ML20149M497
| ML20149M497 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 02/29/1988 |
| From: | Bohn T, Serrano W EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20147E808 | List: |
| References | |
| CON-FIN-D-6034 EGG-PHY-8003, NUDOCS 8802260079 | |
| Download: ML20149M497 (22) | |
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EGG PHY 8003 1
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i TECHNICAL EVALUATION REPORT I
l for the :
l EVALUATION OF OOCH REVISION 1 l i
GULF STATES UTILITIES COMPANY RIVER BEND STATION UNIT 1 l l
NRC Docket NO. 50-458 NRC LICENSE NO. NPF-47 ,
l
. W. Serrano T. S. Bohn l
l Published February 1988 Idaho National Engineering Laboratory l EG1G Idaho, Inc.
Prepared for the l
. U. S. Nuclear Regulatory Connission !
Washington, D.C. 20555 '
Under DOE Contract No. DE-AC07-76ID01570 l FIN No. D6034 j
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- g e ensanq 2a re. l
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ABSTRACT ,
The Offsite Dose Calculation Manual for the River Bend Station Unit I contains current methodology and parameters used in the calculation of offsite doses due to radioactive liquid and gaseous affluents, in the calculation of gaseous and liquid effluent monitoring alarnVtrip setpoints, and in the conduct of the environmental radiological monitoring program. Revision 3 dated August 28, 1985 of the ODCM manual was reviewed and found to be acceptable by the NRC, however, notification of approval was not transmitted to the Licensee. A considerable number of changes were made to the manual and submitted to the NRC as Revision 0. Follow-on changes were made to Revision 0 resulting in Revision 1 which was submitted to the NRC with the January-June 1987 Semiannual Radioactive Effluent Release Report. The NRC. transmitted this revision to EG&G Idaho at the Idaho National Engineering Laboratory (INEL) for review. The ODCM was reviewed and the results are presented in this report. It was determined that 00CM Revision 1 contains methods that are, in general, in agreement with the guidelines of NUREG 0133. However, because of the discrepancies identified in this review, it is recommended that another revision be submitted by the Licensee to address the identified concerns.
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FOREWORD
, This report is submitted as partial fulfillment of the "Review of 4
Radiological Issues for BWRs" 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 Commission funded the work under FIN D6034 and NRC B&R Number 20 19 10 12 2.
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 thia report, or represents that its use by such third party would not infringe privately-owned rights. ,
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- CONTENTS ;
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Page t Abstr4Ct. . . . . . . . . . . . . . . . . . . . . . . . . . . i. [
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Foreword. . . . . . . . . . . . . . . . . . . . . . . . .-.. . 11 !
, 1. Introduction. . . . . . . . . . . . . . . . . . . .-. . . 1 ,
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- 2. Review Criteria . . . . . . . . . . . . . . . . . . . . . 2 .
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- 3. Evaluation. . . . . . . . . . . . . . . . . . . . . . . . 3 ;
- 4. Conclusions . . . . . . . . . . . . . . . . . . . . . . . 13 l i
- 5. References. . . . . . . . . . . . . . . . . . . . . . . . 17 i
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- 1. INTRODUCTION
, Pureose of Rev's This document reports the review and evaluation of a recent version of j the Offsite Dose Calculation Manual (0DCM) submitted by the Gulf States l Utilities Company, thi Licensee for the River Bend Station, Unit 1. The I ODCM is a supplementary document for implementing the Radiological Effluent Technical Specifications (RETS) in compliance with 10 CFR 50, Appendix I requireteents.III l
Plant-Seecific Backettand
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The Gulf States Ut!11 ties Company (GSU) submitted ODCM Revision 3 for the River Bend Station (RBS) Unit I to the Nuclear Regulatory Comission ;
(NRC) with letter dated September 16,1985I23 The NRC reviewed the !
ODCM and found it to be an acceptable reference as stated in an internal NRCletterdatedOctober9,1985.I33 However, notification of approval l was not transmitted to the Licensee. Subsequent changes were made to the ODCM by the Licensee and reported to thrs NRC ire the Semiannual Radioactive Effluent Release Report for the lust half of 1966.[4] ine changes were coalesced into a new revision and submitted to the NRC as Revision 0 with l letter dated April 30,1987.(5) The Licensee made changes to Revision 0 l as reported in the January-June Semiannual Radioactive Effluent Release Report resulting in Revision 1 of the 00CM.l03 Revision 1 of the ODCM was submitted by the NRC to an independent review team at the Idaho National Engineering Laboratory (INEL) for review. The ODCM was re' viewed and the results and conclusions are presented in this report.
