ML20087C772
ML20087C772 | |
Person / Time | |
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Site: | Harris |
Issue date: | 03/08/1984 |
From: | Shearin R, Wilkie W CAROLINA POWER & LIGHT CO. |
To: | |
Shared Package | |
ML20087C753 | List: |
References | |
OL, NUDOCS 8403130164 | |
Download: ML20087C772 (122) | |
Text
{{#Wiki_filter:. y UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
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CAROLINA POWER & LIGHT COMPANY) Docket Nos. 50-400 OL and NORTH CAROLINA EASTERN ) 50-401 OL MUNICIPAL POWER AGENCY )
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(Shearon Harris Nuclear Power ) Plant, Unita 1 and 2) ) AFFIDAVIT OF DR. WILLIAM H. WILKIE AND RONALD L. SHEARIN IN SUPPORT OF APPLICANTS' MOTION FOR
SUMMARY
DISPOSITION OF JOINT INTERVENORS' CONTENTION VI County of Wake )
) SS:
State of North Carolina ) William H. Wilkie, being duly sworn according to law, de-poses and says:
- 1. I am employed by Carolina Power & Light Company (CP&L) as Principal Health Physics Specialist. My business address is Shearon Harris Energy & Environmental Center, Route 1, Box 327, New Hill, North Carolina 27562. A statement of my background and qualifications is affixed hereto as Attachment A. I have a i
Ph.D. degree in nuclear engineering from the Georgia Institute of Technology and extensive professional experience in the area of health physics program development in support of nuclear power operations. Therefore, I have personal knowledge of the matters stated herein and believe them to be true and correct. ( 8403130164 840309 PDR ADOCK 05000400 C PDR
II Ronald L. Shearin, being duly sworn according to law, de-poses and says:
- 2. I am employed by Carolina Power & Light Company as Project Specialist - Environmental for the Radiological & Chem-ical Support Section of the Operations Training & Technical Support Department. My duties have included the development and establishment of the Shearon Harris Radiological Environ-mental Monitoring Program including the selection and placement of sampling equipment, the establishment of sampling proce-dures, and the interpretation of environmental data releted to the radiological nature of the environment. My business ad-dress is Shearon Harris Energy & Environmental Center, Route 1, Box 327, New Hill, North Carolina 27562. A statement of my background and qualifications is affixed hereto as Attachment B. I have personal knowledge of the matters stated herein and believe them to be true and correct.
- 3. We make this affidavit in support of Applicants' Mo-tion for Surmary Disposition of Joint Intervenors' Contention VI. In this affidavit we will: (1) describe the in-plant area, airborne and effluent monitoring systems, and the envi-ronmental monitoring system; (2) show that these systems will provide timely and accurate information necessary to protect the public and plant staff in the event of an emergency; (3) explain the fallacy of Joint Intervenors' contention that
4 c. real-time identification and quantification of specific ra-dionuclides is necessary; and (4) explain why the Joint Inter-venors are incorrect in asserting that pressurized ionization chambers.can provide release rates by specific radionuclide and should be used'for monitoring effluents.
- 4. The Radiation Monitoring System (RMS) for the Shearon Harris Nuclear Power Plant ("SHNPP") consists of:
- a. an Area Radiation Monitoring System,
- b. an Airborne Radiation Monitoring System, and
- c. a Process and Effluent Radiological Monitoring and Sampling System The RMS is designed to meet all the applicable regulatory requirements of 10 C.F.R. Part 20, 10 C.F.R. Part 50, and Gen-eral Design Criteria for Nuclear Power Plants Numbers 60 and 64; and, to the extent practicable, is designed in accordance with the applicable recommendations of NRC Regulatory Guides 1.21, 1.45, 1.97, 1.109, 4.15 (with an exception on standards use), 8.2, and 8.8, NUREG-0472 (" Draft Radiological Effluent Technical Specifications for PWRs," March 1979), NUREG-0737
(" Classification of TMI Action Plan Requirements," November 1980), ANSI /ANS-HPSSC-6.8.1-1981 (" Location and Design Criteria for Area Radiation Monitoring Systems for Light Water Nuclear Reactors," 1981), and ANSI N13.10-1974 (" Specification and
e, 9 Performance of On-Site Instrumentation for Continuously Moni-E toring Radioactivity in Effluents," 1974). In addition, the safety-related portion of the system is designed in accordance
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with specifications in IEEE 279-1971 (" Criteria for Protection Systems for Nuclear Power Generating Stations," 1971), IEEE 308-1974 (" Criteria for Class lE Power Systems for Nuclear Power Generating Stations," 1974),-IEEE 323-1974 (" Qualifying Class lE Equipment for-Nuclear Power Generating Stations," 1974), IEEE 336-1971 (" Installation, Inspection, and Testing
. Requirements for Class lE Instrumentation and Electric Equip-ment at Nuclear Power Generating Stations," 1971), IEEE 344-1975 (" Practices for Seismic Qualifications of Class lE Equipment for Nuclear Power Generating Stations," 1975), and IEEE.384-1974 (" Criteria for Independence of Class lE Equipment
~ and' Circuits," 1974). The Area Radiation Monitoring System is described.in the SENPP Final Safety Analysis Report ("FSAR"), Sections 11.5.2.3 and 12.3.4.1. The Airborne Radiation Monitoring System is de-scribed in FSAR, Sections 11.5.2.3 and 12.3.4.2. The Process and Effluent Radiological Monitoring and Sampling System is de-scribed in FSAR Section 11.5. The above-referenced sections of the FSAR are affixed hereto as Attachment C and are incorpo-rated herein as part of this affidavit.
- 5. The in-plant health physics program includes surveil-
- lance to establish the radiation hazards. The Area Radiation 4
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Monitoring System supports-this program using strategically lo-cated detectors to monitor trends and sudden changes in gamma-radiation fields. The design objectives of the Area Radiation i. Monitoring System are:
- a. to furnish records of radiation exposure rates in specific areas of the plant;
- b. to warn of uncontrolled or inadvertent movement of radioactive material in the plant;
- c. to provide local and remote indication of ambient gamma radiation and local and remote alarms at key points where a substantial change in radiation fields might be of immediate impor-tance to personnel frequenting the area;
- d. to annunciate and warn of possible equipment malfunctions and leaks in specific areas of the plant; and
- e. to provide warnings of unacceptable radiation levels.
- 6. The detectors used in the Area Radiation Monitoring System have the capability of detecting all radionuclides that emit gamma radiation. Gamma radiations deliver all of the sig-nificant external whole body radiation doses to personnel in a nuclear power plant. It would not be useful to design area o
monitors to have the capability for identifying specific ra-
'dionuclides or to be sensitive to neutron, alpha, or beta radi-ation, for the following reasons:
- a. A gamma ray is fully characterized by its ener-gy, and the external radiation hazard from a gamma radiation field is determined completely by establishing the exposure rate. Knowledge of the specific nuclides contributing to tha gamma radiation in the field is irrelevant for as-sessing the external radiation hazard.
- b. Significant neutron radiation is present only within the containment building whenever the re-actor is operating. Hazards to personnel enter-ing containment during reactor operations are determined by surveys at specific locations using portable neutron detectors. This obviates the need for area monitors to detect these neutrons. Neutron fluxes throughout the con-tainment building are directly proportional to reactor power level which is monitored continu-ously with reactor vessel instrumentation.
- c. Alpha radiation poses no external exposure haz-ard because of its very short range (low-penetrating power). Therefore, area monitors sensitive to alpha radiation would provide no useful information.
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- d. Beta radiation can pose a hazard in a nuclear plant. However, this hazard exists only from unshielded radioactive materials and occurs pri-marily at locali.med sources (water leaks, open systems, dietssembled components such as pumps and valves, etc.). It is best evalucted using portable survey instruments and can be minimized with sound work practices and by the use of pro-tective clothing and eye shields. Experience has demonstrated that, when properly controlled, external beta radiation exposures are not a reg-ulatory limitation. It is not reasonable to use area monitors to evaluate beta hazards because of the relatively short range of beta radiation (several meters or less in air) and the local-Jzed sources. As with gamma radiation, detailed knowledge of the specific nuclides emitting the beta particles would provide little useful in-formation for identifying or controlling the ex-ternal beta hazard.
- 7. The Airborne Radiation Monitoring System is designed to support the in-plant health physics program by:
- a. informing operations personnel of airborne particulate, iodine, and gaseous activity trends in the various buildings and structures of the plant;
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- b. alarming on detection of any abnormal increases in the airborne activity;
- c. furnishing records of trends in airborne activi-ty in the various plant areas;
- d. alerting personnel to increases in airborne radioactivity concentrations that may indicate a need for changes in respiratory protection re-quirements;
- e. providino informAtion for evaluating the perfor-mance of all plant systems that function to min-imize the release of airborne radioactivity to accessible areas of the plant and to the envi-ronment;
- f. providing.the capability to alarm and initiate isolation of the normal ventilation systems and actuation of the emergency ventilation systems during postulated accidents and providing opera-tors with information regarding control of the ventilation systems in an emergency; i
- g. providing information for the purpose of reduc-ing radiation exposure via inhalation of air-borne particulates and iodine.
