ML20151C234
| ML20151C234 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/31/1988 |
| From: | Bohn T, Serrano W EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML17279B071 | List: |
| References | |
| CON-FIN-D-6034 EGG-PHY-8032, NUDOCS 8804120250 | |
| Download: ML20151C234 (23) | |
Text
E'iCL35URE EGG PHY 8032 TECHNICAL EVALUATION REPORT for the EVALVATION OF 00CM THROUGH AMEN 0 MENT 4 WASHINGTON PUBLIC POWER SUPPLY SYSTEM WASHINGTON NUCLEAR PLANT NO. 2 NRC Docket NO. 50 397 NRC LICENSE NO. NPF-21 T. S. Bohn W. Serrano Published March 1988 Ideno National Engineering Laboratory EGaG Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761D01570 FIN No. 06034
ACKNOWLEDGMENTS f
The authors wish to express their appreciation to C. R. Amaro and M. R. Winberg for their assistance in proofing the data tables and 1
comparing the Environmental Monitoring Program in the ODCM with the P
requirements of the Technical Specifications.
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i ABSTRACT The Offsite Dose Calculation Manual (ODCM) for the Washington Nuclear plant No. 2 (WNP 2) contains current methodology and parameters used in the calculation of offsite doses dJe to radioactive liquid and gaseous effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the execution of the environmental raciological monitoring program. Amendment 4 dated August 1986 of the ODCM was submitted to the NRC in the Semiannual Effluent Report for January-June 1986. The NRC transmitted Amendment 4 to the Idaho N&tional Engineering Laboratory (INEL) for review. The revised ODCM was reviewed in its entire'ty by EG&G Idaho at the INEL and the results of the review are presented in this report.
It was determined that the WNP-2 ODCM, updated through Amendment 4. 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 and correct 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 contracted by the Idano National Engineering Laboratory for the U. S. Nuclear Regulatory Commissicn, Office of Nuclear Reactor Regulation. The U. S. Nuclear Regulatory Commission funded the work under FIN 06034 and NRC B&R Number 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, not 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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I CONTENTS
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Abstract............................
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. Foreword............................-.
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Introduction............ s...........
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Review Criteria.......................
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Evaluation.........................
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Concl u s i on s.........................
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References.........................
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INTRODUCTION Pureose of Review This document reports the review and evaluation of the Offsite Oose Calculation Manual (ODCM) through Amendment 4 submitted by.the Washington Public Power Supply System (WPPSS), the Licensee for the Washington Nuclear Plant No. 2 (WNP 2). The 00CM is a supplementary document for implementing the Radiolegical Effluent Technical Specifications (RETS) in compliance with 10 CFR 50, Appendix ! requirements.II3 To date, a i
version of an,ODCM has not been approved by the NRC.
Plant-Seecific Backaround WPPSS submitted 00CM Amendment 4, dated August 20, 1986 for WNP-2 to the Nuclear Regulatory Commission (NRC) in the Semiannual Effluent Report for the first half of 1986.(2) Amendment 4 along with an ODCM complete through Amendment 3 was rubmitted by the NRC to an independent review team at the Idaho National Engineering Laboratory (INEL). The INEL team reviewed the 00CM as a whole and the results and conclusions are presented in this report.
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REVIEW CRITERIA Review criteria for the 00CM were provided by the NRC in two documents:
NUREG-0473, RETS for BWRsI33 NUREG 0133, Preparation of RETS for Nuclear Power Plants.E43 l
"General The following NRC guidelines were also used in the 00CM review:
Contents of the Offsite Dose Calculation Manual," Revision 1(53, and Regulatory Guide 1.109 Revision 1.(6)
As specified in NUREG 0473, the 00CM 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 As a minimum, the 00CM should provide equations and methodology
- stems.
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 ef fluent 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 l
solid waste management systems. These flow diagrams should be consistent I
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.
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EVALUATION WNP 2 is presently the only operational unit at the site with the construction of WNP-1 on an indefinite status. The ODCM describes methods of calculating radioactive concentrations in the environment and the potentially resultant personal dose equivalent commitments offsite that are associated with the liquid and gaseous effluents of WNP-2.
