Resolution of Generic Safety Issues: Task III.D.2: Public Radiation Protection Improvement (Rev. 4) ( NUREG-0933, Main Report with Supplements 1–35 )
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The objective of this task was to improve public radiation protection in the event of a nuclear power plant accident by improving (1) radioactive effluent monitoring, (2) the dose analysis for accidental releases of radioiodine, tritium, and carbon-14, (3) the control of radioactivity released into the liquid pathway, (4) the measurement of offsite radiation doses, and (5) the ability to rapidly determine offsite doses from radioactivity release by meteorological and hydrological measurements so that population-protection decisions can be made appropriately.
ITEM III.D.2.1: RADIOLOGICAL MONITORING OF EFFLUENTS
The three parts of this item were combined and evaluated together.
DESCRIPTION
Historical Background
This Three Mile Island (TMI) Action Plan48 item required development and implementation of acceptance criteria for monitors used to evaluate effluent releases under accident and postaccident conditions. Criteria were to be developed for pathways to be monitored (stack, plant vent, steam dump vents) as well as for monitoring instrumentation. This was seen to encompass the requirements in Recommendation 2.1.8-b of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," issued July 1979,57 and Appendix 2 to NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants."224
Liquid effluents were not envisioned as posing a major release pathway because licensees typically had installed, or were installing, adequate storage capacity to prevent discharges. Consequently, existing liquid effluent monitoring systems were considered to be adequate.
Safety Significance
This issue had no impact on core-melt accident frequency.
Possible Solution
The envisioned monitoring system would provide automatic online analysis of airborne effluents, including isotopic analyses of particulate, radioiodine, and gas samples. To prevent saturation of detectors, an automatic sample cartridge changeout feature would be included. The system would include microprocessor control and real-time readouts and would be located in a low postaccident background area. The sampling system would be designed to provide a representative sample under anticipated accident release conditions.
A pressurized-water reactor (PWR) steam-dump sampling and monitoring system would be provided for PWR safety relief and vent valves. Such a system might consist of a noble gas monitor and a radioiodine sampling and monitoring system. The features of such a system would be similar to the above-described airborne effluent monitor with two notable differences: (1) the system would be required to function in a very high humidity (steam-air mixture) environment, and (2) operation would only be required during actual steam venting. Because such venting is usually of a short-term or intermittent duration, the monitoring system activation could be keyed to the opening of the vents.
PRIORITY DETERMINATION
Assumptions
It was assumed that improved radiological monitoring of airborne effluent would result in a reduction of public risk. The following section presents the U.S. Nuclear Regulatory Commission (NRC) staff analysis for prioritizing this issue, which was performed in 1998. This analysis, which includes frequency, consequence, and cost estimates and a value/impact assessment, has not been updated in the 2011 revision of this issue.
Frequency/Consequence Estimate
The magnitude of public risk reduction attributable to improved radiological monitoring of airborne effluents was not certain, but it was estimated by Pacific Northwest Laboratory (PNL)64 to range from 0 to 1 percent, based on the following logic.
Existing radiological monitoring requirements, as contained in NUREG-0737, "Clarification of TMI Acton Plan Requirements,"98 require real-time noble gas monitoring with sampling and laboratory analysis capabilities for radioiodines and particulates. Design-basis conditions defined in NUREG-073798 (100 microcuries per cubic centimeter radioiodines and particulates, 30-minute sample time) indicated that sample collection devices would pose special handling and analysis problems due to very high radioactivity buildup. Consequently, licensees typically provided alternate sample collection and analysis procedures. Execution of those procedures was estimated to require between 2 and 3 hours. During this time, radioiodine and particulate releases would be estimated based on computer-modeled interpretation of noble gas monitor readings, or on previous postaccident containment atmosphere analysis results, if such results were available. Public protective action recommendations would be made based on modeled estimates rather than actual effluent data. It was assumed that these recommendations would err on the conservative side (e.g., evacuate when not really required), due to the conservatism built into the modeled source terms for radioiodine and particulate releases.