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o m? The Airborne Radiation Monitoring System is designed to provide qualitative information for trending and for alarming on detection of sudden increases in concentrations. Detectors I in-this system will monitor groups of radionuclides--noble I gases, particulates, iodines, or all airborne activity depend-i l ing en the detector design and location. It is neither neces-
- sary nor practical for this' system to have the capability for i
monitoring individual nuclides within these groups because the radiation hazard is determined by the nct effect of the compos-ite mixture. Should an alarm signal a high concentration that requires investigation, the air is sampled and is analyzed in a e laboratory to' provide quantitative data for specific nuclides. In emergency situations, however, this laboratory information on. specific airborne radionuclides is not needed in order to protect in-plant workers. In fact, if this detailed informa-tion were instantly available--as the Joint Intervenors contend
- that it should be--it would not provide emergency response per-sonnel with data that would cause them to take any actions dif-ferent from-those indicated by presently available data.
- 8. The Process and Effluent Radiological Monitoring and t
. Sampling System provides the means for monitoring the major liquid and gaseous paths by which radionuclides may be released to the environment, both during normal operating conditions and during emergency conditions. This system provides current and . historical information for indicating both the typos and rates
of change in radioactivity within process and effluent streams. Quantitative data are generated for gaseous and 'iquid effluents which track the activity by nuclide for all signifi-cant radionuclide releases. It is not reasonable to attempt to detset in real time all nuclides listed in 10 U.F.R. Part 20, Appendix B, Table 2, as the Joint Intervenors contend is neces-sary, because many of the nuclides listed are not present in effluent stresms, and most of the remaining nuclides on the list are insignificant for purposes of radiation protection.
- 9. The Process and Effluent Radiological Monitoring Sys-tem detects continuously all gamma emitters in: (1) liquid streams and (2) all noble gases, particulates, and iodines as i
groups of radionuclides in gaseous streams. Although the Pro-cess and Effluent Monitoring and Sampling System provides quan-titative data for each significant radionuclide in effluents, this detail is not useful for either in-plant or off-site emer-gency response personnel. As explained in paragraph 7, initial decisions regarding mitigation of radiological consequences of an accident can be based on knowledge of activity by group; i.e., noble gases, iodines, and particulates rather than on knowledge of detailed data within the groups. Nuclides within these groups have similar properties in plant systems, in the environment, and inside the human body. During the early phases of an emergency it is the plume exposure pathways, i.e. inhaletion and external doses from the plume, that are the
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controlling factors for emergency management decisions. Knowl-edge of the total activity of noble gases, iodines, and particulates being (or having the potential of being) released, along with information regarding release path and meteorologi-cal conditions is sufficient for emergency personnel to calcu-late potential doses from plume exposures for most accident scenarios. These calculations are made on the basis of conser-vative assumptions about the mix of the significant nuclides winnin each group. Release rates for each individual nuclide are calculated using these pre-defined mixes and measured total activity. When tnis figure is multiplied by atmospheric dis-persion factors and dose factors appropriate for each nuclide-pathway combination and than summed, the result is a determina-tion of dose rates. Sensitivity studies demonstrate that the calculated doses per unit activity released do not vary signif-icantly (more than a factor of two) over a wide range of possi-ble mixes and accident scenarios. If a release is occurring, emergency personnel usually will make field measurements to supplement calculated doses. Again, the accuracy of the field
. measurements for evaluating both inhalation and external expo-sure hazards will.not be influenced by a knowledge of the pre-t 'cise mix of nuclides being released.
- 10. A Radiological Environmental Monitoring Program has been established for SHNPP to provide measurements of radiation and radioactive materials in the SENPP site environs and to
shaw that in many cases no detectable contamination exists. The Radiological Environmental Monitoring Program will be con-ducted to verify the effectiveness of in-plant systems used to control the release of radioactive materials and to ensure that unanticipated buildups of radioactivity will not occur in the environment. Secondarily, the environmental monitoring program will identify the effects of releases of radioactivity from all release points whether these are monitored or not and is capa-ble of identifying each radionuclide released. The Radio-logical Environmental Monitoring Program is discussed in the SHNPP Environmental Report (*ER") at Section 6.1.5 (affixed hereto as Attachment D and incorporated herein) and meets the requirements of NRC Regulatory Guide 4.1, Revision 1, and NUREG-0472.
- 11. The SHNPP Radiological Environmental Monitoring Pro-gram will utilize six continuously operating air samplers (Ana-lytical Process Instruments, Inc., Model NRC-2000, Nuclear Air Samplers), three continuously operating water samplers (two Hy-dragard Automatic Liquid Samplers and one Brailsford Battery-Operated Liquid Sampler), and 41 thermoluminescent dosimeters.
I In addition to the continuously operating equipment, periodic samples of environmental media, food crops, and animal products in the SENPP environs are collected as described in the SENPP ER in Table 6.1.5-1. See Attachment D affixed hereto. The samples obtained will be analyzed using analytical procedures and analyses which meet or exceed the sensitivities specified in NUREG 0472.
- 12. In the event of an abnormal release of radioactivity, SHNPP Emergency Environmental Monitoring Field Teams are equipped with instrumentation adequate to provide an analysis of the radiological impact of any such release. The field teams are provided with radiation monitoring equipment with a full range of detection capability for radiation types and en-ergy levels and with air samplers capable of sampling for air-borne particulates-(high-volume air samplers), airborne iodines (low-volume air samplers with silver zeolite collection car-tridges), as well as portable generators to operate the air samplers. The field teams are also equipped with environmental sampling supplies and equipment to enable them to collect water, soil, vegetation, milk, food, and fodder crops. Each field team has the capability to collect and analyze airborne
-8 radlonuclides with a sensitivity of at least 5 x 10 micro l curies per cubic centimeter.
- 13. The field teams are supported by an environmental laboratory located at the Harris Energy & Environmental Center.
This laboratory has the capability to perform detailed analyses on environmental samples for radionuclide content with sensi-I tivities as set forth in the SENPP FSAR at Section 16.2, Table J3.12-2 affixed hereto as Attachment E and incorporated herein I l
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as part of this affidavit. In addition, the field teams are supported by a mobile laboratory with the same analytical capa-bilities of the. central laboratory. To the extent required, these facilities allow the detection of any specific ra-dionuclide in the environment. However, in general, the capa-bility quickly to identify specific radionuclides in the envi-ronment is not used for initial emergency response decisions but rather may be used to provide greater detail should follow-up actions be required. Technical specifications define the operating conditions for the plant, and 40 C.F.R. 9 190 specifies radiation dose limits to members of the public from uranium fuel cycle activities during nornal operation.
- 14. The Joint Intervenors have asserted that pressurized ionization chambers (" PICS") can and should be used to identify specific radionuclides and to establish the release rates of each nuclide. In response to Applicants' interrogatories, Joint Intervenors advised: "they need pressurized ionization monitors to do it as those are the only monitors we know that can rapidly detect specific radionuclides," and " pressurized ionization detectors can detect specific radionuclides being passed through them." " Joint Intervenors' Response to Staff Interrogatories," dated August 31, 1983, Response to Interroga-tory Nos. 9, 31. These statements demonstrate a lack of knowl-edge regarding the theory of operation and the applications of these devices. Radionuclides cannot be " passed through"
3 pressurized. ionization chambers, as the Joint Intervenors ap-pear to believe, because these chambers are sealed. PICS are not spectrometers, and it is impossible to use them to identify specific radionuclides. A PIC incorporates a sealed, pressur-ized chamber within which interactions of gamma radiation with the detection gas occur, producing electrically-charged ions that are collected by an electric field and measured as a cur-rent. This current is' converted to an equivalent exposure rate for readout. Ion pairs produced by gamma rays from one nuclide are identical to ion pairs produced by gamma rays from any other nuclide.
- 15. The Joint Intervenors further state that "[T]he con-tention is that the specific radionuclides and amounts released will not be detected." " Joint Intervenors' Response to Staff
, Interrogatories," dated August 31, 1983, Response to Interroga-tory-No. 29. This assertion is incorrect because the RMS and Radiological Environmental Monitoring Program are designed and will be operated in a manner such that all significant ra-dionuclides in both liquid and gaseous effluents will be ac-counted for, and the activities of each significant ra-dionuclide released and the time of release will be recorded as specified in Regulatory Guide 1.21 and NUREG-0472.