Licuid Effluent PathMjLu WNP-2 is located in the southeast area cf the DOE Hanford Reservation in Benton County, Washington. The site is approximately three miles west of the Columbia River, eight miles north of the City of Richland,12 miles north of the City of Pasco, and 21 miles northwest of the City of Kennewick..The Columbia River supplies makeup water to the circulating water system and receives decant from the cooling tower via the cooling tower blowdown line. Condenser cooling for WNP-2 is provided by water circulated through mechanical draft cooling towers.
The liquid waste management system is designed to collect, segregate, store, and process potentially radioactive liquids generated during normal The plant operation and anticipated operational unusual occurrences.
design of the liquid waste management system incorporates the objectives of maximum recycle and minimum release of radioactive liquid without limiting plant operations or availability. The radioactive liquids to be processed in the radwaste treatment system are collected in the following tanks:
Waste Surge Tank Waste Collector Tank Chemical Waste Tanks (2)
Floor Drain Collector Tank Effluents from the above tanks are processed in the liquid radwaste l
l treatment system and are discharged to the blowdown line as batch releases 3
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from one of the following tanks:
Waste Sample' Tanks (2)
Distillate Tanks (2)
Floor Drain ' Sample Tank According to Section 2.2 of the 00CM, all liquid radwaste effluent meant for release, passes through gng monitored four-inch line (liquid effluent discharge line) which discharges into the 36-inch cooling tower blowdown line for dilution prior to discharge into tha Columbia River.
In addition to the discharges through the above liquid effluent discharge line, there are three liquid release streams that are normally non-radioactive but have a finite possibility of having radioactive material injected into them. These liquid streams are the:
Standby Service Water Turbine Building Service Water
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Turbine Building Sump Water To prevent any discharges of radioactive liquid from these streams, radiation monitoring systems have been installed to detect any increase above the normal background concentration of radioactive material. Of the three, only the Turbine Building Sump Water monitor has an automatic trip.
In case of an alarm, the sunp effluents are automatically routed to the radioactive waste system. The other two streams contain gross radioactivity monitor's which provide an alarm but no automatic termination of release.
Radioactivity released from the liquid radwaste system is the primary concern when assuring compliance to the concentration and the dose limits in the technical specifications. A block diagram describing the liquid waste management system and effluent pathways is missing from the ODCM and must be included. Figure 1 of this report contains a block diagram of the liquid radwaste system as simplified from Figure 3.5-1 in the WNP 2 Environmental Report.
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Diagram of Liquid Radwaste System SUMPS Waste Surge Waste Cooling Radwaste Bldg and Sample Tower Turbine Bldg Collector 3: Tanks (2)
Blowdown Line Reactor Bldg Tanks Drywell S/
u Floor Drain Floor Drain
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Collector
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Sample Tank Tank
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Distillate Tanks (2)
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Misc Waste Detergent chemical
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Reactor Drain
-t Waste Bldg Tanks Tanks Solid Condensate
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Storage Waste Shop Decon Tanks (2)
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Chem Pumps Decon Drain Reactor Bldg gf Plant Turbine Bldg Use j
\\1 Columbia River 9
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Liouid Effluent Monitor Setooint j
Section 2.5.3 of the ODCM contains the methodology used to determine the setpoint for the radiation monitor in the liquid radwaste discharge line, and dilution flow rates for compliance with Technical Specification 3.3.7.11.
The concentration of radioactive nuclides discharged through the liquid radwaste discharge line is monitored and will alarm and terminate any discharge exceeding the concentration determined by the isotopic analysis of a prerelease grab sample.
Section 2.9 of the ODCM contains the methodology for determining the setpoints of 'the radiation monitors for the three non-radioactive liquid The detector's maximum setpoint for each of the three systems is streams.
based on the Maximum Permissible Concentration (NPC) of Cs-137. To ensure,
that the 10 CFR 20 limits are never exceeded, the alarm setpoint is set at 80% of the maximum setpoint plus background.
The methodologies described in Sections 2.5 and 2.9 of the ODCM for determining the setpoints for the radioactive monitors in the radioactive and potentially radioactive liquid effluent streams are, in general, in agreement with the guidelines of NUREG-0133 and are considered acceptable.