Requiring licensees to have more sophisticated airborne effluent monitors would reduce the time required to obtain actual radioiodine and particulate release data to 15 minutes and essentially eliminate reliance on conservative theoretical release models extrapolated from noble gas monitor readings. As projected by the possible solution, real-time isotopic monitoring would save nearly 2 hours in arriving at realistic protective action recommendations based on actual releases.
Under these circumstances, the public risk reduction would be directly attributed to the decrease in public radiation exposure that would result from a more rapid assessment of the radioactive releases (about a 2-hour savings in analysis time). In addition, public risk may be reduced as a result of nonevacuation. The need for evacuation (presumed to exist if release knowledge was based only on noble gas monitor data) could be eliminated as a result of better knowledge of the isotopic releases. Nonevacuation would result in fewer evacuation-related risks (e.g., traffic accidents), the avoidance of which may outweigh the radiation exposure received. However, it was assumed that the public risk reduction would result primarily from the first effect (decrease in exposure due to more rapid assessment).
While protective actions can be recommended based on effluent releases in progress, the probability for a core-melt scenario was such that actions would be recommended based on anticipated releases, before the actual releases themselves. Under this assumption, monitoring effluent releases would have little or no impact on public risk and would be mainly for confirmation and quantification. This possible solution would not impact core-melt accident frequency.
At the time of this evaluation, there were 134 plants affected by the issue: 71 operating (47 PWRs and 24 boiling-water reactors (BWRs)) and 63 planned (43 PWRs and 20 BWRs). It was assumed that the average remaining plant life was 27.4 years for the 44 BWRs and 28.8 years for the 90 PWRs. The dose factors for PWR Release Categories 1 through 7 and BWR Release Categories 1 through 4 were assumed to be affected by the possible solution. From NUREG/CR-2800, "Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development,"64 a 1-percent decrease in the dose factors resulted in an estimated total public risk reduction of 8,500 man-rem for all plants. Assuming a decrease in the dose factors of 0.5 percent for this issue, the estimated public risk reduction was 4,250 man-rem.
Cost Estimate
Industry Cost: The industry cost for equipment development, installation, support facilities, and construction was estimated to be $600,000/plant. Development of procedures, software, and calibration for the equipment was estimated to require 16 man-weeks of effort, with an additional 4 man-weeks for the initial training of all licensee operators and health physics personnel. This was estimated to add $45,400/plant to the implementation cost. Based on an estimated cost of $645,000/plant for labor and equipment, the industry cost for implementing the possible solution was (134 plants)($645,000/plant) or $86.5 million (M).
The recurring industry operation and maintenance costs were estimated at 2 man-weeks/plant-year for retraining, 1 man-week/plant-year for calibration, and a reduction of 1 man-week/plant-year (reduced laboratory analyses due to a fully automated system) for a net increase of 2 man-weeks/plant-year, or an increased cost of $4,540/plant-year. As a result, industry costs for labor and material associated with operation and maintenance of the possible solution were estimated to be $17.2M.
Thus, the total industry cost associated with this issue was $(86.5 + 17.2)M or $103.7M.
NRC Cost: The NRC cost was assumed to be limited to implementation costs for development and plant installation. Because it was assumed that the new radiological monitoring systems would require no periodic inspection effort beyond that required for current systems, no additional NRC operation cost was envisioned. The NRC development cost included 1.5 man-years and $200,000 for research, criteria development, and engineering development, for a total cost of $350,000. The NRC administrative and technical effort associated with the review and approval of licensee submittals was estimated at 0.3 man-week/plant for a total cost of $91,000 for all plants. Therefore, the total NRC cost associated with this issue was $441,000.
Total Cost: The total industry and NRC cost associated with the possible solution was $(103.7 + 0.441)M or $104.1M.
Value/Impact Assessment
Based on an estimated public risk reduction of 4,250 man-rem and a cost of $104.1M for a possible solution, the value/impact score was given by the following:
Other Considerations
It was anticipated that improvement of radiological monitoring of airborne effluents would have no significant impact on occupational risk. The dose required to install equipment would probably not exceed 0.5 man-rem, which was negligible compared to the typical 600 man-rem/year required to operate a plant. Minor man-rem savings might occur under accident conditions due to better direction of field survey teams; however, such savings would be negligible compared to the 19,900 man-rem total associated with response and cleanup following an accident.