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- 16. In summary, it can be stated that the SHNPP RMS and t'he Radiological Environmental Monitoring Program are consis-tent with sound radiation protection procedures, state-of-the-art technology, current industry standard prac-tices, applicable regulations, and recommendations of recog-nized standards development organizations. In our professional
. judgment, the programe and systems as designed will meet the Objectivo of providing accurato and tir.; ly information ::
garding radioactivity releases and potential releases during emergency conditions. Even if accurcte, real-time identifica-tion of specific radionuclides released in the plant or to the environment were feasible, it would not influence the emergency response actions that are dictated by information presently available from the SPNPP RMS. W William H. Wilkie
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Ronald'L. Sh'earin Subscribed and sworn to before me l this 9 tk day of March, 1984. I Notary P de licJ. LA
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) At2achmGn2 A BIOGRAPHICAL DATA DR. WILLIAM H. WILKIE EDUCATION Ale TRAINING 1967-1970 Georgia Institute of Technology, Atlanta, GA
, Nuclear Engineering / Biology / Radiological Physics Specialization Ph.D. 1960-1962 Vanderbilt University, Nashville, TN Physics / Mathematics / Health Physics Specialization ti . S . 1957-1960 North Carolina State University, Raleigh, NC Nuclear Engineering / Engineering Mathematics B.S. (high honors)
- Graduate study one semester 1955-1957 Maryville College, Maryville, TN Mathematics / Physics (no degree) 1984 Chem-Nuclear Systems, Inc., Columbia, SC
" Regulatory Awareness--Radioactive Waste Packaging, Transportation, and Disposal" 1979 Oak Ridge Associated Universi ties 2.nd Department of Energy, Oak Ridge, TN " Health Physics in Radiation Accidents" 1976 Institute for Advanced Technology, Washington, DC " Data Communications Systems" 1967 University of Tennessee, Knoxville, TN Biology 1962 Oak Ridge Institute of Nuclear Studies, Oak Ridge, TN " Advanced Radioisotope Technology" 1971-1981 Tennessee Valley Authority, Muscle Shoals, AL Numerous courses in computer training, management, systems develop-ment, etc.
CERTIFICATION American Board of Health Physics PROFESSIONAL ORGANIZATIONS Health Physics Society International Radiation Protection Association 1 .
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1 Attachmtnt A BIOGRAPHICAL DATA DR. hacLIAM H. WILKIE PRWESSIONAL ORGANIZATIONS (cont.) American Nuclear Society Sigma Xi HONORARY SOCIETIES Pni Kappa Phi Sigma Pi Sigma Tau Beta Pi EXPERIENCE ( Active in the field of health physics for 23 years) 1983- Carolina Power & Licht Company, Raleigh, NC Principal Health Physics Specialist 1981-1983 Electricity Supply Commission, Cape Town, South Africa Regional Senior Health 9hysicist Consul tant 1971-1981 Tennessee Valley Authority, Muscle Shoals, AL Staff Health Physicist 1970-1971 University of Pittsburgh, Pittsburgh, PA Assistant Professor of Health Physics,
- Graduate School of Public Health
! Technical Director, Radiation Medicine Department, Presbyterian-Universi ty Hospital 1968-1970 Technical Analysis Corporation, Atlanta, GA Systems Development Engineer (part time) 1962-1967 Oak Ridge National Laboratory, Oak Ridge, TU Research Associate PREVIOUS CONSULTING Radiological Assessment Systems for Nuclear Power Technology for Energy Corporation Oak Ridge National Laboratory Advanced Research Corporation 2 -__ __. __ _ _ _
D Attachment A 3 BIOGRAPHICAL DATA OR. WILLIAM H. WItKIE BOOKS Am PUBLICATIONS Wilkie, W. H., W. J. Millsap, and J. Walmsley, " Planning Bases for Radiological Emergency Response Near the Koeberg Nuclear Power Station," Escom Report, April 1983
" Operations Manual--Tygerberg Raciation Casualty Facility," Escom/
Tygerberg Hospital Report,1983
" Upgrading Environmental Radiation Data," USEPA Report EPA 520/1-80-012, August 1980 Wilkie, W. H., and M. S. Robinson, " Browns Ferry Nuclear Plant
- Emergency Dose Assessment Procedures for Atmospheric Releases of Radioactivity," TVA Report 0HS-20-80-03, April 1980 Robinson, M. S., and W. H. Wilkie, "3rowns Ferry Nuclear P1 art Emergency Dose Assessment Procedures for Liquid Releases of Radio-activity," TVA Report 0HS-20-80-04, April 1980 Wilkie, W. H., S. M. Nel son, and M. S. Robinson, " Technical Bases for Emergency Dose Assessment Procedures for Liquid Releases of Radioactivity," TVA Report 0HS-20-80-06, May 1980 Wilkie, W. H., and M. S. Robinson, " Technical Bases for Emergency Dose Assessment Procedures for Atmospheric Releases of Radio-activity," TVA Report OHS-20-80-05, May 1980
. Wilkie, W. H., and M. S. Robinson, "Sequoyah Nuclear Plant Emer-gency Dose Assessment Procedures for Atmosoheric Releases of Radio-activity," TVA Report 0HS-20-80-01, Aorii 1980 Robinson, M. S., and W. H. Wilkie, "Sequoyah Nuclear Plant Emer-gency Dose Assessment Procedures for Liquid Releases of Radio-activity," TVA Report OHS-20-80-02, April 1980 "American National Standard for Internal Dosimetry for Mixed Fis-sion and Activation Products," ANSI N343 '978 "The Tennessee Valley Regica Study: Potential Year 2000 Radiologi-cal Dose to Population Resulting from Nuclear Facility Operations,"
00E/ET-0064/2, June 1978 Garry, S. M., and W. H. Wilkie, "Tne use of Environmental Monitor-ing Data in Determining Background Radiation Doses," Pooulation Exposures, USAEC Report CONF-741018 (1974) 3
) '4 Attachment A BIOGRAPHICAL DATA DR. WILLIAM H. WILKIE BOOKS AND PUBLICATIONS (cont.)
Wilkie, W. H. , "The Interdisciplinary Nature of the Radiological Impact of Nuclear Plant Effluents on the Environment," Proceedings of TVA Task Force on Water Resources Research Meeting--The Growing Need for Interdisciplinary Research,1974 Fish, B. R. and W. H. Wilkie, "The Fluid Dynamics of the Spherical Partici e: (1) Tabulation of Settiing Velocity, Reynolds Number, Drag Coefficient, Relaxation Time, and Acceleration-Distance in Air and Water," USAEC Report ORNL-TM-4100 (1973) Wilkie, W. H., "The Spatial and Temporal Capture Distribution for Neutrons in a Coaxial, Two Medium, Liquid Scintillation Detection System," Advanced Research Corporation, Atlanta, GA, February 1970 "Biohazards of Aerospace Nuclear Systems, Final Report," T. G. Clark, B. R. Fish, W. H. Wilkie, J. L. Thompson, R. H. Boyett, and G. W. Royster, Jr. , SC-CR-69-3291 (1969) Wilkie, W. H., and O. S. Harmer, " Theoretical Modulation Transfer Functions and Dosimetry of Fast-Neutron Radiography," Biomedical Sciences Instrumentation-Volume 6, Instrument Society of America, Pittsburgh, PA (1969) Wilkie, W. H. , and B. R. Fi sh , " Scintillation Extrapolation Dosim-etry of Small Beta-Emitting Sources," So1+d State and Chemical Radiation 00simetry in Medicine and Biolo~gy, IAEA, V1enna, Austria
- (1967) l l " Environmental Studies
- Radiological Signi ficance of Nuclear l Rocket Debris," (A series of reports involving classified research) i USAEC Reports ORNL-TM-1053 (1965), ORNL-TM-1159 (1966), ORNL-TM-1686 (1966)
Wilkie, W. H., and R. D. Birkhoff, " Measurement of Spectral Distri-bution of Positron Flux in an Infinite Copper Medium Containing Cu-64," Phys. Rev., g , A1133 (1964) More extensive report on the positron research published as USAEC Report ORht-3469 (1963) l 4
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ATTACHMENT B e BIOGRAPHICAL DATA Ronald L. Shearin Project Specialist - Environmental Education and Training B.S. Degree in Physics from the University of North Carolina, Chapel Hill, North Carolina (1955) M.S.P.H. Degree in Radiological Health from the University of North Carolina, Chapel Hill, North Carolina (1963) Ph.D. Candidate at the University of Florida (1971) - I 1973, no degree) , certificate in Meteorology from Texas A & M (1956) Professional Societies Health Physics (State and National Chapters) Certified Health Physicist by American Board of Health Physics Experience September 1956 to August 1959 - First Lieutenant (Meteorologist) , U.S. Air Force.
' September 1959 to January 1962 - Health Physicist, E.E. DuPont, Aiken, South Carolina. -June 1963 to January 1966 - Health Physics Instructor, U.S.
Public Health Service, Taft Sanitary Engineering Center. January 1966 to June 1968 - Senior Radiological Health Instructor, Southwestern Radiological Health Laboratory, U.S. Public Health Service. June 1968 to July 1970 - USAF Eastern Test Range; Kennedy Space Center (Apollo Program) , ' Liaison Of ficer, U.S._Public Health Service. July 1970 to August 1971 - Chief - Nuclear Facilities Branch, Eastern Environmental Radiation Facility, Environmental Protection Agency, Montgomery, Alabama.
