element transfer parameter in Table 3.1 11 for Te in ecw milk should be 1.0E-03 d/L and the parameter for Fe in goat milk should be 1.3E 04 d/L.

. Data for Mo 99 should be added to Tables 3.1-1 through 3.1-9 (inhalation dose factors) and Tables 3.1-12 through 3.1-30 (Rj values), since Mo-99 is one of the principle gamma emitters identified in the ODCH for which the LLD is specified.

Except for the inconsistencies and errors noted above, the Licensee's methodology to determine doses due to radioactive materials other than noble gases released in gaseous effluents is within the guidelines of NUREG 0133.

3.9 Dose Proiections Technical Specification 4.11.1.3.1 requires that doses due to liquid releases be projected at least once per 31 days in accordance with the methodology and parameters of the ODCM when the liquid Radwaste Treatment Systems are not being fully utilized. Sections 3.2 and C4.1 in the ODCM state that simplified calculations can be used for dose projections. ODCM Sections C4.2.1 and C4.3.1 discuss simplification of the calculations of doses due to liquid effluents and give acceptable dose equations based on releases of Cs-134 and Cs 137. However, the methodology for projecting doses that is requirad for compliance with Technical Specification 4.11.1.3.1 is not included in the 00CM.

Techniral Specification 4.11.2.4.1 requires that doses due to gaseous releases be projected at least once per 31 days in accordance with the methodology and parameters of the ODCM when the Gaseous Radwaste Treatment Systems are not being fully utilized. Sections 3.2 and C4.1 of the ODCH state that simplified calculations can be used for dose projections. Sections C4.2.2 and C4.3.2 discuss simplification of the calculatien of doses due to gaseous ef fluents and give acceptable dose equations based on releases of Xe 133 and I 131, respectively, for doses due to noble gases and doses due to radiotodines and particulates in gaseous effluents. However, the methodology for projecting deses that is required for compliance with Technical Specification 4.11.2.4.1 is not included in the ODCM.

19

The Licensee should add met.' agy for projecting doses due to liquid and gaseous affluents, including des due to anticipated unusual releases.

The Licensee's ODCM is incomplete insofar as the projection of doses is concerned.

3.10 Diaorams of Effluent Pathways The ODCM contains diagrams illustrating the treatment paths and major components of the radioactive liquid radwaste treatment system for both units in Figure C1.0.1. Diagrams illustrating the treatment paths and major components of the gaseous radwaste treatment systems for the two units are shown in 00CM Figure C1.0.2. A simplified diagram illustrating the solid waste treatment system is not included in the ODCH.

3.!! Total Dose Sections 3.3 and C4.4 of the ODCM describe the method used to determine the total dose to any member of the public, to demonstrate compliance with 40 CFR 190. when the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceed any of the quarterly limits of Technical Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. This requirement is actually more conservative that the requirements of Technical Specification 3.11.4.

Section 3.1.3 indicates doses due to direct radiation from Duke Nuclear Power Stations are much lower than 0.01 meem/yr. Section 3.3 discusses other possible fuel cycle sources and gives the conclusion that only doses due to releases of radioactive material in liquid and gaseous effluents from nuclear power production facilities are significant in the vicinities of Duke Nuclear Power Stations.

The Licensee's ODCM states that contributions from McGuire must be included in the fuel cycle calculations since McGuire Nuclear Station is thirty miles NNE of Catawba Nuclear Station. Section C4.4 requires that 20

~

e dose contributions from both McGuire and Catawba stations be calculated when determining the fuel cycle dose for demonstrating compliance with the limits of 40 CFR 190. The Licensee inay wish to consider deleting the McGuire contributions from the total dose calculations, since both the Catawba RETS and NUREG-0472 indicate in the bases statements that only nuclear fuel cycle facilities at the same site or within 8 km need be considered.

Section C4.4 states that the methods in 00CH Sections B.4.3.1, C4.3.1, B4.3.2, and C4.3.2 will be used for the fuel cycle dose calculations. The use of these methods conflicts with the requirements of generic Section 3.3.5, which requires that doses due to liquid and gaseous effluent releases will be calculated using the methodology of Sections 3.1.1 and 3.1.2 (the complete and conservative calculitions in the generic section). Also, use of these simplified dose estimates is not within the guidelines of Section 3.8 of NUREG 0133. Therefore, the methodology of the 00CM for calculating the fuel cycle dose for comparison with 40 CFR 190 limits should be changed.