Based on an estimated occupational dose of 0.5 man-rem/plant for implementation of the possible solution in 71 operating plants, the total risk increase was 36 man-rem for all plants. Inclusion of this factor into the above calculation would reduce the value/impact score.
There was no accident avoidance cost for the resolution of this issue because improved radiological effluent monitoring systems would have no impact on accident frequency or cleanup and refurbishing costs.
CONCLUSION
Based on the risk reduction potential and value/impact score, the issue was given a LOW priority ranking (see Appendix C) in November 1983. NUREG/CR-5382, "Screening of Generic Safety Issues for License Renewal Considerations," issued December 1991,1563 concluded that consideration of a 20-year license renewal period could change the ranking of the issue to medium priority. Further prioritization in 1995, using the conversion factor of $2,000/man-rem approved1689 by the Commission in September 1995, resulted in an impact/value ratio (R) of $24,390/man-rem, which did not change the priority ranking. In 2010, the staff reviewed three parts of this issue in accordance with the SRM 871021A, "Staff Requirements—Briefing on Status of Unresolved Safety/Generic Issues," dated November 6, 1987,1980 which directed the staff to conduct periodic reviews of existing LOW-priority issues to determine whether any new information was available that would necessitate reassessment of the original prioritization evaluations.1964 Based on this review, the status of these issues was changed as described below.
ITEM III.D.2.1(1): EVALUATE THE FEASIBILITY AND PERFORM A VALUE/IMPACT ANALYSIS OF MODIFYING EFFLUENT-MONITORING DESIGN CRITERIA
The overall objective of this issue, which "is to provide assurance that all possible accident effluent-release pathways are monitored and that monitors will perform properly under accident conditions," is covered by General Design Criterion (GDC) 64, "Monitoring Radioactivity Releases," of Appendix A, "General Design Criteria for Nuclear Power Plants," to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities." GDC 64 states that "Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents." Moreover, 10 CFR 50.34(f)(2)(xvii)(E) establishes the requirement for monitoring noble gas effluents and continuous sampling of radioactive iodines and particulates in gaseous effluents. According to this part of the regulation, "each applicant for a light-water-reactor construction permit or manufacturing license whose application was pending as of February 16, 1982," in addition to "each applicant for a design certification, design approval, combined license, or manufacturing license under part 52" of 10 CFR, needs to "Provide instrumentation to measure, record and readout in the control room:…(E) noble gas effluents at all potential, accident release points. Provide for continuous sampling of radioactive iodines and particulates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples." Finally, 10 CFR 50.34(f)(2)(xxvii) and (2)(xxviii) establish requirements for monitoring of inplant radiation and airborne radioactivity for a broad range of routine and accident conditions and for evaluating potential pathways for radioactivity and radiation that may lead to control room habitability problems under accident conditions.
In addition to the regulations stated above, Section 11.5, "Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems," of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (the SRP),11 states that "Provisions should be made for the installation of instrumentation and monitoring equipment and/or sampling and analyses of all normal and potential effluent pathways for release of radioactive materials to the environment, including nonradioactive systems that could become radioactive through interfaces with radioactive systems." Table 1 of Section 11.5 of the SRP11 specifies the gaseous streams or effluent release points that should be monitored and sampled. In addition, for monitoring the effluents during a postulated event, Section 11.5 of the SRP11 states that "Provisions should be made for monitoring instrumentation, sampling, and sample analyses for all identified gaseous effluent release paths in the event of a postulated accident."
As explained earlier, implementation of the proposed solutions has no impact on the core-melt accident frequency. Moreover, "while protective actions can be recommended based on effluent releases in progress, the probability for a core-melt scenario was such that actions would be recommended based on anticipated releases prior to the actual release themselves. Under this assumption, monitoring effluent releases would have little or no impact on public risk and would be mainly for confirmation and quantification."