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- 2. REVIEW CRITERIA' Review criteria for the ODCM were provided by the NRC in three documents:
NUREG-0472, RETS for PWRsE73 NUREG-0473, RETS for BWRs(8) .
NUREG-0133, Preparation of RETS for Nuclear Power Plants.E93 The following NRC guidelines were also used in the 00CM review: "General Contents of the Offsite Dose Calculation Manual," Revision 1(10), and Regulatory Guide 1.109.Illl As specified in NUREG-0472 and NUREG-0473, the ODCM is to be developed by the Licensee to document the methodology and approaches used to calculate offsite doses and maintain the operability of the radioactive effluent systems. As a minimum, the ODCM should provide equations and methodology for the following:
o Alarm and trip setpoints on effluent instrumentation o Liquid effluent concentrations in unrestricted areas o Gaseous effluent dose rates at or beyond the site boundary o Liquid and gaseous effluent dose contributions o Liquid and gaseous affluent dose projections.
In addition, the 00CM should contain flow diagrams, consistent with the systems being used at the station, defining the treatment paths and the components of the radioactive liquid, gaseous, and solid waste management systems. A description and the location of samples in support of the environmental monitoring program are also needed in the ODCM.
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- 3. EVALUATION i I
At the RBS, Unit 1 is presently operational and Unit 2 is not expected to be operational for several years. As stated in the Introduction of the ODCM, the manual contains information and methodologies to be used by the RBS, Unit 1. ,
Liouid Effluent Pathways Condenser cooling for the 940 Net MWe RBS Unit 1 is provided by water circulated through a natural draft cooling tower. The.RBS is located approximately 24 miles north-northwest of Baton Rouge, Louisiana and just east of the Mississippi River which supplies make-up water to the -
circulating water system and receives decant from the cooling tower via the cooling tower blowdown line. The principal sources of liquid radwaste are the following:
Floor Drains Phase separators / Backwash tank subsystem
. Recovery Sample Tanks Reactor Water Cleanup The Liquid Radwaste System was designed to handle all radioactive liquid wastes and there is but one environmental release point at the site for the liquid radwaste. The system is operated as a batch system and the operating procedures used for all liquid radwaste equipment are based on ,
batch processing throughout the radwaste system. The type of operation allows time to sample and check the effluent streams before and after each process step to prevent inadvertent discharge of waste having a radioactivity level above the control limit. Only one of the four
, Recovery Sample tanks of liquid radwaste is released at a time and is considered a batch. Each batch is analyzed prior to release for gross beta / gamma activity, and the resulting specific activity is used to determine the discharge flow rate. Liquids with radioactivity levels exceeding specified limits are recycled for further processing. The batch release process is briefly discussed in ODCM Section 2.2.2.1.
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Prior to discharge into the cooling tower blowdown line, the waste in-these tanks is monitored for radioactivity by a radiation monitor.
Interlocks prevent a release if a high radiation level is detected.
During liquid releases, the flow rate, temperature, and activity level are continuously recorded. According to Section 11.2.2.5.1 of the FSAR, the I radwaste discharge flow is maintained at a predetermined level by a flow control valve operated from the central radwaste control panel.
Therefore, the flow control valve and the radiation monitor are the primary methods for controlling discharges from the liquid radwaste system. The liquid radwaste effluents are released to the cooling tower blowdown line prior to discharge into the Mississippi River.
Liouid Effluent Monitor Setooints Section 2.3.2.1 contains the methodology for detemining the setpoint for the liquid radwaste radiation monitor. The monitor provides both high alarm and automatic termination of release and the alam is set at the i technical specification limit. It may be prudent to add a low alarm with .
a setpoint slightly above the spurious alarm setting. The methodology described in Section 2.3.2 to determine the setpoint for the radiation inonitor in the liquid radwaste system is, in general, in agreement with the guidelincs of NUREG-0133 to provide reasonable assurance that the concentration limits of Technical Specification 3.11.1.1 will not be exceeded.
The liquid cffluent monitoring instrumentation Table 3.3.7.10-1 in the technical specifications identifies a radiation monitor with alarm function only for the cooling tower blowdotn line. The setpoint is set at twice the background which is sufficient to alarm at unusual levels of activity in the blowdown line.
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. . - _ . - . .. - , _ . . - - _a - = .- - - -
. Gaseous Effluent Pathways According to Unit 1 Technical Specification 3.3.7.11, there are .three monitored' environmental gaseous effluent release points for RBS Unit 1: 2!