Gaseous Effluent Pathways According to WNP-2 Technical Specification 3.3.7.12-1, there are three monitored environmental gaseous effluent release points:
Main Plant Vent Turbine Building Vent Radwaste Building Vent The technical specifications identify noble gas monitors, iodine and particulate samplers, and effluent flow measuring devices to monitor gaseous effluent releases.
Sampler collection media are routinely 4.3.7.12-1.
analyzed in accordance with Technical Specification Table 6
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Each release point is continuously monitored during the release of noble gases and all gaseous effluent releases are treated as ground releases.
As stated is section 3.1 of the ODCM, "WNP 2 gaseous effluents are released on a continuous basis; in addition, batch releases also occur when containment and mechanical vacuum pump purges are performed and when the OFF-GAS treatment system operates in the charcoal bypass mode." The ODCM does not contain a simplified diagram of the Gaseous Effluent
.reatment System.
Gaseous Effluent Monitor Setooints Section 3.6.2 of the 00CM contains the methodology used to determine the setpoints of the noble gas radiation monitors at the three gaseous effluent release points as required by Technical Specification 3.3.7.12.
Methodology for calculating the setpoints based on the skin and whole body dose rates are included with the more conservative setpoint being used.
Simultaneous releases from each of the three release points are considered in the equations for determining the setpoints.
The methodology uses the maximum normalized diffusion coefficient at the site boundary due to each release point. The Turbine Building and Radwaste Building values are ' based on average annual ground level values.
The Main Plant vent release values are for mixed mode and may be either short term or average annual values depending upon the type of release.
The monitors for these three vents have no a,utomatic termination of release but p'rovides annunciation only.
In Section 3.6.2 of the 00CM, the frequency of sampling and analysis of radioiodines and radionuclides in particulate form is stated to be performed monthly, however, Table 4.112 of the technical specifications states that the ' Minimum Analysis Frequency" must be weekly. The 00CM requirements are not consistent with the requirements of the technical specifications.
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In Section 3.6.2 of the ODCM, in Equations 22, 24, and 25 the symbols Cjj and C j are defined as representing both measured and calculated T
concentration values for gaseous radionuclides. Different symbols should be used to represent the measured and calculated values.
The methodology for determining the setpoints for the noble gas effluent monitor is, in general, in agreement with the guidelines of hUREG-0133 and is considered acceptable.
Concentrations in Liauid Effluents Sections 2.3, 2.5 and 2.9 of the ODCM contain the methodology for demonstrating that the radionuclide concentrations in the released liquid effluents required by Surveillance Technical Specification 4.11.1.1.2.
The technical specification requires instantaneous compliance to the concentration limits of 10 CFR 20 Appendix B Table II Column 2.
In the first paragraph of Section 2.1 of the 00CM it states that, "a theoretical, continuous concentration of radionuclides at the 10 CFR 20 limits at the point of discharge to the Columbia River will result in compliance with 10 CFR 50, Appendix I limits in the unrestricted areas."
This is not true, since a continuous release of even a single radionuclide at the MPC limit will result in an annual dose far in excess of the annual dose limit identified in 10 CFR 50, Appendix I.
i The methodologies for determining radionuclide concentrations in the released liquid effluents are, in general, within the guidelines of NUREG-0133 and are considered acceptable.
Activity in Outside Temocrary Tanks In Section 2.8 of the 00CM, the use of temporary liquid radwaste holdup tanks at WNP-2 is discussed along with the methodology for 8
f determining their activity limit as required by Technical Specification 3.11.1.4.
Technical Specification 3.11.1.4 states that "the quantity of radioactive material contained in any outside temporary tanks shall be limited to the limits calculated in the CDCM such that a complete release of the tank contents would not result in a concentration at the nearest offsite potable water supply that would exceed the limits specified in 10 CFR 20 Appendix B Table II."
The 00CM contains methodology for determining the tank's allowed activity based on a calculation of the concentration at the WNP-1.well assuming all the WNP-2 tank's contents migrated to the WHP-1 well through the water table.
The 00CM also limits the tank's activity to 10 curies.