3.12 Environmental Monitoring Program Section 5.0 of the 00CM identifies specific parameters of distance and the direction sector from the site and additional information for the samples identified in Environmental Monitoring Table 3.121 of Technical Specification 3.12.1. All sample locations required by Technical Specification 3.12.1 are described with the exception of one milk sample. If enough milk samples meeting the conditions of the technical specifications are not available, this fact should be notea in the table of the ODCH.

3.13 Su--ary The Licensee's 00CM, updated through Revision 20, uses documented and approved methods that are, in general, consistent with the methodology and guidance in NUREG 0133. However, it is recommended that the NRC request

.?.nother revision to address the major and other discrepancies identifie) in this review.

21

o CONCLUSIONS The Licensee's complete ODCM for the Catawba Nuclear Station, updated through Revision 20, dated January 1, 1988 was reviewed. it was determined that the ODCM uses methods that are, in general, consistent with the guidelines of NUREG 0133. However, it is recommended that another revision to the ODCM be submitted to address the aiscrepancies identified in the review.

The following discrepancies should be addressed:

. The introduction to the ODCM states that LADTAP and GASPAR are normally used for calculation of offsite doses. Therefore, all site specific data required as input to these codes should be included in the ODCM, as recommended by NUREG 0133.

. The source of the data and the equations used to' calculate the values in Table 1.2.2 should be identified in the ODCH.

. The equations of Section 3.1.1.2 and the data in Tables 3.1-13 through 3.1 18 should be made consistent either by adding the adjustment factor E (= 0.5) to the equation for the grass cow meat pathway or by doubling the values of the jR 's for iodines in the tables.

.. The exter al dose factors for radiciodines in Table 3.110 should be doublad, since they were apparently calculated using the adjustment f actor E (= 0.5), which is permitted only for the food patt. ways.

. For accuracy, the statements in Sections 3.1.2, 3.1.2.1, and C.4.2.2.1 should be changed to indicate that the doses to be calculated are doses to air and not doses to an individual.

. The equation in Section C2 l.1 that accounts for simultaneous releases via the VC and WL lines should be expanded to account for simultaneous releases from both the Waste Monitor Tanks and the Auxiliary Monitor Tanks.

22

e t t

- . Section C2.2 should specifically require that releases from the  !

Auxiliary Monitor Tank Building be considered when determining offsite dose rates due to gaseous effluents, f

. Section C2.2 should contain requirements for calculating offsite organ dose rates due to radioiodines, etc. in gaseous effluents at frequencies consistent with the frequency of analyses required by Technical Specification Table 4.11 2, *.s required by Technical Specification 4.11.2.1.2.

. The example in Section C3.1.1 of how the setpoints on the waste Liquid

> ifflur,t Line tu be calculated is unnecessary and could be deleted, since the previous sentence identifies the setpoint. Also, following the example to determine the setpoint may permit too high a concentration of untdentified radionuclides to be released to l

unrestricted areas; i.e.,1.0E 07 pCi/ml instead of the

~

! 3.0E-08 gCi/ml allowed by 10 CFR 20.

. The MPC for an unidentified mixture released to an unreuricted area l used in Section C3.1.5 to calculate the setpoint should be 3.0E 08 pCi/ml instead of 1.0E-07 pCi/ml.

. Figure Cl.0.1 should show whether the Convention)1 Waste Water l Treatment System Effluent Line and the liquid Waste Effluent Discharge

) Line release liquids to the unrestricted area at the same point. This

information is necessary to determine if the methodology of Section C3.1.5 ansures that releases are within the limits of 10 CFR 20.

f

. Methcdology should be added to Section C3.1 to ensu.e that the l concentration limit for radionuclides released offsite in liquid i effluents is not exceeded when releases from the Turbine Butiding l Sumps Discharge Line are released directly to the Liquid Waste Effluent Otscharge Line, as permitted by Section C2.1.2.d.

i

. Section C3.2 shauld include instructions for determining the setpoint for the gaseous effluent monitor en the Auxiliary Monitor Tank Building.

23

o e

- . Section C4.1 states that the methodology of the generic section shall be used for calculating dose contributions to the maxinium exposed individual and generi ,; ion 3.1.1 states that doses due to liquid effluents will be calcu)ated for each age group.  !