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Main Plant Exhaust Duct l Fuel Building Exhaust Duct ,
l Radwaste Building Exhaust Duct The technical specifications identify noble gas monitors and iodine-an'd '
particulate samplers. All gaseous effluent releases from the Radwaste Building Exhaust Duct are assumed to be ground level releases.
Consequently, the maximum X/Q and D/Q values for these releases should be !
located at the site boundary. The Main Plant Exhaust Duct routine releases are treated as a wake split (conditionally' elevated) release.
Simultaneous releases from all release points are considered when determining the instantaneous dose rate.
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l Gaseous Effluent Monitor Setooints '
Section 3.3.2 of the ODCM contains the methodology used to determine the setpoints for the noble gas radiation monitors. There are several areas requiring discussion in this section:
The X/Q in Equation 3.3.2.2-1 is defined for "any area at or beyond the unrestricted area buundary" instead of for "any area at or beyond the Site Boundary" as stated in Technical Specification 3.11.2.1. The ODCM definition is consistent with the data in 00CM Table F which identifies the X/Q in the WNW sector at a distance of 994 meters which .
is the approximate distance to the unrestricted area bo.undary as indicated in the site map shown in Figure 1 of the 00CM. It is estimated that the X/Q in this direction sector would be reduced to approximately half the value listed in Table F if it were evaluated at the site boundary. However, the X/Q used in the ODCM is acceptable since it will yield conservative setpoints. If a reevaluation of X/Q 5
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. were made in the WNW sector at the site boundary, the resulting value may not be the maximum X/Q for all sectors.
For all setpoint equations in Section 3.3.2.2, the expressions on each side of the equal signs are not equal due to the 0.8 conservative factor and the . effective dose factor on one side of the equation.
It is not clear what is meant by the "monitor's loop accuracy" in ~
Section 3.3.2.2.a.i.Steo 5.
The HIGH ALARM setpoints are determined by a series of steps in Section 3.3.2.2.a.i. Step 4 identifies the actual high alarm setpoint whereas Step 6 determines another alarm setpoint. It is not clear which is the high alarm setpoint. This same situation exists'in Section 3.3.2.2.a.ii.
With the exception of the conservative X/Q used in the calculation, the methodology in Section 3.3.2.2 is in general, in agreement with the .
guidelines of NUREG-0133 to provide reasonable assurance that the noble l gas dose rate limits of Technical Specification 3.11.2.2 will not be exceeded.
Concentrations in Liouid Effluents Section 2.2.2.1 of the ODCM contains the methodology for demonstrating that the radionuclide concentrations in liquid effluents are in compliance with the technical specification. The methodology is, in general, within the guidelines of NUREG-0133 and should provide reasonable assurance that the concentrations at the point of release are maintained within the limits of Technical Specification 3.11.1.1.
Dose Rates in Gaseous Effluents Section 3.3.1.2 contains the methodology for demonstrating compliance with dose rate Technical Specification 3.11.2.1. It is mentioned that 6
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. l data tabulated in the River Bend Station FSAR are used as the bases for the radionuclide mix. It is not clear why data tabulated in the FSAR are used in lieu of actual data accumulated during plant operation. l l
Equations 3.3.1.2.2-1 and 3.1.2.2-2 are in units of mrem /sec instead
- of mrems/ year as ~specified in Technical Specification 3.11.2.1 since.the equations do not include the sec/ year conversion factor. The X/Q value is-I identified in Table F and may be a conservative value since the parameter is evaluated at the unrestricted area boundary in the W.i sector instead i of using the maximum X/Q determined at or beyond' the site boundary which l may or may not be in the WNW sector. In other words, a value obtained at l
the site boundary would probably be less than a value obtained at the unrestricted area t,oundary. The Licensee should be aware of this added l conservatism in the event the calculated dose rates approach the technical l specification limits.
The methodology for determining the dose rate due to the release of noble gases is, in general, in agreement with the guidelines of NUREG-0133. However, with the omission of the conversion factor and the conservatism in the X/Q value, it is uncertain if the methodology for the dose rates due to the release of noble gases will provide reasonable assurance that the calculated dose rates will be within the limits of Technical Specification 3.11.2.1.