The method presented in the 00CM for determining the total activity in outside temporary tanks appears to be acceptable and satisfies the requirements of the technical specifications. However, a simple demonstration of compliance to the technical specification could be made by multiplying the radioactivity concentration of the liquid in the tank by the volume of liquid in the tank and comparing the resultant activity to the 10 curie limit.
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Dose Rates in Gaseous Effluents Section 3.3.1 of the ODCM contains the methods for determining the noble gas dose rates as required by Surveillance Technical Specification 4.11.2.1.
The total body and skin dose rates due to the release of noble gases are assured to be within the dose rate limits by correctly setting the setpoints for the noble gas monitors as previously l
described in this report.
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l Section 3.3.2 of the ODCM contains the methods for determining the dose rate to any organ due to the release of I-131, I-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days to areas at or beyond the unrestricted area hs required by Surveillance Technical Specification 4.11.2.1.2.
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In Sections 3.3.1 and 3.3.2 of the ODCM, the letter "m" in Equations 1 and 2 is used both for the number of nuclides on the sumation sign and also to indicate a mixed mode release. Different symbols should be used to eliminate the confusion.
The methodology in Section 3.3.2, for determining the dose rate to any organ due to the release of radioiodines, tritium, and radionuclides in particulate form is, in general, in agreement'with the guidelines of NUREG-0133 and is considered acceptable.
Dose Due to Liauid Effluents Section 2.4 of the ODCM contains the methodology for determining the dose or dose commitment to a MEMBER OF THE PUBLIC due to radioactive material released in liquid effluents as required by Surveillance Technical Specification 4.11.1.2.
The dose for the reporting period is determined by computing the accumulated dose comitment to the most exposed organ and total body of the maximum exposed individual, which is assumed to be the adult, whose exposure pathways include potable water and fish consumption.
Calculations are made for those nuclides identified at the point of
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discharge.
th release In equation 5 of section 2.4, the time length of the I period T), is incorrectly defined as the time "over which Cjj and l
l F) are averaged for liquid releases." The value of Cj) is not averaged over T, but is the value determined from the analysis of the 1
pre-release grab sample. Also, it is not clear if the concentration, Cj), in Equation 5 is the diluted or undiluted concentration.
In Section 2.4, the definition of the parameter, F, in Equation 5 1
includes a factor of 100 times the average flow from the site structure.
According to NUREG-0133, for plants with cooling towers, the product of 10 I
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the average blowdown flow to the receiving water body times an applicable factor can be 1000 cfs or less.
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l In Section 2.4 of the ODCM, the dilution factor, D, used in w
Equation 7 is defined as the dilution factor from the near field area to l
l the nearest potable water intake and is given as 200. A reference must be provided to justify the value.
Section 2.7d contains equations for calculating doses to man from each liquid effluent pathway.
In the definition of P, the effective "surface density" for. soil, the reference should be Table E-15 instead of E-1 in Regulatory Guide 1.109, Revision 1.
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The methodology for calculating doses due to the release of radioactivity in liquid effluents is, in general, in agreement with the
- However, guidelines of NUREG-0133 and Regulatory Guide 1.109, Revision 1.
it is uncertain due to the confusion in the definitions of Cj) and F1 if the dose calculations will result in values that can be used to compare with the limits of. Technical Specification 3.11.1.2.
Dose Due to Gaseous EffluentJ Section 3.4.1 of the ODCM-contains the methodology for calculating the cumulative gamma and beta air doses due to the release of radioactive noble gases as required by Surveillanco Technical Specification 4.11.2.2.
The methodology for calculating the air dose due to the release of radioactive noble gases is, in general, in agreement with Regulatory Guide 1.109, Revision 1, and is considered acceptable.
l Section 3.4.2 of the OOCH contains the methodology for calculating the cumulative dose due to the release of I-131,1-133, tritium, and radionuclides in particulate form with half-lives greater than eight days as required by Surveillance Technical Specification 4.11.2.3.
In the first paragraph of Section 3.4.2, tritium is specifically omitted as a 11
radioactive contributor to the dose limits which is inconsistent with the technical specification.