2 Experience cont'd. August 1973 to August 1979 - Health Services Director, Environmental Radiation Studies, Eastern Environmental Radiation Facility, Environmental Protection Agency, Montgomery, Alabama. August 1979 to January 1980 - Employed as a Senior Gener-ation Specialist - Radiation Control in the Generation Services, Harris Energy & Environmental Center, Section of the Generation Departnient located at New Hill, North Carolina. January 1980 to March 1982 - Employed as a Senior Specialist-Eravironmental in the Environmental & Radiation Control Section of the Nuclear Operations Department, located at the Harris Energy & Environmental Center in New Hill, North Carolina. March 1982 to Present - Employed as a Project Specialist-Environmental in the Radiological & Chemical Support Section of the Technical Services Department, lo ated at the Harris Energy & Environmental Center in New Hill, North Carolina. i l I < m
I e SHNPP FSAR ATTACHMENT C 11.5.2.3 Radiation Monitoring SysteL 11.5.2.3.1 System Funweion and Operation The description which follows applies to all radiation monitoring equipment l discussed in Section 5.2. 5, 11.5, and 12.3.4. 1 The major function of the Radiation Monitoring System (RMS) is to provide l plant operations personnel and health physics personnel with both current and historical measurements of radiological conditions in certain areas and plant systems during both normal and design basis conditions. In addition, this system automatically produces alarms to warn plant personnel, and in certain cases exerts control action when unusual radiological conditions or equipment malfunctions occur. All information is presented as efficiently and unambiguously as possible. The RMS is a comprehensive, plant-wide radiation information gathering and control system encompassing the process and effluent monitors and the area and
- airborne monitors. The RMS is a digital, distributed microprocessor-based system in which full functional capability resides locally at the microprocessor controlling each monitor. The RMS is divided into a non-safety related portion and a safety related portion, with all equipment in the latter designed in accordance with IEEE 279-1971, 308-1974, 323-1974, 336-1971, 344-1975 and 384-1974.
The non-safety related portion is composed of the local monitors (Saction 11.5.4.2.1) and four operator's consoles, one located in each of the two Control Rooms, as well as the Radwaste Control Room, and Health Physics Room. Each opeator's console consists of a minicomputer, CRT, and hard copy typer. Each monitor is part of a loop, each loop connecting two operator's consoles. The operator's consoles which are so connected are: Unit 1 and Unit 2 Control Room, and WPB Control Room and Health Physics Room. Communication is in either direction along the loop, thereby assuring redundancy in the event of any single failure. Each microprocessor in a loop is controlled by either of the two operater's console with which it is associated. During the interim period when Unit 1 is operational prior to Unit 2 being operational, the loops shall continue to be routed between the l3 appropriate RMS computers. These computers shall be located in the computer room. During this interim period, no Unit 2 operator's console shall be installed. Readout of all Unit 1 and common monitors shall be via the Unit 1 operator'= console. Information from the monitors is displayed at the CRT. Since all fcur operator's consoles are interconnected, the information is shared among all four operator's consoles and available to all four operator's consoles. .Information includes radiation level in the proper engineering units or counts per minute (cpa), effluent flow histories, monitor status and alarm status. Each monitor has two upscale trips for alert and high radiation, and one downscale trip to indicate monitor failure. Monitor failure includes: low flow, torn filter paper, high dif ferential pressure, 11.5.2-3 Amendmen t No. 3 TL
s SHNPP FSAR and detector failure (low count). Controlled functions include monitor setpoints, purging, checksource activation, and nonitor testing. The safety related portion is composed of the local monitors 2 ( Section 11.5.4.2.1) and four safety related panels, one for each of two electrical divisions A and B, for two Control Rooms. Each monitor is remotely controlled by its own dedicated display and control module located in the appropriate panel. Information is displayed digitally using LED displays. Trending of radiation level is accomplished by recording the radiation level on Seismic Category I, Class IE strip charts driven by a resistive buffered 2 analog output from each display control module. Additionally, the safety related monitors are grouped into loops, each between two non-safety operator's consoles, similar to those of the non-safety monitors, with the exception that all communication ports between each of the safety related monitor; as well as between the safety monitors and non-safety related operator's consoles have properly qualified 1500V, optical isolation buffers, and are used solely for the purpose of transmitting information from
- the monitor to the operator's consoles: no control can be exercised by the non-sefety related portion of the RMS over the safety related monitors. With this technique, information from all the monitors is normally available at the operator's consoles' CRT. Information, control and annunication capabilities of each of the safety related monitors from its display / control module are the same as those capabilities described for the non-safety related monitors.
Furthermore, all information from the Effluent Radiological Monitoring System regarding radiation levels and effluent flows is passed automatically to the report processing computer and used in the preparation of Regulatory df Guide 1.21 reports. This is in conjunction with the meteorological and radiological information inputs from the Process Sampling System. A history 2 of effluent flows is maintained at each microprocessor associated with an effluent detector. Meteorological end ef fluent gross radiation data are collected in real-time modes. The results of isotopic analyses, solid waste data and site specific informattor. (i.e, receptor locations , population data) are collected as they become available. The meteorological sof tware of the report processing computer performs data acquisition, statistical computations, quality checks and data storage. Each meteorological instrument output from the local analog to digital converter is sampled every 10 secor.ds. Every 10 minutes, 10-minute averages are calculated i for each output. For wind speed and wind direction, the minimum and maximum are retained and the standard deviation calculated. Hourly values of these variables are generated and stored for subsequent processing. The effluent gross radiation software collects concentration (i.e., C1/m 3) and process flow data from the operator's consoles, which accumulates hourly averages from each of the monitors in the Ef fluent Radiological Monitoring System. The quality checks made by the microprocessors are used by the operator's consoles to generate status codes which are stored in the report processing computer. Each hourly gross radiation data set will have four indices to indicate what isotopic analyses are associated w'ith that hour. The four 11.5.2-4
SENPP FSAR s indices are associated with isotopic analysis results for nobic gases, iodines, particulates and miscellaneous gases. I The results of isotopic analyses of grab samples performed by multi-channel l analyzers (MCAs), are entered into the report processing computer as they are available. The type of sample, applicable time period, release point, ! concentration of each isotope found and other information are stored. Liquid releases are normally made on a batch basis. The results of isotopic analyses performed prior to a liquid batch release are stored in the report processing computer alo.g with dilution information. Provision is made for continuous liquid release a:nitor data st6 rage. Solid unste and irradiated fuel shipment data required by Regulatory Guide 1.21 are eatered into the report processing computer by the operator for storage. The data collected are stored on the report processing computer's disk. The storage capacity is sufficient to allow time for quarterly and semi-annual report generation while new data are being gathered. Certain data files.are editable prior to their use in computation. All meteorological and radiological data are checked for reasonableness and edited as needed prior to l use in dose computation. Additional files are created during editing to allow [ the subsequent compatations to use edited meteorological and radiological data. This approach preserves the data originally collected yet allows the use of edited data for computations and output. The archiving of data is done on disk. 4 Computation and data reduction will result in a number of smanary reports. These reports are generated upon operator demand. Offsite doses and pre-release dose estimates are generated upon operator demand. Short-term and accumulated doses are calculated in accordance with l Regulatory Guide 1.109, and are available for display at a graphics CRT terminal associated with the report processing computer. l I A schematic of the Radiation Monitoring System is shown on Figure 11.5.2-1. 11.5.2.4 RMS Equipment Description 11.5.2.4.1 Monitor (Cabinet / Skid) Each fluid monitor is a skid mounted scif-contained instrument package i consisting of the requisite number and type of channels, and sampling pumps, valves, interconnecting piping, fittings, flow and pressure indicators, controls and local annunciators, and process fluid pre-conditioners, as required. All skid mounted monitors obtain for analysis a suitable continuous fluid sample of the process of effluent stream and return the sample to the process or effluent stream in a closed loop. Each skid mounted monitor is located as close as practical to the process or effluent stream, such that sample line losses or transit time is minimized. l l 11.5.2-5 l l L _. -_ _ . . _ _ _ _ _ _ _ _ _ . _ . . _ , ____ _ .._, __ [
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-operating floor to close and the normal exhaust fans to stop, while the . eeergency exhaust dampers open and emergency exhaust f ans start up.
There are a total of four safety related triplet detector systems located on the Fuel. Handling Building operating floor. The detectors are located around the appropriate spent fuel pools. The detector locations are shown in Figure 12.3.2-9. The highLradiation clarm setpoints are to be placed at 5 100 nr/hr,.and an actuation will initiate operation of the emergency ventilation, system'and terminate normal ventilation system operation. Since monitors are in a zone II area, the detector alert setpoints are to be ai 2.7 ar/hr. Personnel in the vicinity of the detectors will be warned cf alert levels and high levels of radiation by detector-local inoicator units. 1 Because of a potentially high background, the more sensitive micrcprocessors are removed from the FHB to the RAB. 12.3.4.1.8.4 Other Area Radiation Monitors - 4: I- There are two non-safety area monitors located in the North and South fuel 5 4 pool areas which will also alarm personnel about the presence of unacceptable
.high radiation levels.
All other area radiation' monitors not listed above use the detector types specified in ' Section 11.5.2.5.1 and provide alarms only and exert no control
' action.; These monitors, as well as the ones discssed in i Sections 12.3.4.1.8.1, 12. 3.4.1.8.2, and 12,3.4.1.8.3 are shown in Table 12.3.4-1.