Specific requirements related to some of the factors in the proposed design criteria mentioned in NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," have not been established; however, based on the review of the NRC’s regulations presented above, the staff concluded that the overall objectives of Item III.D.2.1(1) are met by the existing regulations. Moreover, the low safety significance of the issue does not warrant further actions to evaluate and implement the proposed solutions. Therefore, the staff changed the status of Item III.D.2.1(1) and DROPPED this item from further pursuit.1964
ITEM III.D.2.1(2): STUDY THE FEASIBILITY OF REQUIRING THE DEVELOPMENT OF EFFECTIVE MEANS FOR MONITORING AND SAMPLING NOBLE GASES AND RADIOIODINE RELEASED TO THE ATMOSPHERE
In addition to Criterion 64 of Appendix A to 10 CFR Part 50, the regulation at 10 CFR 50.34(f)(2)(xvii) establishes the requirement for monitoring noble gas effluents and continuous sampling of radioactive iodines and particulates in gaseous effluents. According to this part of the regulation, "each applicant for a light-water reactor construction permit or manufacturing license whose application was pending as of February 16, 1982," in addition to "each applicant for a design certification, design approval, combined license, or manufacturing license under part 52" of 10 CFR, needs to "Provide instrumentation to measure, record and readout in the control room:…(E) noble gas effluents at all potential, accident release points. Provide for continuous sampling of radioactive iodines and particulates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples."
Based on the review of the NRC regulations related to this issue presented above and the low safety significance of this issue, the staff concluded that Item III.D.2.1(2) is adequately addressed by the existing regulations. Therefore, the staff changed the status of Item III.D.2.1(2) and DROPPED this item from further pursuit.1964
ITEM III.D.2.1(3): REVISE REGULATORY GUIDES
NUREG-066048 called for this issue to "revise Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, Standard Review Plan Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems, and further revise Regulatory Guide 1.97, as necessary." All of these documents have been updated since the issuance of NUREG-0660.48 Some specific factors of the design criteria mentioned in NUREG-066048 have not been included in these updates. However, the overall objective of the issue has been thoroughly addressed in these updates. As of April 2010, the latest revision of each document is available as follows: Regulatory Guide (RG) 1.21, Revision 2, "Measuring, Evaluating, and Reporting Radioactive Material in Liquid and Gaseous Effluents and Solid Waste," issued June 20091968; SRP11 Section 11.5, issued March 2007; and RG 1.97, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants," Revision 4, issued June 2006.55
Because of the revisions made to RG 1.21,1968 SRP11 Section 11.5, and RG 1.97,55 the staff changed the status of Item III.D.2.1(3) and DROPPED this item from further pursuit.1964
ITEM III.D.2.2: RADIOIODINE, CARBON-14, AND TRITIUM PATHWAY DOSE ANALYSIS
The four parts of this item were combined and evaluated together.
DESCRIPTION
Historical Background
This TMI Action Plan48 item addressed the issue of further research for improving the understanding of radioiodine partitioning in nuclear power reactors and of the environmental behavior of radioiodine, carbon-14, and tritium following an accident and during normal operation.
Iodine isotopes are considered to be major contributors to the occupational and public dose during a loss-of-coolant accident, along with noble gases and fission products. A study in these areas was documented in NUREG-0772, "Technical Bases for Estimating Fission Product Behavior during LWR Accidents," issued June 1981,212 with the following major conclusions: (1) uncertainties in predicting atmospheric release source terms were very large (at least a factor of 10), (2) source terms for certain accident sequences may have been overestimated in past studies; e.g., WASH-1400 (NUREG-75/014), "Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," issued October 1975,16 and (3) cesium iodide should be the predominant chemical form of iodine under severe accident conditions.
Safety Significance
The above conclusions indicated that the methodology and assumptions used for evaluating radioiodine release could result in unrealistic estimates (e.g., RG 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors,"213 and RG 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors"214). Also indicated was that more research in aerosol behavior and fission product chemistry was needed in order to improve and support the calculation methodology concerned with radioiodine partitioning, fission product behavior, and related topics.
Possible Solution
The NRC assumed that further study would improve the understanding of this issue and result in more realistic assumptions and methods for evaluating source terms, releases, and the environmental behavior of radioiodine, carbon-14, and tritium following an accident. This research would not affect accident frequencies at nuclear power plants. However, the NRC assumed that the results of these studies would be used to revise the SRP11 and RGs.