Section 3.3.1.2.3 of the 00CM contains the methodology for i determining "Radioiodine and 8-day Particulate Release Rates" whereas Technical Specification 3.11.2.1 specifies iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days. Tritium is not addressed and "Radioiodine" implies that more j than just I-131 and I-133 are included in the calculation. The bases statement for Technical Specification 3.11.2.1.b states that the release rate of these nuclides restricts at all times the thyroid dose rate to a l child via-the inhalation pathway to less than or equal to
. 1500mrems/ year. However, the methodology in Section 3.3.1.2.3 is for the infant age group which is not the limiting age group and is not consistent i
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- l with the technical-specification. The definition for X/QD references Appendix E instead of the values listed in Appendix F. - Appendix E l contains the dispersion parameters evaluated at the unrestricted area boundary instead of at or beyond the site boundary. The values in Appendix E are acceptable since the values are much larger than the values in Appendix F which results in conservative dose rate estimates.
The. methodology for determining the dose rate due to the release of
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radionuclides in gaseous releases is, in general, in agreement with the guidelines of NUREG-0133. However, with the incorrect age group and the l conservatism created by the use of X/Qs at the unrestricted area boundary, i it is uncertain if the methodology presented in Section 3.3.1.2.3 will I provide reasonable assurance that the calculated dose rate will be within the limits of Technical Specification 3.11.2.1.b.
Dose Ose to liauid Effluents Section 2.4 of the ODCM contains the method for determining the dose to the maximum exposed member of the public due to radionuclides identified in liquid effluents to demonstrate compliance with the dose limits of Technical Specification 3.11.1.2. The doses are calculated for the adult age group using the potable water, aquatic foods, and invertebrate consumption pathways. There are several areas of concern in this section:
Equation 2.4.2-1 apparently is intended to determine the dose to organ r from a single batch release. If this is.the case, it '
should be clearly stated in definitions for D ra and Qg). As written it is not clear if the dose is determined for the release of a single batch or a series of batches. The confusion arises because of At). It is not clear if "1" identifies the single batch released or if it is the 1th batch released in a series of batches released during the specific time interval. If it is'for a single !
batch, then the "1" is not necessary.
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i DF) should be defined as the total dilution volume during the l reporting period (i.e., calendar quarter or year) and not just during ,
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j The primary source of onsite dilution water for RBS is the cooling tower blowdown with a minimum discharge flow of 2200 gal / min (4.9 ft 3/sec). For the dose calculation only, Section 4.3 of NUREG-0133 permits an adjustment factor to increase the dilution. term to no more than 1000 ft3 /sec during the reporting period as long as the receiving body of water can support the adjusted dilution.- This adjustment is permitted for plants not having a once-through condenser cooling system. Without this adjustment factor, the calculated doses at RBS could exceed the dose limits of the technical specifications.
Therefore, it is to the Licensee's advantage to use a dilution volume in accordance with NUREG-0133 instead of the adjusted dilution volume of Equation 2.4.2-1 Equation 2.4.2-2 should use an index other than "i" in order not to
. confuse it with radionuclide "i" of previous Equation 2.4.2-1.
The methodology for calculating doses due to the release of radioactivity in liquid effluents is, in general, in agreement with the guidelines of NUREG-0133. However, without the adjustment allowed by NUREG-0133, it is uncertain if the methodology will provide reasonable assurance that the calculated doses will be within the limits of Technical l Specification 3.11.1.2. I Dose due to Gaseous Effluents l Section 3.4.1.2.a of the ODCM contains the methodology for calculating the cumu,lative dose due to the release of radioactive noble gases in gaseous effluents to demonstrate compliance with the dose limits of
- Technical Specification 3.11.2.2. In Section 3.4.1.2.a of the ODCM, Appendix E is referenced for the X/Q values. It is suspected that this is 9
a typo'and the reference was supposed to be to Appendix F. If it is not a f
typo, then the values for X/Q in Appendix E should be evaluated at or beyond the site boundary for consistency with Technical Specification 3.11.2.2 instead of being evaluated at the unrestricted area ;
boundary. The site boundary map in Figure'1 of the ODCM shows the
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unrestricted area boundary and the site boundary to be in coincident in !
several sectors and there are sectors where the two boundaries are not in coincident. Where th.e two boundaries are not in coincident, the X/Q~ !
values evaluated at or beyond the site boundary should be-less than the l
values listed in Appendix E. The Licensee should be aware of this added l
conservatism incurred by using the Appendix E values instead of the j Appendix F values in the event the calculated dose approaches the technical specification limit. l l
l The methodology for calculating the maximum dose to air due to the -
release of radioactive noble gases is, in general, in agreement with the i guidelines of NUREG-0133. However, with use of conservative X/Q values, it is uncertain if the method will provide reasonable assurance that the -
calculated dose will me within the limits of Technical .