12.3.4.2 Airborne Radiation Monitoring System 12.3.4.2.1 Design Objectives The objectives of the Airborne Radiation Monitoring System during normal operating plant conditions.and anticipated operational occurrences are: a)' To inform operations personnel of airbotne particulate, iodine and gaseous activity trends in the various buildings and structures of.the plant, { b) -to alarm any abnormal increases in the airborna activity levels, i c) to furnish records of gross airborne trends in the various plant areas, d) to help detect identified or unidentified leaks from the reactor coolant pressure boundary (as recommended in Regulatory Guide 1.45, which is
. discussed in -Section 1.8) and other areas of the plant, e) to assist personnel in deciding whether or not breathing apparatus is necessary when entering a high activity area, f) to provide information for evaluation of the performance of all plant systems that function to minimize the release of airborne radioactivity to accessible areas of the plant and to the environment, 12.3.4-5 Amendment No. 5
_ L
SENPP FSAR g) during postulated accidents, to pr9 vide the capability to alarm and initiate isolation of the normal ventilation systems and actuation of the emergency ventilation systems, and to provide operators with information regarding control of the ventilation systems in an emergency, and h) to provide information for the purpose of maintaining low in-plant personnel radiation exposure via inhalation of airborne particulates and iodine, in accordance with 10CFR20 and Regulatory Guide 8.8 (see Section 12.1). 12.3.4.2.2 Criteria for Location of Monitors Considerations for locating the Airborne Radiation Monitoring System monitors are based on the following: a) areas where the airborne radioactivity can abruptly increase and where
- personnel normally have access to the areas, b) in selected ventilation ducts where the monitors can survey the airborne radioactivity level, and c) inside the Containment for the purpose of monitoring unidentified leaks.
The radiation monitors associated with the ventilation system are shown on the flow -diagrams presented as Figures 9.4.0-2 aad following. These monitors are designated by symbols which are defined in Figure 9.4.0-1, Symbols and Abbreviations for HVAC System. 5 An alarm by a ventilation monitor indicates that the airborne contamination is coming from a limited number of enclosures serviced by the monitored ventilation system. Identification of the source enclosure within this group would be made by a systematic program for sampling the air in each enclosure with portable instruments. l 12.3.4.2.3 General System Description The Airborne Radiation Monitoring System provides the means to monitor normal airborne radioactivity levels and to detect and annunciate any abnormal radiation conditions occurring throughout the plant, and as such is an integral part of the Radiation Monitoring System, which is described in detail l in Section 11.5.2. 3. The Airborne Radiation Monitoring System consists of skid or duct mounted gaseous monitors. Each monitor is composed of individual channels, with a channel consisting of a sampling chamber (when required, i.e., off-line, skid mounted monitors), check source, and detector. The detector assembly consists of either gamma or beta sensitive scintillation crystal, a photomultiplier tube and a local amplifier. All channels associated with a monitor are served by a local, dedicated microprocessor, and all channel information is processed through this microprocessor, which is then interrogated by the appropriate radiation monitoring system computer for further processing, indication on a L cathode ray tube (CRT), storage, remote alarming and hard copy production, if so desired. 12.3.4-6 Amendment No. 5
~
SHNPP FSAR The monitors listed in Table 12.3.4-2 of the FSAR comprise the in-Plant _ Airborne Radiation Monitoring System (IPARMS). The IPARMS will consist of 25 chann.As containing two detectors per channel for a total of 50 detectors. Each IPARMS channel will consist of a particulate monitor. iodine monitor, and air sampling system. Each particulate monitor;will be a Beta-sensitive plastic scintillation radiation detector ' coupled to a photomultiplier. tube protected by an - electronaggeticshield. The sensor shall have a minimum detectable limit of 1.0 x'10- pCi/cc concentration for Cesium-137. This is based on sufficient shielding and a sample flow rate of 4 scfm and 4 hour sampling time, in a 1 mr/hr Cs-137 background. Each iodine monitor 'shall be a gamma-sensitive NaI scintillation detector at 5 least two (2) inches.in diameter by two (2) inches in thickness, coupled to a i photomultiplier tube protected by an electro The sensor shall " hava a minimum detectable limit of 2.0 x 15-pygneticpCi/cc shield. concentration for Iodine-131. This.is based on sufficient shielding and a sample flow rate of 4 scfm and 4 hour sampling item, in a 1 mr/hr background.
~ Each IPARMS channel shall be provided with an of f-line air sampling system.
The particulate sampler is a fixed filter assembly and has at least 99 percent efficiency for particles 0.3 microns and larger. The iodine sampler is also a fixed filter assembly and has at least 99% efficiency for iodine. Each sampling system shall be. equipped with a sample probe similar to the isokinetic probe for the Gaseous Effluent Radiation System, except that flow compensation is not required. 12.3.4.2.4 Microprocessor Tra microprocessor is described in Section 12.3.4.1.4. 12.3.4.2.5 Local Annunciation All airborne monitors have . local annunciation, consisting of an audible alarm , rated at 80 dB at 10 feet, and three alarm lights for high radiation, alert radiation and monitor filure. All airborne monitors have .a display at the microprocessor indicating the airborne radioactivity concentration in u-Ci/cc of the particulate isotope (i.e., Iodine 131) or gross particulate or noble gas' activity at the detector,
'12.3.4.2.6 ' Power Supplies Each channel is provided with an independent power supply, designed such that a failure in that channel does not affect any.other channel. Monitors that are identified as safety related are redundant and are supplied power from the station 120V AC safety related buses. The power supplies for these channels are identified in Table 12.3.4-2. Power to the channels that monitor only normal operations is supplied from the station regulated 120V AC instrumentation bus.
12.3.4-7 Amendment No. 5
- 1
m O SHNPP FSAR 12.3.4.2.7 Redundancy, Diversity .and Independence Monitors designate 2 as safety related are part of the safety related portion of the RMS (Section 11.5.2.3) and are designed for redundancy, diversity, and
' independence in accordance with IEEE 279-1971. IEEE 308-1974, IEEE 323-1974, I2EE 336-1971, IEEE 344-1975, and IEEE 384-1974. All monitors which are Seismic Category I are also manufactured and rated to the above standards.
12.3.4.2.8 Airborne Radiation Mcaitors 12.3.4.2.8.1 ' Containment Atmosphere Leak Detection Monitors The containment atmosphere leak detection monitors are part of the safety related portion of the RMS (Section 11.5.2.3) and are designed to provide an indication to operations personnel of the particulate and gaseous radioactivity levels inside the Containment. Radioactivity in the containment indicates the presence of fission products due to a reactor - coolant pressure boundary leak, and as such these aceitors are a ps*t of the Reactor Coolant Pressure Boundary leakage Detection System required by Regulatory Guide 1.45 (see Section 1.8). A detailed description of the sampling system associated with these monitors, operation and monitoring requirements is found in Section 5.2.5. ~ Each monitor uses the airborne particulate and noble gas detector described in Section 11.5.2.6.5. Each of the two Containments has one monitor. Monitors are powered by the Emergency A 5 Bus. A containment isolation actuation signal will isolate these monitors from the Containment. These monitors provide a high radiation alarm when concentrations reach preset limits.- The receipt of these alarms will alert the operator to the presence of low level leakage so that additional sampling can be effected in order to locate the leakage source. ' 12.3.4.2.8.2 Control Room Normal Outside Air Intake Each control room normal outside air intata has two duct mounted beta sensitive monitors, one associated with A Eus, and one with b Bus. These monitors are part of the safety related portion of the RMS (Section 11.5.2.3) and use the ambient gas monitors described in Section 11.5.2.6.1. These monitors provide a high radiation alarm when concentration levels reach preset limits. Upon receipt of the alarm, the monitor closes the normal outside air intake dampers associated with a given unit, shuts off the exhaust fans, closes the exhaust dampers, starts up the blowers, and opens the required dampers to put the air flow into the recirculatory mode. The receipt of these alarms will also alert the operator to check the radiation levels at both emergency outside air intakes, and to open the ir.take at which the radiation level is lower (Section 12.3.4.2.8.3). These normal intake monitors also cause the intake dampers on the electric equicaent protection rooms to close, and then open the recirculation dampers, and close the exhaust dampers. 12.3.4-8 Amendment No. 5 a
.SHNFP FSAR O
In the interim, prior to Unit 2 b'e ing operational, the normal outside air intake monitors for Unit 2 serve as the emergency intake monitors for Unit 1. g10 Section 12.3.4.2.8.3 presents a full discussion of this concept.
- 12.3.4.2.8.3 ' Control Rooms Emergency Outside Air Intake For Units 1 and 1, there are two common emergency. outside air intakes. Each 1, A Bus; intake has four duct-mounted Beta monitors associated with: Unit Unit.2, A Bus; Unit 1, B Bus; Unit 2, B Bus. These monitors are part of the l10 safety related portion of the RMS (Section 11.5.2.3), and use the ambient gas monitors described in Section 11.5.2.6.1.