The NRC also assumed that the RG revisions could result in reducing the size of existing emergency planning zones from a 10-mile radius to a 2-mile radius. This assumption was based on a reduction of source terms in a core-melt accident by a factor of 10. This would result in reducing dose concentration at a particular distance from the nuclear reactor also by a factor of 10. Assuming neutral weather conditions with a 30-meter-high plume, the offsite dose predicted at 2 miles from the accident scene, using the reduced source term assumption, would be the same as that predicted at 10 miles from the reactor.
CONCLUSION
The study of radioiodine, carbon-14, and tritium behavior at Three Mile Island Unit 2 (TMI-2) called for in Item III.D.2.2(1) was completed in June 1981 and documented in NUREG-0771, "Regulatory Impact of Nuclear Reactor Accident Source Term Assumptions," issued June 1981,455 and NUREG-0772.212 Items III.D.2.2(2), (3), and (4) called for a series of studies and evaluations of various radionuclide pathways and models followed, if necessary, by revisions to several SRP11 sections and RGs. As part of the staff’s task to prepare and publish a manual (referred to as the "Offsite Dose Calculation Manual"598) to be used by the NRC and industry to estimate individual and population doses during normal and accident conditions, Items III.D.2.2(2), (3), and (4) were assessed. This Offsite Dose Calculation Manual was prepared under Item III.D.2.5 and fully described each of the theoretical models used to predict radionuclide transport.149 Thus, Items III.D.2.2(2), (3), and (4) were covered under Item III.D.2.5.
ITEM III.D.2.2(1): PERFORM STUDY OF RADIOIODINE, CARBON-14, AND TRITIUM BEHAVIOR
This item was evaluated in Item III.D.2.2 above and was RESOLVED with no new requirements.
ITEM III.D.2.2(2): EVALUATE DATA COLLECTED AT QUAD CITIES
This item was evaluated in Item III.D.2.2 above and was determined to be covered in Item III.D.2.5.
ITEM III.D.2.2.(3): DETERMINE THE DISTRIBUTION OF THE CHEMICAL SPECIES OF RADIOIODINE IN AIR-WATER-STEAM MIXTURES
This item was evaluated in Item III.D.2.2 above and was determined to be covered in Item III.D.2.5.
ITEM III.D.2.2.(4): REVISE SRP AND REGULATORY GUIDES
This item was evaluated in Item III.D.2.2 above and was determined to be covered in Item III.D.2.5.
ITEM III.D.2.3: LIQUID PATHWAY RADIOLOGICAL CONTROL
The four parts of this item were combined and evaluated together.
DESCRIPTION
This TMI Action Plan48 item was concerned with improving public radiation protection in the event of a nuclear power plant accident by improving the control of radioactivity released into the liquid pathway. This control could be accomplished by the application of various interdictive measures at the source of the release and/or along the liquid pathway. Techniques were developed and were being used to evaluate the liquid pathway effects of an accident for each reactor site. Sites that might require interdictive measures related to liquid pathway releases were to be determined. Interdictive measures were to be assessed as to their effectiveness in improving public radiation protection.