Tables 3-Sa through 3-5d contains dose parameters for all age groups and all pathways for the maximum exposed organ. The reviewer was unable to reproduce the dose parameters for the milk (cow), milk (goat),
vegetable, and meat pathways. The inhalation dose parameter for Cs-137 in 3
Table 3-5d should be 1.1E+6 instead of 5.5E+5 mrem m /pci-yr. The inhalation dose parameter value in Table 3-5d.for Ce-144 should be 7.8E+6 3
instead of 8.6E+7 mrem-m /pCi-yr.
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 general, in agreement with '-
the guidelines of NUREG 0133 and Regulatory Guide 1.109, Revision 1.
However, because most of the dose factors in the data tables cannot be reproduced, it is uncertain if the dose calculations will result in values that can be used to compare with the dose limits of Technical Specification 3.11.2.3.
Oose Proiections Section 2.4.1 of the ODCM describes the method used to project doses due to the expected release of radioactive liquids to determine when the liquid radwaste treatment system shall be operated as required by Surveillance Technical Specification 4.11.1.3.1.
The methodology for determining the dose projection due to liquid radwaste effluents is, in general, in agreement with NUREG 0133 and is considered acceptable.
Technical Specification 3.11.2.5 requires dose projections to determine the required use of the Ventilation Exhaust Treatment System.
The dose projection methods are not included in the ODCM as required by Surveillance Technical Specification 4.11.2.5.1 12
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Qjaarams of Effluent Pathways Simplified flow diagrams of the liquid, gaseous, and solid radwaste treatment systems are missing from the 00CM and must be included.
Total Dose Section 4.2 of the OOCH contains the methodology for calculating the total dose including direct radiation as required by Surveillance Technical Specifications 4.11.4.1 and 4.11.4.2.
The methodology described
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in Section 472 for calculating total dose was determined to be acceptable.
Environmental Monitorina Proaram Section 5.0 of the 00CM identifies specific parameters of distance and the direction sector from the site and additional information for most of the samples identified in Environmental Monitoring Table 3.12-1.1 for WNP-2. No direct radiation stations were found in the 6 - 9 km (3.73 -
5.59 mi) range in the NE and SE sectors or in the 9 - 12 km (5.59 -
7.46 mi) range in the S sector as required by Tech 1; cal Specification Table 3.12-1.1.
Both figures containing maps which show the locations of the monitoring samples are labeled Figure 5-1.
Figur0 5-1, showing the monitoring stations inside the ten mile radius is unreadzble and must be replaced.
Summary In summary, the Licensee's 00CM uses documented and approved methods that are generally consistent with the methodology and guidance in NUREG-0133 and Regulatory Guide 1.109. However, because of the discrepancies identified in this review, it is recommended that the NRC request another revision to address the discrepancies.
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CONCLUSIONS The licensee's ODCM for WNP-2, updated through Amentient 4, was It was determined that the ODCM uses methods that are, in reviewed.
general, consistent with the guidelines of Regulatory Guide 1.109, Revision 1 and NUREG-0133 and are acceptable for demonstrating compliance However, it is with the radiological effluent technical specifications.
recommended that a revision to the 00CM be submitted to address the discrepancies identified in the review.
The following are considered to be major discrepancies:
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In Section 2.4, the parameter T) in Equation 5 is defined th release period "over which Cg) and as the length of the 1 Fj are averaged for liquid releases". The value of Cg1 is not averaged over T), it is determined from the pre-release grab sample.
L In Section 2.4, the definition of the parameter, F, in 1
Equation 5 includes a factor of 100 times the average flow from According to NUREG-0133 the product of the the site structure.
average blowdown flow to the receiving water body and an applicable factor can be 1000 cfs or less.
3 Methodology describing projection of doses to determine when the Ventilation Exhaust Treatment System shall be operated is not contained in the 00CM and hence is not in compliance with Technical Specification 3.11.2.5.
Tables 3-5a through 3-5d contain dose parameters for all l
The reviewer age groups and all pathways for the maximum organ.
f was unable to reproduce the dose parameters for the milk (cow).
milk (goat), vegetable, and meat pathways.
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The inhalation dose parameter for Cs-137 in Table 3-5c should 3
be 1.1E+6 instead of 5.5E+5 mrem-m /gCi-yr.
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The inhalation dose parameter value for Ce-144 in Table 3-5d 3
should be 7.8E+6 instead of 8.6E+7 mrem m /pci-yr.