Specification 3.11.2.2. I Section 3.4.2.2 of the 00CM contains 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 to demonstrate compliance with the dose limits of Technical Specification 3.11.2.3. The X/Q and D/Q values used in the calculation of the dose to the maximum exposed individual are listed in Table F.
The methodology for calculating the dose to the maximum exposed individual due to the release of radionuclides in gaseous effluents is, in general, in agreement with the guidelines of NUREG-0133 and should provide reasonable assurance that the dose limits of Technical Specification 3.11.2.3 will not be exceeded, i
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. Dose Proiections Section 2.5 of the ODCM contains the. method to project doses to determine use of the liquid radwaste treatment system as required in Technical Specification 3.11.1.3. However, it appears that the projections will result in unrealistically high doses requiring more frequent use of the radwaste treatment system. It is not clear why XD is defined as the number of hours during which releases occurred instead of the number of hours accumulated to date during the month. The Licensee should reconsider the definition of XD-Section 3.5 of the ODCM contains the method to project doses to determine use of the ventilation exhaust treatment system as required in Technical Specification 3.11.2.5. However, it appears that the projections will result in unrealistically high doses requiring more frequent use of the radwaste treatment system. It is not clear why XD is defined as the number of hours during which the unit was operational instead of the number of hours accumulated to date during the month. The Licensee should recer. sider the definition of XD-Diaorams of Effluent Pathways Simplified diagrams of the liquid and gaseous radwaste treatment systems are contained in Figures 2 and 4, respectively. However, the figures are illegible and should be replaced. A simplified diagram illustrating the solid waste treatment system is not included in the ODCM.
lotal Dose Section 5.0 of the ODCM describes the method used to demonstrate compliance with the total dose technical specification. The method for determining the total dose is an acceptable method for demonstrating compliance with the dose limits of Technical Specification 3.11.4 1 11 I i
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l Environmental Monitorino Procram Table 4.1 in Section 4.0 of the ODCM identifies specific parameters of distance and the direction sector from the site and additional information for each and every sample identified in Environmental Monitoring Table 3.12.1-1 of Technical Specification 3.12.1. Table 4.1 of the ODCM under "Airborne Particulates and Radioiodines" includes "radioiodines" instead of only "I-131". Since radiciodines implies all radioiodines, the Licensee should consider changing to include I-131 only. Figures 1 and 5 l are illegible and should be replaced. !
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In summary, the Licensee's ODCM, as revised, uses documented and j approved methods that are generally consistent with the methodology and j guidance in NUREG-0133. However, because of the discrepancies identified in this review, it is recommended that the NRC request another revision to address the concerns identified in this review.
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. 4. CONCLUSIONS ;
i The Licensee's ODCM Revision 1 for the River Bend Station Unit I was f reviewed.- It was determined that the 00CM uses methods that are, in l
genaral, consistent with the guidelines of NUREG-0133. However, it is j recomended that another revision to the 00CM be submitted to address the i discrepancies. identified in the review. l l
i The following are considered to be major discrepancies: !
o In Section 2.4, the DF) should be defined as the total! dilution !
water volume during the reporting period instead of the dilution )
volume during the release period which may result in overly _ j conservative calculated dosas. ~
o In Section 2.4, the dilution volume should be adjusted in j accordance with the recomendations of Section 4.3 of NUREG-0133. i l
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o In Section 3.3.1.2.1, the X/Q is evaluated at the unrestricted l
area boundary which may result in overly conservative calculated l dose rates for noble gases instead of being evaluated at the site }
boundary.
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5 o In Section 3.3.1.2.3, the dose rate is determined for the infant age group instead of the child age group. The child age group is ,
the limiting age group and is the required age group as defined j in the bases statement of Technical Specification 3.II.2.1.b. i o In Section 3.4.l.2.a the X/Q referenced in Appendix E is evaluated at the unrestricted area boundary (instead of being
, evaluated at the site boundary) which may result in overly conservative calculated doses to air.