Therefore, site related ingestion dose commitment factors should be

added for the teenager, child, and infant age groups.  ;

. Sect'on C4.1 or 3.0 should include a commitment to assign releases  ;

of radioactive material and the resultant doses to the individual reactor units, especially for cases of disproportionate release from

{

the two units. '

. The Licensee should clarify whether Section C4.2 is intended to

, apply to all dose calculations or only to the calculations described I in Section C4.3. If the assumptions of Section C4.2 are intended to ,

apply to calculations required by Section C4.1 there are apparent conflicts with the generic sections of the 00CM. j

. In Section C4.3.1, the value of 37.7 for the dilution factor Og, j should be justified or referenced.

t i

j . Methodology should be added to Sections C4.3.1 and C4.3.2 for l projecting doses, including provisions to account for anticipated (

i- unusual releases,

. Section C4.4 should require that the fuel cycle dose (total dose) be j ulculated using the methodology of Sections 3.1.1 and 3.1.2. I i  :

. Equ hons and parameters used to calculate values in the  :

site specific data table, should either be referenced or given in f the OOCH. l l

. Section C5.0 of the 00CM should include information concerning any 7

- radiological environmental monitoring samples required by the [

] Technical Specification 3.12.1 are not available; e.g., milk j

samples.

I .

l 24 [

. t

e ,

e A chtcs cf data in the 00CM showed the following omissions or errors which should be coriected:

The inhalation and ingestion dose factors in 00CM Tables 3.1-2 through

3. 7 should be corrected to agree with the values from Regulatory Guide 1.109 shown below:

Ace Grouo(Table) Nuclide Kidney Luno GI LLI Adult (3.1 2) Te-127m 2.75E-05 Teenager (3.1 3) Ni 65 5.19E 06 Teenager (3.1-3) Tc 101 8.75E 17 Infant (3.1 5) 1-132 3.76E 06 Adult (3.1 6) Zr 95 2.21E-04 Teenager (3.1 7) Ba-140 2.54E 04

. The value of the stable element transfer parametter in Table 3.1-11 for Te in cow milk should be 1.0E-03 d/L and the parameter for Fe in goat milk should be 1.3E-04 d/L.

1

. In Table 3.11, the bloaccumulation factor for Na in fish should be corrected to 1.0E+02, tr.d bicaccu ulation facters fer P

(for which the 'btst vaines" are now 3000 for fish and 600 for invertebrates) should be added.

l . Ingestion dose factors for P-32 should be added to Tables 3.1 2

through 3.1 5.

l

. Data for Mo-99 should be added to Tables 3.1 1 through 3.1 9 (inhalation dose factors) and Tables 3.1-12 through 3.1-30 (Rg -values) should be added to the 00CM, since Mo 99 is one of the principle gamma emitters identified in the 00CM for which the LLO is specified.

The Licensee should re check the calculations of Rg values in Tables 3.1 16 through 3.1-26 by verifying the input parameters and the equations. The reviewer's values differed from the values in these tables for all radionuclides except H 3, 25

o e

Additionally, the Licensee may want to consider tha following changes:

The Licensee may wish to take advantage of the 1000 cfs per reactor t dilution flow permitted by Section 4.3 of NUREG 0133 for reactors with closed-cycle cooling systems. Using this value would require increasing the value of the recirculation factor (a) and probably reducing the value of Dy for drinking water, but should result in a lower and more realistic calculated dose +.han the present calculational method. As another alternative, the NRC Ctaff  ;

allows the dose for a reporting period (calendar quarter or calendar year) to be calculated using the average dilution flow for the reporting period, if this method is chosen for the dose calculations, the total dilution flow for the reporting period should be included in the semiannual report.

. Although the use of the actual dilution flow (F) averaged over the reportir.g period is acceptable for the calculations of doses de to i liquid effluents in Section 3.1.1, the Licensee may wish to takt advantage of the 1000 cfs per reactor permitted by Section 4.3 of j NUREG 0133 for reactors with closed cycle cooling systems. Using this value would require increasing the value of the recirculation  ;

factor (a) and probably reducing the value of Og for drinking water, but should result in a lower and more realistic calculated ,

dose than the present calculational method.

. The Licensee may wish to include requirements in the 00CM that the l

setpoints of radiation monitors on the liquid and gaseous effluent pathways be set to alarm before offsite dose rate limits are exceeded.  !

. To prevent spurious alarms of the radiation monitor on the Waste Liquid Effluent Line the Licensee may wish to add a requirement to  ;

Section C3.1.1 that the radiation monitor be set some fraction above l the concentration in the line, and to add a requirement to Section j C2.1.1 that the flow rate be set an equal fraction lower than is now required.

26

o .