The following are additional discrepancies:
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In the first paragraph of Section 2.1. of tha ODCM it states that at "a theoretical, continuous concentration of radionuclides at the 10 CFR 20 limits at the point of discharge to the Columbia River will result in compliance with 10 CFR 50, Appendix I limits in the unrestricted areas." This is not true, since a continuous release of a single radionuclide at the MPC limit will result in an annual dose for in excess of the annual dose limit identified in 10 CFR 50, Appendix I.
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In Section 2.4 in the definition of the parameter, Cjj, in Equation 5, it is not clear if the concentration is the diluted or the undiluted concentration.
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In Secticn 2.4 of the OOCM, the dilution factor, D, used w
in Equation 7 h defined as the dilution factor from the near field area to the nearest potable water intake and is given as 200. A reference must be included to support or justify the value.
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In Section 2.7d, for the definition of P, the effective "surface density" for soil, the reference should be Table E-15 instead of Table E-1 in Regulatory Guide 1.109, Revision 1.
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In Section 2.8 of the ODCM, it is not clear why the Licensee recalculates the total allowed activity limit for an outside temporary tank containing liquid radwaste to ensure that in the event of the tank rupture the concentration limits will not be exceeded at the WNP-1 well.
It appears a simple demonstration of
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compliance to Technical Specification 3.11.1.4 could be made by multiplying the radioactivity concentration of the liquid in the tank by the volume of liquid in the tank and comparing the resultant activity to the 10 curie limit identified in the ODCM.
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In t'he first paragraph of Section 3.4.2, tritium is specifically omitted as a radioactive contributor to the dose which is inconsistent with the surveillance requirements of Technical Specification 4.11.2.3.
/8 Simplified flow diagrams of the liquid, gaseous, and solid radwaste treatment systems are missing from the 00CM and must be included.
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In Section 3.6.2 of the 00CM, the symbols Cjj and C j in T
Equations 22, 24 and 25 are interchangeably defined as representing both measured and calculated concentration values for gaseous radionuclides. Different symbols should be used to represent the measured and calculated values of Cjj and Cyj.
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In Section 3.6.2 of the OOCH the sampling and analysis of gaseous effluents is said to be performed monthly, however, Table 4.11-2 of the technical specifications states that "Minimum Analysis Frequency" must be weekly. The ODCM is therefore not consistent with the requirements of the technical specifications.
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The map in Figure 5-1, which shows the environmental monitoring locations inside the ten mile radius, is unreadable and must be replaced.
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In the radiological environmental monitoring program plan there are no direct radiation monitoring locations found in the 6
- 9 km (3.73 - 5.59 mi) range in the NE and SE sectors nor in the 9 - 12 km (5.59 - 7.46 mi) range in the S sector as required by Technical Specification Table 3.12-1.1.
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The two figures containing maps showing the locations of the monitoring samples are both labe' led Figure 5-1.
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In Sections 3.3.1 and 3.3.2 of the ODCM, the letter "m" in Equa'tions 1 and 2 is used both for the number of nuclides on the summation sign and also to indicate a mixed mode release.
Different symbols should be used to eliminate tne confusion..
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REFERENCES 1.
Title 10, Code of Federal Reaulations, Part 50, Appendix I, "Numerical Guides 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 Ef fl uent s. "
2.
Letter from C. M. Powers (WPPSS) to J. B. Martin (NRC),
Subject:
Nuclear Plant No. 2 Semi-Annual Effluent Report, January 1,1986 to June 30, 1986, August 20, 1986.
3.
"Radiological Effluent Technical Specifications for Boiling Water Reactors," Rev. 3, Oraft 7", intended for contractor guidance in reviewing RETS proposals for operating reactors, NUREG-0473, September 1982.
4.
"Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, A Guidance Manual for Users of Standard Technical Specifications," NUREG-0133, October 1978.
5.
General Contents of the Offsite Oose Calculation Manual," Revision 1 Branch Technical Position, Radiological Assessment Branch, NRC, February 8, 1979.
6.
Calculation of Annual Doses of Evaluating Compliance with 10 CFR 50, Appendix I," Regulatory Guide 1.109, Rev.1, October 1977.
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