Thase monitors provide indication to the control room personnel of the radioactivity levels at each emergency air intake, thereby allowing the operator to choose which emergency intake to open (see discussion in Section 12.3.4.2.8.2). These monitors also provide a high radioactivity alarm when concentration levels reach preset'11mits. , In the interim period, prior to Unit 2 being operational, the normal outside
- air intake for Unit 2 serves as the emergency outside air intake for Unit 1.
During this period, the normal air intake radioactivity monitors for Unit 2 serve as the emergency intake monitors for Unit 1, and consequently the required panel mounted display and control modules controlling these monitors are mounted in Unit l's safety related panel. Additionally, all software and t hardware requirements in communicating with the non-safety rel ated portion of the 1005 are completed. . to 12.3.4.2.8.4 Other Airborne Radiation Manitors A non-safety Particulate and Iodine monitor is located on the fuel handling building operating flocr. This monitor takes an ambient air sample from the surrounding air and analyzes it for airborne particulata and iodine concentrations. Alert level alarms are to be set at CS-137 MPC levels for particulate and I-131 MPC levels for Iodine. High level alarms are to be set at 10 MPC' levels. Operators will be warned of when respirators are required to permit occupancy. All other airborne radiation monitors not listed above use the airborne particulate and iodine monitor described in Section 11.5.2.6.3 and provide for alarm only and exert no control action. These monitors, as well as the ones discussed in Sections 12.3.4.2.8.1, 12.3.4.2.8.2, and 12.3.4.2.8.3 are shown in Table 12.3.4-2. I 12.3.4-8a Amendment No. 10 E
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TARLE 12.3.4-1 AREA RADIATION HONITCRS, Typical Subsection High of Senst- Alarm tivity Accu- Set-12.3.4.1.8 Described Range _ CPM / racy points
- Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Power Description Tag # in
.4 CM Tube 10~I-105 10 2 +10 2.5 control Instrumentation Control Room 21RR-3560 SA ~~
Room AC Safety Bus Area (11.5.2.5.1). CONTAINt1ENT 10I -10 7 ISOLATION A 10 3 Around + Unit 1 ICR-3561A SA .1 Ion chamber 1 ~~+15 Reactor g (11.5.2.5.3) Cavity
~ " " " " " + U!
Unit 2 2CR-3S61A SA .1 % h 10 CONTAINMENT ISOLATION B " " " " " "
+
Unit 1 ICR-3561B SB .I
~ " " " *- " +
Unit 2 2CR-3561B SB .I 10
$ + Instrumentation AC Safety Bus g-8 logical alarm points in the absence of data from the operating conditions *encountered These typical alarm setpoints representThe actual alarm points will be adjusted to provide useful alarms based on the radiation y in the plant.-
The operations being conducted and the location of personnel within the plant.
- 1evel within the plant.
! c5
TABLE 12.3.4-l'(continuest) Typical High-Subsection of Sensi- Al a rm 12.3.4.1.8 tivity Accu- Set-Range CPM / racy points Described Power Description Tag # in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Instrumentation CONTAINHENT AC Safety Rus ISOLATION C 3 + Unit 1 ICR-3561C SA .1 Ion chamber 10I -10 7 1 ~~+15 ~ 10 Around Reactor (11.5.2.5.3) Cavity m
" " " " ~ ~ +
Unit 2 2CR-3561C SA .1 10 r. I' CONTAINMENT
- ISOLATION D " " " " " " + U!
i' Unit 1 ICR-3561D SB .1 " " " " "
+ $
- E Unit 2 2CR-3561D SB .1 to i
i i
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TABI.E 12.3.4-1 (continued) Typical Subsection High of _ Sensi- Alarm 12.3.4.1.8 tivity Accu- Set-Described Range CPM / racy points Description Tag # in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) T,ocat ion Pow'e r Post LOCA ~ HONITOR A Unit 1 ICR-3563A SA .2 CM Tube 10~I-105 10 2 ~+10 10 3 Outside + (11.5.2.5.1) containment Unit 2 2CR-3S63A SA .2 shield ws11 + 1 10 vi
;; Post LOCA l
MONITOR B , ,, ,, , ,, ,, m Unit i ICR-3563B SB .2 m i' Unit 2 2CR-3563B SB .2
" " ~ ~ ~ + "
C b 10 3 2 Around FilB + Fuel llandling *lFR-3564 SA .3 CM Tube 10-2-104 10 +10 10 Bldg. South *lFR-3564 SB .3 (11.5.2.5.2)
~ " " " " +
. *lFR-3565 SA .3
+ *lFR-3565 SB .3 +
Fuel llandling *lFR-3566 SA .3
+ *lFR-3566 SB .3 +
g Bldg. North " " " " " "
+
5 *lFR-3567 SA .3
*lFR-3567 SB .3 +
h n . z +1nstrumentation 1C Safety Bus
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TARI.E 12.3.4i1 (continued) Typical Subsection ~ High of Se nsi- Ala rm 12.3.4.1.8 tivity- Accu- Set-Described Range CPff/ racy points Description Tag # in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Power CONTAINMENT SOUTil STAIRWELL, EL.' 261 Instrumentation
.4 2 2 Unit 1 ICR-3575 CM Tube 10-2-105 10 +10 10 Plant AC Rus (11.5.2.5.1)
Unit 2 2CR-3575 .4 " " "
- 10 CONTAINMENT NORTH ui
;; STAIRWELL, EL. 261 w Unit i ICR-3576 .4 " " "
Unit 2 2CR-3576 .4 " " " " ~ n, a us E3 $ 10 3 CONTAINMENT BLDC. IN-CORE INST. CONTROLS Unit i ICR-3577 .4 "
" " " " " ~
l Unit 2 2CR-3577 .4
- 10 t
' a g.
ss y '
- Instrumentation AC Hus
TABLE 12.3.4-1 (continued) Typical
. Subsection High' of Sensi- Al a rm 12.3.4.1.8 tivity Accu- Set-Described -Range CPM / racy points Description Tag #- in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Pow'e r
- Recycle Monitor Tank CM Tube 1A & 2A 21CR-3578A .4 (11.5.2.5.1) 10-2-105 10 2 +3g 39 2 Plant
- IB 5 2B 21CR-3571B .4
- 10 Boric Acid Filters
.4 *- @
[; Unit i ICR-3579 Unit 2 2CR-3579 .4 MS i IO N C: b Valve Gallery " ' Unit 1 ICR-3580 .4 " " " -
" ~
- Unit 2 2CR-3580 .4 10 Access Aisle Between Valve g Calleries a Units 1 & 2 21CR-3581A .4 g- " " " " " "
- 10 e Units 1 & 2 21CR-3581B .4 u
< n
- r.
- Instrumentation AC Bus O
TABLE 12.3.4-1-(continued). Typical Subsection High of Sensi- Alarm'
.12.3.4.1.8 tivity Accu- Set-Described Range CPM / racy points Description Tag i in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Poise r CONTAINMENT NORTH STAIRWELL, EL. 286 CM 'linbe Unit 1 ICR-3582 .4- (11.5.2.5.1)- 10-2-10 5 10 2 -+10 10 2
Plan
- Unit 2 2CR-3582 .4 - " " "
10 CONTAINMENT SOUTH STAIRWELL, EL. 286
" en C Unit 1 ICR-3583 .4 10-I-105 - - - ,
w Unit 2 2CR-3583 .4 " " " " " n **
,L
- 10 3 Pressurizer Unit 1 ICR-3584 .4 " " " "
Unit 2 2CR-3584 .4 " " " " 10 SOUTH SECONDARY CON-TAINMENT, EL. 236 g Unit 1 ICR-3585 .4 " g Unit 2 2CR-3585 .4 " " " " I
$ 10 re N +
- Instrumentation AC Bus o
T
- e TABLR 12.3.4-1 (continued)
Typical Subsection 'High of Sensi- Al arm 12.3.4.1.8 tivity Accu- Set-Described Range CPM / racy points Description Tag # in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Power Primary Cool-ing Ducts CM Tube Unit 1 ICR-3586 .4 (11.5.2.5.1) 10-I-105 10 2 -+10 10 2 Plant
- Unit 2 2CR-3586 .4 - " " " " "
10 Reactor
- Coolant Tank "
P Unit 1 ICR-3587 .4 " " " " " " P Unit 2 2CR-3587 .4 " " " " " " I E U 10 N Personnel lock & Equip. Removal Area Unit 1 ICR-3588 .4 " " " " " " Unit 2 2CR-3588 4 " " " " 10 g Volume Control Tank
- Unit i 1RR-3595 .4 " " " " " "
h Unit 2 2RR-3595 .4 " " " " " " e
$ 10 x
g
- Instrumentation AC Bus
TABLE 12.3.4-1-(continued)- . Typical Subsection _ High of Sensi- Alarm 12.3.4.1.8 tivity Accu- Se t-Described Ra nge CPPf/ racy points Description Tag # in ' Detector Type (mR/h r) (mR/hr) (%) (mR/hr) Location Power f MS & FW Pipe Tunnel Floor CM 'lisbe Unit 1 1RR-3596 .4 (11.5.2.5.1) 10-I-105 10 2 -+10 10 2 Plant
- Unit 2 2RR-3596 .4 " " " " " "
10 4 RHR Pump IB ~ Unit i 1RR-3597 .4 " " " " " " C Unit 2 2RR-3597 .4
" " " - " ~
10 'o T m E RHR Pump 1A Unit 1 1RR-3598 .4 " 10~I-10 5 . Unit 2 2RR-3598 .4 " " " " " "
,10 3
Charging Pump IA 4 ]' Unit i 1RR-3599A .4 " " " " " Charging Pump 2A
, Unit 2 2RR-3599A .4 " " " "
10 i r, 2: '
?