. . i

^d

. The Licensee may wish to follow the recommendations of the bases statements in the CNS Technical Specifications and in NUREG-0472 and  ;

eliminate the calculations for McGuire from the fuel cycle  !

calculations.

. In Section 3.1.2.2 the Licensee may wish to increase the value of  ;

t f, (transport time from pasture to receptor) for the grass-cow meat pathway to the 20 days recommended by Regulatory Guide 1.109.  ;

f

. The Licensee may wish to modify the requirements in Sections 1.2 and  !

C2.2 to match the recommendations in the bases statement for  ;

2 Technical Specification 3.ll.2.1.b for CNS and in NUREG 0472: l 1.e., that the organ dose rate limit may be applied to the thyroid {

of a child via the inhalation pathway. .

l

, . The terms "activity" and "gross activity" are used throughout the 1

ODCM where "concentration" or "activity concentration" should be used. Changes in this notation would improve the accuracy of many  ;

statements in the 00CM.  !

l .

The value of F in the definition of Fj in Section 3.1.1 could be defined as the average dilution flow during the reporting period, [

calendar quarter or calendar year, when calculating the quarterly f and annual doses due to liquid effluents; this procedure for 4

calculating the quarterly and annual doses is acceptable to the NRC  ;

! Staff, if the total volume released in the diluting stream is also f reported in the semiannual reports. I l \

l.

) [

i l

27  !

l i

)i  !

o

) '

~

5. REFERENCES

1. Title 10, Code of Federal Reaulations, Part 50, Appendix 1, "Numerical ,

Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion, 'As low As is Reasonably Achievcble,' for Radioactive Material in Light Water Cooled Nuclear Power Reactor l

Effluents.'

i

2. Letter from H. B. Tucker (DPCo) to H. R. Denton (NRC), Re.: Catawba -

Nuclear Station - McGuire Nuclear Station - Oconee Nuclear Station - ,

Docket Nos. 50 413, 50-414, 50 369, 50 370, 50 269, 50 270, 50 287 -

Offsite Oose Calculational Manual, August 16, 1984. j i

l

3. Letter from E. G. Adensam (NRC) to H. B. Tucker (OPCo) "Offsite Oose  !

Calculation Manual (00CM) - Acceptability and Transmittal of Agenda l Items for a Meeting on Inservice Testing Program for Catawba Unit 1, j l October 19, 1984 j 4. Letter from H. B. Tucker (OPCo) to J. N. Grace (NRC),

Subject:

McGuire  !

Nuclear Station, Docket Nos. 50 369/370 - Semi annual Radioactive l Effluent Release Report, Septemb*r 7,1984, i

5. Letter from H. B. Tucker (DPCo) to H. R. Denton (NRC)," McGuire Nuclear [

l Station Docket Nos. 50-369 and 50 370 - Catawba Nuclear Station - l Occket Nos. 50 413 and 50 414 - Offsite Dose Calculation Manual,"

l l

June 24, 1985. j l i j 6. Letter from H. B. Tucker (DPCo) to H. R. Denton,

Subject:

Oconee Nuclear .

} Station Docket Nos. 50 269, 270, 287 - Catawba Nuclear Station - l j Docket Nos. 50 413 and 50 44 Offsite Dose Calculation Manual, 7 October 2, 1986.  !

I j 7. Lette: from H. B. Tucker (OPCo) to Documcat Control Desk (NRC), (title  !

unknown), January 6, 1987, i

28

o

. /.

0

8. Letter from H. B. Tucker (DPCo) to Document Contr>1 Desk (NRC),

Subject:

Oconee Nuclear Station - Docket Nos. 50-269, 270. -287, - McGuire Nuclear Station - Docket Nos. 50 369, -370, -Catawba Nuclear Station -

Oceket Nos. 50 413, 414, - Offsite Oose Calculation Manual, February 16, 1988.

9. Letter from H. B. Tucker (OPCo) to Document Control Oesk (NRC), Sub.iect:

Catawba Nuclear Station, Units 1 and 2 - Docket Nos. 50-413 and 50 414 -

Offsite Dose Calculation Manual, July 19, 1988.

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

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

12. "General Contents of the Offsite Oose Calculation Manual," Revision 1 Branch Technical Position, Radiological Atsessment Branch, NRC, February 8, 1979.

13. Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix 1," Regulatory Guide 1.109, Rev. 1, October 1977.

14. Title 10, Code of Federal Reculations, Part 20. "Standards for Protection Against Radiation."

29