CONCLUSION
A liquid pathway analysis for Zion Nuclear Power Station was completed by the Office of Nuclear Reactor Regulation’s Division of Engineering in 1980.391 In addition, a liquid pathway analysis was performed for the Indian Point nuclear power plant. Both analyses were used in NUREG-0850, "Preliminary Assessment of Core Melt Accidents at the Zion and Indian Point Nuclear Power Plants and Strategies for Mitigating Their Effects," issued November 1981.390 A simplified analysis for liquid pathway studies (NUREG-1054, "Simplified Analysis for Liquid Pathway Studies,")658 was published in August 1984, and Section 7.1.1 of NUREG-0555, "Environmental Standard Review Plans for the Environmental Review of Construction Permit Applications for Nuclear Power Plants" (the ESRP), issued May 1979,464 was drafted with no new requirements for licensees or applicants.659, 660 ESRP Section 7.1.1 was finally published as NUREG-1165, "Environmental Standard Review Plan for ES Section 7.1.1,"838 in November 1985. Thus, this item was RESOLVED and no new requirements were established.799
ITEM III.D.2.3(1):DEVELOP PROCEDURES TO DISCRIMINATE BETWEEN SITES/PLANTS
This item was evaluated in Item III.D.2.3 above and was RESOLVED with no new requirements.799
ITEM III.D.2.3(2): DISCRIMINATE BETWEEN SITES AND PLANTS THAT REQUIRE CONSIDERATION OF LIQUID PATHWAY INTERDICTION TECHNIQUES
This item was evaluated in Item III.D.2.3 above and was RESOLVED with no new requirements.799
ITEM III.D.2.3(3): ESTABLISH FEASIBLE METHOD OF PATHWAY INTERDICTION
This item was evaluated in Item III.D.2.3 above and was RESOLVED with no new requirements.799
ITEM III.D.2.3(4): PREPARE A SUMMARY ASSESSMENT
This item was evaluated in Item III.D.2.3 above and was RESOLVED with no new requirements.799
ITEM III.D.2.4: OFFSITE DOSE MEASUREMENTS
ITEM III.D.2.4(1):STUDY FEASIBILITY OF ENVIRONMENTAL MONITORS
DESCRIPTION
This TMI Action Plan48 item called for the staff to study the feasibility of environmental monitors capable of measuring real-time rates of exposure to noble gases and radioiodines. Monitors or samplers capable of measuring respirable concentrations of radionuclides and particulates were also considered. This activity supported proposed revisions to RG 1.9755 (see Item II.F.3).
CONCLUSION
The establishment of guidance in RG 1.9755 for fixed monitors to detect unidentified releases was postponed pending the outcome of a feasibility study that was completed in April 1982.188 Using this study as a basis, the staff concluded that environmental monitors of this nature were not practical and that proposed requirements for these monitors should be dropped from consideration.189 Thus, all required action on this item was completed382 and the issue was RESOLVED with no new requirements.
ITEM III.D.2.4(2): PLACE 50 TLDs AROUND EACH SITE
DESCRIPTION
This TMI Action Plan48 item called for Office of Inspection and Enforcement (OIE) to place 50 thermo-luminescent dosimeters (TLDs) around each site in coordination with States and utilities. During normal operation, OIE quarterly reports from these dosimeters were to be provided to NRC, State, and Federal organizations. In the event of an accident, the dosimeters could then be read at a frequency appropriate to the needs of the situation.
The specific objectives of this program were to (1) establish preoperational, historical, baseline radiation dose levels, whenever possible, for each monitored facility, (2) provide ongoing radiation dosimetry data during routine operations, (3) provide postaccident radiation dosimetry to aid in assessment of population exposures and radiological impact, (4) allow for independent verification of the adequacy of NRC licensees’ environmental radiation monitoring programs, (5) provide uniform treatment of dosimeters with respect to handling, shipping, calibrating, reading, and data processing for all monitored facilities in the United States, and (6) provide uniform, consistent environmental radiation monitoring data for use by the Congress, Federal and State agencies, monitored facilities, and the public.
This item addressed improvements in the NRC capability to make independent assessments of safety and, therefore, was considered to be a licensing issue.
CONCLUSION
OIE completed installation of TLDs at all operating reactors in August 1980 in accordance with the TMI Action Plan schedule. A direct radiation monitoring network was established and a program for routine reporting began. The completion of these activities was described in an OIE memorandum.236 With the establishment of the NRC TLD direct radiation monitoring network, the installation of TLDs at all operating reactor sites, and the routine reporting of the TLD measurements, all work required by this item was completed.236, 379 Thus, this licensing issue was resolved.
ITEM III.D.2.5: OFFSITE DOSE CALCULATION MANUAL
DESCRIPTION
Historical Background
This TMI Action Plan48 item called for the Office of Nuclear Reactor Regulation to prepare a manual to be used by the NRC and plant personnel to estimate the maximum individual doses and population doses during an accident.
Safety Significance
This issue did not affect core-melt frequency or the amount of radioactivity released. Instead, it was intended to reduce the consequences of a major release by assuring that licensees have a rapid and sufficiently accurate method of estimating dose, and that communication between licensees and the NRC be expedited by a common standard calculation method used by both.