- Instrumentation AC Bus i
TABLE 12.3.4-1 (continued) Typical Subsection High of Se nsi- Alarm 12.3.4.1.8 tivity Accu- Set-Described Range CPH/ racy points Description Tag # in Detector Type (mR/h r) (mR/hr) (%) (mR/hr) Location Powe r Charging Pump IB Ctl Tube Unit 1 1RR-3599B .4 (11.5.2.5.1) 10-I-105 10 2 +10 10 2 Plant
- Charging Pump 2B Unit 2 2RR-3599B .4 " " " " "
10 [ Charging Pump IC g" w Unit 1 1RR-3599C .4 " " - " "
- G
,, Charging Pump 2C
- Unit 2 2RR-3599C .4 " " " " -
- E!
$5 10 Recycle Evaporator Valve Gallery Unit 1 1RR-3600 .4 " " " " " -
Unit 2 2RR-3600 .4 " " " " " 8 to 8 n f
- Instrumentation AC Bus o
s TABLE 12.3.4-1 (continued) Typical Subsection High of Sensi- Alarm 12.3.4.1.8 tivity Accu- Set-Described Range CPM / racy . points Descrir. tion Tag i in Detector Type (mR/hr) (mR/hr) Q) (mR/hr) Location Power Letdown lix Valve CM 'Ibbe Callery Unit 1 1RR-3601 .4 (11.5.2.5.1) 10-I-105 10 2 +10 2 x 10 3 Plant
- Unit 2 2RR-3601 .4 - " - ~ ~ "
10 Moderating lix Valve Callery Unit i 1RR-3602 .4 " " " " " g Unit 2 2RR-3602 .4 " " " - " "
. 10 'o i - *n.
v CD Boric Acid Pump Valve $ Callery Unit i 1RR-3603 .4 " " " - " Unit 2 2RR-3603 .4 - " " " " " 10 Unit I and 2 Access Aisle 21RR-3604 .4 " " " " 2 - 10 i 10 e 8 n
-
- Instrumentation AC Bus
i
,TABI.E 12.3.4-1 (continued)
Typical Subsection High of Sensi- Alarm 12.3.4.1.8 tivity Accu- Set-Described Range CPtf/ racy points Description Tag i in Detector Type (mR/hr) (mR/hr) (%) . (mR/hr) Location Power Unit i Sample . CM Tube 2 2 Room lA 1RR-3605A .4 (11.5.2.5.1) 10-I-105 10 +10 10 Plant
- Unit 2 Sample Room 2A 2RR-3605A .4 " " " " " "
10 [ Unit 1 Sample g" w Room 1R IRR-3605B .4 " " " " "
- 4 Unit 2 Sample *
)* Room 2B 2RR-3605B .4 " " " "
- M N
10 4 RilR IIx IA 1RR-3606A .4 " " " " " RilR lix 2A 2RR-3606A .4 " " " 10 N' " " " i
$ RilR lix IB IRR-3606B .4 " " "
- g RilR lix 2R 2RR-3606B .4 " " " " " "
i $ n ! m: - 10 g
- Instrumentation AC Bus t
4
-TABLE 12.3.4-1 (cantinued)
Typical Subsection High of Sensi- Alarm
.12.3.4.1.8 tivity Accu- Set-Described Range CPt!/ racy points Description Tag # __
in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Power i Decon Area El. 236 CM habe Unit i 1RR-3607 .4 (11.5.2.5.1) 10-I-105 10 2 +10
~
10 2 Plant
- Unit 2 2RR-3607 .4 " " " " " "
10 Decon Area El. 261 Unit i 1RR-3608 .4 " " " " " Unit 2 2RR-3608 .4 " " " " " " sn h v 10m L Waste Monitor 5 o TK I & 2A 21RR-3609A .4 " " " " " " i N Waste Monitor TK 1 & 2B 21RR-3609B .4 " " " " " "
- 10 Secondary
, Waste Sample a TK 1 & 2 21RR-3610 .4 " " "~ " " " t E 10 n
.E g *Instrementation AC Bus e
TABLE 12.3.4-1 (continued) Typical Subsection . High of Sensi-- Ala rm 12.3.4.1,8 tivity Accu- Set-Described Range CPM / racy points Description Tag # in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) T.ocation Power Recycle Hold up Tank GM Tube Unit 1 1RR-3611 .4 (11.5.2.5.1) 10-I-105 10 2 +10i 102 , Plant
- Unit 2 2RR-3611 .4 " " " " -
- 10 Spent Fuel Pool South IFR-3620 .4 " " " " *
" u)
[; North IFR-3621 .4 " ~ " " " La
* "8 , Spent Fuel d, Shipping Cask IFR-3622 .4 " " " " " "
- g Spent Fuel Pool lix South IFR-3623A .4 " " " " " "
- North IFR-36238 .4 " " " " "
- Access Aisle Between Spent Fuel itx IFR-3625 4
- Purification h.
a Pump Filter N 2A & 3A 32FR-3626 .4
- sc 1A & 4A 41FR-3626 .4 "
.o D""8 *I.istrumentation AC Bus
?
TABLE 12.3.4-1 (continued) Typical' Subsection ilIgh of Sensi- Al a rm 12.3.4.1.8 tivity Accu- Set-Described Range CPtf/ racy points Description Tag # in Detector Type (mR/hr) (mR/hr) -(%) ,mR/hr) ( Location _ Power New Fuel CM Tube Storage Area 1FR-3628 .4 (11.5.2.5.1) 10-I-105 2 2 10 +10 10 Plant '* Filt Backwash Trans. Tank VA Callery IWR-3635 .4 - " " " " + , WPB Filter Part Conc. 'lK 4 - VA Callery IWR-3636A .4 - " " " " - " l0 1 p
- g
- 4. F WPB FiIter Part
- 7 Conc. TK U
1 U VA Callery WR-3636B .4 " " " " " ~ Cas Decay Tank Drain Pump 1o 2 21WR-3637 .4 - " * " " ~ 10 + Backwash Storage TK Pump Valve 9 g Callery IWR-3618A .4 " * * " " " e
, Backwash Storage l'
- , , TK Pump Valve ,
E Callery IWR-3638B .4 " " ' " * " z -
~
- Instrumentation AC Bus O
1 i
~
TABLE 12.3.4-1-(continued) Typical Subsection High of Sensi- Alarm 12.3.4.1.8 tivity Accu- Set-Described Range CPM / racy points Description _ Tag # in _ Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Power Waste Cas Comp. .CH Tube-2 2
- IA & 28 21WR-3639A .4 (11.5.2.5.1) 10-I-105 10 +10 10 Plant IB & 2B 21WR-3639B .4- 10 Waste Evap.
Conc. Valve " Callery 21WR-3640 .4 in
- m k
b 5
" !5 Waste Evapo- 10 rator Feed Pump Valve " " " " " "
Ca11ery 21WR-3641 .4 10 -f D g- Waste Evaporator Valve Callery
$ .4 t' 1A & 2A 21WR-3642A 2:
10 P g
- Instrumentation AC Bus
1 TABLE 12.3.4-1 (continued) Tyoical-Subsection High of Sensi- Alarm 12.3.4.1.8 tivity Accu- Set-Described Range CPH/. racy points Descrip_ tion _ Tag # in Detector _T g (mR/hr) (mR/hr) ,(%) (mR/hr) Location , Power i Waste Evaporator Valve Callery CM Tube IB & 2B 2 21WR-36428 .4 (11.5.2.5.1) 10-I-105 10 +10 10 2 Plant
- 10 Waste Evaporator Sample Room I IWR-3643 .4 " " " " - "
4 Room 2 2WR-3643 .4 " " " " " un
." m . 10
- M N Spent Resin m
! Pumps b 1 - 2A
- LWR-3644A .4 " " " " " "
1 - 2B
- LWR-3644B .4 " " " " " "
i Catalytic , Recombiner , 1 & 2A 21WR-3645A .4 I & 28 21WR-3645B .4 " " " ~ 10 m
@ WP Control Room
- LWR-3646 .4 " " " "
2.5 " r . eg g
- Instrumentation AC Bus i
TABLE 12.3.4-1 (continued) Typical Subsection High of Sensi-- Alarm 12.3.4.1.8 tivity Accu- Set-
. Described Range CPM / racy points Description Tag # in Detector Type (mR/hr) (mR/hr) (%) (mR/hr) Location Power Cas Decay TK Valve Callery CM Tube 1&2 21WR-3647A .4 (11.5.2.5.1) 10-I-105 10 2 +10 10 2 Plant * .