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The following are additional discrepancies:
o In Section 2.4, it is not clear if Equation 2.4.2-1 determines !
i the dose due to liquid effluents for the release of a single batch or for a series of batches.
o In Section 2.4.2, Equation 2.4.2-2 should use an index other than i "i" in order not to confuse it with the radionuclide "i" used in l Equation 2.4.2-1.
o In Sections 2.5 and 3.5, a different definition should be considered for X D , since it appears that the existing definition will result in overly conservative dose projections. I o In Section 3.3.1.2.1 and 3.3.1.2.2, data from the FSAR are used for the gaseous effluent radionuclide mix in the dose rate j calculations instead of using actual plant data. l o In Section 3.3.1.2, Equations 3.3.1.2.2-1 and 3.3.1.2-2 have units of mrem /sec instead of mrems/ year. The mrems/ year are required for consistency with Technical Specification 3.11.2.1.
Therefore, a constant must be included in the equations to adjust the calculated result to mrems/ year.
o Section 3.3.1.2.3 references Table E for X/QD instead of the values from Table F. The data in Table F should be used in the dose calculations due to the release of I-131, I-133, particulates with half lives greater than eight days, and tritium, o In Section 3.3.2.2, the expressions on each side of the equal signs are not equal because of the 0.8 factor and the effective -
' dose factor on one side of the equation.
o In Sections 3.3.2.2.a.1 and 3.3.2.2.a.ii, it is not clear which setpoint is the actual high alarm setpoint.
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i l' o In Section 3.3.2.2.a.1, the X/Q used to determine the noble gas monitor setpoint is defined and evaluated at or beyond the I unrestricted area boundary instead of at or beyond the site
. boundary as required in Technical Specification 3.11.2.1.
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. o In Section 3.3.2.2.a.i.Steo 5, it is not clear what is meant by the "monitor's loop accuracy".
o Figures 2 and 4 illustrating the liquid and gaseous radwaste treatment systems are illegible and should be replaced, o A figure illustrating the solid waste treatment system is not included in the ODCM.
The following are not discrepancies in the 00CM, but are suggestions that should be brought to the attention of the Licensee:
. o In Section 2.3.2.1, it may be prudent to include a low level alarm for the liquid radwaste monitor with a setpoint slightly above the spurious alarm setting.
o In Section 3.3.1.2.3, tritium is not addressed and "Radiciodines-is stated which implies all radiciodines instead of only "I-131 and I-133" as required in Technical Specification 3.11.2.1.
o In Section 4.0, Table 4.1 under "Airborne Particulates and Radioiodines" specifies "radioiodines" instead of only "I-131".
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, 5. REFERENCES
- 1. Title 10, Code of Federal Reaulations, Part 50, Appendix I, "Numerical Guides for Design Objectives and Limiting Conditions for_0peration 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. E. Booker (RBS) to H. R. Denton (NRC),
Subject:
River Bend Station Submittal of Offsite Dose Calculation Manual (00CM) -
Revision 3, September 16,1985.
- 3. Memo from D. R. Muller (NRC) to T. M. Novak (NRC),
Subject:
DSI ACCEPTANCE OF "0FFSITE DOSE CALCULATION MANUAL" (ODCM) FOR RIVER BEND, UNIT 1, October 9, 1985.
- 4. Letter from J. E. Booker (RBS) to U. S. Nuclear Regulatory Commission,
Subject:
Semiannual Radioactive Effluent Release Report, Third &
Fourth Quarters 1986, February 27, 1987.
- 5. Letter from J. E. Booker (RBS) to U. S. Nuclear Regulatory Commission,
Subject:
Rev 0 to Procedure RSP-0008, "River Bend Station, Offsite
. Dase Calculation Manual," April 7;0, 1987.
- 6. Let'.er from J. E. Booker (RBS) to U. S. Nuclear Regulatory Commission,
Subject:
River Bend Station - Unit 1 Docket No. 50-458 Semiannual Radioactive Effluent Release Report for the Period of January 1 to June 30, 1987, RBG-26536, August 31, 1987.
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- 7. "Radiological Effluent Technical Specifications' for Pressurized Water
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Reactors,",Rev. 3, Draft 7", intended for contractor guidance in I
. reviewing RETS proposals for operating reactors, NUREG-0472, j September 1982. !
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. , - . . , _ . , , . , , _ , _ . , _ , , _ , , ,,_,,,.y ,_y__,.,, r "e e---'+-Nre-=n'm--v'---*i-v""' -'C-W'y *7N~ ~'W'
- 8. "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.
- 9. "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, A Guidance Manual for Users of Standard .
Technical Specifications," NUREG-0133, October 1978.
- 10. "General Contents of the Offsite Dose Calculation Manual," Revision 1 Branch Technical Position, Radiological Assessment Branch, NRC, -
February 8, 1979.
- 11. Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Regulatory Guide 1.109, Rev. 1, October 1977.
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