Possible Solution
The proposed manual was expected to include formulations with which to combine source term and meteorological measurements. This would determine offsite dose rates in a manner that would be standard among all parties making decisions on public protection and emergency response. Appendix 2 to NUREG-0654224 established criteria for automated assessment of radiation doses in the event of an accident.
PRIORITY DETERMINATION
Frequency Estimate
Because the proposed solution to the issue did not affect core-melt accident frequency, the frequencies for the various release categories given for Oconee Nuclear Station, Unit 3, and Grand Gulf Nuclear Station, Unit 1, were used unchanged in the value/impact calculation.
Consequence Estimate
In an assessment64 of this issue, PNL experts judged that a 1-percent reduction in public dose (man-rem) might be expected as a result of having an offsite dose calculation manual available. It was estimated that the changes in consequences would be much less (0.01 percent to 0.1 percent). Because all sequences would be affected and the risk from both PWRs and BWRs was about 210 to 250 man-rem/reactor-year (RY), the risk reduction was estimated to be 0.02 to 0.2 man-rem/RY.
At the time of the evaluation of this issue in November 1983, there were 43 PWRs and 27 BWRs operating, with cumulative experience of 350 RY and 260 RY, respectively. Considering the 36 PWRs and 21 BWRs that were under construction and assuming a plant life of 40 years, there were 2,810 PWR-years and 1,660 BWR-years in the future, for a total of 4,470 RY. Therefore, the total risk reduction associated with this issue was (0.2)(4,470)man-rem or 894 man-rem.
Cost Estimate
Industry Cost: For licensees, 4 man-weeks of training for implementation were assumed, since operators were being retrained periodically and this retraining could include dose calculation methods. This different method would not incur additional recurring costs. Thus, the total industry cost was estimated to be $7,700/plant or $0.98M for 127 plants.
NRC Cost: The NRC had already completed work on development of a portable computerized system for dose calculations to be used by the NRC regional offices. This was part of the program for NUREG-0654.224 This program was developed to the point of field trials for the computerized system. Based on the development costs, an additional $125,000 to develop this package into a manual form for use by utilities was assumed. It was estimated that NRC site representatives could spend a minimal amount of time (about 2 days) to evaluate initial utility performance with the package. This was estimated to be $600/plant. Thus, the total NRC cost was approximately $200,000 for all plants.
Total Cost: The total industry and NRC cost associated with the possible solution was $(0.98 + 0.2)M or approximately $1.2M.
Value/Impact Assessment
Based on an estimated public risk reduction of 894 man-rem and a cost of $1.2M for a possible solution, the value/impact score was given by the following:
CONCLUSION
Based on the above value/impact score, the issue would have had a MEDIUM priority ranking (see Appendix C). However, before approval of the prioritization evaluation in November 1983, the Offsite Dose Calculation Manual was published as NUREG/CR-3332, "Radiological Assessment—A Textbook on Environmental Dose Analysis,"599 in September 1983. Thus, the issue was RESOLVED and no new requirements were issued.598
ITEM III.D.2.6: INDEPENDENT RADIOLOGICAL MEASUREMENTS
DESCRIPTION
This TMI Action Plan48 item dealt with independent radiological measurements; i.e., means of collecting data independent of licensees’ programs. An OIE task force developed a plan and requirements for upgrading the capability of regional offices to perform independent radiological measurements during routine inspections and emergency response operations. The objective of the upgrade was to achieve consistent capability among the regional offices, including standardization in major equipment items such as mobile laboratory vans, gamma spectrum analysis equipment, radiation survey instrumentation, and air-sampling and monitoring devices.
Based on the recommendations of the task force, each region was equipped with complete mobile laboratories.235 In some cases, this represented upgrading certain equipment or purchasing new equipment. This action item required that revisions be made to the inspection program to include the upgrading of the independent radiological measurements. The program was included in the routine OIE program for review and revision of the inspection program. As new equipment needs were identified, the program was to be revised and the equipment acquired.
This item addressed improvements in the NRC capability to make independent assessments of safety and, therefore, was considered to be a licensing issue.
CONCLUSION
With the upgrading of independent radiological measurements and the implementation of other recommendations made by the task force, all work required by this item was completed.235, 379 Thus, this licensing issue was resolved.
REFERENCES
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