1&2 21WR-3647B .4 " - " " "
- 10 Reverse Osmosis
'. Module G East IWR-3648 4 " "
w West IWR-3649 .4 " " " " " "
*u v
5 L Reverse Osmosis lT) Module N
,2 21WR-3650 .4 * * " " * " -
3&4 43WR-3650 .4 " " " " " Floor Drain Tank I & 2A 21WR-3651A .4 " " " 3 & 4A 43WR-3651A .4 " " " * " " I & 2B 21WR-36518 .4 " " " " 3 & 4B 43WR-3651B .4 " " " " " " l :s WPB Men's
@ Locker Room IWR'3652A .4 " " "
2.5 "
$ WPB Men's " Locker Room IWR-3652B .4 " " " "
2.5 ' y WPB Women's Locker Room IWR-3652C .4 * " " " 2.5 " o
- Instrumentation AC Bus i
i
TABLE 12.3.4-2 _AIRR1RNE RAD 1 ATION MON 1 TORS Subsection of 12.3.4.2.8 Accu- Typical Des rlbed Ranoe racy High Alarm Description Tao # in Detector Type (pCl/cel Sensitivity (5) Setpoints Location Power Containment # Particulate - 3.6x10'I2- 8.6x10 CPM /lC! 3x10'I2tCl/cc Atmosphere (11.5.2.5.8) & 3.6x10'I Cs-137 & Ou+ side Leak'De- 7 Noble Gas . lx10'#- ' 4.4x10 CPM /lfi/cc ~+10 2x 10'ILCl/cc Contain-tection 4 (11.5.2.5.6) 5x10~3 Xe-133 ment Unit 1 1 T-3502A SA .I
" " " " " " +
Unit 2 2LT-5502A SA .I " " " " " " + w Containment 10 $" m i Atmosphere N Leak De-i tection B h Unit 1 ILT-35028 SB .I
" " " " " " +
Unit 2 2LT-3502B SB 1 " " " " " " + q i 10 Main Control g Rooms Norme1 Beta 5 8 Outstde Air Sensitive 2x10 1.0x10 CPM / Cl/cc i Intake (11.5.2.5.4) 2 x 10'3 Kr-85 +10 2.5 mR/hr Duct e E Unit I IC2-3504A SA 2 " " " " " " + g Unit 2 2CZ-3504A SA .2 " " h " " " +
.u W
o 10
+1nstrumentation AC Saf ety Bus e
. . . . . .~ . ~ . TABLE 12.3.4-2 (continued)
Subsection o' 12.3.4.2.8 Accu- Typical Described Ranoe racy Hlah Alarm Description Tao i _ in Optactor Type gl/ce) Sensitivity (5) .Setpoints Location Power Main Control Rooms Norma 1 Beta Outside Air Sensitive 2xt0 0 1.0x10 CPM /lfi/cc Intake ( 11.5.2.5.4) - 2x10-3 Kr-85 +10 2.5 d /hr Duct Unit I ICZ-35048 SB 2 a a a a a a + unl+ 2 2CZ-35048 SB 2
- a a a a a +
10 " w Main Control ,, p Room Emerg, y to Outside Air Intake Unit 1 ICZ-3505A SA 3
- a a a 10 d /hr a +
Unit 2 2CZ-3505A SA 3 a a a a a a + 10 Main Control Room Emero, p Outside Air k h intake 8 D Unit i ICZ-35058 S8 3 a a * = n . , 2CZ-35058 SFI * .a Unit 2 .3 = . . , , O . 10
+1nstrumentation AC Saf ety Rus i
~
TABLE 12.5.4-2 (continued) Subsection of 12.3.4.2.8 Accu- Typical DescritK d Range racy H igh Alarm Description Tag d in Detector Type (tCl/cc) Sensitivity , ($) Setpoints Location Power Main Control Room Emerg. Beta 8 Outside Air Sensitive 2x10 1.0x10 CPM /pCl/cc intake C (11.5.2.5.4). 2x10-3 Kr-85 +10 10 mR/hr - Duct Unit 1 1CZ-3505A2 SA .3 " " " " " " + Unit 2 2CZ-3505A2 SA 3 " " " " " " + 10 i
~ cn f
w Main Control , i p Room Emargs y N Outside Air M Intake D Unit i 1C2-350582 SB - 3 " " " " " " + 2CZ-350582 SR " " " " " " + Unit 2 3 10 4 ? D
.k 83 re SE o +1pstrumentation AC Saf ety Bus e
O f
TABLE 12.3.4-2 (continued) Subsection of 12.3.4.2.8 Accu- Typical Described Hange racy liigli Alarm De?cription Tag # in Detector Type (DC1/cc) Sensitivity (%) _ Setpoints Location Power Particulate - 3.6x10-I2- 8.6x10 4CPtt/pC1 3x10-IlpC1/cc (!!.5.2.L.4) & 3.6x10-7 Cs-137 & Outside lodine - 6.2x10-12_ g,oxgoSCPrl/p ci 2x10 10 p ci/cc- Contain- , (11.5.2.5.9) 6.5x10-5 1-131 +10 ment llot & chine Sh p V:nt IAV-3533 .4 " " " " " - *
- WPB llP Calibration Room IWV-3548 .4 " " " " " "
FitB Opera-ting kloor IAV-3549 .4 " " " " " " WPB Control Moom lAV-3550 .4 - " " " " " 1 Volume Meduction System IWV-3551 .4 " " " " " " Soliditication System IWV-3552 .4 - " "
- WP3 Personnel liand tlng Fac ility IWV-3553 .4 ,
llo t Lab IWV-3554 .4
- 8814tatrumenLation AC ilus 12.3.4-29 S!!NPP FSAR
?
.. ... - . . ~. .
SHNPP FSAR
REFERENCES:
'SECTION 12.3 > 12.3.1-1 Design, Inspection, Operation, and Maintenance Aspects of the j[ NSSS to Maintain Occupational Radiation Exposures ALARA, WCAP-8872, April-1977.
'12.3.2-1 Rockwell,' T. , " Reactor Shielding Design Manual," USAEC Report 7004, 1956.
12.3.2-2 ISOSHLD, " Kernel Integration Code, General Purpose Isotope Shielding Analysis,"' CCC-79, Oak Ridge RSIC, 1973. 12.3.2-3' SPAN-4', "A Point Farnel Computer Program for Shielding," WAPD-TM-809(L), 0.J. Wallace, October 1973. 12.3.2-4' MORSE-CG, " General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code with Combinational Geometry," CCC-203, E.A. Straker; Oak Ridge, 1970. 12.3.2-5 Engle, W. W., Jr., "A Users Manual for ANISN: A One
' Dimensional Discrete Ordinates ' Transport Code with Anisotropic Scattering", K-1693 (1967). Contained in CCC-254 from the Radiation Shielding Information Center, Oak Ridge National Laboratory.
12.3.2-6 Courtney, J. C. , "A Handbook of Radiation Shielding Data," ANS/SD-76/14, July, 1976. 12.3.2-7 Schaeffer, N. M., " Reactor Shielding for Nuclear Engineers,"
- TID-25951, 1973.-
12.3.2-8 DiNunno, Anderson, Bales, and Anderson " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, Atomic Energy Commission, March 23, 1962. 12.3.2-9 TMI-2' Lessons' Learned Task Force States Report and Short-Term Recommendations, NUREG-0578, U.S. Nuclear Regulatory Commission, July, 1979. 3 12.3.2-10 Theodore Rockwell III, " Reactor Shielding Design Manual," U.S. AEC, 1956. 12.3.2-11 R. L. Engle, J. Greenberg, M. M. Hendrickson, "ISOSHLD - A Computer
. Code for General Purpose Isotope Shielding Analysis", BNWL-236, 1966.
12.3.2-12 Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, NUREG-0588 U.S. Nuclear Regulatory Commission, August, 1979. 12.3.2-13 Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, NUREG-0588 Rev. 1 U.S. Nuclear
' Regulatory Commission, July, 1981.
Amendment No. 3 N
w - SHNPP FSAR
REFERENCES:
- SECTION 12.3 (Continued) 12.3.2-14 Clarification of-IMI Action Plan Requirements. NUREG-0737 U.S.
3 Nuclear Regulatory Commission, November, 1980. F f 4
+
i i i e i 1 1 O a Amendment No. ~l
RADIATION ZONES LEGEND: RADIATION DOSE RATE LEVELS D, EXPECTED NORMAL OPERATION I < 0.25 m rem /hr ll 0.25 $ D < 2.5 Ill 2.55 D < 5.0 IV 5.0 < D < 100 V 2100 () INDICATES AFTER SHUTDOWN
- PATROL ROUTES Y AREA RADIATION MONITOR insenesessee RADIATION ZONE BOUNDARY l
l SHEARON HARRIS FIGURE NUCLCAR POWER PLANT RADI ATION ZONES Carolina Power & Light Company LEGEND 12.3.2 1 FINAL SAFETY ANALYSIS REPORT _ __ ___ _ _ t
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